Text

Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)
	ML15069A232
Person / Time
Site:	Millstone
Issue date:	03/02/2015
From:	Mark D. Sartain; Dominion Nuclear Connecticut
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	14-393C, EA-12-049
	Download: ML15069A232 (82)
	v • d • e

Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 m inion Web Address: www.dom.com March 2, 2015 10 CFR 2.202 EA-12-049 Attention: Document Control Desk Serial No.: 14-393C U.S. Nuclear Regulatory Commission NL&OS/MAE: RO Washington, D.C. 20555-0001 Docket No.: 50-423 License No.: NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 SIX-MONTH STATUS REPORT IN RESPONSE TO MARCH 12, 2012 COMMISSION ORDER MODIFYING LICENSES WITH REGARD TO REQUIREMENTS FOR MITIGATION STRATEGIES FOR BEYOND-DESIGN-BASIS EXTERNAL EVENTS (ORDER NUMBER EA-12-049)

References:

1. NRC Order Number EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, dated March 12, 2012

2. Dominion Nuclear Connecticut, Inc.'s Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049), dated February 28, 2013 (Serial No. 12-1611B)

3. Dominion Nuclear Connecticut, Inc's Six Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events-(Order Number EA-12-049), dated August 28, 2014 (Serial No. 14-393A).

On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued an -order (Reference 1) to Dominion Nuclear Connecticut, Inc. (DNC). Reference .1 was immediately effective and directed DNC to develop, implement, and maintain guidance and strategies to maintain core cooling, containment, and spent fuel pool cooling capabilities in the event of a beyond-design-basis external event.

Reference 1 required submission of an Overall Integrated Plan (OIP) (Reference 2) pursuant to Section IV, Condition C. Reference 1 also required submission of a status report at six-month intervals following submittal of the OIP.

The Attachments to this letter provide: 1) the fourth six-month status report and an update of milestone accomplishments since the submittal of the previous six-month

Serial No. 14-393C Docket Nos. 50-423 Order EA-12-049 Page 2 of 3 status report (Reference 3), including any changes to the compliance method, schedule, or need for relief and the basis, and 2) the responses to the Open Item and Confirmatory Items from the Interim Staff Evaluation for EA-12-049 (ML13338A445) from the NRC, DNC's responses to the items identified in Attachment 4 of the Millstone Power Station Onsite Audit Report dated November 17, 2014 (ML14275A017), plus DNC's responses to additional items requested by the Millstone Power Station NRC Mitigating Strategies Project Manager.

If you have any questions, please contact Ms. Margaret Earle at (804) 273-2768.

Sincerely, Mark D. Sartain Vice President - Nuclear Engineering Attachments (2)

Commitments made by this letter: No new Regulatory Commitments COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain who is Vice President Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of the Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this O- day of ,rz 2015.

My Commission Expires: "rL*'sc/r 31, 2,016 (SEAL)

Notary Public Commonwealth of Virginia Reg.Expires My ommission # 7518653 December 31, 20,&

Serial No. 14-393C Docket Nos. 50-423 Order EA-12-049 Page 3 of 3 cc: Director of Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 13H16M 11555 Rockville Pike Rockville, MD 20852-2738 U. S. Nuclear Regulatory Commission, Region I Regional Administrator 2100 Renaissance Blvd.

Suite 100 King of Prussia, PA 19406-2713 Mr. M. C. Thadani NRC Senior Project Manager Millstone Units 2 and 3 U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 B1 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station

a Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Six-Month Status Report for the Implementation of Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events February 2015 Millstone Power Station Unit 3 Dominion Nuclear Connecticut, Inc. (DNC)

4 Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 1 of 18 Six-Month Status Report for the Implementation of Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events 1 Introduction Dominion Nuclear Connecticut (DNC) developed an Overall Integrated Plan (OIP)

(Reference 1), documenting the diverse and flexible strategies (FLEX) for Millstone Power Station Unit 3 (MPS3), in response to Nuclear Regulatory Commission (NRC)

Order Number EA-12-049 (the Order) (Reference 2). This attachment provides an update of milestone accomplishments and open items since the last status report (Reference 25), including changes to the compliance method, schedule, or need for relief/relaxation and the basis, if any.

2 Milestone Accomplishments The following milestones have been completed since the development of the OIP, and

.are current as of January 31, 2015.

Submit Integrated Plan

" Develop Strategies

" Develop Modifications

" Implement Modifications

Develop Training Plan

Issue FSGs and Associated Procedure Revisions

" Develop Strategies/Contract with NSRC

" Purchase Equipment

Receive Equipment

Validation Walk-throughs or Demonstrations of FLEX Strategies and Procedures

Create Maintenance Procedures

" Outage Implementation 3 Milestone Schedule Status The following table provides an update to Attachment 2A of the OIP. It provides the activity status of each item as of January 31, 2015, and whether the expected completion date has changed. The completion date for 'Implement Training' is discussed in Section 5.

Target Activity Revised Target Milestone Completion Status Completion Date Date Submit Integrated Plan February 2013 Complete Develop Strategies April 2014 Complete

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 2 of 18 Target Activity Revised Target Milestone Completion Status Completion Date

__________________________ Date Develop Modifications July 2014 Complete Implement Modifications November 2014 Complete Develop Training Plan April 2014 Complete Implement Training April 2015 Started Issue FSGs and Associated October 2014 Complete Procedure Revisions Develop Strategies/Contract with August 2014 Complete NSRC*

Purchase Equipment February 2014 Complete Receive Equipment August 2014 Complete Validation Walk-throughs or Demonstrations of FLEX Strategies August 2014 Complete and Procedures Create Maintenance Procedures August 2014 Complete Outage Implementation November 2014 Complete

NSRC is the National SAFER Response Center 4 Changes to Compliance Method By letter dated February 28, 2013, (Reference 1), DNC provided an OIP to address Beyond-Design-Basis (BDB) events at Millstone Power Station Unit 2 (MPS2) and MPS3 as required by Order Number EA-12-049, dated March 12, 2012 (Reference 2).

The first Six-Month Status Update of the OIP for both MPS2 and MPS3 was provided by letter dated August 23, 2013 (Reference 14). The second Six-Month Status Update for MPS3 was provided by letter dated February 28, 2014 (Reference 16). The third Six-Month Status Update for MPS3 was provided by letter dated August 28, 2014 (Reference 25). There are no additional changes to the compliance method information provided in the MPS3 OIP and subsequent updates, therefore the MPS3 BDB FLEX strategy continues to meet Nuclear Energy Institute (NEI) 12-06 guidance (Reference 3).

a Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 3 of 18 5 Need for Relief/Relaxation and Basis for the Relief/Relaxation DNC has complied with the Order implementation date for all items except the completion of BDB operator training. A request to relax Condition A.2 of the Order to allow a delay of the full implementation date until April 30, 2015 was requested in a letter dated May 16, 2014 (Reference 18). The NRC granted approval of the relaxation of the order implementation date by letter dated July 3, 2014 (Reference 19). By letter dated September 22, 2014 (Reference 26) DNC notified the NRC that the final staffing assessment report would be submitted by July 30, 2015 in order to incorporate completion of BDB operator training as necessary.

6 Open Items The NRC has established an audit process to allow the exchange of information between the licensees and the NRC Staff (Reference 21). Between July 21, 2014 and July 25, 2014, MPS2 and MPS3 were the subject of an NRC onsite audit where the site specific aspects of DNC's proposed FLEX Mitigating Strategies were reviewed. During this NRC onsite audit, the staff reviewed site specific documentation and upon completion of the audit, issued an Audit Report "Millstone Power Station, Units 1 and 2

- Report for the Onsite Audit Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Level Instrumentation Related to Orders EA-12-049 and EA-12-051," dated November 17, 2014 (Reference 28). The report indicated that further review of several items was not anticipated as DNC proceeds towards compliance for Orders EA-12-049 and EA-12-051. A status of these items is provided in the following tables in Section 6.

DNC's responses to the Interim Staff Evaluation (ISE) items, plus additional Audit Questions, Licensee Identified Open Items (01) from the OIP, and Safety Evaluation (SE) Review items identified in Attachment 4 of the Audit Report are provided in Attachment 2. Attachment 2 also includes DNC's responses to additional items requested by the Millstone Power Station (MPS) NRC Mitigating Strategies Project Manager.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 4 of 18 6.1. Open Items from Overall Integrated Plan The following table provides a summary of the status of Ols identified by DNC and documented in Attachment 2B of the MPS3 OIP submitted on February 28, 2013 and the status of each item.

Overall Integrated Plan Open Items 01 # Description Status Complete 1 Verify response times listed in timeline (Reference 27) and perform staffing assessment. Response provided in Attachment 2 Complete In the DNC February 28, 2013 OIP submittal for Millstone, the MPS3 Class 1E 125V battery life was "estimated" at between 2 and 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months assuming that load stripping would commence within 45 minutes and be completed within the following 30 minutes. The pending final evaluation of the extended Class 1E station emergency battery life was identified as Open Item No. 2. The calculation with the stated load stripping assumption has been completed. The Evaluation of extended battery life with evaluated time to battery depletion is 2 14.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . The evaluation considered load stripping of all non-essential loads.

that the 301 B battery was stripped of all loads by the assumed time frame and that the 301A battery would carry the necessary instrumentation loads until it reached the minimum voltage for reliable instrument readings in 7.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . At that time, alternate instrument loads for plant monitoring from the 301 B battery would be re-connected and would be available for an additional 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . The combined time available by this approach (14.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ) is sufficient for the implementation of the re-powering strategy for the 120VAC systems as outlined in the OIP submittal, Section

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 5 of 18 Overall Integrated Plan Open Items 01 # Description Status F1.2. (Reference 5)

In the Final Audit report from the July 2014 NRC Onsite Audit, the NRC Staff indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

Preliminary analyses have been Complete performed to determine the time to (Reference 4) steam generator (SG) overfill without operator action to reduce auxiliary In the Final Audit report from the July feedwater (AFW) flow, time to SG 2014 NRC Onsite Audit, the NRC Staff dryout without AFW flow, and time to indicated that further review of this item depletion of the useable volume of the was not anticipated as DNC proceeds demineralized water storage tank towards compliance for Order EA-12-049 (DWST). The final durations will be (Reference 28).

provided when the analyses are completed.

The Phase 3 coping strategy to Complete maintain Containment integrity is under (Reference 16) development. Methods to monitor and evaluate Containment conditions and In the Final Audit report from the July 4 depressurize/cool Containment, if 2014 NRC Onsite Audit, the NRC Staff necessary, will be provided in a future indicated that further review of this item update. was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

Complete The hydraulic calculation performed with Analyses will be performed to develop the FLEX pumps deployed and utilizing fluid components performance their associated hose networks has 5 requirements and confirm fluid confirmed that the primary and alternate hydraulic-related strategy objectives can connections for core cooling/decay heat be met. removal, Reactor Coolant System (RCS)

Inventory, and reactivity control (RCS Injection), and spent fuel pool (SFP) make-up strategies can be satisfactorily

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 6 of 18 Overall Integrated Plan Open Items 01 # Description Status accomplished in response to an Extended Loss of Alternating Current (AC) Power (ELAP)/Loss of Ultimate Heat Sink (LUHS) event. (Reference 8)

Hydraulic calculations have confirmed that the SW flows for the Containment cooling options are adequate.

(Reference 10)

In the Final Audit report from the July 2014 NRC Onsite Audit, the NRC Staff indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

A study is in progress to determine the Complete design features, site location(s), and (References 13 and 17) number of BDB Storage Building(s).

The final design for BDB Storage Building(s) will be based on the A single 10,000 sq. ft. Type 1 building guidance contained in NEI 12-06, has been constructed at Millstone Power Section 11.3, Equipment Storage. A Station for storage of BDB equipment.

supplement to this submittal will be The building is designed to meet the provided equipmentwith the results study.of the plant's design basis for the storage Shutdown Earthquake, high Safe wind hazards, snow, ice and cold conditions, and is located above the flood elevation 6 from the most recent site flood analysis.

The BDB Storage Building is sited south of the railroad bridge, on the west side of the MPS access road, adjacent to the existing northeast contractor parking lot.

In the Final Audit report from the July 2014 NRC Onsite Audit, the NRC Staff indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 1 Page 7 of 18 Overall Integrated Plan Open Items 01 # Description Status FLEX Support Guidelines (FSGs) will be Complete developed in accordance with PWROG Response provided in Attachment 2 guidance. Existing procedures will be revised as necessary to implement FSGs.

Electrical Power Research Institute Complete (EPRI) guidance documents will be used to develop periodic testing and EPRI guidance documents have been preventative maintenance procedures used, where available, to develop the for BDB equipment. Procedures will be testing and preventative maintenance developed to manage unavailability of strategies for the sites. Fleet-wide equipment such that risk to mitigating templates have been developed and strategy capability is minimized, input into the individual site maintenance strategies. Specific Preventative Maintenance (PM) work orders based on these strategies will be implemented prior to the required FLEX equipment maintenance due date.*

A fleet-wide FLEX Strategy Program 8 Document has been developed (Refer to Open Item 9). The program includes the requirement to manage unavailability of equipment such that risk to mitigating strategy capability is minimized. A fleet-wide procedure has been developed to specifically address equipment unavailability. (Reference 22)

In the Final Audit report from the July 2014 NRC Onsite Audit, the NRC Staff indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

An overall program document will be Complete developed to maintain the FLEX (Reference 23) 9 strategies and their bases, and provide configuration control and change management for the FLEX Program. In the Final Audit report from the July 2014 NRC Onsite Audit, the NRC Staff

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 8 of 18 Overall Integrated Plan Open Items 01 # Description Status indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

The DNC Nuclear Training Program will Complete be revised to assure personnel (Reference 21) proficiency in the mitigation of BDB events is developed and maintained.

10 These programs and controls will be In the Final Audit report from the July developed and implemented in 2014 NRC Onsite Audit, the NRC Staff accordance with the Systematic indicated that further review of this item Approach to Training (SAT). was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

Complete the evaluation of turbine Complete driven (TD) auxiliary feedwater (AFW) (References 6 and 7) 1 pump long term operation with < 290 psig inlet steam pressure. Response provided in Attachment 2 Plant modifications will be completed for Complete permanent plant changes required for implementation of FLEX strategies. Plant modifications required to support 12 FLEX strategies have been completed for permanent plant changes required for implementation of FLEX strategies.

Analyses will be performed to develop Complete electrical components performance requirements and confirm electrical Calculations have been completed for loading-related strategy objectives can the sizing and loading analysis of the be met. 120 VAC, 480 VAC, and 4160 VAC generators and confirm the electrical loading-related strategy objectives can 13 be met. (Reference 11)

In the Final Audit report from the July 2014 NRC Onsite Audit, the NRC Staff indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049

_ (Reference28).

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 9 of 18 Overall Integrated Plan Open Items 01 # Description Status An evaluation of all BDB equipment fuel Complete consumption and required re-fill strategies will be developed. An evaluation of all BDB equipment fuel consumption and required refill strategies has been completed and provided as part of the ongoing NRC audit process. (Reference 17) 14 In the Final Audit report from the July 2014 NRC Onsite Audit, the NRC Staff indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

A lighting study will be performed to Complete validate the adequacy of supplemental lighting and the adequacy and A lighting study has been completed practicality of using portable lighting to validating the adequacy of supplemental perform FLEX strategy actions. lighting and the adequacy and practicality of using portable lighting to perform FLEX Strategy actions. This was provided as part of the ongoing 15 NRC audit process. (Reference 17)

In the Final Audit report from the July 2014 NRC Onsite Audit, the NRC Staff indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

A comprehensive study of Complete communication capabilities is being performed in accordance with the A study documenting the commitments made in DNC letter S/N communications strategy has been 12-205F dated October 29, 2012 in completed. The plan concludes that 16 response to Recommendation 9.3 of the FLEX strategies can be effectively 10 CFR 50.54(f) letter dated March 12, implemented with a combination of 2012. The results of this study will satellite phones, hand-held radios and identify the communication means sound powered phones. (Reference 9) available or needed to implement command and control of the FLEX In the Final Audit report from the July strategies at Millstone. Validation of 2014 NRC Onsite Audit, the NRC Staff

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 10 of 18 Overall Intearated Plan Open Items O # Description Status communications required to implement indicated that further review of this item FLEX strategies will be performed as was not anticipated as DNC proceeds part of Open Item No. 1. towards compliance for Order EA-12-049 (Reference 28).

Details of the ventilation strategy are Complete under development and will conform to the guidance given in NEI 12-06. The DNC's Six Month Status Report in details of this strategy will be provided Response to March 12, 2012 at a later date. Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)," dated August 28, 2014 17 (Reference 25), Attachment 2, Provided OIP Section F5 - Safety Functions Support (Ventilation).

In the Final Audit report from the July 2014 NRC Onsite Audit, the NRC Staff indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

Preferred travel pathways will be Complete determined using the guidance (References 12 and 13) contained in NEI 12-06. The pathways 18 will attempt to avoid areas with trees, Response provided in Attachment 2 power lines, and other potential obstructions and will consider the potential for soil liquefaction.

The equipment listed in Table 1 will be Complete received on site.

The equipment listed in Table 1 has been received on site.

19 In the Final Audit report from the July 2014 NRC Onsite Audit, the NRC Staff indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

Revised for consistency with Audit Question #84 response.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 11 of 18 6.2. Open Items from Interim Staff Evaluation (ISE)

The response to the 01 from the ISE for MPS3 (Reference 15) is provided in Attachment 2.

6.3. Confirmatory Items from Interim Staff Evaluation (ISE)

Responses to the Confirmatory Items (CI) from the ISE for MPS3 (Reference 15) are provided in Attachment 2.

6.4. Audit Questions Reviewed During the MPS Unit 3 NRC Onsite Audit Various MPS3 Audit Questions (AQ) were evaluated during the Millstone NRC Onsite Audit. Responses to the AQs identified in Attachment 4 of the MPS Onsite Audit Report and AQs identified by the MPS NRC Mitigating Strategies Project Manager are provided in Attachment 2.

