ML18094A804

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

ML18094A804
Person / Time
Site:	Oyster Creek
Issue date:	04/10/2018
From:	Rajender Auluck Beyond-Design-Basis Engineering Branch
To:	Church C Northern States Power Co
Lee B
References
CAC MF4376, EA-13-109, EPID L-2014-JLD-0052
Download: ML18094A804 (34)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 10, 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- 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:

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 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, 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. The NRC staff reviewed the information provided by the licensee and issued interim staff evaluations (IS Es) for Phase 1 and Phase 2 of Order EA-13-109 for Monticello by letters dated April 2, 2015 (ADAMS Accession No. ML15082A167), and September 6, 2016 (ADAMS Accession No. ML16244A120), 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.

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

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

Sincerely, Rajender Auluck, Senior Project Manager Beyond-Design-Basis Engineering Branch Division of Licensing Projects Office of Nuclear Reactor Regulation Docket No. 50-263

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 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 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 drywall 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 drywall 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, 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 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 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 NRC staff's review of these open items may change if the licensee changes its plans as part of final implementation. Changes in the NRC staff review will be communicated in the ongoing audit process.

Attachments:

1. Table 1 - NRC Staff Audit and Teleconference Participants

2. Table 2 -Audit Documents Reviewed

3. Table 3- ISE Open Item Status Table

Table 1 - NRC Staff Audit and Teleconference Participants Title Team Member Organization T earn Lead/Sr. Project Manager Rajender Auluck NRR/DLP Project Manager Support/Technical Support - Containment I 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 Attachment 1

Table 2 - Audit Documents Reviewed Calculation 16-006, "Hard Pipe Vent D8 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 1OB 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 O 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 LT-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 O 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 Attachment 2

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, "3rd 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. month updates and on the [Staff evaluation to be audit the final sizing evaluation The 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 operation of HCVS components is that the minimum calculated terminal provided by the HCVS 125 VDC voltage at the devices is above the battery and battery charger.

minimum voltage required for each HCVS device while being supplied from the The battery sizing calculation 16-battery. 006, "Hard Pipe Vent D8 Battery HCVS 125VDC Battery The design allows for use of the Diverse Calculation," Revision 1 and Flexible Coping Strategies (FLEX) confirmed that the 125 VDC equipment (i.e. FLEX generator) to power battery has a minimum capacity the system after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The design capable of providing power for 24 incorporates a manual, break-before- hours without recharging, and make transfer switch to transfer the load therefore is adequate.

from the normal HCVS power supply to 250VDC [volts direct current] battery The licensee provided number 16. During an ELAP event, the Engineering Change (EC) 23964 16 battery, through its associated battery - FLEX 480 V Diesel Generator charger, will be connected to and Sizing, which discusses re-powered from the FLEX portable diesel powering of the HCVS 125 VDC generator per procedure. battery charger using the FLEX DG.

An engineering evaluation was performed to demonstrate that the FLEX 480 V No follow-up questions.

Attachment 3

Diesel Generator is of adequate size to support these loads. The evaluation 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 108 and the primary containment isolation valves Calculation 16-011, "Calculation over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 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 design of the new HCVS nitrogen the licensee determined that 2 system is provided in Figure 01 2-1 of the bottles of nitrogen will be needed Sixth 6-Month Status Update submittal. for 8 air operated valves (AOV) actuations for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. An The calculation and drawings for the new additional minimum of 12 nitrogen nitrogen systems were provided to the bottles will be needed for 6 days NRC on the eportal. after the initial 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for more AOV actuations for the HCVS.

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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. By 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, 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 HPV 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 <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> after ELAP. requested clarification that the Additionally, transit paths are acceptable high temperature in the POS for short durations after ventinQ 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 light 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

[degrees Farenheit] at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. By hour (Procedure C.5-4503) instructs 12, supplemental ventilation will be operators to use portable heaters installed per procedure and room as needed to maintain the temperature will then be maintained temperature above 40°F.

below 120°F for the duration of the 7 day period. Room temperature in the winter The licensee concluded the will drop to 35°F after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0°F at summer temperature at the the end of 7 days with no mitigating remote operating station (ROS) actions taken. Procedures direct are not a concern since there are operators to add portable heaters as no heat loads. There is no needed within 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> upon initiation of equipment adversely affected by an ELAP to maintain EFT building third cold temperatures. The ROS is floor temperatures above 40°F. not continuously occupied.

