Text

Correction to the Report for the Audit of Lic. Responses to Interim Staff Eval. Open Items Related to U.S. Nuclear Regulatory Commission Order EA-13-109 to Modify Lic. with Regard to Reliable.....Supersedes ML18094A804
	ML18130A921
Person / Time
Site:	Monticello
Issue date:	05/14/2018
From:	Rajender Auluck; Beyond-Design-Basis Engineering Branch
To:	Church C; Northern States Power Company, Minnesota
	Auluck R, NRR/DLP, 415-1025
References
	CAC MF4376, EPID L-2014-JLD-0052
	Download: ML18130A921 (38)
	v • d • e

Monticell UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 14, 2018 Mr. Christopher R. Church Senior Vice President Northern States Power Company -

Minnesota Monticello Nuclear Generating Plant 2807 West County Road 75 Monticello, MN 55362-9637

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT-CORRECTION TO THE 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 NO. MF4376; EPID L-2014-JLD-0052)

Dear Mr. Church:

By letter dated April 10, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18094A804), the U.S. Nuclear Regulatory Commission (NRC) issued an audit report of the staff's assessment of the status of open items identified in the interm staff evaluations (ADAMS Accession Nos. ML15082A167 and ML16244A120, respectively) of the licensee's Phase 1 and Phase 2 overall integrated plans associated with NRC Order EA-13-109, "Order to Modify Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions," for Monticello Nuclear Generating Plant.

The last four pages (pages 26-29) of the staff's audit report were inadvertently omitted from the electronic version of the document and not included in the final letter that was transmitted on April 10, 2018. The purpose of this letter is to provide the complete audit report as shown in the enclosure. The complete audit report provided herein supersedes the audit report included in the April 10, 2018, letter.

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

Docket No. 50-219

Enclosure:

Audit report cc w/encl: Distribution via Listserv Sincerely, Rajender Auluck, Senior Project Manager Beyond-Design-Basis Engineering Branch Division of Licensing Projects Office of Nuclear Reactor Regulation

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 NORTHERN STATES POWER COMPANY - MINNESOTA MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 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 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. ML14183A412), Northern States Power Company - Minnesota (NSPM, the licensee) submitted its Phase 1 OIP for Monticello Nuclear Generating Plant (MNGP, or Monticello). By letters dated December 16, 2014, June 22, 2015, December 17, 2015 (which included the combined Phase 1 and Phase 2 OIP), June 17, 2016, December 19, 2016, June 14, 2017, and December 21, 2017 (ADAMS Accession Nos.

ML14353A215, ML15173A176, ML15356A120, ML16169A309, ML16354A666, ML17166A051, and ML17355A508, 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 Monticello by letters dated April 2, 2015 (ADAMS Accession No. ML15082A167), and September 6, 2016 (ADAMS Accession No. ML16244A120), respectively. When developing the IS Es, 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 March 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 ISEs. As part of the preparation for this audit call, 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 Monticello. The open items are taken from the Phase 1 and Phase 2 ISEs issued on April 2, 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 Monticello, 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 FIP, 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 NRG staff's review of these open items may change if the licensee changes its plans as part of final implementation. Changes in the NRG staff review will be communicated in the ongoing audit process.

Attachments:

1. Table 1 - NRG 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 Organization Team Lead/Sr. Proiect Manaaer Raiender Auluck NRR/DLP Project Manager Support/Technical Support - Containment / Ventilation Brian Lee NRR/DLP Technical Support - Containment I Ventilation Bruce Heida NRR/DLP Technical Support - Electrical Kerby Scales NRR/DLP Technical Support - Balance of Plant Garry Armstrong NRR/DLP Technical Support - l&C Steve Wyman NRR/DLP Technical Support-Dose John Parillo NRR/DRA

Table 2 - Audit Documents Reviewed Calculation 16-006, "Hard Pipe Vent 08 Battery HCVS 125VDC Battery Calculation," Revision 1 Engineering Change (EC) 23964 - FLEX 480 V Diesel Generator Sizing Calculation 94-017, "Calculation of Alternate Nitrogen System Supply Pressure and Spare Bottle Inventory," Revision 1 OB Calculation 16-011, "Calculation of HPV System Dedicated Nitrogen Supply and Pressure Requirements," Revision OA Calculation 16-055, "Monticello GOTHIC Analysis for the Hardened Contianment Vent Project,"

Revision 0 Calculation 16-054, "MNGP HCVS Radiological Assessment," Revision 0 Calculation 16-019, "Monticello Hardened Containment Vent System (HCVS) Capacity Analysis and Verification of Suppression Pool Decay Heat Capacity," Revision 0 Engineering Evaluation (EE) 26081 Reasonable Protection Evaluation Grade for HCVS Tornado Missile Barrier Calculation 16-032, "Hardened Containment Vent Pipe Supports HPVH1, HPVH2, HPVH3, and HPVH4," Revision 11 Calculation 16-012, "Pipe Stess Analysis of Hard Pipe Vent," Revision 0 Calculation 16-003, "Evaluation of HPV Missile Barrier - Lower Frame," Revision 0 Engineering Change (EC) 28557 - PT-7251 B - Severe Accident Temperature Conditions Engineering Evaluation EC 28582 - BDBEE Environmental Conditions for L T-7338B, Revision 0 Environmental Qualification (EQ)98-039 - Rosemount Pressure Transimitter Series A (DOR),

Revision 0 Environmental Qualification (EQ)08-016 - Rosemount 1154 Transimitters, Revision 1 Engineering Evaluation EC 28546 - BDBEE Environmental Conditions for A0-4539 and AO-4540, Revision 1 Specification NPD-M-39, "Specification for Valve Requirements for Pneumatic Operated Butterfly Valves for the Hard Pipe Vent System," Revision 8 Qualification Summary Report 04518900-QSR - HCVS Radiation Monitoring System (DC & AC Input Power Supplies), Revision C Operations Manual Section B.08.08-01, "Plant Communications Systems," Revision 7 Operations Manual Section A.8-06.02, "Repower PAB PBX Phone System with Portable Generator," Revision 3 Engineering Change (EC) 26083, "Hardened Containment Venting System NRC Order EA 109 Phase 1," Revision 0 Operations Manual Section C.5.-3505, "Venting Primary Containment," Revision 14 Calculation 16-002, "Evaluation of HPV Missile Barrier - Upper & Intermediate Frames,"