6.5. Additional Items Reviewed During the MPS Unit 3 NRC Onsite Audit The following table provides a list of the additional Safety Evaluation (SE) Review items identified and evaluated during the MPS NRC Onsite Audit and the status of each item.

Responses to the SEs identified in Attachment 4 of the MPS Onsite Audit Report are provided in Attachment 2.

Safety Evaluation Review Items SE # Description Status Safety WCAP-17792-P - Provide a detailed discussion on the Complete Evaluation applicability to MPS3 of the recommendations in WCAP- (References 17 and 21)

Review 17792 to vent the RCS while makeup is being provided for Item #1 the mitigating strategies involving RCS makeup and boration. This discussion should include if the MPS3 In the Final Audit report strategy includes venting the RCS, methods of venting, vent from the July 2014 NRC operations criteria, related fluid dynamic analysis, involving Onsite Audit, the NRC instrumentation, and related parameter thresholds. Staff indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 12 of 18 Safety Evaluation Review Items SE # Description Status Safety NSAL-14 On February 10, 2014, Westinghouse issued Complete Evaluation Nuclear Safety Advisory Letter (NSAL)-14-1, informing (Reference 21)

Review licensees of plants with Standard Westinghouse RCP seals Item #2 that 21 gpm may not be a conservative leakage rate for ELAP analysis. This value had been previously used in the Response provided in ELAP analysis referenced for the use in many Attachment 2 Westinghouse pressurized-water-reactors, including the generic reference analysis in WCAP-17601-P. Therefore, please clarify whether the assumption of 21 gpm of seal leakage per RCP (at 550 degrees F, 2250 psia) remains valid in light of the issues identified in NSAL-14-1. In so doing, please identify the specifics of the seal leak off line design in NSAL and #1 seal faceplate material relative to the categories in NSAL-14-1 and identify the corresponding assumed leakage rate from NSAL-14-1 that is deemed applicable.

Safety Time to reflux cooling - Please clarify whether procedural Complete Evaluation guidance for the timing of providing makeup to the reactor (References 21 and 24)

Review coolant system is based on analysis in WCAP-17792-P, Item #3 pages 3-10 through 3-16. If so, provided adequate justification for basing the timing of primary makeup on the In the Final Audit report assumption that reactor coolant pump seal leakage that are from the July 2014 NRC less than expected maximum expected value under ELAP Onsite Audit, the NRC conditions will not increase. Staff indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 13 of 18 Safety Evaluation Review Items SE # Description Status Safety Human factors questions addressed during walkdowns of Complete Evaluation plant equipment as well as in discussion with personnel. (Reference 21)

Review Item #4 In the Final Audit report from the July 2014 NRC Onsite Audit, the NRC Staff indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

Safety Please provide adequate basis that, when considering Complete Evaluation mixing time, there is sufficient flow capacity to support (Reference 21)

Review borated makeup to both units from a single RCS makeup Item #5 pump taking suction from a portable batching tank. In the Final Audit report from the July 2014 NRC Onsite Audit, the NRC Staff indicated that further review of this item was not anticipated as DNC proceeds towards compliance for Order EA-12-049 (Reference 28).

Safety Please provide adequate basis that calculations performed Complete Evaluation with the NOTRUMP code (e.g., those in WCAP-17601 -P, (References 21 and 24)

Review WCAP-17792-P) are adequate to demonstrate that criteria Item #6 associated with the analysis of an ELAP event (e.g.,

avoidance of reflux cooling, promotion of boric acid mixing) Response provided in are satisfied. NRC staff confirmatory analysis suggests that Attachment 2 the need for implementing certain mitigating strategies for providing core cooling and adequate shutdown margin may occur sooner than predicted in NOTRUMP simulations.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 14 of 18 Safety Evaluation Review Items SE # Description Status Safety Security Related Issues. Complete Evaluation (Reference 21)

Review Item #8 Response provided in Attachment 2 Safety Complete Evaluation Please provide adequate justification for the seal leakage rates Review calculated according to the Westinghouse seal leakage model Item #9 that was revised following the issuance of NSAL-14-1. The Response provided in justification should include a discussion of the following Attachment 2 factors:

a. benchmarking of the seal leakage model against relevant data from tests or operating events,

b. discussion of the impact on the seal leakage rate due to fluid temperatures greater than 550°F resulting in increased deflection at the seal interface,

c. clarification whether the second-stage reactor coolant pump seal would remain closed under ELAP conditions predicted by the revised seal leakage model and a technical basis to support the determination, and,

d. justification that the interpolation scheme used to compute the integrated leakage from the reactor coolant pump seals from a limited number of computer simulations (e.g., three) is realistic or conservative.

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 1 Page 15 of 18 Safety Evaluation Review Items SE # Description Status Safety Complete Evaluation The NRC staff understands that Westinghouse has recently Review recalculated seal leakoff line pressures under loss of seal Item #10 cooling events based on a revised seal leakage model and Response provided in additional design-specific information for certain plants. Attachment 2

a. Please clarify whether the piping and all components (e.g.,

flow elements, flanges, valves, etc.) in your seal leakoff line are capable of withstanding the pressure predicted during an ELAP event according to the revised seal leakage model.

b. Please clarify whether operator actions are credited with isolating low-pressure portions of the seal leakoff line, and if so, please explain how these actions will be executed under ELAP conditions.

c. Ifoverpressurization of piping or components could occur under ELAP conditions, please discuss any planned modifications to the seal leakoff piping and component design and the associated completion timeline.

d. Alternately, please identify the seal leakoff piping or components that would be susceptible to overpressurization under ELAP conditions, clarify their locations, and provide justification that the seal leakage rate would remain in an acceptable range if the affected piping or components were to rupture.

7 Potential Safety Evaluation Impacts Section 6.5 provides a list of the additional SE Review items identified and evaluated during the MPS NRC Onsite Audit and the status of each item.

Additionally, DNC is participating in the ongoing industry effort to develop guidance for the Final Integrated Plan that will support NRC preparation of the Safety Evaluation documenting MPS3 compliance with Order EA-12-049. The format of the Final Integrated Plan is consistent with the Safety Evaluation Template provided with the July 1, 2014 Jack Davis memorandum (ML14161A643). (Reference 20)

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 16 of 18 8 References The following references support the updates to the OIP described in this attachment and are available in the Agency-wide Document Access and Management System (ADAMS) or have been provided to the staff for their review.

1. DNC's Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049), dated February 29, 2013 (Serial No. 12-161B).

2. NRC Order Number EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," dated March 12, 2012.

3. NEI 12-06, "Diverse and Flexible Coping Strategies (FLEX) Implementation Guide," Revision 0, dated August 2012.

4. Supplement to Overall Integrated Plan in Response to March 21, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis Events (Order Number EA 049), dated April 30, 2013 (Serial No. 12-161C).

5. Calculation 2013-ENG-04501E3, "MP3 BDB Battery Calculation," Rev. 0 dated May 29, 2013.

6. Calculation 97-014, "MP3 AFW System, Determination of AFW Turbine/Pump Speed and AFW System Flow for Steam Generator Pressures of 185 psig, 600 psig, and 125 psig, and Determination of the Turbine Exhaust Pressure," Rev.1 dated October 6, 2014.

7. Engineering Technical Evaluation ETE-MP-2013-1037, "MP3 Turbine Driven Aux Feedwater Pump Minimum Continuous Operating Speed," Rev. 0 dated March 12, 2013.

8. Calculation 13-015, "MP2 & MP3 FLEX Strategy Hydraulic Calculations," Rev. 2.

9. ETE-CPR-2013-0003, "Beyond Design Basis Communications Strategy/Plan,"

Rev. 2.

10. Calculation 13-015, "MP2 & MP3 FLEX Strategy Hydraulic Calculations" Rev. 3.

11. Calculation 2013-ENG-04503E3, "Millstone Power Station Unit 3 Beyond Design Basis - FLEX Electrical 4160V, 4840V and 120VAC System Loading Analysis,"

Rev. 2.

12. URS Geotechnical Investigation and Engineering Report, "FLEX Storage Building Project, Millstone Power Station, Waterford, Connecticut,' dated January 27, 2014.

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 1 Page 17 of 18 13.Design Change MPG-13-00010, 'BDB Storage Building/Millstone Power Station/Units 2&3."

14.DNC's Six Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049), dated August 23, 2013 (Serial No. 12-161D).

15."Millstone Power Station, Units 2 and 3 - Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigating Strategies),"

dated January 31, 2014. "

16.DNC's Six Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)," dated February 28, 2014 (Serial No. 12-161F).

17.Engineering Technical Evaluation ETE-CPR-12-0008, "Beyond Design Basis -

FLEX Strategy Overall Integrated Plan Basis Document." Rev. 4.

18. DNC Letter to NRC, "Millstone Power Station Unit 3 - Order Modifying License with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events Dated March 12, 2012 - Relaxation Request," dated May 16, 2014, Serial No.14-251.

19. NRC Letter to DNC, "Millstone Power Station, Unit 3 - Relaxation of Schedule Requirements for Order EA-12-049 'Issuance of Order to modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events'," dated July 3, 2014.

20. Memorandum from Jack R. Davis, JLD, Office of NRR, to Stewart N. Bailey, Sheena A, Whaley, and Jeremy S. Bowen, "Supplemental Staff Guidance for the Safety Evaluations for Order EA-12-049 on Mitigation Strategies for Beyond-Design-Basis External Events and Order EA-12-051 on Spent Fuel Pool Instrumentation," dated July 1,2014 (ML14161A643).

21. NRC letter from Jack R. Davis, Director Mitigating Strategies Directorate to All Operating Reactor Licensees and Holders of Construction Permits, "Nuclear Regulatory Commission Audits of Licensee Responses to Mitigating Strategies Order EA-12-049," dated August 28, 2013 (ML13234A503) 22.Procedure CM-AA-BDB-102, "Beyond Design Basis FLEX Equipment Unavailability Tracking," Rev. 1.

23. Procedure CM-AA-BDB-10, "Beyond Design Basis FLEX Program," Rev. 0.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 1 Page 18 of 18 24.Engineering Technical Evaluation ETE-CPR-2012-0150, "Core Cooling Evaluation for Dominion Fleet and Prepared Input for Response to Order EA 049," Rev. 2.

25. DNC's Six Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)," dated August 28, 2014 (Serial No. 14-393A).

26. DNC Letter to NRC, "Millstone Power Station Units 2 and 3 - Response to Request for Additional Information Regarding Phase 2 Staffing Assessment Report, Recommendation 9.3," dated September 22, 2014, Serial No.14-443.

27. Engineering Technical Evaluation ETE-CPR-2014-1008, "Millstone Power Station Unit 2 & 3 Beyond Design Basis FLEX Validation for Time Sensitive Actions (TSA's)," Rev. 1.

28. NRC letter from Stephen Monarque, Project Manager, JLD, Office of NRR, to David A. Heacock, President and Chief Nuclear Officer, Dominion Nuclear Connecticut, Inc., "Millstone Power Station, Units 2 and 3 - Report for the Onsite Audit Regarding Implementation of Mitigating Strategies and Reliable Spend Fuel Instrumentation Related to Orders EA-12-049 and EA-12-051," dated November 17, 2014.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Attachment 2 Responses to Mitigation Strategies Interim Staff Evaluation (ISE) Items, Audit Questions, Licensee Identified Open Items, and Safety Evaluation Review Items for Millstone Power Station Unit 3 Millstone Power Station Unit 3 Dominion Nuclear Connecticut, Inc. (DNC)

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 1 of 59 Response to Mitigation Strategies Interim Staff Evaluation (ISE) Items, Licensee Identified Open Items, Audit Questions, and Safety Evaluation Review Items for Millstone Power Station Unit 3 NOTE: Documents identified in this attachment as having been previously provided to the Nuclear Regulatory Commission (NRC) staff for their review were provided in accordance with NRC letter to All Operating Reactor Licensees and Holders of Construction Permits, "Online Reference Portal for Nuclear Regulatory Commission Review of Fukushima Near-Term Task Force Related Documents," dated August 1, 2013 (ML13206A427).

I. Interim Staff Evaluation (ISE) Open Item Response ISE 01 3.2.1.8.A Core Sub-Criticality - The PWROG submitted to NRC a position paper, dated August 15, 2013, which provides test data regarding boric acid mixing under single-phase natural circulation conditions and outlined applicability conditions intended to ensure that boric acid addition and mixing would occur under conditions similar to those for which boric acid mixing data is available. During the audit process, the licensee informed the NRC staff of its intent to abide by the generic approach discussed above.

The licensee should address the clarifications in the NRC endorsement letter dated January8, 2014.

DNC Response:

In response to industry requests for assistance in resolving the issues raised in Nuclear Safety Advisory Letter (NSAL) NSAL-14-1, Westinghouse prepared report PWROG-14015-P, Revision 0, "No. 1 Seal Flow Rate for Westinghouse Reactor Coolant Pumps Following Loss of All AC Power, Task 2, Determine Seal Flow Rates." Task 1 of the Pressurized Water Reactor Owners Group (PWROG) authorized project, PA-SEE-1 196, evaluated the various seal leakoff line design configurations for the Westinghouse fleet and "binned" the configurations into six different categories. In Task 2, leakage rates vs Reactor Coolant System (RCS) pressure were calculated based on assumptions which conservatively bound the leak rate (i.e., result in high leak rates for each of the six categories). Millstone Power Station Unit 3 (MPS3) falls into Category 1 as defined by Westinghouse based on the seal leakoff line flow orifice diameter, the piping lengths and diameters upstream and downstream of the flow orifice and additional components (valves, bends) in the line.

MPS3 is currently operating with two Westinghouse Seals and two Flowserve N-seals (installed during the fall 2014 refueling outage). The remaining two Westinghouse seals

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 2 of 59 are currently scheduled to be replaced with Flowserve N-seals during the spring 2016 outage.

As discussed in the Flowserve White Paper of February 11, 2014 and the subsequent responses to NRC review questions, the expected response of the Flowserve seal for plants which initiate an early RCS cooldown and depressurization (such as MPS3), is nominal controlled bleedoff (CBO) (i.e., flow proportional to the square root of the total pressure drop across the seals), and negligible third stage leakage for an extended period, (i.e., weeks). Based on this information, DNC has compared the expected leakage performance for the MPS3 fall 2014 configuration to the results of the reference analysis in WCAP-1 7601. The results were as follows.

For the condition of 2-phase natural circulation loop flow becoming < single phase flow:

a. The Reference analysis predicted this condition was reached in 75,000 seconds or 20.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (see Westinghouse letter LTR-LIS-14-79, Revision 0, February 12, 2014).

b. Using the assumed leakage response for Westinghouse seals and RCS pressure response documented in WCAP17601, DNC estimated the integrated RCS leakage at 75,000 seconds to be 230,260 Ibm based on 4 Westinghouse OEM seals (compare to WCAP-17601 Figure 5.2.2-6).

c. Using the projected leakage response for the actual leakoff line configuration documented in Report PWROG-14015-P and the expected Flowserve projected leakage, DNC calculated the time for which the MPS3 integrated leakage would reach 230,260 Ibm to be 34.0 hours0 days 0 hours 0 weeks 0 months with the fall 2014 seal configuration.

d. Since the latest time of deployment of FLEX RCS makeup for MPS3 is 16.0 hours0 days 0 hours 0 weeks 0 months per the FLEX strategy timeline, the condition of 2-phase natural circulation loop flow becoming < single phase flow is avoided with a time margin of 18.0 hours0 days 0 hours 0 weeks 0 months .

Calculations show that, for the target post cooldown steam pressure of 300 psig for MPS3, boron addition to maintain k-effective <0.99 is not needed before approximately 26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months . Since deployment of the FLEX RCS makeup pump occurs no later than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , and will maintain adequate conditions for RCS mixing, the available mixing time is 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

Given this background, each of the clarifications in the NRC Letter of January 8, 2014 are addressed as follows:

Clarification (1): The MPS3 evaluation for boron mixing has considered both the case of maximum RCP seal leakage, as well as the zero leakage case. Westinghouse Letter LTR-FSE-13-46, Rev. 0, "Westinghouse Response to NRC Generic Request for Additional Information (RAI) on Boron Mixing in Support of the Pressurized Water Reactor Owners Group (PWROG)," August 15, 2013 argued that the zero leakage case

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 3 of 59 is more limiting than the high leakage case from the standpoint of mixing because it delays the boron contribution from accumulator injection, resulting in more reliance on pumped injection from the RCS FLEX pump.

However, DNC has not credited boron from accumulator injection in developing its FLEX strategy. Maximizing the remaining RCS inventory at the time of FLEX makeup injection will maximize the required volume of makeup fluid to be added to meet the reactivity requirements. Therefore, zero leakage was assumed in developing the required volumes. Conversely, maximum leakage was considered in demonstrating that the criterion defined by the clarifications in the January 8, 2014 letter are met.

Clarification (2a): Condition 2(a) of the letter was addressed in the previous paragraphs which demonstrate the condition of 2-phase natural circulation loop flow becoming less than single phase flow is avoided with a time margin of 18.0 hours0 days 0 hours 0 weeks 0 months for the fall 2014 RCP seal configuration.

Clarification (2b) states: If loop flow during two-phase natural circulation has decreased below the single phase natural circulation flow rate, then the mixing of any borated primary makeup added to the reactor coolant system is not to be credited until one hour after the flow in all loops has been restored to a flow rate that is greater than or equal to the single-phase natural circulation flow rate. As discussed above, deployment of FLEX RCS makeup by 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months will ensure the condition of 2-phase loop flow less than single phase loop flow is avoided with significant time margin (18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months ). Furthermore, the reactivity analysis demonstrates that boron increases to main k-effective < 0.99 do not need to start until approximately 26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months . If for some reason RCS leakages exceed the projected bounding values for the fall 2014 RCP seal configuration, there is significant time (-10 hours) to recover inventory and restore single phase flow conditions and provide mixing before any boron increase is required.