Operators can perform required Temperature in the Turbine Building 931' actions independent of the local east side corridor (near the ROS) in the ROS temperature.

winter will drop to 29°F after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0°F at the end of 7 days with no mitigating Calculation 16-054, "MNGP actions taken. HCVS equipment in this HCVS Radiological Assessment,"

area can perform its function in these low Revision O was performed to temperature conditions and therefore is determine the integrated radiation acceptable. Summer peak temperatures dose due to HCVS operation.

in this area are not a concern due to a The NRC staff reviewed this lack of heat loads in the area during an calculation and determined that ELAP. the licensee used conservative assumptions and followed the quidance outlined in NEI 13-02

The different pathways between the Rev.1 and HCVS-WP-02 Rev.O.

Reactor Building, EFT Building, and Based on the expected integrated Turbine Building were analyzed and it whole body dose equivalent in the was determined that there are no POS and ROS and the expected substantial heat sources in these areas integrated whole body dose that would cause a significant change in equivalent for expected actions temperature. during the sustained operating period, the NRG staff believes The analyses and supporting information that the order requirements are described has been provided to the NRG met.

in the eportal.

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 NRG staff reviewed the Closed confirms that the modified HCVS information provided in the 6-Make available for NRG 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 per square in containment pressure is The calculation has been provided to the gauge (psig) (PCPL = 62 psig).

restored and then maintained NRG on the eportal. The steady state venting capacity below the primary containment at a torus pressure of 47.9 psig design pressure and the (maximum design pressure in the primary containment pressure drywell and the differential limit. pressure 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-uo auestions.

Phase 1 ISE 01 5 HCVS piping outside the Class I structure The NRG staff reviewed the Closed is designed for tornado/wind loads without information provided in the 6-Make available for NRG 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 [updated safety analysis report], 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 quidance document

HCVS-WP-04 which provides reasoning why protecting the HCVS 30 ft above grade 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 NRG on the eportal.

Phase 1 ISE 01 6 The POS for the HCVS is on the third The NRG staff reviewed the Closed floor of the EFT building and includes the information provided in the 6-Make available for NRG 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 staff's 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 The remote operating station is suppression pool level transmitter on the 931' elevation of the LT7338B is located in the torus room bay Turbine Building. Temperature

9. for these areas evaluated in calc

16-055. The calculation assumed The radiation detector is installed a 95°F outdoor temperature. The adjacent to the pipe above the high calculation determined the ETF pressure coolant injection (HPCI) room at Bldg, 3rd floor peaks at -135°F elevation 935'. The temperature element shortly after start of the event and is installed in the HPCI room adjacent to drops to approximately 100°F the vent pipe at elevation 928'. after mitigating actions are implemented. The temperature The drywell pressure transmitter varies between 110°F and 100°F PT7251 B is located in the Reactor with the daily diurnal temperature Building, elevation 985' south wall. variation.

Radiological Conditions: The main control room was previously evaluated as part of Radiological dose rates resulting from Order EA-12-049.

HCVS venting were determined by calculation for each area using the No follow up questions.

methodology in NEl-13-02, Rev 1 and HCVS-WP-02, Rev 0.

Temperature/ Humidity Conditions:

Temperature conditions for each area have been determined by calculation, using the methodology in NEl-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 NRG in the eportal.

Phase 1 ISE 01 7 The HCVS controls are located on the The NRG staff reviewed the Closed ASDS panel located on the third floor of information provided in the 6-

Make available tor NRG 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 The communication methods are remote HCVS operation temperature, and HCVS radiation the same as accepted in Order locations and HCVS decision indicators are on the panel adjoining the EA-12-049.

makers during ELAP and ASDS panel. These are the indicators severe accident conditions. used by the Operator to monitor the No follow-up questions.

primary containment and HCVS when 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-1 E 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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 eportal.

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 H and 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 system 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 the HCVS instruments and

• Pneumatic Supply Pressure Indications controls (l&C), including a and Manual Valves description of each component and the qualification method. The Qualification Methods: staff's review indicates that the l&C components are consistent The OIP provides the following with the guidance in NEI 13-02 information related to component and its qualifications meet the qualification: order requirements.

"The HCVS instruments, including valve No follow-up questions.

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 provided to the NRG in the eportal.