Revision 2 Calculation 16-067, "HCVS Radiation Detector Support Evaluation," Revision O Calculation 16-059, "Seismic Evaluation of SPOTMOS Panel C-289B," Revision 0 Calculation 16-065, "Seismic Evaluation of Panel C-292," Revision 0 Calculation 03-008, "AOV Component Calculation, Hard Pipe Vent Valves, A0-4539 and AO-4540," Revision 5 EPRI Technical Report 3002003301 - Technical Basis for Severe Accident Mitigating Strategies, Volume 1 Engineering Evaluation 28694 - Evaluation of Radiological Conditions at the Southside of the Radwaste Building during Hard Pipe Vent (HPV) Use As An Optional Location for the Portable Diesel Pump Environemental Qualification (EQ)98-026, "Limitorque Motor Operators (50.49)," Revision 2 Engineering Evaluation 608000000102 - SAWA Flowrates and Torus Water Levels Calculation 16-057, "3rct Floor EFT Exhaust Fan," Revision 0 Calculation 16-022, "Ventilation Requirements for Batteries Located on the Third Floor of theft Building," Revision 0 Specifications for Model EL 2200 Electromagnetic Flow Meter BWROG-TP-008, "Severe Accident Water Addition Timing" BWROG-TP-011, "Severe Accident Water Management Supporting Evaluations"

Monticello Nuclear Generating Plant 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 1 A calculation has been performed that The NRC staff reviewed the Closed confirms that the HCVS battery and information provided in the 6-Make available for NRC staff battery charger are sized adequately. The month updates and on the

[Staff evaluation to be audit the final sizing evaluation results of the analysis show that the ePortal.

included in SE Section for HCVS batteries/Battery battery is adequately sized to supply 3.1.2.6]

charger including incorporation power to the HCVS devices for twenty-The licensee stated that all into. FLEX DG loading four (24) hours following the onset of an electrical power required for calculation.

ELAP. The analysis results also show that operation of HCVS components is the minimum calculated terminal voltage provided by the HCVS 125 voe at the devices is above the minimum battery and battery charger.

voltage required for each HCVS device while being supplied from the battery.

The battery sizing calculation 16-006, "Hard Pipe Vent D8 Battery The design allows for use of the Diverse HCVS 125VDC Battery and Flexible Coping Strategies (FLEX)

Calculation," Revision 1 equipment (i.e. FLEX generator) to power confirmed that the 125 voe the system after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The design battery has a minimum capacity incorporates a manual, break-before-capable of providing power for 24 make transfer switch to transfer the load hours without recharging, and from the normal HCVS power supply to therefore is adequate.

250VDC battery number 16. During an ELAP event, the 16 battery, through its The licensee provided associated battery charger, will be Engineering Change (EC) 23964 connected to and powered from the FLEX

- FLEX 480 V Diesel Generator portable diesel generator per procedure.

Sizing, which discusses re-powering of the HCVS 125 voe An engineering evaluation was performed battery charger using the FLEX to demonstrate that the FLEX 480 V DG.

Diesel Generator is of adequate size to support these loads. The evaluation No follow-uo questions.

determined that the FLEX 480 V Diesel Generator is capable of supplying the battery chargers for the 11, 12, 13, and 16 batteries at current limits. Therefore, the FLEX 480 V Diesel Generator has the required capacity to supply the HCVS loads since it is sized for the full capacity of the battery chargers.

The calculation and evaluations have been provided to the NRC on the eportal.

Phase 1 ISE 01 2 A calculation has been performed that The NRC staff reviewed the Closed confirms that the HCVS two (2) nitrogen information provided in the 6-Make available for NRC staff supply systems that provide pneumatic month updates and on the

[Staff evaluation to be audit documentation of the capacity to the HCVS rupture disc and ePortal.

included in SE Section HCVS nitrogen pneumatic containment isolation valves are sized 3.1.2.6]

system design sizing and adequately. This calculation determined Calculation 94-017, "Calculation location.

that one (1) nitrogen bottle is required to of Alternate Nitrogen System fully burst the HCVS rupture disc and two Supply Pressure and Spare Bottle (2) nitrogen bottles are required to actuate Inventory," Revision 10B and the primary containment isolation valves Calculation 16-011, "Calculation over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

of HPV System Dedicated Nitrogen Supply and Pressure Two (2) new nitrogen supply systems are Requirements," Revision OA installed in the 931' east Turbine Building discusses the pneumatic design with a remote manual operating station and sizing.

located south of the nitrogen bottles near the B Alternate Nitrogen supply.

For rupture disc, the licensee Pneumatic tubing was routed through the determined that one bottle of Turbine Building, Condenser Room, nitrogen can rupture the disc in 12 Reactor Core Isolation Cooling (RCIC) minutes (which is less than the Room, and Torus Room to the HCVS required 15 minutes) to supply rupture disc and containment isolation nitrogen upstream for HCVS valves. The primary location for control of operation. A spare nitrogen bottle the HCVS remains in the third floor will be stored in the Monticello Emergency Filtration Train (EFT) Building warehouse on site.

at the Alternate Shutdown System (ASDS) panel.

For hard pipe vent (HPV) supply, the licensee determined that 2 bottles of nitrogen will be needed The design of the new HCVS nitrogen for 8 air operated valves (AOV) system is provided in Figure 01 2-1 of the actuations for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . An Sixth 6-Month Status Update submittal.

additional minimum of 12 nitrogen bottles will be needed for 6 days The calculation and drawings for the new after the initial 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for more nitrogen systems were provided to the AOV actuations for the HCVS.

NRC on the eportal.

No follow-up questions.

Phase 1 ISE 01 3 The primary operating station (POS) for The NRC staff reviewed the Closed the HCVS is in the third floor of the EFT information provided in the 6-Make available for NRC staff building and includes the controls for the month updates and on the

[Staff evaluation to be audit an evaluation of HCVS as well as the instruments used to ePortal.

included in SE Sections temperature and radiological monitor drywell pressure, suppression 3.1.1.2 and 3.1.1.3]

conditions to ensure that pool level, HCVS radiation, and HCVS Calculation 16-055, "Monticello operating personnel can safely temperature. The remote operating GOTHIC Analysis for the access and operate controls station (ROS) is located in the 931' Hardened Containment Vent and support equipment.

elevation of the turbine building east side.

Project," Revision O indicates that The nitrogen bottle rack, controls, and the temperature in the Emergency indicators are located at the north end of Filtration Train (EFT) building 931' east and the ROS valves are located third floor (location of the primary at the south end of 931 ' east.

operaring station (POS)) would peak at 135°F in the summer at Dose rates due to the Beyond Design 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . By 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , Basis External Event (BDBEE) and the supplemental ventilation will be HCVS order severe accident conditions installed per Procedure C.5-4503.

assumed in the containment atmosphere The supplemental ventilation will during H PV operation were determined by maintain the temperature below calculation using the methodology in NEI-120°F. Figure 7.2-1 indicates the 13-02, Rev 1 and HCVS-WP-02, Rev 0.

