ML18038B606

Report for the Audit of Licensee Responses to Interim Staff Evaluations Open Items Related to NRC Order EA-13-109

ML18038B606
Person / Time
Site:	Browns Ferry
Issue date:	02/21/2018
From:	Rajender Auluck Beyond-Design-Basis Engineering Branch
To:	James Shea Tennessee Valley Authority
Lee B
References
CAC MF4540, CAC MF4541, CAC MF4542, EA-13-109, EPID L-2014-JLD-0044
Download: ML18038B606 (30)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 21, 2018 Mr. Joseph W. Shea Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 4A Chattanooga, TN 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2 AND 3- REPORT FOR THE AUDIT OF LICENSEE RESPONSES TO INTERIM STAFF EVALUATIONS OPEN ITEMS RELATED TO NRC ORDER EA-13-109 TO MODIFY LICENSES WITH REGARD TO RELIABLE HARDENED CONTAINMENT VENTS CAPABLE OF OPERATION UNDER SEVERE ACCIDENT CONDITIONS (CAC NOS. MF4540, MF4541 AND MF4542; EPID L-2014-JLD-0044)

Dear Mr. Shea:

On June 6, 2013 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML13143A334), the U.S. Nuclear Regulatory Commission (NRC) issued Order EA-13-109, "Order to Modify Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Condition," to all Boiling-Water Reactor licensees with Mark I and Mark II primary containments. The order requirements are provided in to the order and are divided into two parts to allow for a phased approach to implementation. The order required licensees to submit for review overall integrated plans (OIPs) that describe how compliance with the requirements for both phases of Order EA-13-109 will be achieved.

By letter dated June 30, 2014 (ADAMS Accession No. ML14181B169), Tennessee Valley Authority (TVA, the licensee) submitted its Phase 1 OIP for Browns Ferry Nuclear Plant, Units 1, 2 and 3 (Browns Ferry). By letters dated December 19, 2014, June 29, 2014, December 29, 2015 (which included the combined Phase 1 and Phase 2 OIP), June 30, 2016, December 22, 2016, June 30, 2017, and December 20, 2017 (ADAMS Accession Nos. ML14353A428, ML15181A338, ML15365A554, ML16182A517, ML16357A577, ML17181A333, and ML17354A250, respectively), the licensee submitted its 6-month updates to the OIP. The NRC staff reviewed the information provided by the licensee and issued interim staff evaluations

( IS Es) for Phase 1 and Phase 2 of Order EA-13-109 for Browns Ferry by letters dated February 11, 2015 (ADAMS Accession No. ML14356A362), and September 6, 2016 (ADAMS Accession No. ML16244A762), respectively. When developing the ISEs, the staff identified open items where the staff needed additional information to determine whether the licensee's plans would adequately meet the requirements of Order EA-13-109.

The NRC staff is using the audit process described in letters dated May 27, 2014 (ADAMS Accession No. ML14126A545), and August 10, 2017 (ADAMS Accession No. ML17220A328),

to gain a better understanding of licensee activities as they come into compliance with the order.

As part of the audit process, the staff reviewed the licensee's closeout of the ISE open items.

J. Shea The NRC staff conducted a teleconference with the licensee on January 25, 2018. The enclosed audit report provides a summary of that aspect of the audit.

If you have any questions, please contact me at 301-415-1025 or by e- mail at Rajender.Auluck@nrc.gov.

Sincerely, Rajender Auluck, Senior Project Manager Beyond-Design-Basis Engineering Branch Division of Licensing Projects 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 AUDIT OF LICENSEE RESPONSES TO INTERIM STAFF EVALUATIONS OPEN ITEMS RELATED TO ORDER EA-13-109 MODIFYING LICENSES WITH REGARD TO RELIABLE HARDENED CONTAINMENT VENTS CAPABLE OF OPERATION UNDER SEVERE ACCIDENT CONDITIONS TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2 AND 3 DOCKET NOS. 50-259, 50-260 AND 50-296 BACKGROUND On June 6, 2013 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML13143A334), the U.S. Nuclear Regulatory Commission (NRC) issued Order EA-13-109, "Order to Modify Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Condition," to all Boiling-Water Reactor (BWR) licensees with Mark I and Mark II primary containments. The order requirements are divided into two parts to allow for a phased approach to implementation.

Phase 1 of Order EA-13-109 requires license holders of BWRs with Mark I and Mark II primary containments to design and install a Hardened Containment Vent System (HCVS), using a vent path from the containment wetwell to remove decay heat, vent the containment atmosphere (including steam, hydrogen, carbon monoxide, non-condensable gases, aerosols, and fission products), and control containment pressure within acceptable limits. The HCVS shall be designed for those accident conditions (before and after core damage) for which containment venting is relied upon to reduce the probability of containment failure, including accident sequences that result in the loss of active containment heat removal capability or extended loss of alternating current (ac) power (ELAP). The order required all applicable licensees, by June 30, 2014, to submit to the Commission for review an overall integrated plan (OIP) that describes how compliance with the Phase 1 requirements described in Order EA-13-109 Attachment 2 will be achieved.

Phase 2 of Order EA-13-109 requires license holders of BWRs with Mark I and Mark II primary containments to design and install a system that provides venting capability from the containment drywell under severe accident conditions, or, alternatively, to develop and implement a reliable containment venting strategy that makes it unlikely that a licensee would need to vent from the containment drywell during severe accident conditions. The order required all applicable licensees, by December 31, 2015, to submit to the Commission for Enclosure

review an OIP that describes how compliance with the Phase 2 requirements described in Order EA-13-109 Attachment 2 will be achieved.

By letter dated June 30, 2014 (ADAMS Accession No. ML141818169), Tennessee Valley Authority (TVA, the licensee) submitted its Phase 1 OIP for Browns Ferry Nuclear Plant, Units 1, 2 and 3 (Browns Ferry). By letters dated December 19, 2014, June 29, 2015, December 29, 2015 (which included the combined Phase 1 and Phase 2 OIP), June 30, 2016, December 22, 2016, June 30, 2017, and December 20, 2017 (ADAMS Accession Nos. ML14353A428, ML15181A338, ML15365A554, ML16182A517, ML16357A577, ML17181A333, and ML17354A250, respectively), the licensee submitted its 6-month updates to the OIP, as required by the order.