Clarification (3) states: In all cases, credit for increases in the reactor coolant system boron concentration should be delayed to account for the mixing of the borated primary makeup with the reactor coolant system inventory. Provided that the flow in all loops is greater than or equal to the corresponding single-phase natural circulation flow rate, the staff considers a mixing delay period of one hour following the addition of the targeted quantity of boric acid to the reactor coolant system to be appropriate. The available mixing time of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months for MPS3 is well in excess of this requirement.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 4 of 59 II. Interim Staff Evaluation (ISE) Confirmatory Item Responses ISE CI 3.1.1.2.A The licensee stated that the haul path from the BDB Storage Building to the MPS3 equipment deployment locations and the building foundation design evaluations are proceeding for Millstone. Confirm that soil liquefaction is not a concern.

DNC Response:

The Millstone Haul Route Evaluation has been completed and incorporated into Section 19.2 of Engineering Technical Evaluations ETE-CPR-2012-0009, Revision 3 and ETE-CPR-2012-0008, Revision 3 for Millstone Power Station Unit 2 (MPS2) and (MPS3),

respectively. The evaluations conclude that, with one (1) exception, adequate on-site equipment is available to clear anticipated debris material/structures and that there are no liquefaction concerns for the designated haul routes. Although debris can generally be cleared or moved out of the way, there are alternate routes around potential obstacles.

Liquefaction has been evaluated for the Beyond-Design-Basis (BDB) Storage Building site and both haul routes from the BDB Storage Building to the station Protected Area (PA) and on the east side of the station for access to the barge slip at the south end of the station. These haul route segments outside of the PA have been evaluated for liquefaction based on geophysical data recently obtained for the Beyond Design Basis project. All areas inside the station PA have also been evaluated for liquefaction concerns. Results of the evaluations indicate there are no liquefaction concerns for any of the above areas.

For the areas inside the PA, a review of the construction specification documents addressing the site work performed during construction indicates that liquefaction will not be a concern with the backfill materials placed in accordance with these specifications. For areas inside the PA that may not have been excavated and backfilled during the construction of the original units, the MPS2 and MPS3 Final Safety Analysis Report (FSAR) documents address the naturally occurring, in-situ materials for the Millstone site. Both FSAR documents state that based on the properties of these materials, the in-situ, naturally occurring materials are not subject to liquefaction during a seismic event. (MPS2 FSAR Section 2.7.6 and MPS3 FSAR Section 2.5.4.8). The PA fence has not been moved out beyond the original fence line except to encompass the Independent Spent Fuel Storage Installation (ISFSI) area. This area is not included in any haul routes used for deployment of BDB equipment.

In summary, the evaluations performed conclude that there are no liquefaction concerns for the designated haul routes, and that adequate on-site equipment is available to clear anticipated debris material/structures.

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 5 of 59 The ETEs referenced in the above response have previously been provided to the NRC staff and are available for their review.

ISE CI 3.1.1.3.A The licensee stated that the review for internal flooding sources that could result from seismic induced failures and engine-driven or gravity-drain water sources has not been completed. Also MPS3 does not have a permanent safety-related groundwaterremoval system installed. However, the Engineered Safety Featuresbuilding does have a sump to control groundwater in-leakage. In ETE-CPR-2012-0008, Section 11.1.3.3, the licensee stated that they also have several small pumps and hoses on site for this purpose. Confirm that the impact of this in-leakage is limited, or can be addressed.

DNC Response:

Areas internal to the plant where BDB strategies require mechanical or electrical connections be made and hose and cable runs be established were reviewed to determine if they could be impacted by large internal flooding sources that are not seismically robust and do not require AC power.

The following summarizes this review's findings:

BDB electrical connection points, all of which are located in the East or West Switchgear Rooms, except for the connection supporting power to the BDB Spent Fuel Pool (SFP) Level Instrumentation system, would not be affected by internal flooding because no significant flooding sources are present in these rooms on a loss of power.

A fire water standpipe is located within the Control Building, but is normally isolated by a valve outside the Control Building.

BDB SFP Instrumentation and power source connections are located at elevation 24'- 6" and elevation 43'- 6" of the Auxiliary Building. These locations would not be affected by internal flooding because no significant non-seismic flooding sources are present on a loss of power.

The Emergency Diesel Generators (EDG's) have no internal flooding sources that are not seismically robust and do not require AC power. Fire water to the A and B EDG Rooms is normally isolated from outside the rooms and drained. Manual action is required to activate both fire suppression systems.

All BDB mechanical connections are made at or near grade in the Engineered Safety Features (ESF) Building except for those associated with manually controlling the Atmospheric Dump Valves (ADVs) by connecting an alternate air supply. The ability to

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 6 of 59 make these connections would not be affected by internal flooding because no significant non-seismic flooding sources are present on a loss of power.

Equipment and connections associated with alternate Steam Generator (SG) fill and manually controlling the ADV's are located within the Main Steam Valve (MSV) Building and would not be affected by internal flooding because no significant non-seismic flooding sources are present on a loss of power.

Review of the Unit 3 Turbine Building found tanks that are not seismically designed.

Given the volume of the Condenser pit, no credible internal flooding source exists that could prevent BDB strategies from being implemented. If a significant volume of water accumulated in the condenser pit and the normal hot well connection could not be made for transfer of water, then the transfer pump could alternatively take suction from the Condenser pit.

A flooding source that can potentially function following a seismic event and a loss of all AC power is the diesel driven Fire Pump in the event that the fire piping is ruptured and the pump auto-starts. MP3 fire piping is considered seismically rugged due to the construction standards it was built to. Therefore, this is not considered a credible flooding source.

Inadvertent actuation of the water based fire suppression systems due to a seismic events is possible, but would not represent a large flooding source. Further, in all cases, termination of the actuation is possible either through isolation or stopping the diesel driven Fire Pump.

BDB actions taken within the ESF Building would not be affected by ground water in-leakage because all mechanical BDB FLEX connections within this building are at the 24'-6" or 21'-6" elevations. Ground water in-leakage accumulates at lower levels of the ESF Building and does not represent a significant flooding source. No credit is needed for the groundwater sump pump in the ESF basement level.

The Control Room is not susceptible to internal flooding.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 7 of 59 ISE Cl 3.1.1.4.A The licensee's plan for implementing the use of off-site resources is not complete. The local assembly areas have not been identified. The licensee is also evaluating the possibilityof boat transportfor personnel.

DNC Response:

National SAFER Response Center (NSRC) local staging areas, access route evaluations, and transportation evaluations to the site have been determined as documented in the SAFER Trip Report for Millstone Power Station, 12-9211132-000, dated November 20, 2012..

The SAFER Response Plan (First Issue) dated June 25, 2014 for Millstone has been issued.

Both SAFER documents have previously been provided to the NRC staff and are available for their review.

The possibility of transporting personnel from Long Island Sound to the MPS site by boat has been evaluated and determined not to be a feasible option.

In response to questions received during the July 2014 MPS BDB Onsite Audit regarding details of the coordination with state and local emergency management organizations, the following information is provided:

DNC's BDB and Emergency Preparedness (EP) groups in conjunction with the SAFER organization, conducted an information sharing session in February 2014 titled, "Industry Approach to Addressing Order EA-12-049: Mitigating Strategies." This session was attended by state emergency responders and was conducted at the Millstone Nuclear Training Center. State and local jurisdiction/EPZ town emergency management directors and other appropriate interested state level emergency management officials were in attendance. The session consisted of a joint presentation by DNC's BDB group and the SAFER organization on what is a Beyond Design Basis External Event (BDBEE). Also addressed was DNC's BDB Mitigating Strategies which included the SAFER organization's response to a BDBEE. The presentation was followed by a question and answer session.

Personnel from the BDB Group scheduled a meeting with the State of Connecticut -

Emergency Management Program Supervisor in August 2014 to discuss the SAFER Response Plan, including the proposed routes for the delivery of emergency equipment from the RRC and the on-site/off-site staging areas. The Emergency Management Program Supervisor is the Millstone's Single-Point-Of-Contact for the State of Connecticut, covering the Department of Emergency Management, State Department of Transportation (DOT), and State National Guard. Following this meeting an additional

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 8 of 59 information exchange occurred in October 2014 with Department of Emergency Management, State DOT, and State Police personnel. This second information exchange occurred with representatives from EP and BDB groups. The Beyond Design Basis group made a presentation to the State of Connecticut at the Department of Emergency Services & Public Protection (DESPP) Headquarters in Middletown, CT.

This presentation covered the Background and Scope of the BDB Project and more specifically discussed the BDB-SAFER Response Plan, including the equipment and logistics for the delivery and set-up of pooled industry equipment to be used for long term response to a BDB event. Implementation of the SAFER plan involves cooperation with many State and Federal agencies, including the Department of Emergency Management and Homeland Security (DEMHS), DOT, State Police, National Guard, and the Federal Aviation Adminstration (FAA). This presentation provided a crucial interface between the Station and the State Emergency Response personnel. All attendees received a copy of the Millstone SAFER Response Plan document.

EP attends periodic meetings with the local officials that ensures offsite agencies are aware of and will support emergency response capabilities such as communications associated with a spectrum of emergency events, including BDB External Events in accordance with Millstone's Emergency Response Plan.

Regarding the coordination of transportation modes for the delivery of equipment from off-site resources, DNC has placed a decision point within procedural guidance document EP-AA-FLX-101, Single Point of Contact, for use by DNC's Single Point of Contact (SPOC). Per the Attachment titled, "Millstone Travel Route Assessment and Debris Removal Guidance," the DNC SPOC is responsible for interfaces with the SAFER organization as the NSRC Phase 3 equipment is traveling towards the station.

The procedural guidance in this attachment includes a step that states:

"IF projected road/haul paths from Bradley International Airport to the station are inaccessible or bridges have not been cleared by the State Emergency Management for NSRC ground transportation, THEN contact SAFER SPOC to dispatch ground transportation to Staging Area C for helicopter operation to the station."

The SAFER Response Plan has steps to simultaneously or concurrently start notification of helicopter support from the multiple commercial helicopter companies, State National Guard via State Emergency Management organizations, and Federal support through the Federal Emergency Management Agency (FEMA).

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 9 of 59 ISE CI 3.2.1.A Confirm that the NOTRUMP analysis provided in Section 5.2.1 of WCAP-17601-P, Revision I is applicable to MPS3 and supports the licensee's sequence of events.

DNC Response:

Engineering Technical Evaluation ETE-NAF-2012-0150, "Evaluation of Core Cooling Coping for Extended Loss of AC Power (ELAP) and Proposed Input for Dominion's Response to NRC Order EA-12-049 for Dominion Fleet", Revision 2, September 25, 2014, documents the review of the applicability of WCAP-17601-P. ETE-NAF-2012-0150 Section 6.2.1 states that Section 5.2.1 of WCAP-17601 provides a Reference Case which assumes standard RCP seal packages to determine the minimum adequate core cooling time with respect to RCS inventory (i.e., core uncovery). The Reference Case models a Westinghouse 4-loop plant with a core height of 12 feet (i.e., a 412 plant), at 3723 MWt, with Model F SGs and Model 93A/A-1 RCPs. This plant model is representative of MPS3.

The MPS3 ELAP analyses are based on the generic NOTRUMP thermal-hydraulic analyses presented in Section 5.2.1 of WCAP-17601. The analysis assumes that the RCP seal leakage will progress from the normal seal leakoff rate of 3 gpm/RCP to 21 gpm/RCP (at normal operating pressure of 2250 psia) after 13 minutes. This is consistent with the upper bound expectation for leakage for Westinghouse RCP seals discussed in the PWROG Position Paper.

Westinghouse issued advisory letter NSAL-14-1 on February 10, 2014. The NSAL indicated that the nominal upper bound RCP seal leak rate of 21 gpm for each RCP, as documented in WCAP-10541 Revision 2, may not be applicable for all plants with Westinghouse RCPs because of the various thermal-hydraulic conditions set up by plant specific seal leakoff piping designs.

In response to industry requests for assistance in resolving the issues raised in NSAL-14-1, Westinghouse prepared report PWROG-14015-P, Revision 0, "No. 1 Seal Flow Rate for Westinghouse Reactor Coolant Pumps Following Loss of All AC Power, Task 2, Determine Seal Flow Rates." Task 1 of the PWROG document authorized project, PA-SEE-1 196, which evaluated the various seal leakoff line design configurations for the Westinghouse fleet and "binned" the configurations into six different categories. In Task 2, leakage rates vs. RCS pressure were calculated based on conservatively bounding assumptions (i.e., result in high leak rates for each of the six categories).

MPS3 falls into Category 1, as defined by Westinghouse based, on the seal leakoff line flow orifice diameter, the piping lengths and diameters upstream and downstream of the flow orifice and additional components (valves, bends) in the line.

Currently, the MPS3 RCP seal configuration is two Westinghouse OEM RCP seals and two Flowserve N-seals installed. The remaining two Westinghouse seals are currently

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 10 of 59 scheduled to be replaced with Flowserve N-seals during the spring 2016 outage. ETE-NAF-2012-0150, Rev. 2 performed an evaluation of the projected times to enter reflux cooling for the RCP seal configuration two Westinghouse Original Equipment Manufacturer (OEM) seals and two Flowserve low-leakage seals. ETE-NAF-2012-0150, Rev. 2 incorporates the Westinghouse RCP OEM seal leakage values from PWROG-14015, Revision 1 for Category 1, and the normal response leakage for the Flowserve seal. The results are discussed in detail in ISE 3.2.1.1.A and 3.2.1.2.A.

ETE-NAF-2012-0150, Rev. 2 concluded that the reference 4-loop NOTRUMP analysis presented in WCAP-17601-P is bounding for the MPS3 RCP seal configuration of two Westinghouse seals and two Flowserve N-seals.

The PWROG has developed a White Paper on NOTRUMP [PWROG-14064]. The purpose of this white paper is to document the applicability of the NOTRUMP code for the evaluation of the ELAP event and application of its results with regards to criteria for boron mixing and reflux cooling for Westinghouse designed PWRs. From PWROG-14064, the comparison of results from the NOTRUMP and TRACE computer codes for the parameters of interest show that the NOTRUMP predicted results agree well or are conservative with respect to the NRC's TRACE predicted results. The comparison shows that NOTRUMP provides a conservative estimate of the required time when the primary makeup pumps are required for an ELAP event. Therefore, it is concluded that NOTRUMP is acceptable for simulation of the ELAP event within the criteria for reflux cooling and boron mixing.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

ISE Cl 3.2.1.1.A Confirm that the use of the NOTRUMP code for the ELAP analysis is limited to the flow conditions prior to reflux condensation initiation. This includes specifying an acceptable definition for reflux condensation cooling.

DNC Response:

As stated in the OIP Six-month Status Report dated August 23, 2013 (S/N 12-161D),

RCS inventory makeup will be initiated within 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months after the start of an ELAP /

LUHS event. PWROG Letter OG-14-60, Generic Information to Support Requests for Additional Information in US NRC Reviews of FLEX Overall Integrated Plans with Regard to Reflux Cooling, LTR-LIS-14-79, (PA-ASC-1 197), February 13, 2014, provided the time of transition to reflux cooling and a definition for reflux condensation cooling for Westinghouse plants as:

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 11 of 59 "Millstone Unit 3 has used generic ELAP analyses performed with the NOTRUMP computer code to support the mitigating strategy in its Overall Integrated Plan (OIP).

The use of NOTRUMP was limited to the thermal-hydraulic conditions before reflux condensation initiates. The initiation of reflux condensation cooling is defined when the one hour centered moving average (CMA) of the flow quality at the top of the SG U-tube bend exceeds 0.1 in any one loop."

For 4-loop plants with an assumed RCP seal leakage rate of 21 gpm/seal at normal operating pressure and temperature along with choked flow conditions with no area change in the seal package or #1 seal leak-off system, reflux cooling was calculated to occur at 17.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Westinghouse issued advisory letter NSAL-14-1 on 2-10-14. The NSAL indicated that the nominal RCP seal leak rate of 21 gpm for each RCP, as documented in WCAP-10541 Revision 2, may not be applicable for all plants with Westinghouse RCPs because of the various thermal-hydraulic conditions set up by plant specific seal leakoff piping designs.

In response to industry requests for assistance in resolving the issues raised in NSAL-14-1, Westinghouse prepared report PWROG-14015-P, Revision 0, "No. 1 Seal Flow Rate for Westinghouse Reactor Coolant Pumps Following Loss of All AC Power, Task 2, Determine Seal Flow Rates." Task 1 of the PWROG document authorized project, PA-SEE-1 196, which evaluated the various seal leakoff line design configurations for the Westinghouse fleet and "binned" the configurations into six different categories. In Task 2, leakage rates vs. RCS pressure were calculated based on conservatively bounding assumptions (i.e., result in high leak rates for each of the six categories).

MPS3 falls into Category 1, as defined by Westinghouse, based on the seal leakoff line flow orifice diameter, the piping lengths and diameters upstream and downstream of the flow orifice and additional components (valves, bends) in the line. The calculations were performed using the two phase flow code ITCHSEAL, previously described in WCAP-10541.

The PWROG has developed a White Paper on NOTRUMP [PWROG-14064]. The purpose of this white paper is to document the applicability of the NOTRUMP code for the evaluation of the ELAP event and application of its results with regards to criteria for boron mixing and reflux cooling for Westinghouse designed PWRs. From PWROG-14064, the comparison of results from the NOTRUMP and TRACE computer codes for the parameters of interest show that the NOTRUMP predicted results agree well or are conservative with respect to the NRC's TRACE predicted results. The comparison shows that NOTRUMP provides a conservative estimate of the required time when the primary makeup pumps are required for an ELAP event. Therefore, it is concluded that NOTRUMP is acceptable for simulation of the ELAP event within the criteria for reflux cooling and boron mixing. Application of the NOTRUMP simulations reference cases require the implementation of the RCS makeup pump at the times in PWROG-14064 Table A (17.0 hours0 days 0 hours 0 weeks 0 months for 4-loop plant), which supersedes information provided in PWROG Letter OG-14-60 (17.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for a 4-loop plant).

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 12 of 59 The results in Table A are based on the above definition of reflux cooling from PWROG-14064. This condition is considered to be conservative since it is defined prior to either the onset of inadequate boron mixing in the RCP suction leg or reflux cooling heat transfer. That is, when this set of circumstances occurs, the dilution process in the RCP suction legs has not yet started. As such, the RCS has not yet reached a stratified state where true reflux cooling heat transfer is possible. Thus, the definition of the onset of reflux cooling is conservative for establishing the time when RCS makeup is desired.

