ML18068A627

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

ML18068A627
Person / Time
Site:	Brunswick
Issue date:	03/22/2018
From:	Rajender Auluck Beyond-Design-Basis Engineering Branch
To:	William Gideon Duke Energy Progress
Lee B
References
CAC MF4467, CAC MF4468, EA-13-109, EPID L-2014-JLD-0041
Download: ML18068A627 (42)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 22, 2018 Mr. William R. Gideon Site Vice President Brunswick Steam Electric Plant 8470 River Rd., SE (M/C BNP001)

Southport, NC 28461

SUBJECT:

BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - 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. MF4467 AND MF4468; EPID L-2014-JLD-0041)

Dear Mr. Gideon:

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 Conditions," to all Boiling-Water Reactor licensees with Mark I and Mark II primary containments. The order requirements are provided in Attachment 2 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 26, 2014 (ADAMS Accession No. ML14191A687), Duke Energy Progress, LLC (Duke, the licensee) submitted its Phase 1 OIP for Brunswick Steam Electric Plant, Units 1 and 2 (BSEP, Brunswick). By letters dated December 17, 2014, June 25, 2015, December 11, 2015 (which included the combined Phase 1 and Phase 2 OIP), June 28, 2016, December 15, 2016, June 19, 2017, and December 20, 2017 (ADAMS Accession Nos. ML14364A029, ML15196A035, ML16020A064, ML16190A111, ML16365A007, ML17171A383, and ML17354A248, 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 (ISEs) for Phase 1 and Phase 2 of Order EA-13-109 for Brunswick by letters dated March 10, 2015 (ADAMS Accession No. ML15049A266), and August 17, 2016 (ADAMS Accession No. ML16223A725), 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.

The NRC staff conducted a teleconference with the licensee on February 22, 2018. The enclosed audit report provides a summary of that aspect of the audit.

W. Gideon 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. 50-325 and 50-324

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 DUKE ENERGY PROGRESS, LLC BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324 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 26, 2014 (ADAMS Accession No. ML14191A687), Duke Energy Progress, LLC (Duke, the licensee) submitted its Phase 1 OIP for Brunswick Steam Electric Plant, Units 1 and 2 (BSEP, Brunswick). By letters dated December 17, 2014, June 25, 2015, December 11, 2015 (which included the combined Phase 1 and Phase 2 OIP), June 28, 2016, December 15, 2016, June 19, 2017, and December 20, 2017 (ADAMS Accession Nos. ML14364A029, ML15196A035, ML16020A064, ML16190A111, ML16365A007, ML17171A383, and ML17354A248, 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 Brunswick by letters dated March 10, 2015 (ADAMS Accession No. ML15049A266), and August 17, 2016 (ADAMS Accession No. ML16223A725), 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 February 22, 2018. 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 IS Es. 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 Brunswick. The open items are taken from the Phase 1 and Phase 2 ISEs issued on March 10, 2015, and August 17, 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 Brunswick, 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, but notes that some open items are designated by the staff to be open or pending as described in Table 3 below.

The status of the NRC staff's review of these open items may change as additional information is provided to the staff, or 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 Suooort - l&C Steve Wyman NRR/DLP Technical Support - Dose John Parillo NRR/DRA Attachment 1

Table 2 - Audit Documents Reviewed Procedure OPLP-01.4, "Fukishima FLEX System Availability, Action, and Surveillance Requirements," Revision 6 Calculation OFLEX-0035, "Flow Capacity of BNP Hardened Wetwell Vent Units 1 & 2 at 1%

Rated Power," Revision 0 Calculation BNP-MECH-FLEX-002, "Brunswick Nuclear Plant Containment Analysis of FLEX Strategies," Revision 0 Calculation 31116-CALC-E-001, "FLEX Diesel Generator Sizing Caculation," Revision 0 EC 289233, "Fukushima: Hardened Vents at BNP," Revision 2 Procedure OEOP-01-FSG-04, "FLEX Diesel Generator Alignment," Revision 3 Calculation BNP-MECH-FLEX-005, "Brunswick Nuclear Plant MAAP 5.02 Analysis to Support SAWA Strategy," Revision 1 Calculation BNP-MECH-FLEX-004, "BNP Hardened Wetwell Check Valve Air lnleakage Evaluation," Revision 0 Calculation BNP-MECH-FLEX-003, "FLEX Control Building GOTHIC Heatup Analysis," Revision 0

Calculation BNP-MECH~FLEX-001, "FLEX Reactor Building GOTHIC Heatup Analysis,"

Revision 0 Calculation BNP-MECH-AOV-DP-CAC, "Differential Pressure Calculations for Yi -CAC-V7-AO, -

V8-AO, -V216-AO - Inboard Suppression Pool Purge Exhaust, Outboard Suppression Pool Purge Exhaust, and Hardened Wetwell Vent Isolation Air-Operated Valves," Revision 0 Calculation BNP-MECH-2-CAC-V216-AO, "Hardened Wetwell Vent Outboard Isolation Valve,"

Revision 1 Calculation BNP-MECH-2-CAC-V?-AO, "Torus Purge Exhaust Valve," Revision 1 Calculation BNP-MECH-2-CAC-V216-AO, "AOV Setup Calculation for 1-CAC-V216-AO Hardened Wetwell Vent Outboard Isolation Valve," Revision 1 Calculation BNP-MECH-2-CAC-V?-AO, "AOV Setup Calculation for 1-CAC-V?-AO Torus Purge Exhaust Valve," Revision 1 Calculation BNP-E-6.125, "24/48 VDC Battery Allowable Discharge Rate for HCVS ELAP,"

Revision 0 Calculation ORNA-001, "Instrument Air Nitrogen Backup System Volume Requirements,"

Revision 4 Calculation OFLEX-0003, "Hydraulic Analysis for Fukushima FLEX Connection Modifications,"

Revision 2 BWROG-TP-008, "Severe Accident Water Addition Timing" BWROG-TP-011, "Severe Accident Water Management Supporting Evaluations" Attachment 2

Brunswick Steam Electric Plant, Units 1 and 2 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 The HCVS out of service and The NRC staff reviewed the Closed compensatory measures were included in information provided in the 6-Make available for NRC staff a revision to OPLP-01.4, Fukushima month updates and on the [Staff evaluation to be audit the site-specific FLEX System Availability, Action, and ePortal. included in SE Section controlling document for HCVS Surveillance Requirements. The OPLP- 3.1.2.13]

out of service and 01.4 revision was issued concurrently with The guidelines and procedures compensatory measures. Revision 3 to the Severe Accident for HCVS out-of-service and Guidelines during the spring 2017 Unit 2 compensatory measures are refueling outage. This procedure will be complete for Unit 2 and consistent revised to incorporate Unit 1 HCVS with the guidance in NEI 13-02.

requirements when that unit's HCVS modifications are installed in accordance Unit 1 procedures will be revised with the milestone schedule reported in following the installation of the the BSEP Overall Integrated Plan. HCVS modifications and will follow the same guidance as Unit The OPLP-01.4 procedure revision that 2, consistent with the guidance in incorporates Unit 2 HCVS is available for NEI 13-02.

review on the ePortal.

No follow-up questions.

