View source for CNS-16-005, Final Notification of Full Compliance with Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events and with Order EA-12-051, Order to Modify.

{{Adams
| number = ML16049A041
| issue date = 02/15/2016
| title = Final Notification of Full Compliance with Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events and with Order EA-12-051, Order to Modify.
| author name = Henderson K
| author affiliation = Duke Energy Carolinas, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000413, 05000414
| license number = NPF-035, NPF-052
| contact person = 
| case reference number = CNS-16-005, EA-1 2-051, EA-12-049
| document type = Letter type:CNS
| page count = 85
| project = 
| stage = Other
}}

=Text=
{{#Wiki_filter:"1
   ~ENERGY%
Kelvin Henderson Vice President Catawba Nuclear Station Duke Energy CN01VP I 4800 Concord Road York, SC 29745 o: 803.701.4251 f: 803.701.3221 10 CER 50.4 10 CFR 2.202(b)
CNS-1 6-005 February 15, 2016 Attention: Document Controi Desk U. S. Nuclear Regulatory Commission Washington, D. C. 20555-001 Duke Energy Carolinas, LLC (Duke Energy)
Catawba Nuclear Station (CNS), Units 1 and 2 Docket Number(s) 50-413 and 50-414 Renewed License Nos. NPF-35 and NPF-52

==Subject:==
Final Notification of Full Compliance with Order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events" and with Order EA-1 2-051, "Order to Modify Licenses With Regard To Reliable Spent Fuel Pool Instrumentation" for Catawba Nuclear Station

==References:==
: 1. Nuclear Regulatory Commission (NRC) Order Number EA-1 2-049, Order Modifying Licensees With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, Revision 0, dated March 12, 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. ML12054A735
: 2. Catawba Nuclear Station (CNS) Overall Integrated Plan Submittal in Response to March 12, 2012, Commission Order to Modify Licenses With Regard To Requirements for Mitigation Strategies for Beyond Design Basis External Events (Order EA-12-049), dated February 28, 2013 (ADAMS Accession No. ML13066A173)
: 3. Catawba Nuclear Station, Units 1 and 2 - Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-1 2-049 (Mitigation Strategies), dated February 6, 2014 (ADAMS Accession No. ML13364A175)
: 4. NRC Order Number EA-12-051, Order to Modify Licenses With Regard To Reliable Spent Fuel Pool Instrumentation, dated March 12, 2012, (ADAMS Accession No. ML12054A679)

              '-4              "4 United States Nuclear Regulatory Commission Page 2 February 15, 2016
: 5. Letter from Duke Energy to NRC Site Overall Integrated Plans in Response to March 12, 2012, Commission Order Modifying Licenses With Regard To Requirements for Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051), dated February 28, 2013 (ADAMS Accession No. ML13086A095)
: 6. Catawba Nuclear Station, Units 1 and 2 - Interim Staff Evaluation and Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-1 2-051, Reliable Spent Fuel Pool Instrumentation, dated October 28, 2013 (ADAMS Accession No. ML13281A562)
: 7. Catawba Nuclear Station, Units 1 and 2, Report for the Audit Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Instrumentation to Orders EA-12-049 and EA-1 2-051, dated February 20, 2015 (ADAMS Accession No. ML15035A679)
: 8. Notification of Full Compliance with Order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events" and with Order EA-1 2-051, "Order to Modify Licenses With Regard To Reliable Spent Fuel Pool Instrumentation"  -  Catawba Nuclear Station Unit 2, dated May 1, 2015 (ADAMS Accession No. ML15126A277)
Ladies and Gentlemen On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued Order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design-Basis External Events" and Order EA-1 2-051, "Order to Modify Licenses With Regard To Reliable Spent Fuel Pool Instrumentation," (Reference 1 and Reference 4, respectively).
The Orders require holders of operating reactor licenses and construction permits issued under Title 10 of the Code of Federal Regulations Part 50 to submit for review, Overall Integrated Plans (OIPs) including descriptions of how compliance with the requirements of each Order will be achieved. By letter dated February 28, 2013 (Reference 2), the OIP for CNS in response to Order EA-12-049 was submitted. In a separate correspondence, the OIP for CNS in response to Order EA-12-051 was submitted by letter dated February 28, 2013 (Reference 5).
Order EA-12-049, Section IV.A.2 and Order EA-12-051, SeCtion IV.A.2 requires completion of full implementation to be no later than two (2) refueling cycles after submittal of the overall integrated plan, as required by Condition C.I .a or December 31, 2016, whichever comes first.
In addition, Section IV.C.3 of Orders EA-12-049 and EA-12-051 require that Licensees and CP holders report to the NRC when full compliance is achieved. For CNS Unit 1, the current requirement for full implementation of NRC Orders EA-12-049 and EA-12-051 was prior to restart from the I1EOC22 refueling outage.
On December 16, 2015, CNS Unit I entered Mode 2 (Startup) following the 1EOC22 refueling outage. As such, December 16, 2015 is the compliance date for CNS Unit I for being in full compliance with Orders EA-12-049 and EA-12-051 as demonstrated by this submittal and any other docketed correspondence concerning these Orders. This determination is based on the best available information and analyses that have been completed as of the date of this letter.

United States Nuclear Regulatory Commission Page 3 February 15, 2016 Notification of full compliance with Orders EA-12-049 and EA-12-051 for CNS Unit 2 was provided by Reference 8. provides a brief summary of the key elements associated with compliance to Orders EA-12-049 and EA-12-051 for CNS Unit 1. Attachment 2 provides the open and pending items from the Audit Report (Reference 7).. For each open and pending item identified in Attachment 2, a brief summary response in support of closure is provided. As such, Duke Energy Carolinas, Inc. (Duke Energy) considers these items complete pending NRC closure. provides all answers to the diverse and flexible strategies Interim Staff Evaluation open and confirmatory items contained in Reference 3. Attachment 4 provides all answers to the spent fuel pool (SEP) instrumentation Interim Staff Evaluation (ISE) Request For Additional Information contained in Reference 6. Attachment 5 provides the bridging document between vendor technical information and CNS specific considerations for SEP instrumentation, which compares CNS assumptions to the vendor's assumptions for the SEP instrumentation. provides the CNS Final Integrated Plan. Attachment 7 provides the CNS RCP Seal Leakage Margin Assessment.
In support of the ongoing NRC Audit process, Duke Energy will continue working with the NRC staff in the issuance of a combined Safety Evaluation (SE) for both the Mitigation Strategies and the Spent Fuel Pool Level Instrumentation Orders.
There are no regulatory commitments contained in this letter or its attachments. Please address any comments or questions regarding this matter to Cecil Fletcher at 803-701-3622.
I declare under penalty of perjury that the foregoing is true and correct. Executed on February 15, 2016.
Sincerely, Vice President, Catawba Nuclear Station Attachments:
: 1. CNS, Unit 1 Summary of Compliance Elements for Orders EA-1 2-049 and EA-1 2-051
: 2. CNS NRC Audit Report Open and Pending Items
: 3. CNS, Response to Diverse and Flexible Strategies Interim Staff Evaluation Open and Confirmatory Items
: 4. CNS, Response to Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order-i12-051, Reliable Spent Fuel Pool Instrumentation
: 5. Design Bridge Document Between Vendor Technical Information and CNS Specific Considerations for Spent Fuel Pool Instrumentation
: 6. CNS, Final Integrated Plan
: 7. CNS, RCP Seal Leakage Margin Assessment

United States Nuclear Regulatory Commission Page 4 February 15, 2016 xc:
C. Haney, Regional Administrator U. S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, GA 30303-1 257 W. M. Dean, Director, Office of Nuclear Reactor Regulation US. Nuclear Regulatory Commission One White Flint North, Mailstop 13 HIdM 11555 Rockville Pike Rockville, MD 20852-2738 J.A. Whited, Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mailstop 8 B1A 11555 Rockville Pike Rockville, MD 20852-2738 P.J. Bamford, Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mailstop 13 F15 11555 Rockville Pike Rockville, MD 20852-2738 G.A. Hutto NRC Senior Resident Catawba Nuclear Station Justin Folkwein American Nuclear Insurers 95 Glastonbury Blvd., Suite 300 Glastonbury, CT 06033-4453

ATTACHMENT 1 CNS, UNIT 1 

==SUMMARY==
OF COMPLIANCE ELEMENTS FOR ORDERS            EA-12-049 AND EA-12-051 The elements identified below for CNS Unit 1, as well as the Overall Integrated Plans (OIP) for Orders EA-1 2-049 and EA-1 2-051 (References 2 and 10, respectively), the 6-Month Status Reports for Orders EA-12-049 and EA-12-051 (References 4 thru 8 and 12 thru 16, respectively),
and any additional docketed correspondence, demonstrate compliance with Orders EA-12-049 and EA-12-051.
STRATEGIES    -  COMPLETE CNS, Unit 1 strategies are in compliance with Order EA-12-049. All strategy related Open Items, Confirmatory Items, or Audit Questions/Audit Report Open Items have been addressed and are considered complete pending NRC closure.
MODIFICATIONS     - COMPLETE The modifications required to support the FLEX strategies for CNS, Unit 1 have been fully implemented in accordance with the station design control process. The design of the Spent Fuel Pool Level Instrumentation installed at CNS, Unit I complies with the requirements specified in the order and described in NEI 12-02, Revision 1, "Industry Guidance for Compliance with NRC Order EA-12-051". The instruments have been installed in accordance with the station design control process.
EQUIPMENT   -   PROCURED AND MAINTENANCE & TESTING      - COMPLETE The equipment required to implement the Mitigation Strategies and Reliable Spent Fuel Pool Level Instrumentation has been procured and is ready for use at CNS, Unit 1.
Testing and Maintenance processes have been established through the use of Industry endorsed Electric Power Research Institute (EPRI) Guidelines and the ONS Preventative Maintenance program such that FLEX equipment reliability is achieved.
Operating and maintenance procedures for the Spent Fuel Pool Instruments for CNS, Unit 1 have been developed, and integrated with existing procedures. These procedures have been verified and are available for use in accordance with the site procedure control program. Site processes have been established to ensure the Spent Fuel Pool Instruments are maintained at their design accuracy.
PROTECTED STORAGE      - COMPLETE The storage facility required to implement the FLEX strategies for CNS, Unit 1 has been completed and provides protection from the applicable site hazards. The equipment required to implement the FLEX strategies for CNS, Unit 1 is stored in its protected configuration and is ready for use.
PROCEDURES      - COMPLETE FLEX Support Guidelines (FSG) and procedures for the maintenance and use of the Spent Fuel Pool Level Instrumentation for ONS, Unit I have been developed in accordance with NEI 12-06, Section 3.2.2 and NEI 12-02, Revision 1, Section 4.2. The FSGs and affected existing procedures have been verified and are available for use in accordance with the site procedure control program.
Page 1 of 4

ATTACHMENT 1 CNS, UNIT 1 

==SUMMARY==
OF COMPLIANCE ELEMENTS FOR ORDERS EA-12-049 AND EA-12-051 TRAINING  -  COMPLETE Training for CNS, Unit I has been completed using the CNS Systematic Approach to Training (SAT) as recommended in NEI 12-06, Revision 0, Section 11.6 and in NEI 12-02, Revision 1, Section 4.1.
STAFFING   - COMPLETE The staffing study for CNS has been completed in accordance with NEI 12-01, Revision 0 and 10OCFR5O.54(f), "Request for Information Pursuant to Title 10 of the Code of Federal Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force review of Insights from the Fukushima Dai-ichi Accident," Recommendation 9.3, dated March 12, 2012 (Reference 18), as documented in letter dated May 20, 2014 (Reference 19) and September 24, 2014 (Reference 20). The staffing study confirmed that CNS has adequate staffing to perform the actions to mitigate beyond design basis events.
NATIONAL SAFER RESPONSE CENTERS        - COMPLETE Duke Energy has established a contract with the Pooled Equipment Inventory Company (PEICo) and has joined the Strategic Alliance for FLEX Emergency Response (SAFER)
Team Equipment Committee for off-site facility coordination. It has been confirmed that PEICo is ready to support CNS with Phase 3 equipment stored in the National SAFER Response Centers in accordance with the site specific SAFER Response Plan.
VALIDATION    - COMPLETE Duke Energy has completed performance of validation in accordance with industry developed guidance to assure required tasks, manual actions and decisions for FLEX strategies are feasible and may be executed within the constraints identified in the Overall Integrated Plans (QIP) for Order EA-12-049.
FLEX PROGRAM DOCUMENT        - ESTABLISHED The FLEX Program Document for CNS has been developed in accordance with the requirements of NEI 12-06, Revision 0.
Page 2 of 4

ATTACHMENT 1 CNS, UNIT 1 

==SUMMARY==
OF COMPLIANCE ELEMENTS FOR ORDERS EA-12-049 AND EA-12-051 REFERENCES
: 1. Nuclear Regulatory Commission Order Number EA-12-049, Order Modifying Licensees With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, Revision 0, dated March 12, 2012, ADAMS Accession No. ML12054A735
: 2. Catawba Nuclear Station, Overall Integrated Plan in Response to March 12, 2012, Commission Order to Modify Licenses With Regard To Requirements for Mitigation Strategies for Beyond Design Basis External Events (Order EA-1 2-049), dated February 28, 2013, ADAMS Accession No. ML13066A173
: 3. Catawba Nuclear Station, Units 1&2 - Interim Staff Evaluation -Relating to Overall Integrated Plan in Response to Order EA-1 2-049 (Mitigation Strategies), dated February 6, 2014, ADAMS Accession No. ML13364A175.
: 4. Catawba Nuclear Station First 6-Month Status Report for the Implementation of Mitigation Strategies for Beyond Design Basis External Events (Order EA-1 2-049), dated August 28, 2013, ADAMS Accession No. ML13298A010.
: 5. Catawba Nuclear Station Second 6-Month Status Report for the Implementation of Mitigation Strategies for Beyond Design Basis External Events (Order EA-1 2-049), dated February 28 2013, ADAMS Accession No. ML14065A038.
: 6. Catawba Nuclear Station Third 6-Month Status Report for the Implementation of Mitigation Strategies for Beyond Design Basis External Events (Order EA-12-049), dated August 28, 2014, ADAMS Accession No. ML14247A232.
: 7. Catawba Nuclear Station Fourth 6-Month Status Report for the Implementation of Mitigation Strategies for Beyond Design Basis External Events (Order EA-1 2-049), dated February 26, 2015, ADAMS Accession No. ML15081A124.
: 8. Catawba Nuclear Station Fifth 6-month Status Report for the Implementation of Mitigation Strategies for Beyond Design Basis External Events (Order EA-1 2-049), dated August 26, 2015, ADAMS Accession No. ML15240A066.
: 9. NRC Order Number EA-12-051, Order to Modify Licenses With Regard To Reliable Spent Fuel Pool Instrumentation, dated March 12, 2012, ADAMS Accession No. MLI12054A679.
: 10. Duke Energy Letter, Duke Energy Carolinas, LLC, (Duke Energy) Overall Integrated Plans in Response to March 12, 2012, Commission Order Modifying Licenses With Regard To Requirements for Reliable Spent Fuel Pool Instrumentation (Order EA        051), dated February 28, 2013, ADAMS Accession No. ML13086A095.
: 11. Catawba Nuclear Station, Units 1 and 2 - Interim Staff Evaluation and Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-12-051, Reliable Spent Fuel Instrumentation, dated October 28, 2013, ADAMS Accession No. ML13281A562.
Page 3 of 4

ATTACHMENT 1 CNS, UNIT 1 

==SUMMARY==
OF COMPLIANCE ELEMENTS FOR ORDERS EA-12-049 AND EA-12-051
: 12. Catawba Nuclear Station First 6-Month SFPLI Status Report in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order EA-12-051), dated October 28, 2013, ADAMS Accession No. ML13242A009.
: 13. Catawba Nuclear Station Second 6-Month SEPLI Status Report in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order EA-12-051), dated February 26, 2014, ADAMS Accession No. ML14063A279.
: 14. Catawba Nuclear Station Third 6-Month SFPLI Status Report in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order EA-12-051 ), dated August 14, 2014, ADAMS Accession No. ML14227A717.
: 15. Catawba Nuclear Station Fourth 6-Month SFPLI Status Report in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order EA-12-051 ), dated February 16, 2014, ADAMS Accession No. ML15051A366.
: 16. Catawba Nuclear Station Fifth 6-month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order EA-12-051), dated August 3, 2015, ADAMS Accession No. MLI1521 7A007.
: 17. Catawba Nuclear Station, Units 1 and 2, Report for the Audit Regarding the Implementation of Mitigation Strategies and Reliable Spent Fuel Pool Instrumentation related to Orders EA-1 2-049 and EA-1 2-051, dated February 20, 2015, ADAMS' Accession No. ML15035A679.
: 18. 10OCFR50.54(f), "Request for Information Pursuant to Title 10 of the Code of Federal Regulations, Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force review of Insights from the Fukushima Dai-ichi Accident", Recommendation 9.3, dated March 12, 2012, ADAMS Accession No. ML12053A340.
: 19. Duke Energy Letter, Phase 1 Staffing Assessment pursuant to 10OCFR50.54(f) regarding NTTF Recommendations 9.3, dated April 30, 2013.
: 20. Catawba Nuclear Station Phase 2 Staffing Assessment pursuant to 10OCFR 50.54(f) regarding NTTF Recommendation 9.3, dated October 28, 2014, ADAMS Accession No. MLI 4303A259.
: 21. NEI 12-06, Revision 0 "Diverse and Flexible Coping Strategies (FLEX) Implementation Guide", ADAMS Accession No. ML12242A378.
: 22. NEI 12-02, Revision 1 "Industry Guidance for Compliance with NRC Order EA-1 2-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", ADAMS Accession No. ML12240A307.
: 23. NEI 12-01, Revision 0 "Guideline for Assessing Beyond Design Basis Accident Response Staffing and Communications Capabilities", ADAMS Accession No. ML12125A412.
Page 4 of 4

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS Duke Energy affirms that CNS is in full compliance with Orders EA-12-049 and EA-12-051 as demonstrated by the docketed correspondences concerning these orders. Briefly, CNS FLEX Interim Staff Evaluation (ISE) Open and Confirmatory Items are complete pending NRC closure; CNS FLEX OIP Open Items are complete pending NRC Closure; CNS FLEX Audit Questions are complete pending NRC closure; CNS FLEX NRC Audit Report Open Items are complete pending NRC closure; and the CNS Request for Additional Information (RAl) provided in the Spent Fuel Pool Level Instrumentation (SFPLI) ISE are complete pending NRC closure.
Duke Energy provides the following response for the Audit Report Open and Pending Items and considers them to be complete pending NRC closure for Catawba Nuclear Station:
Iitem      .. -1,         -,   DescriptionII         ,*'...           Sum mary Response,*.,
Procedure Interfaces   - Seismic           Catawba Response:
Confirm completion of evaluation          RfrneAtcmn               o h of potential Aux Building flooding         repn.
Cl 3.1.1.3.A             and appropriate actions andrepn.
procurement of sump pumps.
Room Temperature Analyses                  C-ab        epne Auxiliary Building, SFP Building           Reference Attachment 3 for the and Control Room response.
Room temperature analyses being CI 3.2.4.1 .A            performed will provide a better idea of the environmental conditions expected during the event. Confirm completion of analyses and appropriate actions.
Freee PotecionCatawba Response:
Evaluations to address the needs           Reference Attachment 3 for the Cl 3.2.4.3.A             for freeze protection are in               response.
progress. Confirm completion of evaluations and appropriate
____     ____    ___   actions.
LighingAnalsesCatawba Response:
Cl 3.2.4.4.A             Confirm evaluations for additional         Reference Attachment 3 for the lighting have been completed and           response.
appropriate actions taken.
Offsite Resources   - Confirm NEI          CtabRepn:
12-06 Section 12.2, Guidelines 2           RfrneAtcmn               o h through 10 are addressed with Cl 3.4.A              SAFER                                      response.
Complete SAFER Site specific
_____      _____    plan.                                     _  _  _ _  _ _ _ _ _ _  _  _ _  _ _
Page 1 of 10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS Catawba Response:
Westinghouse Electric Company submitted letter LTR-RES-1 3-1 53
("Documentation of 7228C 0-Rings at ELAP Conditions") on October 31, 2013. The letter documents a Westinghouse evaluation of compound 7228C RCP 0-rings at ELAP conditions up to 582&deg;F (the same 0-rings in use at Catawba),
and concludes that they will not fail Reactor Coolant Pump 0-rings during an 8-hour ELAP event w/o seal cooling. The 0-rings survived Confirm that, beyond order for an average of 18 hours, with the compliance date, plant will use first failure occurring at 13 AQ.26d             only high-temperature-qualified 0-hours. The Catawba FLEX strategy rings where applicable, or that for RCS make-up post-ELAP is to steam generator relief valve will be cooldown well below 582&deg;F and operated to control temperature to align a FLEX portable RCS make 550 degrees F or below.
up pump within about 12 hours. Therefore, the Catawba RCP 0-ring integrity will be maintained and the assumption of a 21 gpm/pump leak rate remains valid.
Reactor coolant pump seal drawing has been posted to the Catawba Fukushima Sharepoint noting high temperature o-rinp material.
Direct Current (dc) load profile,    CawbRepn:
load shedding, and dc bus voltage. Catawba understands the NRC will AQ.37 The staff will complete a vendor     be auditing the battery vendor to audit of the batteries.              close this item.
Catawba Response:
FLEX Diesel Generator Sizing         EniergChgevauto AQ.47             Sizing calculations need to be       EC401541, FLEX Diesel Generator completed and placed on Catawba      Loading Evaluation has been Fukushima Sharepoint.                completed and placed on the Catawba Fukushima Sharepoint.
Battery Room Ventilation   -         Catawba Response:
TempeatureCatawba                     has purchased sufficient Provide information on the           fans and spot coolers to perform AQ.49             adequacy of the ventilation          Phase 1, 2, and 3 strategies, provided in the battery room to      provide sufficient cooling for protect the batteries from the       equipment operation/personnel
_____________effects         of extreme high and low      habitability, and Hydrogen gas Page 2 of 10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS temnperatu res.                        control based on a GOTHIC analysis performed by Zachry (Reference CNC -1211.00-00-01 46
                                                     - Gothic Analysis For Extended Loss Of All AC Power (ELAPIFLEX)). Procedural guidance has been developed to implement the mitigation recommendations in the Zachry analyses. In addition, per Table 7-1 in the SAFER Response Plan and Table 9-1 in the National SAFER Response Center Equipment Technical Requirements Document (51-9199717-013), Catawba will receive one 3000 cfm ventilation fan and associated ducting per Unit from the NSRC. This ventilation equipment from the NSRC will be used on an as needed/as desired basis to provide additional air flow to any area in the plant.
Calculation CNC-1 2t1.00-00-0146 Gothic Analysis For Extended Loss Of All AC Power (ELAP/FLEX) has been placed on the Catawba Fukushima Sharepoint for review.
Catawba Response:l Diesel Fuel Oil Supply and Quality     Fe     i   osmtosoe Describe plans for supplying fuel      volume, quality, and the need for oil to FLEX equipment (i.e., fuel oil  refueling have been addressed in AQ.50       storage tank volume, supply            calculation CNC-1612.03-00-0001, pathway, etc.). Also, explain how      FLEX Fuel Consumption fuel quality will be assured if stored Calculation. This calculation has for extended periods of time.          been posted to the Catawba Fukushima Sharepoint.
Catawba Response:
Battery Room Ventilation    -          Catawba has purchased sufficient Hydrogen Accumulation Potential        fans and spot coolers to perform Provide a discussion of battery        Phase 1, 2, and 3 strategies, room ventilation to prevent            provide sufficient cooling for AQ.51       hydrogen accumulation while            equipment operation/personnel recargnghebattere in phase 2      habitability, and Hydrogen gas th aeiescontrol rechrgig                             based on a GOTHIC or 3. In your response, include a description of the exhaust path if it  analysis performed by Zachry is different from the design basis.    (Reference CNC -1211.00-00-0146
                                                     - Gothic Analysis For Extended
___________________________________Loss Of All AC Power Page 3 ofl10

ATTACHMENT 2 CNS NRC   AUDIT REPORT OPEN AND PENDING ITEMS (ELAP/FLEX). Procedural guidance has been developed to implement the mitigation recommendations in the Zachry analyses. In addition, per Table 7-1 in the SAFER Response Plan and Table 9-1 in the National SAFER Response Center Equipment Technical Requirements Document (51-9199717-013), Catawba will receive one 3000 cfm ventilation fan and associated ducting per Unit from the NSRC. This ventilation equipment from the NSRC will be used on an as needed/as desired basis to provide additional air flow to any area in the plant.
Calculation CNC-121 1.00-00-0146, Gothic Analysis For Extended Loss Of All AC Power (ELAP/FLEX), has been placed on the Catawba Fukushima Sharepoint for review.
Catawba Response:
The following flow model calculations have been posted to the Catawba Fukushima Sharepoint; CNC-1223.02-00-0025, Cooling Water Flow Model              Flow Model of SNSWP to CA Develop flow model calculations to    Connections For Phase 2 FLEX Strategies, CNC-1 223.02-00-0026, support the various FLEX OIP.32                                       Flow Model of SNSWP to RN strategies and document the Connections and CA for Phase 2 available static water volume in the FLEX Strategies, CNC-1223.02        RN/CA piping.
0027, Flow Model of RN to KC Hxs to Support RHR For Phase 3 FLEX Strategies, and CNC- 1223.02                                             0028, Flow Model For UI/U2 NI Portable Pump Injection to RCS Phase 2 & 3 FLEX Strateaies.
Catawba Response:
Determine Need for Containment Spray                                Calculation DPC-1 552.08-00-0280, Extended Loss of AC Power An analysis is needed to determine   (ELAP)    - Ice Condenser OIP.42 if containment spray for             Containment Response with FLEX temperature/pressure control is not  Mitigation Strategies, has required over the long term.         determined containment spray is not needed for long term temperature/pressure control.
Page 4 of 10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS Calculation has been posted to the Catawba Fukushima Sharepoint.
Catawba Response:
Engineering change EC1 09414 and EC1 10934 were completed during SFP Insrumnts1 eve                      EOC22. EC109414 installed the OIP.43            Provide redundant SEP level                   primary spent fuel pool level instruments                                  indication and EC1 10934 installed the backup spent fuel pool level indication.
Catawba Response:
Staffing                                     Phase 2 staffing assessment OIP.56            NRC will review the Phase 2                  provided October 28, 2014.
Staffing Assessment.                          Reference ADAMS Accession No. ML14303A259.
SFP Level Instrument Mounting                Catawba Response:
SEPI       Complete the "Intermediate SFI4Waveguide Mounting Support"                         Reference Attachment 5, Item 13.
calculation and place on the Catawba Fukushima Sharepoint.
SEP Level Instrument Mounting                CawbRepn:
Complete the "Intermediate                   Calculation CNC-1 139.14-08-0001, SFPI.6            Waveguide Mounting Support"                  Waveguide Antenna Support, has calculation and place on the                 been added to the Catawba Catawba Fukushima Sharepoint.                Fukushima Sharepoint SEP Level Instrument Reliability             Catawba Response:
Complete the following and place             Reference Attachment 5, Items 14, on the Catawba Fukushima                     20 and 22.
Sharepoint for review;
                        -The      justification for the shock and vibration test           PRCD         E deviation
                        -An     assessment of potential        The calibration procedure for the SFPI.7                      susceptibilities of                Spent Fuel Pool Level Back Up electromagnetic\radio              instrument (I P/11A/13120/031, frqenyinterference              Backup Spent Fuel Pool Level) has frEqueF)ncy es                  hr been completed and a copy has theMISFP insarueas wher            been posted on Catawba thekuSFPminstrumentnis located and how to mitigateFushmSaepit those susceptibilities
                        -The      calibration procedure        ARE VA Document # 5 1-9202556-
______________for               the Back-up SFP Level          005, Qualification Analysis of Page 5 ofl10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS Monitoring differential    VEGAPULS 62 ER Through Air pressure transmitter        Radar, has been posted to the Catawba Fukushima Sharepoint as well.
SEP Level Instrument Qualification   Catawba Response:
Complete the calibration procedure   IPI11A131201031, Backup Spent for heSP ackup LvelFuel           Pool Level has been SFPI.8    Monitoring differential pressure    completed and posted to the transmitter and place on Catawba     Catawba Fukushima Sharepoint.
Fukushima Sharepoint for review.
Catawba Response:
SFP Level Instrument Calibration Complete the calibration procedure   IPI11AI3120/031, Backup Spent SFPI.1 1  for the Back-up SEP Level            Fuel Pool Level has been Monitoring differential pressure    Completed and posted to the transmitter and place on Catawba    Catawba Fukushima Sharepoint.
Fukushima Sharepoint for review.
SEP Level Instrument Calibration    Catawba Response:
Complete the calibration procedure SFPI.12   for the Back-up SFP Level            IP/1/A/3120/031, Backup Spent Monitoring differential pressure     Fuel Pool Level has been transmitter and place on Catawba    completed and posted to the Fukushima Sharepoint for review. Catawba Fukushima Sharepoint.
SEP Level Instrument                Catawba Response:
Maintenance and Test IP/11A/3120/031, Backup Spent SFPI.15   Complete the calibration procedure  Fuel Pool Level has been for the Back-up SEP Level            completed and posted to the Monitoring differential pressure    Catawba Fukushima Sharepoint.
transmitter and place on Catawba Fukushima Sharepoint for review.
RCP Seal LeakageCawbRepn:
This is an ongoing generic NRC Please provide adequate             issue related to the ELAP RCP seal justification for the seal leakage   leakage issue and the RCP seal rates calculated according to the    model used in evaluating the LOSC SE.3    Westinghouse seal leakage model      response. To address this item, the that was revised following the       NRC requested sites using issuance of NSAL-14-1 or             standard RCP seal packages to alternative model (e.g., MPR). The   provide a RCP Seal Leakage justification should include a       Margin Assessment paper. This is discussion of the following factors: included as Attachment 7.
Page 6 of 10

.1 ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS a) benchmarking of the seal        Reasonable assurance of leakage model against        compliance with endorsed guidance relevant data from tests or  is achieved via in-house operating events,            evaluations confirming Catawba's FLEX strategies are bounded by b) discussion of the impact on     the WCAP-17601-P, revision 1 the seal leakage rate due     reference case as well subsequent to fluid temperatures         PWROG evaluations. As such, greater than 5500&deg;F           closure of this issue is not a resulting in increased       requirement for Unit startup.
deflection at the seal        Calculation CNC-1223.04-00-01 17, interface,                   MPR RCP Seal Leakoff ELAP/FLEX Analysis has been c) clarification whether the       placed on the Catawba Fukushima second-stage reactor          Sharepoint as supporting coolant pump seal would       documentation.
remain closed under ELAP conditions predicted by the   Duke Fleet Fukushima revised seal leakage model    Response/PWROG continue to and a technical basis to     work with the NRC to close this support the determination,    generic issue.
and, d) justification that the interpolation scheme used to compute the integrated leakage from the reactor coolant pump seals from a limited numb~er of computer simulations (e.g., three) is realistic or conservative.
Catawba Response:
This is an ongoing generic NRC issue related to the ELAP RCP seal leakage issue and the potential RCP Seal Leakoff Piping               rupture of the #1 seal leak-off line.
The current NRC position is that the Perform additional piping and leak-off piping should maintain support analysis to ensure the seal leakoff line temperature and          integrity up to 2500 psia. To SE.4                                       address this item, the NRC pressure seen during the ELAP event will be supported by the        requested sites using standard RCP seal packages to provide a current configuration.
RCP Seal Leakage Margin Assessment paper. This has been included as Attachment 7.
Reasonable assurance of compliance with endorsed auidance Page 7 of 10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS is achieved via in-house evaluations confirming Catawba's FLEX strategies are bounded by the WCAP-1 7601 -P, revision 1 reference case as well subsequent PWROG evaluations. As such, closure of this issue is not a requirement for Unit startup.
Calculation CNC-1223.04-00-01 17, MPR RCP Seal Leakoff ELAP/FLEX Analysis has been placed on the Catawba Fukushima Sharepoint as supporting documentation.
Duke Fleet Fukushima Response/PWROG continue to work with the NRC to close this generic issue.
t                                     +/-
ELAP Calculations with                 Catawba Response:
NOTRUMP Please provide adequate basis         The PWR Owners Group has that calculations performed with       issued report PWROG-14064, the NOTRUMP code (e.g., those in      "Application of NOTRUMP Code WCAP-1 7601 -P, WCAP-1 7792-P)         Results for Westinghouse are adequate to demonstrate that       Designed PWRs in Extended Loss criteria associated with the          of AC Power Circumstances". This analysis of an ELAP event (e.g.,      report compares E LA P avoidance of reflux cooling,          predictions from the NOTRUMP promotion of boric acid mixing) are   and TRACE computer codes. The satisfied. NRC staff confirmatory     results show that the NOTRUMP analysis suggests that the need for   predicted results for the onset implementing certain mitigating       of reflux cooling and the SE.5   strategies for providing core         loop flow under two-phase cooling and adequate shutdown         conditions agree well or are margin may occur sooner than          conservative with respect to the predicted in NOTRUMP                  TRACE predicted results. The simulations.                          comparison shows that NOTRUMP provides a conservative estimate of the required time when the primary make-up pumps are required for an ELAP event as compared to TRACE. Therefore, it is concluded that NOTRUMP is acceptable for simulation of the ELAP event within the constraints listed herein with regards to reflux coolinq and boron mixinq.
Page 8 of 10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS Duke has also performed independent predictions of the ELAP event for both Catawba units using RELAP, and the results are in good agreement with the NOTRUMP results in WCAP-17601-P.
Catawba Response:
The generic 4-loop NSSS model used in WCAP-1 7601 is representative of Catawba Unit I RCS Cooling & RCS Inventory          on all parameters. Each parameter Control Analysis                     of interest in LTR-LIS-14-219 in Table 1C has been reviewed and Justification is needed to show that  no additional assessment is the WCAP 17601-P analysis used        required. As such, the time to for CNS is representative and         reflux cooling provided in PWROG-SE.8 bounding of the actual plant          14027-P is applicable.
conditions for the ELAP event. In    Catawba ELAP parameters table particular the time to reach reflux  has been added to the Catawba cooling needs to be justified.       Fukushima Sharepoint. The table contains Westinghouse analysis values and the corresponding Catawba Unit I parameters.
Supporting calculations have been added to the Catawba Fukushima Sharepoint as well.
Catawba Response:
RCS Cooldown and Prevention of Nitrogen Injection for Safety        CNS uses setpoint 0.08 (setpoint Injection Accumulators               0. 11 was calculated for CNS, but resulted in the same value, Specify whether CNS is using EOP     therefore no benefit in using the Setpoint Number 0-8 or 0-11 for       substitute).
the cooldown. Also, the licensee needs to specify whether the plan     FSG-10 (Cold Leg Accumulator for isolation of the Cold Leg          Isolation) is invoked as part of SE.9 Accumulators occurs prior to           EP/I/A/5000/ECA-0.0, Loss of All cooling down to Emergency              AC Power, and isolates CLAs prior Operating Procedure setpoint 0-12      to reaching EP setpoint or 0-13. Provide the footnote          O.13. Cooldown will be initiated calculations for whichever             only when the CLAs are isolated or setpoints are being used on the        vented. ECA-0.0 has been added Catawba Fukushima Sharepoint           to the Catawba Fukushima for review.                            Sharepoint. CNS uses footnotes 0.08 and 0.13 for target SG pressures durino ELAP recovery.
Page 9 ofl10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS Excerpts from the Catawba Unit 1&2 EP setpoint calculation CNC-1552.08-00-0195, Emergency Procedure Setpoints, has been added to the Catawba Fukushima Sharepoint. Calculations for F.05 and 0.07 were included as they provide input into the evaluation of Footnote 0.08.
Page 10 of 10

ATTACHMENT 3 CNS RESPONSE TO DIVERSE AND FLEXIBLE STRATEGIES INTERIM STAFF EVALUATION OPEN AND CONFIRMATORY ITEMS Duke Energy provides the following response to the Interim Staff Evaluation (ISE) open and confirmatory items contained in NRC Letter, "CNS - Interim Staff Evaluation Relating to the Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies), (Agency-wide Documents Access and Management System (ADAMS) Accession No. ML13364A175).
ISE CI                                                                  Catawba Response:
3.1.1.3.A                                                                 Sump pumps have been procured and strategies Procedure Interfaces  -  Seismic                            developed to address internal flooding issues.
Calculation CNC-1 206.03-00-0001, Flood Levels for Confirm completion of evaluation of potential Aux          Structures Outside of the Reactor Building, has been Building flooding and appropriate actions and               completed and posted to the Catawba Fukushima procurement of sump pumps.                                  Sharepoint. FLEX Support Guidelines FG/OIAICFLX/FSG-20 (Electrical Distribution) and FGIO/AICFLX/FSG-22 (Sump Pump Operation) provide electrical setup and placement of sump pumps. Both are located on the Catawba Fukushima Sharepoint.
ISE Cl                                                                  Catawba Response:
3.2.4.1 .A                                                                Catawba has purchased sufficient fans and spot coolers to perform phase 1, 2, and 3 strategies, provide sufficient cooling for equipment operation/personnel Room Temperature Analyses      - Auxiliary Building, SFP    habitability, and hydrogen gas control based on Gothic Building and Control Room                                   analysis performed by Zachry (Reference CNC-121 1.00-00-0146 - Gothic Analysis for Extended Loss Of All AC Room temperature analyses being performed will              Power (ELAP/FLEX)). Procedural guidance has been provide a better idea of the environmental conditions       developed to implement the mitigation recommendations expected during the event. Confirm completion of            in the Zachry analyses. In addition, per Table 7-1 in the analyses and appropriate actions.                           National SAFER Response Center Equipment Technical Requirements Document (51-9199717-013), Catawba will receive one 3000 cfm ventilation fan and associated ducting per Unit from the NSRC. This ventilation
_________    _____________________________________equipment from the NSRC will be used on as needed/as Page 1 of 3

ATTACHMENT 3 CNS RESPONSE TO DIVERSE AND FLEXIBLE STRATEGIES INTERIM STAFF EVALUATION OPEN AND CONFIRMATORY ITEMS OpenIltem "-           -*!        DesCription -*:'*"'*      ..*"::,-;i:      ::     :-    .'   Response * :-&deg;"          ,   ,,   iil desired basis to provide additional air flow to any area in the plant.
ISE CI                                                               Catawba Response:
3.2.4.3.A Other than components associated with monitoring FWST level, there is no other Flex related instrumentation located in the yard. The FWST level instrumentation is exposed and is normally freeze protected. As noted below, heat tracing equipment is kept in the Flex storage building and can be used to keep the FWST level instrumentation functional in extreme freezing conditions as directed by the FSGs.
Some credited instrumentation components are located in the Doghouses, but they will not require freeze Freeze Protection                                           protection. During cold weather conditions the window Evaluations to address the needs for freeze protection      setosfthDghuesaecvrdbcrai.
are n pogrss.Confrm ompetin ofevauatonsandThis,             combined with heat by steam piping used in the actonfir cmltoofeauinsnd Phase I and 2 response, eliminates freezing of areinprogriaess instruments or impact to supporting equipment. Flex piping connections were reviewed and none were found to be affected by outside freezing conditions. Auxiliary Feedwater and Steam Generator Wet Layup (BW) connections in the Doghouse are "protected" in the same manner as the instrumentation discussed above.
Most Flex piping connections are located in the Auxiliary Building where freezing is not credible. One train of RN connection is located outside, but within a below grade bunker. Inventory used from the FWST enters the plant through the normal system piping. Based on initial conditions and the design of the piping system, freezing Page 2 of 3

ATTACHMENT 3 CNS RESPONSE TO DIVERSE AND FLEXIBLE STRATEGIES INTERIM STAFF EVALUATION OPEN AND CONFIRMATORY ITEMS Openn Item tm                        Description !,J**,!i*    *,* *ii***,***:     i.*:i , ,* /,*:: * *'** Responseecipin es o s s?",*i' /,:*:,*,*:**.*  *!
is not projected to be a problem. Only the piping in the trench leaving the FWST is considered for use in the FLEX response. This piping along with the FWST is judged to be weather protected to the extent that freezing will not occur before FLEX strategies are implemented. The Fuel Building will remain above freezing due to the constant heat source of the spent fuel in the pool.
As a contingency, Catawba has a roll of heating cable stored in the protected storage building. This self-regulating heat trace cable is provided along with the accessories for assembling multiple freeze protection circuits. These heating cables could be powered from any portable DIG supplying 120 VAC power. This equipment could be deployed if an unanticipated freeze protection need developed.
ISE Cl    Lighting Analyses                                              Catawba Response:
3.2.4.4.A Confirm evaluations for additional lighting have been          Lighting assessment has been completed and is completed and appropriate actions taken,                      available on the Catawba Fukushima Sharepoint for review.
ISE CI    Offsite Resources - Confirm NEI 12-06 Section 12.2,            Catawba Response:
3.4.A Guidelines 2 through 10 ar'e addressed with SAFER Complete SAFER Site specific plan.                            Approved CNS SAFER Plan, CNSRP-1612.03-01, has been posted to the Catawba Fukushima Sharepoint for review.
Page 3 of 3

ATTACHMENT 4 CNS   RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING THE OVERALL INTEGRATED PLAN FOR IMPLEMENTATION OF ORDER EA-12-051, RELIABLE SPENT FUEL POOL INSTRUMENTATION The CNS Response to the Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-12-051, Reliable Spent Fuel Pool Instrumentation was provided initially to the NRC in Duke Letter, Duke Energy Carolinas, LLC, (Duke Energy),
Response to Request for Additional Information Regarding Overall Integrated Plan in Response to order EA-12-051, "Reliable Spent Fuel Pool Instrumentation" dated July 23, 2013, (ADAMS Accession No. ML13206A384).
The NRC provided an Interim Staff Evaluation (ISE) and Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-12-051, Reliable Spent Fuel Instrumentation dated October 28, 2013 (ADAMS Accession No. ML13281A562). The Requests For Additional Information (RAls) were renumbered by the ISE.
The ONS Response to this Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-12-051, Reliable Spent Fuel Instrumentation was provided through the Scientech eDocs Web Portal as allowed by NRC Letter, Nuclear Regulatory Commission Audits of Licensee Responses to Reliable Spent Fuel Pool Instrumentation Order EA-12-051, dated March 26, 2014 (ADAMS Accession No. ML14083A620).
No further information is required for this attachment.