The NRC staff reviewed the information provided by the licensee and issued interim staff evaluations (ISEs) for Phase 1 and Phase 2 of Order EA-13-109 for Browns Ferry by letters dated February 11, 2015 (ADAMS Accession No. ML14356A362), and September 6, 2016 (ADAMS Accession No. ML16244A762), respectively. When developing the ISEs, the staff identified open items where the staff needed additional information to determine whether the licensee's plans would adequately meet the requirements of Order EA-13-109.

The NRC staff is using the audit process in accordance with the letters dated May 27, 2014 (ADAMS Accession No. ML14126A545), and August 10, 2017 (ADAMS Accession No. ML17220A328), to gain a better understanding of licensee activities as they come into compliance with the order. The staff reviews submitted information, licensee documents (via ePortals), and preliminary Overall Program Documents (OPDs)/OIPs, while identifying areas where additional information is needed. As part of this process, the staff reviewed the licensee closeout of the ISE open items.

AUDIT

SUMMARY

As part of the audit, the NRC staff conducted a teleconference with the licensee on January 25, 2017. The purpose of the audit teleconference was to continue the audit review and provide the NRC staff the opportunity to engage with the licensee regarding the closure of open items from the ISEs. As part of the preparation for these audit calls, the staff reviewed the information and/or references noted in the OIP updates to ensure that closure of ISE open items and the HCVS design are consistent with the guidance provided in Nuclear Energy Institute (NEI) 13-02, Revision 1 and related documents (e.g. white papers (ADAMS Accession Nos. ML14126A374, ML14358A040, ML15040A038 and ML15240A072, respectively) and frequently asked questions (FAQs), (ADAMS Accession No. ML15271A148)) that were developed and reviewed as part of overall guidance development. The NRC staff audit members are listed in Table 1.

Table 2 is a list of documents reviewed by the staff. Table 3 provides the status of the ISE open item closeout for Browns Ferry. The open items are taken from the Phase 1 and Phase 2 ISEs issued on February 11, 2015, and September 6, 2016, respectively.

FOLLOW UP ACTIVITY The staff continues to audit the licensee's information as it becomes available. The staff will issue further audit reports for Browns Ferry, as appropriate.

Following the licensee's declarations of order compliance, the licensee will provide a final integrated plan (FIP) that describes how the order requirements are met. The NRC staff will

evaluate the FIPs, the resulting site-specific OPDs, as appropriate, and other licensee documents, prior to making a safety determination regarding order compliance.

CONCLUSION This audit report documents the staff's understanding of the licensee's closeout of the ISE open items, based on the documents discussed above. The staff notes that several of these documents are still preliminary, and all documents are subject to change in accordance with the licensee's design process. In summary, the staff has no further questions on how the licensee has addressed the ISE open items, based on the preliminary information. The status of the NRC staff's review of these open items may change if the licensee changes its plans as part of final implementation. Changes in the NRC staff review will be communicated in the ongoing audit process.

Attachments:

1. Table 1 - NRC Staff Audit and Teleconference Participants

2. Table 2 - Audit Documents Reviewed

3. Table 3 - ISE Open Item Status Table

Table 1 - NRC Staff Audit and Teleconference Participants Title Team Member Oraanization Team Lead/Sr. Project Manager Rajender Auluck NRR/DLP Project Manager Support/Technical Support - Containment / Ventilation Brian Lee NRR/DLP Technical Support- Containment/

Ventilation Bruce Heida NRR/DLP Technical Support - Electrical Kerby Scales NRR/DLP Technical Support- Balance of Plant Kevin Roche NRR/DLP Technical Support - l&C Steve Wyman NRR/DLP Technical Support - Dose John Parillo NRR/DRA Attachment 1

Table 2 - Audit Documents Reviewed Caculation MDQ0000642015000351, "Hardened Contatinment System Operator Mission Dose,"

Revision 1 Calculation MDQ0009992014000291, "Temperatue Response of the Reactor Building Following an Extended Loss of AC Power," Revision 2 Procedure 1-EOI Appendix-13, "Emergency Venting Primary Containment," Revision 3 Procedure 2-EOI Appendix-13, "Emergency Venting Primary Containment," Revision 9 Calculation EDQ0009992013000202, "250 DC Unit Batteries 1, 2, & 3 Evaluation for Beyond Design Basis External Event (BDBEE) Extended Loss of AC Power (ELAP)," Revision 6 Procedure O-FSl-3F, "Load Shed of 250V Main Bank Battery 1, 2, 3," Revision 1 Calculation NDQ0000642015000341, "HCVS MAAP Analysis," Revision 0 Procedure O-FSl-4B, "FLEX Communication System Operation," Revision 0 Design Review Report- RAL-70181, Revision 1 - Size 14 Class 150 Wafer Butterfly Valve with Pneumatic Actuator Report No. S1620.0, Revision 1 - Seismic Test Report for an Optima Stantron Cabinet, Absopulse Voltage Converter, and Moore Industries Signal Converter Report No S1619.0, Revision O - Seismic Test Report for a Panasonic Laptop Computer, Matheson Pressure Gauge, Moore Industries Signal Converter Seismic Analysis SA-B150912-2, Revision A - Yi, ANSI Class 1500 Ball Valve of Stainless Steel Construction Schedule 80 Socket Weld Ends, Lever Operated Seismic Analysis SA-B150912-1, Revision A-Yi", ANSI Class 150 Ball Valve of Stainless Steel Construction Schedule 40 Socket Weld Ends, Lever Operated Report No. S1615.0, Revision 1 - Seismic Test Report for an Eaton Circuit Breaker Panel Report No. S1607.0, Revision 1 - Seismic Test Report for an EIZO FlexScan Monitor, Stealth Panel Mount Keyboard, Eaton Disconnect Switches, and Ruhrpumpen/Murphy Controller Panel Test Report No. PR034998SEI-TR16 - Seismic Qualification Test Report for Yokogawa YS1700 Programmable Indicating Controllers Seismic Analysis Report for Yi Inch - ANSI Class 150 Shuttle Valve Design and Seismic Analysis Report for 14 Inch -ANSI Class 150 DRV-Z Nozzle Check Valve Qualification Summary Report 04518900-QSR, Revision B - HCVS Radiation Monitoring System (DC & AC Input Power Supplies)