ETF Building 3rd floor The seven day integrated dose values at temperature varies between the POS and ROS locations are well 110°F and 100°F with the daily within the dose limit of 5 rem. Transit diurnal temperature variation after paths and locations outside of the Reactor supplemental ventilation is and/or HPCI Building have unlimited installed. The NRC staff access up to 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months after ELAP.

requested clarification that the Additionally, transit paths are acceptable high temperature in the POS for short durations after venting has would not hinder operators ability started based on the expected peak dose to take the required actions. The rates. The FLEX Pump and FLEX licensee responded that the work Generator deployment locations were in the POS is classified as liqht evaluated for a 7-day integrated dose and duty and consists of manipulating selected locations are accessible. Dose hand switches and peroidic the operator receives is administratively monitioring light indicators and controlled by health physics personnel to indicator readings. Expected stay ensure set dose rates and dose limits are times are 10 minutes or less.

not exceeded.

Work in high temperature environments is controlled by the Temperature in the EFT building third Monticello Safety Manual.

floor (e.g. POS) during an ELAP in the summer will peak at approximately 135°F In winter, the same procedure at 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . By hour 12, supplemental (Procedure C.5-4503) instructs ventilation will be installed per procedure operators to use portable heaters and room temperature will then be as needed to maintain the maintained below 120°F for the duration temperature above 40°F.

of the 7 day period. Room temperature in the winter will drop to 35°F after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months The licensee concluded the and 0°F at the end of 7 days with no summer temperature at the mitigating actions taken. Procedures remote operating station (ROS) direct operators to add portable heaters are not a concern since there are as needed within 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months upon initiation no heat loads. There is no of an ELAP to maintain EFT building third equipment adversely affected by floor temperatures above 40°F.

cold temperatures. The ROS is not continuously occupied.

Temperature in the Turbine Building 931' Operators can perform required east side corridor (near the ROS) in the actions independent of the local winter will drop to 29°F after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and ROS temperature.

0°F at the end of 7 days with no mitigating actions taken. HCVS equipment in this Calculation 16-054, "MNGP area can perform its function in these low HCVS Radiological Assessment,"

temperature conditions and therefore is Revision O was performed to acceptable. Summer peak temperatures determine the integrated radiation in this area are not a concern due to a dose due to HCVS operation.

lack of heat loads in the area during an The NRC staff reviewed this ELAP.

calculation and determined that the licensee used conservative The different pathways between the assumptions and followed the Reactor Building, EFT Building, and guidance outlined in NEI 13-02 Turbine Building were analyzed and it Rev.1 and HCVS-WP-02 Rev.O.

was determined that there are no Based on the expected integrated substantial heat sources in these areas whole body dose equivalent in the that would cause a significant change in POS and ROS and the expected temperature.

integrated whole body dose equivalent for expected actions The analyses and supporting information during the sustained operating described has been provided to the NRC period, the NRC staff believes in the eportal.

that the order requirements are met.

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 4 A calculation has been performed that The NRC staff reviewed the Closed confirms that the modified HCVS information provided in the 6-Make available for NRC staff configuration with the additional check month updates and on the

[Staff evaluation to be audit analyses demonstrating valve has the capacity to vent the ePortal.

included in SE Section that HCVS has the capacity to steam/energy equivalent of one (1) 3.1.2.1]

vent the steam/energy percent of the current licensed/rated Calculation MNGP 16-019 equivalent of one percent of thermal power of 2004 megawatt thermal Revision 1, "Monticello Hardened licensed/rated thermal power (MWT) while maintaining containment Containment Vent System (unless a lower value is pressure below design and Primary (HCVS) Capacity Analysis and justified}, and that the Containment Pressure Limit (PCPL).

Verification of Suppression Pool suppression pool and the Additionally, this analysis evaluates the Decay Heat Capacity,"

HCVS together are able to capacity of the Suppression Pool (SP) to determined that 1 % of the absorb and reject decay heat, absorb decay heat following a reactor licensed thermal power (2004 such that following a reactor shutdown from full power.

MWt) venting requirement is shutdown from full power 75,718 lbm/hr at 62 psig (PCPL =

containment pressure is The calculation has been provided to the 62 psig). The steady state restored and then maintained NRC on the eportal.

venting capacity at a torus below the primary containment pressure of 47.9 psig (maximum design pressure and the design pressure in the drywell primary containment pressure and the differential pressure limit.

between the drywell and wetwell with the torus completely full of water, is 79,737 lbm/hr (5.3% flow margin to 1 % thermal power requirement). Flow varies from roughly 20,000 lbm/hr at 5 psig to 90,000 lbm/hr at 55 psig.

No follow-up questions.

Phase 1 ISE 01 5 HCVS piping outside the Class I structure The NRC staff reviewed the Closed is designed for tornado/wind loads without information provided in the 6-Make available for NRC staff failure to ensure functionality of the HCVS month updates and on the

[Staff evaluation to be audit the seismic and tornado and safety related systems in the vicinity.

ePortal.

included in SE Section missile final design criteria for HCVS piping up to and including the 3.2.2]

the HCVS stack.

second primary containment isolation Engineering Evaluation (EE) valve is designed to safety related seismic 26081 Reasonable Class 1 requirements. HCVS piping Protection Evaluation Grade for downstream of the second containment HCVS Tornado Missile Barrier, isolation valve, although non-safety evaluated the HCVS stack. The related, is designed to seismic Class 1 as licensee's HCVS design meets it must remain functional following a the assumptions found in seismic event.

guidance document HCVS-WP-

04.

Analysis of the tornado/wind loads and seismic loading is documented in No follow up questions.

calculations performed to support the design of the HCVS piping. The analysis of the modified HCVS piping includes incorporation of wind, tornado, and updated seismic requirements to meet sections 5.1.1.6 and 5.2 of NEI 13-02.

Design basis loading requirements for wind, tornado, and seismic were used as described in the MNGP USAR, Section 12.02.

Portions of the HCVS outside of Class I structures will be protected from tornado missile impact up to 30 feet (ft) above grade. The HCVS design will meet assumptions found in guidance document HCVS-WP-04 which provides reasoning why protecting the HCVS 30 ft above qrade is not required. An Engineering Evaluation validated the guidance is applicable for use at MNGP. Missile barrier design requirements for tornado generated missiles, seismic, and wind loadings were used as described in the MNGP USAR, Section 12.02. Analysis of the missile barrier to these loading requirements is documented in calculations.

The calculations and analyses described above have been provided to the NRC on the eportal.

Phase 1 ISE 01 6 The POS for the HCVS is on the third The NRC staff reviewed the Closed floor of the EFT building and includes the information provided in the 6-Make available for NRC staff controls for the HCVS as well as the month updates and on the

[Staff evaluation to be audit the descriptions of local instruments used to monitor drywell ePortal.

included in SE Section conditions (temperature, pressure, suppression pool level, HCVS 3.1.1.4]

radiation and humidity) radiation, and HCVS temperature.