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION 7TicParameter:                          VedrDein                    Aditommnal    e
                                   .. ,     ,st                    ReferenceTo                         Te, or Analysis Results      -=   ,,         Licensee Evaluation-<         ::
Duke Technical CutmrRequirements                                                                       The vendor instrumentation design was Design                             technicalnt           DouetDR-                    E-20                     N/A               reviewed and determined to adequately meet requirementsn              1336.04-00-0001              NE1-1 2-02 specfictio 02Duk POthefo            Rev                                         specification requirements.
SFPLI 171968 Rev, 003 Qualification is based on a combination of tests and analyses or                                                                                          The vendor qualification documentation was siiart.EA-12-051,                                      Test and analysis results    reviewed and concluded to adequately
:2   Test Strategy                                         Qualification analysis              1.4      meet requirements of EA      demonstrate the instrumentation could reliably Qulfcain          Doc. 51-9202556-005            NE1-12-02,   12-051, JLD-ISG-2012-03,      function in its installed environment(s) during a testsfiatind                                                   3.4        and NEI 12-02 Rev. 1       postulated Beyond Design Basis External analyss are                                                                                          Event (BDBEE).
summarized in qualification analysis report 51-9202556-005.
Tempratre rtin of      The primary channel instrumentation electronics are located outside the SEP area.
Power Control Panel is      Thvedrisumnaoneig qualiicaton Qulifiatio fr                  Anaysis     EI1-02,   in te pael aove mbieT.hxped           venvior nsrmentaltcondiin    d   uringna 3 Electronicsta              Temperature and9F                                             34                               rse     emDoctre n51-92ityl2556-oun enclosureeetoisulfcin with fo ~     meauehumidity an         ulfcto o.005, 5-256-Section Anlss2.3                    NEMA41-0, n4XeNEA       enclosuree ncosreamie    postulatedetepotuatdeviBDBEE.fo BBE.i     Seein eeSection fecio    o this ds7liiof  hi display                                                                                          prevents moisture intrusion,. ouetfrdsuso                 fds iis Radiation withstand       Refer to RAI #7 response posted to Catawba analyzed to 1x1 03 rads     Fukushima Sharepoint.
Page 1 ofl1l

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND                                  CRS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION Measurement capability through saturated steam and smoke.                                               Initial testing (without horn Testing                                                  cover) demonstrated performed to                                             successful measurement demonstrate the                                          capability through steam radar horn cover                                         and smoke. Subsequent was effective at                                         testing of the radar horn and The final radar horn cover qualification report preventing                                               cover demonstrated            was not available when Catawba PAl moisture                                                 adequate operation during     responses were submitted. Areva Report No.
intrusion within    Qualification Analysis               sustained simulated SEP       51-9221032-000 is the "Qualification Analysis Evrnetl               the horn and                 Doc.                        boiling conditions, and that  For Vega Waveguide Horn Cover". The radar tsigfor level           wave guide pipe.       51-9202556-005,                   the horn cover was effective  horn cover qualification testing adequately tstingo                                           Section 2.3, 2.4, 2.5,   EA-12-051, in preventing moisture        demonstrated acceptable operation during sesrRadar                     horn         2.7, Appendix B and          1.4    intrusion within the horn and exposure to simulated SEP boiling conditions.
4opnet            n         cover (fused        supporting references     NEI 12-02, wave guide pipe.
SE ra-                  silic gls)              690860234The                                                        horn cover adhesive manufacturer Saturated steam                   cagas,6-204-234 metal  waveguide       51-9220845-001                    The horn cover adhesive is    radiation test data adequately demonstrated
        &Rdain                pipe and horn          51-9221032-000                    a silicone elastomer          the adhesive would not experience are not                66-9225632-000                    manufactured by Dow           unacceptable degradation for radiation susceptible to                                           Corning (Sylgard 170). The    exposure in excess of that expected for the radiation,                                               adhesive manufacturer         postulated beyond design basis event over the degradation.                                             radiation test data           required mission time.
Manufacturer                                             adequately demonstrates test data                                                the adhesive would not supports                                                 experience unacceptable acceptable                                               degradation for exposures radiation                                                up to 1.64 Xl 0* Rads.
degradation resistance for the radar horn cover
______________adhesive.                    ______________________________
Page 2 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND                             CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION Qualification Analysis Doc. 51-9202556-005, Sections 2.3, 2.5, Appendix A and                    Sno         n    niao          r tsigfrlvl         Temperature and   spotnrerncsdemonstrated                                  to withstand senvionmna                                   3E8-6008-21-3000          E20,      the      manufacturer ratings       Refer to RAI # 7 response posted to Catawba 5 electronics                and analysis of     EN 60529:2000,             3.4   ' 80&deg;C (sensor) and            70&deg;C  Fukushima Sharepoint. See Section 7 of this huig-sensor                 and                                           (indicator),    100%          RH. document for discussion of dose limits.
houtsing SF-niao                             38-9218214-000,Raitowthan otie FPidctr                USNRC Bulletin 79-                     analzdia tion         wxithsr a nds 01B Table C-i,anledtlx0rds NUREG-1 73, Vol.1, Section 3.11.3.2.1, Reg. Guide 1.209 Thermal and radiation aging Radar horn                                                not applicable to metal cover (fused                                             waveguide in SFP area.
silica glass),                                           The horn cover adhesive is         The glass and metallic instrumentation metal waveguide                                           a silicone elastomer              components located within the SEP area are pipe and horn                                             manufactured by Dow               not susceptible to aging due to thermal and/or are not                                                   Corning (Sylgard 170). The        radiation effects.
Thermal &                susceptible to    Qualification analysis     EA1-5,       radiation test data               The horn cover adhesive manufacturer Radiation Aging         rdainDoc.                 51 -9202556-           14       adequately demonstrates           radiation test data adequately demonstrated cmoneants              dgatinn              051-Scto22102-0         NEI 12-02,  the adhesive would not             the adhesive would not experience cmoetin          Horn cover           5192020034                          experience unacceptable            unacceptable degradation for radiation SEP area                 adhesive             66-9225632-000              34      degradation for exposures          exposure in excess of that expected for the manufacturer                                              up to 1.64 xl 0 Rads. The         postulated beyond design basis event over the radiation test                                            silicone adhesive is rated to     required mission time. The horn cover data and                                                 withstand temperatures             adhesive temperature ratings are acceptable temperature                                              extremes of -45 to 200&deg;C,          and readily bound the expected conditions for withstand                                                which adequately bound the         the postulated beyond design basis event.
specifications,                                           postulated temperatures for
                                                 ~sustained                                         SEP boiling conditions.                        _____________________
7    Basis for Dose          SFPLI remote      AREVA Document                         A Catawba calculation              A location specific dose calculation was Requirement             transmitter,      No. 51-9202556-005,        NEI 12-02,  shows that the electronics         performed for the remote electronics, which indicator and     Qualification Analysis         3.4      require periodic                  indicated these components must be replaced
___________power                  control   of VEGAPULS 62 ER                      replacement.                       in less than 6 years. A periodic replacement Page 3 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION
         # I   "' I Topic    Parameter TpcSummary                I      ~~~Reference Vendor iDocument      Design
                                                                #        Additional CommTests Coment eutis Tes orAayssRsut.iene.vauto o nay r____________________________      iese        auto panel                                                                                             for the transmitter, power control panel and lxl10 3 qualified rads      to      Through Air Radar 51-9221032-000                                                          local indicator has been set up. Site specific The horn cover adhesive based on                  66-9225632-000                          manufacturer radiation test     radiation dose data is being collected for a industry                                                          data adequately                 future revision of the dose calculation. Refer operating                                                         demonstrates the adhesive       to RAI #7 response posted to Catawba experience.                                                       would not experience            Fukushima Sharepoint.
unacceptable degradation Based on                                                          for exposures up to 1.64        The horn cover adhesive manufacturer engineering                                                       xl0 8 Rads.                     radiation test data adequately demonstrated judgment, the                                                                                     the adhesive would not experience expected total                                                                                    unacceptable degradation for radiation integrated dose                                                                                   exposure in excess of that expected for the for the radar                                                                                     postulated beyond design basis event over the horn cover                                                                                        required mission time.
adhesive would not exceed 1 x108 over the required mission time for the instrumentation.
                                         -t                                                                         +
Seismic withstand             Qualification analysis Doc. 51-9202556-capability of                                                  VEGAPULS 62 ER sensor,         The vendor instrumentation seismic testing 005, Section 2.1,                       indicator, power control     adequately demonstrates the equipment is VEGAPULS 62 Appendix D and                         panel, mounting brackets,     capable of reliably operating during a seismic Seismic                ER sensor,                                       NEI 12-02, supporting references                      and waveguide pipe are       event.
Qualification       indicators, power                                         3.4 11-9203036-000,                       seismically qualified to RRS control panel, IEEE STD 344-2004,                          levels from EPRI TR-       Refer to RAIs #7 and #8 responses posted to mounting EPRI TR-1 07330,                                  107330             Catawba Fukushima Sharepoint.
brackets, 174-9213558-006 waveQuide pipe NRC RAls                  Sloshing analysis                       Seismic induced sloshing indicated a SEP           was performed by an             NA      analysis concluded that the     Sloshing analysis determined seismic induced seismic induced           alternate vendor than           NA      available SEP free-board        wave would not impact radar horn.
sloshing analysis         the supplier of the      _______readily         enveloped the           _____________________
Page 4 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION Parameter        Vendor Design         Additional
     #Topic.                        Summary             Reference          Comments       Test' or Analysis*   Results                  Licensee...Evaluation Document #               _-__-                                                 .....
is required. If   radar level                           maximum predicted wave wave impact is    instrumentation                       height. The analysis predicted, then   (Reference Catawba                    determined wave impact on the               calculation CNC-                      the radar horn would not hydrodynamic      1336.04-00-0001).                     occur.
forces should be included in the mounting design loading
______________combinations.
SetFePol      FntoaiyVEGA                 Test                                                     The vendor factory acceptance test SetFePol tsigProcedure                     ARE VA                               dmntaetht           demonstrated reliable operation of the SFP isrmnain            performed during                                               the SFPLI met the        lelintu             naioudrnrmlcdtos
]0   system                                                 000,              N/A                                       and under various simulated test conditions functionality test             the factory    Factory Acceptance                        specification functional acceptance test. TsReotAEArequirements.                                               (e.g. steam exposure). The testing procdur   TestReprt See#16 AEVAdemonstrated                                                               the instrumentation met design poeueSe#6                      Doc. 66-9227809-002                                                   accuracy and repeatability specifications.
Waveguide radar horn is not immersed in SFP water and therefore not susceptible to boron accumulation. During      Licensee concurs that the wave guided radar Sloshing analysis                     postulated SFP boiling,         instrumentation located in the SFP area is not was performed by an                   boron is not transported by     susceptible to degradation due to postulated alternate vendor than                  rising steam/vapor,.oo                  ul-p           h    aegie aa o the vendor whom                                                       brnbidu.Tewv                      uddrdrhr 11]  Boron Build-Up                     N/A       supplied the radar          N/A       Seismic induced sloshing        is elevated above the SFP process and would level instrumentation                 analysis concluded that the     not be susceptible to boron build-up on the (Reference Catawba                    available SFP free-board        horn during postulated SFP boiling conditions, calculation CNC-                       readily enveloped the          nor is it credible that boron crystal 1336.04-00-0001).                      maximum predicted wave         accumulation on the perimeter of the SFP height. The analysis            walls would impede the radar signal strength.
determined that a seismic induced wave would not impact the radar horn.         _____________________
Seismic     Qualification analysis                      Sensor brackets and       Refer to RAIs #4 and #5 responses posted to Pool-side                       withstand    Doc. 51 -9202556-005     NEI 12-02,         electronic enclosure      Catawba Fukushima Sharepoint.
12   Bracket Seismic                              and supporting              34mounting               are seismically Analysis                                     reference                                   qualified to EPRI TR-     The test and analysis of the horn cover and 174-9213558-006,                       107330 or site-specific RRS. adhesive demonstrate adequate seismic Page 5 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND                                CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION
* Prmtr                   Vendor Design          Adtoa
    '           Toi
         #* TopicSummary      PrmtrReference           Document #            C Adtommnts
                                                                            ,omns..        Test or Analysis Results                   Licensee Evaluation                 "
Tests and                 Calculations                                                          withstand capability.
analyses were             32-9208751-002,                        Testing and analysis of horn performed for the         51-9221032-000                          cover and adhesive support    The stress analysis does not need to consider horn cover and            66-9225632-000                         the components can tolerate    hydrodynamic sloshing forces in the design of adhesive to               32-9221237-003                             horizontal and vertical    the mounting brackets. The sloshing analysis demonstrate               66-9225469-000                            accelerations up to 100~g   detenrmined seismic induced wave would not adequate                                                           and SEP sloshing loads up    impact radar horn.
seismic                   Sloshing analysis                                 to 3.37 psi.
withstand                 was performed by an capability,               alternate vendor than                     Seismic induced sloshing the vendor whom                          analysis concluded that the Perform seismic           supplied the radar                        available SEP free-board induced sloshing          level instrumentation                       readily enveloped the analysis to               (Reference Catawba                        maximum predicted wave assess                    calculation CNC-                             height. The analysis hydrodynamic              1336.04-00-0001 ).                      determined wave impact on wave force on                                                       the radar horn would not the radar horn.                                                                occur.
Qualification analysis Doc. 51-9202556-Additional                   Seismic             005, Section 2.1,Sesrbaktan Brces(esr        withstand of             Appendix D andSesrraetan 13 Eetoisad           sensor brackets         supporing0reerence         NEI 12-02,        elcrncecoue              Refer to PAls #4 and #8 responses posted to Electronics n           and electronic           11-9203036-000,             3.4        mounting are seismically    Catawba Fukushima Sharepoint.
Enclosure)       13        eclosre mosuntin               P~iT-1 0330qualified 17-23558-006                                        to EPRI TR-107330 or site-specific   RRS.
mountingCalculations 32-9208751-002, 32-922 1237-003 Shock and          Qualification Analysis                 Sensor, displays, and power    The shock and vibration testing performed for vibration           Doc. 51 -9202556-                    control panel have been        the SEP level instrumentation adequately withstand testing         005, Sections 2.2 and                  tested and/or analyzed for     demonstrates the sensor and power control 14    irtoShock
             &           sesrand analysidsplysfOr supporting i__0references,     NEI 12-02, shock and vibration,loainpanel will Thbereliableintuettoin the installedi riildesign power control             MIL-STD-167-1                      The test parameter values      mounted to the Seismic Category I Auxiliary panel              38-9193058-000,                     provided in IEC Standards,     Building wall and would not be subjected to EN 60068-2-27,                     l EC 60068-2-6 (vibration)     any significant shock or vibration during a Tests and             38-9218022-000,                     and lEC 60068-2-27             postulated beyond design basis event, or Page 6 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION T              T   -               - r                 r Parameter        Vendor  Design Reference        Additional    eto      nlyi        eut               'Licensee Evaluation Topic.         Summary                              Comments     Tet.rAnlyi.Rsut Document #
analyses were         EN 60068-2-6,                 (shock), tables are           during normal operation. The instrumentation performed for the     38-9218023-000,                recommendations and not       is located within the Seismic Category I horn cover and         MIL-STD-202                  mandatory testing levels. Auxiliary Building and is protected from adhesive to      51-9221032-000                 The test parameter values     external wind borne missile threats. The demonstrate       66-9225632-000                 were selected to be           instrumentation installed design location is not adequate shock        32-9221237-003                 consistent with previous      susceptible to vibration from surrounding withstand.      66-9225469-000                 shock and vibration testing   rotating equipment. The radar sensor and Additional testing    38-9228047-000                 performed on the VEGA         power control panel design location provides was performed                                       supplied equipment. The       spatial separation from surrounding SSCs, for the power                                     test parameter values         such that potential seismic interaction with control panel                                    specified envelope the        surrounding SSCs is also not a concem.
assembly.                                      expected levels for the equipment installed location, The post modification testing will demonstrate due to the fact that the      reliable operation of the instrumentation, which equipment is mounted to       confirms no damage occurred during shipping, seismic structures within the handling and installation. Similarly, the plant. This approach is       performance of monthly channel functional consistent with similar       comparisons will serve to confirm proper technology used in the        operation of the instrumentation, or provide a same application at other     means of early detection of potential installations.                instrument degradation.
The vibration testing         Refer to RAI #7 response posted to Catawba deviated from the IEC         Fukushima Sharepoint.
60068-2-6 recommended frequency range and           The test and analysis of the homn cover and displacement magnitude for    adhesive demonstrate adequate shock large power plant equipment   withstand capability.
(TABLE C.2). In-lieu of the 10-55 Hz and minimum displacement of 0.15 mm recommended in TABLE C.2, the power control panel vibration testing utilized a narrower frequency band (5-25 Hz) and a more limiting displacement magnitude (1.6 mm). These values were deemed to be acceptable and envelopingq Page 7 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND    CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION Prmtr    Vendor Design   Adtoa
   #Topic          Parameterar      Reference                                                 Liceseetvalutio Summary.       Document #*    Comments  -Test or Analysis Results,         iesevlain        *..
for equipment rigidly mounted to a Seismic Category I structure, based on engineering judgment.
The shock testing deviated from the IEC 60068-2-27 recommended peak acceleration and duration for land-based permanently installed equipment. In-lieu of the 15 g's peak acceleration and duration of 11 in-sec recommended in TABLE A.1, the power control panel vibration testing utilized an acceleration of log with a 6 in-sec duration. These values were deemed to be acceptable and enveloping for equipment rigidly mounted to a seismic Category I structure, based on engineering judgment.
Testing and analysis of horn cover and adhesive support the components can tolerate horizontal and vertical accelerations up to 100~g and SFP sloshing loads up to 3.37 psi.
15  Requirements  Not required by      N/A           N/A                   N/A                          N/A Traceability       order Page 8 ofll1

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND                         CNS    SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION Parameter    Vendor Design           Additional *
      #Topic,       ,         PSmaramete             oumnReference
                                                           *       ,
* omns'-      Test or Analysis Results                 ,, Licensee Evaluation:
Inspection of waveguncidelitestf         VEGA Test                                                              The vendor factory acceptance test pofefntioalityr tof    Procedure AREVA                                                        demonstrated reliable operation of the SFP pwrtaseto      Doc. 38-9219704-                        Test demonstrates that       level instrumentation under normal conditions 1 FAccetonce          battmeryresenso           000, Factory             N/A       specification requirements      and under various simulated test conditions AcpacTet      mascurement             Acceptance Test                               were met.                   (e.g. steam exposure). The testing accurcys ostandeorRVADc                                                                     demonstrated the instrumentation met design an ae n66-9227809-002                                                                    accuracy and repeatability specifications.
waveguide The vendor factory acceptance test demonstrated reliable operation of the SEP Normal and                                              Normal conditions accuracy    level instrumentation under normal conditions accident        AREVA Instruction        EA-12-051,     +1 inch, error due to all   and under various simulated test conditions 17     Channel             conditions SEP        manual Doc. 01-             1.7        effects including 212&deg;F     (e.g. steam exposure). The testing Accuracy                   level           9223080-003,           NEI 12-02,   saturated steam +_3 inches. demonstrated the instrumentation met design measurement            Section 11.6              3.7        Accuracy verified during    accuracy and repeatability specifications.
accuracy                                              factory acceptance testing.
Refer to RAI #11 response posted on Catawba Fukushima Sharepoint.
Based on vendor analysis the battery capacity Qualification Analysis                                               is deemed sufficient to support reliable PwrLftmof                       Doc. 51-9202556-         EA-1 2-051,                                instrument channel operation until off-site 18     Pwr005,                                         Section 2.9,         1.6,      Battery capacity at full load resources can be deployed by the mitigating Cosmtofbteybcuoat p                Instruction Manual       NEl 12-02,      is expected to readily     strategies in response to Order EA-12-049.
fllod01-9223080-003,                    3.6             exceed 7 days.
Section 11.7                                                     Refer to RAI #10 response posted to Catawba Fukushima Sharepoint.
Application-specific information on                                                                          The vendor technical manual has been 19  Mecnical            the installation,     AEADc             1        N/A                    N/A              reviewed, accepted and incorporated in the Mauloperation,                and'       238-0                                                          engineering change package.
maintenance of the SFPLI Page 9 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND                      CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION Parameter    Vendor Design         Adionl         Ts       A1       R    t Topic            armamee             RefrenetiommnalTs                    s or Analysis Rsuls                    Licensee Evaluation
                  , ,,                        Document #          -____
EA-1 2-051, 1.8 NEI 12-02, 3.8 Based on negligible drift rate of VEGA electronics experienced                                 Refer to RAls #12 and #15 responses posted Periodic          EA-12-051, 1.8        over large                                to Catawba Fukushima Sharepoint.
indication        AREVA Doc. 01-        user base, 20  Calibration            checks,           9238-0,                periodic                NAIn-lieu                horn rotation and use of a portable clbain       Sections 7.0 and     calibration is             NAtarget,                periodic verification of proper radar checks,                911not               needed,                               channel functionality can be achieved by calibration              911Functional                                              varying SEP water level (minimum 2 points) verification                               and proper level indication.
can be achieved using cross channel checks and functional checks per vendor manual.
The instrumentation is required to function to provide SEP level indication for a beyond design basis event. Performance of a FMEA 21  Failure Modes                                  N/A                 N/A                  N/A              is not warranted for this type of an application.
and Effects               N/A                                                                             Reasonable assurance that both channels are Analysis (FMEA)                                                                                           not susceptible to a common mode failure is provided by satisfying the NE1 12-02 guidance.
Emissions and     Qualification Analysis                   VEGAPULS 62 ER has         The EMI/RFI susceptibility and emissions susceptibility     Doc. 51-9202556-                     been tested for emissions to teastingeperformedfrtes     waveguide asradarth 22     EMI Testing       testing for      005, Section 2.6 and         N/A      both MIL and lEC standards   trnmtepovdsaqueasrncth VEGAPULS 62       supporting references                  adfr           tl 4 '   EG   instrumentation will be compatible in the ERE-6 004standpards                                              design location. The testing was MIL-STD-461 E,                               sadrsconservatively                 performed with unshielded Page 10 ofil1

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND                              CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION r# .......Topic
       * -Toc   .... "  Parameter Summa             V......
endorDesinc    Additional Reference... Comments                 TsorAalysis Results                        ieseEauto 58-9214362-000,                                                        interconnecting wiring. The Catawba level 38-9219863-000,                                                        channel design included shielded signal 38-9218965-000,                                                        cabling, and grounding of the power control 38-9218966-000,                                                        panel.
38-9219862-000, 38-92 18967-000,                                                       Post-modification testing has demonstrated 38-9218968-000,                                                        acceptable operation in the installed location.
38-9218969-000, 38-9218970-000,                                                        During a postulated BDBEE, it is possible that 38-0218964-000                                                        intermittent UHF radio operation could occur in the vicinity of the radar transmitter.
Successful long-term SFP monitoring capability during a postulated BDBEE would not be inhibited by potential intermittent radio
_____________  __________________transmissiontransissinrinerfeence Page 11 of 11

-4 ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 NRC Order EA-12-049 FLEX FINAL INTEGRATED PLAN Catawba Nuclear Station, Units I & 2 February 2016 Page 1 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 Table of Contents Contents
: 1. Background .....................................................................................                  5
: 2. Order Implementation..........................................................................                    6 2.1    General Elements................................................................................ 6 2.2    Strategies......................................................................................... 8 2.3    Reactor Core Cooling Strategy ................................................................. 9 2.3.1   Phasel1.......................................................................................... 9 2.3.2   Phase 2........................................................................................ 10 2.3.3   Phase 3........................................................................................ 11 2.3.4   Availability of Systems, Structures, and Components ...................................... 11 2.3.5   FLEX Connections............................................................................. 13 2.3.6   Plant Instrumentation.......................................................................... 15 2.3.7   Thermal-Hydraulic Analysis................................................................... 16 2.3.8   Reactor Coolant Pump Seals................................................................. 17 2.3.9   Shutdown Reactivity Analysis ................................................................ 17 2.3.10  FLEX Pumps................................................................................... 17 2.3.11  Electrical Analysis ............................................................................. 19 2.4    Spent Fuel Pool Cooling/Inventory Strategy.................................................. 20 2.4.1   Phase 1I........................................................................................ 20 2.4.2   Phase 2 ........................................................................................                20 2.4.3   Phase 3 ........................................................................................ 20 2.4.4   Availability of Structures, Systems, and Components ...................................... 20 2.4.5   Ventilation ....................................             .................................................. 21 2.4.6   Plant Instrumentation.......................................................................... 21 2.4.7   Thermal-Hydraulic Analysis................................................................... 22 2.4.8   FLEX Pump and Water Supplies............................................................. 22 2.4.9   Electrical Analysis ............................................................................. 22 2.5    Containment Function Strategy................................................................ 23 2.5.1   Phase 1........................................................................................ 23 2.5.2   Phase 2 ........................................................................................ 23 Page 2 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.5.3    Phase 3 ........................................................................................ 23 2.5.4    Availability of Structures, Systems, Components ........................................... 24 2.5.5    Plant Instrumentation.......................................................................... 24 2.5.6    Thermal-Hydraulic Analysis .................................................................. 24 2.5.7    Electrical Analysis ............................................................................. 25 2.6     Characterization of External Hazards......................................................... 25 2.6.1    Seismic Events ................................................................................ 25 2.6.2    External Flooding .............................................................................. 26 2.6.3    Storms such as Hurricanes, High Winds, and Tornadoes.................................. 26 2.6.4    Extreme snow, ice and cold .................................................................. 26 2.6.5    Extreme heat .................................................................................. 26 2.7     Planned Protection of FLEX Equipment ...................................................... 27 2.7.1    FLEX Storage Facility......................................................................... 27 2.8     Planned Deployment of FLEX Equipment.................................................... 27 2.8.1    Haul Paths and Accessibility.................................................................. 27 2.8.2    Deployment of Strategies..................................................................... 29 2.8.3    Fueling of Equipment.......................................................................... 30 2.9     Sequence of Events and Staffing ............................................................. 31 2.9.1    Sequence of Events........................................................................... 31 2.9.2    Staffing ......................................................................................... 34 2.10    Offsite Resources .............................................................................. 35 2.10.1   National SAFER Response Center (NSRC)................................................. 35 2.10.2   Equipment ..................................................................................... 36 2.11    Habitability and Operations.................................................................... 36 2.11.1   Equipment Cooling ............................................................................ 36 2.11.2   Freeze Protection ............................................................................. 36 2.11 .3  Hydrogen Ventilation.......................................................................... 37 2.11.4   Personnel Habitability......................................................................... 37 2.12    Water Sources.................................................................................. 37 2.12.1   SG Feedwater ................................................................................. 37 2.12.2   RCS Makeup................................................................................... 38 2.12.3   SFP Makeup ................................................................................... 38 2.13    Shutdown and Refueling Analysis............................................................. 38 Page 3 of 44

ATTACH MENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.14   Procedures and Training....................................................................... 39 2.14.1  Procedural Guidance.......................................................................... 39 2.14.2  Training ........................................................................................      39
: 3. Acronyms.........................................................................         i............. 41
: 4. References..................................................................................... 43 Page 4 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2

