ML15069A358

Report for the Onsite Audit Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Instrumentation Related to Orders EA-12-049 and EA-12-051

ML15069A358
Person / Time
Site:	Browns Ferry
Issue date:	04/06/2015
From:	Tony Brown Japan Lessons-Learned Division
To:	James Shea Tennessee Valley Authority
References
EA-12-049, EA-12-051, TAC MF0881, TAC MF0882, TAC MF0883, TAC MF0902, TAC MF0903, TAC MF0904
Download: ML15069A358 (18)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 6, 2015 Mr. Joseph W. Shea Vice President, Nuclear Licensing Tennessee Valley Authority 1101 Market Street LP 3D-C Chattanooga, TN 37402

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 - REPORT FOR THE ONSITE AUDIT REGARDING IMPLEMENTATION OF MITIGATING STRATEGIES AND RELIABLE SPENT FUEL INSTRUMENTATION RELATED TO ORDERS EA-12-049 AND EA-12-051 (TAC NOS. MF0902, MF0903, MF0904, MF0881, MF0882, AND MF0883)

Dear Mr. Shea:

On March 12, 2012, the U.S. 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-12-051, "Order to Modify Licenses With Regard To Reliable Spent Fuel Pool Instrumentation," (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML12054A736 and ML12054A679, 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 of each order will be achieved.

By letter dated February 28, 2013, (ADAMS Accession No. ML13064A465), Tennessee Valley Authority (TVA, the licensee) submitted its OIP for Browns Ferry Nuclear Plant, Units 1, 2, and 3 (Browns Ferry) in response to Order EA-12-049. By letters dated August 28, 2013, February 28, 2014, August 28, 2014, and February 27, 2015 (ADAMS Accession Nos. ML13247A284, ML14064A240, ML14248A496, and ML15064A162, respectively), TVA submitted its first four six-month updates to the OIP By letter dated August 28, 2013 (ADAMS Accession No ML13234A503), the NRC notified all licensees and construction permit holders that the staff is conducting audits of their responses to Order EA-12-049 in accordance with NRC Office of Nuclear Reactor Reguiation (NRR) Office Instruction LIC-111, 'Regulatory Audits" (ADAMS Accession No. ML082900195). This audit process led to the issuance of the Browns Ferry interim staff evaluation (ISE) on December 19, 2013 (ADAMS Accession No. ML13353A166),

and continues with in-office and onsite portions of this audit.

By letter dated February 28, 2013 (ADAMS Accession No. ML13063A437}, the licensee submitted its 01 P for Browns Ferry in response to Order EA-12-051. By letter dated June 18, 2013 (ADAMS Accession No. ML13157A164), the NRC staff sent a request for additional information (RAI) to the licensee. By letters dated July 18, 20U, August 28, 2013, February 28, 2014, August 28, 2014, and February 27, 2015 (ADAMS Accession Nos. ML13206A005, ML13247A290, ML14064A240, ML14247A430, and ML15064A188, respectively}, the licensee submitted its RAI responses and first four six-month updates to the OIP.

J. Shea The NRC staff's review led to the issuance of the Browns Ferry ISE and RAI dated November 14, 2013 (ADAMS Accession No. ML13274A657). By letter dated March 26, 2014 (ADAMS Accession No. ML14083A620), the NRC notified all licensees and construction permit holders that the staff is conducting in-office and onsite audits of their responses to Order EA-12-051 in accordance with NRC NRR Office Instruction LIC-111, as discussed above.

The ongoing audit process, to include the in-office and onsite portions, allows the staff to assess whether it has enough information to make a safety evaluation of the Integrated Plans. The audit allows the staff to review open and confirmatory items from the mitigation strategies ISE, RAI responses from the spent fuel pool instrumentation (SFPI) ISE, the licensee's integrated plans, and other audit questions. Additionally, the staff gains a better understanding of submitted and updated information, audit information provided on ePortals, and preliminary Overall Program Documents/Final Integrated Plans while identifying additional information necessary for the licensee to supplement its plan and address staff potential concerns.

In support of the ongoing audit of the licensee's OIPs, as supplemented, the NRC staff conducted an onsite audit at Browns Ferry from January 5-9, 2015, per the audit plan dated November 26, 2014 (ADAMS Accession No. ML14323A295). The purpose of the onsite portion of the audit was to provide the NRC staff the opportunity to continue the audit review and gain key insights most easily obtained at the plant as to whether the licensee is on the correct path for compliance with the Mitigation Strategies and SFPI orders. The onsite activities included detailed analysis and calculation discussion, walk-throughs of strategies and equipment laydown, visualization of portable equipment storage and deployment, review of staging and deployment of offsite equipment, and review of installation details for SFPI equipment.