Qualification Report 04502054-QR, Revision B - 125 voe Input Filter and 24V and 125 voe Input Power Supplies for HCVS RM-1000 System Qualification Report QR-351025195-1, Revision 2 - Battery Charger Test Report AZZ PR051230 QP-351025195 Electromagnetic Interference and Suceptibility Tests Summary of Test Report 06-8680-003, Revision 1 - Nuclear Component Qualification Test Report for the Generic Qualification of Weed Instrument Company Temperature Sensor Assemblies Calculation NDQ0999910030, "Summary of Mild Environmental Conditions for Browns Ferry Nuclear Plant," Revision 11 Calculation MDQ0009992014000447, "Temperatue Response of the Reactor Building Following a Fukushima Type Severe Accident Utilizing the HCVS," Revision 0 Calculation MDQ0000322015000347, "HCVS Nitrogen Sizing Analysis," Revision 1 Calculation MDQ0003602014000222, "BFN ELAP Transient Analysis," Revision 5 Attachment 2

Calculation MDQ0031930018, "BFN Control Bay, Elevation 593.0' and 606.0', and Electric Board Room Analysis," Revision 27 Calculation MDQ0030880213, "Unit 1 and Unit 2 DGB - Central Diesel Information Center Ventilation Requirements," Revision 7 Calculation MDQ0030880208, "U3 DGB Battery Room Ventilation Requirements," Revision 7 Calculation MDN0009992012000027, "Thermal Analysis of Control Bay Rooms, Unit 3 Diesel Generator Building Shutdown Board Rooms and Battery #4 Board Room Following Loss of Cooling," Revision 3 Calculation EDQ3030880318, "Electrical Heat Losses - Zones 13, 14, 15, 16," Revision 12 Calculation EDQ3030880319, "Electrical Heat Losses -Zones 11 and 12," Revision 9 Calculation EDQ0030910058, "Electrical Equipment Heat Losses - Individual Rooms in Units 1, 2 and 3," Revision 45 Calculation EDQ0010642015000349, "Unit 1 HCVS Electrical Design & Equipment Sizing Analysis," Revision 2 Calculation EDQ0000642016000510, "Unit 2 & 3 HCVS DC Electrical Design & Equipment Sizing Analysis," Revision 0 Design Change Technical Evaluation - DCN 71391 (Unit 3)

Design Change Technical Evaluation - DCN 71390 (Unit 2)

Design Change Technical Evaluation - DCN 71389 (Unit 1)

BFN White Paper R06 160315 491 -Validation of NEI White Paper HCVS-WP-04 First Assumption for Missile Protection of Hardened Containment Vent System at Browns Ferry Nuclear Plant Calculation MDN0003602014000233, "Hydraulic Analysis for Fukushima FLEX Connection Modifications," Revision 3 Calculation MDQ0000642015000393, "HCVS Equipment Dose Evaluation," Revision 2 AREVA Document 51-9262174-003- Projected Dose Rate Contour Map of Shine from the HCVS Vent Line Extending Above Refueling Floor (BFNP)

BWROG-TP-008, "Severe Accident Water Addition Timing" BWROG-TP-011, "Severe Accident Water Management Supporting Evaluations"

Browns Ferry Nuclear Plant, Units 1, 2 and 3 Vent Order Interim Staff Evaluation Open Items:

Table 3 - ISE Open Item Status Table ISE Open Item Number Licensee Response - Information NRC Staff Close-out notes Safety Evaluation (SE) provided in 6 month updates and on the status Requested Action ePortal Closed; Pending; Open (need additional information from licensee)

Phase 1 ISE 01 1 An evaluation of temperature and The NRC staff reviewed the Closed radiological conditions was performed to information provided in the 6-Make available for NRC staff ensure that Operating personnel can month updates and on the [Staff evaluation to be audit an evaluation of safely access and operate controls at the ePortal. included in SE Sections temperature and radiological Remote Operating Station located in the 3.1.1.2 and 3.1.1.3]

conditions to ensure that Diesel Buildings and in the Reactor MDQ0003602014000222, "BFN operating personnel can safely building. This evaluation is documented ELAP Transient Temperature access and operate controls in Unit 1 [Design Change] DCN 71389 Analysis" shows with and support equipment. Design Change Technical Evaluation compensatory actions the main (Page 70-73 of 81 ), Unit 2 DCN 71390 control rooms (MCRs) for U-1, U-Design Change Technical Evaluation 2, & U-3 remain below 110°F at (Page 65-68 of 75), and Unit 3 DCN 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

71391 Design Change Technical Evaluation (Page 60-63 of 69). MDQ0030880213, "Unit 1 and MDQ0000642015000351 "HCVS Unit 2 DGB - Central Diesel OPERATOR (MISSION) DOSE Information Center Ventilation CALCULATION", and Requirements," determines MDQ0009992014000291 ventilation requirements for the "TEMPERATURE RESPONSE OF THE remote operation station (ROS)

REACTOR BUILDING FOLLOWING AN area during normal operation.

EXTENDED LOSS OF AC POWER" were This calculation is for determining used to validate the evaluation. normal ventilation requirements.

During ELAP, normal ventilation is not available. However, prior to an ELAP, there is no normal operating equipment which will provide a residual heat load.

Given the mass of concrete construction in the area of the ROS alonq with no major Attachment 3

electrical heat loads or residual heat loads from operating equipment, area temperatures will not be adverse to operators performing their required actions in the ROS.

MDN0009992012000027,

'Thermal Analysis of Control Rooms, Unit 3 DGB SDBRs and BR4BR Following Loss of Cooling," evaluated the Unit-3 ROS in the Unit-3 DG [diesel generator] Shutdown Board Room (SDBR). Temperature remains below 110°F.

MDQ0000642015000351, "HCVS Operator (Mission) Dose Calculation," was performed to determine the integrated radiation dose due to HCVS operation.

Temperature and radiological conditions should not inhibit operator actions needed to initiate and operate the HCVS during an ELAP with severe accident conditions.

No follow-up questions.

Phase 1 ISE 01 2 1-EOI-Appendix 13 revised to include The NRC staff reviewed the Closed venting for loss of DC [direct current] information provided in the 6-Make available for NRC audit power. Revision 3 issued on 10/19/2016. month updates and on the [Staff evaluation to be documentation that procedure ePortal. included in SE Section 1/2/3-EOI Appendix 13 has 2-EOI-Appendix 13 revised to include 5.1]

been revised to include venting for loss of DC power. Revision 9 The guidelines and procedures venting for loss of de power. issued on 3/24/2017. for HCVS operation are complete for Units 1 & 2 and consistent with the guidance in NEI 13-02. -

3-EOI-Appendix 13 will be revised following completion of the installation of The Unit 3 procedure for HCVS Unit 3 Hardened Containment Vent operation will be revised following System. completion of the installation of Unit 3 HCVS and will follow the same guidance as the other two units, consistent with the guidance in NEI 13-02.