Phase 1 ISE 01 2 OFLEX-0035, Flow Capacity of BNP The NRC staff reviewed the Closed Hardened Wetwell Vent Units 1 & 2 at 1% information provided in the 6-Make available for NRC staff Rated Power, provides the calculation month updates and on the [Staff evaluation to be audit analyses demonstrating showing that both units' hardened vents' ePortal. included in SE Section that HCVS has the capacity to flow capacity is greater than 1 % thermal 3.1.2.1]

vent the steam/energy power at design pressure which is lower Calculation OFLEX-0035, "Flow equivalent of one percent of than the primary containment pressure Capacity of BNP Hardened licensed/rated thermal power limit. This is documented in the results Wetwell Vent Units 1 & 2 at 1%

(i.e., unless a lower value is paragraph 4.4 on page 9 of 9 of the Rated Power," Revision 0 justified), and that the calculation. This calculation assumes that assumed a rated thermal power suooression pool and the the new discharqe check valve has a Cv of 2,923 MWt. The calculation Attachment 3

HCVS together are able to of at least 673. The full open Cv of the used the saturated vapor enthalpy absorb and reject decay heat, check valve is approximately 4000. at containment design pressure of such that following a reactor Therefore, the vent pipe will pass at least 62 per sqaure in gauge (psig).

shutdown from full power 1 % thermal power equivalent. BNP- The required vent capacity is containment pressure is MECH-FLEX-002, Brunswick Nuclear 84,420 lb/hr. The calculation restored and then maintained Plant Containment Analysis of FLEX conservatively used a flow of below the primary containment Strategies, is a MAAP [Modular Accident 85,000 lb/hr. The calculation design pressure and the Analysis Program] calculation of the used AFT Arrow, version 4.0 primary containment pressure BSEP response to an extended loss of computer program to model the limit. AC power (ELAP) event initiated from full vent system. The program power. The MAAP results also show that determined the minimum Cv (flow containment pressure is rapidly reduced coefficient) for the check valve is and is maintained below design pressure 673. The full open Cv of the and primary containment pressure limit check valve is approximately (PCPL). This is best seen in the graph on 4000. Therefore, the licensee's page 4 of Appendix 7 (pdf page 55) which HCVS design will meet the 1% of is a plot of Run 1 containment response. rated thermal power requirement.

Run 1 models the BSEP procedural guidance for the FLEX event No follow-up questions.

Support documents are available for review on the ePortal.

Phase 1 ISE 01 3 BNP-MECH-FLEX-0002, provides the The NRC staff reviewed the Closed suppression pool (SP) response to the information provided in the 6-Make available for NRC staff ELAP with operator actions. Initially, in month updates and on the [Staff evaluation to be audit confirmation of the time it this analysis, reactor core isolation ePortal. included in SE Section takes the suppression pool to cooling (RCIC) is aligned to the 3.1.2.1]

reach the heat capacity suppression pool (SP). In this analysis, Calculation BNP-MECH-FLEX-temperature limit during ELAP after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the SP is approaching the 002, "Brunswick Nuclear Plant with RCIC in operation. heat capacity temperature limit (HCTL), Containment Analysis of FLEX although it has not yet reached it. At this Strategies" determined the time point, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the operators begin a for the suppression pool to reach controlled cooldown to 450 psig using one HCTL during an ELAP with RCIC safety relief valve (SRV). This reduces operating to be approximately 3.2 primary pressure while heating up the SP, hours.

but the net result is that the SP stays below the HCTL. No follow-up questions.

At 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br />, the operators further depressurize the reactor pressure vessel

(RPV) to 150-300 psig, which initially maintains the SP below HCTL. The exact time of reaching HCTL depends on the timing of SP heatup and the cycling of RPV pressure between 150 and 300 psig since the actual limit is a function of RPV pressure.

Since pressure is cycled between 150 psig and 300 psig after hour 2, it is conservative to determine the time at which the SP temperature and level reach the HCTL at 300 psig using the OEOP NL, HCTL curve. During this time, SP level is slowly increasing as shown in BNP-MECH-FLEX-0002, but is about -2.4 feet. This puts the HCTL temperature at about 193°F in the SP, which is reached at about 3.2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br />.

Support documents are available for review on the ePortal.

Phase 1 ISE 01 4 The location for the remote operating The NRC staff reviewed the Closed station (ROS) is in the southeast corner of information provided in the 6-Make available for NRC staff the Reactor Building (RB) 50'-0" elevation month updates. [Staff evaluation to be audit a description of the final for Unit 1, and the northeast corner of the included in SE Section ROS location. RB 50'-0" elevation for Unit 2. The ROS The ROS is in a location that is 3.1.2.4]

locations inside the RB are in a corridor readily accessible and appears to just inside a door to the outside of the RB support operation of the HCVS.

that will be blocked open in an ELAP.

This door access provides a direct path to No follow-up questions.

the Main Control Room (MCR) via the Radwaste Building roof.

The evaluation of the ROS for temperature and radiation concerns is contained in the response to ISE open item #10.

Phase 1 ISE 01 5 The primary operating station for the The NRC staff reviewed the Closed HCVS is and remains in the Main Control information provided in the 6-

Make available for NRC staff Room (MCR) with the implementation of month updates and on the [Staff evaluation to be audit documentation that orderEA-13-109. For each unit, the ePortal. included in SE Section demonstrates adequate alternate operating station (also called the 3.1.1.1]

communication between the Remote Operating Station or ROS) is The communication methods are remote HCVS operation located just inside the Reactor Building the same as accepted in Order locations and the HCVS (RB) at the 50-foot elevation, adjacent to EA-12-049.

decision makers during ELAP a door to the outside. The door is in the and severe accident southeast section of the RB for Unit 1 and No follow-up questions.

conditions. in the northeast section for Unit 2. The MCR will direct operators to this alternate control location if required due to an inability to operate the HCVS valves from the MCR. In addition, operators will be dispatched to the backup pneumatic connections on each unit in order to connect the backup air compressor before the 24-hour nitrogen supply is exhausted.

These HCVS activities may require communication with the MCR.

As part of the response to NRC Order EA-12-049, BSEP assumed that permanently installed plant communications systems would not be available during an extended loss of AC power (ELAP).

Instead, BSEP primarily utilizes an 800 MHz [Mega Hertz] radio system consisting of 500 hand-held radios for onsite communications. These radios are stored in reasonably protected buildings, including the FLEX Storage Building, to meet the requirements of EA-12-049.

This information was provided in response to NTTF Recommendation 9.3, by a letter dated October 31, 2012 (ADAMS Accession No. ML12311A299) and supplemented by a letter dated February 22, 2013, Carolina Power &

Light Company's and Florida Power Corooration's Response to Follow-Up

Letter on Technical Issues for Resolution Regarding Licensee Communication Submittals Associated with Near-Term Task Force Recommendation 9.3 (ADAMS Accession No. ML13058A045).

This information was assessed by the NRC staff and a Staff Evaluation was issued for this assessment. This was provided in Brunswick Steam Electric Plant, Units 1 and 2 - Staff Assessment in Response to Information Request Pursuant to 10 CFR 50.54(f)-9.3, Communication Assessment, dated April 4, 2013 (ADAMS Accession No. ML13093A341).

Phase 1 ISE 01 6 HCVS-WP-03, Hydrogen/Carbon The NRC staff reviewed the Closed Monoxide Control Measures (ADAMS information provided in the 6-Provide a description of the Accession No. ML14302A066}, on page month updates and on the [Staff evaluation to be final design of the HCVS to 2, lists the information that licensees shall ePortal. included in SE Section address hydrogen Complete provide with respect to strategies and 3.1.2.11]

detonation and deflagration. options that "ensure the flammability limits The licensee's design is of gases passes through the system are consistent with Option 5 of the not reached." endorsed white paper HCVS-WP-03.

From HCVS-WP-03, page 2:

1. Declare option or options selected No follow-up questions.

(valid for use of Options 3, 4 and/or 5)

2. List any deviations relative to the selected option(s) along with justification

3. Synopsis of venting operation and design

4. Sketch of vent path from associated PCIVs to release point, with delineation of which option applies to each portion of the vent system The information is provided below and was included in the December 2015 six-month uodate to the Overall Integrated

Plan (ADAMS Accession No. ML16020A064).