EC 26083 discusses the anticipated during ELAP and environmental conditions during severe accident for the The ROS is located on the south end of an accident at the locations components (valves, the 931' elevation of the Turbine Building containing instrumentation and instrumentation, sensors, east side. The nitrogen bottle rack, controls (l&C) components. The transmitters, indicators, controls, and pressure indicators are staffs review indicated that the electronics, control devices, located at the north end of the 931' environmental qualification met and etc.) required for HCVS elevation of the Turbine Building east the order requirements.

venting including confirmation side.

that the components are The primary control location is on capable of performing their The primary containment isolation valves the third floor of the EFT building.

functions during ELAP and (PCIVs) and associated solenoid valves Controls for the existing HPV are severe accident conditions.

(SVs) are installed in the vent piping near located on the C-292 Alternate the torus connection in the Reactor Shutdown System (ASDS) panel.

Building elevation 923' above the north east section of the torus. The suppression The remote operating station is pool level transmitter L T7338B is located on the 931' elevation of the in the torus room bay 9.

Turbine Building. Temperature for these areas evaluated in calc The radiation detector is installed 16-055. The calculation assumed adjacent to the pipe above the high a 95°F outdoor temperature. The pressure coolant injection (HPCI) room at calculation determined the ETF elevation 935'. The temperature element Bldg, 3rd floor peaks at -135°F is installed in the HPCI room adjacent to shortly after start of the event and the vent pipe at elevation 928'.

drops to approximately 100°F after mitigating actions are The drywell pressure transmitter implemented. The temperature PT7251B is located in the Reactor varies between 110°F and 100°F Building, elevation 985' south wall.

with the daily diurnal temperature variation.

Radiological Conditions:

The main control room was Radiological dose rates resulting from previously evaluated as part of HCVS venting were determined by Order EA-12-049.

calculation for each area using the methodology in NEl-13-02, Rev 1 and No follow up questions.

HCVS-WP-02, Rev 0.

Temperature/ Humidity Conditions:

Temperature conditions for each area have been determined by calculation, using the methodology in N El-13-02, Rev

1. An additional analysis was performed to determine the severe accident temperature in the torus room.

The calculations determined that key components necessary for HCVS venting are capable of performing their intended functions under ELAP and severe accident conditions.

The analyses and supporting information that support these conclusions have been provided to the NRC in the eportal.

Phase 1 ISE 01 7 The HCVS controls are located on the The NRC staff reviewed the Closed ASDS panel located on the third floor of information provided in the 6-Make available tor NRC staff the EFT building. Primary containment month updates and on the

[Staff evaluation to be audit documentation that pressure and suppression pool level ePortal.

included in SE Section demonstrates adequate indicators are located on the ASDS panel.

3.1.1.1]

communication between the Suppression pool temperature, HCVS remote HCVS operation temperature, and HCVS radiation The communication methods are locations and HCVS decision indicators are on the panel adjoining the the same as accepted in Order makers during ELAP and ASDS panel. These are the indicators EA-12-049.

severe accident conditions.

used by the Operator to monitor the primary containment and HCVS when No follow-up questions.

making decisions regarding use of the HCVS during severe accident conditions.

When dispatched from the control room, the Operator sent to the ASDS panel will have been given a containment pressure control band by the Control Room Supervisor per procedure. Procedural guidance for operating the HCVS is maintained both in the control room and at the ASDS panel. Therefore, the Operator actuating the HCVS from the ASDS panel requires no further communication.

Should actuation of the HCVS from the ASDS panel fail, the HCVS can be actuated by an Operator manipulating manual valves at the ROS, located on the east side of the 931 foot elevation of the Turbine Building. This Operator will be in communication with a second Operator who is at the ASDS panel monitoring the primary containment and HCVS. These Operators will be in communication via the telephone system. There is a phone on the ASDS panel and a phone in the Turbine Building, a short distance from the HCVS ROS.

The MNGP phone system is powered by the Non-1E Uninterruptable Power Supply (Y91), which is powered from the site non-essential 250 volt battery. A calculation determined that the non-essential 250 volt battery will maintain power to the portion of the site phone system supplied from Y91 energized for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months following an ELAP event.

Phones that remain energized include the phone at the ASDS panel, the Control Room Supervisor's phone in the Main Control Room, and the phone in the Turbine Building near the HCVS ROS.

In response to NRG Order EA-12-049 (Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design Basis External Events), NSPM developed and implemented FLEX Support Guidelines (FSGs) to provide pre-planned procedures to improve the stations capability to cope with beyond design basis events. As part of the FLEX response, MNGP has an FSG procedure to stage a 120 volt portable diesel generator and a procedure to use this generator to repower the phone system.

Timing studies performed as part of FLEX implementation have shown the phone system can be repowered from the portable diesel generator within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Since the phones required for communication at the ASDS panel and the HCVS ROS will be repowered from a portable diesel generator before power is lost from the site non-essential 250 volt battery, the phone system remains available at all times for communication between the Operator at the HCVS ROS and the Operator at the ASDS panel.

The calculation and procedures described in this response have been provided to the NRC on the eoortal.

Phase 1 ISE 01 8 The risk of hydrogen detonation and The NRC staff reviewed the Closed deflagration has been mitigated in the information provided in the 6-Provide a description of the design of the month updates and on the

[Staff evaluation to be final design of the HCVS to MNGP HCVS system by use of the ePortal.

included in SE Section address hydrogen detonation following elements:

3.1.2.11]

and deflagration.

The licensee's design is

- A check valve will be installed on the consistent with Option 5 of the HCVS piping at the reactor building roof NRC staff endorsed white paper to prevent ingress of air when venting HCVS-WP-03.

stops and the steam condenses. This will prevent a flammable mixture of gasses No follow-up questions.

from potentially building up within the piping upstream of the check valve.

Piping downstream of the check valve will be at a length less than the recommended run up distance in order to rule out detonation loading in this portion of the piping. HCVS piping where the check valve is installed will be routed slightly over the reactor building roof to allow for maintenance/testing accessibility, then routed upwards to direct effluent away from plant structures. This is consistent with Option 5 of NEI 13-02, Appendix H.

- A check valve will be utilized on the rupture disc pneumatic supply connection to the HCVS piping to prevent backflow to the remote operating station. With exception of the rupture disc supply, the HCVS piping is designed to have no interfaces with other plant systems. In addition, HCVS pneumatic system valves that open external to the system are designed to the system operating conditions. With these design features, the HCVS meets the requirement for minimizing the potential for hydrogen gas migration and ingress into site buildings.

This is consistent with the guidance provided in NEI 13-02, Section 4.1.2, Appendix H, HCVS-FAQ-05 and HCVS-WP-03.