The enclosed audit report provides a summary of the activities for the onsite audit portion.

Additionally, this report contains an attachment listing all open audit items currently under NRC staff review.

J. Shea If you have any questions, please contact me at 301-415-1924 or by e-mail at Tony.Brown@nrc.gov.

Tony Br n, Project Manager Orders Management Branch Japan Lessons-Learned Division Office of Nuclear Reactor Regulation Docket Nos.: 50-259, 50-260, and 50-296

Enclosure:

Audit report cc w/encl: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 AUDIT REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO ORDERS EA-12-049 AND EA-12-051 MODIFYING LICENSES WITH REGARD TO REQUIREMENTS FOR MITIGATION STRATEGIES FOR BEYOND-DESIGN-BASIS EXTERNAL EVENTS AND RELIABLE SPENT FUEL POOL INSTRUMENTATION TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2. AND 3 DOCKET NOS. 50-259, 50-260, and 50-296 BACKGROUND AND AUDIT BASIS On March 12, 2012, the U.S. 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-12-051, "Order to Modify Licenses With Regard To Reliable Spent Fuel Pool Instrumentation," (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML12054A736 and ML12054A679, respectively). Order EA-12-049 directs licensees to develop, implement, and maintain guidance and strategies to maintain or restore core cooling, containment, and spent fuel pool (SFP) cooling capabilities in the event of a beyond-design-basis external event (BDBEE). Order EA-12-051 requires, in part, that all operating reactor sites have a reliable means of remotely monitoring wide-range SFP levels to support effective prioritization of event mitigation and recovery actions in the event of a BDBEE. 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 Attachment 2 of each order will be achieved.

By letter dated February 28, 2013, (ADAMS Accession No. ML13064A465), Tennessee Valley Authority (TVA, the licensee) submitted its OIP for Browns Ferry Nuclear Plant, Units 1, 2, and 3 (Browns Ferry) in response to Order EA-12-049. By letters dated August 28, 2013, February 28, 2014, August 28, 2014, and February 27, 2015 (ADAMS Accession Nos. ML13247A284, ML14064A240, ML14248A496, and ML15064A162, respectively), TVA submitted its first four six-month updates to the OIP. By letter dated August 28, 2013 (ADAMS Accession No. ML13234A503), the NRC notified all licensees and construction permit holders that the staff is conducting audits of their responses to Order EA-12-049 in accordance with NRC Office of Nuclear Reactor Regulation (NRR) Office Instruction LIC-111, "Regulatory Audits" (ADAMS Enclosure

Accession No. ML082900195). This audit process led to the issuance of the Browns Ferry interim staff evaluation (ISE) on December 19, 2013 (ADAMS Accession No. ML13353A166),

and continues with in-office and onsite portions of this audit.

By letter dated February 28, 2013 (ADAMS Accession No. ML13063A437), the licensee submitted its 01 P for Browns Ferry in response to Order EA-12-051. By letter dated June 18, 2013 (ADAMS Accession No. ML13157A164), the NRG staff sent a request for additional information (RAI) to the licensee. By letters dated July 18, 2013, August 28, 2013, February 28, 2014, August 28, 2014, and February 27, 2015 (ADAMS Accession Nos. ML13206A005, ML13247A290, ML14064A240, ML14247A430, and ML15064A188, respectively), the licensee submitted its RAI responses and first four six-month updates to the OIP. The NRG staff's review led to the issuance of the Browns Ferry ISE and RAI dated November 14, 2013 (ADAMS Accession No. ML13274A657). By letter dated March 26, 2014 (ADAMS Accession No. ML14083A620), the NRG notified all licensees and construction permit holders that the staff is conducting in-office and onsite audits of their responses to Order EA-12-051 in accordance with NRG NRR Office Instruction LIC-111, as discussed above.

The ongoing audit process, to include the in-office and onsite portions, allows the staff to assess whether it has enough information to make a safety evaluation of the Integrated Plans. The audit allows the staff to review open and confirmatory items from the mitigation strategies (MS)

ISE, RAI responses from the SFP instrumentation(SFPI) ISE, the licensee's integrated plans, and other audit questions. Additionally, the staff gains a better understanding of submitted and updated information, audit information provided on ePortals, and preliminary Overall Program Documents (OPDs)/Final Integrated Plans {FIPs) while identifying additional information necessary for the licensee to supplement its plan and address staff potential concerns.

In support of the ongoing audit of the licensee's OIPs, as supplemented, the NRG staff conducted an onsite audit at Browns Ferry from January 5-9, 2015, per the audit plan dated November 26, 2014 (ADAMS Accession No. ML14323A295). The purpose of the onsite portion of the audit was to provide the NRG staff the opportunity to continue the audit review and gain key insights most easily obtained at the plant as to whether the licensee is on the correct path for compliance with the Mitigation Strategies and SFPI orders. The onsite activities included detailed analysis and calculation discussion, walk-throughs of strategies and equipment laydown, visualization of portable equipment storage and deployment, review of staging and deployment of offsite equipment, and review of installation details for SFPI equipment.