1. BSEP has chosen option 5 which is to install a downstream check valve to prevent air from being drawn into the vent pipe when venting is stopped.

2. BSEP is not planning any deviations relative to option 5.

3. BSEP procedures contain guidance to open the hardened vent if plant conditions require it to prevent containment pressure exceeding the Primary Containment Pressure Limit. The vent will remain open until alternate reliable containment heat removal is established unless there is some condition or event that would require it be closed. There are no procedure steps that direct the vent be cycled to maintain a certain containment pressure band. The vent design is described in the BSEP OIP.

4. The sketch of the vent path with delineation of which option applies is available for review on the ePortal. Piping downstream of the second containment isolation valve, CAC-V216, is protected by the check valve (Option 5).

The final HCVS design installs a check valve in the piping slightly below the Reactor Building roof as discussed in item 5 of the table on page 12 of HCVS-WP-

03. The check valve will be mounted near the roof to minimize seismic effects, and will be less than 30 pipe diameters from the end as discussed in HCVS-WP-03 page 35. The BSEP check valve will minimize leakage of air into the HCVS piping such that a flammable mixture will

not occur while venting has been stopped without alternate containment heat removal. Just downstream of the check valve, BSEP will install a low pressure, 13 psig, rupture disk that will allow check valve testing, but will not prevent containment venting to avoid the primary containment pressure limit (PCPL).

As part of the modifications, the new check valve will have test ports above and below it that will allow testing to verify that the valve opens and allow testing to verify that the valve leaks less than an amount that would allow a combustible mixture to occur in the pipe.

Support documents, including a sketch of the vent path with delineation of which option applies, are available for review on the ePortal.

Phase 1 ISE 01 7 BSEP evaluated the HCVS stack for all The NRC staff reviewed the Closed Beyond-Design-Basis-External Events in information provided in the 6-Make available for NRC staff Engineering Change (EC) 299559, month updates and on the [Staff evaluation to be audit seismic and tornado Evaluation of the Hardened Wetwell Vent ePortal. included in SE Section missile final design criteria for for Beyond-Design-Basis External Events, 3.2.2]

the HCVS stack. attachment Z01 RO. This evaluation is Engineering Change (EC) available for review on the ePortal. This 299559, "Evaluation of the evaluation was provided as part of the Hardened Wetwell Vent for BSEP FLEX audit in 2014, and was Beyond Design-Basis External accepted for the tornado missile hazard Events," Revision 0, evaluated disposition in Section 3.4 of Brunswick the HCVS stack. The hardened Steam Electric Plant, Units 1 and 2 - wetwell vent (HWV) is routed Report for the Audit Regarding through a seismic isolation space Implementation of Mitigating Strategies between the Reactor Building and and Reliable Spent Fuel Pool the Turbine Building. The HWV is Instrumentation Related to Orders EA protected from tornado missiles 049 and EA-12-051, March 31, 2015 by the Reactor Building and the (ADAMS Accession No. ML15082A155). Turbine Building. The HWV reenters the Reactor Building and

The HCVS stack is part of the Hardened exits through the roof. The Wetwell Vent system that is evaluated in Reactor Building provides tornado EC 299559. Sections 3.1.1 and 3.1.2 missile protection. The licensee state the seismic design input and hazard looked at potential wind-driven criteria. Section 3.2.1 dispositions the missiles and concluded that none seismic hazard. Section 3.1.5 states the of the potential missiles could be design criteria for the tornado missile accelerated to speeds sufficient to hazard (along with the high wind hazard). damage the HWV.

Section 3.2.4 dispositions the tornado missile hazard (along with the high wind No follow up questions.

hazard).

Phase 1 ISE 01 8 Calculation ORNA-0001, "Instrument Air The NRC staff reviewed the Closed Nitrogen Backup System Volume information provided in the 6-Make available for NRC staff Requirements," provides the backup month updates and on the [Staff evaluation to be audit documentation of the nitrogen system usage calculation and ePortal. included in SE Section HCVS nitrogen pneumatic adequacy verification. On pages 5 and 6, 3.1.2.6]

system design including sizing the base calculation determines usage by Calculation ORNA-0001 discusses and location. the Safety-Relief Valves (SRVs), the the pneumatic design and sizing.

Reactor Building to Suppression Chamber This calculation determined the vacuum breaker valves, the Hardened required number of nitrogen Wetwell Vent Valves, and leakage. The cylinders needed in the backup total usage is determined to be 91 O nitrogen system for vent operation standard cubic feet against an available for sustained operation for each volume of 961 cubic feet (page 4). unit, respectively. The number of However, this usage was over a 22-hour nitrogen cylinders installed in period, vice the 24-hour period required each unit and available are by EA-13-109. sufficient to operate the HCVS for 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />.

As part of the BSEP response to EA 109, 2 bottles were added to each unit's No follow-up questions.

Backup Nitrogen System, on each of 2 divisions. Appendix A of this calculation was created to demonstrate that the system has 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> worth of capacity.

Appendix A shows that, with the additional bottles being added, there is enough nitrogen in Division 2 alone to supply 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> of nitrogen including leakaqe assumptions, HCVS valve

cycling, SRV cycling, and containment vacuum breaker cycling.

The safety-related Backup Nitrogen System bottles are located in seismically qualified racks (sections B.5.5 and B.5.10 of ECs 290410, Hardened Containment Vent System - Backup Nitrogen Bottles Unit 2, and 292338, Hardened Containment Vent System - Backup Nitrogen Bottles Unit 1) on the 50' elevation of the Reactor Building. The locations are shown on drawing F-02503 for Unit 2 and F-25003 for Unit 1. These bottles are always lined up to supply the HCVS vent valves if required so that no operator actions are required at the bottle racks. If the HCVS valves cannot be operated electrically, the operators can open them from the Remote Operating Station, located as shown on drawing F-02503, without approaching primary containment or the vent valves themselves (which are approximately 60 feet below the ROS, and in the area of the vent pipe, across the Reactor Building from the ROS).

For pneumatic makeup after the backup bottles are depleted (later than 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />),

the FLEX air compressor will be connected to the Backup Nitrogen System.

Phase 1 ISE 01 9 As described in 31116-CALC-E-001, The NRC staff reviewed the Closed "FLEX Diesel Generator Sizing information provided in the 6-Make available for NRC staff Calculation," the bounding expected load month updates and on the [Staff evaluation to be audit documentation of HCVS for the FLEX diesel generators (DGs) is ePortal. included in SE Section incorporation into the FLEX 367.4 kW. Taking a 25% margin, the 3.1.2.6]

diesel generator loading required maximum output of the FLEX DG The licensee stated that all calculation. must be at least 460 kW. A nominal 500 electrical power required for

kW FLEX DG meets this requirement. As operation of HCVS components is discussed in this calculation, the major provided by the Division 2 24/48 load is the battery chargers, and they VDC battery and battery have completed re-charging the batteries chargers.

within 11 hours0.458 days <br />0.0655 weeks <br />0.0151 months <br />. The battery chargers represent 288 KW of the 367.4 kW The battery sizing calculation maximum load. BNP-E-6.125 confirmed that the 24/48 VDC battery has a The majority of the loads initially aligned minimum capacity capable of to the FLEX DGs are battery chargers providing power for 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> and UPS, as described in Calculation without recharging, and therefore 31116-CALC-E-001. The FLEX DGs are is adequate.

oversized for the load profile which helps minimize any effects from non-linear The licensee provided 31116-loading. This can be seen since the CALC-E-001, which discusses re-diesel generators are rated 500kW, but powering of the 24/48 battery the maximum draw, for FLEX critical chargers using the FLEX DG.

loads, including the non-linear loading will be less than 380kW. The non-linear No follow-up questions.

loading from the battery chargers quickly drops off after batteries are fully recharged. This can be seen from the load profiles in Calculation 31116-CALC-E-001 where the power draw to the chargers drops below 20% of rated load after 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br />.