- The HCVS release point will be modified from its current location of 3 ft above the Reactor Building roof and plenum exhaust to be vertical over the reactor building roof. The modified release point is at an elevation higher than the adjacent power block structures, which is approximately 145 ft off the ground. The existing "T" type exhaust at the top of the vent pipe will be replaced with a vertical exit to direct the effluent away from site structures and away from ventilation system intake and exhaust openings. A weather cap will be installed at the release point for protection of the pipe and newly installed check valve during normal operation, and will be designed to blow off if the vent is operated. The weather cap will be designed to blow off at a minimal interior pipe pressure to not impede the initial venting. This design allows for no permanently added resistance to piping for effluent flow. This is consistent with the guidance provided in NEI 13-02 Section 4.1.5, Appendix Hand HCVS-FAQ-04.

- With the exception of the rupture disc supply, the HCVS piping is designed to have no interfaces with other plant systems. In addition, HCVS pneumatic

~stem valves that open external to the system are designed to the system operating conditions. With these design features, the HCVS meets the requirement of minimizing unintended cross flow within the unit. MNGP is a single unit site, so cross flow between units is not a concern. This is consistent with the guidance provided in NEI 13-02, Sections 4.1.2, 4.1.4 and 4.1.6 and HCVS-FAQ-05.

The engineering change describing the above design elements has been provided to the NRC on the eportal.

Phase 1 ISE 01 9 The HCVS utilizes a dedicated The NRC staff reviewed the Closed penetration from the torus to HCVS information provided in the 6-Provide a description of the piping, which is routed through the month updates and on the

[Staff evaluation to be strategies for hydrogen control Reactor Building. The HCVS piping does ePortal.

included in SE Section that minimizes the potential for not pass through other buildings thus 3.1.2.12]

hydrogen gas migration and eliminating the potential for migration of The NRC staff's review of the ingress into the reactor hydrogen gas from the HCVS into other proposed system indicates that building or other buildings.

buildings.

the licensee's design appears to meet the requirement for A check valve is provided on the rupture minimizing the potential for disc pneumatic supply connection to the hydrogen gas migration and HCVS piping to prevent backflow to the ingress into the Reactor Building remote operating station. With exception or other site buildings.

of the rupture disc pneumatic supply, the HCVS piping is designed to have no No follow-up questions.

interfaces with other plant systems, and all valves that open external to the system are designed to the system operating conditions. Once the rupture disk is burst the pneumatic supply will be isolated to prevent migration of hydrogen gas into the pneumatic supply system.

Initial and periodic testing of the HCVS will be performed in accordance with manufacturer instructions and the NEI 13-02 guidance. This includes leak tests which will ensure leak tightness of the HCVS to prevent hydrogen gas ingress into the Reactor Building.

Finally, the HCVS outlet is above plant structures, and is designed to direct the vent discharge away from structures and ventilation inlets and outlets.

With these design features, the HCVS meets the requirement for minimizing the potential for hydrogen gas migration and ingress into the Reactor Building or other site buildings.

The design documents and procedures described in this response have been provided to the NRC on the eportal.

Phase 1 ISE 01 10 Reguired Instrumentation and Controls:

The NRC staff reviewed the Closed information provided in the 6-Make available for NRC staff As documented in the MNGP Overall month updates and on the

[Staff evaluation to be audit descriptions of all Integrated Plan (OIP), the following ePortal.

included in SE Section instrumentation and controls instrumentation and controls are required 3.1.2.8]

(i.e., existing and planned) for order compliance:

The existing plant instuments necessary to implement this required for HCVS (i.e. wetwell order including qualification

Valve Position Indication level instruments and drywell methods.

Effluent Discharge Radioactivity pressure instruments) meet the

Effluent Temperature requirements of Regulatory Guide

Containment Pressure (RG) 1.97.

Wetwell Level

Electrical Power The licensee provided analyses

Remote Operating Station Valves and/or supporting information of

Pneumatic Supply Pressure Indications the HCVS instruments and and Manual Valves controls (l&C), including a description of each component Qualification Methods:

and the qualification method. The staff's review indicates that the l&C components are consistent with the quidance in NEI 13-02 The OIP provides the following and its qualifications meet the information related to component order requirements.

qualification:

No follow-up questions.

"The HCVS instruments, including valve position indication, process instrumentation, radiation monitoring, and support system monitoring, will be qualified by using one or more of the three methods described in the ISG, which includes:

1. Purchase of instruments and supporting components with known operating principles from manufacturers with commercial quality assurance programs (e.g., IS09001) where the procurement specifications include the applicable seismic requirements, design requirements, and applicable testing.

2. Demonstration of seismic reliability via methods that predict performance described in IEEE 344-2004.

3. Demonstration that instrumentation is substantially similar to the design of instrumentation previously qualified."

All components were determined to have acceptable qualifications to meet the HCVS order requirements.

The analyses and supporting information that support these conclusions have been orovided to the NRC in the eoortal.

Phase 1 ISE 01 11 A calculation was performed that The NRC staff reviewed the Closed determined that the HCVS primary information provided in the 6-Make available for NRC staff containment isolation valves, A0-4539 month updates and on the audit documentation of an and A0-4540, will open under the ePortal.

evaluation verifying the maximum differential pressure expected

[Staff evaluation to be existing containment isolation during Beyond Design Basis External The NRC staff reviewed included in SE Section valves, relied upon for the Event (BDBEE) suppression pool venting calculation 03-088, "AOV 3.2.1]

HCVS, will open under the with greater than 20% margin. The valves Component Calculation, Hard maximum expected differential have been shown to open against a Pipe Vent Valves, A0-4539 and pressure during BDBEE and maximum expected differential pressure A0-4540," which discusses the severe accident wetwell of 76.7 psid.

valve/actuator information for the venting.

PCIVs.

The calculation has been provided to the NRC on the eportal.

The calculation determined the full opening maximum torque was 252 foot-pounds and the corresponding actuator capability at that required valve toque is 304 foot-pounds.

The NRC staff verified the actuator can develop greater torque than PCIV's unseating torque.

No follow-up questions.

Phase 2 ISE 01 1 NEI 13-02 Section 4.1.1.2 provides the The NRC staff reviewed the Closed following guidance in determining the information provided in the 6-Licensee to provide the plant maximum flow capacity:

month updates and on the

[Staff evaluation to be specific justification for SAWA ePortal.

included in SE Section

[Severe Accident Water 4.1.1.2.1 Sites may use SAWA 4.1.1.3]

Addition] flow capacity less capacity at 500 GPM based on SAWA provides cooling of core than specified in the guidance the generic analysis per reference debris limiting the drywell in NEI 13-02, Section 4.1.1.2.