The PWROG has documented leakage rates for Westinghouse OEM RCP Seals in PWROG-14015, Revision 1 using bounding plant configurations. The initial leakage information provided in PWROG-14015, Rev. 0, consisted of three points: initial leakage at normal operating temperature and normal operating pressure (NOT/NOP); peak leakage at 1500 psia; and leakage at the cooled-down, depressurized conditions.

Additional studies have been documented in PWROG-14015, Revision 1 to evaluate the linear assumptions between points and the effect of minimal subcooling for Category 1 seals. The intermediate flow rates are slightly above what is predicted by the linear assumption for seal leakage between the points identified in PWROG-14015, Rev. 0.

The seal leakage flow rate is almost unaffected from the change from 5°F of sub-cooling to less than 1OF of sub-cooling. For each pressure analyzed, these points are within 0.1 gpm.

As discussed in the Flowserve White Paper of February 11, 2014 and the subsequent responses to NRC review questions, the expected response of the Flowserve seal for plants which initiate an early RCS cooldown and depressurization (such as MPS3), is nominal controlled bleedoff (CBO) flow, and negligible 3rd stage leakage for an extended period (i.e., several weeks).

Engineering Technical Evaluation ETE-NAF-2012-0150, Rev. 2 performed an evaluation of the projected times to enter reflux cooling for the RCP seal configuration which will apply at the time of FLEX implementation for MPS3 (i.e., two Westinghouse OEM seals and two Flowserve low-leakage seals). ETE-NAF-2012-0150, Rev. 2 incorporates the Westinghouse RCP OEM seal leakage values from PWROG-14015, Revision 1 for Category 1, and the normal response leakage for the Flowserve seal.

Based on that assessment, entry into reflux cooling would be extended from the 17.0 hours0 days 0 hours 0 weeks 0 months noted in PWROG-14064 Table A to 23.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months . Since the latest time of deployment of FLEX RCS makeup is 16.0 hours0 days 0 hours 0 weeks 0 months per the OIP timeline, and the RCS makeup rate of 45 gpm significantly exceeds the expected leakage rate at 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , the reflux cooling condition is avoided with a time margin of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for the fall 2014 RCP seal configuration.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 13 of 59 ISE Cl 3.2.1.2.A If the RCP seal leakage rates used in the plant-specific ELAP analyses are less than the upper bound expectation for the seal leakage rate discussed in the PWROG position paper addressing the RCP seal leakage (ADAMS Accession No. ML13235A151 (Non-Publicly Available)) or justification should be provided for use of a lower value. If the seals are changed to non-Westinghouse seals, the acceptability of the use of non-Westinghouse seals should be addressed, and the RCP seal leakage rates for use in the ELAP analysis should be justified.

DNC Response:

MPS3 RCPs currently have two Westinghouse shaft seals and two Flowserve N-seals installed. The remaining two Westinghouse seals are currently scheduled to be replaced with Flowserve N-seals during the spring 2016 outage.

The MPS3 ELAP analyses are based on the generic NOTRUMP thermal-hydraulic analyses presented in Section 5.2.1 of WCAP-17601. The analysis assumes that the RCP seal leakage will progress from the normal seal leakoff rate of 3 gpm/RCP to 21 gpm/RCP (at normal operating pressure of 2250 psia) after 13 minutes. This is consistent with the upper bound expectation for leakage for Westinghouse RCP seals discussed in the PWROG Position Paper.

In response to industry requests for assistance in resolving the issues raised in NSAL-14-1, Westinghouse prepared report PWROG-14015-P, Revision 0, No. 1 Seal Flow Rate for Westinghouse Reactor Coolant Pumps Following Loss of All AC Power, Task 2, Determine Seal Flow Rates. Task 1 of the PWROG document authorized project, PA-SEE-1196, which evaluated the various seal leakoff line design configurations for the Westinghouse fleet and "binned" the configurations into six different categories. In Task 2, leakage rates vs. RCS pressure were calculated based on conservatively bounding assumptions (i.e., result in high leak rates for each of the six categories).

MPS3 falls into Category 1, as defined by Westinghouse based, on the seal leakoff line flow orifice diameter, the piping lengths and diameters upstream and downstream of the flow orifice and additional components (valves, bends) in the line. The calculations were performed using the two phase flow code ITCHSEAL, previously described in WCAP-10541.

As discussed in the Flowserve White Paper of February 11, 2014 and the subsequent responses to NRC review questions, the expected response of the Flowserve seal for plants which initiate an early RCS cooldown and depressurization (such as MPS3), is nominal controlled bleedoff (CBO) flow and negligible third stage leakage for an extended period (i.e., several weeks).

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 14 of 59 Engineering Technical Evaluation ETE-NAF-2012-0150, Rev. 2 performed an evaluation of the projected times to enter reflux cooling for the RCP seal configuration which will apply at the time of FLEX implementation for MPS3 (i.e., two Westinghouse OEM seals and two Flowserve low-leakage seals). ETE-NAF-2012-0150, Rev. 2 incorporates the Westinghouse RCP OEM seal leakage values from PWROG-14015, Revision 1 for Category 1, and the normal response leakage for the Flowserve seal.

Based on that assessment, entry into reflux cooling would be extended from the 17.0 hours0 days 0 hours 0 weeks 0 months noted in Table A above (see Response to ISE Cl 3.2.1.1.A) to 23.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months , providing a margin of approximately 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months from the expected time of deployment of the FLEX RCS injection pump, or 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

DNC Calculation MISC-11790 showed that, for the target post cooldown steam pressure of 300 psig, boron addition to maintain k-effective < 0.99 is not needed before approximately 26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months . Since initiation of the FLEX RCS makeup occurs no later than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , and will maintain adequate conditions for RCS mixing as demonstrated above, the available mixing time is 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . This meets Criterion 3 of the NRC's January 8, 2014 letter:

(3) In all cases, credit for increases in the reactor coolant system boron concentration should be delayed to account for the mixing of the borated primary makeup with the reactor coolant system inventory. Provided that the flow in all loops is greater than or equal to the corresponding single-phase natural circulation flow rate, the staff considers a mixing delay period of one hour following the addition of the targeted quantity of boric acid to the reactor coolant system to be appropriate.

Based on the assessment above, the reference 4-loop NOTRUMP analysis presented in WCAP-17601-P is bounding for the MPS3 RCP seal configuration of two RCPs with Westinghouse seals and two RCPs with Flowserve seals. The applicability has been demonstrated using Westinghouse seal performance which is specific to the MPS3 seal leakoff line configuration as calculated in PWROG-14015-P, Revision 0. The analysis also demonstrates that initiation of FLEX RCS makeup by no later than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months will avoid the reflux cooling condition and provide for adequate boron mixing conditions with significant time margins.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

ISE CI 3.2.1.2.B For Westinghouse Reactor Coolant Pump (RCP) seals, a discussion (including the applicable analysis and relevant seal leakage testing data) should be provided to justify that (1) the integrity of the associated 0-rings will be maintained at the temperature

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 15 of 59 conditions experienced during the ELAP event, and (2) the seal leakage rate of 21 gpm/seal used in the ELAP is acceptable.

DNC Response:

(1) It has been noted that some assumptions regarding the potential temperatures in the seal compartment in the pre-cooldown stages of the ELAP event [WCAP-17601, Section 4.4.1.1] were not bound by the testing performed to date [WCAP-10541, Rev. 2, Supplement 1]. This deviation has been noted and is being addressed by an on-going Westinghouse testing program. It should be noted that the NRC currently accepts the improvement of seal conditions and seal leakage for Westinghouse design plants with respect to the same failure mode and effects [WCAP-1 5603 Rev. 1-A] and, as such, it is anticipated that the higher temperature qualification required will be determined to be acceptable based on the on-going Westinghouse testing program.

(2) For a discussion of the acceptability of the 21 gpm seal leakage rate used in the analysis in light of the issues raised in Nuclear Safety Advisory Letter NSAL 14-1, subsequent analysis of RCP seal leakoff rates for a leakoff line configuration which bounds MPS3 and the expected fall 2014 RCP seal configuration of two Westinghouse Seals and two Flowserve Seals, see the response to ISE Cl 3.2.1.2.A.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

ISE Cl 3.2.1.2.C If the seals are changed to the newly designed Generation 3 SHIELD seals, or non-Westinghouse seals, justify the acceptability of the use of the newly designed Generation 3 SHIELD seals or non-Westinghouse seals and the RCP seal leakages rates for use in the ELAP analysis. (prior to on-set of reflux boiling with adequate margin.)

DNC Response:

MPS3 RCPs currently have two Westinghouse shaft seals and two Flowserve N-seals installed. The remaining two Westinghouse seals are currently scheduled to be replaced with Flowserve N-seals during the spring 2016 outage. (See also the response to ISE Cl 3.2.1.2.A.)

Flowserve has provided information on the leakage of the N-Seal RCP seal in the Flowserve document "White Paper on the Response for the N-Seal Reactor Coolant Pump (RCP) Seal Package to Extended Loss of AC Power (ELAP)," Revision 0, dated February 11, 2014 (Proprietary). MPS3 plans to initiate a cooldown/depressurization no

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 16 of 59 later than two hours after declaration of an ELAP event. At that time, the maximum temperature of the RCS will be 572°F. In accordance with existing Emergency Operating Procedures, cooldown will proceed at 70-100°F per hour, ending at an RCS temperature of -419'F. The Flowserve paper documents a test that generally exceeds the temperature for the MPS3 RCP N-Seal seals during approximately the first 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months of the ELAP. The integration of the temperature difference between the test temperature and the seal temperature during the ELAP, when converted using the Arrhenius equation to estimate material degradation, shows that the N-Seal seals can remain at 419°F for more than a day before leakage increases. The initial leakage is minimal and improves the RCS response significantly. The additional leakage also occurs well after the RCS makeup pump is deployed and installed for RCS inventory makeup. An evaluation was performed which determined that the total integrated leakage for MPS3 with two Flowserve N-seals and two Westinghouse OEM seals would be less than the leakage assumed by the Westinghouse analysis which supported the deployment of the RCS makeup pump deployment at 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months which was prior to on-set of reflux boiling with adequate margin. Therefore, with the low leakage Flowserve seals installed, the margin increases.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

ISE Cl 3.2.1.3.A Confirm that the licensee has addressed the applicabilityof assumption 4 on page 4-13 of WCAP-17601-P, and confirm that the values used for the requested parameters in the Westinghouse calculations that were performed using the ANS 5.1 1979 +2 sigma decay heat model bound initial condition 3.2.1.2(1) of NEI12-06, Section 3.2.1.2.

DNC Response:

Westinghouse Letter LTR-LIS-13-515 [1], Attachment 1, page 4 of 5, addresses the applicability of assumption 4 on page 4-13 of WCAP-17601-P, and concludes that the values used for the requested parameters in the Westinghouse calculations that were performed using the ANS 5.1 1979 + 2 sigma decay heat model bound initial condition 3.2.1.2(1) of NEI 12-06, Section 3.2.1.2.

The Westinghouse NSSS calculations documented in WCAP-17601-P using the NOTRUMP code were performed with the ANS 5.1 1979 + 2 sigma decay heat model and assumed the reactor is initially operating at 100% power (NOTRUMP reference case core power is 3723 MWt). Implementation of this model includes fission product decay heat resulting from the fission of U-235, U-238, and Pu-239 and actinide decay heat from U-239 and Np-239. The power fractions are typical values expected for each of the three fissile isotopes through a three region burn-up with an enrichment based on typical fuel cycle feeds that approach 5%. With that, a conversion ratio of 0.65 was used

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 17 of 59 to derive the decay power of the two actinides U-239 and Np-239. Fission product neutron capture is treated per the ANS standard. The decay heat calculation utilizes a power history of three 540 day cycles separated by two 20 day outages that bounds initial condition 3.2.1.2 (1) of the NEI document NEI 12-06, Section 3.2.1.2 (minimum assumption of NEI 12-06 is that the reactor has been operated at 100% power for at least 100 days prior to event initiation). Therefore, the decay heat curve assumed in the Westinghouse calculations is representative of MPS3.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

ISE Cl 3.2.1.6.A The licensee stated that for Action Item l Ithe portable boric acid batching tank will be deployed at 12 - 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months , if the RWST tank is not available. Confirm that the deployment time of 12 - 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months is acceptable.

DNC Response:

In the initial assessment for RCS Make-Up, it was determined that adding water to the RCS would not be required for at least 26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months . At that time the deployment of the portable boric acid batching tank in the 12 to 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months timeframe was deemed acceptable. Subsequently, the RCS Inventory strategy has been revised to require the injection of borated water at the 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months mark. In order to meet this requirement, the portable batching tank and BDB RCS Injection pump will be deployed from the BDB Storage Building in the 10 to 14 hour1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months timeframe. The Storage Building is approximately

/4 mile from the area where the tank and pump will be staged. By this time, debris will have been cleared due to deployment of other Phase 2 equipment. This deployment schedule will provide a minimum two hour window to set up the tank, connect the tank to the RCS Injection pump, and batch the boric acid in time to inject into the RCS at 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

ISE CI 3.2.3.A The strategy for containment cooldown and depressurization will be completed per the schedule given in the August 23, 2013 6- Month Status Update. The detailed validation analysis will be completed later this year and the results will be provided in the February 2014 6-Month Status Update. Confirm that the analysis and the strategy to maintain the containmentparameterswithin acceptable limits is satisfactory.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 18 of 59 DNC Response:

Overall Integrated Plan (OIP) Open Item No. 4 was completed and documented as "Complete" in the Six-Month Status Update Report dated February 28, 2014 (SN: 12-161F). Attachment 2 of the update report provided the Containment cooling strategy.

OIP Open Item No. 5 addresses the thermal and hydraulic calculations, which confirm that the Containment strategies are adequate. Calculation 13-015, "MP2 and MP3 FLEX Proto-Flo Model and Analysis," Rev. 3 documents hydraulic analyses that confirm FLEX fluid hydraulic design and coping strategy objectives can be met for MPS3 NSRC High Flow/Low Pressure Pump network confirming ultimate heat sink (UHS) coolant flow rates into the service water (SW) system to support the Phase 3 coping strategy objectives. Specifically, the Phase 3 coping strategy includes restoring Reactor Plant Component Cooling Water (RPCCW) heat exchanger coolant flow to support decay heat removal via residual heat removal (RHR) system operation.

Calculation 13-015, "MP2 and MP3 FLEX Proto-Flo Model and Analysis," Rev. 3 has previously been provided to the NRC staff and is available for their review.

ISE CI 3.2.4.2.A Analyses to evaluate the effects of loss of ventilation in various areas are currently underway. Upon completion of these analyses, detailed strategies and operatoraction timelines will be developed for the implementation of compensatory measures to maintain the area temperatures below the applicable design limits, if necessary. The results will be provided in the February 2014 6-month update. Confirm that the analyses and the compensatory measures show that room temperaturesare acceptable to maintain functionality of the equipment needed to carry out the mitigation strategies.

DNC Response:

Detailed evaluations of the effects of loss ventilation during an ELAP/LUHS have been completed and documented in the MPS3 FLEX basis document (Engineering Technical Evaluation ETE-CPR-2012-0008 Rev. 4, Section 10.4.2). The purpose of these evaluations was threefold:

" To identify areas containing electrical and/or mechanical equipment heat loads that would be energized and or credited during the various phases of ELAP/LUHS coping strategies.

" To determine the steady state ambient air temperatures for these areas during the Loss of Ventilation (LOV) following an ELAP.

To identify short and long-term compensatory cooling measures, if necessary and the timing for the implementation of these measures to maintain area temperatures below the acceptable levels for both equipment and personnel habitability.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 19 of 59 Typically, the results show that opening doors early in ELAP/LUHS scenario can significantly limit temperature increases due to loss of forced ventilation, therefore, maintaining room temperatures generally less than 120 OF.

Ventilation of the SFP area for removal of steam from SFP boil-off is addressed separately in Section 6.2.2 of ETE-CPR-2012-0008. Ventilation is established by opening five doors (three roll-up doors and two personnel doors) at elevations below and above the SFP surface creating a chimney effect in the SFP area.

During the July 2014 NRC BDB Onsite Audit at MPS, it was identified that the DNC Main Steam Valve House (MSVH) habitability analysis (NAI-1731-001, Rev 000, addendum A) did not consider a missile shield that is positioned outside and directly in front of Door SV-71-1 (Security Door # 376) and could affect ventilation and heat load in the MSVH Atmospheric Dump Valve (ADV) room. The MSP3 MSVH habitability analysis has been revised to evaluate the impact of this previously unaccounted for reinforced concrete missile shield. This door is located on the 71' elevation of the MSVH and connects the MSVH to the outside atmosphere. The opening of door #376, plus four (4) other doors, at the one hour mark post-ELAP, is credited to reduce the area temperatures prior to a second entry into the area, at two-hours into the ELAP event for manual operation of the ADV bypass valves by the Operations staff.

As expected, the results of the revised analysis (NAI-1731-001, Revision 00 Addendum B) indicate that after opening of the five MSVH doors, the temperature in the vicinity of the valves at the two-hour mark drops to -199 0 F, compared to - 176 0 F temperature that was calculated in the Addendum A analysis. This is primarily due to the inclusion of the missile shield in front of door #376 which introduces an additional restriction that must be overcome by the buoyant hot air exiting the MSVH to the outside atmosphere. This temperature, although higher than the previous analysis results, continues to remain bounded by the protection that is provided by the required protective gear identified in Section 10.4.2.2.3 of the MPS3 Beyond Design Basis - FLEX Strategy Overall Integrated Plan Basis Document (ETE-CPR-2012-0008, Rev. 4).

With regards to long-term equipment qualification in the MSVH, following the restoration of 480 VAC (around 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ), the operation staff is directed per FSG-4, Attachment 3 to energize one of the exhaust fans in the MSVH to supplement the natural ventilation to rapidly reduce the area temperatures.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 20 of 59 ISE Cl 3.2.4.2.B Confirm that the habitability limits of the main control room will be maintained in all Phasesof an ELAP.