While the exact loading of the HCVS has not been incorporated into the FLEX DG loading calculations above, inspection of the HCVS power supply demonstrates that the HCVS load is insignificant to the FLEX DGs given the amount of load margin available. Calculation BNP-E-6.076-ICC-001, "Hardened Containment Vent System - Unit 2 Power Distribution,"

adds the HCVS Radiation Monitor to the loading of the associated battery, 1.124 amps at 24 VDC [volts direct current) as shown on oaae 3 of Attachment 1 of BNP-

E-6.076-ICC-001. Calculation BNP-E-6.125, "24/48 voe Battery Allowable Discharge Rate for HCVS during an ELAP," contains the additional loading of the three instrument loops that will be powered by the HCVS distribution. These three instrument loops total 0.06 amps at 24 VDC as shown on page 1-1 of BNP-E-6.125. Therefore, the total load of the HCVS distribution is approximately 1.184 amps at 24 VDC or a little more than 28 watts. The 28 watts is insignificant to the FLEX DG load since the margin available in the FLEX DGs, even when the safety-related battery chargers are in service is approximately 132.6 K>N.

The full one-hour load on the Division 2 24/48 voe batteries is approximately 20 amps per BNP-E-6.076-ICC-0001. This represents a load of 20 Ax 24 VDC = 480 watts. Assuming the FLEX DGs are required to carry the full load of the Division 2 24/48 VDC batteries through the charger, the additional 480 watts is also insignificant to the 132.6 kW of available capacity. Therefore, the FLEX DGs are fully capable of carrying the HCVS loads at any time they are energized.

Phase 1 ISE 01 10 Operator actions for HCVS may be The NRC staff reviewed the Open required at the following operating information provided in the 6-Make available for NRC staff locations during an ELAP (see "Operator month updates and on the [Staff evaluation to be audit an evaluation of Action Maps.pdf' available for review on ePortal. included in SE Sections temperature and radiological the ePortal): 3.1.1.2 and 3.1.1.3]

conditions to ensure that Calculation RWA-L-1312-003 operating personnel can safely 1. Main Control Room (MCR) (primary contains the Control Building access and operate control operating location) GOTHIC room heatup analysis for and support equipment. 2. Control Building 49' elevation (location the ELAP event. This calculation of HCVS power supply transfer switches) assumes selected doors open at

3. Outside of the RB, FLEX instrument air the start of the event. The supply and refueling of FLEX compressor Control Room temperature peaks for long-term pneumatic supply at 116°F. After compensatory

4. Outside of the RB, FLEX Diesel actions of opening select doors Generator (DG) enclosure to refuel the and installing a portable fan, the FLEX DGs for long-term electrical supply temperature drops below 95°F.

5. Reactor Building (RB) - 50' elevation at For the remainder of the 72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> the Remote Operating Station (ROS) modeled, the Control Room temperature cycles between Main Control Room and Control Building roughly 95°F and 105°F following 49' - Temperature Evaluation outdoor air diurnal temperature variation.

Calculation RWA-L-1312-003, BNP Control Building (CB) FLEX Room Heat- Calculation BNP-MECH-FLEX-up Analysis, contains a Control Building 0001 contains the Reactor GOTHIC room heatup analysis for the Building GOTHIC room heatup ELAP event. This analysis takes no credit analysis for the ELAP event. This for operator action for the first six hours calculation determines the (other than opening panel doors) at which maximum temperature at the time the outside doors to the Control ROS to be 121°F. The operator Building are opened and fans are started actions will take place near the to force outside air through the building. 50' door near the ROS so the This is a FLEX action evaluated as local temperature at this location acceptable in response to NRC Order EA- will be close to ambient outside 12-049. This action is represented by the RB. Stay time in ROS will be Case 4 as shown in Table 7 on page 18 limited. Procedures identify of 51. The results are tabulated on page requirements for hot area work.

19 of 51 in Table 8. The results show Ice vests will be available as ambient temperatures being maintained need.

below 124°F for all spaces in which there may be operator actions for HCVS. Per The MCR and its boundary are HCVS-FAQ-06 in NEI 13-02 Appendix J, acceptable for radiological FLEX strategies that are not specific to conditions without further HCVS can be credited as previously evaluation for HCVS actions per evaluated for FLEX. This temperature is NEI 13-02, Rev.1, HCVS-FAQ-01.

judged acceptable for the simple and non-physical operator actions (i.e., switch For the ROS, the licensee manipulation, meter reading) required for provided a qualitative argument HCVS operation during an ELAP event. as to why the dose is not a concern. During the audit call on

Main Control Room and Control Building February 22, 2018, the NRC staff 49' - Radiation Evaluation requested that the licensee perform a dose calculation.

The MCR and CB 49' (49' is adjacent to the MCR and inside the MCR boundary) This item will remain open until are acceptable for radiological conditions the licensee provides the NRC without further evaluation for HCVS staff a dose calculation, which actions per N El 13-02, Rev. 1, HCVS- shows the integrated radiation FAQ-01. dose due to HCVS operation should not inhibit operator actions Outside Areas for Pneumatic Makeup, needed to initiate and operate the Electrical Supply, and Refueling Activities HCVS during an ELAP with severe accident conditions.

Pneumatic makeup location The pneumatic supply for the first 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> of the ELAP event comes from the safety-related Backup Nitrogen System.

No operator actions are required to supply pneumatics in the first 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />. On both units, there is a makeup station for the backup nitrogen system in the seismic isolation space between the Reactor Building (RB) and Turbine Building (TB)

(see Operator Action Maps.pdf available for review on the ePortal). Per the response to order EA-12-049, portable FLEX compressors will be moved to outside locations near these makeup stations. Since the locations are outside the RB, there is no possible effect from RB heatup due to the ELAP. The compressors can be safely operated and refueled from this outside location as they will be shielded from the vent pipe by at least two of the RB concrete walls (three feet thick each) and no actions are required in the RB to supply the long-term pneumatic supply.

The makeup connections in the seismic isolation spaces are near the vent pipes (more so on Unit 2 than Unit 1) and possibly subject to gamma dose from the pipe once venting starts. Therefore, the connections of hose to the makeup stations in the seismic isolation space will be made before venting starts at approximately 17. 7 hours0.292 days <br />0.0417 weeks <br />0.00959 months <br />.

Electrical makeup location The HCVS electrical supply for the first 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> is from the station 24/48 voe battery system. This backup power supply is aligned at the 49-foot elevation of the Control Building adjacent to the MCR. As previously stated, this location is in the Control Building inside the MCR boundary and is acceptable for the duration of the event.

The long-term electrical supply for the HCVS is from the FLEX Diesel Generators which can repower the normal supply buses to the HCVS controls and instruments or re-power the 24/48 voe battery chargers. The FLEX Diesel Generators are located in the FLEX DG enclosure which is east of the RBs and the Emergency Diesel Generator (EOG) building (see Operator Action Maps.pdf available for review on ePortal). The location is on the opposite side of the RBs from the HCVS pipes and outside the RBs so that there are no concerns with operation of the FLEX DGs including refueling operation. No electrical lineups needs to be made in the RB for the FLEX DG to supply the needed HCVS

components, only inside the EOG Building which is not a dose or temperature concern area.

Remote Operating Station - Temperature Evaluation Calculation BNP-MECH-FLEX-0001 documents the Reactor Building Heatup Analysis under ELAP conditions in which all ventilation, heating and cooling are deenergized. This analysis was used for development of the FLEX actions per order EA-12-049, but since the same Extended Loss of AC Power (ELAP) conditions apply to the EA-13-109 order, this analysis can be used to estimate the temperature at the ROS for HCVS purposes. Even though EA-13-109 requires the consideration of a severe accident, the existence of core damage and possible vessel breach will have no effect on the temperature at the ROS.