27.

temperature. SAWA permits venting containment through the 4.1.1.2.2 Sites may use a SAWA wetwell vent without the necessity capacity equivalent to the site of having a drywell vent (see specific RCIC design flow rate if discussion for Phase 1 ISE 4 for less than 500 GPM (e.g., some wetwell vent capacity). SAWM sites have a RCIC design flow manages the water addition into rate of 400 or 450 GPM).

the wetwell such that the wetwell vent does not become blocked by 4.1.1.2.3 SAWA capacity less the water level and remains than specified in 4.1.1.2.1 or operational. SAWA and SAWM 4.1.1.2.2 should be supported by industry study (The EPRI study plant specific design (i.e., SAWA (Technical Basis for Severe flow rate determined by scaling, a Accident Mitigating Strategies, ratio of the plant thermal power 3002003301) assumes a 500 rating over the reference plant gpm SAWA injection flow) was power level multiplied by 500 based on a reference plant which GPM).

has the most limiting containment heat capacity in the US fleet and NEI 13-02 Appendix C describes the therefore is conservative.

basis for the reference plant SAWA flowrates (500 gpm initial flowrate, and NSPM used the SAWA injection then reduced to 100 gpm for remainder of flow rate for the reference plant the mission time). Guidance is provided prorated for the difference for determining plant specific flow rates between the reactor thermal based on scaling, using the ratio of the power level and the licensed specific plant thermal power to the reactor thermal power for reference plant thermal power.

Monticello.

Additional basis for determining the No follow-up questions.

reference plant SAWA flow rates is provided in Electiric Power Research Institute (EPRI) Technical Report 3002003301. The EPRI Report in turn references the State-of-the-Art Reactor Consequence Analyses (SOARCA) which provides the Peach Bottom (reference plant) specific analysis.

Based on the established guidance, the MNGP plant specific flowrates are determined using the scaling method:

Reference 12lant values:

Rated thermal power= 3514 MWth SAWA flow= 500 gpm MNGP calculation:

SAWA = 500 gpm * (2004/ 3514) = 285 gpm SAWM = 100 gpm * (2004/ 3514) = 57 gpm It should be noted that these values are different than those provided in the Phase 2 OIP. The original calculation used a reference plant thermal power of 3293 MWth, resulting is SAWA/SAWM values of 305/61 gpm.

The analyses and supporting information described above were provided to the NRC in the eportal.

Phase 2 ISE 01 2 Plant instrumentation for SAWM that is The NRC staff reviewed the Closed qualified to RG 1.97 or equivalent is information provided in the 6-Licensee to evaluate the considered qualified for the sustained month updates and on the

[Staff evaluation to be SAWA equipment and operating period without further ePortal.

included in SE Sections controls, as well as the ingress evaluation. The following plant 4.5.1.1, 4.5.1.2 and and egress paths for the instruments are qualified to RG 1.97:

The drywell pressure and torus 4.5.1.3]

expected severe accident level indications are RG 1.97 conditions (temperature, Pl-72518 (PT-7251 B) Primary compliant and are acceptable as humidity, and radiation) for the Containment Wide Range Pressure qualified.

sustained operating period.

Ll-73388 (L T-73388)

Calculation 16-054, "MNGP Suppression Pool Level HCVS Radiological Assessment,"

Revision O shows that radiological Passive components that do not need to conditions should not inhibit change state after initially establishing operator actions or SAWA SAWA flow do not require evaluation equipment and controls needed to beyond the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , at which time initiate and operate the HCVS they are expected to be installed and during an ELAP with severe ready for use to support SAWA/SAWM.

accident conditions.

The following additional equipment The NRC staff reviewed performing an active SAWA/SAWM calculation 16-054, "MNGP HCVS function is considered:

Radiological Assessment," and determined that the licensee used SAWA/SAWM flow instrument conservative assumptions and followed the guidance outlined in NEI 13-02 Rev.1 and HCVS-WP-The environmental (temperature) capability of the flow instrument has been documented in ISE Open item 7.

The deployment location for the flowmeter is inside the Turbine Building, east side, 931' elevation. In this location, the Turbine Building provides environmental protection from external events, and substantial radiation shielding from the HCVS vent line. Dose calculations performed determine that peak severe accident dose rate in this area is 0.186 R/hrwith a 7-day integrated dose of 15.5 R. This radiation level is not expected to have any adverse effect on operation of the flowmeter.

SAWNSAWM pump (Flex Pump)

The deployment and staging for the portable diesel pump is the same as FLEX strategies. The deployment routes and environmental operating conditions (temperature) have previously been addressed for FLEX. Planned staging locations are near the Intake Structure, Discharge Canal, or Cooling Tower Basins.

Dose calculations performed determine the peak accident dose rates and integrated 7-day dose in these areas:

Intake Structure-3.1 R/hr, 261 R (7-day integrated dose)

Discharge Canal- 0.15 R/hr, 122 R (7-day integrated dose)

Cooling Tower Basin (not calculated, but similar to Discharge canal) 02 Rev.O.

Based on the expected integrated whole body dose equivalent in the MCR and ROS and the expected integrated whole body dose equivalent for expected actions during the sustained operating period, the NRC staff believes that radiological conditions should not inhibit operator actions or SAWA equipment and controls needed to initiate and operate the HCVS during an ELAP with severe accident conditions.

The temperature evaluation addressed in Phase 1 Open Item

6 bounds the SAWNSAWM operation. For operation of equipment located outdoors, existing plant work controls remain applicable.

No follow-up questions.

An alternate staging location for a flood event requires suction from the Condensate Storage Tanks (CST). An engineering evaluation was performed to determine dose rates in a staging location south of the Radwaste Building. This evaluation concludes that the dose rates would be similar to the FLEX Diesel Generator south location, which are negligible.

These radiological conditions in the planned staging locations are not expected to affect pump operation.

SAWA/SAWM generator (FLEX generator)

Deployment and staging of the 480VAC portable diesel generator is the same as FLEX strategies. This is required to provide the power supply to the low pressure coolant injection (LPCI) valve via the LPCI swing bus. The deployment routes and environmental operating conditions (temperature) have previously been addressed for FLEX. Planned staging locations are near the Plant Administration Building (PAB) south entrance or east entrance.

Dose calculations determined the peak accident dose rates and integrated 7-day dose in these areas:

PAB south, negligible dose rate and 7-day dose

PAB east-negligible dose rate and 7-day dose These radiological conditions are not ex_Qected to affect generator operation.

Ingress and Egress Instrumentation (Pl-7251 Band Ll-7338B):

These instruments are located on the ASDS Panel in the EFT Building 3rd Floor. Dose calculations performed determine the peak accident dose rate in this area is 1.75mR/ hr. Access to this area will not be affected by the radiological conditions.

SAWA/SAWM flow instrument Dose calculations determined the peak dose rate associated with the transit path to the flow instrument (Turbine Building 931' east side) is approximately 5 R/hr.

Since the transit times to the area are short, ingress and egress are not expected to be impacted.