===1. Background===
In 2011, an earthquake-induced tsunami caused Beyond-Design-Basis (BDB) flooding at the Fukushima Dai-ichi Nuclear Power Station in Japan. The flooding caused the emergency power supplies and electrical distribution systems to be inoperable, resulting in an extended loss of AC power (ELAP) in five of the six units on the site. The ELAP led to (1) the loss of core cooling, (2) loss of spent fuel pool cooling capabilities, and (3) a significant challenge to maintaining containment integrity. All DC power was lost early in the event on Units 1 & 2 and after some period of time at the other units. Core damage occurred in three of the units along with a loss of containment integrity resulting in a release of radioactive material to the surrounding environment.
The U.S. Nuclear Regulatory Commission (NRC) assembled a Near-Term Task Force (NTTF) to advise the Commission on actions the U.S. nuclear industry should take to preclude core damage and a release of radioactive material after a natural disaster such as that seen at Fukushima. The NTTF report (Reference 1) contained many recommendations to fulfill this charter, including assessing extreme external event hazards and strengthening station capabilities for responding to beyond design basis external events (BDBEEs).
Based on NTTF Recommendation 4.2, the NRC issued Order EA-12-049 (Reference 2) on March 12, 2012 to implement mitigation strategies for BDBEEs. The order provided the following requirements for strategies to mitigate BDBEEs:
* Licensees shall develop, implement, and maintain guidance and strategies to maintain or restore core cooling, containment, and Spent Fuel Pool (SFP) cooling capabilities following a BDBEE.
          *These strategies must be capable of mitigating a simultaneous loss of all AC power and loss of normal access to the ultimate heat sink (UHS) and have adequate capacity to address challenges to core cooling, containment and SFP cooling capabilities at all units on a site subject to the Order.
          *Licensees must provide reasonable protection for the associated equipment from external events. Such protection must demonstrate that there is adequate capacity to address challenges to core cooling, containment, and SEP cooling capabilities at all units on a site subject to the Order.
* Licensees must be capable of implementing the strategies in all modes.
* Full compliance shall include procedures, guidance, training, and acquisition, staging or installing of equipment needed for the strategies.
The order specifies a three-phase approach for strategies to mitigate BDBEEs:
* Phase 1 - Initially cope relying on installed equipment and on-site resources.
* Phase 2 - Transition from installed plant equipment to on-site FLEX equipment.
* Phase 3 - Obtain additional capability and redundancy from off-site equipment and resources until power, water, and coolant injection systems are restored or commissioned.
NRC Order EA-1 2-049 (Reference 2) required licensees of operating reactors to submit an overall integrated plan, including a description of how compliance with these requirements would be achieved. The Order also required licensees to complete implementation of the requirements no later than two refueling cycles after submittal of the overall integrated plan or December 31, 2016, whichever comes first.
Page 5 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 The Nuclear Energy Institute (NEI) developed NEI 12-06 (Reference 3), which provides guidelines for nuclear stations to assess extreme external event hazards and implement the mitigation strategies specified in NRC Order EA-12-049. The NRC issued Interim Staff Guidance JLD-ISG-2012-01 (Reference 4), dated August 29, 2012, which endorsed NEI 12-06 with clarifications on determining baseline coping capability and equipment quality.
Duke Energy (Duke) declared that Catawba Nuclear Station (CNS) Unit 2 was in compliance with Order EA-1 2-049 on March 31, 2015 following the 2EOC20 refueling outage, which is within two refueling cycles of the submittal of the OIP dated February 28, 2013 (Reference 5). Duke declared that CNS Unit 1 was in compliance with Order EA-12-049 on December 16, 2015 following the I E0C22 refueling outage, also within two refueling cycles of the QIP submittal (Reference 6).
NRC Order EA-12-051 (Reference 7) required licensees to install reliable SEP instrumentation with specific design features for monitoring SEP water level. This order was prompted by NTTF Recommendation 7.1 (Reference 1).
NEI 12-02 (Reference 8) provided guidance for compliance with Order EA-12-051. The NRC determined that, with the exceptions and clarifications provided in JLD-ISG-2012-03 (Reference 9),
conformance with the guidance in NEI 12-02 is an acceptable method for satisfying the requirements in Order EA-12-051.
Duke declared that CNS Unit 2 was in compliance with Order EA-12-051 on March 31, 2015 following the 2E0C20 refueling outage, which is within two refueling cycles of the submittal of the OIP dated February 28, 2013 (Reference 5). Duke declared that CNS Unit 1 was in compliance with Order EA-12-051 on December 16, 2015 following the 1 EOC22 refueling outage, also within two refueling cycles of the OIP submittal (Reference 6).
: 2. Order Implementation 2.1      General Elements The assumptions used for the evaluations of an ELAP/Loss of Ultimate Heat Sink (LUHS) event and the development of diverse and flexible coping strategies (FLEX strategies) are stated below.
Initial conditions and boundary conditions consistent with NEI 12-06 were established to support development of FLEX strategies, as follows:
* The reactor is initially operating at power, unless there are procedural requirements to shut down due to the impending event. The reactor was operating at 100% power for the past 100 days.
                 *The reactor is successfully shut down when required (i.e., all rods inserted, no Anticipated Transient Without Scram (ATWS)). Steam release to maintain decay heat removal upon shutdown functions normally, and reactor coolant system (RCS) overpressure protection valves respond normally, if required by plant conditions, and reseat.
* On-site staff is at site administrative minimum shift staffing levels.
* All personnel on-site are available to support site response.
* The reactor and supporting plant equipment are either operating within normal ranges for pressure, temperature and water level, or available to operate, at the time of the event consistent with the design and licensing basis.
Page 6 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 No specific initiating event is used. The initial condition is assumed to be a loss of off-site power (LOOP) with installed sources of emergency on-site AC power and station blackout (S80) alternate AC power sources unavailable with no prospect for recovery.
All AC power supplies were considered to fail concurrently at the beginning of the event.
       *Cooling and makeup water inventories contained in systems or structures with designs that are robust with respect to seismic events, floods, and high winds and associated missiles are available.
      *Normal access to the ultimate heat sink (UHS) is lost, but the water inventory in the UHS remains available and robust piping connecting the UHS to plant systems remains intact.
The motive force for UHS flow, i.e., pumps, is assumed to be lost with no prospect for recovery.
       *Permanent plant equipment that is contained in structures with designs that are robust with respect to seismic events, floods, and high winds and associated missiles, are available. SSCs were considered seismically robust if seismic requirements were imposed by licensing requirements. The portion of the fire protection system that is robust with respect to seismic events, floods, and high winds and associated missiles is available as a water source. Installed electrical distribution systems, including inverters and battery chargers, remain available since they are protected.
       *Fuel for FLEX equipment stored in structures with designs that are robust with respect to seismic events, floods and high winds and associated missiles, remains available.
      *No additional accidents, events, or failures are assumed to occur immediately prior to or during the event, including security events.
      *Reactor coolant inventory loss consists of unidentified leakage at the upper limit of Technical Specifications, reactor coolant letdown flow (until isolated), and reactor coolant pump seal leak-off at normal maximum rate.
      *For the SFP, all boundaries (e.g., liner, gates) and the SFP cooling system are assumed to be intact. The SEP heat load is assumed to be the maximum design basis heat load.
In addition, inventory loss from sloshing during a seismic event does not preclude access to the pool area.
Additional key assumptions associated with design and implementation of FLEX strategies are as follows:
      *Additional deployment resources are assumed to begin arriving at 6 hours and the site Emergency Response Organization (ERO) will be fully staffed at 24_ hours after the event.
      *The plant Technical Specifications contain the limiting conditions for normal unit operations to ensure that design safety features are available to respond to a design basis accident and direct the required actions to be taken when the limiting conditions are not met. The result of the BDBEE may place the plant in a condition where it cannot comply with certain Technical Specifications and/or with its Security Plan, and, as such, may warrant invocation of 10 CER 50.54(x) and/or 10 CFR 73.55(p). (Reference 10)
      *All installed AC power supplies (emergency on-site and SBO Alternate AC power sources as defined by 10CFR50.2) were considered not available and not immediately recoverable.
Page 7 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 &.2
           *Where non-safety, non-seismically designed, permanently installed equipment is used for FLEX strategies, SSCs were considered seismically robust if:
o   Seismic Qualification Utility Group (SQUG) methods are applied per the existing plant licensing basis.
o   Testing, analysis or experience-based methods are applied for the equipment class at design basis seismic levels.
o    Methodologies in EPRI 1019199, Experience Based Seismic Verification Guidelines for Piping and Tubing Systems, can be successfully applied relative to the Safe Shutdown Earthquake (SSE).
o   Other industry-recognized codes such as AWWA DI00 are applied to demonstrate functionality at SSE level ground motion.
o   High Confidence of a Low Probability of Failure (HCLPF) capacities are determined (e.g., EPRI NP-6041 Rev 1) conservative compared to the SSE.
* Personnel access to and qualification of equipment that forms a part of the FLEX strategy assumed no core damage.
* Per NEI 12-06 Section 3.2.1.8, maximum environmental room temperatures for habitability or equipment operation may be based on NUMARO 87-00 guidance if other design basis information or industry guidance is not available. Extreme high temperatures are not expected to impact the utilization of offsite resources or the ability of personnel to implement the required FLEX strategies.
           *Access through security fences, doors, and other barriers will be unimpeded and not require additional resources.
2.2 Stratecqies The objective of the FLEX strategies is to establish indefinite coping capability in order to:
* Prevent damage to the fuel in the reactors
* Maintain the containment function
* Maintain cooling and prevent damage to fuel in the SFP This indefinite coping capability will address an ELAP   - loss of off-site power, emergency diesel generators (EDGs) and any alternate AC source, but not the loss of AC power to buses fed by station batteries through inverters  -  with a simultaneous LUHS. This condition could arise following external events that are within the existing design basis with additional failures and conditions that could arise from a BDBEE.
The plant indefinite coping capability is attained through the implementation of pre-determined strategies (FLEX strategies) that are focused on maintaining or restoring key plant safety functions. The FLEX strategies are not tied to any specific damage state or mechanistic assessment of external events. Rather, the strategies are developed to maintain the key plant safety functions based on the evaluation of plant response to the coincident ELAP/LUHS event.
A safety function-based approach provides consistency with, and allows coordination with, existing plant emergency operating procedures (EOPs). FLEX strategies are implemented in support of EOPs using FLEX Support Guidelines (FSGs).
Page 8 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 The strategies for coping with the plant conditions that result from an ELAP/LUHS event involve a three-phase approach:
* Phase 1 - Initially cope by relying on installed plant equipment and on-site resources.
* Phase 2   - Transition from installed plant equipment to on-site FLEX equipment.
* Phase 3   - Obtain additional capability and redundancy from off-site equipment and resources until power, water, and coolant injection systems are restored.
The transitions to Phase 2 and Phase 3 will occur at different times for different portions of the FLEX strategies.
The strategies described below are capable of mitigating an ELAP/LUHS resulting from a BDBEE by providing adequate capability to maintain or restore core cooling, containment, and SEP cooling capabilities at CNS. Though specific strategies have been developed, due to the inability to anticipate all possible scenarios, the strategies are also diverse and flexible to encompass a wide range of possible conditions. These pre-planned strategies developed to protect public health and safety are integrated into EOPs in accordance with established change processes, and their impact to the design basis capabilities of the unit are evaluated under 10 CFR 50.59.
2.3    Reactor Core Coolingq Strategqy 2.3.1        Phase I Following reactor and turbine trips of each operating unit, the Turbine Driven Auxiliary Feedwater Pump (TDAFWP) supplies water to the Steam Generators (SGs) secondary side to cool down the Reactor Coolant System (RCS). CNS will limit cool down to a rate near 100&deg;F/hr to minimize RCS system inventory loss, while cooling the Reactor Coolant Pump (RCP) seals in a controlled manner. The SGs will be depressurized to 240 psig and RCS temperature will be maintained above 280&deg;F. Maintaining SG pressure at this level prevents nitrogen injection from the Cold Leg Accumulators (CLAs) into the RCS The TDAFWP will draw suction from the Upper Surge Tank (UST), the Auxiliary Feedwater Condensate Storage Tank (CACST), or the hotwell, which are condensate grade water sources. However, these sources are susceptible to damage from wind and seismic events and may not be available. The TDAFWP can also draw suction from buried piping in the Condenser Circulating Water (RC) system, which has been evaluated to be seismically robust and can supply sufficient water to support decay heat removal for at least 48 hours.
The valve to enable TDAFW pump suction from embedded RC piping opens on a loss of control power or instrument air, and will provide flow when suction pressure from the UST or the hotwell drop5s below the RC static pr'essure. CNS will control SG level by throttling CA system flow control valves from the Control Room or starting and stopping the TDAFWP.
To remove heat, CNS will discharge steam through the SG Power Operated Relief Valves (PORVs), which can be operated from the control room using vital battery power and safety grade nitrogen backup for the valve actuators. The SG PORVs can also be manually operated.
The vital station batteries provide DC power for essential instrumentation. Vital battery load shedding will be initiated 2.5 hours into the event. The CNS load shedding strategy will maintain power supply from the A and 0 vital batteries for approximately 10 hours, and the B and C vital batteries for approximately 24 hours.
Page 9 of44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 No action is necessary for managing RCS inventory or reactivity during Phase 1. CNS-specific analysis concluded that core uncovery will not occur until at least 72 hours into the event. Additionally, CNS determined that boration of the RCS to control reactivity is not needed until FLEX equipment can be deployed during Phase 2.
2.3.2      Phase 2 The Phase 2 core cooling strategy continues to use the SGs as the heat sink. CNS has multiple strategies for providing feedwater to the SGs using FLEX equipment:
            *If the TDAFWP is still available after the RC piping static inventory has been depleted, CNS can use a portable low pressure pump to establish a new suction source. The low pressure pump will draw suction from the Standby Nuclear Service Water Pond (SNSWP) to a connection in the service water system (RN) piping. CNS will align valves to direct flow to the auxiliary feedwater (CA) system and the TDAFWP.
            *When the TDAFWP is no longer available (e.g., steam pressure becomes too low), a portable, low pressure pump will provide feedwater by taking suction from the SNSWP and discharging into portable, medium pressure pumps located near the Exterior Doghouses of each Unit. The medium pressure pumps will discharge into connections in the Doghouses to establish flow to the SGs. CNS can establish a flow path using only hoses; fire protection system piping can also be used to reduce hose deployment.
             *SGs may be fed directly from the FLEX Raw Water Distribution system (i.e., without the medium pressure pump). This approach requires depressurizing the SGs below the discharge pressure of the FLEX Raw WaterDistribution System (approximately 150 psig). The option to use only low pressure pumps is less preferred because it provides less capacity and it requires feeding and steaming SGs that are close to dry. Additionally, this strategy is not preferred prior to CLA isolation, because of the potential for injection of nitrogen into the RCS, which may occur at 140 psig. CNS can establish a flow path using only hoses; fire protection system piping can also be used to reduce hose deployment.
CNS will deploy 600V diesel generators (DGs) to provide power for the FLEX strategies within 9 hours of the event.
            *A back-feed receptacle connection on selected motor control centers (MCCs) allows re-powering key installed equipment, including battery chargers. The A vital battery charger will be re-powered within 11.5 hours and the D vital battery charger will be re-powered within 12.5 hours. Battery chargers will be aligned to the vital batteries prior to battery voltage dropping below acceptable values. The spare battery charger will provide redundancy as a backup for either safety train on the applicable unit.
The B and C vital battery chargers will be repowered within 24 hours. This strategy allows all channels of vital instrumentation to be maintained.
            *The 600V FLEX DGs will also be connected to the plant MCCs for re-powering via the FLEX 'Backbone,' which consists of permanently installed cables, portable panelboards, and transformers. If normal power supplies are not available, they will be disconnected to allow connection of alternate supply cables. The alternate cable(s) will be plugged directly into the MCCs via FLEX receptacles for the component. Cables will be routed out of the Auxiliary Building for connection to the FLEX DGs.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 &.2 CNS will provide Phase 2 RCS makeup using a portable high pressure pump. Initially, CNS will use the reactor head vent valves to establish venting/letdown of the RCS. These valves will be re-powered using the MCC backfeed strategy and FLEX DGs. The high-pressure pump will draw suction from the Refueling Water Storage Tank (FWST) via a connection on the piping for makeup to the SEP. The high pressure pump will discharge into safety injection pump discharge piping. Sufficient borated water will be added to maintain the core sub-critical, in a xenon-free condition, at 350&deg;F.
The CLA isolation valves will be closed to prevent nitrogen from being injected into the RCS prior to final cooldown and depressurization. The CLA isolation valves will be re-powered using the FLEX DGs. After the CLAs are isolated or vented, CNS will depressurize SGs to 160 psig at a rate that will limit cooldown to less than 1 00&deg;F per hour.
A portable FLEX sump pump is placed in each CA TDAFWP pit to pump out normal drains input from the TDAFWP to the room sump before flooding impacts operation of the TDAFWP. The flooding impact could occur 7 hours into the event. Additionally, Groundwater Drainage System (WZ) sump pumps must be in operation by 10.6 hours into the event to prevent installed sump pump motors from being flooded. CNS can also deploy FLEX sump pumps in various locations (in addition to the TDAFWP pits) to manage groundwater intrusion.
2.3.3       Phase 3 After CNS has decreased RCS temperature to less than 350&deg;F and RCS pressure to less than 385 psig, CNS will transition to cooling by Residual Heat Removal (RHR). CNS will repower the Component Cooling (KC) system pumps to provide cooling water to the RHR pumps.
The National SAFER Response Center (NSRC) will deliver equipment to CNS to establish and align RHR and KC systems.
oTwo portable pumps will take suction from the SNSWP and discharge by hose to the supply piping of the RN system (one portable pump per RN train). The pressurized RN system will be manually aligned to supply flow to the KG system heat exchangers.
o    Large portable DGs from the NSRC will power KC and RHR system pumps.
The RHR pumps can provide borated makeup water to the core by taking normal suction from the FWST. If needed, mobile boration units from the NSRC can be used to refill the FWST through existing vents.
The NSRC will deliver a water treatment skid that can provide a cleaner water source than the SNSWP.
Additional diesel fuel for portable equipment will be brought in from off-site resources when required.
2.3.4       Availability of Systems. Structures, and Components The FLEX strategy for core cooling relies on various installed systems, structures, and components (SSCs). These SSCs are protected in regard to the applicable extreme external hazards as discussed below.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.3.4.1  Structures The FLEX strategy relies on selected site structures to provide protection from the applicable extreme external hazards. Specifically, the FLEX strategy relies on Containment, the Auxiliary Building, the Nuclear Service Water System Pumphouse, the SG Doghouses, and the Fuel Handling Building, which are all Seismic Category I structures.
2.3.4.2  Pipingq and Fittinqs The FLEX strategy relies on installed piping and fittings in various plant systems to deliver water for core cooling. Such piping is either designed for safety-related service or was analyzed as part of development of the FLEX strategies to confirm availability following any of the applicable hazards (e.g., the RC piping).
Primarily, CNS relies on piping and components from the Reactor Coolant System (RCS), Auxiliary Feedwater (CA) system, Safety Injection (NI) system, Residual Heat Removal (RHR) system, the Nuclear Service Water (RN) system, the Component Cooling (KC) system, the Condenser Circulating Water (RC) system, the Main Steam to Auxiliary Equipment (SA) system, the Main Steam Vent to Atmosphere (SV) system, the Steam Generator Wet Layup Recirculation (BW) system, the Spent Fuel Pool Cooling (KF) system, and the Refueling Water (FW) system.
2.3.4.3 Turbine Driven AFW Pump (TDAFWP)
The CNS FLEX strategy relies on the TDAFWP to provide feedwater for the SGs during Phase 1. The TDAFWP and its flow control valVes (FCVs) are safety-related, seismically-qualified components that are located in the Auxiliary Building, which is a Seismic Category I structure. These components are therefore protected from the applicable hazards.
2.3.4.4 Steam Generator Power Operated Relief Valves (PORVs)
The CNS FLEX strategy relies on the SG PORVs to remove heat during SG cooling, because cooling from the main condenser is not available. The SG PORVs are safety-related, seismically-qualified components located inside the SG Doghouses, which are Seismic Category I structures. These components are therefore protected from the applicable hazards.
2.3.4.5 Vital Station Batteries The CNS FLEX strategy relies on vital station batteries to power vital instrumentation.
The vital station batteries and associated DC distribution systems are located within the Auxiliary Building, which is a Seismic Category I structure. The vital batteries are therefore protected from the applicable hazards.
2.3.4.6 Electrical Distribution System CNS uses selected plant electrical distribution equipment to repower installed components credited for the FLEX strategy. Electrical distribution components used for the FLEX strategy are located within Seismic Category I structures and will therefore be available following the applicable hazards.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.3.4.7      Groundwater Drainagqe (WZ') Pumps The CNS FLEX strategy relies on the WZ pumps to manage ground water in-leakage.
These pumps will be re-powered by the 600V FLEX DGs. CNS has six WZ pumps (i.e.,
A1I,A2, B1,B2, and Cl,C2 pumps), which can all be powered simultaneously. The C groundwater sump and associated components are not seismically qualified; however, if the C WZ sump is intact, it is the preferred source. If the C WZ sump pump cannot be used, the A and B WZ sump pumps have adequate capacity to manage groundwater in-leakage. The A and B WZ sump pumps are protected from all hazards and will be available following a BDBEE.
2.3.4.8      Condenser Circulatingi Water (RC) Pipingq Although the preferred water source for SG feedwater is a clean source (e.g., the UST or the hotwell), the CNS FLEX strategy credits the RC piping for cooling water because it is protected from all applicable hazards. The Unit I RC system has a total usable volume of approximately 982,000 gallons and the Unit 2 RC system has a total usable volume of approximately 1,280,000 gallons. These water inventories are sufficient to provide decay heat removal capability for at least 48 hours.
2.3.4.9      Standby Nuclear Service Water Pond ('SNSWP')
After depletion of the inventory in the RC piping, CNS can use the SNSWP as a suction source for SG feedwater via the FLEX Raw Water Distribution System. The SNSWP is nuclear safety-related, seismically-protected, and contains sufficient inventory to support the FLEX strategy for an essentially indefinite duration (i.e., well into Phase 3).
2.3.4.10     Refueling Water Storaqe Tank (FWST)
The FWST is the credited source of makeup for the RCS. The minimum inventory of the intact FWST is 377,537 gallons and contains boron in accordance with CNS technical specifications. The FWST is seismically-qualified and the bottom portion is protected by a missile wall. If the top portion of the FWST is not damaged by the BDBEE, the initial inventory will be sufficient for RCS makeup for the duration of the event. If the top portion of the FWST is damaged, makeup to the FWST may be required within 52 hours.
            *If necessary, makeup to the FWST can be accomplished using the NSRC-supplied mobile boration skid or other borated water sources (e.g., trucking from McGuire Nuclear Station). Less preferred options for RCS makeup include recovery of borated water from the FWST annulus (if it was damaged) or preparing solution in a FLEX Portable Mixing Tank using borated water from the Boric Acid Tank (BAT) or barrels of dry boric acid.
2.3.5          FLEX Connections FLEX connections for water and power are installed on various plant systems to facilitate use of portable FLEX equipment. Primary and alternate connections are available for each system used for FLEX. These connection points are protected from the applicable extreme external hazards as discussed below.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.3.5.1  FLEX SG Feedwater Connections CNS can provide SG feedwater using portable FLEX equipment and hoses to connect to installed systems. Connection points used to establish a flow path include the following:
               *The primary SG feedwater connections are located on CA system piping in the SG Doghouses downstream of the containment isolation valves. A set of four connection points provides one connection for each of the four SGs. At each of the four connection points, CNS will replace a blind flange with a hose adapter to establish a flow path. These connections are located inside a Category I structure (i.e., the SG Doghouses) and are protected from the applicable hazards.
               *The alternate SG feedwater connections are located on BW system piping, and are located in the SG doghouses. A set of four connection points provides one connection for each of the four SGs. At each of the alternate connection points, CNS will replace the blind flange with a hose adapter to establish a flow path.
These alternate connections are located inside a Category I structure (i.e., the SG Doghouses) and are protected from the applicable hazards.
2.3.5.2 FLEX RN Connections CNS can provide water through connections on the RN system supply headers to provide feedwater to the TDAFWP or to support long-term RHR during Phase 3.
Connection points used to establish a flow path include the following:
                *The primary RN connection is on RN Train B at a fill valve that is located in the RN Pumphouse. The RN Pumphouse is a Category I structure that provides protection to this connection point for all applicable hazards.
                *The alternate RN connection is on RN Train A at an access plug in the yard.
The plug is inside the protected area fence near the RN Pumphouse.
2.3.5.3 FWST Supply Connection for RCS Makeupi The suction supply for the portable high pressure pump will come from a connection on the FWST supply line for makeup to the SEP. This connection point is in the Auxiliary Building, which is a Seismic Category I structure. Therefore, the supply connection is protected from all applicable hazards.
The suction strategy will designate the FWST of one Unit as the primary common suction source for both Units and the other Unit's FWST as a spare. CNS will use a gated wye assembly to facilitate this approach.
2.3.5.4 FLEX RCS Connections CNS will provide borated water to the RCS using a high pressure makeup pump.
Connection points supporting this strategy are located on the NI pump discharge piping in the Auxiliary Building and can be aligned for cold leg or hot leg injection. Connections are available on both Train A and Train B, which provide primary and alternate connection points. These connections are located on safety-related piping and are in a Seismic Category I structure. Therefore, these connections are protected from the applicable hazards.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.3.5.5      FLEX Electrical Connections The CNS FLEX strategy relies on DGs to charge batteries, maintain vital instrumentation, and repower plant equipment. The following connections are available to support this strategy:
* Back-feed receptacle connections are available on selected MCCs, which can feed train A and B and spare battery chargers, thereby providing primary and alternate connections for battery charging.
                    *Connections to normal plant MCCs will enable re-powering via the FLEX Backbone, which consists of permanently installed cables, portable panelboards, and transformers. The alternate approach is to disconnect normal power supplies to enable connection of alternate supply cables, which can be connected at the portable DG.
To support power for vital instrumentation and other equipment, permanent cabling, receptacles, back-feed breakers and MCC's will be seismically qualified and located above anticipated flood level within Category I structures.
2.3.6          Plant Instrumentation The following instrumentation is relied upon to support the FLEX core cooling strategy:
* RCS Hot Leg Temperature (Thot)
* RCS Wide Range (WR) Pressure
* SG Narrow Range (NR) Level
* Core Exit Thermocouple Temperature
* Pressurizer Level
* Reactor Vessel Level Indicating System
* TDAFW Pump Flow
* SG Pressure
* DC Bus Voltage
* Neutron Flux
* FWST Level
* Containment Sump Wide Range Level
* SG and Pressurizer Cavity Temperature
* RCP Seal Leakoff Flow If the primary instrument is not available or cannot be powered via normal means, CNS will provide an alternate power source or use a portable instrument. CNS can dispatch operators to monitor parameters locally (e.g., CA flow, SG pressure via temporary gauges) or portable test equipment may be used to monitor parameters from inside the Process Control System 7300 cabinet (e.g., SG narrow range level, RCS pressure and temperature, Pressurizer level, FWST level). In addition, CNS has 120 VAC DGs that may be used to provide power for instrumentation ifthe power is not available by other means.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.3.7        Thermal-Hydraulic Analysis CNS performed Fathom analyses to demonstrate that the FLEX pumps have sufficient capability to supply water for the various demands associated with the FLEX strategies.
Specific conclusions from these analyses included the following:
              *A conservative bounding flow rate for SG feedwater demand at 6 hours into the event is 50 gpm per loop (total of 200 gpm per Unit). A conservative bounding flow for SG feedwater demand at 24 hours is 33 gpm per loop (total of 132 gpm per Unit).
              *One FLEX low pressure pump, taking suction from the SNSWP, can supply (by hose) a FLEX medium pressure pump for each Unit for direct feedwater supply to all four SGs (at 50 gpm each; total of 400 gpm for both Units) and four control room spot coolers (at 10 gpm each; total of 40 gpm). SG feedwater supply is sufficient using either the primary or alternate connections.
              *After 24 hours, one FLEX low pressure pump, taking suction from the SNSWP, can supply (by hose) a medium pressure pump for each Unit, which then directly feeds all four SGs (at 33 gpm each), boration equipment for each Unit (at 160 gpm total) and control room spot coolers (at 40 gpm total).
              *One FLEX low pressure pump, taking suction from the SNSWP and discharging by hose to the RN system, can supply feedwater for all four SGs at both Units (maximum of 50 gpm each; 400 gpm total) and the SFPs at a rate significantly exceeding the most conservative steaming rate (96.8 gpm).
              *During Phase 3, two NSRC low pressure pumps, taking suction from the SNSWP and discharging by hose to the RN system, can supply a KG Heat Exchanger for each unit at 3,457 gpm (via A Train) and 3,111 gpm (via B train).
* One FLEX high pressure pump, taking suction from a common FWST and discharging by hose connections into the Unit I and Unit 2 NI system piping, can supply the Unit 1 and Unit 2 RCS at least 40 gpm with back-pressure of 450 psia.
* The CNS FLEX strategies require approximately 489,000 gallons of water per Unit in the first 48 hours of the event.
Condensate grade water is required for SG makeup instead of raw water within 96 hours for Unit 2 based on sludge buildup and 275 hours for Unit 1 based on reaching the corrosion limit. (Unit 1 and Unit 2 have different types of SGs, and design differences result in the large difference in allowable service times with raw water.) CNS will deploy water treatment equipment as part of the Phase 3 FLEX strategy to provide a higher purity water source, if long-term SG feedwater is needed.
The minimum required SG pressure to operate the TDAFW pump is 125 psig. CNS procedures permit SG cool down resulting in a SG pressure of 160 psig following closure of the CLAs, which is well into Phase 2. Therefore, SG pressure will be sufficient to power the TDAFWP at least until the FLEX pumps are deployed.
CNS analysis concluded that a conservatively high inflow to the TDAFW pump sumps is 15 gpm. Accumulation of water at this rate would result in flooding the pump after 7 hours.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.3.8         Reactor Coolant Pump Seals CNS performed a modification to each of the RCP #1 seal leak-off lines to replace the flow orifices with a thicker component that would be suitable for higher differential pressure.
Analyses of this modified configuration show that leak-off flow rates stay within the maximum limits of current WCAP-17601-P, Revision 1 assumptions (i.e., 21 gpm / seal).
See Attachment 7 of the Final Compliance Submittal (Reference 6), "CNS Reactor Coolant Pump Seal Leakage Margin Assessment" for additional discussion on RCP seal leakage.
2.3.9         Shutdown Reactivity Analysis CNS performed a shutdown reactivity analysis that incorporated the guidance provided in the Westinghouse position paper entitled "Westinghouse Response to NRC Generic Request for Additional Information (RAI) on Boron Mixing in Support of the Pressurized Water Reactor Owners Group (PWROG)" (ADAMS Accession Number ML13235A135) with the clarifications specified in the NRC endorsement of this approach (Reference 11). The CNS analysis included a one-hour mixing delay, in accordance with those clarifications.
The CNS analysis concluded that the FLEX high pressure pump, delivering FWST water at 40 gpm, must be started by 13.85 hours into the ELAP event to provide and maintain the necessary 1% shutdown margin and prevent a potential re-criticality during cooldown. To meet this criterion, CNS plans to deploy the high pressure pump prior to 9 hours into the event. For boration after initial cool down of both units, ONS requires delivery of a total of
       *28,100 gallons of FWST water into the RCS, which is well within the nominal FWST inventory of 377,537 gallons, even if damaged above the protective wall.
For the latest NOTRUMP reference case, the PWROG-14027-P, Revision 3 report indicates that reflux cooling may begin in four-loop plants like CNS after 15.6 hours. Addition of borated RCS makeup must occur before reflux cooling to ensure adequate mixing. The CNS timeline for deploying the high pressure pump satisfies this requirement.
2.3.10        FLEX Pumps 2.3.10.1     FLEX Low Pressure Pump After the TDAFWP is secured, the CNS FLEX strategy relies on a portable, diesel-driven, low pressure, high volume pump to supply water from the SNSWP to the FLEX Raw Water Distribution System.
The FLEX low pressure pumps are portable, diesel-driven pumps that can supply a design flow of 3,000 gpm. The FLEX low pressure pump discharge pressure will be maintained at 250 to 275 psig. As discussed in Section 2.3.7, hydraulic analysis shows that the FLEX low pressure pumps have sufficient capacity to support the CNS FLEX strategies.
CNS has two portable FLEX low pressure pumps to satisfy the N+1 requirement.
2.3.10.2    FLEX Medium Pressure Pump After the TDAFWP is secured, the CNS FLEX strategy relies on a portable medium pressure pump to provide makeup water to the SGs. The FLEX Medium Pressure Makeup Pump can also be used to provide RCS makeup if the event occurs when the plant is in modes 5 or 6.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 The FLEX medium pressure pumps are portable, diesel-driven centrifugal pumps that can supply 300 gpm at a maximum pressure of 400 psig. As discussed in Section 2.3.7, hydraulic analysis shows that these pumps have sufficient capacity to support the CNS FLEX strategies.
CNS has three FLEX medium pressure pumps to satisfy the N+1 requirement.
The FLEX low pressure pump provides water to the FLEX medium pressure pump via hoses (and also fire protection piping, if desired).
2.3.10.3  FLEX Higqh Pressure Pump For an ELAP event initiating in Modes 1 - 4, the CNS FLEX strategy relies on a high pressure pump to provide RCS makeup and boration. The FLEX high pressure pumps are portable, diesel-driven, centrifugal pumps that can supply 40 gpm at 1700 psig, which is adequate to support the reactivity control and RCS system make-up requirements for the FLEX strategy.
CNS has three portable FLEX high pressure pumps to satisfy the N+1 requirement.
The credited water supply for the FLEX high pressure pump is the FWST.
2.3.10.4  CA TDAFW Pit Portable Sump Pumps A portable sump pump is placed in the CA TDAFW pit to pump out normal drains input from the TDAFW pump to the room sump before flooding impacts operation of the TDAFWP. Each sump pump is electrically powered and requires 120 V, which can be supplied either via the FLEX Electrical Distribution System, or a 120V DG. The sump pumps can deliver 15 gpm flow at over 30 feet of head, which is adequate to prevent flooding of the TDAFW pit sum ps.
CNS has three 120V sump pumps, which are sufficient to support the CNS FLEX strategies.
2.3.10.5 Auxiliary Building Portable Sump Pumps CNS can deploy portable 600V sump pumps to manage internal flooding from potential pipe breaks. Each 600V sump pump can be powered by one of the 600V DGs, although a single DG may not have sufficient capacity for operation of all three sump pumps simultaneously.
CNS has three 600V sump pumps, which are sufficient to support the CNS FLEX strategies.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 12& 2 2.3.11     Electrical Analysis 2.3.11.1  Diesel Generators CNS relies on DC systems for necessary electrical coping power during Phase I of the ELAP. To extend the coping capability of the vital station batteries, CNS will complete load shedding within 3.5 hours of the event to reduce battery discharge to only essential loads (e.g., vital instrumentation). This action will extend the functional capability of the vital station batteries to at least 11.5 hours for the limiting battery. The CNS FLEX strategy for repowering battery chargers with FLEX DGs will provide electric power prior to the end of battery life.
For longer term electrical power, CNS will deploy portable FLEX DGs (one for each Unit), and associated support equipment to establish the FLEX Electrical Distribution System. CNS has three FLEX 600 VAC DGs to satisfy the N+I requirement. CNS performed an analysis, which verified that the FLEX DGs and planned cable routing arrangement were adequate to support the required loads. The rating of each 600V DG exceeds its total running load during FLEX deployment.
2.3.11.2 Liahtina Post-Fire Safe Shutdown lighting is available in many areas where manual actions are necessary. The Post-Fire Safe Shutdown lights have self-contained batteries with an 8-hour life. Additional portable lighting will be provided for use in the yard.
Lighting units included in the FLEX strategy are as follows:
I. LED tripod mounting Quad Pod LED lights. Units are rated at 20,000 lumens.
: 2. LED string lights
: 3. Miscellaneous helmet lights and flashlights. Typical units and approximate quantities include 50 D cell LED flashlights, 50 helmet lights, and 25 stand spot lights.
LED-mounted tripods will be deployed in many areas including, the Control Room, CA Pump Room(s), Motor Generator (MG) Set Room(s), Interior and Exterior Doghouses, Electrical Penetration Rooms, Battery Room(s), Technical Support Center, and General floor areas in the Auxiliary Building. String lights will also be used in the Control Room and the Technical Support Center. The lighting plan includes a total of 27 LED-mounted tripods and 3 string lights.
Four hard hats with lights are located at the Main Control Room (MCR) exit doors S400 and $406.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.4    Spent Fuel Pool Coolingq/Inventory Strateoqy 2.4.1         Phase I No actions are required during ELAP Phase 1 for SEP make-up because the time to boil is sufficient to enable deployment of Phase 2 equipment.
                   *For the worst case heat load scenario, the SEP will begin to boil after an ELAP/LUHS event in 8.8 hours. This scenario assumes a maximum starting SFP temperature of 125&deg;F and a full core offload during an outage after 6 days.
                   *During normal operation (21 days after the beginning of a refueling outage and initial SEP temperature of 125&deg;F), SEP level can be maintained at least 10 feet above the top of the fuel (above which personnel access to the SFP operating deck is still viable) for 202.9 hours with no makeup. For normal operating conditions, boiling will begin in 37.0 hours.
CNS will monitor SEP water level using SFP level instrumentation.
2.4.2         Phase 2 To compensate for SEP boil-off, CNS will provide makeup water by pumping raw water from the SNSWP using the FLEX low pressure pump.
The primary FLEX strategy for SEP cooling is to pressurize the RN system and then open an installed RN valve and an installed KF valve to establish a flow path to the SEP using installed piping.
The alternate FLEX strategy for SEP cooling is to pressurize the RN system and use a hose to connect RN piping to a valve in the KF skimmer loop. CNS will attach an adapter on the KF valve to facilitate the hose connection.
CNS does not have a FLEX strategy for spray makeup to the SEP. The CNS SFPs are in Category I structures and are connected to safety-related systems, so they are not susceptible to being drained. In accordance with NEI 12-06, Table D-3, spray makeup capability is not required for sites that have SEPs that cannot be drained.
2.4.3         Phase 3 Long term SEP cooling will be accomplished by re-powering the installed KE and KC pumps using a portable generator obtained from the NSRC to provide cooling via normal means.
Cooling of KC will be from the RN system that is being supplied from the portable diesel driven pump located at the SNSWP. Additional diesel fuel for portable equipment will be brought in from off-site resources when required.
2.4.4        Availability of Structures. Systems. and Components 2.4.4.1     Structures The CNS SEP cooling strategy relies on the SEP Building, the Auxiliary Building, and the RN Pumphouse, which are addressed in Section 2.3.4.1. All relevant structures are robust to the applicable external hazards.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.4.4.2      Systems The CNS SEP cooling strategy relies on the KE system, the KC system, and the RN system, which are addressed in Section 2.3.4.2. All relevant systems are robust to the applicable external hazards.
2.4.4.3      SNSWP The CNS SFP cooling strategy provides SEP makeup water from the SNSWP, which is addressed in Section 2.3.4.9. The SNSWP is robust to the applicable external hazards and has sufficient inventory for long-term SEP cooling (among other demands).
2.4.4.4      FLEX Connections for SEP Makeup The CNS SEP cooling strategy includes primary and alternate connections for delivering water, which meets the requirements of NEI 12-06.
                *The primary strategy for SEP makeup relies on pressurizing the RN system using the FLEX low pressure pump. As discussed in Section 2.3.5.2, the B Train connection is located inside the RN Pumphouse, which is a Seismic Category I structure that is protected from all applicable hazards. The A Train connection provides a backup and is in the yard. RN and KE valves to be manually operated are located inside the Auxiliary Building, which is a Seismic Category I structure.
                *The alternate strategy also relies on pressurizing the RN system, but uses a jumper hose to connect to the KE system skimmer loop. The skimmer loop flows through a manifold line around the SEP with a series* of discharge points controlled by manual valves.
2.4.5          Ventilation CNS will vent the SEP to outside to minimize the impact of condensed steam on auxiliary building habitability. Fuel building air temperature and humidity will begin to increase shortly after event initiation, and the building may become uninhabitable within a few hours. To limit
        *pressure in the SEP building, selected doors will be opened within 6 hours. Additionally, CNS will take actions to prevent condensed steam from entering other parts of the Auxiliary Building outside of the selected vent path, such as securing doors and sealing door cracks.
2.4.6          Plant Instrumentation The key parameter for the SEP cooling/inventory function is SEP level. Instrumentation is capable of identifying the following SEP water level conditions, in accordance with NRC Order EA-12-051:
* Level that is adequate to support operation of the normal fuel pool cooling system.
* Level that is adequate to provide substantial radiation shielding for a person standing on the SEP operating deck.
* Level where fuel remains covered and action to implement makeup water addition should no longer be deferred.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 The SEP level instrumentation equipment was designed to perform its design function at temperature, humidity, and radiation levels consistent with SEP water at saturation conditions for an extended period following a BDBEE. Sensing components and cables located at the SFP were qualified to withstand peak and total integrated dose radiation levels for their installed locations based on post-event SEP water level equal to Level 3 for an extended period of time. (Level 3 is the water level where fuel remains covered but actions to implement make-up water addition should no longer be deferred based on the accuracy of the instrumentation and the highest point of any fuel rack.)
2.4.7       Thermal-Hydraulic Analysis CNS performed thermal-hydraulic analysis to address the SEP cooling/inventory function under the most limiting conditions and configuration. Key conclusions from the analysis include the following:
* The worst case heat load corresponding to a full core off load 6 days after a shutdown will produce evaporation rates of 96.8 gpm per SEP.
* Eor the worst case heat load scenario, the SEP will begin to boil after an ELAP/LUHS event in 8.8 hours. This scenario assumes a maximum starting SEP temperature of 125&deg;F and a full core offload during an outage after 6 days.
             *During normal operation (21 days after the beginning of a refueling outage and initial SEP temperature of 125&deg;F), SEP level can be maintained at least 10 feet above the top of the fuel (above which personnel access to the SEP operating deck is still viable) for 202.9 hours with no makeup. Eor normal operating conditions, boiling will begin in 37.0 hours.
* A ELEX low pressure pump, taking suction from the SNSWP and discharging by hose to connections on the RN system, can supply each of the SEPs at a rate significantly exceeding the most conservative steaming rate in addition to the all SGs at 50 gpm each.
In addition to the thermal-hydraulic analyses, the CNS SEP has been analyzed to remain subcritical with zero boron credited, i.e., a complete dilution scenario as long as the fuel assemblies remain covered and the parameters in CNS Technical Specification 4.3.1 remain in effect.
2.4.8       ELEX Pump and Water Supplies The CNS ELEX strategy relies on FLEX low pressure pumps to supply raw water from the SNSWP via RN piping. See Section 2.3.10.1 for description of the pumps.
As discussed in Section 2.4.7, the ELEX low pressure pumps have more than enough capacity to maintain SEP levels.
2.4.9       Electrical Analysis SEP level will be monitored by instrumentation installed to satisfy Order EA-12-051.
The instrumentation is normally powered from 120 V AC and upon an ELAP, the instrument will continue to operate upon switchover to a battery backup. The manufacturer-provided battery backup will last a minimum of 7 days from loss of off-site power, and those batteries can be replaced if necessary. SEP level instrumentation includes a backup channel that is powered from the vital inverters. FLEX strategies to restore power to vital battery chargers will ensure continued availability of the backup channel.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.5    Containment Function Strateoqy 2.5.1        Phase 1 The CNS Reactor Building includes a metal containment vessel and annulus region between the metal containment and a reinforced concrete enclosure. The containment vessel design pressure is 15 psig.
Following an ELAP/LUHS, the CNS Containment is initially cooled by an ice condenser.
Steam escaping the primary and/or secondary systems is cooled as it rises from lower containment through the ice condenser and into upper containment.
CNS performed a containment analysis, which demonstrated that containment pressure is expected to remain below the design pressure during Phase 1.
2.5.2       Phase 2 CNS will use the FLEX Electrical Distribution System to enable Phase 2 actions for the FLEX strategy to maintain Containment integrity.
                 *CNS will start a Hydrogen Skimmer Fan within 24 hours of event occurrence to limit the temperature increase in the SG and Pressurizer compartments. The SG and Pressurizer compartment temperature limits will be challenged before overall containment temperature limits. Abnormally high temperatures in these compartments could affect associated level indications due to reference leg flashing.
                 *Per NEI 12-06, Revision 0, plants with ice condenser containment designs such as*
CNS are required to repower hydrogen igniters to prevent buildup of hydrogen in case the ELAP event degrades to core damage. One train of hydrogen igniters will be re-powered and restored to service in Phase 2 using the back feed portable power strategy. The opposite train of hydrogen igniters can be repowered as an alternate.
* Manual containment isolation will also be Completed by the end of Phase 2.
2.5.3       Phase 3 Following deployment of an NSRC 480V generator and energizing of the 600V MO~s, CNS will start two Lower Containment Ventilation Units (LCVUs) within 48 hours to limit the temperature increase in the SG and Pressurizer compartments. Additionally, one Containment Air Return Fan (CARF) will be started within 52 hours of the event to establish an air flow path through the ice condenser, reduce containment pressure, and limit further heatup of the SG and pressurizer compartments.
With the CARE operating, SG and Pressurizer compartment temperature will be maintained below 200&deg;F and containment pressure will be less than 6 psig.
CNS wijl complete transition to RHR system cooling and cooldown to Mode 5 within 6 days of ELAP initiation to prevent challenging containment temperature and pressure limits following ice bed depletion.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.5.4        Availability of Structures. Systems. Components 2.5.4.1     Structures The FLEX strategy relies on site structures to provide protection for components, fluid and electrical connections, and deployment paths from applicable extreme external hazards. Specifically, the FLEX strategy for maintaining Containment integrity relies on the Reactor Building/Containment Vessel, along with MCCs located in the Auxiliary Building. The Reactor Building/Containment Vessel and Auxiliary Building are Seismic Category I structures that are designed to provide protection from the applicable extreme external hazards.
2.5.4.2     Components Inside Containment CNS relies on repowering a set of fans (Hydrogen Skimmer fans, LCVUs, and CAR~s) to maintain Containment temperature and pressure below acceptable limits. Hydrogen igniters are available to maintain hydrogen concentration below acceptable limits as defense in depth if the ELAP event degrades to core damage. All of these components are located inside the Reactor Building/Containment vessel, which is a Seismic Category I structure that protects equipment from external hazards.
2.5.4.3     Spray Strategqy Containment spray capability will not be required. An analysis was performed to validate that containment spray for temperature/pressure control is not required over the long term.
2.5.5        Plant Instrumentation The key parameter for the Containment integrity function is containment wide range pressure. Instrumentation will be powered by the FLEX Electrical Distribution System or a portable 120 VAC generator.
2.5.6        Thermal-Hydraulic Analysis CNS performed a thermal-hydraulic analysis to assess containment integrity using a GOTHIC model. Key conclusions from this analysis are as follows:
              *Containment pressure is expected to remain below the design pressure during the Phase 1 response.
               *The maximum pressure in containment will exceed the design pressure (15 psig) for a period of 12 hours (maximum of 17.5 psig). The Containment Vessel at CNS was initially pressure tested to 17.25 psig and has been analyzed to withstand a maximum of 72 psi. The temporary Containment pressure excursion beyond 15 psig during a BDBEE calculated by the GOTHIC model is therefore not significant.
              *The maximum temperature inside the SG and Pressurizer compartments will remain at or below the 2800&deg;F limit for the duration of the event, considering the planned ventilation actions that are part of the CNS FLEX strategy. The 280&deg;F limit was selected to ensure that reference leg flashing did not occur in the associated level indications.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2
                 *The annulus portion of the containment does not increase in pressure or temperature above design limits for the long term since it passively relieves through the annulus HVAC system exhaust dampers which fail open. The annulus calculation ensures the annulus pressure does not exceed the containment pressure and also ensures the annulus region can be accessed by personnel if necessary.
2,5.7       Electrical Analysis Containment pressure instrumentation will be powered by the FLEX Electrical Distribution System or a portable 120 VAC generator.
Hydrogen Skimmer fans used for containment ventilation and hydrogen igniters will be powered by the FLEX Electrical Distribution System. The LCVU fans and CARE will be powered from NSRC equipment. The repowering strategy is addressed in Section 2.3.11.1.
CNS performed an analysis to ensure that the 600V FLEX DGs had sufficient capacity to support the Phase 2 FLEX strategies. The analysis included electrical loads relevant for maintaining Containment integrity, such as battery chargers, hydrogen igniters, and the Hydrogen Skimmer fans. CNS performed comparable analyses for the 480V and 4160V NSRC DGs and the planned loads during Phase 3. This analysis concluded that the DGs and planned cable routing arrangement were adequate to support operation of the required equipment.
2.6    Characterization of External Hazards The following extreme external hazards were assessed for applicability for CNS:
* Seismic events
* External flooding
* Storms such as hurricanes, high winds, and tornadoes
* Extreme snow, ice, and cold
* Extreme heat 2.6.1.      Seismic Events The seismic hazard is applicable for CNS. Per NEI 12-06, Table 4-2, all sites will consider seismic events.
The CNS Updated Final Safety Analysis Report (UFSAR) states that the safe shutdown earthquake (SSE) has a ground acceleration design value of 0.15g acting horizontally and 0.10g acting vertically, and the operating basis earthquake (OBE) has a ground acceleration design value of 0.08g acting horizontally and 0.0533g acting vertically (FSAR, Section 3.1).
ONS is currently involved in the Expedited Seismic Evaluation Process (ESEP). The NRC has scheduled CNS to submit a Seismic Probabilistic Risk Assessment (SPRA) in 2019 to address the new Ground Motion Response Spectra (GMRS).
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.6.2      External Floodingq The external flooding hazard is applicable for CNS.
CNS is subject to external flooding from Probable Maximum Floods (PMFs) resulting from Probable Maximum Precipitation (PMP) events, Standard Project Floods (SPF) equal to 1/2 of the PMF, failures of upstream dams, a combination of dam failures and SPF's, seiche, hurricanes, and storm surge.
The ONS FLEX strategies reflect the site's current licensing basis. NTTF 2.3 flooding walkdowns and updated information on the flood hazard have resulted in a re-evaluation of the CNS flood hazard. CNS will assess potential impacts on the FLEX strategy resulting from this re-evaluation separately from this Final Integrated Plan (FIP).
CNS is susceptible to groundwater intrusion that may necessitate use of sump pumps to remove ground water from plant spaces.
2.6.3      Storms such as Hurricanes. Hi qh Winds, and Tornadoes The high wind hazard is applicable for CNS.
As described in UFSAR Section 2.1.1, the ONS site is located at latitude 35&deg;3'5" north and longitude 81&deg;4'10''. According to NEI 12-06, Revision 0, the location of CNS has a peak gust wind speed of 150-1 60 mph and a recommended tornado wind design speed of 175 mph. Based on the potential for winds in excess of 130 mph, the CNS site is susceptible to damage from severe winds from a hurricane or tornado.
2.6.4      Extreme snow, ice and cold The extreme cold (including snow and ice) hazard is applicable for ONS.
The CNS UFSAR contains environmental data from 1940 to 1980 obtained from the National Oceanic and Atmosphere Administration. The minimum temperature from the Catawba site was -5&deg;F. Because Catawba is in a temperate climate, storage and operation of the FLEX equipment in low temperatures is not considered to be an issue.
However, movement of FLEX equipment in snow and ice is a concern. CNS is located above the 35th parallel and is therefore subject to snowfall accumulation and extreme low temperatures per NEI 12-06, Revision 0. Based on NEI 12-06, the CNS site is also subject to the existence of an extreme amount of ice, and thus the potential for severe power line damage.
Based on UFSAR Section 2.4, water temperatures in the UHS will not reach a point where ice formation is an issue.    =
2.6.5      Extreme heat NEI 12-06, Revision 0 states that virtually every state in the lower 48 contiguous United States has experienced temperatures in excess of I110&deg;F and many in excess of 120&deg;F. In accordance with NEI 12-06, all sites will address high temperatures.
The CNS UFSAR contains environmental data from 1940 to 1980 obtained from the National Oceanic and Atmosphere Administration. The maximum temperature from the Catawba site was 104&deg;F. Because Catawba is in a temperate climate, storage and operation of the FLEX equipment in high temperatures is not considered to be an issue.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.7     Planned Protection of FLEX Equipment Storage and protection of FLEX equipment is discussed in this section. CNS evaluated the applicability of external hazards and addressed implementation considerations associated with each including:
* protection of FLEX equipment
* deployment of FLEX equipment
* procedural interfaces
* utilization of off-site resources CNS has one structure (i.e., the FLEX Storage Facility) for storing FLEX equipment that is located on-site and protects FLEX equipment from all applicable hazards. The FLEX Storage Facility contains sufficient equipment to satisfy redundancy requirements of NEI 12-06, Revision 0 for reliability and availability.
2.7.1         FLEX Storagqe Facility The FLEX Storage Facility at CNS is a single, large building located in the northwest area of the plant in a portion of the material lay-down storage area. This area is well above the flood level. The building is designed to resist seismic, wind forces, and tornado missiles of a magnitude that bounds all design basis hazards. The FLEX Storage Facility was seismically designed in accordance with the considerations presented in ASCE 7-10, and also meets the CNS SSE criteria.
The FLEX Storage Facility is a dome structure that has an outside diameter of 144 ft. It has two equipment doors on opposite sides of the building and two personnel access doors.
The building is located outside of the protected area approximately 500 feet west of the Shipping and Receiving Warehouse.
To ensure the dome storage area is functional, a monthly walkdown will be conducted in accordance with CNS procedures. Any issues identified as a result of the periodic walkdowns will be addressed in the site corrective action program.
FLEX equipment stored in the FLEX Storage Facility includes low pressure pumps, medium pressure pumps, high pressure pumps, 120V DGs, 600V DGs, hose trailers, a fuel transfer trailer, a CAT 924K, a pickup truck, portable spot coolers, 600V sump pumps, 120V sump pumps, portable transformers, portable panelboards, ventilation fans, and 230v sump pumps.
2.8    Planned Deployment of FLEX Equipment 2.8.1         Haul Paths and Accessibility CNS will use pre-defined deployment routes to transport FLEX equipment to the staging areas. The identified paths and deployment equipment positions will be accessible during all modes of operation. CNS will periodically perform walkdowns of the deployment paths to ensure pathways remain clear.
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ATTACH MENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1.& 2 Transmission lines can impede deployment of FLEX equipment along the pre-determined haul paths. In this case, CNS will ground the power lines and move them out of the deployment path. For other debris potentially impeding FLEX deployment, CNS has a CAT 924 Loader that can be used for debris clearing on the deployment path.
Once deployment paths are clear, CNS will dispatch deployment teams to transport Phase 2 FLEX equipment from the FLEX Storage Facility to deployment locations. One team will use the CAT 924 Loader and the other will use a Dodge 5500 truck. CNS has developed guidance on an appropriate sequence of deployment actions for each team.
The deployment routes for FLEX equipment begin with passage through the Vehicle Barrier Access Port (VBAP) and proceed along five primary on-site deployment paths as follows:
o  From the VBAP proceeding East along the southern bank of the SNSWP to the FLEX pump ramp access to the SNSWP.
o  From the SNSWP ramp through Gate #47 next to the ISFSI Area into the Protected Area and then South to the East side of the Auxiliary and Reactor Buildings. Pump discharge hoses may be deployed along this path for deployment strategies requiring a pump taking suction from the SNSWP and discharging through hoses to the suction of a second pump located outside the Auxiliary Building (e.g., for SG feedwater to support core cooling).
o  From the VBAP to the Vehicle Access Port (VAP) into the Protected Area proceeding East along the North side of Unit 2 connecting to the roadway along the east side of the Auxiliary Building.
oFrom the  VBAP proceeding directly south along the interior road circling the site on the south side of Unit 1 to the east side of the Auxiliary Building. Enter rear Gate #17  on the northeast side of plant into the Protected Area to the path along the east side of the Auxiliary Building and/or other established routes.
o  From the roadway on the East side of the Auxiliary Building inside the Protected Area, there are two paths that run west to FLEX equipment positions as follows:
* West to the north side of Unit 2 Diesel Generator Building
* West to the south side of Unit 1 Diesel Generator Building The deployment paths were evaluated for seismic stability and liquefaction and determined to be acceptable. A flooding event may inundate portions of the site, but flood waters will recede from all deployment paths and staging areas in time for FLEX deployment to meet all time constraints. Any changes to the CLB resulting from the Flood Hazard Re-evaluation will be addressed separate from this FIP. =...
CNS has instituted administrative measures to ensure that deployment paths and FLEX equipment deployment sites (including paths for portable cables, hoses, etc.) are maintained clear of other equipment or interferences. CNS performs general rounds of various plant areas on a daily shift basis and will identify FLEX interferences. Additionally, CNS will periodically perform specific inspections of FLEX-related areas and ensure they are maintained clear.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.8.2        Deployment of Strategqies CNS identified pre-determined deployment and staging locations for FLEX equipment necessary to support the FLEX strategies as discussed in the subsections below.
Human performance aids such as labels will be attached to specific portable FLEX equipment to identify the equipment with a unique number that agrees with FSGs.
2.8.2.1     Core Coolinq Makeup water for the CA system to the SGs via the TDAFWP will be supplied by the FLEX low pressure pump, which will be staged on a ramp at the edge of the SNSWP.
Suction hoses with strainers will be attached to the pump and placed into the SNSWP.
Water temperature in the SNSWP will not decrease to the point where access will be challenged due to ice formation.
SG feedwater will be delivered by a FLEX medium pressure pump that draws suction from the discharge of the low pressure pump. The Unit I medium pressure pump is normally deployed south of the Unit I Exterior Doghouse door. The Unit 2 medium pressure pump is normally deployed north of the Unit 2 FWST and west of the Unit 2 Containment Mechanical Equipment Building (CMEB). Discharge hoses from the FLEX medium pressure pump will be routed to the CA piping connections located in the Exterior and Interior Doghouses.
For Phase 3 deployment, procedures and FSGs were developed to connect NSRC equipment to station equipment. Use of portable pumps to supply feedwater to the SGs remains the same as Phase 2 until the RHR system is restored to service with supplemental power from the NSRC DGs or off-site power is restored.
2.8.2.2     RCS Makeup Two FLEX high pressure pumps (one for each unit) with hoses/adapters will be deployed to support RCS boration and inventory control. For Unit 1, the normal staging location for the FLEX high pressure pump is south and west of door AX658A (Unit 1 Electrical Penetration Room). For Unit 2, the normal staging location for the FLEX high pressure pump is north of door AX656B (Unit 2 Electrical Penetration Room). Suction hoses will be deployed to connect the portable pump to KF system connections from the FWST. A single FWST will supply borated water to the portable pumps for both units using a gated wye connection. Discharge hoses from the FLEX high pressure pumps will be connected to the discharge piping of NI pumps in the Auxiliary Building to supply injection makeup to the RCS.
2.8.2.3    FLEX Electrical Distribution System Two 600V portable DGs and cabling will be deployed to an area near the FWSTs of each Unit (one generator per Unit). The normal staging area for the Unit 1 FLEX 600V DG is east of the Unit 1 Turbine Building, in close proximity to the Unit 1 MG set room door. The normal staging area for the Unit 2 FLEX 600 V DG is east of the Unit 2 Turbine Building, in close proximity to the Unit 2 MG set room door. Alternate staging areas are near the waste shipping/auxiliary access point area doors, the hot machine shop area doors, and the Unit I RC Pit / Safe Shutdown Facility Area.
The portable 600V DGS will be connected to the normal plant MCCs for re-powering equipment via the FLEX Backbone using a combination of permanently installed cables, portable panelboards, and transformers. Cables and connectors of the FLEX 600V distribution system are color coded to ensure proper phase rotation.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 During Phase 3, DGs from the NSRC will be available to provide additional power. For Unit 1, the NSRC 41 60V DGs and switchgear will be staged south of the Diesel Generator Building and the NSRC 480VI DGs will be staged south and west of the Auxiliary Building, near the FLEX high pressure pump. For Unit 2, the NSRC 4160V DGs and switchgear and the NSRC 480V DGs will be staged near the west side of the Diesel Generator Building.
2.8.2.4      Sump Pumps CNS can deploy FLEX sump pumps to ensure that equipment for core cooling is not compromised by flooding. The normal FLEX Sump Pump staging area is at the rear of the Auxiliary Building near the Clean Trash Room. Alternate staging areas include the area outside the Unit 1 MG set room and the area outside the Unit 2 MG set room.
2.8.2.5      Small Diesel Generators Eleven portable DGs (6 kW) are available to be deployed from the FLEX Storage Facility to the location where they are needed. These 6 kW DGs will be used to power battery chargers for hand-held radios and portable satellite phones, fans, small sump pumps, and other identified loads.
2.8.3         Fuelina of EauiDment Diesel Fuel Oil (DFO) for the FLEX equipment will be obtained from the safety-related, underground Emergency Diesel Generator (EDG) fuel oil storage tanks. The diesel fuel will be pumped out of the underground tanks and transferred to the portable Diesel Fuel Storage Tank using a diesel driven Portable Transfer Pump or will be gravity drained into small containers in the EDG room and hand carried to the location needed. Connections on the piping from the EDG tanks that will be used for removing fuel oil are seismically qualified and located above the flood level in the yard. The portable fuel tank will be pulled by the FLEX Pickup Truck or other available vehicles to the various site locations for refueling portable diesel generators.
CNS performed a fuel consumption calculation to determine the estimated total fuel required to support FLEX equipment operation. The estimated fuel required to support FLEX equipment for 24 hours is 2,027 gallons. FLEX equipment will be stored full of DFO and refueled as necessary. The CNS fuel consumption calculation determined that 485 gallons of DFO would be required for refueling in the first 24 hours, which is well within the capacity of the EDO DFO storage tanks.
To avoid multiple pieces of FLEX equipment running out of fuel at the same time, CNS will attempt to maintain individual fuel tanks greater than half full. The small DGs may require refueling prior to the FLEX Diesel Fuel Transfer Trailer being available. If necessary, CNS will dispatch personnel to transport required fuel by hand.
During Phase 3, additional diesel fuel will be brought onsite from outside resources as required.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.9Seuence* of Events* and Staffingq 2.9.1      Sequence of Events The Table below presents a Sequence of Events (SOE=) Timeline for an ELAP/LUHS event at CNS. CNS Simulator Observation was conducted to time-validate EOP and FSG actions.
For other actions, time-validations involved combinations of local FSG time-validations, local (EOP) action time-validations, timed table-top procedure readings, and timed repetitive actions, such as cable and hose deployments. The analyses were conducted in incremental fashion, typically beginning with an assumed time of EOP or FSG implementation, and progressing through completion of the action or procedure. The analyses utilized ratio analysis to project times to complete actions such as deployment of multiple sections of cabling, hose, couplings, etc. Elements documented sequentially included observed and projected times to complete actions, time for procedural progression, decision making, necessary, briefings for implementers, and time required to perform manual deployments and repetitive actions, such as walking to required destinations, towing equipment to required destinations, carrying and deploying cable and hose segments, etc.
All FSGs for FLEX strategies were validated per normal procedural requirements to ensure hands-on equipment/component verification and the mechanics of executing procedure steps were practical and accurate.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 &.2
                                                        ~~Sequence of Events Timeline                                 ,
Ato                   Str ieCompletion trTie        Time                                    'Remarks / APplicability Acin(hours)                          (Hours),
Event Starts                                                0         N/A         Plant @100% power TDAFWP starts on SBO                                        0         N/A         Installed plant equipment. Automatic Action.
TDAFWP automatically aligns to embedded RC                                        Action occurs if normal water sources are damaged. Normal water sources header piping if normal water sources are                   0          48         are not robust to all hazards, so conservative timeline aligns to RC damaged.                                                                          immediately. Embedded RC system has at least 48 hours of inventory.
Control SG level by starting / stopping TDAFWP with Trip and Throttle Valve from                   0.5     As needed      Maintain SG level. Control board manipulation.
Control Room.
Initial cooldown and depressurization of RCS                                      Establish conditions for reactor makeup, reduce seal leakage, preserve (to a temperature and pressure that does not               0.6         2.6        sel.Isaedpntqumn.CorlBadmnilto.
result in nitrogen injection)ses.IsaldpnteipetCorlBarmnputo.
Isolteontanmet nstrmen Airto                1            .~Time          constraint (38 hr). Analysis shows containment pressure, remains below design for the first 38 hours even if Instrument Air is not isolated.
Dateiscnectalnnciiallasfosia                               2.5        2.75        Time constraint (3.5 hr). Required to preserve Vital Batteries.
Time constraint (6 hr). Provides SEP vent path to outside of building to OpenSEPbay doo ersnne                     3           .04       avoid adverse Auxiliary Building conditions due to SFP steam.
Provide pumping capacity to control level in              36           .6         Tm      osrit(        r TDAFWP put sump Align charging to Channel A and D Vital                     61.5                  Tm      osrit(15h          o   ,1. rfrD     4h o          )
Batteries (B and C included)                                61.5                  Tm      osrit(15h          o    ,1. rfrD     4h o           )
Align portable injection pump from FWST to NI System to provide RCS makeup and                        6.5        11.6        Time constraint (13.8 hr) boration.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2
* Sequence of Events Timeline Action..                  Start Time       Time                                  "RemarksIApplicability        "...          *     ,
Action     *         ~    ~(hours)    (Hours)....                                                                                r Ample handheld/headlamp lighting would be available at 0 hr with outside Provide portable lighting (beyond head and               8+         As needed    resources available to provide additional lighting and replacement batteries hand lamps and installed battery lighting)                                       after 6 hr. Portable lighting would be placed into service as time and resources allow.
Re-power H2 igniters.                                     9            9.35      This action is required by NEI 12-06 as a contingency.
Power A and B WZ (Groundwater Drainage)                   101.5                  Time constraint (10.6 hr). Prevents ground water from flooding Aux Bldg El.
Sump Pumps                                                101.5                  543' and below.
Install portable fans in Control Room and113.3                                   Tmcosrit(0h)
Battery Rooms Rehag cmuncain yte     adsaelie12                        14       Time constraint (24 hr). Portable DG will be used to recharge batteries.
phone system Install 600 V sump pump on the 522' elevation             12           16.6      Time constraint (36 hr)
Fan            2023.32                 Time constraint (24 hr). Analysis of containment heat loads is basis for Start Hydrogen Skimmer                                                           these2actions Connect FLEX low pressure pump through RN piping to SEP skimmer loop to provide a               243.3                   Time constraint (40 hr). Critical pool level assuming worst case heat means to make up to the SEP without entering             243.3                   load/evaporation rate is reached after 40 hours.
the SEP area.
Install NSRC DGs.                                      24 -48       As needed Isolate the CLAs.                                        42            44.1      Required prior to depressurizing the SGs below 240 psig to prevent gas intrusion to ROS that would disrupt natural recirculation cooling.
Evaluate need to provide freeze protection for instrumentation located in Doghouses and                 48         As needed    Time constraint (48 hr) yard.
Start Lower Containment Ventilation Units                48             48       Time constraint (48 hr)
Start Containment Return Fan                             48             52       Time constraint (52 hr)
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ATTACH MENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.9.2        Staffingq Using the methodology of NEI 12-01, Guideline for Assessing Beyond Design Basis Accident Response Staffing and Communications Capabilities,an assessment of the capability of the on-shift staff and augmented ERO to respond to a BDBEE was performed; The assumptions for the NEI 12-01 Phase 2 scenario postulate that the BDBEE involves a large-scale external event that results in:
* An extended loss of AC power
* An extended loss of access to ultimate heat sink
* Impact on units (all units are in operation at the time of the event)
* Impeded access to the units by off-site responders as follows:
o   0 to 6 Hours Post Event   -  No site access.
o   6 to 24 Hours Post Event - Limited site access. Individuals may access the site by walking, personal vehicle or via alternate transportation capabilities (e.g.,
private resource providers or public sector support).
o   24+ Hours Post Event    -  Improved site access. Site access is restored to a near-normal status and/or augmented transportation resources are available to deliver equipment, supplies and large numbers of personnel.
The on-shift ERO analysis concluded that current CNS on-shift staffing present for the "no site access" 6-hour time period is sufficient to perform the EOP, FSG, and emergency response tasks.
The augmented ERO analysis concluded that sufficient personnel resources exist in the current CNS augmented ERO to fill positions for the expanded ERO functions. Thus, ERO resources and capabilities necessary to implement coping strategies after 6 hours exist in the current program.
To conduct the on-shift portion of the assessment, a team of subject matter experts from Operations, Maintenance, Security, Radiation Protection, Chemistry, Engineering, Emergency Preparedness, and industry consultants conducted tabletop exercises. The participants reviewed the assumptions and existing procedural guidance, including applicable draft FSGs for coping with a BDBEE using minimum on-shift staffing. Particular attention was given to the sequence and timing of each procedural step, jits duration, and the on-shift individual performing the step to account for both the task and time motion analyses of NEI 10-05, Assessment of On-Shift Emergency Response Organization Staffing and CaPabilities.
In the event of a natural or any other type disaster that renders roads and passage ways unusable, CNS essential staff will be airlifted from two preplanned landing zones. The northern landing zone would be the Duke Energy Shelby Operations Center. The southwest Landing zone would be the Duke Energy Chester Operations Center. The two landing zones are Duke Energy properties and can provide shelter, communications and storage capacity for necessary gear. The helicopter provider will be under a contract or agreement letter to provide support.
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ATTACH MENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.10    Offsite Resources The Strategic Alliance for FLEX Emergency Response (SAFER) team is contracted by the nuclear industry through Pooled Equipment Inventory Corporation (PEICo) to establish National SAFER Response Centers (NSRC) operated by Pooled Inventory Management (PIM) and in collaboration with AREVA to purchase, store, maintain and deliver emergency response equipment in the case of a major nuclear accident or BDBEE in the United States.
The NRC letter dated September 26, 2014 (ADAMS Ascension No. MLI 4265A1 07) titled "Staff Assessment of National SAFER Response Centers Established in Response to Order EA         049" (Reference 12) endorsed NEI's White Paper titled "National SAFER Response Centers" (Reference 13). NRC concluded that SAFER procured equipment, implemented appropriate processes to maintain the equipment, and developed plans to deliver the equipment needed to support site responses to BDBEEs, consistent with NEI 12-06 guidance and the SAFER Response Plan to meet Phase 3 requirements of Order EA-12-049.
CNS relies on equipment stored off-site for Phase :3 of the FLEX strategy. Equipment may be provided from NSRCs. Another nuclear plant may also provide Phase 3 equipment, if response would be faster than from the NSRCs.
2.10.1       National SAFER Response Center (NSRC)
The SAFER Response Plan for CNS (Reference 14) contains (1) SAFER control center procedures, (2) National SAFER Response Center procedures, (3) logistics and transportation procedures, (4) staging area procedures, which includes travel routes between staging areas to the site, (5) guidance for site interface procedure development, and (6) a listing of site-specific equipment (generic and non-generic) to be deployed for FLEX Phase 3.
Two NSRC's are strategically located across the country in Memphis, TN and Phoenix, AZ.
If possible, NSRC equipment would be delivered to CNS by trucks from the Memphis location.
If possible, NSRC equipment will be delivered to Staging Area C, which is the Kings Mountain Training Center (34 miles away from the CNS site by driving). When CNS is ready, NSRC equipment will then be delivered to Staging Area B, which is an on-site parking lot at the CNS site outside the protected area. NSRC Equipment can also be delivered directly to Staging Area B, if requested by CNS.
Primary and alternate driving routes from Staging Area C to Staging Area B have been identified. CNS will coordinate with local and state authorities to assess the condition of roads and bridges along the travel path. If ground transportation from Staging Area C to Staging Area B is not feasible, NSRC equipment can be delivered to Staging Area B by helicopter airlift.
Two access routes from Concord Road to Staging Area B have been identified: the primary access location is through the Main Entrance, Northwest Gate; the secondary access path is through the Main Entrance, Southwest Gate.
The SAFER Response Plan for CNS does not include a Staging Area D.
The first arriving equipment will be delivered to the site within 24 hours from initial contact and remaining equipment will be delivered within 72 hours from initial contact.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.10.2       Equipment The NSRC will provide equipment as listed in the response plan. The NSRC will deliver the first pieces of equipment within 24 hours from initial contact. Such priority equipment includes Medium Voltage Generators, a Low Voltage Generator, 4160 VAC Distribution System, a Low Voltage Step-Up Transformer, a Low Pressure / High Flow Dewatering Pump, Mobile Lighting Towers, a Diesel Fuel Transfer Pump, a Water Treatment System, and a Mobile Boration Unit. The generic set of NSRC equipment as identified in the plan provides back up to on-site FLEX equipment (e.g., pumps, DGs) and will be provided as lower priority items to arrive within 72 hours from initial contact. NSRC equipment connections to applicable hoses and/or plant equipment are compatible or necessary adapters are available.
Other offsite resources may be obtained as needed to support the event which may include diesel fuel oil, equipment from other nuclear plants, and equipment from vendors.
2.11    Habitability and Operations 2.11.1       Equipment Coolincq CNS performed HVAC analysis to evaluate temperature increases resulting from an ELAP /
LUHS in the control room, TDAFWP room, and other rooms of the Auxiliary Building.
Considering the actions planned to provide ventilation as part of the FLEX strategies (discussed below), these analyses concluded that temperatures would not challenge essential equipment.
To enhance MCR temperature margins and personnel comfort, various doors will be opened to provide an air exchange flow path with the outside. Additional cooling will be provided by four portable, water-cooled coolers powered from the FLEX Electrical Distribution System.
The cooling water supply will be provided by a branch from the FLEX Raw Water Distribution System.
Analyses validated TDAFWP room temperatures remain acceptable with no additional action.
Portable fans will be placed in the vital battery area during battery charging to ensure hydrogen accumulation does not exceed flammability limits. Additional ventilation needs, such as fans, will be provided from the NSRC and used if required.
2.11.2       Freeze Protection If freezing weather exists, CNS will take appropriate actions to ensure protection of the FLEX strategies from cold conditions. As necessary, CNS will start FLEX diesel-powered equipment early due to the potential for difficult starting and CNS will establish trickle drains from idle equipment and hoses to prevent formation of ice plugs.
CNS can also provide heat tracing for equipment in the FLEX Storage Facility, FWST level instrumentation, and pressure instruments in the Doghouses. Other than components associated with monitoring FWST level, there is no FLEX-credited instrumentation in the yard or exposed to cold temperature to the extent that freezing would be expected. Heat tracing equipment may be deployed from the FLEX Storage Building to maintain the FWST level instrumentation functional. Heat tracing may also be deployed for other instrumentation as a contingency. Heat tracing cables could be powered by any portable DG supplying 120V power.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.11.3        Hvdroaqen Ventilation The minimum concentration of hydrogen gas to result in an explosive mixture is 4%.
Without mitigation, CNS determined that hydrogen concentrations in the battery rooms would not exceed 4% (i.e., the lower explosive limit) for at least 21 days. Regardless, CNS plans to deploy fans in the battery room areas as part of the FLEX strategy as time permits.
Battery rooms are ventilated by pulling outside air through the U2 MG set room, down the Unit 2 Auxiliary Building stairwell, through the battery rooms, and up the Service Building stairwell to outdoors through an open door, creating a chimney effect. CNS will open a series of doors to create this air flow path, all of which will be opened within 24 hours.
2.11.4        Personnel Habitability CNS will set up spot coolers in the control room to maintain temperature less than 90&deg;F.
Cooling water will be provided to the spot coolers using 5-inch raw water distribution hose, the fire protection header, or the "B" train service water system. Coolers will start operating when the first source of water becomes available (most likely the 5-inch hose or the fire protection header).
For providing water from the 5-inch FLEX hose or the fire protection header, CNS will route hosing through door AX(657 (Unit 2 Penetration Room) and AX657F (Control Room access to Unit 2 Auxiliary Building Corridor at 594'). For providing water from the "B" train RN system, CNS will route hose through AX(657G (Control Room rear access). Cooling water that has been used by the room coolers will be routed through a water discharge manifold to a yard drain. CNS will also deploy a condensate drain manifold and hosing to route condensation from the spot coolers to a yard drain.
2.12        Water Sources Discussion of credited water sources for the FLEX strategies is included in the sections above for each individual strategy.
2.12.1        SG Feedwater For SG feedwater, CNS will provide water from any of the following sources:
* USTs
* CACST
* Hotwell
* RC system piping embedded volume
* SNSWP The embedded RC system captured volume and the SNSWP are the credited sources of water because of their robustness to the applicable hazards. The USTs and the hotwell are not protected from external hazards. These tanks are normally aligned as a TDAFWP suction source, but automatic realignment to embedded RC system captured volume is provided if the default sources are not available.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 The UST and hotwell have condensate grade water that will not degrade the SGs. CNS may switch to a raw water source for a limited duration, if necessary. In this case, water purification equipment from the NSRC will eventually be deployed to establish a clean water source prior to excessive exposure of the SGs to raw water. ONS will also consider actions to clean up the SGs including blowdowns or chemical additions.
2.12.2        RCS Makeup For NC system boration during Phase 2, CNS will provide borated water from one or both of the following sources:
* FWSTs
* CLAs Alternate sources of borated water include the following options:
* NSRC-supplied mobile boration skid
* Trucking from an off-site source (e.g., McGuire Nuclear Station)
* Recovery of borated water from the FWST annulus (if damaged)
* Portable FLEX drop tanks mixing boron and raw water 2.12.3        SEP Makeup For inventory control of the SFP, CNS uses raw water from the RN system, which is pressurized by the FLEX low pressure pump using the SNSWP as the suction source. The SNSWP will be available following the applicable extreme external hazards.
During Phase 3, CNS may transition to a clean water source (e.g., NSRC-supplied water purification unit) when available.
2.13    Shutdown and Refuelingq Analysis Order EA-12-049 requires that licensees must be capable of implementing the mitigation strategies in all modes. In general, the discussion above focuses on a BDBEE occurring during power operations. This is appropriate, as plants typically operate at power for 90 percent or more of the year. If the BDBEE occurs with the plant at power, the mitigation strategy initially focuses on the use of a pump coupled to a steam-powered turbine to provide the water initially needed for decay heat removal. If all or most of the fuel has been placed in the SFP, there is a shorter timeline to implement the strategy for providing SEP makeup water. However, this is balanced by the fact that if immediate cooling is not required for the fuel in the reactor vessel, the operators can concentrate on providing makeup to the SEP and the number of personnel on-site is much greater during an outage. ONS analysis shows that following a full core offload to the SFP, about 76 hours are available to implement makeup before boil-off results in the water level in the SFP dropping far enough to uncover fuel assemblies. About 48 hours are available prior to level decreasing to 10 feet above the top of the fuel assemblies. As previously discussed, CNS can provide SEP makeup in advance of these times.
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ATTACHMENT 6 FINAL INTEGRATED PLAN catawba Nuclear Station, Units 1 & 2 When a plant is in a shutdown mode and steam is not available to operate the steam-powered pump, another strategy must be used for decay heat removal. On September 18, 2013, NEI submitted to the NRC a position paper entitled "Shutdown Refueling Modes" (Reference 15),
which described methods to ensure plant safety in shutdown modes. By letter dated September 30, 2013 (Reference 16), the NRC staff endorsed this position paper as a means of meeting the requirements of the Order. In the third six-month update dated August 28, 2014 (Reference 17),
CNS committed to follow the guidance in this position paper.
CNS's FLEX strategy in Modes 5 and 6 relies on a FLEX Medium Pressure Pump to provide borated makeup to the RCS. This pump will be staged locally at one of four potential locations around the Auxiliary Building. The suction supply for the portable pump will come from a connection on the FWST supply line for KF/FW makeup to the SEP (see Section 2.3.5.3). The discharge from the portable pump will be into the Train A NI pump discharge piping that feeds the RCS hot or cold legs (alternate connection on the Train B NI pump discharge piping)
(see Section 2.3.5.4).
If the reactor vessel head is installed, RCS system depressurization will be initiated through the reactor vessel head vents.
FWST inventory may only last 40 hours if RCS feed and bleed cooling is utilized. FWST makeup, use of alternate borated water strategies, or RHR pump restoration will be provided prior to loss of FWST level.
For the containment integrity function, CNS would open the equipment hatch to establish an emergency vent path from containment, if possible. The equipment hatch would only be opened if core cooling is expected to be maintained.
2.14    Procedures and Training 2.14.1         Procedural Guidance The inability to predict actual plant conditions that require the use of BDBEE equipment makes it impossible to provide specific procedural guidance. As such, the FSGs provide guidance that can be employed for a variety of conditions. FSGs, to the extent possible, provide pre-planned FLEX strategies for accomplishing specific tasks in support of EOPs and Abnormal Operating Procedures (AOPs). FSGs are used to supplement (not replace) the existing procedure structure that establishes command and control for the event.
Procedural interfaces were incorporated into ECA-0.0, "Loss of All AC Power" to the extent necessary to include appropriate reference to FSGs and provide command and control for the ELAP.
2.14.2        Trainingq Programs and controls were established to assure personnel proficiency in the mitigation of BDBEE is developed and maintained. The Systematic Approach to Training (SAT) process was utilized to evaluate, develop and implement training for applicable personnel.
Initial training was provided and continuing periodic training will be provided to site emergency response leaders on BDBEEs, emergency response strategies, and implementing guidelines. Personnel assigned to direct the execution of mitigation strategies for BDBEEs received the necessary training to ensure familiarity with the associated tasks,
            .considering available job aids, instructions, and mitigating strategy time constraints.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 &.2 Care was taken to not give undue weight (in comparison with other training requirements) for operator training for BDBEE accident mitigation. The testing/evaluation of operator knowledge and skills in this area was similarly weighted.
A table top analysis was conducted by the Duke fleet ERO trainers using a graded Systematic Approach to Training. The analysis determined no impact on the existing ERO position specific Job Task Analyses (JTAs). Staffing resources for the expanded response capability ERO are provided from a qualified depth of ERO personnel. ERO depth is outlined and maintained per NSD1 17 (ERO Staffing Training and Responsibilities.) ETQS 3100 section 5.2 states, "Job Task Analysis (JTA) should be completed when no previous job data exist, is incomplete, or when revising a job." This analysis referenced the Duke Energy NEI 12-01, Phase 2 staffing assessment report. As a result of the analysis, no new skills, knowledge or tasks, were identified since current qualified ERO resources will be augmented for expanded response capability.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2
: 3. Acronyms AOP - Abnormal Operating Procedure ATWS - Anticipated Transient Without Scram AWWA - American Water Works Association BAT - Boric Acid Tank BOB - Beyond Design Basis BDBEE - Beyond Design Basis External Event BW - Steam Generator Wet Layup Recirculation System CA - Auxiliary Feedwater System CACST - Auxiliary Feedwater Condensate Storage Tank CARF - Containment Air Return Fan CFR - Code of Federal Regulations CLA - Cold Leg Accumulator CMEB - Containment Mechanical Equipment Building CNS - Catawba Nuclear Station DG - Diesel Generator EDG - Emergency Diesel Generator ELAP - Extended Loss of AC Power EOP - Emergency Operating Procedure EPRI - Electric Power Research Institute ERO - Emergency Response Organization ESEP - Expedited Seismic Evaluation Process FCV - Flow Control Valve FIP - Final Integrated Plan FLEX - Diverse Flexible Coping Strategies FSG - FLEX Support Guideline FW - Refueling Water System FWST - Refueling Water Storage Tank GMRS - Ground Motion Response Spectra HCPLF - High Confidence of Low Probability of Failure HVAC - Heating Ventilation and Air Conditioning JTA - Job Task Analysis KC - Component Cooling System KF - Spent Fuel Pool Cooling System LCVU - Lower Containment Ventilation Unit LOOP - Loss of Offsite Power LUHS - Loss of Ultimate Heat Sink MCR - Main Control Room MG - Motor Generator NEI - Nuclear Energy Institute NI - Safety Injection System NR - Narrow Range NRC - Nuclear Regulatory Commission Page 41 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 NSRC - National SAFER Response Center NTTF - Near-Term Task Force OBRE - Operating Basis Earthquake PEiCo - Pooled Equipment Inventory Corporation PIM - Pooled Inventory Management PMF - Probable Maximum Flood PMP - Probable Maximum Precipitation PORV - Power Operated Relief Valve RC - Condenser Circulating Water System RCP - Reactor Coolant Pump RCS - Reactor Coolant System RHR - Residual Heat Removal System RN - Nuclear Service Water System SA - Main Steam to Auxiliary Equipment System SAFER - Strategic Alliance for FLEX Emergency Response SAT - Systematic Approach to Training SBO - Station Blackout SQE - Sequence of Events SG - Steam Generator SFP - Spent Fuel Pool SNSWP - Standby Nuclear Service Water Pond SQUG - Seismic Qualification Utility Group SFF - Standard Project Flood SPRA - Seismic Probabilistic Risk Assessment SSE - Safe Shutdown Earthquake SV - Main Steam Vent to Atmosphere System TDAFWP - Turbine Driven Auxiliary Feedwater Pump TIA - Task Interface Agreement TS - Technical Specifications UFSAR - Updated Final Safety Analysis Report UHS - Ultimate Heat Sink UST - Upper Surge Tank WR - Wide Range Page 42 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2
: 4. References
: 1. Recommendations for Enhancing Reactor Safety in the 21 st Century; The Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, July 12, 2011
: 2. NRC Order EA-12-049, Issuance of Order to Modify Licenses with regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, March 12, 2012. (ML12054A735)
: 3. NEI 12-06, Rev. 0, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, August 2012.
: 4. NRC Interim Staff Guidance JLD-ISG-2012-01, Compliance with Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events. (MLI12229A1 74)
: 5. Duke Energy (Henderson) letter CNS-1 5-035 dated May 1, 2015, "Notification of Full Compliance with Order EA-12-049, 'Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events' and with Order EA-1 2-051, 'Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation' - Catawba Nuclear Station Unit 2." (ML15126A277 is NRC ADAMS Database)
: 6. Duke Energy (Henderson) letter CNS-16-005, "Final Notification of Full Compliance with Order EA-12-049, 'Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events' and with Order EA-1 2-051, 'Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation' for Catawba Nuclear Station."
: 7. NRC Order EA-1 2-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation.
: 8. NEI 12-02, Rev. 1, Industry Guidance for Compliance with NRC Order EA-1 2-051 to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, August 2012.
: 9. NRC Interim Staff Guidance JLD-ISG-2012-03, Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation.
: 10. NRC letter dated September 12, 2006, "Final Response to Task Interface Agreement (TIA) 2004-04, 'Acceptability of Proceduralized Departures from Technical Specifics (TSs) Requirements at the Surry Power Station,' (TAO NOs. MC4331 and MC4332)." (ML060590273 in NRC ADAMS Database)
: 11. NRC (Davis) letter to PWROG (Stringfellow), dated January 8, 2014. (ML13276A183 in NRC ADAMS database)
: 12. NRC (Davis) letter to NEI (Pollock), dated September 26, 2014, "Staff Assessment of National SAFER Response Centers Established in Response to Order EA-12-049." (ML14265A107 in NRC ADAMS Database)
: 13. NEI (Pollock) letter to NRC (Davis), dated September 11, 2014, "National SAFER Response Center Operational Status," with Enclosure "White Paper; National SAFER Response Centers."
(ML14259A222 & ML14259A2.23 in NRC ADAMS Database)
: 14. Areva, Inc. Engineering Information Record 51-9233066-0002, "Catawba SAFER Response Plan."
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2
: 15. NE! Position Paper, "Shutdown / Refueling Modes", Rev. 0, dated September 18, 2013.
(ML13273A514 in NRC ADAMS Database)
: 16. NRC (Davis) letter to NEi (Pollock), dated September 30, 2013. (ML13267A382 in NRC ADAMS Database)
: 17. Duke Energy letter CNS-14-086 dated August 28, 2014, "Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)." (ML14247A232 in NRC ADAMS database)
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ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT
: 1.      Background and Purpose NRC Order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," required licensees to develop, implement, and maintain guidance and strategies to maintain or restore core cooling, containment and spent fuel pool cooling capabilities following a beyond-design-basis external event. To develop strategies for maintaining/restoring core cooling, licensees evaluated reactor coolant system (RCS) leakage from reactor coolant pump (RCP) seals during an extended loss of all AC power (ELAP).
NSAL-1 4-1, Revision 1 was issued by Westinghouse on September 8, 2014 and it documents that the nominal RCP seal leakage rate of 21 gallons per minute (gpm), as documented in WCAP-1 0541, Revision 2, may be not be applicable for all plants using Westinghouse RCPs with standard seal designs because of the various thermal-hydraulic conditions set up by plant-specific seal leak-off piping designs.
PWROG-14015-P, Revision 2 was issued by the PWR Owner's Group in April 2015 to determine revised No. 1 RCP seal leak-off flow rates following an ELAP.
PWROG-1 4027-P, Revision 3 was issued by the PWR Owner's Group in April 2015 to evaluate the time to enter reflux cooling and the time at which the core uncovers based on the revised seal leak-off flow rates during an ELAP.
Following issuance of the Watts Bar Mitigating Strategies Safety Evaluation dated March 27, 2015, the NRC requested via e-mail dated March 31, 2015 that licensees with standard Westinghouse RCP seal packages review the technical content therein and provide information addressing similar issues. Specifically, the NRC communication stated:
"At the present time the NRC staff is unable to conclude that Westinghouse's analytical modeling of ROP seal leakage is acceptable on its own merits. However, for the purposes of mitigating strategies, the staff can balance the modeling uncertainties and deficiencies of the model with the unique aspect of FLEX. To expedite individual plant resolution, licensees could provide a brief discussion about the margin for RCS makeup time, based on the favorable aspects of individual site mitigating strategies."
In addition,* the NRC provided examples of pertinent information to include in the Margin Assessment.
The purpose of this Margin Assessment is to provide a discussion regarding the margin for RCS makeup time, specifically addressing the examples of pertinent information regarding seal leakage provided by the NRC.
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J ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT
: 2.        RCP Seal Leak-Off Line Configuration Catawba is a four-loop Westinghouse PWR utilizing Model 93A reactor coolant pumps, using standard Westinghouse seal packages. The Catawba RCS loops utilize inverted U-tube type steam generators. Catawba's Mitigating Strategies (FLEX) response is based on the established RCP seal leakage profile as identified in WCAP-1 7601-F, revision 1 "Reactor Coolant System Response to the Extended Loss of AC Power Event for Westinghouse, Combustion Engineering and Babcock & Wilcox NSSS Designs".
In late 2014 and early 2015, as a result of Westinghouse NSAL-1 4-1, Catawba contracted with MPR Associates to have the existing RCP No. 1 and No. 2 seals and the associated No. 1 seal leak-off piping evaluated for an extended loss of seal cooling event, such as an ELAP. Due to time constraints, this effort was performed in parallel with the on-going PWROG initiative to resolve issues associated with the established RCP seal leak-off rates during a Loss Of Seal Cooling (LOSC) event. The MPR RCP seal model is different from the Westinghouse seal model being used in the PWROG work, in that the MPR model accommodates a transient analysis for evaluation of known pressure spikes during the early stages of the LOSC event.
The Westinghouse RCP seal model does not currently allow for evaluation of transient behavior.
As a result of the MPR seal analyses, Catawba determined that a modification to the No. 1 RCP seal leak-off piping 0.254 inch bore restriction orifice positioned downstream of the seal exit was required. This modification installed a thicker orifice plate to accommodate the higher differential pressures associated with an extended LOSC event. This modification did not change Catawba's leak-off configuration categorization discussed in the PWROG-14015-P, revision 2 report dated April 2015 (Catawba remained a Category 1 plant).
Catawba is not crediting the MPR analyses for ELAP response or compliance with Order EA   049 even though the officially credited PWROG work to resolve remaining open issues is not yet complete. The MPR analysis results for both seal leak rate and the attendant leak-off piping pressure-temperature conditions during an ELAP/LOSC event show Catawba seal leak rates (post orifice plate modification) to be bounded by the documented Westinghouse leakage results as identified in the PWROG-14027-P revision 3 report, dated April 2015. Additionally, in-house piping stress/support and component evaluations of the RCP No. I seal leak-off lines shows the modified system retains its integrity throughout the transient predicted by the MPR seal model, as well as at more extreme conditions.
During the October 2014 NRC FLEX Audit, the above information was discussed with the NRC audit team and with ONRR, and subsequently an information package was placed on the Catawba e-Portal (which has now been moved to a SharePoint) for technical staff information/use. Catawba has installed thicker 0.254 inch bore restriction orifices in all four RCP No.1 seal leak-off lines on each Unit and declared Unit 2 to be in compliance with Order EA-12-049 in March 2015 and Unit I to be in compliance in December 2015. The status of the Catawba response to elevated RCP No. 1 seal leak-off line conditions and orifice plate modification was updated in the EA-1 2-049 Fourth Six-month Status Report dated February 28, 2015.
In March 2015, specific transient conditions potentially requiring further evaluation of the RCP No. 1 seal leak-off piping were identified by PWROG via Westinghouse NSAL-1 5-2. This NSAL Page 2 of 8

ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT formally identifies the existence of a potential 2045 psia pressure spike that occurs at the No. 1 seal exit early in the LOSC transient and its potential effect on the seal leak-off line, a transient the Westinghouse seal model cannot specifically evaluate as noted previously. As a result of this model limitation, the NSAL recommends Licensees assume a conservatively high seal exit pressure and temperature in the leak-off piping to account for the pressure spike for evaluation of system response to an ELAP. While the current Catawba MPR ELAP transient analysis predicts lower pressure and temperature conditions than those recommended by NSAL-15-2, an additional analysis case was run by MPR with a 2045 psia pressure (the NSAL-15-2 recommendation) as a forced input at the seal exit and another analysis at 2500 psia. This analytical approach removes reliance on the MPR seal model entirely and allows for an independent thermal-hydraulic evaluation of the RCP No. 1 seal leak-off line. Based on the MPR evaluation and resulting piping and support analyses performed by Duke, the ROP No.1 seal leak-off piping remains adequately protected from this extreme pressure transient with the thicker restriction orifice plate installed. As such, the published PWROG Category 1 ELAP leak off rates still apply for Catawba.
: 3.       Margin Assessment This margin assessment was performed using the examples of pertinent information regarding seal leakage provided by NRC via e-mail dated March 31, 2015. This assessment highlights the favorable aspects of the Catawba FLEX strategy and identifies areas with margin.
3.1      Early RCS Cooldown Per ECA-0.0 response, symmetric RCS cooldown/depressurization at Catawba is started within 1-2 hours of ELAP onset to minimize RCS inventory loss and protect the ROP seal packages.
Post-event initiation, RCS conditions at Catawba will peak at 2485 psig and 5680 F until cooldown commences. The Catawba RCP Model 93A seal packages contain 0-rings made from 7228C elastomer material, which has been evaluated to withstand up to 582&deg;F for eight hours. Early initiation of RCS cooldown therefore provides assurance the RCP seals will continue their function to limit leak off flow.
Additional Favorable Cooldown Information Along with facilitating RCS conditions favorable for passive injection of highly borated water from the Cold Leg Accumulators, the Catawba RCS cooldown strategy over the first 24 hours of the ELAP event specifically supports the integrity of the RCP No. 2 seal consistent with Westinghouse guidance. Based on discussions from the October 27, 2015 audit between the NRC, AREVA, and the PWROG, it was agreed that No. 1 seal corrosion does not invalidate the conservatisms in the PWROG-1 401 5-P, revision 2 leakage analysis when an early RCS cooldown is performed. Specifically, an RCS cooldown as specified in Appendix D of WCAP-1 7792-P, revision 0 is acceptable to mitigate corrosion of the No. 1 seal. Based on the results of the NRC audit on No. 1 seal corrosion, the first cooldown specified in TB-I15-1 will not change from 450&deg;F within 4 hours. The second cooldown requirement for cold leg parameters will be revised to 400&deg;F and 800 psig within 24 hours. This revision allows margin to continue to rely on the Turbine Driven Auxiliary Feedwater pump for at least 24 hours following a loss of seal cooling. A temperature of 4000&deg;F also provides a favorable long term condition to prevent continued corrosion of the No. I seal.
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ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT 3.2       Early RCS Makeup In order to identify margin associated with the RCS makeup strategy, two characteristics related to RCS behavior are addressed: 1) Adequate boration capability/adequate boron mixing during two-phase natural circulation in the RCS to prevent a return to criticality, and 2) The predicted time to reflux cooling in the steam generators.
Adequate Boration Capability and Boron Mixing For an ELAP scenario initiating while in Modes 1-4, RCS boration at Catawba will begin around 11.6 hours after the start of the event and is based on preventing a potential return to criticality calculated to be 13.85 hours rather than the predicted onset of reflux cooling in the SG tubes, which occurs later.
As noted previously, after the initiation of an ELAP event, the operators will cooldown and depressurize the RCS to approximately 420&deg;F/400 psia within the first several hours in order to minimize RCP seal leakage and inventory loss. Operators will then ensure sufficient boration has been completed before continuing to cooldown further. The Catawba high pressure FLEX makeup pump has sufficient performance (40 gpm at 1700 psig) to ensure injection flow is greater than RCP seal leak off flow at the time of alignment to the RCS.
Endorsed NEI 12-06 guidance indicates normal plant operational parameters can be assumed prior to onset of an ELAP, in lieu of the more restrictive limits of a design basis analysis. In performing the in-house RELAP5 Catawba ELAP boration evaluation, credit for parameters in their nominally expected ranges was not generally taken (i.e., more limiting assumptions were made), which provides for a qualitative margin assessment as follows:
* For this evaluation all four RCP seal packages are assumed not to leak (i.e., they seal perfectly), minimizing RCS letdown and maximizing the boron injection requirement
* Boration requirements for Catawba RCS cooldown are based on an ELAP event occurring after a 500-day EFPD reactor run (EOC), with the most limiting equilibrium Xenon characteristics
* The assumed required final RCS boron concentration after FLEX makeup pump injection is conservatively high, which increases the amount of borated water volume injection to meet shutdown requirements
* Assumed decay heat is representative of EOC
* Minimum boron concentration allowed by Technical Specifications (TS) is assumed in the Cold Leg Accumulators
* Minimum cover pressure is assumed in the Cold Leg Accumulators
* Minimum boron concentration allowed by TS is assumed in the Refueling Water Storage Tank Page 4 of 8

ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT
* An hour is subtracted from the actual time to re-criticality (and hence the response time) to ensure adequate boron mixing occurs during FLEX pump make-up
* The time to start the FLEX make-up pump is calculated based on the required boron curve at an ROS temperature of 350&deg;F; during boration activities operators would maintain the plant near 420&deg;F. This conservatively requires the FLEX make-up pump to start earlier than necessary
* The Pressurizer is assumed to only be filled to 60% level prior to requiring RCS letdown through the RV head vents. Controlling the injection pump to RCS pressure in lieu of Pressurizer level would reduce the total boration time (and delay the boration start setpoint) by allowing additional RCS injection.
The margin inherent in the boration calculation assumptions/inputs therefore shows that any return to criticality during an ELAP event would reasonably be expected to occur well beyond the maximum 13.85 hour RCS make-up setpoint in the documented Catawba FLEX response.
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ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT Time to Re flux Cooling For the latest NOTRUMP reference case, the PWROG-1 4027-P, revision 3 report dated April 2015 for 4-loop TCoId plants in Category 1, identifies that Catawba will enter reflux cooling at 15.6 hours with time to uncover the core at 43.9 hours during an ELAP event. Initiating RCS boration around 11.6 hours and no later than 13.85 hours after event initiation at Catawba therefore ensures that boration will occur with acceptable loop flow conditions.
Catawba initially performed an in-house analysis of the time to reflux cooling using the RELAP5 code to establish a setpoint for RCS boration during ELAP, using the seal leakage profile from WCAP-1 7601 -P, revision 1. Subsequent to that analysis, sensitivity cases were also run in-house to evaluate the new seal leakage rates identified in PWROG-14015-P. Margin in the calculation of the predicted time to reflux cooling in the steam generator U-tubes is qualitatively identified in these analyses as noted below:
      *For this evaluation, all four RCP seals are assumed to leak at their maximum flow rate, minimizing the time to reflux cooling in the steam generators
      *Catawba-specific ELAP mass-energy release evaluation (RELAP5) shows a predicted time to reflux cooling well beyond the Westinghouse timeframe
      *Catawba has installed new thicker restriction orifice plates in the RCP No. 1 seal leak-off lines. The thicker orifice plates ensure system integrity at the higher postulated pressures so that Catawba remains in PWROG Category 1.
     *Catawba-specific ELAP mass-energy release sensitivity cases (RELAP5) adjusted for the revised PWROG Category 1 RCP seal leak-off profile show that the predicted time to reflux cooling in the steam generator tubes is still considerably delayed as compared to the reference case
      *MPR analysis flowrates for the Catawba leak-off piping configuration (MPR site-specific models) have shown peak values less than those shown in the WCAP-17601-P, revision 1 or the PWROG-14015-P, revision 2 reference cases. Therefore, cumulative RCP seal leakage will likely be lower than identified for PWROG Category 1 plants
     *Assumed decay heat is representative of EOC Page 6 of 8

ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT 3.3      Possessing the Capability to Initiate RCS Makeup within "X"Hours (Shorter than Planned Time)
RCS makeup during an ELAP event is a prioritized action per ECA-0.0, and relies on diesel-driven injection pumps that don't require FLEX electrical distribution to be set up first. Catawba stores the N+I pumps in a single FLEX Building (dome) constructed to NEI 12-06 requirements to protect FLEX response capability. The maximum RCS injection time start setpoint is 13.85 hours based on DPC-1 552.08-00-0278 revision 2. Pump deployment from the dome assumes maximum debris removal and pump deployment times (i.e., 3 hours for debris removal and deployment of pump starting at 6.5 hours). Current guidance located in Catawba's FSG-05 "Initial Assessment and FLEX Equipment Staging" directs responding Operators to identify availability of FLEX resources early in the event. This serves to minimize deployment times of prioritized actions such as RCS injection. Based on the Unit 1 and 2 Flex Strategy Timing Validation Study (Letter to File CN-1612.03), a reasonable timeframe for initiating boration following an ELAP is around 11.6 hours with additional margin if deployment can be started sooner than 6.5 hours based on site conditions and available resources.
3.4       Having an Abundant Supply of B orated Coolant Onsite and/or Having a Relatively Large Capacity for Injecting Coolant Catawba has adequate onsite borated makeup capacity for at least 72 hours following the onset of an ELAP event in Modes 1-4. Catawba's FSG-08 "Alternate NC System Boration" directs responding Operators to utilize the borated inventory available in the FWST (Refueling Water Storage Tank) for RCS makeup (approximately 6 days worth if FWST is undamaged). Should the FWST be damaged by a wind-borne missile above the protective wall, further boration capability is afforded by using water in the protected Boric Acid Tanks (BATs) and/or mixing powdered boric acid with water from the Standby Nuclear Service Water Pond (Ultimate Heat Sink) in portable mixing tanks. Beyond 72 hours, equipment from the Memphis or Phoenix NSRC (i.e., mobile boration skid) is available.
Additionally, Catawba's Standby Nuclear Service Water Pond remains available as a clean (i.e.,
< 16 ppm TSS) un-borated water source, and its use is proceduralized later in the ELAP event.
Use of the 40 gpm makeup pump and the FWST/BATs provides sufficient boration to reach the reactivity objective. The supply of borated coolant onsite provides several (> 3) days of boration capacity.
3.5       Having a High Capacity and/or High Pressure RCS FLEX Makeup Pump The Catawba FLEX High Pressure Pump (diesel driven) has a capacity of 40 gpm at 1700 psig, and has a variable speed control for flow and pressure. The diesel driver provides the ability to increase injection flowrates by approximately 50% during RCS depressurization if needed without changing to a different pump.
Page 7 of 8

ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT 3.6      Having the Ability to Monitor RCS Inventory during the Event and Attempting to Implement Makeup More Rapidly If Signs of Increased Leakage Were Detected ECA-O.O "Loss of All AC Power" lists the critical instruments required to be maintained during the ELAP transient. Available instrumentation related to monitoring RCS inventory includes:
* RCS wide range pressure
* RCS wide range hot leg temperature
* Core exit thermocouples
* RVLIS
* Pressurizer level
* Neutron flux Current guidance located in ECA-0.0 "Loss of All AC Power" and FSG-8 "Alternate NC System Boration" instructs responding Operators to prioritize RCS injection and respond more quickly if high RCS leakage is suspected.
3.7     NSAL-15-2 Leak-Off Line Break As noted previously, Catawba has evaluated the leak-off piping/components for a transient pressure spike at the No. 1 seal exit up to 2500 psia per the suggestion of the NRC to ensure system integrity is maintained. Results show the RCP No.1 seal leak-off piping remains adequately protected following installation of the thicker flow measurement orifice plates. As such, the published PWROG Category 1 ELAP leak-off rates still apply for Catawba.
3.8     Additional Considerations There is additional qualitative margin associated with the PWROG analyses performed in support of this issue. The following items were specifically noted by the NRC in the Watts Bar Mitigating Strategies Safety Evaluation dated March 27, 2015:
* The PWROG's generic ITCHSEAL calculations contain known conservatisms such as the comparison of the results of the generic analysis to the Montereau test data and in the application of the generic leak-off line configuration assumptions for each maximum leakage analysis category to individual plants' leak-off lines.
* Although entry into reflux cooling is undesirable and has not been fully analyzed in the context of an ELAP event, the use of this threshold as an acceptance criterion provides significant margin to uncover and severely damage the core.
Page 8 of 8

"1
   ~ENERGY%
Kelvin Henderson Vice President Catawba Nuclear Station Duke Energy CN01VP I 4800 Concord Road York, SC 29745 o: 803.701.4251 f: 803.701.3221 10 CER 50.4 10 CFR 2.202(b)
CNS-1 6-005 February 15, 2016 Attention: Document Controi Desk U. S. Nuclear Regulatory Commission Washington, D. C. 20555-001 Duke Energy Carolinas, LLC (Duke Energy)
Catawba Nuclear Station (CNS), Units 1 and 2 Docket Number(s) 50-413 and 50-414 Renewed License Nos. NPF-35 and NPF-52

==Subject:==
Final Notification of Full Compliance with Order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events" and with Order EA-1 2-051, "Order to Modify Licenses With Regard To Reliable Spent Fuel Pool Instrumentation" for Catawba Nuclear Station

==References:==
: 1. Nuclear Regulatory Commission (NRC) Order Number EA-1 2-049, Order Modifying Licensees With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, Revision 0, dated March 12, 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. ML12054A735
: 2. Catawba Nuclear Station (CNS) Overall Integrated Plan Submittal in Response to March 12, 2012, Commission Order to Modify Licenses With Regard To Requirements for Mitigation Strategies for Beyond Design Basis External Events (Order EA-12-049), dated February 28, 2013 (ADAMS Accession No. ML13066A173)
: 3. Catawba Nuclear Station, Units 1 and 2 - Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-1 2-049 (Mitigation Strategies), dated February 6, 2014 (ADAMS Accession No. ML13364A175)
: 4. NRC Order Number EA-12-051, Order to Modify Licenses With Regard To Reliable Spent Fuel Pool Instrumentation, dated March 12, 2012, (ADAMS Accession No. ML12054A679)

              '-4              "4 United States Nuclear Regulatory Commission Page 2 February 15, 2016
: 5. Letter from Duke Energy to NRC Site Overall Integrated Plans in Response to March 12, 2012, Commission Order Modifying Licenses With Regard To Requirements for Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051), dated February 28, 2013 (ADAMS Accession No. ML13086A095)
: 6. Catawba Nuclear Station, Units 1 and 2 - Interim Staff Evaluation and Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-1 2-051, Reliable Spent Fuel Pool Instrumentation, dated October 28, 2013 (ADAMS Accession No. ML13281A562)
: 7. Catawba Nuclear Station, Units 1 and 2, Report for the Audit Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Instrumentation to Orders EA-12-049 and EA-1 2-051, dated February 20, 2015 (ADAMS Accession No. ML15035A679)
: 8. Notification of Full Compliance with Order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events" and with Order EA-1 2-051, "Order to Modify Licenses With Regard To Reliable Spent Fuel Pool Instrumentation"  -  Catawba Nuclear Station Unit 2, dated May 1, 2015 (ADAMS Accession No. ML15126A277)
Ladies and Gentlemen On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued Order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design-Basis External Events" and Order EA-1 2-051, "Order to Modify Licenses With Regard To Reliable Spent Fuel Pool Instrumentation," (Reference 1 and Reference 4, respectively).
The Orders require holders of operating reactor licenses and construction permits issued under Title 10 of the Code of Federal Regulations Part 50 to submit for review, Overall Integrated Plans (OIPs) including descriptions of how compliance with the requirements of each Order will be achieved. By letter dated February 28, 2013 (Reference 2), the OIP for CNS in response to Order EA-12-049 was submitted. In a separate correspondence, the OIP for CNS in response to Order EA-12-051 was submitted by letter dated February 28, 2013 (Reference 5).
Order EA-12-049, Section IV.A.2 and Order EA-12-051, SeCtion IV.A.2 requires completion of full implementation to be no later than two (2) refueling cycles after submittal of the overall integrated plan, as required by Condition C.I .a or December 31, 2016, whichever comes first.
In addition, Section IV.C.3 of Orders EA-12-049 and EA-12-051 require that Licensees and CP holders report to the NRC when full compliance is achieved. For CNS Unit 1, the current requirement for full implementation of NRC Orders EA-12-049 and EA-12-051 was prior to restart from the I1EOC22 refueling outage.
On December 16, 2015, CNS Unit I entered Mode 2 (Startup) following the 1EOC22 refueling outage. As such, December 16, 2015 is the compliance date for CNS Unit I for being in full compliance with Orders EA-12-049 and EA-12-051 as demonstrated by this submittal and any other docketed correspondence concerning these Orders. This determination is based on the best available information and analyses that have been completed as of the date of this letter.