No follow-up questions.

Phase 1 ISE 01 3 Calculation EDQ0009992013000202, The NRC staff reviewed the Closed 250V DC Unit Batteries, 1, 2, & 3 information provided in the 6-Make available for NRC staff Evaluation for the Beyond Design Basis month updates and on the [Staff evaluation to be audit documentation External Event (BDBEE) Extended Loss ePortal. included in SE Section demonstrating that all load of AC Power (ELAP), has been issued to 3.1.2.6]

sheds will be accomplished determine load shedding impact on the EDQ0009992013000202, "250V within one hour of event unit batteries. The performance of the DC Unit Batteries, 1, 2, & 3 initiation and will occur in an load shed is directed by O-FSl-1 (Page 1) Evaluation for the Beyond Design area not impacted by a "FLEX Support Instruction" and performed Basis External Event (BDBEE) possible radiological event. in accordance with O-FSl-3F "Load Shed Extended Loss of AC Power of 250V Main Bank Battery 1,2,3." The (ELAP)," Rev. 6. Coping time of load shed is performed in the Control Bay Unit batteries extended from 8 and Electrical Board rooms only and will hours to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

not require entry into areas that are impacted by a possible radiological event. FLEX procedure O-FSl-3F, "Load Shed of 250V Main Bank Battery 1, 2, 3," Rev. 1. The load shed is performed in the Control Bay and Electrical Board rooms and will not require entry into areas that are impacted by a possible radiological event.

No follow-up questions.

Phase 1 ISE 01 4 A conceptual meeting was held in The NRC staff reviewed the Closed November 2014, and a staging plan was information provided in the 6-Make available for NRC staff used to separate the existing HWWV from month updates and on the [Staff evaluation to be audit documentation that the HCVS. The Hardened Containment ePortal. included in SE Section demonstrates that operating Vent System has been implemented on 3.1.2.3]

units that have not Unit 1 and Unit 2. Unit 3 remains capable

implemented the order will be of using the existing HWWI/ system until Units are not interconnected.

able to vent through the completion of the Hardened Containment There are no cross-ties between existing vent system Vent System which will be performed in units.

unaffected by the the Spring of 2018.

implementation of HCVS on No follow-up questions.

other units.

Phase 1 ISE 01 5 The existing wetwell vent and the HCVS The NRC staff reviewed the Closed have been designed for 1 percent of rated information provided in the 6-Make available for NRC staff thermal power at extended power uprate month updates and on the [Staff evaluation to be audit analyses demonstrating (EPU) (3952 MWt) conditions. This ePortal. included in SE Section that HCVS has the capacity to analysis is available and documented in 3.1.2.1]

vent the steam/energy Calculation NDQ0000642015000341. NDQ0000642015000341, "HCVS equivalent of one percent of MAAP Analysis" uses the licensed/rated thermal power Computer Code MAAP, Version (unless a lower value is 4. 0. 7 to confirm the vent design.

justified), and that the The HCVS can be opened at 56 suppression pool and the pounds per square inch gauge HCVS together are able to (psig) and closed at a lower absorb and reject decay heat, pressure.

such that following a reactor shutdown from full power Based on the preliminary study, a containment pressure is vent size of 14" Schedule 40 was restored and then maintained chosen.

below the primary containment design pressure and the At 70.7 per square inch absolute primary containment pressure (psia) (56psig), the corresponding limit. required rate of flow is: 1%

• Rated Power /(hg(@70.7psia) =

0.01 *3952MW (947.817BTU /(s.

mega watt(MW)) (lbm/1101 BTU)

= 31.7 lbm/sec Upon examination of the GOTHIC analysis (Reference (12]), it can be seen that GOTHIC calculates the HCVS flow rate capability at this pressure (70.7 psia, design pressure of the drywell and wetwell vent conditions) to be 58.4 lbm/sec. Since the

calculation satisfying the regulatory requirement set forth in EA 13-109 (1 % decay heat removal capability requirement) requires a minimum of 31.7 lbm/sec, the current design of the HCVS safely satisfies this requirement.

No follow-up questions.

Phase 1 ISE 01 6 A communication system has been The NRG staff reviewed the Closed implemented (DCN 70852) that uses information provided in the 6-Make available for NRG staff hand held radios for communication month updates and on the [Staff evaluation to be audit documentation that between the MGR and the ROS. This ePortal. included in SE Section demonstrates adequate Radio System consists of a UHFNHF 3.1.1.1]

communication between the trunked system and an independent VHF The communication methods are remote HCVS operation channel (F4). The In-plant Radio System the same as accepted in Order locations and HCVS decision is accessed by handheld radios. The In- EA-12-049.

makers during ELAP and plant Radio System has normal and severe accident conditions. emergency diesel generator backed No follow-up questions.

power supply. The radio system is powered from two Class 1E redundant power sources, 480V DG Auxiliary Boards A and B. Primary power source will be from 480V DG Auxiliary Board A via second 480-208/120 transformer /

distribution center. In the event of loss of primary power source, power to radio equipment will be automatically transferred to backup source via transfer switches located in each cabinet, with exception of cabinet 4, which receives power via cabinet 1 transfer switch.

Backup power source includes UPS with battery capacity to supply four (4) UHF channels for three hours. Therefore in this configuration capacity is reduced from five simultaneous conservations to three.

The loads supplied via UPS can be

alternatively supplied from portable generator via transfer switch (O-FSl-48).

UPS Conservation can be accomplished by switching off one of the two UPSs until such time the active UPS reaches "low level". Then the UPS previously switched off can be returned to service extending the overall time the radio system can remain operable without portable generator power to approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

BFN maintains a large number of handheld radios, batteries and charging units. The FLEX program does not maintain dedicated handheld radios.

These units, spare batteries and chargers will be gathered if not readily available in the control rooms.

Handheld Radios can additionally be operated in "Radio-to-Radio" mode enabling communications not affected by shielding or distance.