SAWA/SAWM pump (FLEX Pump)

As documented above, the radiological conditions for the deployment and staging locations are relatively low. The dose rates at the Intake Structure location could preclude access to that area; in that case, one of the alternate locations would be used. Access for operation and refueling of the pump would not be impacted by the radiological conditions.

SAWA/SAWM generator (FLEX generator)

As documented above, the radiological conditions for the deployment and staging locations are negligible. Access for operation and refueling of the generator would not be impacted by the radiological conditions.

[Note: The dose calculation performed does not consider radiation shine from the external radioactive plume. Station procedures will direct plant staff to monitor the radiological conditions in and around the plant during an emergency.

Based on the specific site conditions, equipment locations, transport paths, and stay times would be altered as necessary to minimize personnel dose.]

The analyses and supporting information described above were provided to the NRC in the eportal.

Phase 2 ISE 01 3 Egui~ment and Controls:

The NRC staff reviewed the Closed information provided in the 6-Licensee to demonstrate how The following instrumentation and month updates and on the

[Staff evaluation to be instrumentation and equipment equipment has been evaluated for the ePortal.

included in SE Sections being used for SAWA and expected temperature and radiological 4.4.1.3 and 4.5.1.2]

supporting equipment is conditions (Reference the response to The NRC staff confirmed the Pl-capable to perform for the Phase 2 Open Item 2):

7251 B Primary Containment Wide sustained operating period Range Pressure and Ll-7338B under the expected Pl-7251 B Primary Containment Suppression Pool Level are temperature and radiological Wide Range Pressure previously qualified for R.G. 1.97 conditions.

Ll-7338B Suppression Pool Level accident monitoring. The flow SAWA/SAWM flow instrument instrument qualification is SAWA/SAWM pump (FLEX discussed in Phase 2 Open Item pump)

7 below.

SAWA/SAWM generator (FLEX generator)

The NRC staff reviewed calculation 16-054, "MNGP HCVS This equipment is capable of performing Radiological Assessment," and during the sustained operating period in determined that the licensee used the expected environmental conditions.

conservative assumptions and followed the guidance outlined in NEI 13-02 Rev.1 and HCVS-WP-One additional active component requires review, M0-2014 Residual Heat Removal (RHR) Division 1 LPCI Inboard Injection Valve. This valve would be electrically opened from the Main Control Room in order to establish the reactor pressure valve (RPV) injection path. The valve is located in the Reactor Building, 931' elevation, East Shutdown Cooling Room.

The motor operated valve would be cycled within the first eight hours of the event.

Temperature:

A calculation determined environmental temperature profiles for various locations in the Reactor Building. The temperature in the East Shutdown Cooling Room is not calculated. It is conservative to assume this room is at the same temperature as the Torus room (highest value in the Reactor Building}, which reaches approximately 170°F at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for the severe accident case.

The Environmental Qualification (EQ)

Report applicable to M0-2014 specifies a peak qualification temperature of 343°F, with test temperatures at or above 251 °F for 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months . Based on this, there is high confidence the valve can be electrically opened in the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of the accident.

Radiation:

A dose rate calculation determined dose rates and total 7-day integrated dose for various locations, including the Reactor Building. The dose rates in the East 02 Rev.O.

Based on the expected integrated whole body dose equivalent in the MGR and ROS and the expected integrated whole body dose equivalent for expected actions during the sustained operating period, the NRG staff believes that the order requirements are met.

No follow-up questions.

Shutdown Cooling Room were not calculated. It is conservative to assume this room has the same radiological conditions as the Torus room, which is the compartment below this area (does not account for any shielding effect from 931' floor slab). The peak dose rate in the Torus room (near CV4539/ CV4540) is 2.7E5 R/hr. The 7-day integrated dose is 1.14E7 R.

The environmental qualification (EQ) report applicable to M0-2014 specifies a demonstrated total equivalent gamma dose of 2.04E8 Rad. Assuming that 1 Rem

= 1 Rad for this case, the qualified dose exceeds the calculated accident dose.

Based on this, there is high confidence the valve can be electrically opened in the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of the accident.

The analyses and supporting information described above were provided to the NRC in the eportal.

Phaes 2 ISE 01 4 The SAWA/SAWM strategy requires The NRC staff reviewed the Closed demonstration that the wetwell vent will information provided in the 6-Licensee to demonstrate that remain available for the 7-day mission month updates and on the

[Staff evaluation to be containment failure as a result time (i.e. water level does not rise above ePortal.

included in SE Section of overpressure can be the elevation of the vent connection on 4.2]

prevented without a drywell the torus). An Engineering Evaluation has BWROG-TP-15-008 vent during severe accident been performed to determine wetwell demonstrates adding water to the conditions.

water level during the event. The reactor vessel within 8-hours of evaluation determines the SAWA and the onset of the event will limit the SAWM flowrates; the RPV injection rate is peak containment drywell specified as 285 gpm for four hours, then temperature significantly reducing 57 gpm for the remainder of the 7 days.

the possibility of containment The resulting wetwell water level at 7 failure due to temperature.

days is approximately 24.2 feet (elevation Drywell pressure can be 922.95 feet), which is below the wetwell controlled by venting the vent elevation of 925.21 feet (upper limit on water level instrument is 925 feet). The suppression chamber through the analysis is conservative since no mass suppression pool.

loss through the HPV is credited. Based on this analysis, the wetwell vent BWROG-TP-011 demonstrates capability is maintained for a 7-day that starting water addition at a mission time.

high rate of flow and throttling after approximately 4-hours will The wetwell vent has been designed and not increase the suppression pool installed to meet NEI 13-02 Rev 1 level to that which could block the guidance, which ensures that it is suppression chamber HCVS.

adequately sized to prevent containment overpressure under severe accident As noted under Phase 1 open conditions. The SAWM strategy will item #4, the vent is sized to pass ensure that the wetwell vent remains a minimum steam flow equivalent functional for the period of sustained to 1 % rated core power. This is operation. MNGP will follow the guidance sufficient permit venting to (flow rate and timing) for SAWA/SAWM maintain containment below the described in BWROG-TP-15-008 and lower of PCPL or of design BWROG-TP-15-011. The wetwell vent pressure.

will be opened prior to exceeding the No follow-up questions.

PCPL value of 62 PSIG. Therefore, containment over pressurization is prevented without the need for a drywell vent.

The analyses and supporting information described above were provided to the NRC in the eportal.

Phase 2 ISE 01 5 NEI 13-02 Appendix C provides a The NRC staff reviewed the Closed description of the Severe Accident Water information provided in the 6-Licensee to demonstrate how Management strategy, and recognizes month updates and on the

[Staff evaluation to be the plant is bounded by the insights gained from EPRI Technical ePortal.

included in SE Section reference plant analysis that Report 3002003301.