United States Nuclear Regulatory Commission Page 3 February 15, 2016 Notification of full compliance with Orders EA-12-049 and EA-12-051 for CNS Unit 2 was provided by Reference 8. provides a brief summary of the key elements associated with compliance to Orders EA-12-049 and EA-12-051 for CNS Unit 1. Attachment 2 provides the open and pending items from the Audit Report (Reference 7).. For each open and pending item identified in Attachment 2, a brief summary response in support of closure is provided. As such, Duke Energy Carolinas, Inc. (Duke Energy) considers these items complete pending NRC closure. provides all answers to the diverse and flexible strategies Interim Staff Evaluation open and confirmatory items contained in Reference 3. Attachment 4 provides all answers to the spent fuel pool (SEP) instrumentation Interim Staff Evaluation (ISE) Request For Additional Information contained in Reference 6. Attachment 5 provides the bridging document between vendor technical information and CNS specific considerations for SEP instrumentation, which compares CNS assumptions to the vendor's assumptions for the SEP instrumentation. provides the CNS Final Integrated Plan. Attachment 7 provides the CNS RCP Seal Leakage Margin Assessment.
In support of the ongoing NRC Audit process, Duke Energy will continue working with the NRC staff in the issuance of a combined Safety Evaluation (SE) for both the Mitigation Strategies and the Spent Fuel Pool Level Instrumentation Orders.
There are no regulatory commitments contained in this letter or its attachments. Please address any comments or questions regarding this matter to Cecil Fletcher at 803-701-3622.
I declare under penalty of perjury that the foregoing is true and correct. Executed on February 15, 2016.
Sincerely, Vice President, Catawba Nuclear Station Attachments:
: 1. CNS, Unit 1 Summary of Compliance Elements for Orders EA-1 2-049 and EA-1 2-051
: 2. CNS NRC Audit Report Open and Pending Items
: 3. CNS, Response to Diverse and Flexible Strategies Interim Staff Evaluation Open and Confirmatory Items
: 4. CNS, Response to Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order-i12-051, Reliable Spent Fuel Pool Instrumentation
: 5. Design Bridge Document Between Vendor Technical Information and CNS Specific Considerations for Spent Fuel Pool Instrumentation
: 6. CNS, Final Integrated Plan
: 7. CNS, RCP Seal Leakage Margin Assessment

United States Nuclear Regulatory Commission Page 4 February 15, 2016 xc:
C. Haney, Regional Administrator U. S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, GA 30303-1 257 W. M. Dean, Director, Office of Nuclear Reactor Regulation US. Nuclear Regulatory Commission One White Flint North, Mailstop 13 HIdM 11555 Rockville Pike Rockville, MD 20852-2738 J.A. Whited, Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mailstop 8 B1A 11555 Rockville Pike Rockville, MD 20852-2738 P.J. Bamford, Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mailstop 13 F15 11555 Rockville Pike Rockville, MD 20852-2738 G.A. Hutto NRC Senior Resident Catawba Nuclear Station Justin Folkwein American Nuclear Insurers 95 Glastonbury Blvd., Suite 300 Glastonbury, CT 06033-4453

ATTACHMENT 1 CNS, UNIT 1 

==SUMMARY==
OF COMPLIANCE ELEMENTS FOR ORDERS            EA-12-049 AND EA-12-051 The elements identified below for CNS Unit 1, as well as the Overall Integrated Plans (OIP) for Orders EA-1 2-049 and EA-1 2-051 (References 2 and 10, respectively), the 6-Month Status Reports for Orders EA-12-049 and EA-12-051 (References 4 thru 8 and 12 thru 16, respectively),
and any additional docketed correspondence, demonstrate compliance with Orders EA-12-049 and EA-12-051.
STRATEGIES    -  COMPLETE CNS, Unit 1 strategies are in compliance with Order EA-12-049. All strategy related Open Items, Confirmatory Items, or Audit Questions/Audit Report Open Items have been addressed and are considered complete pending NRC closure.
MODIFICATIONS     - COMPLETE The modifications required to support the FLEX strategies for CNS, Unit 1 have been fully implemented in accordance with the station design control process. The design of the Spent Fuel Pool Level Instrumentation installed at CNS, Unit I complies with the requirements specified in the order and described in NEI 12-02, Revision 1, "Industry Guidance for Compliance with NRC Order EA-12-051". The instruments have been installed in accordance with the station design control process.
EQUIPMENT   -   PROCURED AND MAINTENANCE & TESTING      - COMPLETE The equipment required to implement the Mitigation Strategies and Reliable Spent Fuel Pool Level Instrumentation has been procured and is ready for use at CNS, Unit 1.
Testing and Maintenance processes have been established through the use of Industry endorsed Electric Power Research Institute (EPRI) Guidelines and the ONS Preventative Maintenance program such that FLEX equipment reliability is achieved.
Operating and maintenance procedures for the Spent Fuel Pool Instruments for CNS, Unit 1 have been developed, and integrated with existing procedures. These procedures have been verified and are available for use in accordance with the site procedure control program. Site processes have been established to ensure the Spent Fuel Pool Instruments are maintained at their design accuracy.
PROTECTED STORAGE      - COMPLETE The storage facility required to implement the FLEX strategies for CNS, Unit 1 has been completed and provides protection from the applicable site hazards. The equipment required to implement the FLEX strategies for CNS, Unit 1 is stored in its protected configuration and is ready for use.
PROCEDURES      - COMPLETE FLEX Support Guidelines (FSG) and procedures for the maintenance and use of the Spent Fuel Pool Level Instrumentation for ONS, Unit I have been developed in accordance with NEI 12-06, Section 3.2.2 and NEI 12-02, Revision 1, Section 4.2. The FSGs and affected existing procedures have been verified and are available for use in accordance with the site procedure control program.
Page 1 of 4

ATTACHMENT 1 CNS, UNIT 1 

==SUMMARY==
OF COMPLIANCE ELEMENTS FOR ORDERS EA-12-049 AND EA-12-051 TRAINING  -  COMPLETE Training for CNS, Unit I has been completed using the CNS Systematic Approach to Training (SAT) as recommended in NEI 12-06, Revision 0, Section 11.6 and in NEI 12-02, Revision 1, Section 4.1.
STAFFING   - COMPLETE The staffing study for CNS has been completed in accordance with NEI 12-01, Revision 0 and 10OCFR5O.54(f), "Request for Information Pursuant to Title 10 of the Code of Federal Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force review of Insights from the Fukushima Dai-ichi Accident," Recommendation 9.3, dated March 12, 2012 (Reference 18), as documented in letter dated May 20, 2014 (Reference 19) and September 24, 2014 (Reference 20). The staffing study confirmed that CNS has adequate staffing to perform the actions to mitigate beyond design basis events.
NATIONAL SAFER RESPONSE CENTERS        - COMPLETE Duke Energy has established a contract with the Pooled Equipment Inventory Company (PEICo) and has joined the Strategic Alliance for FLEX Emergency Response (SAFER)
Team Equipment Committee for off-site facility coordination. It has been confirmed that PEICo is ready to support CNS with Phase 3 equipment stored in the National SAFER Response Centers in accordance with the site specific SAFER Response Plan.
VALIDATION    - COMPLETE Duke Energy has completed performance of validation in accordance with industry developed guidance to assure required tasks, manual actions and decisions for FLEX strategies are feasible and may be executed within the constraints identified in the Overall Integrated Plans (QIP) for Order EA-12-049.
FLEX PROGRAM DOCUMENT        - ESTABLISHED The FLEX Program Document for CNS has been developed in accordance with the requirements of NEI 12-06, Revision 0.
Page 2 of 4

ATTACHMENT 1 CNS, UNIT 1 

==SUMMARY==
OF COMPLIANCE ELEMENTS FOR ORDERS EA-12-049 AND EA-12-051 REFERENCES
: 1. Nuclear Regulatory Commission Order Number EA-12-049, Order Modifying Licensees With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, Revision 0, dated March 12, 2012, ADAMS Accession No. ML12054A735
: 2. Catawba Nuclear Station, Overall Integrated Plan in Response to March 12, 2012, Commission Order to Modify Licenses With Regard To Requirements for Mitigation Strategies for Beyond Design Basis External Events (Order EA-1 2-049), dated February 28, 2013, ADAMS Accession No. ML13066A173
: 3. Catawba Nuclear Station, Units 1&2 - Interim Staff Evaluation -Relating to Overall Integrated Plan in Response to Order EA-1 2-049 (Mitigation Strategies), dated February 6, 2014, ADAMS Accession No. ML13364A175.
: 4. Catawba Nuclear Station First 6-Month Status Report for the Implementation of Mitigation Strategies for Beyond Design Basis External Events (Order EA-1 2-049), dated August 28, 2013, ADAMS Accession No. ML13298A010.
: 5. Catawba Nuclear Station Second 6-Month Status Report for the Implementation of Mitigation Strategies for Beyond Design Basis External Events (Order EA-1 2-049), dated February 28 2013, ADAMS Accession No. ML14065A038.
: 6. Catawba Nuclear Station Third 6-Month Status Report for the Implementation of Mitigation Strategies for Beyond Design Basis External Events (Order EA-12-049), dated August 28, 2014, ADAMS Accession No. ML14247A232.
: 7. Catawba Nuclear Station Fourth 6-Month Status Report for the Implementation of Mitigation Strategies for Beyond Design Basis External Events (Order EA-1 2-049), dated February 26, 2015, ADAMS Accession No. ML15081A124.
: 8. Catawba Nuclear Station Fifth 6-month Status Report for the Implementation of Mitigation Strategies for Beyond Design Basis External Events (Order EA-1 2-049), dated August 26, 2015, ADAMS Accession No. ML15240A066.
: 9. NRC Order Number EA-12-051, Order to Modify Licenses With Regard To Reliable Spent Fuel Pool Instrumentation, dated March 12, 2012, ADAMS Accession No. MLI12054A679.
: 10. Duke Energy Letter, Duke Energy Carolinas, LLC, (Duke Energy) Overall Integrated Plans in Response to March 12, 2012, Commission Order Modifying Licenses With Regard To Requirements for Reliable Spent Fuel Pool Instrumentation (Order EA        051), dated February 28, 2013, ADAMS Accession No. ML13086A095.
: 11. Catawba Nuclear Station, Units 1 and 2 - Interim Staff Evaluation and Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-12-051, Reliable Spent Fuel Instrumentation, dated October 28, 2013, ADAMS Accession No. ML13281A562.
Page 3 of 4

ATTACHMENT 1 CNS, UNIT 1 

==SUMMARY==
OF COMPLIANCE ELEMENTS FOR ORDERS EA-12-049 AND EA-12-051
: 12. Catawba Nuclear Station First 6-Month SFPLI Status Report in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order EA-12-051), dated October 28, 2013, ADAMS Accession No. ML13242A009.
: 13. Catawba Nuclear Station Second 6-Month SEPLI Status Report in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order EA-12-051), dated February 26, 2014, ADAMS Accession No. ML14063A279.
: 14. Catawba Nuclear Station Third 6-Month SFPLI Status Report in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order EA-12-051 ), dated August 14, 2014, ADAMS Accession No. ML14227A717.
: 15. Catawba Nuclear Station Fourth 6-Month SFPLI Status Report in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order EA-12-051 ), dated February 16, 2014, ADAMS Accession No. ML15051A366.
: 16. Catawba Nuclear Station Fifth 6-month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order EA-12-051), dated August 3, 2015, ADAMS Accession No. MLI1521 7A007.
: 17. Catawba Nuclear Station, Units 1 and 2, Report for the Audit Regarding the Implementation of Mitigation Strategies and Reliable Spent Fuel Pool Instrumentation related to Orders EA-1 2-049 and EA-1 2-051, dated February 20, 2015, ADAMS' Accession No. ML15035A679.
: 18. 10OCFR50.54(f), "Request for Information Pursuant to Title 10 of the Code of Federal Regulations, Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force review of Insights from the Fukushima Dai-ichi Accident", Recommendation 9.3, dated March 12, 2012, ADAMS Accession No. ML12053A340.
: 19. Duke Energy Letter, Phase 1 Staffing Assessment pursuant to 10OCFR50.54(f) regarding NTTF Recommendations 9.3, dated April 30, 2013.
: 20. Catawba Nuclear Station Phase 2 Staffing Assessment pursuant to 10OCFR 50.54(f) regarding NTTF Recommendation 9.3, dated October 28, 2014, ADAMS Accession No. MLI 4303A259.
: 21. NEI 12-06, Revision 0 "Diverse and Flexible Coping Strategies (FLEX) Implementation Guide", ADAMS Accession No. ML12242A378.
: 22. NEI 12-02, Revision 1 "Industry Guidance for Compliance with NRC Order EA-1 2-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", ADAMS Accession No. ML12240A307.
: 23. NEI 12-01, Revision 0 "Guideline for Assessing Beyond Design Basis Accident Response Staffing and Communications Capabilities", ADAMS Accession No. ML12125A412.
Page 4 of 4

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS Duke Energy affirms that CNS is in full compliance with Orders EA-12-049 and EA-12-051 as demonstrated by the docketed correspondences concerning these orders. Briefly, CNS FLEX Interim Staff Evaluation (ISE) Open and Confirmatory Items are complete pending NRC closure; CNS FLEX OIP Open Items are complete pending NRC Closure; CNS FLEX Audit Questions are complete pending NRC closure; CNS FLEX NRC Audit Report Open Items are complete pending NRC closure; and the CNS Request for Additional Information (RAl) provided in the Spent Fuel Pool Level Instrumentation (SFPLI) ISE are complete pending NRC closure.
Duke Energy provides the following response for the Audit Report Open and Pending Items and considers them to be complete pending NRC closure for Catawba Nuclear Station:
Iitem      .. -1,         -,   DescriptionII         ,*'...           Sum mary Response,*.,
Procedure Interfaces   - Seismic           Catawba Response:
Confirm completion of evaluation          RfrneAtcmn               o h of potential Aux Building flooding         repn.
Cl 3.1.1.3.A             and appropriate actions andrepn.
procurement of sump pumps.
Room Temperature Analyses                  C-ab        epne Auxiliary Building, SFP Building           Reference Attachment 3 for the and Control Room response.
Room temperature analyses being CI 3.2.4.1 .A            performed will provide a better idea of the environmental conditions expected during the event. Confirm completion of analyses and appropriate actions.
Freee PotecionCatawba Response:
Evaluations to address the needs           Reference Attachment 3 for the Cl 3.2.4.3.A             for freeze protection are in               response.
progress. Confirm completion of evaluations and appropriate
____     ____    ___   actions.
LighingAnalsesCatawba Response:
Cl 3.2.4.4.A             Confirm evaluations for additional         Reference Attachment 3 for the lighting have been completed and           response.
appropriate actions taken.
Offsite Resources   - Confirm NEI          CtabRepn:
12-06 Section 12.2, Guidelines 2           RfrneAtcmn               o h through 10 are addressed with Cl 3.4.A              SAFER                                      response.
Complete SAFER Site specific
_____      _____    plan.                                     _  _  _ _  _ _ _ _ _ _  _  _ _  _ _
Page 1 of 10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS Catawba Response:
Westinghouse Electric Company submitted letter LTR-RES-1 3-1 53
("Documentation of 7228C 0-Rings at ELAP Conditions") on October 31, 2013. The letter documents a Westinghouse evaluation of compound 7228C RCP 0-rings at ELAP conditions up to 582&deg;F (the same 0-rings in use at Catawba),
and concludes that they will not fail Reactor Coolant Pump 0-rings during an 8-hour ELAP event w/o seal cooling. The 0-rings survived Confirm that, beyond order for an average of 18 hours, with the compliance date, plant will use first failure occurring at 13 AQ.26d             only high-temperature-qualified 0-hours. The Catawba FLEX strategy rings where applicable, or that for RCS make-up post-ELAP is to steam generator relief valve will be cooldown well below 582&deg;F and operated to control temperature to align a FLEX portable RCS make 550 degrees F or below.
up pump within about 12 hours. Therefore, the Catawba RCP 0-ring integrity will be maintained and the assumption of a 21 gpm/pump leak rate remains valid.
Reactor coolant pump seal drawing has been posted to the Catawba Fukushima Sharepoint noting high temperature o-rinp material.
Direct Current (dc) load profile,    CawbRepn:
load shedding, and dc bus voltage. Catawba understands the NRC will AQ.37 The staff will complete a vendor     be auditing the battery vendor to audit of the batteries.              close this item.
Catawba Response:
FLEX Diesel Generator Sizing         EniergChgevauto AQ.47             Sizing calculations need to be       EC401541, FLEX Diesel Generator completed and placed on Catawba      Loading Evaluation has been Fukushima Sharepoint.                completed and placed on the Catawba Fukushima Sharepoint.
Battery Room Ventilation   -         Catawba Response:
TempeatureCatawba                     has purchased sufficient Provide information on the           fans and spot coolers to perform AQ.49             adequacy of the ventilation          Phase 1, 2, and 3 strategies, provided in the battery room to      provide sufficient cooling for protect the batteries from the       equipment operation/personnel
_____________effects         of extreme high and low      habitability, and Hydrogen gas Page 2 of 10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS temnperatu res.                        control based on a GOTHIC analysis performed by Zachry (Reference CNC -1211.00-00-01 46
                                                     - Gothic Analysis For Extended Loss Of All AC Power (ELAPIFLEX)). Procedural guidance has been developed to implement the mitigation recommendations in the Zachry analyses. In addition, per Table 7-1 in the SAFER Response Plan and Table 9-1 in the National SAFER Response Center Equipment Technical Requirements Document (51-9199717-013), Catawba will receive one 3000 cfm ventilation fan and associated ducting per Unit from the NSRC. This ventilation equipment from the NSRC will be used on an as needed/as desired basis to provide additional air flow to any area in the plant.
Calculation CNC-1 2t1.00-00-0146 Gothic Analysis For Extended Loss Of All AC Power (ELAP/FLEX) has been placed on the Catawba Fukushima Sharepoint for review.
Catawba Response:l Diesel Fuel Oil Supply and Quality     Fe     i   osmtosoe Describe plans for supplying fuel      volume, quality, and the need for oil to FLEX equipment (i.e., fuel oil  refueling have been addressed in AQ.50       storage tank volume, supply            calculation CNC-1612.03-00-0001, pathway, etc.). Also, explain how      FLEX Fuel Consumption fuel quality will be assured if stored Calculation. This calculation has for extended periods of time.          been posted to the Catawba Fukushima Sharepoint.
Catawba Response:
Battery Room Ventilation    -          Catawba has purchased sufficient Hydrogen Accumulation Potential        fans and spot coolers to perform Provide a discussion of battery        Phase 1, 2, and 3 strategies, room ventilation to prevent            provide sufficient cooling for AQ.51       hydrogen accumulation while            equipment operation/personnel recargnghebattere in phase 2      habitability, and Hydrogen gas th aeiescontrol rechrgig                             based on a GOTHIC or 3. In your response, include a description of the exhaust path if it  analysis performed by Zachry is different from the design basis.    (Reference CNC -1211.00-00-0146
                                                     - Gothic Analysis For Extended
___________________________________Loss Of All AC Power Page 3 ofl10

ATTACHMENT 2 CNS NRC   AUDIT REPORT OPEN AND PENDING ITEMS (ELAP/FLEX). Procedural guidance has been developed to implement the mitigation recommendations in the Zachry analyses. In addition, per Table 7-1 in the SAFER Response Plan and Table 9-1 in the National SAFER Response Center Equipment Technical Requirements Document (51-9199717-013), Catawba will receive one 3000 cfm ventilation fan and associated ducting per Unit from the NSRC. This ventilation equipment from the NSRC will be used on an as needed/as desired basis to provide additional air flow to any area in the plant.
Calculation CNC-121 1.00-00-0146, Gothic Analysis For Extended Loss Of All AC Power (ELAP/FLEX), has been placed on the Catawba Fukushima Sharepoint for review.
Catawba Response:
The following flow model calculations have been posted to the Catawba Fukushima Sharepoint; CNC-1223.02-00-0025, Cooling Water Flow Model              Flow Model of SNSWP to CA Develop flow model calculations to    Connections For Phase 2 FLEX Strategies, CNC-1 223.02-00-0026, support the various FLEX OIP.32                                       Flow Model of SNSWP to RN strategies and document the Connections and CA for Phase 2 available static water volume in the FLEX Strategies, CNC-1223.02        RN/CA piping.
0027, Flow Model of RN to KC Hxs to Support RHR For Phase 3 FLEX Strategies, and CNC- 1223.02                                             0028, Flow Model For UI/U2 NI Portable Pump Injection to RCS Phase 2 & 3 FLEX Strateaies.
Catawba Response:
Determine Need for Containment Spray                                Calculation DPC-1 552.08-00-0280, Extended Loss of AC Power An analysis is needed to determine   (ELAP)    - Ice Condenser OIP.42 if containment spray for             Containment Response with FLEX temperature/pressure control is not  Mitigation Strategies, has required over the long term.         determined containment spray is not needed for long term temperature/pressure control.
Page 4 of 10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS Calculation has been posted to the Catawba Fukushima Sharepoint.
Catawba Response:
Engineering change EC1 09414 and EC1 10934 were completed during SFP Insrumnts1 eve                      EOC22. EC109414 installed the OIP.43            Provide redundant SEP level                   primary spent fuel pool level instruments                                  indication and EC1 10934 installed the backup spent fuel pool level indication.
Catawba Response:
Staffing                                     Phase 2 staffing assessment OIP.56            NRC will review the Phase 2                  provided October 28, 2014.
Staffing Assessment.                          Reference ADAMS Accession No. ML14303A259.
SFP Level Instrument Mounting                Catawba Response:
SEPI       Complete the "Intermediate SFI4Waveguide Mounting Support"                         Reference Attachment 5, Item 13.
calculation and place on the Catawba Fukushima Sharepoint.
SEP Level Instrument Mounting                CawbRepn:
Complete the "Intermediate                   Calculation CNC-1 139.14-08-0001, SFPI.6            Waveguide Mounting Support"                  Waveguide Antenna Support, has calculation and place on the                 been added to the Catawba Catawba Fukushima Sharepoint.                Fukushima Sharepoint SEP Level Instrument Reliability             Catawba Response:
Complete the following and place             Reference Attachment 5, Items 14, on the Catawba Fukushima                     20 and 22.
Sharepoint for review;
                        -The      justification for the shock and vibration test           PRCD         E deviation
                        -An     assessment of potential        The calibration procedure for the SFPI.7                      susceptibilities of                Spent Fuel Pool Level Back Up electromagnetic\radio              instrument (I P/11A/13120/031, frqenyinterference              Backup Spent Fuel Pool Level) has frEqueF)ncy es                  hr been completed and a copy has theMISFP insarueas wher            been posted on Catawba thekuSFPminstrumentnis located and how to mitigateFushmSaepit those susceptibilities
                        -The      calibration procedure        ARE VA Document # 5 1-9202556-
______________for               the Back-up SFP Level          005, Qualification Analysis of Page 5 ofl10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS Monitoring differential    VEGAPULS 62 ER Through Air pressure transmitter        Radar, has been posted to the Catawba Fukushima Sharepoint as well.
SEP Level Instrument Qualification   Catawba Response:
Complete the calibration procedure   IPI11A131201031, Backup Spent for heSP ackup LvelFuel           Pool Level has been SFPI.8    Monitoring differential pressure    completed and posted to the transmitter and place on Catawba     Catawba Fukushima Sharepoint.
Fukushima Sharepoint for review.
Catawba Response:
SFP Level Instrument Calibration Complete the calibration procedure   IPI11AI3120/031, Backup Spent SFPI.1 1  for the Back-up SEP Level            Fuel Pool Level has been Monitoring differential pressure    Completed and posted to the transmitter and place on Catawba    Catawba Fukushima Sharepoint.
Fukushima Sharepoint for review.
SEP Level Instrument Calibration    Catawba Response:
Complete the calibration procedure SFPI.12   for the Back-up SFP Level            IP/1/A/3120/031, Backup Spent Monitoring differential pressure     Fuel Pool Level has been transmitter and place on Catawba    completed and posted to the Fukushima Sharepoint for review. Catawba Fukushima Sharepoint.
SEP Level Instrument                Catawba Response:
Maintenance and Test IP/11A/3120/031, Backup Spent SFPI.15   Complete the calibration procedure  Fuel Pool Level has been for the Back-up SEP Level            completed and posted to the Monitoring differential pressure    Catawba Fukushima Sharepoint.
transmitter and place on Catawba Fukushima Sharepoint for review.
RCP Seal LeakageCawbRepn:
This is an ongoing generic NRC Please provide adequate             issue related to the ELAP RCP seal justification for the seal leakage   leakage issue and the RCP seal rates calculated according to the    model used in evaluating the LOSC SE.3    Westinghouse seal leakage model      response. To address this item, the that was revised following the       NRC requested sites using issuance of NSAL-14-1 or             standard RCP seal packages to alternative model (e.g., MPR). The   provide a RCP Seal Leakage justification should include a       Margin Assessment paper. This is discussion of the following factors: included as Attachment 7.
Page 6 of 10

.1 ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS a) benchmarking of the seal        Reasonable assurance of leakage model against        compliance with endorsed guidance relevant data from tests or  is achieved via in-house operating events,            evaluations confirming Catawba's FLEX strategies are bounded by b) discussion of the impact on     the WCAP-17601-P, revision 1 the seal leakage rate due     reference case as well subsequent to fluid temperatures         PWROG evaluations. As such, greater than 5500&deg;F           closure of this issue is not a resulting in increased       requirement for Unit startup.
deflection at the seal        Calculation CNC-1223.04-00-01 17, interface,                   MPR RCP Seal Leakoff ELAP/FLEX Analysis has been c) clarification whether the       placed on the Catawba Fukushima second-stage reactor          Sharepoint as supporting coolant pump seal would       documentation.
remain closed under ELAP conditions predicted by the   Duke Fleet Fukushima revised seal leakage model    Response/PWROG continue to and a technical basis to     work with the NRC to close this support the determination,    generic issue.
and, d) justification that the interpolation scheme used to compute the integrated leakage from the reactor coolant pump seals from a limited numb~er of computer simulations (e.g., three) is realistic or conservative.
Catawba Response:
This is an ongoing generic NRC issue related to the ELAP RCP seal leakage issue and the potential RCP Seal Leakoff Piping               rupture of the #1 seal leak-off line.
The current NRC position is that the Perform additional piping and leak-off piping should maintain support analysis to ensure the seal leakoff line temperature and          integrity up to 2500 psia. To SE.4                                       address this item, the NRC pressure seen during the ELAP event will be supported by the        requested sites using standard RCP seal packages to provide a current configuration.
RCP Seal Leakage Margin Assessment paper. This has been included as Attachment 7.
Reasonable assurance of compliance with endorsed auidance Page 7 of 10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS is achieved via in-house evaluations confirming Catawba's FLEX strategies are bounded by the WCAP-1 7601 -P, revision 1 reference case as well subsequent PWROG evaluations. As such, closure of this issue is not a requirement for Unit startup.
Calculation CNC-1223.04-00-01 17, MPR RCP Seal Leakoff ELAP/FLEX Analysis has been placed on the Catawba Fukushima Sharepoint as supporting documentation.
Duke Fleet Fukushima Response/PWROG continue to work with the NRC to close this generic issue.
t                                     +/-
ELAP Calculations with                 Catawba Response:
NOTRUMP Please provide adequate basis         The PWR Owners Group has that calculations performed with       issued report PWROG-14064, the NOTRUMP code (e.g., those in      "Application of NOTRUMP Code WCAP-1 7601 -P, WCAP-1 7792-P)         Results for Westinghouse are adequate to demonstrate that       Designed PWRs in Extended Loss criteria associated with the          of AC Power Circumstances". This analysis of an ELAP event (e.g.,      report compares E LA P avoidance of reflux cooling,          predictions from the NOTRUMP promotion of boric acid mixing) are   and TRACE computer codes. The satisfied. NRC staff confirmatory     results show that the NOTRUMP analysis suggests that the need for   predicted results for the onset implementing certain mitigating       of reflux cooling and the SE.5   strategies for providing core         loop flow under two-phase cooling and adequate shutdown         conditions agree well or are margin may occur sooner than          conservative with respect to the predicted in NOTRUMP                  TRACE predicted results. The simulations.                          comparison shows that NOTRUMP provides a conservative estimate of the required time when the primary make-up pumps are required for an ELAP event as compared to TRACE. Therefore, it is concluded that NOTRUMP is acceptable for simulation of the ELAP event within the constraints listed herein with regards to reflux coolinq and boron mixinq.
Page 8 of 10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS Duke has also performed independent predictions of the ELAP event for both Catawba units using RELAP, and the results are in good agreement with the NOTRUMP results in WCAP-17601-P.
Catawba Response:
The generic 4-loop NSSS model used in WCAP-1 7601 is representative of Catawba Unit I RCS Cooling & RCS Inventory          on all parameters. Each parameter Control Analysis                     of interest in LTR-LIS-14-219 in Table 1C has been reviewed and Justification is needed to show that  no additional assessment is the WCAP 17601-P analysis used        required. As such, the time to for CNS is representative and         reflux cooling provided in PWROG-SE.8 bounding of the actual plant          14027-P is applicable.
conditions for the ELAP event. In    Catawba ELAP parameters table particular the time to reach reflux  has been added to the Catawba cooling needs to be justified.       Fukushima Sharepoint. The table contains Westinghouse analysis values and the corresponding Catawba Unit I parameters.
Supporting calculations have been added to the Catawba Fukushima Sharepoint as well.
Catawba Response:
RCS Cooldown and Prevention of Nitrogen Injection for Safety        CNS uses setpoint 0.08 (setpoint Injection Accumulators               0. 11 was calculated for CNS, but resulted in the same value, Specify whether CNS is using EOP     therefore no benefit in using the Setpoint Number 0-8 or 0-11 for       substitute).
the cooldown. Also, the licensee needs to specify whether the plan     FSG-10 (Cold Leg Accumulator for isolation of the Cold Leg          Isolation) is invoked as part of SE.9 Accumulators occurs prior to           EP/I/A/5000/ECA-0.0, Loss of All cooling down to Emergency              AC Power, and isolates CLAs prior Operating Procedure setpoint 0-12      to reaching EP setpoint or 0-13. Provide the footnote          O.13. Cooldown will be initiated calculations for whichever             only when the CLAs are isolated or setpoints are being used on the        vented. ECA-0.0 has been added Catawba Fukushima Sharepoint           to the Catawba Fukushima for review.                            Sharepoint. CNS uses footnotes 0.08 and 0.13 for target SG pressures durino ELAP recovery.
Page 9 ofl10

ATTACHMENT 2 CNS NRC AUDIT REPORT OPEN AND PENDING ITEMS Excerpts from the Catawba Unit 1&2 EP setpoint calculation CNC-1552.08-00-0195, Emergency Procedure Setpoints, has been added to the Catawba Fukushima Sharepoint. Calculations for F.05 and 0.07 were included as they provide input into the evaluation of Footnote 0.08.
Page 10 of 10

ATTACHMENT 3 CNS RESPONSE TO DIVERSE AND FLEXIBLE STRATEGIES INTERIM STAFF EVALUATION OPEN AND CONFIRMATORY ITEMS Duke Energy provides the following response to the Interim Staff Evaluation (ISE) open and confirmatory items contained in NRC Letter, "CNS - Interim Staff Evaluation Relating to the Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies), (Agency-wide Documents Access and Management System (ADAMS) Accession No. ML13364A175).
ISE CI                                                                  Catawba Response:
3.1.1.3.A                                                                 Sump pumps have been procured and strategies Procedure Interfaces  -  Seismic                            developed to address internal flooding issues.
Calculation CNC-1 206.03-00-0001, Flood Levels for Confirm completion of evaluation of potential Aux          Structures Outside of the Reactor Building, has been Building flooding and appropriate actions and               completed and posted to the Catawba Fukushima procurement of sump pumps.                                  Sharepoint. FLEX Support Guidelines FG/OIAICFLX/FSG-20 (Electrical Distribution) and FGIO/AICFLX/FSG-22 (Sump Pump Operation) provide electrical setup and placement of sump pumps. Both are located on the Catawba Fukushima Sharepoint.
ISE Cl                                                                  Catawba Response:
3.2.4.1 .A                                                                Catawba has purchased sufficient fans and spot coolers to perform phase 1, 2, and 3 strategies, provide sufficient cooling for equipment operation/personnel Room Temperature Analyses      - Auxiliary Building, SFP    habitability, and hydrogen gas control based on Gothic Building and Control Room                                   analysis performed by Zachry (Reference CNC-121 1.00-00-0146 - Gothic Analysis for Extended Loss Of All AC Room temperature analyses being performed will              Power (ELAP/FLEX)). Procedural guidance has been provide a better idea of the environmental conditions       developed to implement the mitigation recommendations expected during the event. Confirm completion of            in the Zachry analyses. In addition, per Table 7-1 in the analyses and appropriate actions.                           National SAFER Response Center Equipment Technical Requirements Document (51-9199717-013), Catawba will receive one 3000 cfm ventilation fan and associated ducting per Unit from the NSRC. This ventilation
_________    _____________________________________equipment from the NSRC will be used on as needed/as Page 1 of 3

ATTACHMENT 3 CNS RESPONSE TO DIVERSE AND FLEXIBLE STRATEGIES INTERIM STAFF EVALUATION OPEN AND CONFIRMATORY ITEMS OpenIltem "-           -*!        DesCription -*:'*"'*      ..*"::,-;i:      ::     :-    .'   Response * :-&deg;"          ,   ,,   iil desired basis to provide additional air flow to any area in the plant.
ISE CI                                                               Catawba Response:
3.2.4.3.A Other than components associated with monitoring FWST level, there is no other Flex related instrumentation located in the yard. The FWST level instrumentation is exposed and is normally freeze protected. As noted below, heat tracing equipment is kept in the Flex storage building and can be used to keep the FWST level instrumentation functional in extreme freezing conditions as directed by the FSGs.
Some credited instrumentation components are located in the Doghouses, but they will not require freeze Freeze Protection                                           protection. During cold weather conditions the window Evaluations to address the needs for freeze protection      setosfthDghuesaecvrdbcrai.
are n pogrss.Confrm ompetin ofevauatonsandThis,             combined with heat by steam piping used in the actonfir cmltoofeauinsnd Phase I and 2 response, eliminates freezing of areinprogriaess instruments or impact to supporting equipment. Flex piping connections were reviewed and none were found to be affected by outside freezing conditions. Auxiliary Feedwater and Steam Generator Wet Layup (BW) connections in the Doghouse are "protected" in the same manner as the instrumentation discussed above.
Most Flex piping connections are located in the Auxiliary Building where freezing is not credible. One train of RN connection is located outside, but within a below grade bunker. Inventory used from the FWST enters the plant through the normal system piping. Based on initial conditions and the design of the piping system, freezing Page 2 of 3

ATTACHMENT 3 CNS RESPONSE TO DIVERSE AND FLEXIBLE STRATEGIES INTERIM STAFF EVALUATION OPEN AND CONFIRMATORY ITEMS Openn Item tm                        Description !,J**,!i*    *,* *ii***,***:     i.*:i , ,* /,*:: * *'** Responseecipin es o s s?",*i' /,:*:,*,*:**.*  *!
is not projected to be a problem. Only the piping in the trench leaving the FWST is considered for use in the FLEX response. This piping along with the FWST is judged to be weather protected to the extent that freezing will not occur before FLEX strategies are implemented. The Fuel Building will remain above freezing due to the constant heat source of the spent fuel in the pool.
As a contingency, Catawba has a roll of heating cable stored in the protected storage building. This self-regulating heat trace cable is provided along with the accessories for assembling multiple freeze protection circuits. These heating cables could be powered from any portable DIG supplying 120 VAC power. This equipment could be deployed if an unanticipated freeze protection need developed.
ISE Cl    Lighting Analyses                                              Catawba Response:
3.2.4.4.A Confirm evaluations for additional lighting have been          Lighting assessment has been completed and is completed and appropriate actions taken,                      available on the Catawba Fukushima Sharepoint for review.
ISE CI    Offsite Resources - Confirm NEI 12-06 Section 12.2,            Catawba Response:
3.4.A Guidelines 2 through 10 ar'e addressed with SAFER Complete SAFER Site specific plan.                            Approved CNS SAFER Plan, CNSRP-1612.03-01, has been posted to the Catawba Fukushima Sharepoint for review.
Page 3 of 3

ATTACHMENT 4 CNS   RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING THE OVERALL INTEGRATED PLAN FOR IMPLEMENTATION OF ORDER EA-12-051, RELIABLE SPENT FUEL POOL INSTRUMENTATION The CNS Response to the Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-12-051, Reliable Spent Fuel Pool Instrumentation was provided initially to the NRC in Duke Letter, Duke Energy Carolinas, LLC, (Duke Energy),
Response to Request for Additional Information Regarding Overall Integrated Plan in Response to order EA-12-051, "Reliable Spent Fuel Pool Instrumentation" dated July 23, 2013, (ADAMS Accession No. ML13206A384).
The NRC provided an Interim Staff Evaluation (ISE) and Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-12-051, Reliable Spent Fuel Instrumentation dated October 28, 2013 (ADAMS Accession No. ML13281A562). The Requests For Additional Information (RAls) were renumbered by the ISE.
The ONS Response to this Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-12-051, Reliable Spent Fuel Instrumentation was provided through the Scientech eDocs Web Portal as allowed by NRC Letter, Nuclear Regulatory Commission Audits of Licensee Responses to Reliable Spent Fuel Pool Instrumentation Order EA-12-051, dated March 26, 2014 (ADAMS Accession No. ML14083A620).
No further information is required for this attachment.