Following the licensee's declarations of order compliance, the NRG staff will evaluate the OIPs, as supplemented; the resulting site-specific OPDs/FIPs; and, as appropriate, other licensee submittals based on the requirements in the orders. For Order EA-12-049, the staff will make a safety determination using the Nuclear Energy Institute (NEI) developed guidance document NEI 12-06, "Diverse and Flexible Coping Strategies (FLEX) Implementation Guide" issued in August 2012 (ADAMS Accession No. ML12242A378), as endorsed by NRG Japan Lessons-Learned Directorate (JLD) interim staff guidance (ISG) 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"' (ADAMS Accession No. ML12229A174).

For Order EA-12-051, the staff will make a safety determination using the NEI developed guidance document NEI 12-02, Revision 1, "Industry Guidance for Compliance with NRG Order EA-12-051, 'To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation"'

(ADAMS Accession No. ML12240A307), as endorsed, with exceptions and clarifications, by NRC ISG JLD-ISG-2012-03 "Compliance with Order EA-12-051, 'Reliable Spent Fuel Pool Instrumentation"' (ADAMS Accession No. ML12221A339) as providing one acceptable means of meeting the order requirements. Should the licensee propose an alternative strategy for compliance, additional staff review will be required to evaluate the alternative strategy in reference to the applicable order.

AUDIT ACTIVITIES The onsite audit was conducted at the Browns Ferry facility from January 5, 2015, through January 9, 2015. The NRC audit team staff was as follows:

Title Team Member Organization Team Lead/Project Manager Tony Brown NRR/JLD Technical Support - Balance of Plant Garry ArmstronQ NRR/JLD Technical Support - Reactor Systems Joshua Miller NRR/JLD Technical Support - Electrical Matthew McConnell NRR/JLD Technical Support - l&C Stephen Wyman NRR/JLD Technical Support - Hazards John Bowen Contractor Technical Support - Containment Brett Titus NRR/JLD The NRC staff executed the onsite portion of the audit per the three part approach discussed in the November 26, 2014, plan, to include conducting a tabletop discussion of the site's integrated mitigating strategies compliance program, a review of specific technical review items, and discussion of specific program topics. Activities that were planned to support the above included detailed analysis and calculation discussions, walk-throughs of strategies and equipment laydown, visualization of portable equipment storage and deployment, staging and deployment of offsite equipment, and physical sizing and placement of SFPI equipment.

AUDIT

SUMMARY

1.0 Entrance Meeting (January 5, 2015)

At the audit entrance meeting, the NRC staff audit team introduced itself followed by introductions from the licensee's staff. The NRC audit team provided a brief overview of the audit's objectives and anticipated schedule.

2.0 Integrated Mitigating Strategies Compliance Program Overview Per the audit plan and as an introduction to the site's program, the licensee provided a presentation to the NRC audit team describing the site's strategies to meet the NRC orders.

The licensee reviewed its strategy to maintain core cooling, containment, and SFP cooling in the event of a BDBEE, and the plant modifications being done in order to implement the strategies.

Also reviewed was the design and location of the storage facilities for the FLEX equipment, the interface with the National Strategic Alliance for FLEX Emergency Response (SAFER)

Response Center including staging areas, the SFP level indication modification, the modifications planned to enhance emergency communications, preventative maintenance plans

for the FLEX equipment, procedural enhancements such as development of FLEX support guidelines, and operator training.

3.0 Onsite Audit Technical Discussion Topics Based on the audit plan, and with a particular emphasis on the Part 2 "Specific Technical Review Items," the NRG staff technical reviewers conducted interviews with licensee technical staff, site walk-downs, and detailed document review for the items identified in the plan. Results of these technical reviews and any additional review items needed from the licensee are documented in the audit item status table in Attachment 3, as discussed in the Conclusion section below.

3.1 Reactor Systems Technical Discussions and Walk-Downs NRG staff met with licensee staff to discuss the amount of leakage from the recirculation pump (RCP) seals, the timing of the makeup of water to the Reactor Core Isolation Cooling (RCIC) system used for core cooling, and the availability of water sources. NRG staff reviewed the analsysis and flow calculations along with applicable procedures. NRG staff noted that the Modular Accident Analysis Program (MAAP) 4 analysis is not finalized and requested to review it once completed. NRG staff reviewed the licensee's strategy for utilizing raw water sources (Tennessee River), including water filtration and monitoring of core parameters to ensure adequate cooling. NRG staff also walked down the licensee's strategies and reviewed plant procedures for implementing the core cooling and makeup strategies.