Phase 1 ISE 01 7 An evaluation was performed and The NRG staff reviewed the Closed concluded that the containment isolation information provided in the 6-Make available for NRG staff valves will open under the maximum month updates and on the [Staff evaluation to be audit documentation of an expected differential pressure and is ePortal. included in SE Section evaluation verifying the documented in FLOWSERVE Report 3.2.1]

existing containment isolation RAL-70181, Design Review Report of RAL-70181 Rev 1 Design Review valves, relied upon for the Size 14 Class 150 Wafer Butterfly Valve Report "Size 14 Class 150 Wafer HCVS, will open under the with Pneumatic Actuator Rev. 1. Butterfly Valve with Pneumatic maximum expected differential Actuator Drawing: 94-15972,"

pressure during BDBEE and shows a maximum stem torque is severe accident wetwell 6000 in-lbf. Maximum pressure venting. differential= 70.7 psi. The operating torque at the seat is expected to increase approximately 11 % due to the

increase in differential pressure from 56 psi to 70.7 psi. Using current methods and parameters the calculated required torque to start open increases from 4944 in-lbs to 5508 in-lbs. The actuator is a Bettis model NCB725-SR80-MCW pneumatic quarter turn actuator with an internal coil spring to fail close and air pressure to open. The actuator start to open output torque varies from 5064 in-lbs at 70 psig actuator air pressure to 6395 in-lbs at 80 psig air pressure. An air pressure of 75 psig will provide a start to open torque of approximately 5730 in-lbs sufficient to open the valve at the higher differential pressure.

No follow-up questions.

Phase 1 ISE 01 8 Seismic/Qualification Reports for Units 1, The NRC staff reviewed the Closed 2, & 3 HCVS Components tables, file information provided in the 6-Make available for NRC staff HCVS Phase 1 ISE 01-8, and documents month updates and on the [Staff evaluation to be audit documentation of a are attached. From the tables, use the ePortal. included in SE Section seismic qualification evaluation Seismic Report# column to find the 3.2.2]

of HCVS components. reports. The licensee provided several reports which demonstrate the seismic adequacy of the HCVS components required for HCVS venting remain functional following a design-basis earthquake.

No follow-up questions.

Phase 1 ISE 01 9 Seismic/Qualification Reports for Units 1, The NRC staff reviewed the Closed 2, & 3 HCVS Components tables, file information provided in the 6-Make available for NRC staff HCVS Phase 1 ISE 01-9. From the month updates and on the audit descriptions of all ePortal.

instrumentation and controls tables, use the Qualification Report# [Staff evaluation to be (existing and planned) column to find the reports. The existing plant instuments included in SE Section necessary to implement this required for HCVS (i.e. wetwell 3.1.2.8]

order including qualification level instruments and drywell methods. pressure instruments) meet the requirements of Regulatory Guide (RG) 1.97.

The HCVS components tables and associated qualification reports provide the qualifications for new HCVS l&C components.

The staff's review indicated that the qualification met the order requirements.

No follow-up questions.

Phase 1 ISE 01 10 The DCN determined that only three The NRC staff reviewed the Closed components per unit of the HCVS would information provided in the 6-Make available for NRC staff experience harsh post-event month updates and on the [Staff evaluation to be audit the descriptions of local environmental conditions (high ePortal. included in SE Section conditions (temperature, temperatures and humidity). These three 3.1.1.4]

radiation and humidity) components were procured with the The DCN discusses the anticipated during ELAP and requirements specified and the vendors environmental conditions during severe accident for the provided qualification reports an accident at the locations components (valves, documenting the capabilities of each containing instrumentation and instrumentation, sensors, component. controls (l&C) components. The transmitters, indicators, staff's review indicated that the electronics, control devices, BFN HCVS components subjected to post environmental qualification met and etc.) required for HCVS event harsh environments were specified the order requirements.

venting including confirmation and procured with capabilities exceeding that the components are those requirements. No follow-up questions.

capable of performing their functions during ELAP and severe accident conditions.

Phase 1 ISE 01 11 From DCN 71389 (Unit 1) Design Change The NRC staff reviewed the Closed Technical Evaluation (Page 36 of 81): information provided in the 6-Make available for NRC staff BFN Calculation EDQ0010642015000349 month updates and on the [Staff evaluation to be audit the final sizing evaluation was performed to size and perform ePortal. included in SE Section for HCVS batteries/battery electrical analysis on the HCVS batteries 3.1.2.6]

charger including incorporation and battery charger. BFN Calculation The licensee stated that all into FLEX DG loading EDQ0010642015000349 also performs electrical power required for calculation. electrical analysis on the new 250 voe operation of HCVS components is

[volt direct current] power Distribution provided by the 250 voe Panel (1-BDDD-064-0001), the new 250V battery/battery charger.

DC to DC Converter (1-CNV-064-0001),

and determines parameters for isolation The battery sizing calculations diodes 1-Dl0-064-221A, 1-DI0-064- (EDQ0010642015000349 - Unit 1 221 B, 1-D10-064-222A, and 1-010-064- & EDQ0000642016000510 - Unit 2228. Battery sizing was performed in 2 and 3) confirmed that the HCVS accordance with IEEE [Institute of batteries have a minimum Electrical and Electronics Engineers] Std. capacity capable of providing 485-2010 and electrical equipment sized power for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without in accordance with IEEE Std. 946-2004. recharging, and therefore is adequate.

From DCN 71390 (Unit 2) Design Change Technical Evaluation (Page 35 of 75): There is no incorporation required into the FLEX DG loading BFN Calculation EDQ0000642016000510 calculation for any Unit due to was performed to size and perform there are no plans or electrical analysis on the HCVS batteries requirements to recharge the and battery charger. BFN Calculation HCVS battery after depletion.

EDQ0000642016000510 also performs The HCVS electrical loads would electrical analysis on the new 250 VDC be aligned back to their normal power Distribution Panel (0-8000-064- power supply which is the Unit 0001 ), the new 250V DC to DC Converter Battery.

(2-CNV-064-0001 ), and determines parameters for isolation diodes (2-DIO- No follow-up questions.

064-221A, 2-010-064-221 B, 2-DI0-064-222A, and 2-010-064-2228). Battery sizing was performed in accordance with IEEE Std. 485-2010 and electrical equipment sized in accordance with IEEE Std. 946-2004. Note the HCVS battery and battery charger installed by this DCN have been sized to accommodate both Unit 2 and Unit 3 HCVS electrical loads.