Phase 1 ISE 01 11 A calculation was performed that The NRG staff reviewed the Closed determined that the HCVS primary information provided in the 6-Make available for NRG staff containment isolation valves, A0-4539 month updates and on the [Staff evaluation to be audit documentation of an and A0-4540, will open under the ePortal. included in SE Section evaluation verifying the maximum differential pressure expected 3.2.1]

existing containment isolation during Beyond Design Basis External The NRG staff reviewed valves, relied upon for the Event (BDBEE) suppression pool venting calculation 03-088, "AOV HCVS, will open under the with greater than 20% margin. The Component Calculation, Hard maximum expected differential valves have been shown to open against Pipe Vent Valves, A0-4539 and pressure during BDBEE and a maximum expected differential pressure A0-4540," which discusses the severe accident wetwell of 76.7 psid [per square inch differential]. valve/actuator information for the venting. PCIVs.

The calculation has been provided to the NRG 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 NRG 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 NRG 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 [gallons per flow rate for the reference plant minute] for remainder of the mission prorated for the difference time). Guidance is provided for between the reactor thermal determining plant specific flow rates power level and the licensed based on scaling, using the ratio of the reactor thermal power for specific plant thermal power to the Monticello.

reference plant thermal power.

No follow-up questions.

Additional basis for determining the 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 giant 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 SAW A/SAWM [severe accident water management] values of 305/61 gpm.

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

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

sustained operating period.

. Ll-73388 (LT-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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, at which time initiate and operate the HCVS they are expected to be installed and during an ELAP with severe readvfor use to support SAWA/SAWM. accident conditions.

The following additional equipment The NRG 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 The environmental (temperature) NEI 13-02 Rev.1 and HCVS-WP-capability of the flow instrument has been 02 Rev.a. Based on the documented in ISE Open item 7. expected integrated whole body dose equivalent in the MGR and The deployment location for the flowmeter ROS and the expected integrated is inside the Turbine Building, east side, whole body dose equivalent for 931' elevation. In this location, the expected actions during the Turbine Building provides environmental sustained operating period, the protection from external events, and NRG staff believes that substantial radiation shielding from the radiological conditions should not HCVS vent line. Dose calculations inhibit operator actions or SAWA performed determine that peak severe equipment and controls needed to accident dose rate in this area is 0.186 initiate and operate the HCVS R/hr with a 7-day integrated dose of 15.5 during an ELAP with severe R. This radiation level is not expected to accident conditions.

have any adverse effect on operation of the flowmeter. The temperature evaluation addressed in Phase 1 Open Item SAWA/SAWM pump (Flex Pump) #6 bounds the SAWA/SAWM operation. For operation of The deployment and staging for the equipment located outdoors, portable diesel pump is the same as existing plant work controls FLEX strategies. The deployment routes remain applicable.

and environmental operating conditions (temperature) have previously been No follow-up questions.

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)

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 expected 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 radioloqical 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-73388 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-73388 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 The NRC staff reviewed generator) calculation 16-054, "MNGP HCVS Radiological Assessment," and This equipment is capable of performing determined that the licensee used during the sustained operating period in conservative assumptions and the expected environmental conditions. followed the guidance outlined in NEI 13-02 Rev.1 and HCVS-WP-One additional active component requires 02 Rev.a. Based on the review, M0-2014 Residual Heat Removal expected integrated whole body (RHR) Division 1 LPCI Inboard Injection dose equivalent in the MCR and Valve. This valve would be electrically ROS and the expected integrated opened from the Main Control Room in whole body dose equivalent for order to establish the reactor pressure expected actions during the valve (RPV) injection path. The valve is sustained operating period, the located in the Reactor Building, 931' NRC staff believes that the order elevation, East Shutdown Cooling Room. requirements are met.

The motor operated valve would be cycled within the first eight hours of the No follow-up questions.

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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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 <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />. Based on this, there is high confidence the valve can be electrically oQ_ened in the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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 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 1Rem = 1Rad 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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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 BWROG-TP-15-008 vent during severe accident has 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 suppression chamber through the on water level instrument is 925 feet). suppression pool.

The analysis is conservative since no mass loss through the HPV is credited. BWROG-TP-011 demonstrates Based on this analysis, the wetwell vent that starting water addition at a capability is maintained for a 7- day high rate of flow and throttling mission time. after approximately 4-hours will not increase the suppression pool The wetwell vent has been designed and level to that which could block the installed to meet NEI 13-02 Rev 1 suppression chamber HCVS.

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

BWROG-TP- 15-011. The wetwell vent No follow-up questions.

will be opened prior to exceeding the 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 NRG 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. 4.2.1.11

ML18094A804 OFFICE NRR/DLP/PBEB/PM NRR/DLP/PBMB/LA NRR/DLP/PBEB/BC (A) NRR/DLP/PBEB/PM NAME RAuluck Slent TBrown RAuluck DATE 4/9/18 4/5/18 4/10/18 4/10/18