DNC Response:

The Control Room Complex habitability analysis has been documented in the MPS3 FLEX basis document (Engineering Technical Evaluation ETE-CPR-2012-0008 Rev. 4, Section 10.4.2) As discussed in Sub Section 10.4.2.2.1, Calculation MISC-11802, Rev.

0, demonstrated that no compensatory cooling measures are required to maintain the area temperatures below the habitability limit during all phases of the ELAP/LUHS scenario.

For defense in depth, FSG-5 includes guidance to monitor area temperatures and to implement contingency compensatory measures, such as opening doors or placement of fans, if area temperatures continue to indicate increasing trends.

During the July 2014 MPS3 BDB Onsite Audit, the NRC staff requested DNC to provide the specific MPS3 Main Control Room (MCR) habitability limits used as the basis of the response to ISE Cl 3.2.4.2.B.

Accordingly, the MCR habitability limit is 11 0°F and the equipment operability limit in the MCR is 120'F. These limits are consistent with NUMARC 87-00, Revision 1, "Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors."

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

ISE Cl 3.2.4.4.A Confirm the adequacy of existing lighting and the adequacy of portable lighting to perform FLEX strategy actions.

DNC Response:

In order to validate the adequacy of supplemental lighting and the adequacy and practicality of using portable lighting to perform FLEX strategy actions, an evaluation of the tasks to be performed and the available battery powered emergency lighting in the designated task areas was completed. The results are documented in Section 10.5 of Engineering Technical Evaluation ETE-CPR-2012-0008. Tasks evaluated included traveling to/from the various areas necessary to implement the FLEX strategies, making

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 21 of 59 required mechanical and electrical connections, performing instrumentation monitoring, and component manipulations.

Some of the battery powered emergency lights are designated as Appendix R lights.

These Appendix R lights are designed and periodically tested to insure the battery pack will provide a minimum of eight hours of lighting with no external AC power sources.

The remainder of the emergency lights are designed and tested for 90 minutes of lighting.

Battery powered emergency lights were determined to provide adequate lighting for all interior travel pathways needed to access the BDB connection points. However, in some areas supplemental lighting will be required to safely perform the required electrical and mechanical connections. Supplemental lighting may also be required to return from the required areas depending on when and how long these actions are required.

Supplemental lighting consists of ready to use flashlights and Remote Area Lighting Systems (RALS). The RALS will be deployed to support the FLEX strategy tasks. These RALS's are rechargeable light-emitting diode (LED) lighting systems designed to power the LED lights for seven hours at 6000 Lumens and 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months at 500 lumens.

FSG-5 directs the operators to use the flashlights and RALs as part of their deployment.

Accordingly, a number of RALS and flashlights are stored in a fully charged condition in a location nearby the MPS3 Control Room (Control Building El. 64'-6") for easy access.

BDB supplemental lighting equipment, will be accounted for and maintained (except for flashights) per the site Preventative Maintenance (PM) program.

There are no emergency lighting fixtures in the yard outside the protected area to provide necessary lighting in those areas where portable BDB equipment is to be deployed. Therefore, the diesel powered pumps and generators are outfitted with light plants that are powered from their respective diesels to support connection and operation. In addition to the lights installed on the portable BDB equipment, portable light plants are included in the FLEX response strategies. These portable diesel powered light plants can be deployed from the BDB Storage Building as needed to support night time operations. Additional portable light plants will be available from the National SAFER Response Center.

In addition to installed emergency DC lighting, flashlights and the stored RALSs and portable light plants, the BDB Storage Building contains a ready to use stock of flashlights and head lights to further assist the staff responding to a BDB event during low light conditions.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 22 of 59 ISE Cl 3.2.4.4.B Confirm that upgrades to the site's communications systems have been completed DNC Response:

The study documenting the communications strategy has been completed.

Subsequently, Overall Integrated Plan (OIP) Open Item No. 16 was documented as "Complete" in the Six-Month Status Report dated February 28, 2014 (SN: 12-161F).

The plan concluded that for MPS3, FLEX strategies can be effectively implemented with a combination of sound powered phones, satellite phones and hand-held radios.

Although the overall communications plan has not changed, the details regarding the components to be used and the number of components have continued to evolve. At this time, the quantity of components needed to implement the communications strategy has been determined to be 40 satellite phones, 40 hand-held radios, and 10 additional, dedicated sets of sound powered phone headsets and extension cords. The communications equipment is common to both MPS2 and MPS3, with the exception of the sound-powered phones and five (5) of the satellite phones located in the MPS2 Control Room. Distribution of the satellite phones includes both Main Control Rooms (MCR), the Technical Support Center (TSC), the Local Emergency Operations Facility (LEOF), and the surrounding county Offsite Response Organizations (OROs). The hand-held radios are for command and control of the FLEX mitigating strategies and include 10 spare radios per unit and three batteries per device.

The required communications equipment has been received and tested onsite and all required modifications to implement the MPS3 communications strategy were completed prior to the end of the fall 2014 refueling outage.

The MCR and TSC satellite phones were installed per plant design change DC MPG-14-01080. Hand-held satellite phones will be available for initial notifications of state, federal, and local authorities. Afterward, Security personnel will deploy a satellite phone antennae setup. This antennae setup will be established with a fiber optics cable from an outdoor portable dish antennae to the installed inside "desk sets." This portion of the communications strategy is intended to suffice for approximately the first six hours.

Once augmented staff arrives on site a mobile communications trailer designed to handle both satellite voice and data traffic, as well as to function as a radio repeater to enhance on-site communications, will be deployed from the BDB Storage Building by available personnel.

In response to specific questions received during the July 2014 NRC Onsite Audit, the following additional information is provided:

Question 1: Complete the assessment of the TA-312 sound powered phones for all five hazards (survivability), including potential challenges in the routing of the phones.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 23 of 59 The Battery Operated Field Phones are stored in a locker in the MPS3 Control Building and a locker in the MPS1 Control Room (for the MPS2 phones). The lockers each contain two Battery Operated Field Phones (TA-312), a DR-8 reel of WD-1 wire with two RL1 59 Reeling Machines, a roll of duct tape, packages of "D" cell batteries, reel wire connectors, wire nuts, roll of electrical tape, wire stripper/cutter, and a portable radio. These lockers are protected from flooding, high winds/missiles, extreme temperatures, and are secured/supported in order to be seismically protected.

Upon the loss of communication between the MCRs, the Battery Operated Field Phones and replacement batteries will be moved from their lockers to each of the Unit Supervisor's Desks and the WD-1 wire will be deployed between the MPS3 Control Room and MPS2 Control Room along the west side of the station (primary path). This is the shortest distance and offers the best opportunity for the person deploying the wire to communicate via portable radios, if necessary. If the west side of the station is inaccessible, an alternate path is to deploy the wire on the east side of the station which is the longest cable route. If outside conditions make deployment outside the site structures hazardous or impractical, there is an additional alternate deployment path inside the plant through the Turbine Buildings and the Condensate Polishing Facility.

The cable has been tested through closed doors, including water tight doors, and no damage to the wire or the wire jacket was observed as well as no reduction in sound quality.

Question 2: Look at implementing proceduraland emergency plan changes needed to allow each Unit to individually declare an ELAP, in the event communications between the CRs can be not effectively establishedin the requiredtimeframe. The impact on off-site organizationsneeds to be considered.

MPS has two separate control rooms which present a unique challenge for onsite communications. Each unit's control room operates independent of the other and each enter their own unit-specific procedures in response to a loss of AC power event. Site procedures currently establish the Shift Manager (SM) in the MPS3 Control Room as the Site Emergency Manager for matters that impact the entire site and for notifications to offsite federal, state, and local agencies.

However, in accordance with the E-Plan, each unit can declare an emergency situation to offsite agencies using the satellite phones available in each control room. In the event that Unit 2 is at minimum staffing when a BDB external event occurs, they may not have an available designated E-Plan communicator and would then rely on a "runner" to notify Unit 3 of the Unit 2 situation. In this case, Unit 3 would provide the required offsite notifications for both units.

For this minimum staffing scenario, a MPS2 Chemistry Technician will report to the MPS2 Control Room upon the loss of A/C power event. If communications between the control rooms is not available, the MPS2 SM will direct the Chemistry Technician to

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 24 of 59 immediately function as the "runner" to communicate the status of MPS2 to the MPS3 Control Room SM. This dialogue will occur upon declaration of the emergency on MPS2 and will allow the MPS3 SM to make appropriate notifications in accordance with the requirement of the Millstone Emergency Plan. Accordingly, there should be no impact on the notification of offsite organizations since they would be initiated as required.

After briefing the MPS3 SM, the MPS2 Chemistry Technician will then support the MPS3 Chemistry Technician in the process of deploying the Battery Operated Field Phones (TA-312) as required, as discussed in the response to Question 1, to establish direct communications between the Control Rooms. This deployment should be complete within 45 minute from the initiation of the event, which corresponds to the anticipated time at which each unit would be declaring that an ELAP has occurred.

However, if needed, the Chemistry Technicians can continue to function as "runners" in the event the Battery Operated Field Phones are not deployed and operating. Unit 2 would require information on the status of the site SBO diesel from the Unit 3 SM in order to declare an ELAP. If communication between the Control Rooms has not been established with the sound-powered field phones, this information would be communicated by a "runner."

By letter dated September 22, 2014, DNC responded to a Request for Additional Information regarding the MPS Phase 2 Staffing Report. In this response, the role of Security in the deployment of communications equipment was identified. Essentially, four Security Officers have been designated to assist with debris removal and staging FLEX and communications equipment. The use of security personnel to perform BDB functions was based upon a tiered approach to minimize the impact to security response capability.

With regard to communications equipment, Security will deploy the RapidCom satellite phone antennae arrangement. Later, they will deploy the Communications On Wheels (COW) trailer with a self powered satellite phone antennae setup and a radio repeater to improve site radio reception.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 25 of 59 ISE Cl 3.2.4.7.A Westinghouse is currently performing an analysis to determine the consequences of usage of impure water sources in the steam generators. The results of the analysis are expected to provide the allowed time limits on usage of these sources. The RRC will provide equipment to initiate residualheat removal and water treatment equipment such that heat removal can be ensured for extended durations. Confirm that the analysis results and resultantstrategiesare acceptable.

DNC Response:

The onsite water sources have a wide range of associated chemical compositions.

Therefore, extended periods of operation with the addition of these various onsite water sources to the SGs have been evaluated for impact on long term SG performance and SG material (e.g., tube) degradation (Westinghouse Calculation CALC-WEST-CN-CDME-13-12, Rev. 0, Addendum 0, "Supporting Chemistry Calculations for Alternate Cooling Source Usage During Extended Loss of All AC Power at Plant Millstone Unit 3,"

March 26, 2014."). The analysis provides guidance for times that the various onsite water sources can be used for core cooling, but do not define failure limits for BDB event response or for the SGs.

In addition to the onsite tanks, the city water supply would be available as a clean water source for the SGs. The analyses show that the city water could be used for approximately 301 hours0.00348 days 0.0836 hours 4.976852e-4 weeks 1.145305e-4 months after Demineralized Water Storage Tank (DWST) depletion before the SG design corrosion limit would be expected to be reached, or if a conservative Total Suspended Solids (TSS) level of 500 ppm is assumed, for about 219 hours0.00253 days 0.0608 hours 3.621032e-4 weeks 8.33295e-5 months after Condensate Storage Tank (CST) depletion before the limiting SG precipitation level would be expected to be reached. This evaluation also applies to the site fire systems since their water source is from the city water supply.

The onsite freshwater pond can be credited for all site hazards except for flooding. This water source could be used for approximately 97 hours0.00112 days 0.0269 hours 1.603836e-4 weeks 3.69085e-5 months after DWST depletion before the SG design corrosion limit is reached. The corrosion evaluation bounds the precipitation evaluation regarding the limiting time period. In the event of flooding due to storm surge and hurricane conditions, the onsite freshwater pond can become contaminated with saltwater. In these conditions, however, several onsite tanks containing clean water suitable for use in the SGs for extended periods of time are available.

Exceeding the expected time to reach the SG design corrosion limit would have an impact on SG tube integrity but an insignificant impact on the ability of the SGs to remove core decay heat from the RCS at significantly lower than design SG temperature/pressure conditions. However, reaching the limiting SGs precipitation levels could potentially impact/reduce SG heat transfer capabilities.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 26 of 59 The RWST could be used for approximately 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months after DWST depletion before the SG design corrosion limit is reached. Also, once the borated Refueling Water Storage Tank (RWST) is introduced into the SGs, the pH of the SG fluid is lowered, causing active corrosion. Thus, the SG corrosion rates resulting from the RWST introduction would not decrease until the concentration of boric acid is reduced by a method such as feed and bleed. Boric acid precipitation is not a concern based on the more limiting corrosion evaluation and elevated SG temperatures (the analyses were performed at a SG temperature of 420'F).

The evaluation shows that the Long Island Sound (LIS) seawater could cause through-wall pitting of the SG tubes within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . On this basis, water from LIS should not be considered for use except as a last resort.

During Phase 3, Reverse Osmosis (RO) / Ion Exchange (IX) equipment (up to 300 gpm capacity) is delivered from the NSRC and deployed to the site to remove impurities from the onsite water sources. Once the RO/IX equipment is in operation, the onsite water sources can provide for an indefinite supply of purified water.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

ISE Cl 3.4.A The licensee's plans for the use of off-site resources conform to the minimum capabilities specified in NEI 12-06 Section 12.2, with regard to the capability to obtain equipment and commodities to sustain and backup the site's coping strategies (item 1).

Confirm the licensee addresses the remaining items (2 through 10), or provides an appropriatealternative.

DNC Response:

Considerations 2 through 10 in Section 12.2 of NEI 12-06 are, in general, considerations applicable to the third party organization handling the Phase 3 portion of the FLEX Mitigating Strategies. This organization, SAFER, has prepared a White Paper addressing these nine considerations. This White Paper was formally transmitted to the NRC for endorsement on September 11, 2014, (ADAMS Accession No. ML14259A222), and endorsed by the NRC by letter dated September 26, 2014 (ADAMS Accession No.ML14265A107).

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 27 of 59 Ill. Licensee Identified Overall Integrated Plan (OIP) Open Item Responses Licensee Identified Open Item 1 Verify response times listed in timeline and perform staffing assessment DNC Response:

MPS3 BDB FLEX Validation for Time Sensitive Actions (TSA's), has been documented in Engineering Technical Evaluation ETE-CPR-2014-1008, Revision 1. This report documents the process that was used to reasonably assure required tasks, manual actions, and decisions for FLEX strategies are feasible and may be executed within the constraints identified in Engineering Technical Evaluation ETE-CPR-2012-0008 "Beyond Design Basis - FLEX Strategy Overall Integrated Plan Basis Document" and the Overall Integrated Plan (OIP) /Final Integrated Plan (FIP).

By letter dated June 12, 2014, DNC submitted the Phase 2 Staffing Assessment Report for MPS2 and MPS3. In a letter dated August 21, 2014, the NRC transmitted a request for additional information (RAI) related to the submittal. By letter dated September 22, 2014 (Reference 28) DNC provided a response to the RAI and notified the NRC that the final staffing assessment report would be submitted by July 30, 2015 in order to incorporate BDB operator training scheduled to be complete by April 30, 2015. By letter dated December 15, 2014, the NRC notified DNC that the staff review of the Phase 2 staffing study concluded that it adequately addresses the response strategies needed to respond to a BDBEE using DNC procedures and guidelines.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Licensee Identified Open Item 7 FSGs will be developed in accordance with PWROG guidance. Existing procedures will be revised as necessary to implement FSGs.

DNC Response:

FLEX Support Guidelines (FSGs) have been developed in accordance with PWROG guidance. Existing procedure ECA-0.0, "Loss of All AC Power ," has been revised to diagnose and declare an ELAP and implement FSGs. The MPS3 FSGs and associated Emergency Operating Procedures (EOPs) required for implementation of the MPS3 FLEX Mitigating Strategies have been approved and issued for use:

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 28 of 59 ECA-0.0, Loss of All AC Power FSG-1, Long Term RCS Inventory Control FSG-2, Alternate AFW Suction Source FSG-3, Alternate Low Pressure Feedwater FSG-4, ELAP DC Bus Load Shed / Management FSG-5, Initial Assessment and FLEX Equipment Staging FSG-6, Alternate DWST Makeup FSG-7, Loss of Vital Instrumentation or Control Power FSG-8, Alternate RCS Boration FSG-9, Low Decay Heat Temperature Control FSG-10, SI Accumulator Isolation FSG-1 1, Alternate SFP Makeup and Cooling FSG-12, Alternate Containment Cooling FSG-13, Transition From FLEX Equipment AOP-2583, Loss of All AC Power During Shutdown Conditions FSG-14, Shutdown RCS Makeup (new FSG from PWROG 6/23/14)

FSG-15, 4160V Repowering Using RRC Generator The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 29 of 59 Licensee Identified Open Item 11 Complete the evaluation of turbine driven (TD) auxiliary feedwater (AFW) pump long term operation with < 290 psig inlet steam pressure.

DNC Response:

TDAFW pump operation and adequate AFW flow to the SGs at SG pressures < 290 psig has been confirmed. Calculation 97-014, "MP3 AFW System, Determination of AFW Turbine/Pump Speed and AFW System Flow for Steam Generator Pressures of 185 psig, 600 psig, and 125 psig, and Determination of the Turbine Exhaust Pressure,"

April 2, 1997 Rev. 0, Change Notice No. 3 dated January 28, 2002 and Engineering Technical Evaluation ETE-MP-2013-1037, "MP3 Turbine Driven Aux Feedwater Pump Minimum Continuous Operating Speed," dated March 12, 2013, have previously been provided to the NRC staff and are available for their review.

Licensee Identified Open Item 18 Preferred travel pathways will be determined using the guidance contained in NEI 12-

06. The pathways will attempt to avoid areas with trees, power lines, and other potentialobstructions and will consider the potentialfor soil liquefaction.