3.2 Electrical Technical Discussions and Walk-Downs

a. NRG staff reviewed the calculations on extending battery life based on load shedding, and walked down the battery rooms to evaluate strategies for hydrogen and temperature control. NRG staff also walked down panels used for load shedding to evaluate feasibility and timing.

b. NRG staff walked down connection points and location for FLEX electrical generators.

The staff noted that there is only one pre-determined location for staging the Phase 2 FLEX generators, which is located adjacent to structures that are not fully protected from all hazards. The staff requested the licensee provide justification that the generator can be placed into service in the required time in the event the area is severely obstructed (Attachment 3). The staff reviewed the licensee's load and sizing calculations for the Phase 2 FLEX generators and also requested the sizing calculations for the Phase 3 4.16kV generators that will be received from the National SAFER Response Center.

The staff also walked down the storage locations for the FLEX diesel generators (DGs).

3.3 SFPI Technical Discussions and Walk-Downs NRG staff walked down instrument, transmitter, electronics, and display locations for the SFP level instrumentation, along with the associated cable runs. No concerns were identified during the walkdown. NRG staff also reviewed the associated calibration, maintenance and test procedures for the SFP level instrumentation.

3.4 Other Technical Discussion Areas and Walk-Downs

a. NRC staff toured the designated location of the FLEX equipment storage building (FESB) and reviewed the building plans and noted that it will be a robust building (as defined in NEI 12-06). The staff walked down equipment haul routes from the storage building to the designated deployment sites, and walked down haul routes from designated staging areas for equipment that will be delivered from the National SAFER Response Center.

b. NRC staff walked down the FLEX strategies for core cooling, reactor coolant system inventory, and SFP cooling functions. This included the point of deployment for the portable FLEX pumps, hose routing and deployment connection points (primary and alternate). The licensee's strategy relies upon portable FLEX pumps that will be staged east of the forebay to take suction from the Tennessee River to supply water to the Residual Heat Removal Service Water and Emergency Equipment Cooling Water permanent piping systems to maintain core cooling, containment cooling and spent fuel pool cooling functions. The staff also reviewed the licensee's hydraulic analyses to ensure pump sizing and location were sufficient to provide the necessary water flow to perform the associated functions.

c. NRC staff reviewed the strategy that will be implemented by the licensee to refuel the portable diesel-powered FLEX equipment. The NRC staff reviewed the instructions for refueling the equipment as well as the equipment needed to perform the refueling.

Additionally, the staff reviewed the licensee's procedures for ensuring adequate fuel quality.

d. NRC staff walked down the licensee's strategy for ventilating the main control rooms and battery board rooms to ensure equipment reliability and personnel habitability. The staff noted that the licensee is relying upon the new hardened containment vent system (HCVS), as required by Order EA-13-109, as part of its strategy to meet Order EA-12-049. The NRC has granted relaxation to compliance with Order EA-12-049, for Browns Ferry, Units 2 and 3, to coincide with the licensee's compliance with Order EA-13-109. As such, the staff has requested that the licensee provide final design documentation for the HCVS, along with the associated calculations and procedures to ensure equipment reliability and personnel habitability in the reactor building.

e. The staff reviewed the licensee's plans to ensure adequate communications, lighting, personnel access, and equipment access, to successfully implement the strategies. The staff interviewed plant personnel responsible for these areas, and observed lighting and communication needs during plant walkdowns.

4.0 Exit Meeting (January 9, 2015)

The NRC staff audit team conducted an exit meeting with licensee staff following the closure of onsite audit activities. The NRC staff highlighted items reviewed and noted that the results of the onsite audit trip will be documented in this report. The NRC staff also discussed the remaining open items with the licensee and information needed for closure. The open items are listed in Attachment 3 of this report.

CONCLUSION The NRC staff completed all three parts of the November 26, 2014, onsite audit plan. The audit items identified in Part 2 of the plan was reviewed by NRC staff members while on site. In addition to the list of NRC and licensee onsite audit staff participants in Attachment 1, provides a list of documents reviewed during the onsite audit portion.

In support of the continuing audit process as the licensee proceeds towards orders compliance for this site, Attachment 3 provides the status of all open audit review items that the NRC staff is evaluating in anticipation of issuance of a combined safety evaluation for both the Mitigation Strategies and Spent Fuel Pool Level Instrumentation orders. The five sources for the audit items referenced below are as follows:

a. Interim Staff Evaluation (ISE) Open Items (Ols) and Confirmatory Items (Cls)

b. Audit Questions (AQs)

c. Licensee-identified OIP Open Items (Ols)

d. Spent Fuel Pool Level Instrumentation (SFPLI) RAls

e. Additional Safety Evaluation (SE) needed information The attachments provide audit information as follows:

a. Attachment 1: List of NRC staff and licensee staff audit participants

b. Attachment 2: List of documents reviewed during the onsite audit

c. Attachment 3: MS/SFPI SE Audit Items currently under NRC staff review (licensee input needed as noted)

While this report notes the completion of the onsite portion of the audit per the audit plan dated November 26, 2014, the ongoing audit process continues as per the letters dated August 28, 2013, and March 26, 2014, to all licensees and construction permit holders for both orders.