From DCN 71391 (Unit 3) Design Change Technical Evaluation (Page 32 of 69):

BFN Calculation ED00000642016000510 was performed to size and perform electrical analysis on the HCVS batteries and battery charger. BFN Calculation EDQ0000642016000510 also performs electrical analysis on the new 250 VDC power Distribution Panel (0-BDDD-064-0001 ), the new 250V DC to DC Converter (3-CNV-064-0001 ), and determines parameters for isolation diodes (3-DIO-064-221A, 2-010-064-221 B, 3-010-064-222A, and 3-Dl0-064-222B). Battery sizing was performed in accordance with IEEE Std. 485-2010 and electrical equipment sized in accordance with IEEE Std. 946-2004. Note the HCVS battery and battery charger installed by this DCN have been sized to accommodate both Unit 2 and Unit 3 HCVS electrical loads.

There is no incorporation required into the FLEX DG loading calculation for any Unit due to there are no plans or requirements to recharge the HCVS battery after depletion. The HCVS electrical loads would be aligned back to their normal power supply which is the Unit Battery.

Phase 1 ISE 01 12 Evaluation has been completed and The NRC staff reviewed the Closed documented in DCN 71389 for Unit 1, information provided in the 6-Make available for NRC staff DCN 71390 for Unit 2, DCN 71391 for month updates and on the [Staff evaluation to be audit documentation of the Unit 3 and calculation ePortal. included in SE Section HCVS nitrogen pneumatic MDQ0000322015000347 "HCVS 3.1.2.6]

system design including sizing NITROGEN SIZING ANALYSIS". As The DCN and calculation and location. documented in DCN 71389 Design M DQ000032201500034 7 Change Technical Evaluation (Page 29 of discusses the pneumatic design

81) there are 9 Nitrogen Cylinders and sizing. DCN 71389, DCN required for Unit 1 for 7 days of Hardened 71390, and DCN 71391 Vent operation. There are 5 Nitrogen discusses the required number of Cylinders installed to support Hardened nitrogen cylinders needed for vent

Vent operation for Unit 1. As documented operation for sustained operation in DCN 71390 Design Change Technical for each unit, respectively. The Evaluation (Page 28 of 75) there are 10 number of nitrogen cylinders Nitrogen Cylinders required for Unit 2 for installed in each unit and 7 days of Hardened Vent operation. As available are sufficient to operate documented in DCN 71391 Design the HCVS for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Change Technical Evaluation (Page 26 of

69) there are 9 Nitrogen Cylinders No follow-up questions.

required for Unit 3 for 7 days of Hardened Vent operation. (24 Nitrogen Cylinders are required for Unit 2 and 3 for 7 days simultaneous operation. There are 5 Nitrogen Cylinders installed to support Hardened Vent operation for Unit 2 and 3.

There are 6 Nitrogen Cylinder carts with 6 Nitrogen Cylinders on each cart available in the FLEX Storage building with no other committed use of them.

Phase 1 ISE 01 13 Tornado and seismic missile criteria are The NRC staff reviewed the Closed located in System 64A Design Control information provided in the 6-Make available for NRC staff Document (DCD). As part of DCN 71389 month updates and on the [Staff evaluation to be audit the seismic and tornado (Unit 1) and DCN 71390 (Unit 2) the DCD ePortal. included in SE Section missile final design criteria for was revised. For DCN 71391 (Unit 3) a 3.2.2]

the HCVS stack. markup reflecting these changes has DCN 71389, DCN 71390, and been generated and will be incorporated DCN 71391 addresses the HCVS in the next revision of the DCD per TVA seismic qualification and tornado process. missile design for each unit, respectively. .

As stated in the OIP the HCVS design provides missile protection from ground The licensee evaluated the entire level to 30 feet in accordance with NRC HCVS system to Seismic Regulatory Guide 1. 76 based on a site- Catergory I, which is consistent specific tornado missile evaluation. with the plants seismic design-Above 30 feet the exposed vent piping will basis.

be robustly designed in accordance with HCVS-WP-04. This is a design For the tornado missile design, consideration using reasonable protection the licensee's design is consistent features for the screened in hazards from with the endorsed white paper NEI 12-06. (reference HCVS-FAQ-04; HCVS-WP-04 and meets all of HCVS-WP-04). Browns Ferry has utilized the aoolicable tornado missile

NEI HCVS-WP-04 "Missile Evaluation for assumptions identified in the HCVS Components 30 Feet Above white paper.

Grade" and meet the assumptions in NEI HCVS-WP-04 as stated below: No follow-up questions.

1. Piping and components external to any missile-protected structure and less than 30 feet above grade are evaluated and, unless otherwise justified in plant-specific OIPs, protected from large and small missiles.

How met: The various BFN site areas were reviewed for their potential to create missiles, defined by NRC Regulatory Guide 1. 76, Revision 1, dated March 2007, which may strike unprotected HCVS piping and components located less than 30 feet above grade. The review was performed to validate the first assumption from NEI White Paper HCVS-WP-04. It has been determined that it is not credible that any tornado borne commodities within the scope of the first assumption will strike and jeopardize function of the HCVS. This review and conclusions are documented in BFN White Paper "Validation of NEI White Paper HCVS-WP-04 First Assumption for Missile Protection of Hardened Containment Vent System at Browns Ferry Nuclear Plant"

2. Piping and components external to any missile-protected structure and greater than 30 feet above grade conform to the following:

a. The target area of the HCVS components is less than 300 ft2,

How met: As stated in DCN 71389 (Unit

1) DCN 71390 (Unit 2) and DCN 71391 (Unit 3) DCD 7064A Mark-up the HCVS pipe is 14 inch diameter and has non-protected pipe runs not in excess of 250' which results in a potential missile target area less than the 300 ft2 limit specified in HCVS-WP-04

b. The size and robustness of the exposed HCVS s are substantial (e.g.,

steel piping versus small tubing or plastic piping),

How met: As stated in DCN 71389 (Unit

1) DCN 71390 (Unit 2) and DCN 71391 (Unit 3) DCD 7064A Mark-up the HCVS piping is constructed of 14" schedule 40 carbon steel piping providing substantial robustness, thereby satisfying HCVS-WP-04.

c. There is no source of obvious potential missiles in the proximity of the exposed HCVS components (such as an unrestrained material lay down area).

How met:The Reactor Building roof contains a limited amount of rigidly mounted permanent plant equipment with non-anchored temporary equipment being stored on the roof at a low frequency.