DNC Response:

The Haul Route Evaluation is addressed in ISE CI 3.1.1.2.A.

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 30 of 59 IV. Audit Question Responses Audit Question I NEI 12-06 Section 5.3.2 consideration 1 requires that equipment deployment routes to be traveled should be reviewed for potential soil liquefaction that could impede equipment movement following a severe seismic event. Dominion did not provide a definite conclusion regarding the potential for liquefaction along deployment routes or if liquefaction was an issue at MPS2. Dominion identified that liquefaction may be a problem but no analysis was provided to evaluate this potential deployment issue.

Provide a discussion regarding the potential for seismic event liquefaction that clearly defines this deployment hazard for MSP3. (Reference Item 3.1.1.2.A)

DNC Response:

The Millstone Haul Route Evaluation has been completed and incorporated into Engineering Technical Evaluation ETE-CPR-2012-0008, Revision 4, Section 19.2. The evaluation concludes that there are no liquefaction concerns for the haul route and that adequate on-site equipment is available to clear anticipated debris material/structures.

The Haul Route Evaluation and specifics on liquefaction are addressed further in ISE Cl 3.1.1.2.A.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Audit Question 2 NEI 12-06 Section 5.3.3 Consideration 1 requires that seismically qualified electrical equipment can be affected by beyond-design-basisseismic events, therefore guidance should be available for determining instrument reading for both main control room (MCR) and non-control room readouts regarding how and where to measure key instrument readings (e.g. at containment penetrationsfor in-containmentsensors, using a portable instruments). Dominion's integratedplan did not include providing guidance for this situation. Provide a discussion of how plant staff will determine required key instrument readings if MCR instrumentationis not functioning following a seismic event.

(Reference Item 3.1.1.3.A)

DNC Response:

FLEX Support Guideline 7, "Loss Of Vital Instrumentation Or Control Power," has been developed to enable plant personnel to obtain instrument readings locally at the Containment penetrations. The guideline indicates the penetration number and cable contacts to be used to determine a parameter's value. Portable meters are used to produce a display, which can then be compared to a conversion chart included in the guideline to determine the converted parametric value of the readout. Key

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 31 of 59 instrumentation identified in NEI 12-06, Section 3.2.1.10, that is required to implement the FLEX strategies can be accessed using this method. The guideline includes conditions required to access the areas needed to get the readings and special tools and equipment required to take the readings.

During the July 2014 NRC Onsite Audit, a specific question was asked regarding the number of hand-held meters, the number of FSG procedure copies, and where both of these items would be stored. The following information responds to this question:

Fluke Meters (or equivalent) - A total of six dedicated BDB meters will be available. There are two meters with extra batteries stored in each Control Room. There are also two spare meters with extra batteries stored in the BDB Storage Building.

" Procedure Storage- Hard copies (and quantity) of FSGs to be maintained as follow:

MPS2 MPS3 Shift Manager's Office (1) Shift Manager's Office (1)

Unit Supervisor's Desk (1) Unit Supervisor's Desk (1)

Control Room File Cabinet (10) Control Room File Cabinet(10)

Tech Support Center (1) Tech Support Center (1)

EOF (1) EOF (1)

BDB Storage Building (5) BDB Storage Building (5)

In addition, sets of BDB Distribution drawings printed on large paper will be kept in the drawing desk in each Control Room, the TSC and EOF for ease in describing detail during briefs.

Audit Question 3 NEI 12-06 Section 5.3.3 consideration 2 and 3 require providing guidance regarding seismic hazards related to large internal flooding sources that are not seismically robust and do not require ac power, and the use of ac power to mitigate ground water in critical locations. Dominion did provide information regarding these issues in the integrated plan. Provide a discussion regarding of the need for any guidance needed to deal with potential large internal flooding sources and the potential need for ac power to mitigate ground water intrusion. (Reference Item 3.1.1.3.B)

DNC Response:

The review for internal flooding sources that could result from seismic induced failures and engine-driven or gravity-drain water sources has been completed. The results of

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 32 of 59 this review are included in Engineering Technical Evaluation ETE-CPR-2012-0008 and response to ISE CI 3.1.1.3.A.

MPS3 does not have a permanent safety-related groundwater removal system installed.

However, the ESF building does have a sump to control groundwater in-leakage. This is discussed in ETE-CPR-2012-0008, Section 11.1.3.3. and response to ISE Cl 3.1.1.3.A.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Audit Question 10 NEI 12-06, Section 9.3.2 states that the FLEX equipment should be transported to different locations, even during the extreme conditions applicable to the site. The potentialimpact of high or low temperatures on the storage of equipment should also be considered, e.g., expansion of sheet metal, swollen door seals, etc. Although Dominion addressed accessibility issues regarding loss of power to normal access points, no information was provided regardingany plans for managing access through doors/gates under extreme high temperature conditions. Provide a discussion regarding deployment of FLEX equipment considering the potential effects of extreme high temperatureson accessibilitynoted above. (Reference Item 3.1.5.2.A)

DNC Response:

The BDB Storage Building was designed considering the potential impact of high temperatures on the equipment and equipment access doors consistent with NEI 12-06 Section 9.3.2. Both of the large steel equipment access doors are each constructed of two halves that open to facilitate deployment of BDB equipment from the BDB Storage Building. These equipment access doors are not subject to binding in extreme temperature conditions. Specifically, each equipment access door has sufficient gap between the two closed halves (approximately four inches) such that binding due to thermal expansion is not credible. The gap is covered by a plate attached to the outside of one of the door halves.

Plant doors that are needed for access to implement FLEX strategies are also not expected to be significantly impacted by extreme heat conditions. Tolerances on roll-up doors are generally not tight and personnel doors generally have sufficient gaps to accommodate thermal expansion. In the event that an access door may be jammed, sufficient tools will be available in the BDB Storage Building or in the plant shops to free or remove a door to allow access.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 33 of 59 Audit Question 16 Section 4.4.1 of WCAP-17601-P states, in part, that, "The NRC Information Notice (IN) 2005-14 has accepted the use of a 21 gpm assumption in deterministic analyses to develop coping analyses to show compliance with Appendix R. Given that the 50.63 SBO transient is similar with regard to seal performance, the 21 gpm should also be acceptable for developing ELAP strategies;this has not been called into question by the NRC in inspections (e.g., Component Design Basis Inspections)."

It is stated in IN 2005-14 that, "Forthe Westinghouse RCP seals, as discussed in a recently submitted document on RCP seal performance, a leakage rate of 21 gpm per RCP may be assumed in the licensee's safe shutdown assessment following the loss of all RCP seal cooling. Assumed leakage rates greaterthan 21 gpm are only warrantedif the increase seal leakage is postulated as a result of deviations from seal vendor recommendations."

It is also stated in IN 2005-14 that, "Even if seal cooling is not reestablished, degradation of the seals for leakage rate to significantly increase is not expected for an indefinite period of time if the RCPs are secured before the seal temperature exceeds 235 degrees F. Restoration of seal cooling may result in cold thermal shock of the seal and possibly cause increased seal leakage." Address the applicability of the above statements from IN 2005-14 to the ELAP analysis. (Reference Item 3.2.1.2.D)

DNC Response:

Back-ground Discussion Westinghouse issued Nuclear Safety Advisory Letter NSAL-14-1, Impact of Reactor Coolant Pump Seal Leakoff Piping on Reactor Coolant Pump Seal Leakage During a Loss of All Seal Cooling, in February of 2014. Westinghouse had determined that the number 1 seal leakage rate calculated in Westinghouse Report WCAP-10541, Rev. 2, "Westinghouse Owner's Group Report Reactor Coolant Pump Seal Performance Following a Loss of All AC Power," November 1986, was based on what was considered to be a representative leak-off line layout. The calculated value, 21 gpm at normal reactor coolant system (RCS) operating temperature and pressure, was then used in analyses to support station blackout (SBO), Appendix R, probabilistic risk assessment (PRA), and extended loss of all ac power (ELAP per WCAP-17601-P).

More recent investigations determined that the leak-off line configuration used in WCAP-10541 is not applicable for all plants. Revised flow rates for normal RCS conditions, as well as flow rates for lower pressures and temperatures, were needed to determine the impact on other analyses (e.g., to support industry FLEX strategies for responding to an ELAP).

As a result, Westinghouse performed extensive thermal hydraulic analyses of seal performance for a variety of seal leakoff line configurations. This was done by

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 34 of 59 surveying the industry and "binning" plants into six groups and analyzing a configuration considered bounding for each group. These analyses are documented in Westinghouse Report PWROG-14015-P, Rev 1, No. 1 Seal Flow Rate for Westinghouse Reactor Coolant Pumps Following Loss of All AC Power, Task 2:

Determine Seal Flow Rates, September 2014 (Westinghouse Proprietary Class 2). The MPS3 configuration was determined to be bounded by the Category 1 configuration (2-inch piping, 0.254 in flow element bore). Pressure and temperature-dependent flows were calculated. The results showed seal leakage values ranging from a maximum of 17.5 gpm to a minimum of 0.7 gpm, depending on the upstream RCS condition.

DNC used the results from PEROG-14015-P to evaluate the performance of MPS3 with respect to the relevant criteria for ELAP (e.g., adequate RCS loop flow for boron mixing, avoidance of reflux cooling) in Engineering Technical Evaluation ETE-NAF-2012-0150, Revision 2, "Evaluation of Core Cooling Coping for Extended Loss of AC Power (ELAP) and Proposed Input for Dominion's Response to NRC Order EA-12-049 for Dominion Fleet," dated September 25, 2014. ETE-NAF-2012-0150 concluded that deployment of the FLEX RCS makeup capability by no later than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months into an ELAP event (the latest expected time of deployment) ensures that these criteria are met. The evaluation was based on the MPS3 configuration of two Westinghouse seals and two Flowserve N-9000 low leakage seals.

Evaluation of Statements in IN-2005-14 "Forthe Westinghouse RCP seals, as discussed in a recently submitted document on RCP seal performance, a leakage rate of 21 gpm per RCP may be assumed in the licensee's safe shutdown assessment following the loss of all RCP seal cooling.

Assumed leakage rates greater than 21 gpm are only warranted if the increase seal leakage is postulated as a result of deviations from seal vendor recommendations."

As stated in the Background Discussion, the 21 gpm seal leakage applied to safe shutdown analyses in the Information Notice (IN) has been superseded by more detailed plant-specific results obtained in PWROG-14015-P, Rev 1 and applied to MPS3 in ETE-NAF-2012-0150.

"Even if seal cooling is not reestablished,degradation of the seals for leakage rate to significantly increase is not expected for an indefinite period of time if the RCPs are secured before the seal temperature exceeds 235 degrees F.

For the ELAP event, this condition will be met as power to the RCP's is terminated coincident with the loss of seal cooling.

"Restorationof seal cooling may result in cold thermal shock of the seal and possibly cause increasedseal leakage."

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 35 of 59 During the onsite audit, it was noted that seal cooling would be isolated and would not be restored in order to avoid thermally shocking the seals.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Audit Question 25 Specify the required cooldown completion time that is supportable by adequate analysis. Discuss the required action to complete the cooldown and justify that the all the required actions can be accomplished within the completion time. (Reference Item 3.2.1.6.F)

DNC Response:

The reference analysis presented in Section 5.2.1 of WCAP-17601-P is applicable to MPS3 as stated in Engineering Technical Evaluation ETE-NAF-2012-0150. The analysis models a cooldown starting at two hours after initiation of the event and a cooldown rate of -7 0 °F/hr. The cooldown is terminated when secondary steam pressure reaches the ECA-0.0 target of -300 psia. This occurs at approximately four hours after initiation of the event, when the RCS average temperature reaches -425°F.

This is directly comparable to the MPS3 case. The ECA-target SG pressure for termination of the cooldown is 290 psig or approximately 305 psia. The ECA guidance specifies cooldown at a rate not to exceed 100°F per hour.

Two hours is considered more than adequate time to dispatch an operator to the Main Steam Valve Building (MSVB), establish communication with the control room and initiate the cooldown. Personnel will be able to access the MSVB and initiate cooldown prior to two hours. Steam release from the SGs will be controlled locally within the MSVB.

Initiation of the cooldown at a time earlier than two hours would not invalidate the conclusions of the generic analysis. This would result in more rapid depressurization of the RCS and less leakage through the RCP seals. As stated in ETE-NAF-2012-0150, the commencement and termination times of the cooldown have no impact on the core cooling coping time.

The MPS3 deployment of a BDB RCS Injection pump for RCS make-up is within 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , which is well before the transition to reflux condensation mode of cooling at 33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months in the reference case (see WCAP-17601-P, Table 5.2.2-1). The 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months allows movement of the pump from the storage building to the location within the protected area using augmented staff and after debris removal activities are completed. The

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 36 of 59 commencement is expected at approximately 10 - 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months and completed prior to 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

During the July 2014 BDB Onsite Audit, the conditions in the MSVB/ADV area (coincident with the timing of the required operator actions for plant cooldown) were reviewed. A modeling discrepancy was identified and has subsequently been corrected (addressed in response to ISE Cl 3.2.4.2.A). Although the revised evaluation increased the temperature in the area of the ADVs, the resulting temperature is still within the capability of the required personnel protection gear (Engineering Technical Evaluation ETE-MP-2014-1030). Thus, the operators are able to make the necessary connections for the use of air bottles to facilitate using the ADVs or use the handwheels installed on the motor-operated ADBVs in the MSVB room to cooldown the plant.

Subsequent to the onsite audit, ETE-NAF-2012-0150, Revision 2, has been issued to provide the latest assessment of core cooling based on the latest seal leakage data available. ETE-MP-2014-1030 and ETE-NAF-2012-0150, Revision 2 have previously been provided to the NRC staff and are available for their review.

Audit Question 37 NEI 12-06, Section 3.2.1.7 and JLD-ISG-2012-01, Section 2.1, requires strategies that have a time constraint to be successful should be identified and a basis provided that the time can be reasonably met. NEI 12-06, Table 3-2 and Appendix D provide some examples of acceptable approaches for demonstrating the baseline capability of the containment strategies to effectively maintain containment functions during all phases of an ELAP. One of these acceptable approachesis by analysis using a computer code.

Dominion's plans for maintaining containment integrity are under development under open item 4. Dominion provided evaluations and calculations that show no strategies are required in Phase 1 or 2 to maintain containment temperature and pressure below design limits and that key parameter instruments subject to the containment environment will remain functional for at least 7 days. Dominion did not provide any supporting details regarding actual containment pressures and temperatures to be experienced during the ELAP based on these calculations. Dominion provided a reference "Dominion Nuclear Engineering Calculation MISC-11793" that noted the GOTHIC computer code was used in the analysis. When complete, provide a discussion and the supporting details regarding actual containment pressures and temperatures vs. time functions to be experienced during the ELAP based on these calculations,to validate the assumptions that no strategiesare requiredfor Phase I and 2 to maintain containment functions. (Reference Item 3.2.3.A)

DNC Response:

As indicated in section 5.1.2 of Engineering Technical Evaluation ETE-CPR-2012-0008, the long term containment pressure and temperature analysis for MPS3 has been

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 37 of 59 documented in calculation MISC-11793. Pages 23 and 24 of the calculation document the long term pressure and temperature profiles for the initial seven days of the post ELAP scenario. As documented on page 16 of the calculation, at the end of seven days, the MPS3 containment pressure and temperature are calculated to be 28.46 psia and 203.1 F, respectively. This is well below the containment design pressure and temperature limits of 45 psig and 260 0 F, respectively.

The containment response analysis has been performed utilizing the same approved GOTHIC licensing model and methodology that was used for FSAR Chapter 6 containment integrity analysis. The DNC containment analysis methodology is documented in topical report DOM-NAF-3-0.0-P-A. This topical report describes the assumptions to be used and the mathematical formulations employed for containment integrity analysis for the Dominion/DNC fleet. The NRC has approved the use of the GOTHIC code and the analysis methodology described in this topical report in a letter dated August 30, 2006.

Licensee Identified Open Item #4 was to provide the containment cooldown and depressurization strategy when it was finalized. This strategy was addressed in of the MPS3 Six-Month Status Update letter dated February 27, 2014.

Additionally, Section 5.3 of ETE-CPR-2012-0008, Rev. 4, provides more detail regarding the various containment cooling strategy options.. Both ETE-CPR-2012-0008, Rev. 4 and MISC-11793 have previously been provided to the NRC staff and are available for their review.

Audit Question 39 NEI 12-06, Section 3.2.2, Paragraph(3) provides that plant procedures/guidanceshould specify actions necessary to assure that equipment functionality can be maintained (including support systems or alternate method) in an ELAP/[LNUHS] or can perform without ac power or normal access to the UHS. Cooling functions provided by such systems as auxiliary building cooling water, service water, or component cooling water may normally be used in order for equipment to perform their function. Dominion did not provide sufficient information regarding these cooling functions when AC power is lost during the ELAP for Phase I and 2. For example, the potential need for cooling water for the TDAFW pump bearings was not discussed. Provide a discussion regarding the need for cooling functions for systems affected by the ELAP. (Reference Item 3.2.4. 1.A)

DNC Response:

The MPS3 Turbine Driven Auxiliary Feedwater (TDAFW) pump bearings are cooled by circulating auxiliary feedwater (i.e., pumpage from the TDAFW pump) fluid through the TDAFW pump oil cooler, and the pump does not rely on cooling support systems. Other

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 38 of 59 than general room ventilation requirements which have been addressed separately, FLEX-credited plant equipment does not rely on the cooling functions provided by cooling support systems in order to provide either their Phase 1 or Phase 2 FLEX functions.

Audit Question 41 Regarding NRC Question 40, the NRC's has identified the following issues regarding habitabilityof the MCR during the ELAP. Without ventilation the MCR would most likely heat up. If temperaturesapproach a steady-state condition of I 100 F, the environmental conditions within the main control room would remain at the uppermost habitability temperature limit defined in NUMARC 87-00 for efficient human performance.