Additionally, while Attachment 3 provides a list of currently open items, the status and progress of the NRC staff's review may change based on licensee plan changes, resolution of generic issues, and other NRC staff concerns not previou/sly documented. Changes in the NRC staff review will be communicated in the ongoing audit process.

Attachments:

1. NRC and Licensee Staff Onsite Audit Participants

2. Onsite Audit Documents Reviewed

3. MS/SFPI Audit Items currently under NRC staff review

Onsite Audit Participants NRC Staff:

Tony Brown NRR/JLD/JOMB Stephen Wyman NRR/JLD/JHMB Garry Armstrong NRR/JLD/JCBB Brett Titus NRR/JLD/JCBB Joshua Miller NRR/JLD/JERB John Bowen Contractor Matthew McConnell NRR/JLD/JERB Browns Ferry and TVA Staff:

Keith Polson BFN Site Vice President Steve Bono BFN Plant Manaqer Jim Kent Site Projects, Sr. Project Manaqer Gerald Doyle Director, EPU/PUP Scott Hunnewell Engineering Director Mike Oliver Site Licensing, Sr. Program Manager Phillip Summers Plant Support Director John Kulisek Emergency Preparedness Manager Aaron Bergeron Training Director Neil Gannon Director, Fukushima Strategies Pat Parker Site Security Manager Micah Nash FLEX Response Anthony Feltman FLEX Response Wes Howard Site Security Support Manager Greg Burgess Site Security Coordinator David Matherly Site VP Technical Advisor Amy Reagan Communications D.L. Hughes Operations Director Jeff Barker Operations Support Superintendent Brandon Grav Project Manager, BFN Fukushima Project Victoria Dennis Work Management Kevin Casey Corporate Licensing, Fukushima Project Sr. Program Manager Attachment 1

Documents Reviewed

• AMEC Environment & Infrastructure, Inc., AMEC Project 3043121002, "Report Of Geotechnical Exploration Commercial Grade "Bunker" Building TVA Browns Ferry Nuclear Plant," December 11, 2012

• "Addendum To Report Of Geotechnical Exploration Ground Motion And Liquefaction Studies Commercial Grade "Bunker" Building TVA Browns Ferry Nuclear Plant",

AMEC Environment & Infrastructure, Inc., March 7, 2013

• AMEC Environment & Infrastructure, Inc., AMEC Project 3050140243, "Report Of Geotechnical Exploration Alternative Travel Path Analysis TVA Browns Ferry Nuclear Plant," June 19, 2014

• AMEC Environment & Infrastructure, Inc., AMEC Project 3050-14-0240, "Report of State Route Study Emergency Equipment Mobilization Routes TVA Browns Ferry Nuclear Plant," July 11, 2014

• AMEC Environment & Infrastructure, Inc., AMEC Project 3050140251, "Report Of Geotechnical Exploration Additional Haul Paths Analysis TVA Browns Ferry Nuclear Plant," August 19, 2014

• General Criteria Document No. BFN-50-7360, RIMS#: R14 141114 102, "FLEX Mitigation System," Rev. 3

• CDQ0003602013000136, "FLEX Equipment Storage Building - Structural Calculations," Rev. 0

• CDQ0003602013000135, "FLEX Equipment Storage Building - Civil Design Basis and Criteria," Rev. O

• CDQ007920030261, "Soil Structure Interaction Analysis for BFN ISFSI Pad (Holtec Report No. Hl-2022963 For The Dry Cask Storage Project)," Rev. 0

• CDQ0009992014000268, "BFN Expedited Seismic Evaluation Process (ESEP)

HCLPF Capacity Evaluations," Rev. 001

• Design Change Notice 707 45

• Design Change Notice 70807

• Design Change Notice 70852

• Design Change Notice 71159

• Design Change Notice 71160

• Design Change Notice 71161

• Design Change Notice 71329

• Design Change Notice 71335

• Design Change Notice 71386

• Design Change Notice 71387

• Design Change Notice 71389

• Design Change Notice 71390

• Design Change Notice 71391

• Design Change Notice 71405 Attachment 2

• Design Change Notice 71454

• Design Change Notice 71470

• EDQ0003602014000281, "Electrical Evaluation for Portable Power Supply for Unit Battery Chargers," Rev. 000