The applicable case in BNP-MECH-FLEX-0001 is case 1 which models the operator actions in an ELAP. The GOTHIC analysis results in Table 4 (page

23) show that the maximum temperature on the 50' elevation is 121°F. The actions at the ROS will be to open or close a maximum of three % inch valves so that they are expected to take less than 5 minutes. Furthermore, the operator will be entering the RB through the 50' door near the ROS so that the local temperature will be close to ambient outside the RB. These temperatures, coupled with the short duration of action, are judged acceptable.

Remote Operating Station - Radiation Evaluation The bottom of the active core region is at 51' elevation. Therefore, an operator would be roughly at core elevation while at the ROS. The shielding provided by the vessel, bio-shield, Primary Containment (PC), and distance from the core results in the 50' door location being a low-dose-rate-waiting area during normal full-power operation. The Primary Containment wall alone provides six feet of concrete shielding. Since the core is shutdown for the ELAP event, the dose rates from the core area will be lower than during operation.

The existence of core damage with possible reactor pressure vessel breach will not raise the dose levels at the ROS.

If the core were to melt through the lower vessel head, there would be loss of shielding from the vessel, however there would be additional distance to the ROS and additional concrete shielding provided by the pedestal. Any gap release to the suppression pool will contribute to RB dose rates, however the ROS is on the 50' elevation, two floors above the torus.

Therefore, the dose rate at the ROS due to the torus will be insignificant due to the 5 feet of concrete below the ground floor as well as the additional concrete and distance afforded by the location being on the 50-foot elevation. Likewise, any gap release that migrates back to the Primary Containment, will be shielded from the ROS by the 6' thick Primary Containment

wall. In addition, the ROS is approximately 50 feet away from the PC wall.

Support documents, including the Operator Action Maps.pdf, are available for review on the ePortal.

Phase 1 ISE 01 11 A list of instruments and controls The NRC staff reviewed the Closed necessary to implement EA-13-109 with information provided in the 6-Make available for NRC staff their descriptions and qualification month updates and on the [Staff evaluation to be audit descriptions of all methods is provided in the December ePortal. included in SE Section instrumentation and controls 2016 Six-Month Status Report by letter 3.1.2.8]

(i.e., existing and planned) dated December 15, 2016 (i.e., ADAMS The existing plant instuments necessary to implement this Accession No. ML16365A007). required for HCVS (i.e. wetwell order including qualification level instruments and drywell methods. pressure instruments) meet the requirements of Regulatory Guide (RG) 1.97.

The licensee provided a list of HCVS instruments and controls (l&C) a brief description of each component, the component identification number, the component make and model number and the qualification method. The staff's review indicates that the l&C components are consistent with the guidance in NEI 13-02 and its qualifications meet the order requirements.

No follow-up questions.

Phase 1 ISE 01 12 Existing equipment installed prior to RG The NRC staff reviewed the Closed

1. 97 is qualified in accordance with information provided in the 6-Clarify whether the seismic original licensing basis and IEEE 344- month updates and on the [Staff evaluation to be reliability demonstration of 1971. Equipment installed after RG 1.97 ePortal. included in SE Section instruments, including valve is qualified to IEEE 344-1975. Therefore, 3.1.1.4]

position indication, vent pipe the BSEP HCVS instruments will be

temperature instrumentation, qualified to IEEE 344-1971 or 1975. The The licensee stated, in part, that radiation monitoring, and exception is the new 24 VDC voltmeter the seismic qualification method support system monitoring will being installed for EA-13-109 response is was dependent on the time of (be) via methods that predict qualified to IEEE-344-2004 as this vendor original installation and that most performance described in only provides qualification to that version. l&C equipment is pre-existing

[Institute of Electrical and except the voltmeter for the new Electronics Engineers] IEEE- See ISE Open Item #14 for more details HCVS battery. The staff 344-2004 or provide on the instrument qualifications. confirmed the individual justification for using a component seismic qualification different revision of the methods in the l&C component standard. list provided.

No follow-up questions.

Phase 1 ISE 01 13 While NEI 13-02 paragraph 4.2.4.5 The NRC staff reviewed the Closed provides an acceptable approach for information provided in the 6-Make available for NRC staff HCVS monitoring that includes vent pipe month updates and on the [Staff evaluation to be audit a justification for not pressure, BSEP has not included HCVS ePortal. included in SE Section monitoring HCVS system vent pipe pressure. If the HCVS is not in 3.1.2.8) pressure as described in NEI service, a vent pipe pressure indicator The NRC staff determined that 13-02. would not provide useful information. If HCVS valve position indication, the HCVS is placed in service, BSEP has HCVS pipe temperature, HCVS several indicators that will reliably indicate radiation level and drywell the status of containment and of the pressure are sufficient indication HCVS. The following indicators are to determine the vent is operating already qualified for post-accident as expected. Supression pool conditions or are qualified per the level is needed to determine the requirements of EA-13-109. amount of freeboard before operating the vent.

1. Drywell pressure

2. HCVS valve position indication No follow-up questions.

3. HCVS pipe temperature

4. HCVS pipe radiation level

5. Suppression Pool level These five instruments provide sufficient information for the operators to monitor the status of the vent system without the addition of a vent pipe pressure indicator.

Phase 1 ISE 01 14 The list of components and their The NRC staff reviewed the Pending evaluations is available for review on the information provided in the 6-

Make available for NRC staff ePortal in spreadsheet HCVS ISE OPEN month updates and on the [Staff evaluation to be audit the descriptions of local ITEM 14.xlsx. The components in the ePortal. included in SE Section conditions (i.e., temperature, table boxes with no background color are 3.1.1.4]

radiation and humidity) new for EA-13-109 compliance. The The NRC staff reviewed the table anticipated during ELAP and components with the blue background provided on the eportal. The staff severe accident for the color are existing plant equipment that anticipates a similar table in the components (e.g., valves, meet the current design basis of the plant. FIP (on the docket) to close the instrumentation, sensors, All components are evaluated for item.

transmitters, indicators, temperature, humidity integrated electronics, control devices, radiation, and seismic adequacy. No follow-up questions.

etc.) required for HCVS venting including confirmation The estimates of temperature in the that the components are Reactor Building at the various locations capable of performing their are based on GOTHIC analyses of the functions during ELAP and ELAP event for the Reactor Building.

severe accident conditions. Reactor Building humidity is assumed to rise to 100% due to boiling from the spent fuel pool in less than 7 days. For the Control Building, the temperature estimates are based on a GOTHIC analysis that assumes zero humidity, thereby maximizing temperature response. The Control Building humidity is assumed to reach a maximum of 91 %

which is based on the historic maximum humidity of the ambient air (used to ventilate the Control Building in the FLEX strategies) from UFSAR Table 2-24.

All components are either seismically qualified to IEEE-344-1975 (the battery voltmeter is new and is qualified to IEEE-344-2004) or have been evaluated as seismically rugged so that they will perform their function following a seismic event. The estimate of radiation dose at any component is based on the results presented in Table 2 of HCVS-WP-02 as scaled to BSEP plant specifics. For this evaluation of components, the 4-hour time

step was chosen for the pipe dose rates and the dose rate is held constant rather than accounting for decay. In the BSEP MAAP analysis 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br /> is before the vent would be opened to avoid PCPL. Since the dose rate decreases after the 4-hour time step, it is conservative to use this dose rate.