NUMARC 87-00 provides the technical basis for this habitabilitystandardas MIL-STD-1472C, which concludes that I 10°F is tolerable for light work for a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period while dressed in conventional clothing with a relative humidity of -30%. Add the effect of higher humidity at same or even lower temperature because higher humidity is more oppressive. Include a discussion of the above considerations regarding any expected high temperaturesaffecting MCR habitabilityin the appropriateupdate to the integrated plan. (Reference Item 3.2.4.2.B)

DNC Response:

The Control Room Complex habitability analysis has been documented in the MPS3 FLEX basis document (Engineering Technical Evaluation ETE-CPR-2012-0008 Rev. 4, Section 10.4.2). As discussed in Sub Section 10.4.2.2.1, Calculation MISC-11802, Rev.

0, demonstrated that no compensatory cooling measures are required to maintain the area temperatures below the habitability limit during all phases of the ELAP scenario.

The habitability and equipment operability limits for the MPS3 Control Room are discussed in Section 10.4.2.1 of ETE-CPR-2012-0008, Revision 4. The Control Room habitability limit is 1 10°F and the equipment operability limit is 120 0 F. These limits are consistent with NUMARC 87-00, Revision 1, "Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors."

Specifically, the results of a detailed GOTHIC analyses which utilized both the site maximum outdoor design temperature of 86 0F and the maximum outdoor statistical summer temperature of 93°F, indicate that the temperatures in the MCR and the Instrument Rack Room (IRR) will remain well below the acceptance criteria for both habitability and equipment operability limits (Refer to Table 1 below).

In regards to MCR, the relative humidity is low (in a controlled environment) prior to the ELAP event and there is no reason for the humidity to abruptly increase. DNC considers the habitability aspects of humidity to be a secondary and long-term consideration

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 39 of 59 which, if needed, can be alleviated by establishing natural or forced ventilation or the use of other portable cooling equipment in the MCR.

For defense in depth, FSG-5 includes guidance to monitor area temperatures and to implement contingency compensatory cooling measures, such as opening doors or placement of fans, if area temperatures continue to indicate increasing trends.

Table 1 - MISC-1 1802, Rev 0 Calculation Results Case Description Max. Steady State Temp (10 Days,0 F)

Control Room / Instrument Rack room All MCR Doors Remain Closed (Outside Air at 860F) -98 / -95 All MCR Doors Remain Closed (Outside Air at 93 0 F) -104 1 -101 The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Audit Question 43 NEI 12-06, Section 3.2.2, Paragraph (12) provides that: plant procedures/guidance should consider loss of heat tracing effects for equipment required to cope with an ELAP. Alternate steps, if needed, should be identified to supplement planned action. In the integratedplan Dominion did not discuss the effects of loss of power to heat tracing.

Provide a discussion and analysis of the effects of the loss of heat tracing for equipment required to cope with an ELAP. (Reference Item 3.2.4.3.A)

DNC Response:

Heat trace is used to provide two protection functions:

- Heat trace is used to maintain highly concentrated soluble boron solutions above the temperature where the soluble boron will precipitate out of solution.

- Heat trace is also used to protect piping systems and components from freezing in extreme cold weather conditions.

FLEX strategies developed do not depend on highly concentrated soluble boron solutions. FLEX strategies developed use borated water sources with boron concentrations below 4000 PPM. At these levels boron precipitation is not expected to occur.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 40 of 59 FLEX strategies have been developed to protect piping systems and components from freezing. Commercially available Heat Tape and insulation rolls have been procured and are maintained in the BDB Storage Building for use on piping systems and component, both installed and portable, that will be used during an ELAP event where freezing is a concern in extreme cold weather conditions. The heat tape and heat tape insulation are listed in the table presented in Section 16.1.3, "Table of BDB Equipment (Complete Inventory List)" in Engineering Technical Evaluation ETE-CPR-2012-0008, Revision 4.

In addition, major components being procured for FLEX strategies have been provided with cold weather packages and small electrical generators to power the heat tape circuits as well as protect the equipment from damage due to extreme cold weather and help assure equipment reliability.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Audit Question 54 Discuss Dominion's position on each of the recommendations discussed in Section 3.1 of WCAP-17601-P for developing the FLEX mitigation strategies. List the recommendations that are applicable to the plant, provide rationale for the applicability, address how the applicable recommendations are considered in the ELAP coping analysis, and discuss the plan to implement the recommendations. Also, provide rationale for each of the recommendations that are determined to be not applicable to the plant.

DNC Response:

The following verbiage is taken from WCAP-17601-P for the recommendation associated with each objective:

Obiective #1 Develop a Reference Case which assumes standard RCP seal packages to determine the minimum adequate core cooling time with respect to inventory.

Recommendation With a time of 55 hours6.365741e-4 days 0.0153 hours 9.093915e-5 weeks 2.09275e-5 months , it is possible that even without low-leakage or safe shut-down seals, nearer term on-site make-up capability is not required with regard to RCS inventory. In addition, current ECA-0.0 guidance instructs the operations staff to perform a cooldown of the RCS to reduce RCP seal leakage. This step appears in the up-front portion of the guideline. It is recommended that this step be preserved for a

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 41 of 59 host of reasons beyond the idea of reducing RCS leakage from the RCP seals. That is, regardless of the RCP seal package configuration and its leakage characteristics under ELAP conditions, a cooldown of the RCS should be performed when time and resources permit.

Response

Operators are directed to initiate a RCS cooldown using the installed SG atmospheric dump valves (ADVs) and the installed TDAFW at a rate not to exceed 100°F/hr until the pressure in each intact SG reaches 290 psig. Cooling the reactor at that rate to the prescribed minimum target SG pressure (290 psig) minimizes RCS leakage.

Reactor sub-criticality is procedurally monitored during the cooldown and the cooldown is halted if required. Boron additions are made as required to ensure the unit is sub-critical. Sub-criticality (shutdown margin) is monitored using excore nuclear instruments.

Obiective #2 Develop inventory coping times beyond the Reference Case Recommendation Some RCP models may show a significant benefit relative to RCS leakage, which may indicate in those cases that a SDS (safe shut-down)/Iow-leakage design may be required. However, there are many advantages to SDS/low leakage seals, such as Appendix R/NFPA-0805 scenarios that make them warranted. In addition, limiting inventory loss in the ELAP event reduces complications that may burden the operators and /or technical support staff, such as additional impacts on containment heat-up.

Response

MPS3 replaced two Westinghouse RCP seals in 2014 with Flowserve N-seals. The remaining two Westinghouse seals are currently scheduled to be replaced with Flowserve N-seals during the spring 2016 outage. DNC has evaluated the post-ELAP performance of MPS3 with the current configuration of 2 Westinghouse and two Flowserve N-seals in Engineering Technical Evaluation ETE-NAF-2012-0150, Revision 2, "Evaluation of Core Cooling Coping for Extended Loss of AC Power (ELAP) and Proposed Input for Dominion's Response to NRC Order EA-12-049 for Dominion Fleet,"

dated September 25, 2014. The evaluation included the effects of updated Westinghouse leakages considered bounding for the MPS3 seal leakoff line configuration, as documented in Westinghouse Report PWROG-14015-P, Rev 1, No. 1 Seal Flow Rate for Westinghouse Reactor Coolant Pumps Following Loss of All AC Power, Task 2: Determine Seal Flow Rates, September 2014 (Westinghouse

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 42 of 59 Proprietary Class 2). The evaluation in ETE-NAF-2012-0150 showed that, with no makeup, the criteria for ensuring adequate RCS loop flow to provide boron mixing and avoidance of reflux cooling are not violated for first -24 hours of the event. Therefore, the Phase 1 strategy for RCS makeup for the first 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months after the initiation of the ELAP/LUHS event is to rely on the RCP seals to maintain adequate RCS water inventory.

Within 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months after the start of an ELAP/LUHS event, a portable diesel driven BDB RCS Injection pump will be deployed from the on-site BDB Storage Building and positioned for delivery of RCS inventory makeup from the RWST or another borated suction source for the remainder of the event.

Obiective #3 Develop high-level list of instrumentation for the RCS in order to confirm/maintain adequate core cooling Recommendation This list forms the basis for additional instrument recommendations which will be provided through the PSC-0965 project.

Response

Instrumentation includes:

AFW flow

" SG level for each SG (wide range and narrow range)

" SG pressure for each SG

" RCS Hot Leg Temperature

" RCS Cold Leg Temperature Indication of AFW flow, SG level and SG pressure will provide the necessary information for manually controlling the TDAFW pump / AFW system for proper AFW delivery to the SGs. Indication of RCS temperature will be used to control steam release from the SGs.

Obiective #4 Study reactor sub-critical aspects under ELAP conditions

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 43 of 59 Recommendation For maintaining a subcritical condition in the reactor core, for the Westinghouse plants, it is recommended that a set of curves be developed on a plant-specific basis based on realistic conditions. It is assumed that all control rods insert, which is consistent with NEI 12-06 guidance.

Response

Operators are directed to initiate a RCS cooldown using the installed SG atmospheric relief bypass valves at a rate not to exceed 100°F/hr until the pressure in each intact SG reaches 290 psig. Reactor sub-criticality is procedurally monitored during the cooldown and the cooldown is halted if required. Boron additions are made as required to ensure the unit is sub-critical. Sub-criticality (shutdown margin) is monitored using excore nuclear instruments.

DNC has performed detailed studies for recent MPS3 cores which develop the required boron concentration to maintain K-effective < 0.99 as a function of time after trip and RCS temperature (Steam Generator Pressure). These studies form the basis for the guidance in FLEX Support Guideline FSG-8, Alternate RCS Boration. DNC has also developed core reload checks to perform ongoing validation of the procedural guidance for new cores.

Obiective #5 Investigate various aspects of RCS make-up with regard to maintaining core cooling and a sub-critical state in the reactor core.

Recommendation Because evaluation indicates that RCS makeup with regard to inventory will not be of immediate concern, and that many plants have, or will have, low-leakage RCP seals, boration of the RCS becomes the primary driver for RCS makeup needs. An RCS makeup pump that can supply 40 gpm at 1500 psia should be sufficient. That would meet the RCS inventory makeup need and the pressure is high enough to allow for accommodation of voiding should saturation in the head or other portions of the RCS be reached. This can allow effective boration without the need for significant cooldown/depressurization when a means of letdown is available. More analysis work may be required to ensure the adequacy of the pump size and performance characteristics.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 44 of 59

Response

For MPS3, the required delivery pressure for the RCS Injection pump is approximately the saturation pressure of 561°F (1143 psia) plus 100 psi or 1243 psia. The RCS Injection pump is capable of delivering at least 40 gpm at a discharge pressure of at least 2,000 psig. Hydraulic analysis of the BDB RCS Injection pump with the associated hoses and installed piping systems confirms appropriate system pressure distribution.

Deployment of the BDB RCS Injection pump prior to 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months meets the requirements with substantial margin. Two RCS Injection pumps will be available on-site to provide batch injection.

Evaluations of sub-criticality shows that, because of post trip Xenon buildup and decay, for the limiting EOC case, assuming cooldown to the steam generator pressure of 290 psig, boration is not needed before about 25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months . Deployment of BDB RCS Injection pump for makeup and boration will begin at approximately 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months so that injection can be initiated by 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months ; therefore, adequate margin will be maintained. The expected flow from the BDB RCS Injection pump, approximately 45 gpm, will provide additional Shutdown Margin.

The RWST, if available, has a usable volume of approximately 1,000,000 gallons of borated water at a concentration between 2700 and 2900 ppm.

A 1000 gallon portable tank for batching boron is also available as an alternate suction supply of borated water for RCS makeup. The portable tank will be deployed to a location adjacent to the BDB RCS injection pump.

Obiective #6 Quantify RCS response with SDS/Iow-leakage seals.

Recommendation The use of the SDS or low-leakage seals reduces the complexity of the transient, and eases operator burden.

Response

MPS3 had two Westinghouse RCP seals replaced with Flowserve N-seals in 2014. The remaining two Westinghouse seals are currently scheduled to be replaced with Flowserve N-seals during the spring 2016 outage. DNC has evaluated the post-ELAP performance of MPS3 with the current configuration of two Westinghouse and two Flowserve N-seals in ETE-NAF-2012-0150, Revision 2, "Evaluation of Core Cooling Coping for Extended Loss of AC Power (ELAP) and Proposed Input for Dominion's

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 45 of 59 Response to NRC Order EA-12-049 for Dominion Fleet," dated September 25, 2014.

The evaluation included the effects of updated Westinghouse leakages considered bounding for the MPS3 seal leakoff line configuration, as documented in Westinghouse Report PWROG-14015-P, Rev 1, No. 1 Seal Flow Rate for Westinghouse Reactor Coolant Pumps Following Loss of All AC Power, Task 2: Determine Seal Flow Rates, September 2014 (Westinghouse Proprietary Class 2). The evaluation in ETE-NAF-2012-0150 showed that, with no makeup, the criteria for ensuring adequate RCS loop flow to provide boron mixing and avoidance of reflux cooling are not violated for first approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the event. Therefore, the Phase 1 strategy for RCS makeup for the first 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months after the initiation of the ELAP/LUHS event is to rely on the RCP seals to maintain adequate RCS water inventory.

Within 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months after the start of an ELAP/LUHS event, a portable diesel driven BDB RCS Injection pump will be deployed from the on-site BDB Storage Building and positioned for delivery of RCS inventory makeup from the RWST or another borated suction source for the remainder of the event.

Obiective #7 Quantify (on a rough order) what acceptable amount of time exists for feedwater flow interruption early in the transient.

Recommendation There is some tolerance to feedwater flow interruptions as the transient progresses but that time frame is not unlimited. The PSC-0965 guidance development may need to consider the prioritization of pre-staging a FLEX strategy for alternative feedwater additions when time and resources permit.

Response

DNC has performed calculations to estimate the time to steam generator dryout following a flow interruption for several conditions:

For failure of the TDAFWP to start at the initiation of the event, time to steam generator dryout is approximately one hour. This is considered sufficient time to deploy personnel to locally start the pump. For a flow interruption following depletion of the Demineralized Water Storage Tank (DWST) at 22.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , about seven hours are available before SG dryout to deploy an alternate supply. These considerations have been incorporated into the FLEX support guidelines.

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 46 of 59 Objective #8 Develop more proof of concept of feeding a single steam generator with a low-pressure portable pump. Also, develop high-level functional requirements for the feed pump.

Recommendation In summary, it is recommended (as a start) that the portable feedwater system be capable of delivering 300 gpm at 300 psig at the SG feedring or AFW inlet in those SG's which introduce AFW in the lower portions of the tube bundle. Precautions must be taken with regard to asymmetric RCS temperatures, loop stagnation, and cooldown in order to preserve RCP seal components and maintain sub-critical margins. It should be noted that in some cases these requirements (300 gpm at 300 psig at the SG feedring) may vary from one plant to another. Project PSC-0965 will issue more formal requirements for consideration with regard to this.

Response

Heat removal from the RCS is accomplished by supplying feed water from the demineralized water storage tank (DWST) to the SGs using the TDAFW pump. Feed rate can be controlled by local manual operation of the SG feed line manual operated isolation valves. Throttling of the feed flow within approximately 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months will avoid overfilling the steam generators at the initiation of the ELAP.

The Phase 2 strategy for core heat removal is to indefinitely extend AFW suction supply by deploying the portable diesel driven BDB High Capacity pump within 22.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months . For MPS3, the BDB AFW pump has been designed to deliver a minimum of 300 gpm to the SGs at a pressure of 300 psig at the feedring. This flow rate is adequate to establish adequate core cooling in the event the TDAFW pump is unavailable.

Capability exists to monitor levels and flow rates for the steam generators individually.

Obiective #9 Quantification of accumulator makeup capability and isolation/venting to prevent cover gas injection.

Recommendation Evaluate the following strategy in PA-PSC-0965:

Align ELAP mitigation equipment as soon as practical for accumulator isolation or venting. This is in anticipation of boration or the need to cool down below the current ERG SG pressure setpoint to possibly mitigate an inadequate core cooling situation. In

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 47 of 59 the latter, should a means of venting or isolation not exist at that time, it is recommended that the additional cooldown be performed as the consequences of the nitrogen injection are second order compared to the eminent need to return the core to a stable temperature.

For boration purposes, using accumulator level, determine the amount of mass that has been delivered to the RCS. Note that reactor vessel head indicating system (RVLIS) and pressurizer level are not necessarily required here since an assumed full RCS will always dictate the highest volume to be injected to yield a particular boron concentration. When accumulator inventory, as monitored via level, indicates sufficient mass to be injected, no further action is required. If further injection is determined to be necessary, perform a plant cooldown to 100 psia on the SG's and use the upper head vent to deliver the desired amount of inventory as determined for the shutdown reactivity margin curves described in Section 5.8.1. Accumulator level should be monitored as a continuous step and isolation or venting should be executed when the appropriate low-low level is reached.

Response

In accordance with ECA-0.0, maintaining steam generator pressure above 190 psig will preclude the injection of the nitrogen cover gas into the RCS. The controlling setpoint for MPS3 procedures is 290 psig to account for instrument uncertainties and containment temperature increases.

In preparation for further cooldown, procedural guidance provides for either isolating or venting the accumulators to preclude nitrogen injection into the RCS. FLEX Support Guideline FSG-10 provides direction to deploy equipment that is needed to re-energize accumulator "A" and "C" isolation valves so that they can be closed. FSG-10 also provides direction for venting the accumulators. The associated provision of electrical power to those isolation valves is included in the 480-volt recovery plan.

FLEX Support Guideline FSG-8 provides direction for using the reactor vessel head vent for RCS depressurization that may be needed to accommodate additional boration.

Objective #10 Evaluate and quantify the effects of TDAFW pump heat load and ambient heat loss on the ability to maintain the Nuclear Steam Supply Systems (NSSS) at normal operating temperature.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 48 of 59 Recommendation The PSC-0965 guideline development will need to consider the prioritization of staging portable equipment that may be required to isolate/vent the accumulators when certain cooldown maneuvers are necessitated for other reasons.