• EDQ0009992013000202, "250V DC Unit Batteries 1, 2 & 3 Evaluation for Beyond Design Basis External Event (BDBEE) Extended Loss of AC Power (ELAP)," Rev. 001

• O-AOl-57-1A, "Station Blackout," Rev. Draft N

• O-FSl-1, "FLEX Response Instruction," Rev. Draft H

• O-FSl-2E, "FLEX Mechanical Hose Connections from EECW," Draft Rev. O

• O-FSl-3A, "480V FLEX Diesel Generator Setup and Operation," Draft Rev. 0

• O-FSl-3C, "4KV FLEX Generator Setup and Operation," Draft Rev. 0

• O-FSl-3F, "Load Shed of 250V Main Bank Battery 1, 2, 3," Draft Rev. O

• O-FSl-4A, "Localized Temporary Lighting, Ventilation and Communications," Draft Rev.

0

• O-FSl-5A, "Control Bay, Reactor Building Lighting and Communications from 4KV Distribution System," Draft Rev. 0

• O-FSl-5D, "MISC Electrical Alignments from 4KV Distribution System," Draft Rev.

OOOOB2

• O-FSl-6A, "Damage Assessment," Draft Rev. O

• O-FSl-6B, "FLEX - Long Term Fueling Operations," Draft Rev. O

• O-FSl-6C, "Key Instrument Readings During Loss of DC Power, Draft Rev. O

• 2-EOI Appendix 16K, "Bypassing RCIC High Temperature Isolation," Rev. 0002

• 2-EOI Appendix 20M, "RCIC Operations During Station Blackout," Rev. 0000

• 2-EOl-1, "RPV Control, Draft Rev. H3

• 2-01-35, "Generator Hydrogen Cooling System, Rev. 65

• MDN0003602014000233, "Hydraulic Analysis for Fukushima FLEX Connections Modifications, Rev. 001

• MDQ0003602014000222, "ELAP Transient Temp Analysis," Rev. 001

• NDQ0031890069, "Hydrogen Concentration in the Control Bay 250-Volt Station Battery Rooms," Rev. 1

• NPG-SPP-06.2, "Preventive Maintenance," Rev. 0009

• NPG-SPP-09.22.2, "NPG Programs and Processes, Diverse and Flexible Coping Strategies [FLEX] Program Document," Rev. 0000-A

• OPDP-1, " Conduct of Operations," Rev. 0033

• OPDP-8, "Operability Determination Process and Limiting Conditions for Operation Tracking," Rev. 0017

• PIP-02-03, "AC Electrical Distribution System Browns Ferry Nuclear Plant," 4/15/2014

• TRN-30, "Radiological Emergency Preparedness Training," Rev. 29

• AREVA Inc., Engineering Information Record, Document No.: 51-9233061-000, "Browns Ferry Nuclear Plant SAFER Response Plan, DRAFT, 1/30/2014

Mitigation Strategies/Spent Fuel Pool Instrumentation Safety Evaluation Audit Items:

Audit Items Currently Under NRC Staff Review, Requiring Licensee Input As Noted Audit Item Description Licensee Input Needed Item Reference ISECI From the June 2013 position paper, benchmarks must The staff requests that the licensee 3.2.1.1.A be identified and discussed which demonstrate that make available for audit the final MAAP MAAP4 is an appropriate code for the simulation of an analysis when complete and ensure ELAP event at your facility. this item is discussed.

ISECI The collapsed level must remain above Top of Active The staff requests that the licensee 3.2.1.1.B Fuel (TAF) and the cool down rate must be within make available for audit the final MAAP technical specification limits. analysis when complete and ensure this item is discussed.

ISECI MAAP4 must be used in accordance with Sections 4.1, The staff requests that the licensee 3.2.1.1.C 4.2, 4.3, 4.4, and 4.5 of the June 2013 position paper. make available for audit the final MAAP analysis when complete and ensure this item is discussed.

ISECI In using MAAP4, the licensee must identify and justify The staff requests that the licensee 3.2.1.1.D the subset of key modeling parameters cited from make available for audit the final MAAP Tables 4-1 through 4-6 of the "MAAP4 Application analysis when complete and ensure Guidance, Desktop Reference for Using MAAP4 this item is discussed.

Software, Revision 2" (Electric Revision 1 Page 64 of 71 2013-12-18 Power Research Institute Report 1 020236). This should include response at a plant-specific level regarding specific modeling options and parameter choices for key models that would be expected to substantially affect the ELAP analysis performed for that licensee's plant.

ISECI The specific MAAP4 analysis case that was used to The staff requests that the licensee 3.2.1.1.E validate the timing of mitigating strategies in the make available for audit the final MAAP integrated plan must be identified and should be analysis when complete and ensure available on the ePortal for NRC staff to view. this item is discussed.