Therefore it is not credible that an item located on the roof of the Reactor Building will cause loss of function of the HCVS piping. Signs have been posted on the south side of the Reactor Building in area of exposed HCVS piping to prohibit the storage or placing of equipment within the

proximity of the exposed HCVS piping and this restriction is documented in O-TPP-ENG-632(Bases) Diverse and Flexible Coping Strategies (FLEX)

Program Bases Document.

3. Licensees consider guidance in FLEX, or other procedures, to restore venting capability in the event the HCVS is damaged. Restoration could include cutting pipe below damaged section. This location may have to be below the release height requirements otherwise imposed.

How met: As stated in DCN 71389 (Unit

1) DCN 71390 (Unit 2) and DCN 71391 (Unit 3) DCD 7064A Mark-up appropriate assessment and restoration of venting capacity in the event the HCVS is damage is conducted under procedure O-FS1-6A "Damage Assessment." The HCVS piping was designed to provide a Tee in the piping in the event that the piping is damaged and will require cutting.

4. Licensees verify that if hurricanes are screened in for FLEX (see NEl-12-06},

that the site procedures recommend a plant shut down prior to hurricane arrival on-site.

As stated in DCN 71389 (Unit 1) DCN 71390 (Unit 2) and DCN 71391 (Unit 3)

DCD 7064A Mark-up in accordance with NEI 12-06, BFN is not susceptible to winds exceeding 130 mph from hurricanes. As such, Browns Ferry Nuclear Plant screens out for hurricanes and HCVS-WP-04 assumption 4 does not a_EQly.

Phase 1 ISE 01 14 A description of the final design of the The NRC staff reviewed the Closed HCVS to address hydrogen detonation information provided in the 6-Provide a description of the and deflagration is contained in the month updates and on the [Staff evaluation to be final design of the HCVS to Design Change Technical Evaluation ePortal. included in SE Section address hydrogen detonation (section 6.4.8) of Unit 1 DCN 71389, Unit 3.1.2.11]

and deflagration. 2 DCN 71390, and Unit 3 DCN 71391. The licensee's design concept for use of the check valve is that after venting, steam and Hydrogen/CO are isolated in the pipe volume between the downstream check valve and the upstream primary containment isolation valve. As the vented steam cools a vacuum forms in the HCVS piping. The check valve prevents this vacuum from pulling oxygen into the vent pipe and creating a combustible mixture. For this approach check valve 1(2,3)-CKV-064-0802 is located less than the run-up distance to DDT which is determined by 30 LID (NEI 13-02, Ref 4.1.6). Based on the internal pipe diameter of 14" schedule 40 piping, the check valve must be located within 35 feet of the vent opening which is located at the 741 '-6" elevation. The location of this check valve is at approximately elevation 722' just above the Unit 1(2,3) Reactor Building roof, approximately 20 feet from the vent opening, which is less than the minimum run-up distance to DDT.

The design also addressed the potential for oxygen entering a condensate drain downstream of check valve 1(2,3)-FCV-064-022.

The drain line has two check valves in series which will greatly restrict in leakage of oxygen ECP 71389, 71390, and 71391 indicates that HCVS piping for each unit is routed independently up the southern exterior wall of the Reactor Building.

The licensee's design is consistent with Option 5 of the endorsed white paper HCVS-WP-03.

No follow-up questions.

Phase 1 ISE 01 15 A description of the strategies for The NRC staff reviewed the Closed hydrogen control that minimizes the information provided in the 6-Provide a description of the potential for hydrogen gas migration and month updates and on the [Staff evaluation to be strategies for hydrogen control ingress into the reactor building or other ePortal. included in SE Section that minimizes the potential for buildings is contained in the Design 3.1.2.12]

hydrogen gas migration and Change Technical Evaluation (section ECP 71389, 71390, and 71391 ingress into the reactor 6.4.8) of Unit 1 DCN 71389, Unit 2 DCN indicate that HCVS piping for building or other buildings. 71390, and Unit 3 DCN 71391. each unit is routed independently up the southern exterior wall of the Reactor Building.

No follow-up questions.

Phase 1 ISE 01 16 A description of the design details that The NRC staff reviewed the Closed minimize unintended cross flow of vented information provided in the 6-Provide design details that fluids within a unit and between units on month updates and on the [Staff evaluation to be minimize unintended cross the site is contained in the Design ePortal. included in SE Section flow of vented fluids within a Change Technical Evaluation (section 3.1.2.3]

unit and between units on the 6.4.10) of Unit 1 DCN 71389, Unit 2 DCN The licensee's design appears to site. 71390, and Unit 3 DCN 71391. minimize the unintended cross flow of vented fluids.

No follow-up questions.

Phase 2 ISE 01 1 Hydraulic Analysis calculation The NRC staff reviewed the Closed MDN0003602014000233 was revised to information provided in the 6-include a boundinQ case that concluded

Licensee to perform a that a single FLEX pump (with booster month updates and on the [Staff evaluation to be hydraulic evaluation to ensure pump) can provide 500 gpm [gallons per ePortal. included in SE Section flow adequacy can be met for minute] to Unit 1 RPV, 500 gpm to Unit 2 4.1.1.2]

all 3 units using 1 FLEX pump RPV, and 500 gpm to Unit 3 RPV (each at The hydraulic analysis, to support SAWA flow RPV pressure of 106 psig) in response to "MDN0003602014000233" shows requirement. a SAWA [severe accident water addition] that the required SAWA flowrate event. of 500 gpm for each unit, is within the capacity of the single FLEX pump (with booster pump).

No follow-up questions.

Phase 2 ISE 01 2 Eguigment and Controls The NRC staff reviewed the Closed information provided in the 6-Licensee to evaluate the Plant instrumentation for SAWM [severe month updates and on the [Staff evaluation to be SAWA [severe accident water accident water management] that is ePortal. included in SE Sections addition] equipment and qualified to RG 1.97 or equivalent is 4.5.1.2 and 4.5.1.3]

controls, as well as ingress considered qualified for the sustained AREVA document 51-9262174-and egress paths for the operating period without further 003 shows that radiological expected severe accident evaluation. The following plant conditions should not inhibit conditions (temperature, instruments are qualified to RG 1.97: operator actions or SAWA humidity, radiation) for the equipment and controls needed to sustained operating period. OW Pressure 1,2,3-Pl-64-678 initiate and operate the HCVS during an ELAP with severe Suppression Pool Level 1,2,3-Ll-64-159A accident conditions.