4.2.1.1]

shows the SAWM [Severe Engineering Evaluation Accident Water Management]

EPRI Technical Report 3002003301 608000000102, "SAWA Flowrates strategy is successful in performs a comprehensive analysis of two and Torus Water Levels,"

making it unlikely that a reference plants. The approach develops demonstrates that the initial water drywell vent is needed.

several cases using various water injection rate of 285 gpm for 4 addition/ venting strategies, and a range hours followed by 57 gpm for the of boundinq plant parameters. Each case remainder of the event (7 days) determines whether the strategy is successful in preventing primary containment failure. In order to demonstrate that the reference plant analyses are applicable to the Mark I fleet, plant-to-plant variability was assessed. This is presented in section 4 of the report. Plant specific data were reviewed to determine if there were variations that would influence the overall conclusions from the technical analysis.

Some of the potential plant variations were investigated further to confirm that the overall conclusions using the reference plant would be applicable to the other Mark I plants. The following table provides the parameters that were reviewed, including the MNGP specific values:

jP*ammtr I Man:"fAtet l,l"NGP i COf!;,&~-G<"!t '""Sa,; CaMC,~)'

~:,,;1:0;Yjd,i; ciJbi<: f~t i,,.c.'.'<-,c,,:,:'>."""'"'i~'Ms.C<',",t, j 6~6 k.y f:

,.,..,.1.,,,...,-i*,.... '"~

-*-~.....................,..,,...,.,...

l...JOC>OO G""fj{.i.;jt}:);:i g.1ficn~ T?'2-,inh~ i).3':<:~,s.

'J'!i :o wv, $pi e*1e* ~ight ! 7 ~ to 3 if"i::""H

! 6 ::: ncnes

J:,,:;::...,~

c.,-rnt<:!

,,1.a\\.,

Based on the results, plant-to-plant variations would not be expected to significantly influence the overall conclusions. Therefore, MNGP is bounded by the reference plant analysis.

Additional evaluation of the severe accident water management strategy was performed by the BWROG (TP-15-011 ).

The purpose of the evaluation is to demonstrate that the Mark I (and Mark II) fleet is bounded by the reference plant analyses. This study addressed how along with the minimum available freeboard at the start of the event will not result in the water level increasing to block the wetwell vent even if operation action is not taken to monitor Torus water level and adjust water flow as needed.

No follow-up questions.

suppression pool level control could be achieved in a manner that maintains long term function of the wetwell vent, and determined if there would be adverse effects by controlling (limiting) flow rate.

The study concludes that plants with Mark I containments, with injection into the RPV, can maintain containment cooling and preserve the wetwell vent without a plant specific analysis. Since this is the planned strategy, MNGP is bounded by the conclusions of the BWROG evaluation.

The analyses and supporting information described above were provided to the NRC in the eportal.

Phase 2 ISE 01 6 The severe accident response strategies The NRC staff reviewed the Closed require coordinated communications information provided in the 6-Licensee to demonstrate that between the Main Control Room, ASDS month updates and on the

[Staff evaluation to be there is adequate panel for HCVS operation (EFT third ePortal.

included in SE Section communication between the floor), FLEX manual valve for SAWA flow 4.1]

MCR and the operator at the control (Turbine Building 930 east), and The communication methods are FLEX pump during severe the FLEX pump staging location (Intake the same as accepted in Order accident conditions.

area, discharge canal, or cooling towers).

EA-12-049.

Communication methods are the same as No follow-up questions.

accepted in Order EA-12-049 for FLEX strategies (Final Integrated Plan section 8.3). Communications necessary to provide on-site command and control of the response strategies can be effectively implemented with a combination of the power block Private Branch Exchange (PBX}, sound powered phones, satellite phones, and hand-held radios. These items will be powered and remained powered using the same methods as evaluated under EA-12-049 for the period of sustained operation.

The analyses and supporting information described above were provided to the NRC in the eportal.

Phae 2 ISE 01 7 MNGP has two types of flowmeters The NRC staff reviewed the Closed available for use (one in each FLEX information provided in the 6-Licensee to demonstrate the storage location). These are a Siemens month updates and on the

[Staff evaluation to be SAWM flow instrumentation Sitrans F M MAG 8000, product number ePortal.

included in SE Section qualification for the expected 7ME681-4.4.1.3]

environmental conditions.

4BJ31-2AA1 and Flow Technologies Inc.

The licensee provided FTI EL2200-125, with MC608B environmental conditions for electronics.

radiation and temperature as well Each flowmeter has 5" hose adapters as the qualified temperature which facilitate installation in-line on the 5" range for the flow instrument.

pump discharge hose. Plant procedures provide the deployment instructions for The NRC staff found the the portable diesel pump, hoses, and instrument appears to be qualified flowmeter. As described in the procedure, for the anticipated conditions the flowmeter is installed in the common during an ELAP for the proposed 5" discharge hose, between the final two Turbine Building East elevation sections of hose just before reaching the 931' location.

FLEX valve (RHRSW-68) (Turbine Building, east side, 931' elevation).

No follow-up questions.

The deployment location for the flowmeter is inside the Turbine Building, east side, 931' elevation. In this location, the Turbine Building provides environmental protection from external events, and substantial radiation shielding from the HCVS vent line. Dose calculations determined that peak severe accident dose rate in this area is 0.186 R/hr with a 7-day integrated dose of 15.5 R. This radiation level is not expected to have any adverse effect on operation of the flowmeter. The peak dose rate associated with the transit path to the area is approximately 5 R/hr. Since the transit times to the area are short, ingress and egress are not expected to be impacted.

The selected instruments are designed for the expected flow rate, temperature and pressure for SAWA over the period of sustained operation.

~- r""' w.,...-nt~U11,.,,.,,,....,.,

upectoo~-*~

Siem""' F M Mag 1IQOO fl1E!.221(M2S P-,o,.,..1trRong,,

5C-2000 GPM

,e. 19'.4 GPM

57. 2S5 GPIV 176~ GPt,,i

.J.'F to+ 140'F 40'; to *176'F

+()' F rnir-imw'fl No :rrnXH'ah..1rn specified for E.LAP e\\1ent

< 50= class,i:.~.../Sl 1ti:,

150# c!ass ANb! <c :>

Ota <t50PS!G flange rating flange ratr'."lg The analyses and supporting information described above were provided to the NRC in the eportal.

ML18130A921 OFFICE NRR/DLP/PBEB/PM NRR/DLP/PBMB/LA NAME RAuluck Slent DATE 5/14/18 5/11/18 RidsRgn3MailCenter Resource BTitus, NRR RAuluck, NRR Blee, NRR NRR/DLP/PBEB/BC(A)

NRR/DLP/PBEB/PM BTitus RAuluck 5/14/18 5/14/18