Alternately, a comparable level of information may be included in the supplemental response. In either case, the analysis should include a plot of the collapsed vessel level to confirm that TAF is not reached (the elevation of the TAF should be provided) and a plot of the temperature cool down to confirm that the cool down is within technical specification limits.

ISECI There is insufficient information provided to determine No additional input from the licensee is 3.2.1.2.A the adequacy of the determination of recirculation needed at this time. Further review of pump seal or other sources of leakage used in the the information provided is required by ELAP analysis. the NRC staff.

Attachment 3

ISECI On page 10 of the Integrated Plan, the licensee stated The NRC staff requests that the 3.2.1.3.A that the Main Steam Relief Valve (MSRV) control is licensee demonstate that the MSRVs maintained from the control room with sufficient de will continue to operate as needed

[direct current] power and pneumatic pressure to during the ELAP event, assuming that operate the MSRVs throughout Phase 1 and Phase 2. calculated drywell temperatures The licensee describes that SRV actuation may exceed the environmental qualification require a higher than nominal de voltage to actuate the of the MSRVs for the duration of the MSRVs due to higher containment temperature with a event.

longer duration event than an existing SBO [station blackout] coping time. The SRV pilot solenoid coil electrical resistance will increase due to a higher containment temperature with a longer duration event than an existing SBO coping time. The licensee is planning to evaluate MSRV qualification against the predicted containment response with FLEX implementation to ensure there will be sufficient de bus voltage during the ELAP event. The licensee also provides that if required, there will be a modification to increase voltage as necessary to achieve the necessary coil current, or modifications will be made to reduce the coil resistance under higher temperature conditions. Because the MSRV control system will be exhausting control gas to the containment and containment pressure will be higher, the licensee is evaluating methods to establish any required increases in pneumatic supply pressure and modifications that may be required to ensure a supply of control gas for the MSRVs over the longer ELAP interval. These two questions were asked during the audit process and the licensee stated that the analysis/evaluation has not vet been completed.

ISECI The licensee has not provided finalized calculations Once the design of the system is 3.2.3.A which support the primary strategy timeline by finalized, the NRC staff requests that concluding that venting or other heat removal activities the licensee make available for audit will not be required during the first eight hours of the any formal calculations that will be event, maintaining a suppression pool temperature low performed for the configuration.

enough to support continued RCIC [Reactor Core Isolation Coolinql operation for this time period.

ISECI The licensee did not provide details regarding the The NRC staff requests the licensee 3.2.4.2.A effects of loss of ventilation in the HPCI [high-pressure clarify the timeline during which RCIC coolant injection] /RCIC pump rooms to conclude that is relied upon and when it is not the equipment in the HPCl/RCIC pump rooms would needed. Additionally, the NRC staff perform its function and assist in core cooling requests the licensee make available throughout all Phases of an ELAP. During the audit for audit the reactor building process, the licensee stated that preliminary analysis temperature calculation, has been performed, but the calculations have not MDQ0009992014000291, and also been finalized. Based on preliminary analysis, the clearly identify any actions to address RCIC room would reach 140 degrees at 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />. personnel habitability or room cooling.

The electronic governor module (EGM) for RCIC could fail with temperatures at 150 degrees Fahrenheit.

RCIC has steam isolation at 165 degrees in the room or torus area. EOI [emergency operating instruction]

Appendix SC, Injection System Lineup RCIC, allows bypassing of the high temperature isolation using booted contacts in accordance with EOI Appendix 16K, Bypassing RCIC High Temperature Isolation.

Core Spray room cooler strategy is being evaluated to aid in cooling of the Core Spray/RCIC room. A detailed summary of the analysis and/or technical evaluation performed to demonstrate the adequacy of the ventilation provided in the HPCl/RCIC pump rooms to support equipment operation throughout all phases of an ELAP is requested.

ISECI RCIC Room Habitability and RHR/CS Room The NRC staff requests the licensee 3.2.4.6.B Habitability make available for audit the reactor building temperature calculation, MDQ0009992014000291. The resolution of this item is related to the successful closure of ISE Cl 3.2.4.2.A.

ISECI The Integrated Plan lacked information regarding The NRC staff requests the licensee 3.2.4.10.A battery availability, and lack of availability to review the make available for audit any battery load shed analysis, there is insufficient documentation to support the basis for information presented in the integrated plan to the minimum required voltage.

conclude that the requirements of NEI 12-06, Section 3.2.2, consideration 6, regarding load reduction to conserve de power will be implemented. During the audit process, the licensee provided a listing of the loads that would be part of the initial load shed that extended the battery availability to twelve hours. The licensee also stated that the shedding of these loads was determined to have no detrimental effects on unit safety and that the described load shedding would be included in a future revision to O-AOl-57-1 A, Blackout Station Procedure.