For valves and other components that are in or on the pipe, the 1' dose is used and integrated over a 7-day period. For other components in the Reactor Building such as pressure transmitters, the 3' dose is used. This is conservative because these instruments are, in fact, not near the vent pipe and are shielded from the vent pipe by the 3' thick Reactor Building wall or the Primary Containment wall. The Primary Containment wall also shields these instruments from the airborne activity in the Primary Containment. As with the 1' dose components, this 3' dose is integrated over the 168-hour period with no allowance for decay. The resulting total integrated dose (TIO) are then compared to the qualification total integrated dose for each susceptible component.

All components are confirmed capable of performing their functions during ELAP and severe accident conditions. Support documents, including spreadsheet HCVS ISE OPEN ITEM 14.xlsx, are available for review on the ePortal.

Phase 1 ISE 01 15 BSEP procedure OEOP-02-PCCP, The NRC staff reviewed the Closed Primary Containment Control Procedure, information provided in the 6-Make available for NRC staff directs opening the hardened wetwell vent month updates and on the audit documentation of an valves before reachinq the primary ePortal.

evaluation verifying the containment pressure limit (PCPL) of 70 [Staff evaluation to be existing containment isolation psig. Therefore, the maximum opening The NRC staff reviewed BNP- included in SE Section valves, relied upon for the dip is 70 psid (containment to MECH-AOV-DP-CAC, 3.2.1]

HCVS, will open under the atmosphere). Calculation BNP-MECH- "Differential Pressure Calculations maximum expected differential AOV-DP-CAC, in the table in for Y:z - CAC-V7-AO, -V8-AO, -

pressure during BDBEE and Section 4.0, page 12 of this calculation, V216-AO Inboard Suppression severe accident wetwell confirms that 70 psid is the maximum Pool Purge Exhaust, Outboard venting. expected opening differential pressure. Suppression Pool purge Exhaust, and Hardened Wetwell Vent BNP-MECH-1-CAC-V7-AO and BNP- Isolation Air-Operated Valves,"

MECH-2-CAC-V7-AO contain the Air Revision 0, which discusses the Operated Valve (AOV) calculations for the valve/actuator information for the inboard wetwell purge valve on each unit. PCIVs. The NRC staff verified Section 4.1.1 contains a table of minimum the actuator can develop greater margins for these valves. The minimum torque than PCIV's unseating opening margin for 1-CAC-V7 is 12. 7%, torque.

and for 2-CAC-V7 is 19.5%.

No follow-up questions.

BNP-MECH-1-CAC-V216-AO and BNP-MECH-2-CAC-V216-AO contain the AOV calculations for the hardened wetwell vent valve on each unit. Section 4.1.1 contains a table of minimum margins for these valves. The minimum opening margin for 1-CAC-V216 is 33.8%, and for 2-CAC-V? is 25. 7%.

Phase 1 ISE 01 16 As shown and described in the BSEP The NRC staff reviewed the Closed OIP, the HCVS pipe taps off a 20-inch information provided in the 6-Provide a description of the torus purge pipe, is routed outside the month updates and on the [Staff evaluation to be strategies for hydrogen control Reactor Building (RB) into the seismic ePortal. included in SE Section that minimizes the potential for isolation space between the RB and 3.1.2.12]

hydrogen gas migration and Turbine Building (TB), is routed up the The wetwell vent for each unit ingress into the reactor outside of the RB, re-enters the RB at the utilizes CAC system valves CAC-building or other buildings 120' elevation, then exits through the RB V7 and CAC-V216 for roof. There is no penetration into any containment isolation. CAC other building. system containment isolation valves CAC-V8 and CAC-V172 The only interface between HCVS and are the only functional boundary any other system is through valves CAC- valves between the HCVS and VB and CAC-V172. These two valves the downstream SBGT system.

connect the purge system to the Standby These valves are tested, and will Gas Treatment System (located inside the continue to be tested, for leakage RB) and are primary containment isolation under 10 CFR 50 Appendix J as valves (PCIV). Since they are PCIVs they part of the containment boundary are tested for leakage per 10 CFR 50 in accordance with HCVS-FAQ-Appendix J. This testing methodology 05. The NRC staffs review of the has been endorsed as an acceptable proposed system indicates that testing means in HCVS-FAQ-05. the licensee's design appears to Therefore, it is expected that the potential maintain hydrogen below for hydrogen gas migration to the SBGT flammability limits.

system, which could lead to leakage into the RB, is minimized. No follow-up questions.

As shown in Sketch 1 of the BSEP OIP (ADAMS Accession No. ML14191A687),

the HCVS pipe is routed to the RB roof without any further connections to the RB atmosphere, to any system other than SBGT, or to any other building. This portion of the HCVS shall be leak tested in accordance with NEI 13-02. Since the pipe does not enter any other station building, there is no possibility of hydrogen gas migration into any other building.

The HCVS piping is constructed of seamless type 304 stainless steel piping.

The piping joints are a combination of welded and flanged connections. The piping was designed, fabricated and installed in accordance with ANSI 831.1 and is tested per NEI 13-02. Therefore, the HCVS piping can be considered leak tight and there is minimal potential for hydrogen to leak into the Reactor Building.

The BSEP HCVS pipe is a connection off the wetwell purqe line. The other branch

connections from this purge line contain automatic, fail-closed, containment isolation valves that are tested as part of 10CFR50, Appendix J, testing to ensure leakage is within limits (per HCVS-FAQ-05). The rest of the HCVS pipe is not connected to any other system and does not traverse any building other than the same unit Reactor Building. The HCVS pipe is sealed with flanges and closed valves, was pressure tested when initially installed and will additionally be tested after modifications for EA-13-109 compliance to ensure it is leak-tight.

Phase 2 ISE 01 1 Operator actions may be required at the The NRC staff reviewed the Open following operating locations during an information provided in the 6-Licensee to confirm through ELAP (see "Operator Action Maps.pdf' month updates and on the [Staff evaluation to be analysis, the temperature and located on the ePortal): ePortal. included in SE Sections radiological conditions to 3.1.1.2 and 3.1.1.3]

ensure that operating 1. Main Control Room (MCR) (i.e., See comments for Phase 1 open personnel can safely access primary operating location) item #10 for NRC staff's review and operate controls and 2. Control Building 49' level (i.e., location considerations.

support equipment. of HCVS power supply transfer switches)

3. Outside of the Reactor Building (RB),

FLEX instrument air supply and refueling of FLEX compressor for long-term pneumatic supply

4. Outside of the RB, at the FLEX Diesel Generator (DG) enclosure to refuel the FLEX DGs for long-term electrical supply

5. Reactor Building (RB) - 50' level at the Remote Operating Station (ROS)

6. East of the RB near the Condensate Storage Tanks (CST) to connect a hose for FLEX/Severe Accident Water Addition (SAWA) and stage the FLEX/SAWA pump.

7. At the outside wall of the RB (i.e., north for Unit 1 and south for Unit 2) at the

FLEX core bore for FLEX/SAWA pump discharge

8. Inside the RB at the 20' elevation (i.e.,

ground)

Main Control Room and Control Building 49' -Temperature Evaluation:

Vendor calculation RWA-L-1312-003, BNP CB FLEX Room Heat-up Analysis, contains a Control building GOTHIC room heatup analysis for the ELAP event (i.e.,

Reference EC 289577 Attachment Z52).

This analysis takes no credit for operator action for the first six hours (i.e., other than opening panel doors) at which time the outside doors to the control building are opened and fans are started to force outside air through the building. This is a FLEX action evaluated as acceptable in response to NRC Order EA-12-049. This action is represented by Case 4 as shown in Table 7 on Page 18 of 51. The results are tabulated on Page 19 of 51 in Table B. The results show ambient temperatures being maintained below 120°F for all spaces except the electrical equipment rooms in which there are no actions for HCVS. Per HCVS-FAQ-06 in NEI 13-02 Appendix J, FLEX strategies that are not specific to HCVS can be credited as previously evaluated for FLEX. This temperature is judged acceptable for operator action during an ELAP event.