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION 7TicParameter:                          VedrDein                    Aditommnal    e
                                   .. ,     ,st                    ReferenceTo                         Te, or Analysis Results      -=   ,,         Licensee Evaluation-<         ::
Duke Technical CutmrRequirements                                                                       The vendor instrumentation design was Design                             technicalnt           DouetDR-                    E-20                     N/A               reviewed and determined to adequately meet requirementsn              1336.04-00-0001              NE1-1 2-02 specfictio 02Duk POthefo            Rev                                         specification requirements.
SFPLI 171968 Rev, 003 Qualification is based on a combination of tests and analyses or                                                                                          The vendor qualification documentation was siiart.EA-12-051,                                      Test and analysis results    reviewed and concluded to adequately
:2   Test Strategy                                         Qualification analysis              1.4      meet requirements of EA      demonstrate the instrumentation could reliably Qulfcain          Doc. 51-9202556-005            NE1-12-02,   12-051, JLD-ISG-2012-03,      function in its installed environment(s) during a testsfiatind                                                   3.4        and NEI 12-02 Rev. 1       postulated Beyond Design Basis External analyss are                                                                                          Event (BDBEE).
summarized in qualification analysis report 51-9202556-005.
Tempratre rtin of      The primary channel instrumentation electronics are located outside the SEP area.
Power Control Panel is      Thvedrisumnaoneig qualiicaton Qulifiatio fr                  Anaysis     EI1-02,   in te pael aove mbieT.hxped           venvior nsrmentaltcondiin    d   uringna 3 Electronicsta              Temperature and9F                                             34                               rse     emDoctre n51-92ityl2556-oun enclosureeetoisulfcin with fo ~     meauehumidity an         ulfcto o.005, 5-256-Section Anlss2.3                    NEMA41-0, n4XeNEA       enclosuree ncosreamie    postulatedetepotuatdeviBDBEE.fo BBE.i     Seein eeSection fecio    o this ds7liiof  hi display                                                                                          prevents moisture intrusion,. ouetfrdsuso                 fds iis Radiation withstand       Refer to RAI #7 response posted to Catawba analyzed to 1x1 03 rads     Fukushima Sharepoint.
Page 1 ofl1l

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND                                  CRS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION Measurement capability through saturated steam and smoke.                                               Initial testing (without horn Testing                                                  cover) demonstrated performed to                                             successful measurement demonstrate the                                          capability through steam radar horn cover                                         and smoke. Subsequent was effective at                                         testing of the radar horn and The final radar horn cover qualification report preventing                                               cover demonstrated            was not available when Catawba PAl moisture                                                 adequate operation during     responses were submitted. Areva Report No.
intrusion within    Qualification Analysis               sustained simulated SEP       51-9221032-000 is the "Qualification Analysis Evrnetl               the horn and                 Doc.                        boiling conditions, and that  For Vega Waveguide Horn Cover". The radar tsigfor level           wave guide pipe.       51-9202556-005,                   the horn cover was effective  horn cover qualification testing adequately tstingo                                           Section 2.3, 2.4, 2.5,   EA-12-051, in preventing moisture        demonstrated acceptable operation during sesrRadar                     horn         2.7, Appendix B and          1.4    intrusion within the horn and exposure to simulated SEP boiling conditions.
4opnet            n         cover (fused        supporting references     NEI 12-02, wave guide pipe.
SE ra-                  silic gls)              690860234The                                                        horn cover adhesive manufacturer Saturated steam                   cagas,6-204-234 metal  waveguide       51-9220845-001                    The horn cover adhesive is    radiation test data adequately demonstrated
        &Rdain                pipe and horn          51-9221032-000                    a silicone elastomer          the adhesive would not experience are not                66-9225632-000                    manufactured by Dow           unacceptable degradation for radiation susceptible to                                           Corning (Sylgard 170). The    exposure in excess of that expected for the radiation,                                               adhesive manufacturer         postulated beyond design basis event over the degradation.                                             radiation test data           required mission time.
Manufacturer                                             adequately demonstrates test data                                                the adhesive would not supports                                                 experience unacceptable acceptable                                               degradation for exposures radiation                                                up to 1.64 Xl 0* Rads.
degradation resistance for the radar horn cover
______________adhesive.                    ______________________________
Page 2 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND                             CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION Qualification Analysis Doc. 51-9202556-005, Sections 2.3, 2.5, Appendix A and                    Sno         n    niao          r tsigfrlvl         Temperature and   spotnrerncsdemonstrated                                  to withstand senvionmna                                   3E8-6008-21-3000          E20,      the      manufacturer ratings       Refer to RAI # 7 response posted to Catawba 5 electronics                and analysis of     EN 60529:2000,             3.4   ' 80&deg;C (sensor) and            70&deg;C  Fukushima Sharepoint. See Section 7 of this huig-sensor                 and                                           (indicator),    100%          RH. document for discussion of dose limits.
houtsing SF-niao                             38-9218214-000,Raitowthan otie FPidctr                USNRC Bulletin 79-                     analzdia tion         wxithsr a nds 01B Table C-i,anledtlx0rds NUREG-1 73, Vol.1, Section 3.11.3.2.1, Reg. Guide 1.209 Thermal and radiation aging Radar horn                                                not applicable to metal cover (fused                                             waveguide in SFP area.
silica glass),                                           The horn cover adhesive is         The glass and metallic instrumentation metal waveguide                                           a silicone elastomer              components located within the SEP area are pipe and horn                                             manufactured by Dow               not susceptible to aging due to thermal and/or are not                                                   Corning (Sylgard 170). The        radiation effects.
Thermal &                susceptible to    Qualification analysis     EA1-5,       radiation test data               The horn cover adhesive manufacturer Radiation Aging         rdainDoc.                 51 -9202556-           14       adequately demonstrates           radiation test data adequately demonstrated cmoneants              dgatinn              051-Scto22102-0         NEI 12-02,  the adhesive would not             the adhesive would not experience cmoetin          Horn cover           5192020034                          experience unacceptable            unacceptable degradation for radiation SEP area                 adhesive             66-9225632-000              34      degradation for exposures          exposure in excess of that expected for the manufacturer                                              up to 1.64 xl 0 Rads. The         postulated beyond design basis event over the radiation test                                            silicone adhesive is rated to     required mission time. The horn cover data and                                                 withstand temperatures             adhesive temperature ratings are acceptable temperature                                              extremes of -45 to 200&deg;C,          and readily bound the expected conditions for withstand                                                which adequately bound the         the postulated beyond design basis event.
specifications,                                           postulated temperatures for
                                                 ~sustained                                         SEP boiling conditions.                        _____________________
7    Basis for Dose          SFPLI remote      AREVA Document                         A Catawba calculation              A location specific dose calculation was Requirement             transmitter,      No. 51-9202556-005,        NEI 12-02,  shows that the electronics         performed for the remote electronics, which indicator and     Qualification Analysis         3.4      require periodic                  indicated these components must be replaced
___________power                  control   of VEGAPULS 62 ER                      replacement.                       in less than 6 years. A periodic replacement Page 3 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION
         # I   "' I Topic    Parameter TpcSummary                I      ~~~Reference Vendor iDocument      Design
                                                                #        Additional CommTests Coment eutis Tes orAayssRsut.iene.vauto o nay r____________________________      iese        auto panel                                                                                             for the transmitter, power control panel and lxl10 3 qualified rads      to      Through Air Radar 51-9221032-000                                                          local indicator has been set up. Site specific The horn cover adhesive based on                  66-9225632-000                          manufacturer radiation test     radiation dose data is being collected for a industry                                                          data adequately                 future revision of the dose calculation. Refer operating                                                         demonstrates the adhesive       to RAI #7 response posted to Catawba experience.                                                       would not experience            Fukushima Sharepoint.
unacceptable degradation Based on                                                          for exposures up to 1.64        The horn cover adhesive manufacturer engineering                                                       xl0 8 Rads.                     radiation test data adequately demonstrated judgment, the                                                                                     the adhesive would not experience expected total                                                                                    unacceptable degradation for radiation integrated dose                                                                                   exposure in excess of that expected for the for the radar                                                                                     postulated beyond design basis event over the horn cover                                                                                        required mission time.
adhesive would not exceed 1 x108 over the required mission time for the instrumentation.
                                         -t                                                                         +
Seismic withstand             Qualification analysis Doc. 51-9202556-capability of                                                  VEGAPULS 62 ER sensor,         The vendor instrumentation seismic testing 005, Section 2.1,                       indicator, power control     adequately demonstrates the equipment is VEGAPULS 62 Appendix D and                         panel, mounting brackets,     capable of reliably operating during a seismic Seismic                ER sensor,                                       NEI 12-02, supporting references                      and waveguide pipe are       event.
Qualification       indicators, power                                         3.4 11-9203036-000,                       seismically qualified to RRS control panel, IEEE STD 344-2004,                          levels from EPRI TR-       Refer to RAIs #7 and #8 responses posted to mounting EPRI TR-1 07330,                                  107330             Catawba Fukushima Sharepoint.
brackets, 174-9213558-006 waveQuide pipe NRC RAls                  Sloshing analysis                       Seismic induced sloshing indicated a SEP           was performed by an             NA      analysis concluded that the     Sloshing analysis determined seismic induced seismic induced           alternate vendor than           NA      available SEP free-board        wave would not impact radar horn.
sloshing analysis         the supplier of the      _______readily         enveloped the           _____________________
Page 4 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION Parameter        Vendor Design         Additional
     #Topic.                        Summary             Reference          Comments       Test' or Analysis*   Results                  Licensee...Evaluation Document #               _-__-                                                 .....
is required. If   radar level                           maximum predicted wave wave impact is    instrumentation                       height. The analysis predicted, then   (Reference Catawba                    determined wave impact on the               calculation CNC-                      the radar horn would not hydrodynamic      1336.04-00-0001).                     occur.
forces should be included in the mounting design loading
______________combinations.
SetFePol      FntoaiyVEGA                 Test                                                     The vendor factory acceptance test SetFePol tsigProcedure                     ARE VA                               dmntaetht           demonstrated reliable operation of the SFP isrmnain            performed during                                               the SFPLI met the        lelintu             naioudrnrmlcdtos
]0   system                                                 000,              N/A                                       and under various simulated test conditions functionality test             the factory    Factory Acceptance                        specification functional acceptance test. TsReotAEArequirements.                                               (e.g. steam exposure). The testing procdur   TestReprt See#16 AEVAdemonstrated                                                               the instrumentation met design poeueSe#6                      Doc. 66-9227809-002                                                   accuracy and repeatability specifications.
Waveguide radar horn is not immersed in SFP water and therefore not susceptible to boron accumulation. During      Licensee concurs that the wave guided radar Sloshing analysis                     postulated SFP boiling,         instrumentation located in the SFP area is not was performed by an                   boron is not transported by     susceptible to degradation due to postulated alternate vendor than                  rising steam/vapor,.oo                  ul-p           h    aegie aa o the vendor whom                                                       brnbidu.Tewv                      uddrdrhr 11]  Boron Build-Up                     N/A       supplied the radar          N/A       Seismic induced sloshing        is elevated above the SFP process and would level instrumentation                 analysis concluded that the     not be susceptible to boron build-up on the (Reference Catawba                    available SFP free-board        horn during postulated SFP boiling conditions, calculation CNC-                       readily enveloped the          nor is it credible that boron crystal 1336.04-00-0001).                      maximum predicted wave         accumulation on the perimeter of the SFP height. The analysis            walls would impede the radar signal strength.
determined that a seismic induced wave would not impact the radar horn.         _____________________
Seismic     Qualification analysis                      Sensor brackets and       Refer to RAIs #4 and #5 responses posted to Pool-side                       withstand    Doc. 51 -9202556-005     NEI 12-02,         electronic enclosure      Catawba Fukushima Sharepoint.
12   Bracket Seismic                              and supporting              34mounting               are seismically Analysis                                     reference                                   qualified to EPRI TR-     The test and analysis of the horn cover and 174-9213558-006,                       107330 or site-specific RRS. adhesive demonstrate adequate seismic Page 5 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND                                CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION
* Prmtr                   Vendor Design          Adtoa
    '           Toi
         #* TopicSummary      PrmtrReference           Document #            C Adtommnts
                                                                            ,omns..        Test or Analysis Results                   Licensee Evaluation                 "
Tests and                 Calculations                                                          withstand capability.
analyses were             32-9208751-002,                        Testing and analysis of horn performed for the         51-9221032-000                          cover and adhesive support    The stress analysis does not need to consider horn cover and            66-9225632-000                         the components can tolerate    hydrodynamic sloshing forces in the design of adhesive to               32-9221237-003                             horizontal and vertical    the mounting brackets. The sloshing analysis demonstrate               66-9225469-000                            accelerations up to 100~g   detenrmined seismic induced wave would not adequate                                                           and SEP sloshing loads up    impact radar horn.
seismic                   Sloshing analysis                                 to 3.37 psi.
withstand                 was performed by an capability,               alternate vendor than                     Seismic induced sloshing the vendor whom                          analysis concluded that the Perform seismic           supplied the radar                        available SEP free-board induced sloshing          level instrumentation                       readily enveloped the analysis to               (Reference Catawba                        maximum predicted wave assess                    calculation CNC-                             height. The analysis hydrodynamic              1336.04-00-0001 ).                      determined wave impact on wave force on                                                       the radar horn would not the radar horn.                                                                occur.
Qualification analysis Doc. 51-9202556-Additional                   Seismic             005, Section 2.1,Sesrbaktan Brces(esr        withstand of             Appendix D andSesrraetan 13 Eetoisad           sensor brackets         supporing0reerence         NEI 12-02,        elcrncecoue              Refer to PAls #4 and #8 responses posted to Electronics n           and electronic           11-9203036-000,             3.4        mounting are seismically    Catawba Fukushima Sharepoint.
Enclosure)       13        eclosre mosuntin               P~iT-1 0330qualified 17-23558-006                                        to EPRI TR-107330 or site-specific   RRS.
mountingCalculations 32-9208751-002, 32-922 1237-003 Shock and          Qualification Analysis                 Sensor, displays, and power    The shock and vibration testing performed for vibration           Doc. 51 -9202556-                    control panel have been        the SEP level instrumentation adequately withstand testing         005, Sections 2.2 and                  tested and/or analyzed for     demonstrates the sensor and power control 14    irtoShock
             &           sesrand analysidsplysfOr supporting i__0references,     NEI 12-02, shock and vibration,loainpanel will Thbereliableintuettoin the installedi riildesign power control             MIL-STD-167-1                      The test parameter values      mounted to the Seismic Category I Auxiliary panel              38-9193058-000,                     provided in IEC Standards,     Building wall and would not be subjected to EN 60068-2-27,                     l EC 60068-2-6 (vibration)     any significant shock or vibration during a Tests and             38-9218022-000,                     and lEC 60068-2-27             postulated beyond design basis event, or Page 6 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION T              T   -               - r                 r Parameter        Vendor  Design Reference        Additional    eto      nlyi        eut               'Licensee Evaluation Topic.         Summary                              Comments     Tet.rAnlyi.Rsut Document #
analyses were         EN 60068-2-6,                 (shock), tables are           during normal operation. The instrumentation performed for the     38-9218023-000,                recommendations and not       is located within the Seismic Category I horn cover and         MIL-STD-202                  mandatory testing levels. Auxiliary Building and is protected from adhesive to      51-9221032-000                 The test parameter values     external wind borne missile threats. The demonstrate       66-9225632-000                 were selected to be           instrumentation installed design location is not adequate shock        32-9221237-003                 consistent with previous      susceptible to vibration from surrounding withstand.      66-9225469-000                 shock and vibration testing   rotating equipment. The radar sensor and Additional testing    38-9228047-000                 performed on the VEGA         power control panel design location provides was performed                                       supplied equipment. The       spatial separation from surrounding SSCs, for the power                                     test parameter values         such that potential seismic interaction with control panel                                    specified envelope the        surrounding SSCs is also not a concem.
assembly.                                      expected levels for the equipment installed location, The post modification testing will demonstrate due to the fact that the      reliable operation of the instrumentation, which equipment is mounted to       confirms no damage occurred during shipping, seismic structures within the handling and installation. Similarly, the plant. This approach is       performance of monthly channel functional consistent with similar       comparisons will serve to confirm proper technology used in the        operation of the instrumentation, or provide a same application at other     means of early detection of potential installations.                instrument degradation.
The vibration testing         Refer to RAI #7 response posted to Catawba deviated from the IEC         Fukushima Sharepoint.
60068-2-6 recommended frequency range and           The test and analysis of the homn cover and displacement magnitude for    adhesive demonstrate adequate shock large power plant equipment   withstand capability.
(TABLE C.2). In-lieu of the 10-55 Hz and minimum displacement of 0.15 mm recommended in TABLE C.2, the power control panel vibration testing utilized a narrower frequency band (5-25 Hz) and a more limiting displacement magnitude (1.6 mm). These values were deemed to be acceptable and envelopingq Page 7 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND    CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION Prmtr    Vendor Design   Adtoa
   #Topic          Parameterar      Reference                                                 Liceseetvalutio Summary.       Document #*    Comments  -Test or Analysis Results,         iesevlain        *..
for equipment rigidly mounted to a Seismic Category I structure, based on engineering judgment.
The shock testing deviated from the IEC 60068-2-27 recommended peak acceleration and duration for land-based permanently installed equipment. In-lieu of the 15 g's peak acceleration and duration of 11 in-sec recommended in TABLE A.1, the power control panel vibration testing utilized an acceleration of log with a 6 in-sec duration. These values were deemed to be acceptable and enveloping for equipment rigidly mounted to a seismic Category I structure, based on engineering judgment.
Testing and analysis of horn cover and adhesive support the components can tolerate horizontal and vertical accelerations up to 100~g and SFP sloshing loads up to 3.37 psi.
15  Requirements  Not required by      N/A           N/A                   N/A                          N/A Traceability       order Page 8 ofll1

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND                         CNS    SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION Parameter    Vendor Design           Additional *
      #Topic,       ,         PSmaramete             oumnReference
                                                           *       ,
* omns'-      Test or Analysis Results                 ,, Licensee Evaluation:
Inspection of waveguncidelitestf         VEGA Test                                                              The vendor factory acceptance test pofefntioalityr tof    Procedure AREVA                                                        demonstrated reliable operation of the SFP pwrtaseto      Doc. 38-9219704-                        Test demonstrates that       level instrumentation under normal conditions 1 FAccetonce          battmeryresenso           000, Factory             N/A       specification requirements      and under various simulated test conditions AcpacTet      mascurement             Acceptance Test                               were met.                   (e.g. steam exposure). The testing accurcys ostandeorRVADc                                                                     demonstrated the instrumentation met design an ae n66-9227809-002                                                                    accuracy and repeatability specifications.
waveguide The vendor factory acceptance test demonstrated reliable operation of the SEP Normal and                                              Normal conditions accuracy    level instrumentation under normal conditions accident        AREVA Instruction        EA-12-051,     +1 inch, error due to all   and under various simulated test conditions 17     Channel             conditions SEP        manual Doc. 01-             1.7        effects including 212&deg;F     (e.g. steam exposure). The testing Accuracy                   level           9223080-003,           NEI 12-02,   saturated steam +_3 inches. demonstrated the instrumentation met design measurement            Section 11.6              3.7        Accuracy verified during    accuracy and repeatability specifications.
accuracy                                              factory acceptance testing.
Refer to RAI #11 response posted on Catawba Fukushima Sharepoint.
Based on vendor analysis the battery capacity Qualification Analysis                                               is deemed sufficient to support reliable PwrLftmof                       Doc. 51-9202556-         EA-1 2-051,                                instrument channel operation until off-site 18     Pwr005,                                         Section 2.9,         1.6,      Battery capacity at full load resources can be deployed by the mitigating Cosmtofbteybcuoat p                Instruction Manual       NEl 12-02,      is expected to readily     strategies in response to Order EA-12-049.
fllod01-9223080-003,                    3.6             exceed 7 days.
Section 11.7                                                     Refer to RAI #10 response posted to Catawba Fukushima Sharepoint.
Application-specific information on                                                                          The vendor technical manual has been 19  Mecnical            the installation,     AEADc             1        N/A                    N/A              reviewed, accepted and incorporated in the Mauloperation,                and'       238-0                                                          engineering change package.
maintenance of the SFPLI Page 9 of 11

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND                      CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION Parameter    Vendor Design         Adionl         Ts       A1       R    t Topic            armamee             RefrenetiommnalTs                    s or Analysis Rsuls                    Licensee Evaluation
                  , ,,                        Document #          -____
EA-1 2-051, 1.8 NEI 12-02, 3.8 Based on negligible drift rate of VEGA electronics experienced                                 Refer to RAls #12 and #15 responses posted Periodic          EA-12-051, 1.8        over large                                to Catawba Fukushima Sharepoint.
indication        AREVA Doc. 01-        user base, 20  Calibration            checks,           9238-0,                periodic                NAIn-lieu                horn rotation and use of a portable clbain       Sections 7.0 and     calibration is             NAtarget,                periodic verification of proper radar checks,                911not               needed,                               channel functionality can be achieved by calibration              911Functional                                              varying SEP water level (minimum 2 points) verification                               and proper level indication.
can be achieved using cross channel checks and functional checks per vendor manual.
The instrumentation is required to function to provide SEP level indication for a beyond design basis event. Performance of a FMEA 21  Failure Modes                                  N/A                 N/A                  N/A              is not warranted for this type of an application.
and Effects               N/A                                                                             Reasonable assurance that both channels are Analysis (FMEA)                                                                                           not susceptible to a common mode failure is provided by satisfying the NE1 12-02 guidance.
Emissions and     Qualification Analysis                   VEGAPULS 62 ER has         The EMI/RFI susceptibility and emissions susceptibility     Doc. 51-9202556-                     been tested for emissions to teastingeperformedfrtes     waveguide asradarth 22     EMI Testing       testing for      005, Section 2.6 and         N/A      both MIL and lEC standards   trnmtepovdsaqueasrncth VEGAPULS 62       supporting references                  adfr           tl 4 '   EG   instrumentation will be compatible in the ERE-6 004standpards                                              design location. The testing was MIL-STD-461 E,                               sadrsconservatively                 performed with unshielded Page 10 ofil1

ATTACHMENT 5 BRIDGING DOCUMENT BETWEEN VENDOR TECHNICAL INFORMATION AND                              CNS SPECIFIC CONSIDERATIONS FOR SPENT FUEL POOL INSTRUMENTATION r# .......Topic
       * -Toc   .... "  Parameter Summa             V......
endorDesinc    Additional Reference... Comments                 TsorAalysis Results                        ieseEauto 58-9214362-000,                                                        interconnecting wiring. The Catawba level 38-9219863-000,                                                        channel design included shielded signal 38-9218965-000,                                                        cabling, and grounding of the power control 38-9218966-000,                                                        panel.
38-9219862-000, 38-92 18967-000,                                                       Post-modification testing has demonstrated 38-9218968-000,                                                        acceptable operation in the installed location.
38-9218969-000, 38-9218970-000,                                                        During a postulated BDBEE, it is possible that 38-0218964-000                                                        intermittent UHF radio operation could occur in the vicinity of the radar transmitter.
Successful long-term SFP monitoring capability during a postulated BDBEE would not be inhibited by potential intermittent radio
_____________  __________________transmissiontransissinrinerfeence Page 11 of 11

-4 ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 NRC Order EA-12-049 FLEX FINAL INTEGRATED PLAN Catawba Nuclear Station, Units I & 2 February 2016 Page 1 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 Table of Contents Contents
: 1. Background .....................................................................................                  5
: 2. Order Implementation..........................................................................                    6 2.1    General Elements................................................................................ 6 2.2    Strategies......................................................................................... 8 2.3    Reactor Core Cooling Strategy ................................................................. 9 2.3.1   Phasel1.......................................................................................... 9 2.3.2   Phase 2........................................................................................ 10 2.3.3   Phase 3........................................................................................ 11 2.3.4   Availability of Systems, Structures, and Components ...................................... 11 2.3.5   FLEX Connections............................................................................. 13 2.3.6   Plant Instrumentation.......................................................................... 15 2.3.7   Thermal-Hydraulic Analysis................................................................... 16 2.3.8   Reactor Coolant Pump Seals................................................................. 17 2.3.9   Shutdown Reactivity Analysis ................................................................ 17 2.3.10  FLEX Pumps................................................................................... 17 2.3.11  Electrical Analysis ............................................................................. 19 2.4    Spent Fuel Pool Cooling/Inventory Strategy.................................................. 20 2.4.1   Phase 1I........................................................................................ 20 2.4.2   Phase 2 ........................................................................................                20 2.4.3   Phase 3 ........................................................................................ 20 2.4.4   Availability of Structures, Systems, and Components ...................................... 20 2.4.5   Ventilation ....................................             .................................................. 21 2.4.6   Plant Instrumentation.......................................................................... 21 2.4.7   Thermal-Hydraulic Analysis................................................................... 22 2.4.8   FLEX Pump and Water Supplies............................................................. 22 2.4.9   Electrical Analysis ............................................................................. 22 2.5    Containment Function Strategy................................................................ 23 2.5.1   Phase 1........................................................................................ 23 2.5.2   Phase 2 ........................................................................................ 23 Page 2 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.5.3    Phase 3 ........................................................................................ 23 2.5.4    Availability of Structures, Systems, Components ........................................... 24 2.5.5    Plant Instrumentation.......................................................................... 24 2.5.6    Thermal-Hydraulic Analysis .................................................................. 24 2.5.7    Electrical Analysis ............................................................................. 25 2.6     Characterization of External Hazards......................................................... 25 2.6.1    Seismic Events ................................................................................ 25 2.6.2    External Flooding .............................................................................. 26 2.6.3    Storms such as Hurricanes, High Winds, and Tornadoes.................................. 26 2.6.4    Extreme snow, ice and cold .................................................................. 26 2.6.5    Extreme heat .................................................................................. 26 2.7     Planned Protection of FLEX Equipment ...................................................... 27 2.7.1    FLEX Storage Facility......................................................................... 27 2.8     Planned Deployment of FLEX Equipment.................................................... 27 2.8.1    Haul Paths and Accessibility.................................................................. 27 2.8.2    Deployment of Strategies..................................................................... 29 2.8.3    Fueling of Equipment.......................................................................... 30 2.9     Sequence of Events and Staffing ............................................................. 31 2.9.1    Sequence of Events........................................................................... 31 2.9.2    Staffing ......................................................................................... 34 2.10    Offsite Resources .............................................................................. 35 2.10.1   National SAFER Response Center (NSRC)................................................. 35 2.10.2   Equipment ..................................................................................... 36 2.11    Habitability and Operations.................................................................... 36 2.11.1   Equipment Cooling ............................................................................ 36 2.11.2   Freeze Protection ............................................................................. 36 2.11 .3  Hydrogen Ventilation.......................................................................... 37 2.11.4   Personnel Habitability......................................................................... 37 2.12    Water Sources.................................................................................. 37 2.12.1   SG Feedwater ................................................................................. 37 2.12.2   RCS Makeup................................................................................... 38 2.12.3   SFP Makeup ................................................................................... 38 2.13    Shutdown and Refueling Analysis............................................................. 38 Page 3 of 44

ATTACH MENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.14   Procedures and Training....................................................................... 39 2.14.1  Procedural Guidance.......................................................................... 39 2.14.2  Training ........................................................................................      39
: 3. Acronyms.........................................................................         i............. 41
: 4. References..................................................................................... 43 Page 4 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2