AQ35 Page 54 of the Integrated Plan indicates that items The NRC staff requests the licensee such as hoses will be deployed to the refuel floor early make available for audit the calculation in the ELAP event "to avoid having to access the refuel which supports the closure of this issue deck while the SFP is boiling." Additionally, pages 70 for review.

and 71 of the Integrated Plan provide a discussion of how ventilation will be achieved to avoid overpressurization in the SFP area. However, no analysis or technical justification was provided to verify the adequacy of the ventilation and/or support the conclusion of habitability in the spent fuel pool area.

Clarify whether any actions will be required in the SFP area and provide additional information relative to the ventilation and/or habitability analyses which provide the bases for concluding that the proposed strategies in this area will be successful as specified in NEI 12-06 Section 3.2.2 consideration 11.

AQ39 For RCIC room habitability in phases 1 and 2 and The NRC staff requests the licensee RHR/CS Room Habitability, page 66 of the Integrated make available for audit the calculation Plan discusses the use of ice vests or other personnel for the RCIC room for review.

protective measures implemented in accordance with Site Administrative and Safety Procedures and Processes. Several hours into an ELAP condition, the availability of ice vests may be questionable. Clarify what areas of the plant are anticipated to require personnel entry (including for the Alternate Phase 2 Strategy specified on page 29), what the anticipated habitability conditions will be in those environments at the anticipated time of personnel entry, and what personnel protection measures will be required to carry out those actions in high heat environments several hours into an ELAP.

AQ 52 Open Items 3.2.1.6.C and 3.2.1.6.D. The NRC staff requests the licensee make available for audit the sizing Provide a summary of the sizing calculation for the caclulation for the 4.16 kV FLEX FLEX generators to show that they can supply the generators that will be received from loads assumed in phases 2 and 3. the National SAFER Response Center.

OIP 01 #1 Flood and seismic re-evaluations pursuant to the 10 The NRC staff requests the licensee CFR 50.54{f) letter of March 12, 2012, are not make available for audit the flood and completed and therefore not assumed in this seismic hazard re-evaluations.

submittal. As the re-evaluations are completed, appropriate issues will be entered into the corrective action system and addressed OIP 01 #7 Browns Ferry will take actions as necessary to assure The NRC staff requests the licensee RCIC can operate at elevated temperatures. make available for audit the design of the hardened containment vent system, and the corresponding calculations and procedures.

SFPI RAI Please provide information indicating what will be the The NRC staff requests the licensee

1. 6 maximum expected ambient temperature in the room make available for audit calculation in which the associated transmitter (electronics MDQ0009992014000291, package) will be located under BOB [beyond-design- "Temperature Response of the Reactor basis] conditions in which there is no AC power Building Following an Extended Loss of available to run Heating Ventilation and Air AC Power."

Conditioning (HVAC) systems.

SE#4 Verification of phase rotation for FLEX DGs The NRC staff requests the licensee make available for audit information on how it will verify phase rotation after connecting the DGs to the equipment.

SE #6 Final SAFER Playbook The NRC staff requests the licensee make available for audit the Browns Ferry SAFER Response Plan after the document is finalized.

SE#? Browns Ferry Nuclear Plant plans to diagnose ELAP The NRC staff requests the licensee and complete the load shed for the three units within make available for audit validation of one hour of the initiation of the event. Please provide the ability to enter ELAP and complete validation that the timeframe that is required by the the ELAP load shed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of analysis is able to be completed by the operators. the event.

SE #8 The staff walked down the lay down site of the 480v The NRC staff walked down the lay generators. Due to the non-fully protected structures down site of the 480v generators. Due around the lay down area, the staff requests that to the non-fully protected structures Browns Ferry Nuclear Plant provide justification that around the lay down area, the NRC the 480v generator will be able to be positioned, staff requests the licensee make connecected and put into operation in the time available for audit justification that the required. This should include justification of the ability 480v generator will be able to be to clear debris that may be in the way, access to the positioned, connected and placed into hallway to connect to the cables, and the the voltage operation in the time required. The drop across the cables if a longer cable run was justification should include the ability to necessary would still be acceptable. clear debris that may be in the way, access to the hallway to connect to the cables, and any impact of the voltage drop across the cables if a longer cable run was necessary.

J. Shea If you have any questions, please contact me at 301-415-1924 or by e-mail at Tony.Brown@nrc.gov.

Sincerely, IRA/

Tony Brown, Project Manager Orders Management Branch Japan Lessons-Learned Division Office of Nuclear Reactor Regulation Docket Nos.: 50-259, 50-260, and 50-296

Enclosure:

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