Main Control Room, Control Building 49',

and Control Building (Battery Rooms) -

Radiation Evaluation:

The MCR and CB 49' (i.e., 49' is adjacent to the MCR and inside the MCR boundary) are acceptable for radiological conditions without further evaluation for HCVS actions per NEI 13-02, Rev.1, HCVS-FAQ-01.

Areas for SAWA Injection, Pneumatic Makeup, Electrical Supply, and Refueling Activities SAWA Locations The source of water for SAWA is the Condensate Storage Tanks (CST), which were qualified for all external hazards for EA-12-049 response. Operators will connect a suction hose, stored with the pump in the FLEX Storage Building (FSB), to the tank and then to the pump.

The pumps are staged outside and east of the RBs. From the pumps, hoses are run to the FLEX core bores on the outside of the RBs (north wall for Unit 1 and south wall of Unit 2). The Severe Accident Water Mitigation (SAWM) flow instrumentation is mounted on the pump itself.

Since the pumps are outside the RB, there are no ambient temperature concerns for personnel action. Also since the pumps are outside the RB and on the opposite side of the RBs from the vent pipes, there is no concern for the radiation levels at the pumps from the vent pipe or damaged core which would still be inside the primary containment.

Likewise, the long term makeup for the CST is from the discharge canal weir.

The pump staging location and hose runs for this makeup path are far from the RBs and the vent pipes, so there are no radiological concerns for CST makeup.

Inside the RB, the SAWA pipe is aligned by opening three valves at the 20' (i.e.,

ground) level of the RB. These valves will be opened within the first hour after the start of the ELAP before there is any core damage or significant RB heatup. The use of the hard pipe eliminates the need for the operators to run hoses inside the RB during an event. In the case of a seismically-initiated event, the operators will also close one valve on the 80' level of the RB, so at most four valves need be operated to align the SAWA flow path inside the RB.

Pneumatic makeup location The pneumatic supply for the first 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> of the ELAP event comes from the safety-related backup nitrogen system.

No operator actions are required to supply pneumatics in the first 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> as the backup nitrogen system automatically aligns itself. On both units, there is a makeup station for the backup nitrogen system in the seismic isolation space between the RB and Turbine Building (TB) (i.e., see Operator Action Maps.pdf).

Per the response to order EA-12-049, portable FLEX compressors will be moved to outside locations near these makeup stations. Since the locations are outside the RB, there is no ambient

temperature concern for personnel actions. The staging location is shielded from the vent pipe by a minimum of six feet of concrete, so that the vent pipe radiation is mitigated. Therefore, the compressors can be safely operated and refueled from this outside location. No actions are required in the RB to supply the long-term pneumatic supply.

The makeup connections in the seismic isolation spaces are near the vent pipes (more so on Unit 2 than Unit 1) and possibly subject to gamma dose from the pipe once venting starts. Therefore, the hose connections to the makeup stations in the seismic isolation space will be made before venting starts at approximately 7-8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br />. The hose connections will be fitted with quick disconnects to aid in making this a simple action during event response. The hose connection timing will be validated per HCVS-FAQ-13.

Electrical makeup location The HCVS electrical supply for the first 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> is from the station 24/48 voe battery system. This backup power supply is aligned at the 49 foot level of the control building at the same panel as the HCVS Radiation Monitor, adjacent to the MCR. As previously stated, this location is in the control building inside the MCR boundary and is acceptable for the duration of the event. The long-term electrical supply for the HCVS is from the FLEX Diesel Generators which can re-power the normal suQ_ply buses to the

HCVS controls and instruments or re-power the 24/48 VDC battery chargers.

The FLEX generators are located in the FLEX DG enclosure which is east of the RBs and the Emergency Diesel Generator (EOG) building (i.e., see Operator Action Maps.pdf). The location is on the opposite side of the RBs from the HCVS pipes and outside the RBs so that there are no concerns with operation of the FLEX DGs including refueling operation.

No electrical lineups need be made in the RB for the FLEX DG to supply the needed HCVS components, only inside the EDG building which is not a dose or temperature concern area.

Remote Operating Station - Temperature Evaluation:

Calculation BNP-MECH-FLEX-0001 documents the Reactor Building Heatup Analysis under ELAP conditions in which all ventilation, heating and cooling are de-energized. This analysis was used for development of the FLEX actions per order EA-12-049, but since the same Extended Loss of AC Power (ELAP) conditions apply to the EA-13-109 order, this analysis can be used to estimate the temperature at the ROS for HCVS purposes. Even though EA-13-109 requires the consideration of a severe accident, the existence of core damage and possible vessel breach will have no effect on the temperature at the ROS.

The applicable case in BNP-MECH-FLEX-0001 is Case 1 which models the operator actions in an ELAP. The

GOTHIC analysis results in Table 4 (i.e.,

Page 23 of 76) show that the maximum temperature on the 50' elevation is 121°F.

The actions at the ROS will be to open or close a maximum of three ~ inch valves so that they are expected to take less than 5 minutes. Furthermore, the operator will be entering the RB from outside through the 50' airlock door near the ROS so that the local temperature will be close to ambient outside the RB.

These temperatures, coupled with the short duration of action, are judged acceptable.

Remote Operating Station - Radiation Evaluation:

The bottom of the active core region is at 51' elevation. Therefore an operator would be roughly at core elevation while at the ROS. The shielding provided by the vessel, bio-shield, primary containment (PC), and distance from the core results in the 50' door location being a low dose rate area during normal full-power operation. The primary containment wall alone provides six feet of concrete shielding (i.e., drawing F-01132). Since the core is shutdown for the ELAP event, the dose rates from the core area will be lower than during operation.

The existence of core damage with possible reactor pressure vessel breach will not raise the dose levels at the ROS.

If the core were to melt through the lower vessel head, there would be loss of shielding from the vessel; however, there

would be additional distance to the ROS and additional concrete shielding provided by the pedestal. Any gap release to the suppression pool will contribute to RB dose rates, however the ROS is on the 50' level, two floors above the torus.

Therefore, the dose rate at the ROS due to the torus will be insignificant due to the five feet of concrete below the ground floor (i.e., drawing F-01787) as well as the additional concrete and distance afforded by the location being on the 50' elevation.

Likewise, any gap release that migrates back to the primary containment, will be shielded from the ROS by the six foot thick PC wall. In addition, the ROS is some 50 feet away from the PC wall.

Phase 2 ISE 01 2 BNP-MECH-FLEX-0005 documents the The NRC staff reviewed the Closed BSEP specific MAAP evaluation that information provided in the 6-Licensee to provide the site- verifies that an initial SAWA flow rate of month updates and on the [Staff evaluation to be specific MAAP evaluation that 300 gpm [gallons per minute] is sufficient ePortal. included in SE Section establishes the initial SAWA to protect containment as described in 4.1.1.3]

flow rate. NEI 13-02, Rev. 1. Cases 1 and 2 are the Calculation BNP-MECH-FLEX-cases that compare the containment 0005 is the BSEP specific MAAP response with a 300 gpm initial flow rate evaluation that verifies that an to the response with an initial flow rate of initial SAWA flow rate of 300 gpm 500 gpm which is the maximum flow rate is sufficient to protect required by NEI 13-02. Section 7 starting containment. The Electric Power on Page 25 of 178 presents the results of Research Institute (EPRI) study both analyses and compares them in (Technical Basis for Severe Table 7-1 and Figures 7-1 to 7-5. As can Accident Mitigating Strategies, be seen in these results, there is no 3002003301) assumes a 500 significant difference between the two gpm SAWA injection flow. BNP initial flow rates. evaluated a 300 gpm SAWA flow.