===1. Background===
In 2011, an earthquake-induced tsunami caused Beyond-Design-Basis (BDB) flooding at the Fukushima Dai-ichi Nuclear Power Station in Japan. The flooding caused the emergency power supplies and electrical distribution systems to be inoperable, resulting in an extended loss of AC power (ELAP) in five of the six units on the site. The ELAP led to (1) the loss of core cooling, (2) loss of spent fuel pool cooling capabilities, and (3) a significant challenge to maintaining containment integrity. All DC power was lost early in the event on Units 1 & 2 and after some period of time at the other units. Core damage occurred in three of the units along with a loss of containment integrity resulting in a release of radioactive material to the surrounding environment.
The U.S. Nuclear Regulatory Commission (NRC) assembled a Near-Term Task Force (NTTF) to advise the Commission on actions the U.S. nuclear industry should take to preclude core damage and a release of radioactive material after a natural disaster such as that seen at Fukushima. The NTTF report (Reference 1) contained many recommendations to fulfill this charter, including assessing extreme external event hazards and strengthening station capabilities for responding to beyond design basis external events (BDBEEs).
Based on NTTF Recommendation 4.2, the NRC issued Order EA-12-049 (Reference 2) on March 12, 2012 to implement mitigation strategies for BDBEEs. The order provided the following requirements for strategies to mitigate BDBEEs:
* Licensees shall develop, implement, and maintain guidance and strategies to maintain or restore core cooling, containment, and Spent Fuel Pool (SFP) cooling capabilities following a BDBEE.
          *These strategies must be capable of mitigating a simultaneous loss of all AC power and loss of normal access to the ultimate heat sink (UHS) and have adequate capacity to address challenges to core cooling, containment and SFP cooling capabilities at all units on a site subject to the Order.
          *Licensees must provide reasonable protection for the associated equipment from external events. Such protection must demonstrate that there is adequate capacity to address challenges to core cooling, containment, and SEP cooling capabilities at all units on a site subject to the Order.
* Licensees must be capable of implementing the strategies in all modes.
* Full compliance shall include procedures, guidance, training, and acquisition, staging or installing of equipment needed for the strategies.
The order specifies a three-phase approach for strategies to mitigate BDBEEs:
* Phase 1 - Initially cope relying on installed equipment and on-site resources.
* Phase 2 - Transition from installed plant equipment to on-site FLEX equipment.
* Phase 3 - Obtain additional capability and redundancy from off-site equipment and resources until power, water, and coolant injection systems are restored or commissioned.
NRC Order EA-1 2-049 (Reference 2) required licensees of operating reactors to submit an overall integrated plan, including a description of how compliance with these requirements would be achieved. The Order also required licensees to complete implementation of the requirements no later than two refueling cycles after submittal of the overall integrated plan or December 31, 2016, whichever comes first.
Page 5 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 The Nuclear Energy Institute (NEI) developed NEI 12-06 (Reference 3), which provides guidelines for nuclear stations to assess extreme external event hazards and implement the mitigation strategies specified in NRC Order EA-12-049. The NRC issued Interim Staff Guidance JLD-ISG-2012-01 (Reference 4), dated August 29, 2012, which endorsed NEI 12-06 with clarifications on determining baseline coping capability and equipment quality.
Duke Energy (Duke) declared that Catawba Nuclear Station (CNS) Unit 2 was in compliance with Order EA-1 2-049 on March 31, 2015 following the 2EOC20 refueling outage, which is within two refueling cycles of the submittal of the OIP dated February 28, 2013 (Reference 5). Duke declared that CNS Unit 1 was in compliance with Order EA-12-049 on December 16, 2015 following the I E0C22 refueling outage, also within two refueling cycles of the QIP submittal (Reference 6).
NRC Order EA-12-051 (Reference 7) required licensees to install reliable SEP instrumentation with specific design features for monitoring SEP water level. This order was prompted by NTTF Recommendation 7.1 (Reference 1).
NEI 12-02 (Reference 8) provided guidance for compliance with Order EA-12-051. The NRC determined that, with the exceptions and clarifications provided in JLD-ISG-2012-03 (Reference 9),
conformance with the guidance in NEI 12-02 is an acceptable method for satisfying the requirements in Order EA-12-051.
Duke declared that CNS Unit 2 was in compliance with Order EA-12-051 on March 31, 2015 following the 2E0C20 refueling outage, which is within two refueling cycles of the submittal of the OIP dated February 28, 2013 (Reference 5). Duke declared that CNS Unit 1 was in compliance with Order EA-12-051 on December 16, 2015 following the 1 EOC22 refueling outage, also within two refueling cycles of the OIP submittal (Reference 6).
: 2. Order Implementation 2.1      General Elements The assumptions used for the evaluations of an ELAP/Loss of Ultimate Heat Sink (LUHS) event and the development of diverse and flexible coping strategies (FLEX strategies) are stated below.
Initial conditions and boundary conditions consistent with NEI 12-06 were established to support development of FLEX strategies, as follows:
* The reactor is initially operating at power, unless there are procedural requirements to shut down due to the impending event. The reactor was operating at 100% power for the past 100 days.
                 *The reactor is successfully shut down when required (i.e., all rods inserted, no Anticipated Transient Without Scram (ATWS)). Steam release to maintain decay heat removal upon shutdown functions normally, and reactor coolant system (RCS) overpressure protection valves respond normally, if required by plant conditions, and reseat.
* On-site staff is at site administrative minimum shift staffing levels.
* All personnel on-site are available to support site response.
* The reactor and supporting plant equipment are either operating within normal ranges for pressure, temperature and water level, or available to operate, at the time of the event consistent with the design and licensing basis.
Page 6 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 No specific initiating event is used. The initial condition is assumed to be a loss of off-site power (LOOP) with installed sources of emergency on-site AC power and station blackout (S80) alternate AC power sources unavailable with no prospect for recovery.
All AC power supplies were considered to fail concurrently at the beginning of the event.
       *Cooling and makeup water inventories contained in systems or structures with designs that are robust with respect to seismic events, floods, and high winds and associated missiles are available.
      *Normal access to the ultimate heat sink (UHS) is lost, but the water inventory in the UHS remains available and robust piping connecting the UHS to plant systems remains intact.
The motive force for UHS flow, i.e., pumps, is assumed to be lost with no prospect for recovery.
       *Permanent plant equipment that is contained in structures with designs that are robust with respect to seismic events, floods, and high winds and associated missiles, are available. SSCs were considered seismically robust if seismic requirements were imposed by licensing requirements. The portion of the fire protection system that is robust with respect to seismic events, floods, and high winds and associated missiles is available as a water source. Installed electrical distribution systems, including inverters and battery chargers, remain available since they are protected.
       *Fuel for FLEX equipment stored in structures with designs that are robust with respect to seismic events, floods and high winds and associated missiles, remains available.
      *No additional accidents, events, or failures are assumed to occur immediately prior to or during the event, including security events.
      *Reactor coolant inventory loss consists of unidentified leakage at the upper limit of Technical Specifications, reactor coolant letdown flow (until isolated), and reactor coolant pump seal leak-off at normal maximum rate.
      *For the SFP, all boundaries (e.g., liner, gates) and the SFP cooling system are assumed to be intact. The SEP heat load is assumed to be the maximum design basis heat load.
In addition, inventory loss from sloshing during a seismic event does not preclude access to the pool area.
Additional key assumptions associated with design and implementation of FLEX strategies are as follows:
      *Additional deployment resources are assumed to begin arriving at 6 hours and the site Emergency Response Organization (ERO) will be fully staffed at 24_ hours after the event.
      *The plant Technical Specifications contain the limiting conditions for normal unit operations to ensure that design safety features are available to respond to a design basis accident and direct the required actions to be taken when the limiting conditions are not met. The result of the BDBEE may place the plant in a condition where it cannot comply with certain Technical Specifications and/or with its Security Plan, and, as such, may warrant invocation of 10 CER 50.54(x) and/or 10 CFR 73.55(p). (Reference 10)
      *All installed AC power supplies (emergency on-site and SBO Alternate AC power sources as defined by 10CFR50.2) were considered not available and not immediately recoverable.
Page 7 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 &.2
           *Where non-safety, non-seismically designed, permanently installed equipment is used for FLEX strategies, SSCs were considered seismically robust if:
o   Seismic Qualification Utility Group (SQUG) methods are applied per the existing plant licensing basis.
o   Testing, analysis or experience-based methods are applied for the equipment class at design basis seismic levels.
o    Methodologies in EPRI 1019199, Experience Based Seismic Verification Guidelines for Piping and Tubing Systems, can be successfully applied relative to the Safe Shutdown Earthquake (SSE).
o   Other industry-recognized codes such as AWWA DI00 are applied to demonstrate functionality at SSE level ground motion.
o   High Confidence of a Low Probability of Failure (HCLPF) capacities are determined (e.g., EPRI NP-6041 Rev 1) conservative compared to the SSE.
* Personnel access to and qualification of equipment that forms a part of the FLEX strategy assumed no core damage.
* Per NEI 12-06 Section 3.2.1.8, maximum environmental room temperatures for habitability or equipment operation may be based on NUMARO 87-00 guidance if other design basis information or industry guidance is not available. Extreme high temperatures are not expected to impact the utilization of offsite resources or the ability of personnel to implement the required FLEX strategies.
           *Access through security fences, doors, and other barriers will be unimpeded and not require additional resources.
2.2 Stratecqies The objective of the FLEX strategies is to establish indefinite coping capability in order to:
* Prevent damage to the fuel in the reactors
* Maintain the containment function
* Maintain cooling and prevent damage to fuel in the SFP This indefinite coping capability will address an ELAP   - loss of off-site power, emergency diesel generators (EDGs) and any alternate AC source, but not the loss of AC power to buses fed by station batteries through inverters  -  with a simultaneous LUHS. This condition could arise following external events that are within the existing design basis with additional failures and conditions that could arise from a BDBEE.
The plant indefinite coping capability is attained through the implementation of pre-determined strategies (FLEX strategies) that are focused on maintaining or restoring key plant safety functions. The FLEX strategies are not tied to any specific damage state or mechanistic assessment of external events. Rather, the strategies are developed to maintain the key plant safety functions based on the evaluation of plant response to the coincident ELAP/LUHS event.
A safety function-based approach provides consistency with, and allows coordination with, existing plant emergency operating procedures (EOPs). FLEX strategies are implemented in support of EOPs using FLEX Support Guidelines (FSGs).
Page 8 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 The strategies for coping with the plant conditions that result from an ELAP/LUHS event involve a three-phase approach:
* Phase 1 - Initially cope by relying on installed plant equipment and on-site resources.
* Phase 2   - Transition from installed plant equipment to on-site FLEX equipment.
* Phase 3   - Obtain additional capability and redundancy from off-site equipment and resources until power, water, and coolant injection systems are restored.
The transitions to Phase 2 and Phase 3 will occur at different times for different portions of the FLEX strategies.
The strategies described below are capable of mitigating an ELAP/LUHS resulting from a BDBEE by providing adequate capability to maintain or restore core cooling, containment, and SEP cooling capabilities at CNS. Though specific strategies have been developed, due to the inability to anticipate all possible scenarios, the strategies are also diverse and flexible to encompass a wide range of possible conditions. These pre-planned strategies developed to protect public health and safety are integrated into EOPs in accordance with established change processes, and their impact to the design basis capabilities of the unit are evaluated under 10 CFR 50.59.
2.3    Reactor Core Coolingq Strategqy 2.3.1        Phase I Following reactor and turbine trips of each operating unit, the Turbine Driven Auxiliary Feedwater Pump (TDAFWP) supplies water to the Steam Generators (SGs) secondary side to cool down the Reactor Coolant System (RCS). CNS will limit cool down to a rate near 100&deg;F/hr to minimize RCS system inventory loss, while cooling the Reactor Coolant Pump (RCP) seals in a controlled manner. The SGs will be depressurized to 240 psig and RCS temperature will be maintained above 280&deg;F. Maintaining SG pressure at this level prevents nitrogen injection from the Cold Leg Accumulators (CLAs) into the RCS The TDAFWP will draw suction from the Upper Surge Tank (UST), the Auxiliary Feedwater Condensate Storage Tank (CACST), or the hotwell, which are condensate grade water sources. However, these sources are susceptible to damage from wind and seismic events and may not be available. The TDAFWP can also draw suction from buried piping in the Condenser Circulating Water (RC) system, which has been evaluated to be seismically robust and can supply sufficient water to support decay heat removal for at least 48 hours.
The valve to enable TDAFW pump suction from embedded RC piping opens on a loss of control power or instrument air, and will provide flow when suction pressure from the UST or the hotwell drop5s below the RC static pr'essure. CNS will control SG level by throttling CA system flow control valves from the Control Room or starting and stopping the TDAFWP.
To remove heat, CNS will discharge steam through the SG Power Operated Relief Valves (PORVs), which can be operated from the control room using vital battery power and safety grade nitrogen backup for the valve actuators. The SG PORVs can also be manually operated.
The vital station batteries provide DC power for essential instrumentation. Vital battery load shedding will be initiated 2.5 hours into the event. The CNS load shedding strategy will maintain power supply from the A and 0 vital batteries for approximately 10 hours, and the B and C vital batteries for approximately 24 hours.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 No action is necessary for managing RCS inventory or reactivity during Phase 1. CNS-specific analysis concluded that core uncovery will not occur until at least 72 hours into the event. Additionally, CNS determined that boration of the RCS to control reactivity is not needed until FLEX equipment can be deployed during Phase 2.
2.3.2      Phase 2 The Phase 2 core cooling strategy continues to use the SGs as the heat sink. CNS has multiple strategies for providing feedwater to the SGs using FLEX equipment:
            *If the TDAFWP is still available after the RC piping static inventory has been depleted, CNS can use a portable low pressure pump to establish a new suction source. The low pressure pump will draw suction from the Standby Nuclear Service Water Pond (SNSWP) to a connection in the service water system (RN) piping. CNS will align valves to direct flow to the auxiliary feedwater (CA) system and the TDAFWP.
            *When the TDAFWP is no longer available (e.g., steam pressure becomes too low), a portable, low pressure pump will provide feedwater by taking suction from the SNSWP and discharging into portable, medium pressure pumps located near the Exterior Doghouses of each Unit. The medium pressure pumps will discharge into connections in the Doghouses to establish flow to the SGs. CNS can establish a flow path using only hoses; fire protection system piping can also be used to reduce hose deployment.
             *SGs may be fed directly from the FLEX Raw Water Distribution system (i.e., without the medium pressure pump). This approach requires depressurizing the SGs below the discharge pressure of the FLEX Raw WaterDistribution System (approximately 150 psig). The option to use only low pressure pumps is less preferred because it provides less capacity and it requires feeding and steaming SGs that are close to dry. Additionally, this strategy is not preferred prior to CLA isolation, because of the potential for injection of nitrogen into the RCS, which may occur at 140 psig. CNS can establish a flow path using only hoses; fire protection system piping can also be used to reduce hose deployment.
CNS will deploy 600V diesel generators (DGs) to provide power for the FLEX strategies within 9 hours of the event.
            *A back-feed receptacle connection on selected motor control centers (MCCs) allows re-powering key installed equipment, including battery chargers. The A vital battery charger will be re-powered within 11.5 hours and the D vital battery charger will be re-powered within 12.5 hours. Battery chargers will be aligned to the vital batteries prior to battery voltage dropping below acceptable values. The spare battery charger will provide redundancy as a backup for either safety train on the applicable unit.
The B and C vital battery chargers will be repowered within 24 hours. This strategy allows all channels of vital instrumentation to be maintained.
            *The 600V FLEX DGs will also be connected to the plant MCCs for re-powering via the FLEX 'Backbone,' which consists of permanently installed cables, portable panelboards, and transformers. If normal power supplies are not available, they will be disconnected to allow connection of alternate supply cables. The alternate cable(s) will be plugged directly into the MCCs via FLEX receptacles for the component. Cables will be routed out of the Auxiliary Building for connection to the FLEX DGs.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 &.2 CNS will provide Phase 2 RCS makeup using a portable high pressure pump. Initially, CNS will use the reactor head vent valves to establish venting/letdown of the RCS. These valves will be re-powered using the MCC backfeed strategy and FLEX DGs. The high-pressure pump will draw suction from the Refueling Water Storage Tank (FWST) via a connection on the piping for makeup to the SEP. The high pressure pump will discharge into safety injection pump discharge piping. Sufficient borated water will be added to maintain the core sub-critical, in a xenon-free condition, at 350&deg;F.
The CLA isolation valves will be closed to prevent nitrogen from being injected into the RCS prior to final cooldown and depressurization. The CLA isolation valves will be re-powered using the FLEX DGs. After the CLAs are isolated or vented, CNS will depressurize SGs to 160 psig at a rate that will limit cooldown to less than 1 00&deg;F per hour.
A portable FLEX sump pump is placed in each CA TDAFWP pit to pump out normal drains input from the TDAFWP to the room sump before flooding impacts operation of the TDAFWP. The flooding impact could occur 7 hours into the event. Additionally, Groundwater Drainage System (WZ) sump pumps must be in operation by 10.6 hours into the event to prevent installed sump pump motors from being flooded. CNS can also deploy FLEX sump pumps in various locations (in addition to the TDAFWP pits) to manage groundwater intrusion.
2.3.3       Phase 3 After CNS has decreased RCS temperature to less than 350&deg;F and RCS pressure to less than 385 psig, CNS will transition to cooling by Residual Heat Removal (RHR). CNS will repower the Component Cooling (KC) system pumps to provide cooling water to the RHR pumps.
The National SAFER Response Center (NSRC) will deliver equipment to CNS to establish and align RHR and KC systems.
oTwo portable pumps will take suction from the SNSWP and discharge by hose to the supply piping of the RN system (one portable pump per RN train). The pressurized RN system will be manually aligned to supply flow to the KG system heat exchangers.
o    Large portable DGs from the NSRC will power KC and RHR system pumps.
The RHR pumps can provide borated makeup water to the core by taking normal suction from the FWST. If needed, mobile boration units from the NSRC can be used to refill the FWST through existing vents.
The NSRC will deliver a water treatment skid that can provide a cleaner water source than the SNSWP.
Additional diesel fuel for portable equipment will be brought in from off-site resources when required.
2.3.4       Availability of Systems. Structures, and Components The FLEX strategy for core cooling relies on various installed systems, structures, and components (SSCs). These SSCs are protected in regard to the applicable extreme external hazards as discussed below.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.3.4.1  Structures The FLEX strategy relies on selected site structures to provide protection from the applicable extreme external hazards. Specifically, the FLEX strategy relies on Containment, the Auxiliary Building, the Nuclear Service Water System Pumphouse, the SG Doghouses, and the Fuel Handling Building, which are all Seismic Category I structures.
2.3.4.2  Pipingq and Fittinqs The FLEX strategy relies on installed piping and fittings in various plant systems to deliver water for core cooling. Such piping is either designed for safety-related service or was analyzed as part of development of the FLEX strategies to confirm availability following any of the applicable hazards (e.g., the RC piping).
Primarily, CNS relies on piping and components from the Reactor Coolant System (RCS), Auxiliary Feedwater (CA) system, Safety Injection (NI) system, Residual Heat Removal (RHR) system, the Nuclear Service Water (RN) system, the Component Cooling (KC) system, the Condenser Circulating Water (RC) system, the Main Steam to Auxiliary Equipment (SA) system, the Main Steam Vent to Atmosphere (SV) system, the Steam Generator Wet Layup Recirculation (BW) system, the Spent Fuel Pool Cooling (KF) system, and the Refueling Water (FW) system.
2.3.4.3 Turbine Driven AFW Pump (TDAFWP)
The CNS FLEX strategy relies on the TDAFWP to provide feedwater for the SGs during Phase 1. The TDAFWP and its flow control valVes (FCVs) are safety-related, seismically-qualified components that are located in the Auxiliary Building, which is a Seismic Category I structure. These components are therefore protected from the applicable hazards.
2.3.4.4 Steam Generator Power Operated Relief Valves (PORVs)
The CNS FLEX strategy relies on the SG PORVs to remove heat during SG cooling, because cooling from the main condenser is not available. The SG PORVs are safety-related, seismically-qualified components located inside the SG Doghouses, which are Seismic Category I structures. These components are therefore protected from the applicable hazards.
2.3.4.5 Vital Station Batteries The CNS FLEX strategy relies on vital station batteries to power vital instrumentation.
The vital station batteries and associated DC distribution systems are located within the Auxiliary Building, which is a Seismic Category I structure. The vital batteries are therefore protected from the applicable hazards.
2.3.4.6 Electrical Distribution System CNS uses selected plant electrical distribution equipment to repower installed components credited for the FLEX strategy. Electrical distribution components used for the FLEX strategy are located within Seismic Category I structures and will therefore be available following the applicable hazards.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.3.4.7      Groundwater Drainagqe (WZ') Pumps The CNS FLEX strategy relies on the WZ pumps to manage ground water in-leakage.
These pumps will be re-powered by the 600V FLEX DGs. CNS has six WZ pumps (i.e.,
A1I,A2, B1,B2, and Cl,C2 pumps), which can all be powered simultaneously. The C groundwater sump and associated components are not seismically qualified; however, if the C WZ sump is intact, it is the preferred source. If the C WZ sump pump cannot be used, the A and B WZ sump pumps have adequate capacity to manage groundwater in-leakage. The A and B WZ sump pumps are protected from all hazards and will be available following a BDBEE.
2.3.4.8      Condenser Circulatingi Water (RC) Pipingq Although the preferred water source for SG feedwater is a clean source (e.g., the UST or the hotwell), the CNS FLEX strategy credits the RC piping for cooling water because it is protected from all applicable hazards. The Unit I RC system has a total usable volume of approximately 982,000 gallons and the Unit 2 RC system has a total usable volume of approximately 1,280,000 gallons. These water inventories are sufficient to provide decay heat removal capability for at least 48 hours.
2.3.4.9      Standby Nuclear Service Water Pond ('SNSWP')
After depletion of the inventory in the RC piping, CNS can use the SNSWP as a suction source for SG feedwater via the FLEX Raw Water Distribution System. The SNSWP is nuclear safety-related, seismically-protected, and contains sufficient inventory to support the FLEX strategy for an essentially indefinite duration (i.e., well into Phase 3).
2.3.4.10     Refueling Water Storaqe Tank (FWST)
The FWST is the credited source of makeup for the RCS. The minimum inventory of the intact FWST is 377,537 gallons and contains boron in accordance with CNS technical specifications. The FWST is seismically-qualified and the bottom portion is protected by a missile wall. If the top portion of the FWST is not damaged by the BDBEE, the initial inventory will be sufficient for RCS makeup for the duration of the event. If the top portion of the FWST is damaged, makeup to the FWST may be required within 52 hours.
            *If necessary, makeup to the FWST can be accomplished using the NSRC-supplied mobile boration skid or other borated water sources (e.g., trucking from McGuire Nuclear Station). Less preferred options for RCS makeup include recovery of borated water from the FWST annulus (if it was damaged) or preparing solution in a FLEX Portable Mixing Tank using borated water from the Boric Acid Tank (BAT) or barrels of dry boric acid.
2.3.5          FLEX Connections FLEX connections for water and power are installed on various plant systems to facilitate use of portable FLEX equipment. Primary and alternate connections are available for each system used for FLEX. These connection points are protected from the applicable extreme external hazards as discussed below.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.3.5.1  FLEX SG Feedwater Connections CNS can provide SG feedwater using portable FLEX equipment and hoses to connect to installed systems. Connection points used to establish a flow path include the following:
               *The primary SG feedwater connections are located on CA system piping in the SG Doghouses downstream of the containment isolation valves. A set of four connection points provides one connection for each of the four SGs. At each of the four connection points, CNS will replace a blind flange with a hose adapter to establish a flow path. These connections are located inside a Category I structure (i.e., the SG Doghouses) and are protected from the applicable hazards.
               *The alternate SG feedwater connections are located on BW system piping, and are located in the SG doghouses. A set of four connection points provides one connection for each of the four SGs. At each of the alternate connection points, CNS will replace the blind flange with a hose adapter to establish a flow path.
These alternate connections are located inside a Category I structure (i.e., the SG Doghouses) and are protected from the applicable hazards.
2.3.5.2 FLEX RN Connections CNS can provide water through connections on the RN system supply headers to provide feedwater to the TDAFWP or to support long-term RHR during Phase 3.
Connection points used to establish a flow path include the following:
                *The primary RN connection is on RN Train B at a fill valve that is located in the RN Pumphouse. The RN Pumphouse is a Category I structure that provides protection to this connection point for all applicable hazards.
                *The alternate RN connection is on RN Train A at an access plug in the yard.
The plug is inside the protected area fence near the RN Pumphouse.
2.3.5.3 FWST Supply Connection for RCS Makeupi The suction supply for the portable high pressure pump will come from a connection on the FWST supply line for makeup to the SEP. This connection point is in the Auxiliary Building, which is a Seismic Category I structure. Therefore, the supply connection is protected from all applicable hazards.
The suction strategy will designate the FWST of one Unit as the primary common suction source for both Units and the other Unit's FWST as a spare. CNS will use a gated wye assembly to facilitate this approach.
2.3.5.4 FLEX RCS Connections CNS will provide borated water to the RCS using a high pressure makeup pump.
Connection points supporting this strategy are located on the NI pump discharge piping in the Auxiliary Building and can be aligned for cold leg or hot leg injection. Connections are available on both Train A and Train B, which provide primary and alternate connection points. These connections are located on safety-related piping and are in a Seismic Category I structure. Therefore, these connections are protected from the applicable hazards.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.3.5.5      FLEX Electrical Connections The CNS FLEX strategy relies on DGs to charge batteries, maintain vital instrumentation, and repower plant equipment. The following connections are available to support this strategy:
* Back-feed receptacle connections are available on selected MCCs, which can feed train A and B and spare battery chargers, thereby providing primary and alternate connections for battery charging.
                    *Connections to normal plant MCCs will enable re-powering via the FLEX Backbone, which consists of permanently installed cables, portable panelboards, and transformers. The alternate approach is to disconnect normal power supplies to enable connection of alternate supply cables, which can be connected at the portable DG.
To support power for vital instrumentation and other equipment, permanent cabling, receptacles, back-feed breakers and MCC's will be seismically qualified and located above anticipated flood level within Category I structures.
2.3.6          Plant Instrumentation The following instrumentation is relied upon to support the FLEX core cooling strategy:
* RCS Hot Leg Temperature (Thot)
* RCS Wide Range (WR) Pressure
* SG Narrow Range (NR) Level
* Core Exit Thermocouple Temperature
* Pressurizer Level
* Reactor Vessel Level Indicating System
* TDAFW Pump Flow
* SG Pressure
* DC Bus Voltage
* Neutron Flux
* FWST Level
* Containment Sump Wide Range Level
* SG and Pressurizer Cavity Temperature
* RCP Seal Leakoff Flow If the primary instrument is not available or cannot be powered via normal means, CNS will provide an alternate power source or use a portable instrument. CNS can dispatch operators to monitor parameters locally (e.g., CA flow, SG pressure via temporary gauges) or portable test equipment may be used to monitor parameters from inside the Process Control System 7300 cabinet (e.g., SG narrow range level, RCS pressure and temperature, Pressurizer level, FWST level). In addition, CNS has 120 VAC DGs that may be used to provide power for instrumentation ifthe power is not available by other means.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.3.7        Thermal-Hydraulic Analysis CNS performed Fathom analyses to demonstrate that the FLEX pumps have sufficient capability to supply water for the various demands associated with the FLEX strategies.
Specific conclusions from these analyses included the following:
              *A conservative bounding flow rate for SG feedwater demand at 6 hours into the event is 50 gpm per loop (total of 200 gpm per Unit). A conservative bounding flow for SG feedwater demand at 24 hours is 33 gpm per loop (total of 132 gpm per Unit).
              *One FLEX low pressure pump, taking suction from the SNSWP, can supply (by hose) a FLEX medium pressure pump for each Unit for direct feedwater supply to all four SGs (at 50 gpm each; total of 400 gpm for both Units) and four control room spot coolers (at 10 gpm each; total of 40 gpm). SG feedwater supply is sufficient using either the primary or alternate connections.
              *After 24 hours, one FLEX low pressure pump, taking suction from the SNSWP, can supply (by hose) a medium pressure pump for each Unit, which then directly feeds all four SGs (at 33 gpm each), boration equipment for each Unit (at 160 gpm total) and control room spot coolers (at 40 gpm total).
              *One FLEX low pressure pump, taking suction from the SNSWP and discharging by hose to the RN system, can supply feedwater for all four SGs at both Units (maximum of 50 gpm each; 400 gpm total) and the SFPs at a rate significantly exceeding the most conservative steaming rate (96.8 gpm).
              *During Phase 3, two NSRC low pressure pumps, taking suction from the SNSWP and discharging by hose to the RN system, can supply a KG Heat Exchanger for each unit at 3,457 gpm (via A Train) and 3,111 gpm (via B train).
* One FLEX high pressure pump, taking suction from a common FWST and discharging by hose connections into the Unit I and Unit 2 NI system piping, can supply the Unit 1 and Unit 2 RCS at least 40 gpm with back-pressure of 450 psia.
* The CNS FLEX strategies require approximately 489,000 gallons of water per Unit in the first 48 hours of the event.
Condensate grade water is required for SG makeup instead of raw water within 96 hours for Unit 2 based on sludge buildup and 275 hours for Unit 1 based on reaching the corrosion limit. (Unit 1 and Unit 2 have different types of SGs, and design differences result in the large difference in allowable service times with raw water.) CNS will deploy water treatment equipment as part of the Phase 3 FLEX strategy to provide a higher purity water source, if long-term SG feedwater is needed.
The minimum required SG pressure to operate the TDAFW pump is 125 psig. CNS procedures permit SG cool down resulting in a SG pressure of 160 psig following closure of the CLAs, which is well into Phase 2. Therefore, SG pressure will be sufficient to power the TDAFWP at least until the FLEX pumps are deployed.
CNS analysis concluded that a conservatively high inflow to the TDAFW pump sumps is 15 gpm. Accumulation of water at this rate would result in flooding the pump after 7 hours.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.3.8         Reactor Coolant Pump Seals CNS performed a modification to each of the RCP #1 seal leak-off lines to replace the flow orifices with a thicker component that would be suitable for higher differential pressure.
Analyses of this modified configuration show that leak-off flow rates stay within the maximum limits of current WCAP-17601-P, Revision 1 assumptions (i.e., 21 gpm / seal).
See Attachment 7 of the Final Compliance Submittal (Reference 6), "CNS Reactor Coolant Pump Seal Leakage Margin Assessment" for additional discussion on RCP seal leakage.
2.3.9         Shutdown Reactivity Analysis CNS performed a shutdown reactivity analysis that incorporated the guidance provided in the Westinghouse position paper entitled "Westinghouse Response to NRC Generic Request for Additional Information (RAI) on Boron Mixing in Support of the Pressurized Water Reactor Owners Group (PWROG)" (ADAMS Accession Number ML13235A135) with the clarifications specified in the NRC endorsement of this approach (Reference 11). The CNS analysis included a one-hour mixing delay, in accordance with those clarifications.
The CNS analysis concluded that the FLEX high pressure pump, delivering FWST water at 40 gpm, must be started by 13.85 hours into the ELAP event to provide and maintain the necessary 1% shutdown margin and prevent a potential re-criticality during cooldown. To meet this criterion, CNS plans to deploy the high pressure pump prior to 9 hours into the event. For boration after initial cool down of both units, ONS requires delivery of a total of
       *28,100 gallons of FWST water into the RCS, which is well within the nominal FWST inventory of 377,537 gallons, even if damaged above the protective wall.
For the latest NOTRUMP reference case, the PWROG-14027-P, Revision 3 report indicates that reflux cooling may begin in four-loop plants like CNS after 15.6 hours. Addition of borated RCS makeup must occur before reflux cooling to ensure adequate mixing. The CNS timeline for deploying the high pressure pump satisfies this requirement.
2.3.10        FLEX Pumps 2.3.10.1     FLEX Low Pressure Pump After the TDAFWP is secured, the CNS FLEX strategy relies on a portable, diesel-driven, low pressure, high volume pump to supply water from the SNSWP to the FLEX Raw Water Distribution System.
The FLEX low pressure pumps are portable, diesel-driven pumps that can supply a design flow of 3,000 gpm. The FLEX low pressure pump discharge pressure will be maintained at 250 to 275 psig. As discussed in Section 2.3.7, hydraulic analysis shows that the FLEX low pressure pumps have sufficient capacity to support the CNS FLEX strategies.
CNS has two portable FLEX low pressure pumps to satisfy the N+1 requirement.
2.3.10.2    FLEX Medium Pressure Pump After the TDAFWP is secured, the CNS FLEX strategy relies on a portable medium pressure pump to provide makeup water to the SGs. The FLEX Medium Pressure Makeup Pump can also be used to provide RCS makeup if the event occurs when the plant is in modes 5 or 6.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 The FLEX medium pressure pumps are portable, diesel-driven centrifugal pumps that can supply 300 gpm at a maximum pressure of 400 psig. As discussed in Section 2.3.7, hydraulic analysis shows that these pumps have sufficient capacity to support the CNS FLEX strategies.
CNS has three FLEX medium pressure pumps to satisfy the N+1 requirement.
The FLEX low pressure pump provides water to the FLEX medium pressure pump via hoses (and also fire protection piping, if desired).
2.3.10.3  FLEX Higqh Pressure Pump For an ELAP event initiating in Modes 1 - 4, the CNS FLEX strategy relies on a high pressure pump to provide RCS makeup and boration. The FLEX high pressure pumps are portable, diesel-driven, centrifugal pumps that can supply 40 gpm at 1700 psig, which is adequate to support the reactivity control and RCS system make-up requirements for the FLEX strategy.
CNS has three portable FLEX high pressure pumps to satisfy the N+1 requirement.
The credited water supply for the FLEX high pressure pump is the FWST.
2.3.10.4  CA TDAFW Pit Portable Sump Pumps A portable sump pump is placed in the CA TDAFW pit to pump out normal drains input from the TDAFW pump to the room sump before flooding impacts operation of the TDAFWP. Each sump pump is electrically powered and requires 120 V, which can be supplied either via the FLEX Electrical Distribution System, or a 120V DG. The sump pumps can deliver 15 gpm flow at over 30 feet of head, which is adequate to prevent flooding of the TDAFW pit sum ps.
CNS has three 120V sump pumps, which are sufficient to support the CNS FLEX strategies.
2.3.10.5 Auxiliary Building Portable Sump Pumps CNS can deploy portable 600V sump pumps to manage internal flooding from potential pipe breaks. Each 600V sump pump can be powered by one of the 600V DGs, although a single DG may not have sufficient capacity for operation of all three sump pumps simultaneously.
CNS has three 600V sump pumps, which are sufficient to support the CNS FLEX strategies.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 12& 2 2.3.11     Electrical Analysis 2.3.11.1  Diesel Generators CNS relies on DC systems for necessary electrical coping power during Phase I of the ELAP. To extend the coping capability of the vital station batteries, CNS will complete load shedding within 3.5 hours of the event to reduce battery discharge to only essential loads (e.g., vital instrumentation). This action will extend the functional capability of the vital station batteries to at least 11.5 hours for the limiting battery. The CNS FLEX strategy for repowering battery chargers with FLEX DGs will provide electric power prior to the end of battery life.
For longer term electrical power, CNS will deploy portable FLEX DGs (one for each Unit), and associated support equipment to establish the FLEX Electrical Distribution System. CNS has three FLEX 600 VAC DGs to satisfy the N+I requirement. CNS performed an analysis, which verified that the FLEX DGs and planned cable routing arrangement were adequate to support the required loads. The rating of each 600V DG exceeds its total running load during FLEX deployment.
2.3.11.2 Liahtina Post-Fire Safe Shutdown lighting is available in many areas where manual actions are necessary. The Post-Fire Safe Shutdown lights have self-contained batteries with an 8-hour life. Additional portable lighting will be provided for use in the yard.
Lighting units included in the FLEX strategy are as follows:
I. LED tripod mounting Quad Pod LED lights. Units are rated at 20,000 lumens.
: 2. LED string lights
: 3. Miscellaneous helmet lights and flashlights. Typical units and approximate quantities include 50 D cell LED flashlights, 50 helmet lights, and 25 stand spot lights.
LED-mounted tripods will be deployed in many areas including, the Control Room, CA Pump Room(s), Motor Generator (MG) Set Room(s), Interior and Exterior Doghouses, Electrical Penetration Rooms, Battery Room(s), Technical Support Center, and General floor areas in the Auxiliary Building. String lights will also be used in the Control Room and the Technical Support Center. The lighting plan includes a total of 27 LED-mounted tripods and 3 string lights.
Four hard hats with lights are located at the Main Control Room (MCR) exit doors S400 and $406.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.4    Spent Fuel Pool Coolingq/Inventory Strateoqy 2.4.1         Phase I No actions are required during ELAP Phase 1 for SEP make-up because the time to boil is sufficient to enable deployment of Phase 2 equipment.
                   *For the worst case heat load scenario, the SEP will begin to boil after an ELAP/LUHS event in 8.8 hours. This scenario assumes a maximum starting SFP temperature of 125&deg;F and a full core offload during an outage after 6 days.
                   *During normal operation (21 days after the beginning of a refueling outage and initial SEP temperature of 125&deg;F), SEP level can be maintained at least 10 feet above the top of the fuel (above which personnel access to the SFP operating deck is still viable) for 202.9 hours with no makeup. For normal operating conditions, boiling will begin in 37.0 hours.
CNS will monitor SEP water level using SFP level instrumentation.
2.4.2         Phase 2 To compensate for SEP boil-off, CNS will provide makeup water by pumping raw water from the SNSWP using the FLEX low pressure pump.
The primary FLEX strategy for SEP cooling is to pressurize the RN system and then open an installed RN valve and an installed KF valve to establish a flow path to the SEP using installed piping.
The alternate FLEX strategy for SEP cooling is to pressurize the RN system and use a hose to connect RN piping to a valve in the KF skimmer loop. CNS will attach an adapter on the KF valve to facilitate the hose connection.
CNS does not have a FLEX strategy for spray makeup to the SEP. The CNS SFPs are in Category I structures and are connected to safety-related systems, so they are not susceptible to being drained. In accordance with NEI 12-06, Table D-3, spray makeup capability is not required for sites that have SEPs that cannot be drained.
2.4.3         Phase 3 Long term SEP cooling will be accomplished by re-powering the installed KE and KC pumps using a portable generator obtained from the NSRC to provide cooling via normal means.
Cooling of KC will be from the RN system that is being supplied from the portable diesel driven pump located at the SNSWP. Additional diesel fuel for portable equipment will be brought in from off-site resources when required.
2.4.4        Availability of Structures. Systems. and Components 2.4.4.1     Structures The CNS SEP cooling strategy relies on the SEP Building, the Auxiliary Building, and the RN Pumphouse, which are addressed in Section 2.3.4.1. All relevant structures are robust to the applicable external hazards.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.4.4.2      Systems The CNS SEP cooling strategy relies on the KE system, the KC system, and the RN system, which are addressed in Section 2.3.4.2. All relevant systems are robust to the applicable external hazards.
2.4.4.3      SNSWP The CNS SFP cooling strategy provides SEP makeup water from the SNSWP, which is addressed in Section 2.3.4.9. The SNSWP is robust to the applicable external hazards and has sufficient inventory for long-term SEP cooling (among other demands).
2.4.4.4      FLEX Connections for SEP Makeup The CNS SEP cooling strategy includes primary and alternate connections for delivering water, which meets the requirements of NEI 12-06.
                *The primary strategy for SEP makeup relies on pressurizing the RN system using the FLEX low pressure pump. As discussed in Section 2.3.5.2, the B Train connection is located inside the RN Pumphouse, which is a Seismic Category I structure that is protected from all applicable hazards. The A Train connection provides a backup and is in the yard. RN and KE valves to be manually operated are located inside the Auxiliary Building, which is a Seismic Category I structure.
                *The alternate strategy also relies on pressurizing the RN system, but uses a jumper hose to connect to the KE system skimmer loop. The skimmer loop flows through a manifold line around the SEP with a series* of discharge points controlled by manual valves.
2.4.5          Ventilation CNS will vent the SEP to outside to minimize the impact of condensed steam on auxiliary building habitability. Fuel building air temperature and humidity will begin to increase shortly after event initiation, and the building may become uninhabitable within a few hours. To limit
        *pressure in the SEP building, selected doors will be opened within 6 hours. Additionally, CNS will take actions to prevent condensed steam from entering other parts of the Auxiliary Building outside of the selected vent path, such as securing doors and sealing door cracks.
2.4.6          Plant Instrumentation The key parameter for the SEP cooling/inventory function is SEP level. Instrumentation is capable of identifying the following SEP water level conditions, in accordance with NRC Order EA-12-051:
* Level that is adequate to support operation of the normal fuel pool cooling system.
* Level that is adequate to provide substantial radiation shielding for a person standing on the SEP operating deck.
* Level where fuel remains covered and action to implement makeup water addition should no longer be deferred.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 The SEP level instrumentation equipment was designed to perform its design function at temperature, humidity, and radiation levels consistent with SEP water at saturation conditions for an extended period following a BDBEE. Sensing components and cables located at the SFP were qualified to withstand peak and total integrated dose radiation levels for their installed locations based on post-event SEP water level equal to Level 3 for an extended period of time. (Level 3 is the water level where fuel remains covered but actions to implement make-up water addition should no longer be deferred based on the accuracy of the instrumentation and the highest point of any fuel rack.)
2.4.7       Thermal-Hydraulic Analysis CNS performed thermal-hydraulic analysis to address the SEP cooling/inventory function under the most limiting conditions and configuration. Key conclusions from the analysis include the following:
* The worst case heat load corresponding to a full core off load 6 days after a shutdown will produce evaporation rates of 96.8 gpm per SEP.
* Eor the worst case heat load scenario, the SEP will begin to boil after an ELAP/LUHS event in 8.8 hours. This scenario assumes a maximum starting SEP temperature of 125&deg;F and a full core offload during an outage after 6 days.
             *During normal operation (21 days after the beginning of a refueling outage and initial SEP temperature of 125&deg;F), SEP level can be maintained at least 10 feet above the top of the fuel (above which personnel access to the SEP operating deck is still viable) for 202.9 hours with no makeup. Eor normal operating conditions, boiling will begin in 37.0 hours.
* A ELEX low pressure pump, taking suction from the SNSWP and discharging by hose to connections on the RN system, can supply each of the SEPs at a rate significantly exceeding the most conservative steaming rate in addition to the all SGs at 50 gpm each.
In addition to the thermal-hydraulic analyses, the CNS SEP has been analyzed to remain subcritical with zero boron credited, i.e., a complete dilution scenario as long as the fuel assemblies remain covered and the parameters in CNS Technical Specification 4.3.1 remain in effect.
2.4.8       ELEX Pump and Water Supplies The CNS ELEX strategy relies on FLEX low pressure pumps to supply raw water from the SNSWP via RN piping. See Section 2.3.10.1 for description of the pumps.
As discussed in Section 2.4.7, the ELEX low pressure pumps have more than enough capacity to maintain SEP levels.
2.4.9       Electrical Analysis SEP level will be monitored by instrumentation installed to satisfy Order EA-12-051.
The instrumentation is normally powered from 120 V AC and upon an ELAP, the instrument will continue to operate upon switchover to a battery backup. The manufacturer-provided battery backup will last a minimum of 7 days from loss of off-site power, and those batteries can be replaced if necessary. SEP level instrumentation includes a backup channel that is powered from the vital inverters. FLEX strategies to restore power to vital battery chargers will ensure continued availability of the backup channel.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.5    Containment Function Strateoqy 2.5.1        Phase 1 The CNS Reactor Building includes a metal containment vessel and annulus region between the metal containment and a reinforced concrete enclosure. The containment vessel design pressure is 15 psig.
Following an ELAP/LUHS, the CNS Containment is initially cooled by an ice condenser.
Steam escaping the primary and/or secondary systems is cooled as it rises from lower containment through the ice condenser and into upper containment.
CNS performed a containment analysis, which demonstrated that containment pressure is expected to remain below the design pressure during Phase 1.
2.5.2       Phase 2 CNS will use the FLEX Electrical Distribution System to enable Phase 2 actions for the FLEX strategy to maintain Containment integrity.
                 *CNS will start a Hydrogen Skimmer Fan within 24 hours of event occurrence to limit the temperature increase in the SG and Pressurizer compartments. The SG and Pressurizer compartment temperature limits will be challenged before overall containment temperature limits. Abnormally high temperatures in these compartments could affect associated level indications due to reference leg flashing.
                 *Per NEI 12-06, Revision 0, plants with ice condenser containment designs such as*
CNS are required to repower hydrogen igniters to prevent buildup of hydrogen in case the ELAP event degrades to core damage. One train of hydrogen igniters will be re-powered and restored to service in Phase 2 using the back feed portable power strategy. The opposite train of hydrogen igniters can be repowered as an alternate.
* Manual containment isolation will also be Completed by the end of Phase 2.
2.5.3       Phase 3 Following deployment of an NSRC 480V generator and energizing of the 600V MO~s, CNS will start two Lower Containment Ventilation Units (LCVUs) within 48 hours to limit the temperature increase in the SG and Pressurizer compartments. Additionally, one Containment Air Return Fan (CARF) will be started within 52 hours of the event to establish an air flow path through the ice condenser, reduce containment pressure, and limit further heatup of the SG and pressurizer compartments.
With the CARE operating, SG and Pressurizer compartment temperature will be maintained below 200&deg;F and containment pressure will be less than 6 psig.
CNS wijl complete transition to RHR system cooling and cooldown to Mode 5 within 6 days of ELAP initiation to prevent challenging containment temperature and pressure limits following ice bed depletion.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.5.4        Availability of Structures. Systems. Components 2.5.4.1     Structures The FLEX strategy relies on site structures to provide protection for components, fluid and electrical connections, and deployment paths from applicable extreme external hazards. Specifically, the FLEX strategy for maintaining Containment integrity relies on the Reactor Building/Containment Vessel, along with MCCs located in the Auxiliary Building. The Reactor Building/Containment Vessel and Auxiliary Building are Seismic Category I structures that are designed to provide protection from the applicable extreme external hazards.
2.5.4.2     Components Inside Containment CNS relies on repowering a set of fans (Hydrogen Skimmer fans, LCVUs, and CAR~s) to maintain Containment temperature and pressure below acceptable limits. Hydrogen igniters are available to maintain hydrogen concentration below acceptable limits as defense in depth if the ELAP event degrades to core damage. All of these components are located inside the Reactor Building/Containment vessel, which is a Seismic Category I structure that protects equipment from external hazards.
2.5.4.3     Spray Strategqy Containment spray capability will not be required. An analysis was performed to validate that containment spray for temperature/pressure control is not required over the long term.
2.5.5        Plant Instrumentation The key parameter for the Containment integrity function is containment wide range pressure. Instrumentation will be powered by the FLEX Electrical Distribution System or a portable 120 VAC generator.
2.5.6        Thermal-Hydraulic Analysis CNS performed a thermal-hydraulic analysis to assess containment integrity using a GOTHIC model. Key conclusions from this analysis are as follows:
              *Containment pressure is expected to remain below the design pressure during the Phase 1 response.
               *The maximum pressure in containment will exceed the design pressure (15 psig) for a period of 12 hours (maximum of 17.5 psig). The Containment Vessel at CNS was initially pressure tested to 17.25 psig and has been analyzed to withstand a maximum of 72 psi. The temporary Containment pressure excursion beyond 15 psig during a BDBEE calculated by the GOTHIC model is therefore not significant.
              *The maximum temperature inside the SG and Pressurizer compartments will remain at or below the 2800&deg;F limit for the duration of the event, considering the planned ventilation actions that are part of the CNS FLEX strategy. The 280&deg;F limit was selected to ensure that reference leg flashing did not occur in the associated level indications.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2
                 *The annulus portion of the containment does not increase in pressure or temperature above design limits for the long term since it passively relieves through the annulus HVAC system exhaust dampers which fail open. The annulus calculation ensures the annulus pressure does not exceed the containment pressure and also ensures the annulus region can be accessed by personnel if necessary.
2,5.7       Electrical Analysis Containment pressure instrumentation will be powered by the FLEX Electrical Distribution System or a portable 120 VAC generator.
Hydrogen Skimmer fans used for containment ventilation and hydrogen igniters will be powered by the FLEX Electrical Distribution System. The LCVU fans and CARE will be powered from NSRC equipment. The repowering strategy is addressed in Section 2.3.11.1.
CNS performed an analysis to ensure that the 600V FLEX DGs had sufficient capacity to support the Phase 2 FLEX strategies. The analysis included electrical loads relevant for maintaining Containment integrity, such as battery chargers, hydrogen igniters, and the Hydrogen Skimmer fans. CNS performed comparable analyses for the 480V and 4160V NSRC DGs and the planned loads during Phase 3. This analysis concluded that the DGs and planned cable routing arrangement were adequate to support operation of the required equipment.
2.6    Characterization of External Hazards The following extreme external hazards were assessed for applicability for CNS:
* Seismic events
* External flooding
* Storms such as hurricanes, high winds, and tornadoes
* Extreme snow, ice, and cold
* Extreme heat 2.6.1.      Seismic Events The seismic hazard is applicable for CNS. Per NEI 12-06, Table 4-2, all sites will consider seismic events.
The CNS Updated Final Safety Analysis Report (UFSAR) states that the safe shutdown earthquake (SSE) has a ground acceleration design value of 0.15g acting horizontally and 0.10g acting vertically, and the operating basis earthquake (OBE) has a ground acceleration design value of 0.08g acting horizontally and 0.0533g acting vertically (FSAR, Section 3.1).
ONS is currently involved in the Expedited Seismic Evaluation Process (ESEP). The NRC has scheduled CNS to submit a Seismic Probabilistic Risk Assessment (SPRA) in 2019 to address the new Ground Motion Response Spectra (GMRS).
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.6.2      External Floodingq The external flooding hazard is applicable for CNS.
CNS is subject to external flooding from Probable Maximum Floods (PMFs) resulting from Probable Maximum Precipitation (PMP) events, Standard Project Floods (SPF) equal to 1/2 of the PMF, failures of upstream dams, a combination of dam failures and SPF's, seiche, hurricanes, and storm surge.
The ONS FLEX strategies reflect the site's current licensing basis. NTTF 2.3 flooding walkdowns and updated information on the flood hazard have resulted in a re-evaluation of the CNS flood hazard. CNS will assess potential impacts on the FLEX strategy resulting from this re-evaluation separately from this Final Integrated Plan (FIP).
CNS is susceptible to groundwater intrusion that may necessitate use of sump pumps to remove ground water from plant spaces.
2.6.3      Storms such as Hurricanes. Hi qh Winds, and Tornadoes The high wind hazard is applicable for CNS.
As described in UFSAR Section 2.1.1, the ONS site is located at latitude 35&deg;3'5" north and longitude 81&deg;4'10''. According to NEI 12-06, Revision 0, the location of CNS has a peak gust wind speed of 150-1 60 mph and a recommended tornado wind design speed of 175 mph. Based on the potential for winds in excess of 130 mph, the CNS site is susceptible to damage from severe winds from a hurricane or tornado.
2.6.4      Extreme snow, ice and cold The extreme cold (including snow and ice) hazard is applicable for ONS.
The CNS UFSAR contains environmental data from 1940 to 1980 obtained from the National Oceanic and Atmosphere Administration. The minimum temperature from the Catawba site was -5&deg;F. Because Catawba is in a temperate climate, storage and operation of the FLEX equipment in low temperatures is not considered to be an issue.
However, movement of FLEX equipment in snow and ice is a concern. CNS is located above the 35th parallel and is therefore subject to snowfall accumulation and extreme low temperatures per NEI 12-06, Revision 0. Based on NEI 12-06, the CNS site is also subject to the existence of an extreme amount of ice, and thus the potential for severe power line damage.
Based on UFSAR Section 2.4, water temperatures in the UHS will not reach a point where ice formation is an issue.    =
2.6.5      Extreme heat NEI 12-06, Revision 0 states that virtually every state in the lower 48 contiguous United States has experienced temperatures in excess of I110&deg;F and many in excess of 120&deg;F. In accordance with NEI 12-06, all sites will address high temperatures.
The CNS UFSAR contains environmental data from 1940 to 1980 obtained from the National Oceanic and Atmosphere Administration. The maximum temperature from the Catawba site was 104&deg;F. Because Catawba is in a temperate climate, storage and operation of the FLEX equipment in high temperatures is not considered to be an issue.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.7     Planned Protection of FLEX Equipment Storage and protection of FLEX equipment is discussed in this section. CNS evaluated the applicability of external hazards and addressed implementation considerations associated with each including:
* protection of FLEX equipment
* deployment of FLEX equipment
* procedural interfaces
* utilization of off-site resources CNS has one structure (i.e., the FLEX Storage Facility) for storing FLEX equipment that is located on-site and protects FLEX equipment from all applicable hazards. The FLEX Storage Facility contains sufficient equipment to satisfy redundancy requirements of NEI 12-06, Revision 0 for reliability and availability.
2.7.1         FLEX Storagqe Facility The FLEX Storage Facility at CNS is a single, large building located in the northwest area of the plant in a portion of the material lay-down storage area. This area is well above the flood level. The building is designed to resist seismic, wind forces, and tornado missiles of a magnitude that bounds all design basis hazards. The FLEX Storage Facility was seismically designed in accordance with the considerations presented in ASCE 7-10, and also meets the CNS SSE criteria.
The FLEX Storage Facility is a dome structure that has an outside diameter of 144 ft. It has two equipment doors on opposite sides of the building and two personnel access doors.
The building is located outside of the protected area approximately 500 feet west of the Shipping and Receiving Warehouse.
To ensure the dome storage area is functional, a monthly walkdown will be conducted in accordance with CNS procedures. Any issues identified as a result of the periodic walkdowns will be addressed in the site corrective action program.
FLEX equipment stored in the FLEX Storage Facility includes low pressure pumps, medium pressure pumps, high pressure pumps, 120V DGs, 600V DGs, hose trailers, a fuel transfer trailer, a CAT 924K, a pickup truck, portable spot coolers, 600V sump pumps, 120V sump pumps, portable transformers, portable panelboards, ventilation fans, and 230v sump pumps.
2.8    Planned Deployment of FLEX Equipment 2.8.1         Haul Paths and Accessibility CNS will use pre-defined deployment routes to transport FLEX equipment to the staging areas. The identified paths and deployment equipment positions will be accessible during all modes of operation. CNS will periodically perform walkdowns of the deployment paths to ensure pathways remain clear.
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ATTACH MENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1.& 2 Transmission lines can impede deployment of FLEX equipment along the pre-determined haul paths. In this case, CNS will ground the power lines and move them out of the deployment path. For other debris potentially impeding FLEX deployment, CNS has a CAT 924 Loader that can be used for debris clearing on the deployment path.
Once deployment paths are clear, CNS will dispatch deployment teams to transport Phase 2 FLEX equipment from the FLEX Storage Facility to deployment locations. One team will use the CAT 924 Loader and the other will use a Dodge 5500 truck. CNS has developed guidance on an appropriate sequence of deployment actions for each team.
The deployment routes for FLEX equipment begin with passage through the Vehicle Barrier Access Port (VBAP) and proceed along five primary on-site deployment paths as follows:
o  From the VBAP proceeding East along the southern bank of the SNSWP to the FLEX pump ramp access to the SNSWP.
o  From the SNSWP ramp through Gate #47 next to the ISFSI Area into the Protected Area and then South to the East side of the Auxiliary and Reactor Buildings. Pump discharge hoses may be deployed along this path for deployment strategies requiring a pump taking suction from the SNSWP and discharging through hoses to the suction of a second pump located outside the Auxiliary Building (e.g., for SG feedwater to support core cooling).
o  From the VBAP to the Vehicle Access Port (VAP) into the Protected Area proceeding East along the North side of Unit 2 connecting to the roadway along the east side of the Auxiliary Building.
oFrom the  VBAP proceeding directly south along the interior road circling the site on the south side of Unit 1 to the east side of the Auxiliary Building. Enter rear Gate #17  on the northeast side of plant into the Protected Area to the path along the east side of the Auxiliary Building and/or other established routes.
o  From the roadway on the East side of the Auxiliary Building inside the Protected Area, there are two paths that run west to FLEX equipment positions as follows:
* West to the north side of Unit 2 Diesel Generator Building
* West to the south side of Unit 1 Diesel Generator Building The deployment paths were evaluated for seismic stability and liquefaction and determined to be acceptable. A flooding event may inundate portions of the site, but flood waters will recede from all deployment paths and staging areas in time for FLEX deployment to meet all time constraints. Any changes to the CLB resulting from the Flood Hazard Re-evaluation will be addressed separate from this FIP. =...
CNS has instituted administrative measures to ensure that deployment paths and FLEX equipment deployment sites (including paths for portable cables, hoses, etc.) are maintained clear of other equipment or interferences. CNS performs general rounds of various plant areas on a daily shift basis and will identify FLEX interferences. Additionally, CNS will periodically perform specific inspections of FLEX-related areas and ensure they are maintained clear.
Page 28 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.8.2        Deployment of Strategqies CNS identified pre-determined deployment and staging locations for FLEX equipment necessary to support the FLEX strategies as discussed in the subsections below.
Human performance aids such as labels will be attached to specific portable FLEX equipment to identify the equipment with a unique number that agrees with FSGs.
2.8.2.1     Core Coolinq Makeup water for the CA system to the SGs via the TDAFWP will be supplied by the FLEX low pressure pump, which will be staged on a ramp at the edge of the SNSWP.
Suction hoses with strainers will be attached to the pump and placed into the SNSWP.
Water temperature in the SNSWP will not decrease to the point where access will be challenged due to ice formation.
SG feedwater will be delivered by a FLEX medium pressure pump that draws suction from the discharge of the low pressure pump. The Unit I medium pressure pump is normally deployed south of the Unit I Exterior Doghouse door. The Unit 2 medium pressure pump is normally deployed north of the Unit 2 FWST and west of the Unit 2 Containment Mechanical Equipment Building (CMEB). Discharge hoses from the FLEX medium pressure pump will be routed to the CA piping connections located in the Exterior and Interior Doghouses.
For Phase 3 deployment, procedures and FSGs were developed to connect NSRC equipment to station equipment. Use of portable pumps to supply feedwater to the SGs remains the same as Phase 2 until the RHR system is restored to service with supplemental power from the NSRC DGs or off-site power is restored.
2.8.2.2     RCS Makeup Two FLEX high pressure pumps (one for each unit) with hoses/adapters will be deployed to support RCS boration and inventory control. For Unit 1, the normal staging location for the FLEX high pressure pump is south and west of door AX658A (Unit 1 Electrical Penetration Room). For Unit 2, the normal staging location for the FLEX high pressure pump is north of door AX656B (Unit 2 Electrical Penetration Room). Suction hoses will be deployed to connect the portable pump to KF system connections from the FWST. A single FWST will supply borated water to the portable pumps for both units using a gated wye connection. Discharge hoses from the FLEX high pressure pumps will be connected to the discharge piping of NI pumps in the Auxiliary Building to supply injection makeup to the RCS.
2.8.2.3    FLEX Electrical Distribution System Two 600V portable DGs and cabling will be deployed to an area near the FWSTs of each Unit (one generator per Unit). The normal staging area for the Unit 1 FLEX 600V DG is east of the Unit 1 Turbine Building, in close proximity to the Unit 1 MG set room door. The normal staging area for the Unit 2 FLEX 600 V DG is east of the Unit 2 Turbine Building, in close proximity to the Unit 2 MG set room door. Alternate staging areas are near the waste shipping/auxiliary access point area doors, the hot machine shop area doors, and the Unit I RC Pit / Safe Shutdown Facility Area.
The portable 600V DGS will be connected to the normal plant MCCs for re-powering equipment via the FLEX Backbone using a combination of permanently installed cables, portable panelboards, and transformers. Cables and connectors of the FLEX 600V distribution system are color coded to ensure proper phase rotation.
Page 29 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 During Phase 3, DGs from the NSRC will be available to provide additional power. For Unit 1, the NSRC 41 60V DGs and switchgear will be staged south of the Diesel Generator Building and the NSRC 480VI DGs will be staged south and west of the Auxiliary Building, near the FLEX high pressure pump. For Unit 2, the NSRC 4160V DGs and switchgear and the NSRC 480V DGs will be staged near the west side of the Diesel Generator Building.
2.8.2.4      Sump Pumps CNS can deploy FLEX sump pumps to ensure that equipment for core cooling is not compromised by flooding. The normal FLEX Sump Pump staging area is at the rear of the Auxiliary Building near the Clean Trash Room. Alternate staging areas include the area outside the Unit 1 MG set room and the area outside the Unit 2 MG set room.
2.8.2.5      Small Diesel Generators Eleven portable DGs (6 kW) are available to be deployed from the FLEX Storage Facility to the location where they are needed. These 6 kW DGs will be used to power battery chargers for hand-held radios and portable satellite phones, fans, small sump pumps, and other identified loads.
2.8.3         Fuelina of EauiDment Diesel Fuel Oil (DFO) for the FLEX equipment will be obtained from the safety-related, underground Emergency Diesel Generator (EDG) fuel oil storage tanks. The diesel fuel will be pumped out of the underground tanks and transferred to the portable Diesel Fuel Storage Tank using a diesel driven Portable Transfer Pump or will be gravity drained into small containers in the EDG room and hand carried to the location needed. Connections on the piping from the EDG tanks that will be used for removing fuel oil are seismically qualified and located above the flood level in the yard. The portable fuel tank will be pulled by the FLEX Pickup Truck or other available vehicles to the various site locations for refueling portable diesel generators.
CNS performed a fuel consumption calculation to determine the estimated total fuel required to support FLEX equipment operation. The estimated fuel required to support FLEX equipment for 24 hours is 2,027 gallons. FLEX equipment will be stored full of DFO and refueled as necessary. The CNS fuel consumption calculation determined that 485 gallons of DFO would be required for refueling in the first 24 hours, which is well within the capacity of the EDO DFO storage tanks.
To avoid multiple pieces of FLEX equipment running out of fuel at the same time, CNS will attempt to maintain individual fuel tanks greater than half full. The small DGs may require refueling prior to the FLEX Diesel Fuel Transfer Trailer being available. If necessary, CNS will dispatch personnel to transport required fuel by hand.
During Phase 3, additional diesel fuel will be brought onsite from outside resources as required.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.9Seuence* of Events* and Staffingq 2.9.1      Sequence of Events The Table below presents a Sequence of Events (SOE=) Timeline for an ELAP/LUHS event at CNS. CNS Simulator Observation was conducted to time-validate EOP and FSG actions.
For other actions, time-validations involved combinations of local FSG time-validations, local (EOP) action time-validations, timed table-top procedure readings, and timed repetitive actions, such as cable and hose deployments. The analyses were conducted in incremental fashion, typically beginning with an assumed time of EOP or FSG implementation, and progressing through completion of the action or procedure. The analyses utilized ratio analysis to project times to complete actions such as deployment of multiple sections of cabling, hose, couplings, etc. Elements documented sequentially included observed and projected times to complete actions, time for procedural progression, decision making, necessary, briefings for implementers, and time required to perform manual deployments and repetitive actions, such as walking to required destinations, towing equipment to required destinations, carrying and deploying cable and hose segments, etc.
All FSGs for FLEX strategies were validated per normal procedural requirements to ensure hands-on equipment/component verification and the mechanics of executing procedure steps were practical and accurate.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 &.2
                                                        ~~Sequence of Events Timeline                                 ,
Ato                   Str ieCompletion trTie        Time                                    'Remarks / APplicability Acin(hours)                          (Hours),
Event Starts                                                0         N/A         Plant @100% power TDAFWP starts on SBO                                        0         N/A         Installed plant equipment. Automatic Action.
TDAFWP automatically aligns to embedded RC                                        Action occurs if normal water sources are damaged. Normal water sources header piping if normal water sources are                   0          48         are not robust to all hazards, so conservative timeline aligns to RC damaged.                                                                          immediately. Embedded RC system has at least 48 hours of inventory.
Control SG level by starting / stopping TDAFWP with Trip and Throttle Valve from                   0.5     As needed      Maintain SG level. Control board manipulation.
Control Room.
Initial cooldown and depressurization of RCS                                      Establish conditions for reactor makeup, reduce seal leakage, preserve (to a temperature and pressure that does not               0.6         2.6        sel.Isaedpntqumn.CorlBadmnilto.
result in nitrogen injection)ses.IsaldpnteipetCorlBarmnputo.
Isolteontanmet nstrmen Airto                1            .~Time          constraint (38 hr). Analysis shows containment pressure, remains below design for the first 38 hours even if Instrument Air is not isolated.
Dateiscnectalnnciiallasfosia                               2.5        2.75        Time constraint (3.5 hr). Required to preserve Vital Batteries.
Time constraint (6 hr). Provides SEP vent path to outside of building to OpenSEPbay doo ersnne                     3           .04       avoid adverse Auxiliary Building conditions due to SFP steam.
Provide pumping capacity to control level in              36           .6         Tm      osrit(        r TDAFWP put sump Align charging to Channel A and D Vital                     61.5                  Tm      osrit(15h          o   ,1. rfrD     4h o          )
Batteries (B and C included)                                61.5                  Tm      osrit(15h          o    ,1. rfrD     4h o           )
Align portable injection pump from FWST to NI System to provide RCS makeup and                        6.5        11.6        Time constraint (13.8 hr) boration.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2
* Sequence of Events Timeline Action..                  Start Time       Time                                  "RemarksIApplicability        "...          *     ,
Action     *         ~    ~(hours)    (Hours)....                                                                                r Ample handheld/headlamp lighting would be available at 0 hr with outside Provide portable lighting (beyond head and               8+         As needed    resources available to provide additional lighting and replacement batteries hand lamps and installed battery lighting)                                       after 6 hr. Portable lighting would be placed into service as time and resources allow.
Re-power H2 igniters.                                     9            9.35      This action is required by NEI 12-06 as a contingency.
Power A and B WZ (Groundwater Drainage)                   101.5                  Time constraint (10.6 hr). Prevents ground water from flooding Aux Bldg El.
Sump Pumps                                                101.5                  543' and below.
Install portable fans in Control Room and113.3                                   Tmcosrit(0h)
Battery Rooms Rehag cmuncain yte     adsaelie12                        14       Time constraint (24 hr). Portable DG will be used to recharge batteries.
phone system Install 600 V sump pump on the 522' elevation             12           16.6      Time constraint (36 hr)
Fan            2023.32                 Time constraint (24 hr). Analysis of containment heat loads is basis for Start Hydrogen Skimmer                                                           these2actions Connect FLEX low pressure pump through RN piping to SEP skimmer loop to provide a               243.3                   Time constraint (40 hr). Critical pool level assuming worst case heat means to make up to the SEP without entering             243.3                   load/evaporation rate is reached after 40 hours.
the SEP area.
Install NSRC DGs.                                      24 -48       As needed Isolate the CLAs.                                        42            44.1      Required prior to depressurizing the SGs below 240 psig to prevent gas intrusion to ROS that would disrupt natural recirculation cooling.
Evaluate need to provide freeze protection for instrumentation located in Doghouses and                 48         As needed    Time constraint (48 hr) yard.
Start Lower Containment Ventilation Units                48             48       Time constraint (48 hr)
Start Containment Return Fan                             48             52       Time constraint (52 hr)
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ATTACH MENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.9.2        Staffingq Using the methodology of NEI 12-01, Guideline for Assessing Beyond Design Basis Accident Response Staffing and Communications Capabilities,an assessment of the capability of the on-shift staff and augmented ERO to respond to a BDBEE was performed; The assumptions for the NEI 12-01 Phase 2 scenario postulate that the BDBEE involves a large-scale external event that results in:
* An extended loss of AC power
* An extended loss of access to ultimate heat sink
* Impact on units (all units are in operation at the time of the event)
* Impeded access to the units by off-site responders as follows:
o   0 to 6 Hours Post Event   -  No site access.
o   6 to 24 Hours Post Event - Limited site access. Individuals may access the site by walking, personal vehicle or via alternate transportation capabilities (e.g.,
private resource providers or public sector support).
o   24+ Hours Post Event    -  Improved site access. Site access is restored to a near-normal status and/or augmented transportation resources are available to deliver equipment, supplies and large numbers of personnel.
The on-shift ERO analysis concluded that current CNS on-shift staffing present for the "no site access" 6-hour time period is sufficient to perform the EOP, FSG, and emergency response tasks.
The augmented ERO analysis concluded that sufficient personnel resources exist in the current CNS augmented ERO to fill positions for the expanded ERO functions. Thus, ERO resources and capabilities necessary to implement coping strategies after 6 hours exist in the current program.
To conduct the on-shift portion of the assessment, a team of subject matter experts from Operations, Maintenance, Security, Radiation Protection, Chemistry, Engineering, Emergency Preparedness, and industry consultants conducted tabletop exercises. The participants reviewed the assumptions and existing procedural guidance, including applicable draft FSGs for coping with a BDBEE using minimum on-shift staffing. Particular attention was given to the sequence and timing of each procedural step, jits duration, and the on-shift individual performing the step to account for both the task and time motion analyses of NEI 10-05, Assessment of On-Shift Emergency Response Organization Staffing and CaPabilities.
In the event of a natural or any other type disaster that renders roads and passage ways unusable, CNS essential staff will be airlifted from two preplanned landing zones. The northern landing zone would be the Duke Energy Shelby Operations Center. The southwest Landing zone would be the Duke Energy Chester Operations Center. The two landing zones are Duke Energy properties and can provide shelter, communications and storage capacity for necessary gear. The helicopter provider will be under a contract or agreement letter to provide support.
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ATTACH MENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.10    Offsite Resources The Strategic Alliance for FLEX Emergency Response (SAFER) team is contracted by the nuclear industry through Pooled Equipment Inventory Corporation (PEICo) to establish National SAFER Response Centers (NSRC) operated by Pooled Inventory Management (PIM) and in collaboration with AREVA to purchase, store, maintain and deliver emergency response equipment in the case of a major nuclear accident or BDBEE in the United States.
The NRC letter dated September 26, 2014 (ADAMS Ascension No. MLI 4265A1 07) titled "Staff Assessment of National SAFER Response Centers Established in Response to Order EA         049" (Reference 12) endorsed NEI's White Paper titled "National SAFER Response Centers" (Reference 13). NRC concluded that SAFER procured equipment, implemented appropriate processes to maintain the equipment, and developed plans to deliver the equipment needed to support site responses to BDBEEs, consistent with NEI 12-06 guidance and the SAFER Response Plan to meet Phase 3 requirements of Order EA-12-049.
CNS relies on equipment stored off-site for Phase :3 of the FLEX strategy. Equipment may be provided from NSRCs. Another nuclear plant may also provide Phase 3 equipment, if response would be faster than from the NSRCs.
2.10.1       National SAFER Response Center (NSRC)
The SAFER Response Plan for CNS (Reference 14) contains (1) SAFER control center procedures, (2) National SAFER Response Center procedures, (3) logistics and transportation procedures, (4) staging area procedures, which includes travel routes between staging areas to the site, (5) guidance for site interface procedure development, and (6) a listing of site-specific equipment (generic and non-generic) to be deployed for FLEX Phase 3.
Two NSRC's are strategically located across the country in Memphis, TN and Phoenix, AZ.
If possible, NSRC equipment would be delivered to CNS by trucks from the Memphis location.
If possible, NSRC equipment will be delivered to Staging Area C, which is the Kings Mountain Training Center (34 miles away from the CNS site by driving). When CNS is ready, NSRC equipment will then be delivered to Staging Area B, which is an on-site parking lot at the CNS site outside the protected area. NSRC Equipment can also be delivered directly to Staging Area B, if requested by CNS.
Primary and alternate driving routes from Staging Area C to Staging Area B have been identified. CNS will coordinate with local and state authorities to assess the condition of roads and bridges along the travel path. If ground transportation from Staging Area C to Staging Area B is not feasible, NSRC equipment can be delivered to Staging Area B by helicopter airlift.
Two access routes from Concord Road to Staging Area B have been identified: the primary access location is through the Main Entrance, Northwest Gate; the secondary access path is through the Main Entrance, Southwest Gate.
The SAFER Response Plan for CNS does not include a Staging Area D.
The first arriving equipment will be delivered to the site within 24 hours from initial contact and remaining equipment will be delivered within 72 hours from initial contact.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.10.2       Equipment The NSRC will provide equipment as listed in the response plan. The NSRC will deliver the first pieces of equipment within 24 hours from initial contact. Such priority equipment includes Medium Voltage Generators, a Low Voltage Generator, 4160 VAC Distribution System, a Low Voltage Step-Up Transformer, a Low Pressure / High Flow Dewatering Pump, Mobile Lighting Towers, a Diesel Fuel Transfer Pump, a Water Treatment System, and a Mobile Boration Unit. The generic set of NSRC equipment as identified in the plan provides back up to on-site FLEX equipment (e.g., pumps, DGs) and will be provided as lower priority items to arrive within 72 hours from initial contact. NSRC equipment connections to applicable hoses and/or plant equipment are compatible or necessary adapters are available.
Other offsite resources may be obtained as needed to support the event which may include diesel fuel oil, equipment from other nuclear plants, and equipment from vendors.
2.11    Habitability and Operations 2.11.1       Equipment Coolincq CNS performed HVAC analysis to evaluate temperature increases resulting from an ELAP /
LUHS in the control room, TDAFWP room, and other rooms of the Auxiliary Building.
Considering the actions planned to provide ventilation as part of the FLEX strategies (discussed below), these analyses concluded that temperatures would not challenge essential equipment.
To enhance MCR temperature margins and personnel comfort, various doors will be opened to provide an air exchange flow path with the outside. Additional cooling will be provided by four portable, water-cooled coolers powered from the FLEX Electrical Distribution System.
The cooling water supply will be provided by a branch from the FLEX Raw Water Distribution System.
Analyses validated TDAFWP room temperatures remain acceptable with no additional action.
Portable fans will be placed in the vital battery area during battery charging to ensure hydrogen accumulation does not exceed flammability limits. Additional ventilation needs, such as fans, will be provided from the NSRC and used if required.
2.11.2       Freeze Protection If freezing weather exists, CNS will take appropriate actions to ensure protection of the FLEX strategies from cold conditions. As necessary, CNS will start FLEX diesel-powered equipment early due to the potential for difficult starting and CNS will establish trickle drains from idle equipment and hoses to prevent formation of ice plugs.
CNS can also provide heat tracing for equipment in the FLEX Storage Facility, FWST level instrumentation, and pressure instruments in the Doghouses. Other than components associated with monitoring FWST level, there is no FLEX-credited instrumentation in the yard or exposed to cold temperature to the extent that freezing would be expected. Heat tracing equipment may be deployed from the FLEX Storage Building to maintain the FWST level instrumentation functional. Heat tracing may also be deployed for other instrumentation as a contingency. Heat tracing cables could be powered by any portable DG supplying 120V power.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 2.11.3        Hvdroaqen Ventilation The minimum concentration of hydrogen gas to result in an explosive mixture is 4%.
Without mitigation, CNS determined that hydrogen concentrations in the battery rooms would not exceed 4% (i.e., the lower explosive limit) for at least 21 days. Regardless, CNS plans to deploy fans in the battery room areas as part of the FLEX strategy as time permits.
Battery rooms are ventilated by pulling outside air through the U2 MG set room, down the Unit 2 Auxiliary Building stairwell, through the battery rooms, and up the Service Building stairwell to outdoors through an open door, creating a chimney effect. CNS will open a series of doors to create this air flow path, all of which will be opened within 24 hours.
2.11.4        Personnel Habitability CNS will set up spot coolers in the control room to maintain temperature less than 90&deg;F.
Cooling water will be provided to the spot coolers using 5-inch raw water distribution hose, the fire protection header, or the "B" train service water system. Coolers will start operating when the first source of water becomes available (most likely the 5-inch hose or the fire protection header).
For providing water from the 5-inch FLEX hose or the fire protection header, CNS will route hosing through door AX(657 (Unit 2 Penetration Room) and AX657F (Control Room access to Unit 2 Auxiliary Building Corridor at 594'). For providing water from the "B" train RN system, CNS will route hose through AX(657G (Control Room rear access). Cooling water that has been used by the room coolers will be routed through a water discharge manifold to a yard drain. CNS will also deploy a condensate drain manifold and hosing to route condensation from the spot coolers to a yard drain.
2.12        Water Sources Discussion of credited water sources for the FLEX strategies is included in the sections above for each individual strategy.
2.12.1        SG Feedwater For SG feedwater, CNS will provide water from any of the following sources:
* USTs
* CACST
* Hotwell
* RC system piping embedded volume
* SNSWP The embedded RC system captured volume and the SNSWP are the credited sources of water because of their robustness to the applicable hazards. The USTs and the hotwell are not protected from external hazards. These tanks are normally aligned as a TDAFWP suction source, but automatic realignment to embedded RC system captured volume is provided if the default sources are not available.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 The UST and hotwell have condensate grade water that will not degrade the SGs. CNS may switch to a raw water source for a limited duration, if necessary. In this case, water purification equipment from the NSRC will eventually be deployed to establish a clean water source prior to excessive exposure of the SGs to raw water. ONS will also consider actions to clean up the SGs including blowdowns or chemical additions.
2.12.2        RCS Makeup For NC system boration during Phase 2, CNS will provide borated water from one or both of the following sources:
* FWSTs
* CLAs Alternate sources of borated water include the following options:
* NSRC-supplied mobile boration skid
* Trucking from an off-site source (e.g., McGuire Nuclear Station)
* Recovery of borated water from the FWST annulus (if damaged)
* Portable FLEX drop tanks mixing boron and raw water 2.12.3        SEP Makeup For inventory control of the SFP, CNS uses raw water from the RN system, which is pressurized by the FLEX low pressure pump using the SNSWP as the suction source. The SNSWP will be available following the applicable extreme external hazards.
During Phase 3, CNS may transition to a clean water source (e.g., NSRC-supplied water purification unit) when available.
2.13    Shutdown and Refuelingq Analysis Order EA-12-049 requires that licensees must be capable of implementing the mitigation strategies in all modes. In general, the discussion above focuses on a BDBEE occurring during power operations. This is appropriate, as plants typically operate at power for 90 percent or more of the year. If the BDBEE occurs with the plant at power, the mitigation strategy initially focuses on the use of a pump coupled to a steam-powered turbine to provide the water initially needed for decay heat removal. If all or most of the fuel has been placed in the SFP, there is a shorter timeline to implement the strategy for providing SEP makeup water. However, this is balanced by the fact that if immediate cooling is not required for the fuel in the reactor vessel, the operators can concentrate on providing makeup to the SEP and the number of personnel on-site is much greater during an outage. ONS analysis shows that following a full core offload to the SFP, about 76 hours are available to implement makeup before boil-off results in the water level in the SFP dropping far enough to uncover fuel assemblies. About 48 hours are available prior to level decreasing to 10 feet above the top of the fuel assemblies. As previously discussed, CNS can provide SEP makeup in advance of these times.
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ATTACHMENT 6 FINAL INTEGRATED PLAN catawba Nuclear Station, Units 1 & 2 When a plant is in a shutdown mode and steam is not available to operate the steam-powered pump, another strategy must be used for decay heat removal. On September 18, 2013, NEI submitted to the NRC a position paper entitled "Shutdown Refueling Modes" (Reference 15),
which described methods to ensure plant safety in shutdown modes. By letter dated September 30, 2013 (Reference 16), the NRC staff endorsed this position paper as a means of meeting the requirements of the Order. In the third six-month update dated August 28, 2014 (Reference 17),
CNS committed to follow the guidance in this position paper.
CNS's FLEX strategy in Modes 5 and 6 relies on a FLEX Medium Pressure Pump to provide borated makeup to the RCS. This pump will be staged locally at one of four potential locations around the Auxiliary Building. The suction supply for the portable pump will come from a connection on the FWST supply line for KF/FW makeup to the SEP (see Section 2.3.5.3). The discharge from the portable pump will be into the Train A NI pump discharge piping that feeds the RCS hot or cold legs (alternate connection on the Train B NI pump discharge piping)
(see Section 2.3.5.4).
If the reactor vessel head is installed, RCS system depressurization will be initiated through the reactor vessel head vents.
FWST inventory may only last 40 hours if RCS feed and bleed cooling is utilized. FWST makeup, use of alternate borated water strategies, or RHR pump restoration will be provided prior to loss of FWST level.
For the containment integrity function, CNS would open the equipment hatch to establish an emergency vent path from containment, if possible. The equipment hatch would only be opened if core cooling is expected to be maintained.
2.14    Procedures and Training 2.14.1         Procedural Guidance The inability to predict actual plant conditions that require the use of BDBEE equipment makes it impossible to provide specific procedural guidance. As such, the FSGs provide guidance that can be employed for a variety of conditions. FSGs, to the extent possible, provide pre-planned FLEX strategies for accomplishing specific tasks in support of EOPs and Abnormal Operating Procedures (AOPs). FSGs are used to supplement (not replace) the existing procedure structure that establishes command and control for the event.
Procedural interfaces were incorporated into ECA-0.0, "Loss of All AC Power" to the extent necessary to include appropriate reference to FSGs and provide command and control for the ELAP.
2.14.2        Trainingq Programs and controls were established to assure personnel proficiency in the mitigation of BDBEE is developed and maintained. The Systematic Approach to Training (SAT) process was utilized to evaluate, develop and implement training for applicable personnel.
Initial training was provided and continuing periodic training will be provided to site emergency response leaders on BDBEEs, emergency response strategies, and implementing guidelines. Personnel assigned to direct the execution of mitigation strategies for BDBEEs received the necessary training to ensure familiarity with the associated tasks,
            .considering available job aids, instructions, and mitigating strategy time constraints.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 &.2 Care was taken to not give undue weight (in comparison with other training requirements) for operator training for BDBEE accident mitigation. The testing/evaluation of operator knowledge and skills in this area was similarly weighted.
A table top analysis was conducted by the Duke fleet ERO trainers using a graded Systematic Approach to Training. The analysis determined no impact on the existing ERO position specific Job Task Analyses (JTAs). Staffing resources for the expanded response capability ERO are provided from a qualified depth of ERO personnel. ERO depth is outlined and maintained per NSD1 17 (ERO Staffing Training and Responsibilities.) ETQS 3100 section 5.2 states, "Job Task Analysis (JTA) should be completed when no previous job data exist, is incomplete, or when revising a job." This analysis referenced the Duke Energy NEI 12-01, Phase 2 staffing assessment report. As a result of the analysis, no new skills, knowledge or tasks, were identified since current qualified ERO resources will be augmented for expanded response capability.
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2
: 3. Acronyms AOP - Abnormal Operating Procedure ATWS - Anticipated Transient Without Scram AWWA - American Water Works Association BAT - Boric Acid Tank BOB - Beyond Design Basis BDBEE - Beyond Design Basis External Event BW - Steam Generator Wet Layup Recirculation System CA - Auxiliary Feedwater System CACST - Auxiliary Feedwater Condensate Storage Tank CARF - Containment Air Return Fan CFR - Code of Federal Regulations CLA - Cold Leg Accumulator CMEB - Containment Mechanical Equipment Building CNS - Catawba Nuclear Station DG - Diesel Generator EDG - Emergency Diesel Generator ELAP - Extended Loss of AC Power EOP - Emergency Operating Procedure EPRI - Electric Power Research Institute ERO - Emergency Response Organization ESEP - Expedited Seismic Evaluation Process FCV - Flow Control Valve FIP - Final Integrated Plan FLEX - Diverse Flexible Coping Strategies FSG - FLEX Support Guideline FW - Refueling Water System FWST - Refueling Water Storage Tank GMRS - Ground Motion Response Spectra HCPLF - High Confidence of Low Probability of Failure HVAC - Heating Ventilation and Air Conditioning JTA - Job Task Analysis KC - Component Cooling System KF - Spent Fuel Pool Cooling System LCVU - Lower Containment Ventilation Unit LOOP - Loss of Offsite Power LUHS - Loss of Ultimate Heat Sink MCR - Main Control Room MG - Motor Generator NEI - Nuclear Energy Institute NI - Safety Injection System NR - Narrow Range NRC - Nuclear Regulatory Commission Page 41 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2 NSRC - National SAFER Response Center NTTF - Near-Term Task Force OBRE - Operating Basis Earthquake PEiCo - Pooled Equipment Inventory Corporation PIM - Pooled Inventory Management PMF - Probable Maximum Flood PMP - Probable Maximum Precipitation PORV - Power Operated Relief Valve RC - Condenser Circulating Water System RCP - Reactor Coolant Pump RCS - Reactor Coolant System RHR - Residual Heat Removal System RN - Nuclear Service Water System SA - Main Steam to Auxiliary Equipment System SAFER - Strategic Alliance for FLEX Emergency Response SAT - Systematic Approach to Training SBO - Station Blackout SQE - Sequence of Events SG - Steam Generator SFP - Spent Fuel Pool SNSWP - Standby Nuclear Service Water Pond SQUG - Seismic Qualification Utility Group SFF - Standard Project Flood SPRA - Seismic Probabilistic Risk Assessment SSE - Safe Shutdown Earthquake SV - Main Steam Vent to Atmosphere System TDAFWP - Turbine Driven Auxiliary Feedwater Pump TIA - Task Interface Agreement TS - Technical Specifications UFSAR - Updated Final Safety Analysis Report UHS - Ultimate Heat Sink UST - Upper Surge Tank WR - Wide Range Page 42 of 44

ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2
: 4. References
: 1. Recommendations for Enhancing Reactor Safety in the 21 st Century; The Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, July 12, 2011
: 2. NRC Order EA-12-049, Issuance of Order to Modify Licenses with regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, March 12, 2012. (ML12054A735)
: 3. NEI 12-06, Rev. 0, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, August 2012.
: 4. NRC Interim Staff Guidance JLD-ISG-2012-01, Compliance with Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events. (MLI12229A1 74)
: 5. Duke Energy (Henderson) letter CNS-1 5-035 dated May 1, 2015, "Notification of Full Compliance with Order EA-12-049, 'Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events' and with Order EA-1 2-051, 'Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation' - Catawba Nuclear Station Unit 2." (ML15126A277 is NRC ADAMS Database)
: 6. Duke Energy (Henderson) letter CNS-16-005, "Final Notification of Full Compliance with Order EA-12-049, 'Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events' and with Order EA-1 2-051, 'Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation' for Catawba Nuclear Station."
: 7. NRC Order EA-1 2-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation.
: 8. NEI 12-02, Rev. 1, Industry Guidance for Compliance with NRC Order EA-1 2-051 to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, August 2012.
: 9. NRC Interim Staff Guidance JLD-ISG-2012-03, Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation.
: 10. NRC letter dated September 12, 2006, "Final Response to Task Interface Agreement (TIA) 2004-04, 'Acceptability of Proceduralized Departures from Technical Specifics (TSs) Requirements at the Surry Power Station,' (TAO NOs. MC4331 and MC4332)." (ML060590273 in NRC ADAMS Database)
: 11. NRC (Davis) letter to PWROG (Stringfellow), dated January 8, 2014. (ML13276A183 in NRC ADAMS database)
: 12. NRC (Davis) letter to NEI (Pollock), dated September 26, 2014, "Staff Assessment of National SAFER Response Centers Established in Response to Order EA-12-049." (ML14265A107 in NRC ADAMS Database)
: 13. NEI (Pollock) letter to NRC (Davis), dated September 11, 2014, "National SAFER Response Center Operational Status," with Enclosure "White Paper; National SAFER Response Centers."
(ML14259A222 & ML14259A2.23 in NRC ADAMS Database)
: 14. Areva, Inc. Engineering Information Record 51-9233066-0002, "Catawba SAFER Response Plan."
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ATTACHMENT 6 FINAL INTEGRATED PLAN Catawba Nuclear Station, Units 1 & 2
: 15. NE! Position Paper, "Shutdown / Refueling Modes", Rev. 0, dated September 18, 2013.
(ML13273A514 in NRC ADAMS Database)
: 16. NRC (Davis) letter to NEi (Pollock), dated September 30, 2013. (ML13267A382 in NRC ADAMS Database)
: 17. Duke Energy letter CNS-14-086 dated August 28, 2014, "Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)." (ML14247A232 in NRC ADAMS database)
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ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT
: 1.      Background and Purpose NRC Order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," required licensees to develop, implement, and maintain guidance and strategies to maintain or restore core cooling, containment and spent fuel pool cooling capabilities following a beyond-design-basis external event. To develop strategies for maintaining/restoring core cooling, licensees evaluated reactor coolant system (RCS) leakage from reactor coolant pump (RCP) seals during an extended loss of all AC power (ELAP).
NSAL-1 4-1, Revision 1 was issued by Westinghouse on September 8, 2014 and it documents that the nominal RCP seal leakage rate of 21 gallons per minute (gpm), as documented in WCAP-1 0541, Revision 2, may be not be applicable for all plants using Westinghouse RCPs with standard seal designs because of the various thermal-hydraulic conditions set up by plant-specific seal leak-off piping designs.
PWROG-14015-P, Revision 2 was issued by the PWR Owner's Group in April 2015 to determine revised No. 1 RCP seal leak-off flow rates following an ELAP.
PWROG-1 4027-P, Revision 3 was issued by the PWR Owner's Group in April 2015 to evaluate the time to enter reflux cooling and the time at which the core uncovers based on the revised seal leak-off flow rates during an ELAP.
Following issuance of the Watts Bar Mitigating Strategies Safety Evaluation dated March 27, 2015, the NRC requested via e-mail dated March 31, 2015 that licensees with standard Westinghouse RCP seal packages review the technical content therein and provide information addressing similar issues. Specifically, the NRC communication stated:
"At the present time the NRC staff is unable to conclude that Westinghouse's analytical modeling of ROP seal leakage is acceptable on its own merits. However, for the purposes of mitigating strategies, the staff can balance the modeling uncertainties and deficiencies of the model with the unique aspect of FLEX. To expedite individual plant resolution, licensees could provide a brief discussion about the margin for RCS makeup time, based on the favorable aspects of individual site mitigating strategies."
In addition,* the NRC provided examples of pertinent information to include in the Margin Assessment.
The purpose of this Margin Assessment is to provide a discussion regarding the margin for RCS makeup time, specifically addressing the examples of pertinent information regarding seal leakage provided by the NRC.
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J ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT
: 2.        RCP Seal Leak-Off Line Configuration Catawba is a four-loop Westinghouse PWR utilizing Model 93A reactor coolant pumps, using standard Westinghouse seal packages. The Catawba RCS loops utilize inverted U-tube type steam generators. Catawba's Mitigating Strategies (FLEX) response is based on the established RCP seal leakage profile as identified in WCAP-1 7601-F, revision 1 "Reactor Coolant System Response to the Extended Loss of AC Power Event for Westinghouse, Combustion Engineering and Babcock & Wilcox NSSS Designs".
In late 2014 and early 2015, as a result of Westinghouse NSAL-1 4-1, Catawba contracted with MPR Associates to have the existing RCP No. 1 and No. 2 seals and the associated No. 1 seal leak-off piping evaluated for an extended loss of seal cooling event, such as an ELAP. Due to time constraints, this effort was performed in parallel with the on-going PWROG initiative to resolve issues associated with the established RCP seal leak-off rates during a Loss Of Seal Cooling (LOSC) event. The MPR RCP seal model is different from the Westinghouse seal model being used in the PWROG work, in that the MPR model accommodates a transient analysis for evaluation of known pressure spikes during the early stages of the LOSC event.
The Westinghouse RCP seal model does not currently allow for evaluation of transient behavior.
As a result of the MPR seal analyses, Catawba determined that a modification to the No. 1 RCP seal leak-off piping 0.254 inch bore restriction orifice positioned downstream of the seal exit was required. This modification installed a thicker orifice plate to accommodate the higher differential pressures associated with an extended LOSC event. This modification did not change Catawba's leak-off configuration categorization discussed in the PWROG-14015-P, revision 2 report dated April 2015 (Catawba remained a Category 1 plant).
Catawba is not crediting the MPR analyses for ELAP response or compliance with Order EA   049 even though the officially credited PWROG work to resolve remaining open issues is not yet complete. The MPR analysis results for both seal leak rate and the attendant leak-off piping pressure-temperature conditions during an ELAP/LOSC event show Catawba seal leak rates (post orifice plate modification) to be bounded by the documented Westinghouse leakage results as identified in the PWROG-14027-P revision 3 report, dated April 2015. Additionally, in-house piping stress/support and component evaluations of the RCP No. I seal leak-off lines shows the modified system retains its integrity throughout the transient predicted by the MPR seal model, as well as at more extreme conditions.
During the October 2014 NRC FLEX Audit, the above information was discussed with the NRC audit team and with ONRR, and subsequently an information package was placed on the Catawba e-Portal (which has now been moved to a SharePoint) for technical staff information/use. Catawba has installed thicker 0.254 inch bore restriction orifices in all four RCP No.1 seal leak-off lines on each Unit and declared Unit 2 to be in compliance with Order EA-12-049 in March 2015 and Unit I to be in compliance in December 2015. The status of the Catawba response to elevated RCP No. 1 seal leak-off line conditions and orifice plate modification was updated in the EA-1 2-049 Fourth Six-month Status Report dated February 28, 2015.
In March 2015, specific transient conditions potentially requiring further evaluation of the RCP No. 1 seal leak-off piping were identified by PWROG via Westinghouse NSAL-1 5-2. This NSAL Page 2 of 8

ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT formally identifies the existence of a potential 2045 psia pressure spike that occurs at the No. 1 seal exit early in the LOSC transient and its potential effect on the seal leak-off line, a transient the Westinghouse seal model cannot specifically evaluate as noted previously. As a result of this model limitation, the NSAL recommends Licensees assume a conservatively high seal exit pressure and temperature in the leak-off piping to account for the pressure spike for evaluation of system response to an ELAP. While the current Catawba MPR ELAP transient analysis predicts lower pressure and temperature conditions than those recommended by NSAL-15-2, an additional analysis case was run by MPR with a 2045 psia pressure (the NSAL-15-2 recommendation) as a forced input at the seal exit and another analysis at 2500 psia. This analytical approach removes reliance on the MPR seal model entirely and allows for an independent thermal-hydraulic evaluation of the RCP No. 1 seal leak-off line. Based on the MPR evaluation and resulting piping and support analyses performed by Duke, the ROP No.1 seal leak-off piping remains adequately protected from this extreme pressure transient with the thicker restriction orifice plate installed. As such, the published PWROG Category 1 ELAP leak off rates still apply for Catawba.
: 3.       Margin Assessment This margin assessment was performed using the examples of pertinent information regarding seal leakage provided by NRC via e-mail dated March 31, 2015. This assessment highlights the favorable aspects of the Catawba FLEX strategy and identifies areas with margin.
3.1      Early RCS Cooldown Per ECA-0.0 response, symmetric RCS cooldown/depressurization at Catawba is started within 1-2 hours of ELAP onset to minimize RCS inventory loss and protect the ROP seal packages.
Post-event initiation, RCS conditions at Catawba will peak at 2485 psig and 5680 F until cooldown commences. The Catawba RCP Model 93A seal packages contain 0-rings made from 7228C elastomer material, which has been evaluated to withstand up to 582&deg;F for eight hours. Early initiation of RCS cooldown therefore provides assurance the RCP seals will continue their function to limit leak off flow.
Additional Favorable Cooldown Information Along with facilitating RCS conditions favorable for passive injection of highly borated water from the Cold Leg Accumulators, the Catawba RCS cooldown strategy over the first 24 hours of the ELAP event specifically supports the integrity of the RCP No. 2 seal consistent with Westinghouse guidance. Based on discussions from the October 27, 2015 audit between the NRC, AREVA, and the PWROG, it was agreed that No. 1 seal corrosion does not invalidate the conservatisms in the PWROG-1 401 5-P, revision 2 leakage analysis when an early RCS cooldown is performed. Specifically, an RCS cooldown as specified in Appendix D of WCAP-1 7792-P, revision 0 is acceptable to mitigate corrosion of the No. 1 seal. Based on the results of the NRC audit on No. 1 seal corrosion, the first cooldown specified in TB-I15-1 will not change from 450&deg;F within 4 hours. The second cooldown requirement for cold leg parameters will be revised to 400&deg;F and 800 psig within 24 hours. This revision allows margin to continue to rely on the Turbine Driven Auxiliary Feedwater pump for at least 24 hours following a loss of seal cooling. A temperature of 4000&deg;F also provides a favorable long term condition to prevent continued corrosion of the No. I seal.
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ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT 3.2       Early RCS Makeup In order to identify margin associated with the RCS makeup strategy, two characteristics related to RCS behavior are addressed: 1) Adequate boration capability/adequate boron mixing during two-phase natural circulation in the RCS to prevent a return to criticality, and 2) The predicted time to reflux cooling in the steam generators.
Adequate Boration Capability and Boron Mixing For an ELAP scenario initiating while in Modes 1-4, RCS boration at Catawba will begin around 11.6 hours after the start of the event and is based on preventing a potential return to criticality calculated to be 13.85 hours rather than the predicted onset of reflux cooling in the SG tubes, which occurs later.
As noted previously, after the initiation of an ELAP event, the operators will cooldown and depressurize the RCS to approximately 420&deg;F/400 psia within the first several hours in order to minimize RCP seal leakage and inventory loss. Operators will then ensure sufficient boration has been completed before continuing to cooldown further. The Catawba high pressure FLEX makeup pump has sufficient performance (40 gpm at 1700 psig) to ensure injection flow is greater than RCP seal leak off flow at the time of alignment to the RCS.
Endorsed NEI 12-06 guidance indicates normal plant operational parameters can be assumed prior to onset of an ELAP, in lieu of the more restrictive limits of a design basis analysis. In performing the in-house RELAP5 Catawba ELAP boration evaluation, credit for parameters in their nominally expected ranges was not generally taken (i.e., more limiting assumptions were made), which provides for a qualitative margin assessment as follows:
* For this evaluation all four RCP seal packages are assumed not to leak (i.e., they seal perfectly), minimizing RCS letdown and maximizing the boron injection requirement
* Boration requirements for Catawba RCS cooldown are based on an ELAP event occurring after a 500-day EFPD reactor run (EOC), with the most limiting equilibrium Xenon characteristics
* The assumed required final RCS boron concentration after FLEX makeup pump injection is conservatively high, which increases the amount of borated water volume injection to meet shutdown requirements
* Assumed decay heat is representative of EOC
* Minimum boron concentration allowed by Technical Specifications (TS) is assumed in the Cold Leg Accumulators
* Minimum cover pressure is assumed in the Cold Leg Accumulators
* Minimum boron concentration allowed by TS is assumed in the Refueling Water Storage Tank Page 4 of 8

ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT
* An hour is subtracted from the actual time to re-criticality (and hence the response time) to ensure adequate boron mixing occurs during FLEX pump make-up
* The time to start the FLEX make-up pump is calculated based on the required boron curve at an ROS temperature of 350&deg;F; during boration activities operators would maintain the plant near 420&deg;F. This conservatively requires the FLEX make-up pump to start earlier than necessary
* The Pressurizer is assumed to only be filled to 60% level prior to requiring RCS letdown through the RV head vents. Controlling the injection pump to RCS pressure in lieu of Pressurizer level would reduce the total boration time (and delay the boration start setpoint) by allowing additional RCS injection.
The margin inherent in the boration calculation assumptions/inputs therefore shows that any return to criticality during an ELAP event would reasonably be expected to occur well beyond the maximum 13.85 hour RCS make-up setpoint in the documented Catawba FLEX response.
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ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT Time to Re flux Cooling For the latest NOTRUMP reference case, the PWROG-1 4027-P, revision 3 report dated April 2015 for 4-loop TCoId plants in Category 1, identifies that Catawba will enter reflux cooling at 15.6 hours with time to uncover the core at 43.9 hours during an ELAP event. Initiating RCS boration around 11.6 hours and no later than 13.85 hours after event initiation at Catawba therefore ensures that boration will occur with acceptable loop flow conditions.
Catawba initially performed an in-house analysis of the time to reflux cooling using the RELAP5 code to establish a setpoint for RCS boration during ELAP, using the seal leakage profile from WCAP-1 7601 -P, revision 1. Subsequent to that analysis, sensitivity cases were also run in-house to evaluate the new seal leakage rates identified in PWROG-14015-P. Margin in the calculation of the predicted time to reflux cooling in the steam generator U-tubes is qualitatively identified in these analyses as noted below:
      *For this evaluation, all four RCP seals are assumed to leak at their maximum flow rate, minimizing the time to reflux cooling in the steam generators
      *Catawba-specific ELAP mass-energy release evaluation (RELAP5) shows a predicted time to reflux cooling well beyond the Westinghouse timeframe
      *Catawba has installed new thicker restriction orifice plates in the RCP No. 1 seal leak-off lines. The thicker orifice plates ensure system integrity at the higher postulated pressures so that Catawba remains in PWROG Category 1.
     *Catawba-specific ELAP mass-energy release sensitivity cases (RELAP5) adjusted for the revised PWROG Category 1 RCP seal leak-off profile show that the predicted time to reflux cooling in the steam generator tubes is still considerably delayed as compared to the reference case
      *MPR analysis flowrates for the Catawba leak-off piping configuration (MPR site-specific models) have shown peak values less than those shown in the WCAP-17601-P, revision 1 or the PWROG-14015-P, revision 2 reference cases. Therefore, cumulative RCP seal leakage will likely be lower than identified for PWROG Category 1 plants
     *Assumed decay heat is representative of EOC Page 6 of 8

ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT 3.3      Possessing the Capability to Initiate RCS Makeup within "X"Hours (Shorter than Planned Time)
RCS makeup during an ELAP event is a prioritized action per ECA-0.0, and relies on diesel-driven injection pumps that don't require FLEX electrical distribution to be set up first. Catawba stores the N+I pumps in a single FLEX Building (dome) constructed to NEI 12-06 requirements to protect FLEX response capability. The maximum RCS injection time start setpoint is 13.85 hours based on DPC-1 552.08-00-0278 revision 2. Pump deployment from the dome assumes maximum debris removal and pump deployment times (i.e., 3 hours for debris removal and deployment of pump starting at 6.5 hours). Current guidance located in Catawba's FSG-05 "Initial Assessment and FLEX Equipment Staging" directs responding Operators to identify availability of FLEX resources early in the event. This serves to minimize deployment times of prioritized actions such as RCS injection. Based on the Unit 1 and 2 Flex Strategy Timing Validation Study (Letter to File CN-1612.03), a reasonable timeframe for initiating boration following an ELAP is around 11.6 hours with additional margin if deployment can be started sooner than 6.5 hours based on site conditions and available resources.
3.4       Having an Abundant Supply of B orated Coolant Onsite and/or Having a Relatively Large Capacity for Injecting Coolant Catawba has adequate onsite borated makeup capacity for at least 72 hours following the onset of an ELAP event in Modes 1-4. Catawba's FSG-08 "Alternate NC System Boration" directs responding Operators to utilize the borated inventory available in the FWST (Refueling Water Storage Tank) for RCS makeup (approximately 6 days worth if FWST is undamaged). Should the FWST be damaged by a wind-borne missile above the protective wall, further boration capability is afforded by using water in the protected Boric Acid Tanks (BATs) and/or mixing powdered boric acid with water from the Standby Nuclear Service Water Pond (Ultimate Heat Sink) in portable mixing tanks. Beyond 72 hours, equipment from the Memphis or Phoenix NSRC (i.e., mobile boration skid) is available.
Additionally, Catawba's Standby Nuclear Service Water Pond remains available as a clean (i.e.,
< 16 ppm TSS) un-borated water source, and its use is proceduralized later in the ELAP event.
Use of the 40 gpm makeup pump and the FWST/BATs provides sufficient boration to reach the reactivity objective. The supply of borated coolant onsite provides several (> 3) days of boration capacity.
3.5       Having a High Capacity and/or High Pressure RCS FLEX Makeup Pump The Catawba FLEX High Pressure Pump (diesel driven) has a capacity of 40 gpm at 1700 psig, and has a variable speed control for flow and pressure. The diesel driver provides the ability to increase injection flowrates by approximately 50% during RCS depressurization if needed without changing to a different pump.
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ATTACHMENT 7 CATAWBA NUCLEAR STATION REACTOR COOLANT PUMP SEAL LEAKAGE MARGIN ASSESSMENT 3.6      Having the Ability to Monitor RCS Inventory during the Event and Attempting to Implement Makeup More Rapidly If Signs of Increased Leakage Were Detected ECA-O.O "Loss of All AC Power" lists the critical instruments required to be maintained during the ELAP transient. Available instrumentation related to monitoring RCS inventory includes:
* RCS wide range pressure
* RCS wide range hot leg temperature
* Core exit thermocouples
* RVLIS
* Pressurizer level
* Neutron flux Current guidance located in ECA-0.0 "Loss of All AC Power" and FSG-8 "Alternate NC System Boration" instructs responding Operators to prioritize RCS injection and respond more quickly if high RCS leakage is suspected.
3.7     NSAL-15-2 Leak-Off Line Break As noted previously, Catawba has evaluated the leak-off piping/components for a transient pressure spike at the No. 1 seal exit up to 2500 psia per the suggestion of the NRC to ensure system integrity is maintained. Results show the RCP No.1 seal leak-off piping remains adequately protected following installation of the thicker flow measurement orifice plates. As such, the published PWROG Category 1 ELAP leak-off rates still apply for Catawba.
3.8     Additional Considerations There is additional qualitative margin associated with the PWROG analyses performed in support of this issue. The following items were specifically noted by the NRC in the Watts Bar Mitigating Strategies Safety Evaluation dated March 27, 2015:
* The PWROG's generic ITCHSEAL calculations contain known conservatisms such as the comparison of the results of the generic analysis to the Montereau test data and in the application of the generic leak-off line configuration assumptions for each maximum leakage analysis category to individual plants' leak-off lines.
* Although entry into reflux cooling is undesirable and has not been fully analyzed in the context of an ELAP event, the use of this threshold as an acceptance criterion provides significant margin to uncover and severely damage the core.
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