Passive components that do not need to The temperature evaluation change state after initially establishing addressed in Phase 1 Open Item SAWA flow do not require evaluation #1 bounds the SAWNSAWM beyond the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, at which time operation.

they are expected to be installed and ready for use to support SAWNSAWM. No follow-up questions.

The following additional equipment performing an active SAWNSAWM function is considered:

SAWA/SAWM flow instrument.

SAWA/SAWM pump

FLEX generator SAWA Throttle valve These components will be used at a remote location (outside reactor building) and have been evaluated for the environmental conditions applicable at those locations.

Ingress and Egress A quantitative evaluation of expected dose rates AREVA document 51-9262174-003 "Projected Dose Rate Contour Map of Shine from the HCVS Vent Line Extending Above Refueling Floor (BFNP)" has been performed per HCVS-WP-02 and found the dose rates at deployment locations including ingress/egress paths are acceptable.

Phase 2 ISE 01 3 The SAWA throttle valve will be used at a The NRC staff reviewed the Closed remote location (outside reactor building) information provided in the 6-Licensee to demonstrate how and has been evaluated for the month updates and on the [Staff evaluation to be SAWA flow is capable to environmental conditions applicable at ePortal. included in SE Section perform its intended function that location. 4.5.1.2]

for the sustained operating AREVA document 51-9262174-period under the expected A validation was performed on the SAWA 003 shows that radiological temperature and radiological flowmeter that determined thru discussion conditions should not inhibit the conditions. with the vendor (Fire Research operation of the SAWA throttle Corporation) that the ambient temperature valve during an ELAP with severe limits of the flowmeter are a minimum of - accident conditions.

20 degrees F and a maximum temperature of 150 degrees F when an The temperature evaluation external power source is being used. addressed in Phase 1 Open Item Browns Ferry Nuclear Plant has a #1 bounds the location of the minimum ambient extreme daily SAWA throttle valve.

temperature record of -12 degrees F and a maximum ambient extreme daily No follow-up questions.

temperature of 108 degrees Fas

documented in the EA-12-049 Overall Integrated Plan. The BFN Flowmeter is using external batteries (FLEX pump diesel) to power flowmeter which allow operations within the extreme daily temperatures of -12 degrees F and 108 Deqrees F.

Phase 2 ISE 01 4 The wetwell vent has been designed and The NRC staff reviewed the Closed installed to meet NEI 13-02 Rev 1 information provided in the 6-Make available for NRC staff guidance which will ensure that it is month updates and on the [Staff evaluation to be supporting documentation adequately sized to prevent containment ePortal. included in SE Section demonstrating that overpressure under severe accident 4.2]

containment failure as a result conditions. BWROG-TP-15-008 of overpressure can be demonstrates adding water to the prevented without a drywell The SAWM strategy will ensure that the reactor vessel within 8-hours of vent during severe accident wetwell vent remains functional for the the onset of the event will limit the conditions. period of sustained operation. Browns peak containment drywell Ferry Nuclear Plant will follow the temperature significantly reducing guidance (flow rate and timing) for the possibility of containment SAWA/SAWM described in [Boiling-Water failure due to temperature.

Reactors Owners Group] BWROG-TP Drywell pressure can be 008 and BWROG-TP-15-011. These controlled by venting the documents have been posted for NRC suppression chamber through the staff review. The wetwell vent will be suppression pool.

opened prior to exceeding the PCPL

[Primary Containment Pressure Limit] BWROG-TP-15-011 value of 62 PSIG. Therefore, demonstrates that starting water containment over pressurization is addition at a high rate of flow and prevented without the need for a drywell throttling after approximately 4-vent. hours will not increase the suppression pool level to that which could block the suppression chamber HCVS.

As noted under Phase 1, the vent is sized to pass a minimum steam flow equivalent to 1% rated thermal power. This is sufficient to permit ventinq to maintain

containment below the lower of PCPL or design pressure.

No follow-up questions.

Phase 2 ISE 01 5 Reference Plant The NRG staff reviewed the Closed Torus freeboard volume is 525,000 information provided in the 6-Make available for the NRG gallons, SAWA flow is 500 GPM [gallons month updates and on the [Staff evaluation to be staff a description of how the per minute] at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> followed by 100 ePortal. included in SE Section plant is bounded by the GPM from 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />, 4.2.1.1]

reference plant analysis that The staff reviewed the shows the SAWM strategy is Browns Ferry parameters from the reference successful in making it unlikely Torus freeboard volume is 757,544 plant to those of Browns Ferry.

that a drywell vent is needed. gallons, SAWA flow is 500 GPM at 8 The staff concurs that it is unlikely hours followed by 100 GPM from 12 the suppression chamber HCVS hours to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />. could become blocked leading to a successful SAWNSAWM The above parameters for Browns Ferry strategy. Therefore, it is unlikely compared to the reference plant that a drywell vent would be required determine success of the SAWM strategy to maintain containment integrity.

demonstrate that the reference plant values are bounding. Therefore, the No follow-up questions.

SAWM strategy implemented at Browns Ferry makes it unlikely that a drywell vent is needed to prevent containment overpressure related failure.

Note:

Determined Torus Freeboard volume by using values from O-Tl-394 and subtracting the air volume above the Vent line (26.3 Ft) from the air volume from the top of the normal Suppression pool water level (15.08 Ft).

Using values from O-Tl-394, the air volume above 26.3 is 192765.2 gallons.

The air volume at -1" (15.08) is 950,309.8.

950,309.8 - 192765.2 = 757544.6

Phase 2 ISE 01 6 Browns Ferry Nuclear Plant utilizes the The NRG staff reviewed the Closed Harris Radio System to communicate information provided in the 6-Make available for NRG staff between the MGR and the operator at the month updates and on the [Staff evaluation to be documentation that FLEX pump. This communication method ePortal. included in SE Section demonstrates adequate is the same as accepted in Order EA 4.1]

communication between the 049. These items will be powered and The communication methods are MGR and the operator at the remained powered using the same the same as accepted in Order FLEX pump during severe methods as evaluated under Order EA- EA-12-049.

accident conditions.12-049 and continued for the period of sustained operation. No follow-up questions.

ML180386606 OFFICE NRR/DLP/PBEB/PM NRR/DLP/PBMB/LA NRR/DLP/PBEB/BC NRR/DLP/PBEB/PM NAME RAuluck Slent TBrown RAuluck DATE 2/12/18 2/9/18 2/20/18 2/21/18