The 300 gpm SAWA flow starts at The site-specific MAAP evaluation for 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br />. At 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> SAWA flow SAWA, BNP-MECH-FLEX-0005, is is reduced to 100 gpm.

available for review on the ePortal. The calculations assume a 2923 MWt. The calculation concludes that with the SAWA flow

indicated, wetwell venting is maintained, the peak suppression pool air space pressure is 84.7 per square inch absolute (psia)

(70 psig) which is :5 70 psig acceptance pressure, and peak suppression pool water level is 17.1 feet.

No follow-up questions.

Phase 2 ISE 01 3 The instrumentation required for SAWA The NRG staff reviewed the Pending performance and monitoring consists of information provided in the 6-Licensee to demonstrate how drywell pressure, torus level and the month updates and on the [Staff evaluation to be instrumentation and equipment SAWA flow instrument. The only ePortal. included in SE Sections being used for SAWA and operating equipment for SAWA is the 4.5.1.2 and 4.5.1.3]

supporting equipment is SAWApump. The drywell pressure and torus capable to perform for the level indications are RG 1.97 sustained operating period The drywell pressure instruments are compliant and are acceptable as under the expected CAC-PT-1230 (i.e., one for each unit). qualified.

temperature and radiological The torus level instruments are CAC-LT-conditions. 2601 (i.e., one for each unit). These The SAWA flow instrument instruments are safety-related and qualificaitons for temperature and qualified to IEEE-323-1974 and IEEE- radiation need to be included in 344-1975. They are qualified to 148.8°F Table 1 of the FIP.

but are expected to see no more than 132°F during the ELAP. They are No follow-up questions.

qualified to 8.36E6 R total integrated dose and expected to receive no more than 9.13E5 R total integrated dose in the first seven days after the event starts.

Therefore, both instruments are qualified for the conditions they would see during a severe accident.

The drywell pressure and torus level instruments are powered from safety-related instrument buses that will be re-powered by the FLEX generators so that they will remain in service without offsite power or Emergency Diesel Generators.

In the case that the FLEX generators are not initially available, both instruments can be repowered from the station's division two 24 VDC battery for at least 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />.

The SAWA flow instrument is mounted on the FLEX/SAWA pump and is powered by the pump's generator. The SAWA pumps are stored in the FLEX Storage Building so that they will be available after the event. The SAWA pump is moved to the area outside of the RB, near the CST.

This area is on the opposite side of the RB from the vent pipe so that radiation is not a concern. Additionally, since the pump is outside, it will not be in an area of excessive temperature due to the accident.

The FLEX/SAWA pumps are refueled in accordance with BSEP's FLEX Support Guidelines developed in response to EA-12-049.

Phaes 2 ISE 01 4 BNP-MECH-FLEX-0005 documents the The NRC staff reviewed the Closed MAAP calculation of containment information provided in the 6-Licensee to demonstrate that response over a 7 day period during a month updates and on the [Staff evaluation to be containment failure as a result severe accident with only Severe ePortal. included in SE Section of overpressure can be Accident Water Addition and the wetwell 4.2]

prevented without a drywell vent in service for containment protection. Calculation BNP-MECH-FLEX-vent during severe accident This analysis demonstrates that the 0005, "Brunswick Nuclear Plant conditions. containment pressure remains below the MAAP 5.02 Analysis to Support Primary Containment Pressure Limit SAWA Strategy," Revision 1 (PCPL} and the temperatures remain low shows the MAAP calculation of enough that failure due to high containment response over a 7 temperatures is avoided. day period during a severe accident with only SAWA and the The base case is Case 1 in this report. wetwell in service for containment Table 7-1 on Page 26 of 178 contains a protection. The EPRI study table of results. The peak drywell air (Technical Basis for Severe

space pressure is shown as 89.9 psia. Accident Mitigating Strategies, This is the pressure at the time of vent 3002003301) assumes a 500 opening. The MAAP program was set to gpm SAWA injection flow. BNP initiate the vent based on wetwell evaluated a 300 gpm SAWA flow.

pressure at PCPL of 84.7 psia. The 300 gpm SAWA flow starts at 8 drywell pressure is slightly higher based hours. At 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> SAWA flow is on the differential pressure across the reduced to 100 gpm. The wetwell water. However, it is clear from calculation assumed a 2923 MWt.

the graph of drywell pressure (PRB(2)) on The calculation concludes that Page 66 of 178 that the pressure drops drywell temperature peaks at rapidly when the vent is opened and roughly 581°F and drops below never returns to a pressure near the 300° after SAWA flow is initiated.

PCPL. With 300 gpm SAWA followed by 100 gpm SAWA flow starting at The peak drywell temperature is shown 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> and continuing for 168 as 581°F on Table 7-1 on Page 26 of 178. hours will not result in a large As seen in the Figure 7-2 graph on Page increase in the suppression pool 28 of 178, the drywell temperature water lever that could potentially (TGRB(2)) rapidly decreases when the challenge the operation of the vent is opened and never increases for HWV.

the rest of the 7 day analysis period.

No follow-up questions.

The site-specific MAAP evaluation for SAWA, BNP-MECH-FLEX-0005, is available for review on the ePortal.

Phase 2 ISE 01 5 During a severe accident in which the The NRC staff reviewed the Closed SAWA/FLEX pump is in use to add water information provided in the 6-Licensee to demonstrate that to containment, the pump will be located month updates and on the [Staff evaluation to be there is adequate between the CST and the Reactor ePortal. included in SE Section communication between the Building (RB). The primary accident 4.1]

MCR and the operator at the communication means will be the 800 The communication methods are FLEX pump during severe MHz radio system. the same as accepted in Order accident conditions. EA-12-049.

As part of the response to NRC Order EA-12-049, BSEP assumed that permanently No follow-up questions.

installed plant communications systems would not be available during an ELAP.

Instead, BSEP primarily utilizes an 800 MHz radio system consisting of 500 hand held radios for onsite

communications. These radios are stored in reasonably protected buildings, including the FLEX Storage Building, to meet the requirements of EA-12-049.

This information was provided in response to NTTF Recommendation 9.3, by a letter dated October 31, 2012 (ADAMS Accession No. ML12311A299) and supplemented by a letter dated February 22, 2013, Carolina Power &

Light Company's and Florida Power Corporation's Response to Follow-Up Letter on Technical Issues for Resolution Regarding Licensee Communication Submittals Associated With Near-Term Task Force Recommendation 9.3 (ML 1305BA045). This information was assessed by the NRC staff and a Staff Evaluation was issued for this assessment. This was provided in ML13093A341, Brunswick Steam Electric Plant, Units 1 and 2 - Staff Assessment in Response to Information Request Pursuant to 1 0 CFR 50.54(f) - 9.3, Communication Assessment.

The radios will enable the MCR to communicate with operators in the field at the SAWA/FLEX pump locations.

Phae 2 ISE 01 6 The SAWM flow instrumentation (i.e., The NRC staff reviewed the Pending sensor and meter) will be permanently information provided in the 6-Licensee to demonstrate the installed on each SAWA/FLEX pump. month updates and on the [Staff evaluation to be SAWM flow instrumentation Electrical power is provided by the pump's ePortal. included in SE Section qualification for the expected 12 VDC electrical system. 4.4.1.3]

environmental conditions. The SAWA flow instrument The flow instrumentation is qualified for qualificaitons for temperature and use on fire pumps in outside ambient radiation need to be included in conditions. SAWA pumps are staged Table 1 of the FIP.

outside for use, between the CST and Reactor Buildino. No follow-up questions.

ML18068A627 OFFICE NRR/DLP/PBEB/PM NRR/DLP/PBMB/LA NRR/DLP/PBEB/BC NRR/DLP/PBEB/PM NAME RAuluck Slent TBrown RAuluck DATE 3/13/18 3/12/18 3/22/18 3/22/18