Response

FLEX Support Guideline FSG-10 verifies the deployment of the BDB 480 VAC DG, via FSG-4, to re-energize accumulator "A" and "C" isolation valves so that the valves can be closed. The 480 VAC DG is the only staged equipment needed to re-energize these valves. FSG-1 0 also provides direction for venting the accumulators.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Audit Question 73 The Order requires mitigating beyond-design-basisexternal events. NEI 12-06, section 4.2 indicates characterization of a hazard for a site includes the functional threats caused by the hazard, e.g., equipment that may be inundated. As part of the discussion of external flooding, the licensee states that seiche-related flooding is not addressed in the FSAR. The Overall Integrated Plan does not discuss the possibility of flooding from a seiche. The licensee is requested to discuss why a beyond-design-basis external event such as a seiche is not possible on Long Island Sound. If a seiche is possible, please discuss why the licensee does not consider a seiche as a beyond-design-basis external event applicable to MPS3.

DNC Response:

As stated, the MPS3 seiche is not specifically addressed in the FSAR. However, the MPS3 FSAR does include seiche conditions in the evaluation of storm surge, but states that the Probable Maximum Hurricane (PMH) surge is the more significant flooding event at the MPS site. Although this statement is made for MPS3, it is applicable to both units at the MPS site.

Based on the above, DNC does not consider a seiche as a beyond-design-basis external event applicable to MPS3.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 49 of 59 Audit Question 84 Generic Open Item: The licensees' plans for equipment maintenance and testing which endorses the EPRI industry program for maintenance which is currently under development does not provide reasonable assurance that guidance and strategies developed and implemented under them will conform to the guidance of NEI 12-06, Section 11.5 with respect to maintenance and testing. Please provide details of the EPRI industry program for maintenance and testing of FLEX electrical equipment such as batteries, cables, and diesel generators.

DNC Response:

NEI 12-06 "Diverse and Flexible Coping Strategies (FLEX) Implementation Guide" Section 11.5 requires in part:

"Portable equipment that directly performs a FLEX mitigation strategy for the core, containment, or SFP should be subject to maintenance and testing guidance provided in INPO AP 913, Equipment Reliability Process, to verify proper function. The maintenance program should ensure that the FLEX equipment reliability is being achieved. Standard industry templates (e.g., EPRI) and associated bases will be developed to define specific maintenance and testing .... "

EPRI has completed and has issued "Preventive Maintenance Basis for FLEX Equipment-Project Overview Report" (Report 3002000623). Preventative Maintenance Templates for FLEX equipment have also been developed.

Typical PM task lists that have been developed are listed below:

- Periodic Static Inspections - Monthly walkdown

- Fluid analysis (Yearly)

- Periodic operational verifications - Quarterly starts

- Periodic functional verifications with performance tests - Annual 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months run with pump flow and head verifications The EPRI PM Templates for FLEX equipment conform to the guidance of NEI 12-06 providing assurance the FLEX equipment is being properly maintained and tested.

EPRI Templates are used for most equipment. However, in the event EPRI templates are not available / applicable to a piece of equipment, Preventative Maintenance (PM) actions are developed based on manufacturer provided information/ recommendations.

Site procedural implementation for the EPRI maintenance and testing guidance in cludes the following:

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 50 of 59 Fleet-wide templates have been developed and input into the site maintenance work management systems. Specific Preventative Maintenance (PM) work orders based on these strategies will be implemented prior to the required FLEX equipment maintenance due date. FLEX equipment will be maintained through vendor service contracts. As such, maintenance checklists, created in lieu of maintenance procedures, will be attached to the PM work order for specific FLEX equipment. Quarterly runs of FLEX equipment as recommended in the EPRI templates have been changed to semi-annual runs with adequate basis.

- Operations three year functional test procedures have been developed for FLEX equipment.

A fleet-wide FLEX Strategy Program Document has been developed (CM-AA-BDB-10). The program includes the requirement to manage unavailability of equipment such that risk to mitigating strategy capability is minimized in accordance with the requirements of NEI 12-06, Section 11.5. A fleet-wide procedure has been developed to specifically address equipment unavailability (CM-AA-BDB-102). The two referenced procedures have previously been provided to the NRC staff and are available for their review.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 51 of 59 V. Additional Safety Evaluation (SE) Review Item Responses Safety Evaluation Review Item 2 On February 10, 2014, Westinghouse issued Nuclear Safety Advisory Letter (NSAL) 1, which informed licensees of plants with standard Westinghouse RCP seals that 21 gpm may not be a conservative leakage rate for ELAP analysis. This value had been previously used in the ELAP analysis referenced by many Westinghouse PWRs, including the generic reference analysis in WCAP-17601-P. Therefore, please provide the following information:

a. Clarify whether the assumption of 21 gpm of seal leakage per RCP (at 550 degrees F, 2250 psia) remains valid in light of the issues identified in NSAL-14-1.

b. Identify the corresponding leakage rate from NSAL-14-1 or other associated documents (e.g., PWROG-14015-P, PWROG-14027-P)that is deemed applicable.

c. Provide the plant-specific design parametersassociatedwith the seal leakoff line and confirm whether they are bounded by each of the model input parametersin Table 2 of PWROG-14015-P for the appropriateanalysis category. If any parameters in Table 2 are not bounded, please provide justification that the generically calculated leakage rate and maximum pressure are applicable.

d. Confirm that the #1 seal faceplate material is silicon nitride for all RCPs. Alternately, if one or more RCPs use a different material, please identify the material used and provide justification for the leakage rate assumed to apply to these RCPs.

e. Provide the set pressure and flow area associated with the relief valve on the #1 seal leakoff line common headerpiping.

f. Provide an estimate of the piping diameter, length, and number and type of components for the seal leakoff line common headerpiping.

g. If plant modifications will be undertaken to move the plant to a more favorable category relative to RCP seal leakage, please identify the applicable modifications and discuss the associatedcompletion timeline.

DNC Response:

a.) The assumption of 21 gpm of seal leakage per RCP is no longer valid. The PWROG has documented new leakage rates for Westinghouse RCP Original Equipment Manufacturer (OEM) Seals in PWROG-14015, Revision 1 using bounding plant configurations. The initial leakage information consisted of three points: initial leakage at normal operating temperature and normal operating pressure (NOT/NOP);

peak leakage at 1500 psia; and leakage at the cooled-down, depressurized conditions.

Additional studies have been documented in PWROG-14015, Revision 1 to evaluate the

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 52 of 59 linear assumptions between points and the effect of minimal subcooling for Category 1 seals. The intermediate flow rates are slightly above what is predicted by the linear assumption for seal leakage between the Revision 0 points. The seal leakage flow rate is almost unaffected from the change from 5°F of sub-cooling to less than V0 F of sub-cooling. For each pressure analyzed, these points are within 0.1 gpm.

b.) The RCP seal leakage rates for the Westinghouse OEM seals are documented in Attachment G of Engineering Technical Evaluation ETE-NAF-2012-0150, Revision 2.

These leakage rates are taken from PWROG-14015.

c.) Relative to Westinghouse PWROG 14015-P , MPS3 is considered a Category 1 plant. Westinghouse PWROG 14015-P, Table 2 reports the piping configuration for an individual seal return line. A comparison between PWROG-14015-P (Table 2) and MPS3 specific values was performed in the document titled "SE Review #2.c - Seal Return Line Comparison Tables," (previously provided to the NRC staff and available for their review). These tables show that all plant-specific parameters are bounded by the model input parameters except for number of downstream fittings. The assumed PRWOG-14015-P pressure analysis downstream fitting count (i.e., none) was not bounding, but conservative pipe diameter and pipe length analysis assumptions ensure a bounding overall hydraulic resistance.

d.) The #1 seal faceplate material for the two Westinghouse OEM RCP seals is silicon nitride.

e.) The main relief valve is 3CHS*RV8121, which has a 1.838 sq.in. flow area and a 150 psig set pressure.

f.) Engineering Record Correspondence (ERC) 25212-ER-04-0004 transmitted analysis inputs for Calculation 11-0188-TR-001, "MP3 RCP Seal Return Line Analysis" which is a transient analysis for the RCP seal return line. Document "SE Review #2.f - ERC 25212-ER-04-0004" (previously provided to the NRC staff and available for their review) contains the requested component information.

Document titled "SE Review #2.f - Calculation 00-109 excerpts,"(previously provided to the NRC staff and available for their review) contains applicable pages from Calculation 00-109, "Development of MP3 CHS Proto-Flo Model", R1, which summarizes pipe lengths and fittings for the subject common return line piping. Specifically, Pipe #s 2, 37, and 117 through 124 are in the seal leak-off return flow common piping back to the VCT as identified in the nodal diagram included in Calculation 00-109.

The Question SE 2.c response provides information on the four %-inch diameter individual seal return line branches from the pump seal leak-off nozzle to the common line.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 53 of 59 g.) As discussed in Engineering Technical Evaluation ETE-CRP-2012-0008, Rev. 4, Section 3.1.1, two Westinghouse OEM RCP seals have been replaced with Flowserve seals during the fall 2014 refueling outage. The remaining two Westinghouse seals are currently scheduled to be replaced with Flowserve N-seals during the spring 2016 outage.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Safety Evaluation Review Item 6 Please provide adequate basis that calculations performed with the NOTRUMP code (e.g., those in WCAP-17601-P, WCAP-17792-P) are adequate to demonstrate that criteria associated with the analysis of an ELAP event (e.g., avoidance of reflux cooling, promotion of boric acid mixing) are satisfied. NRC staff confirmatory analysis suggests that the need for implementing certain mitigating strategies for providing core cooling and adequate shutdown margin may occur sooner than predicted in NOTRUMP simulations.

DNC Response:

The PWROG has documented the applicability of the NOTRUMP code for the evaluation of the ELAP event and application of its results with regards to criteria for boron mixing and reflux cooling for Westinghouse designed PWRs in PWROG-14064.

PWROG-14064 provides a comparison of results from the NOTRUMP and NRC's TRACE computer codes for the parameters of interest and shows that the NOTRUMP predicted results agree well or are conservative with respect to the TRACE predicted results. For example, the comparison shows that NOTRUMP provides a conservative estimate of the required time when the primary make-up pumps are required for an ELAP event as compared to TRACE. Therefore, it is concluded that NOTRUMP is acceptable for simulation of the ELAP event for a 4-loop Westinghouse unit such as MPS3 when used within the criteria for reflux cooling and boron mixing.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 54 of 59 Safety Evaluation Review Item 8 Security Issues.

DNC Response:

DNC was requested to provide plan changes reflecting documentation, tools, and equipment needed to support personnel and vehicle access. Accordingly, the following information is provided:

The five BDB External Events (BDBEEs) were reviewed for each of DNC's nuclear sites by DNC's Security Department and the BDB group. Security discussed their general response to external events and the equipment and procedures needed to accomplish their mission and to support BDB's mission in responding to these external events.

Hostile action coincident with a BDBEE was not assumed. Security evaluated the impact of the loss of the site security facilities; in particular, the storage requirements (e.g., keys & access logs) that would be needed for Security's response to a BDBEE per NEI 12-06.

The site Security buildings are not designed to survive a BDB external event. For MPS, it was concluded that the loss of station Security vital door keys is not a concern since the Operations Department has access to an additional, redundant set of keys for vital area doors which are maintained in the Main Control Room (MCR), which is a structure protected from the design basis external hazards. Operations will issue these vital access keys, as necessary, to operators being dispatched to the vital areas to implement FLEX strategies based on the guidance contained in the FLEX Support Guidelines.

With regard to access control, the Access list is generated daily and will be stored in the MPS3 Main Control Room. Security has agreed to re-locate controlled copies of the procedures/documents that are needed for their response to BDBEE and General Emergency situations to a protected building.

Tools and equipment necessary to support personnel or vehicle access (fence cutters, hand tools, etc) have been procured and are stored in the BDB Storage Building.

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 55 of 59 Safety Evaluation Review Item 9 Please provide adequate justification for the seal leakage rates calculated according to the Westinghouse seal leakage model that was revised following the issuance of NSAL-14-1. The justification should include a discussion of the following factors:

a. benchmarking of the seal leakage model againstrelevant data from tests or operating

events,

b. discussion of the impact on the seal leakage rate due to fluid temperatures greater than 550'Fresulting in increased deflection at the seal interface,

c. clarification whether the second-stage reactor coolant pump seal would remain closed under ELAP conditions predicted by the revised seal leakage model and a technical basis to support the determination, and,

d. justification that the interpolation scheme used to compute the integrated leakage from the reactor coolant pump seals from a limited number of computer simulations (e.g., three) is realisticor conservative.

DNC Response:

a.) The PWROG is performing a benchmark of the EDF 7" seal. Testing of the EDF 7" seal occurred in the mid-1980's. WCAP-10541, Revision 2 includes a summary of the test. The benchmark will use the analysis methodology described in the response to Item b below.

WCAP-10541, Revision 2 documented the No. 1 seal leakage rates of an 8" Westinghouse RCP seal following a loss of all Alternating Current (AC) power for a reference case. MPR Associates was contracted by Westinghouse to perform independent calculations of RCP seal leakage, the results of which are provided in MPR-797. In addition, the United States Nuclear Regulatory Commission (NRC) contracted with Energy Technology Engineering Center (ETEC) to perform an independent investigation of the seal leakage rates for Westinghouse RCPs following a loss of all seal cooling. The ETEC work is summarized in WCAP-10541, Revision 2.

These analyses used the same overall analysis methodology which is summarized in the response to Item b below. The seal flow rate predicted by ETEC at low RCS pressure is approximately one-half of that predicted by Westinghouse. A review of the three areas in the analysis methodology indicated that the difference is not expected to be related to differences in the seal deformation model or in the seal hydraulic model.

However, there are sufficient differences in the seal leak-off line hydraulic model. The work by ETEC utilized a different two-phase flow correction factor and void fraction correlation. Therefore, the Westinghouse method is considered to provide acceptable results.

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 56 of 59 The estimated times to reach reflux cooling for MPS3 is 23.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months which provides a margin to the established time of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months for the implementation of the RCS makeup of 7.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months . This margin is judged to be sufficient based on the comparison of the independent evaluations of the RCP leakage in WCAP-10541, Revision 2. This will be confirmed by the benchmark analysis.

b.) PWROG-14013, Revision 1 "Summary of Validation of Seal Flow Calculations at Reduced Reactor Coolant System Pressures Report," provides a description of the overall analysis methodology which is summarized in Section 2.4 as follows:

- Determination of seal deformation due to pressure and temperature gradients using a finite element evaluation,

- Determination of seal flow and seal gap as a function of seal inlet and outlet pressures based on assumed seal gap, and

- Determination of seal outlet pressure due to No. 1 seal leak-off line pressure drop.

Westinghouse performed a finite element evaluation to determine the effect of pressure and temperature boundary conditions on the mechanical deformation of the No. 1 seal.

The Westinghouse seal flow rates provided in PWROG-14015, Revision 2, used the ITCHSEAL code to determine the seal flow rate. The ITCHSEAL code is a merger of two programs: the Itch program, which is used to model general thermal hydraulic transients, and the Seal program, which solves the hydraulic and force balances for the film riding seals. The ITCHSEAL code uses the results of the finite element analysis as input.

Therefore, the impact on the seal leakage rate due to fluid temperatures greater than 550'F (increased deflection) has been incorporated into the ITCHSEAL analyses. The results of these calculations are documented in PWROG-14015, Revision 2.

c.) Section 3.1.1 of PWROG-14017, Revision 1, provides a discussion of the evaluation of the No. 2 seal performance from WCAP-10541, Revision 2. Section 4.3 of WCAP-10541, Revision 2, states that Westinghouse performed thermal-hydraulic and mechanical finite element analysis for the pressure and temperature conditions on the No. 2 seal ring and runner assemblies similar to the No. 1 seal described in Section 4.1 of WCAP-10541. The analysis indicates that the temperature gradient across the No. 2 seal ring face during a loss of all seal cooling would offset any high pressure deflections in situations where the No. 1 seal remains functional. Further, it is stated that the thermal face rotation is more than an order of magnitude greater than the pressure induced deflection. Testing of the 7" EDF seal was discussed in Section 7 of WCAP-10541. Therein, it is stated that the No. 2 seal established the thermal gradients and pressure loadings which forced the seal faces to be diverging resulting in the closure of the No. 2 seal in the manner predicted by the analysis of the 8-inch standard and 8-inch cartridge seal. Therefore, it was concluded that the No. 2 seal is considered to be

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 57 of 59 functioning, but is modeled as a tightly closed obstruction which does not allow flow to pass.

d.) The PWROG has documented leakage rates for Westinghouse RCP Original Equipment Manufacturer (OEM) Seals in PWROG-14015, Revision 2 using bounding plant configurations. The initial leakage information consisted of three points: initial leakage at normal operating temperature and normal operating pressure (NOT/NOP);

peak leakage at 1500 psia; and leakage at the cooled-down, depressurized conditions.

Additional studies have been documented in PWROG-14015, Revision 2 to evaluate the linear assumptions between points and the effect of minimal subcooling for Category 1 seals. The intermediate flow rates are slightly above what is predicted by the linear assumption for seal leakage between the Revision 0 points. The seal leakage flow rate is almost unaffected from the change from 50F of sub-cooling to less than 1F of sub-cooling. For each pressure analyzed, these points are within 0.1 gpm. PWROG-14015, Revision 2 transmits the results of the calculation of the seal flow rate for each category of plant identified in PWROG-14008, Revision 1. Westinghouse has demonstrated the reasonableness of linear interpolation between points, that the peak leakage occurs at 1500 psia, and to include minimal subcooling.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

Safety Evaluation Review Item 10 The NRC staff understands that Westinghouse has recently recalculated seal leakoff line pressuresunder loss of seal cooling events based on a revised seal leakage model and additionaldesign-specific information for certain plants.

a. Please clarify whether the piping and all components (e.g., flow elements, flanges, valves, etc.) in your seal leakoff line are capable of withstanding the pressure predicted during an ELAP event according to the revised seal leakage model.

b. Please clarify whether operator actions are credited with isolating low-pressure portions of the seal leakoff line, and if so, please explain how these actions will be executed under ELAP conditions.

c. If overpressurization of piping or components could occur under ELAP conditions, please discuss any planned modifications to the seal leakoff piping and component design and the associatedcompletion timeline.

d. Alternately, please identify the seal leakoff piping or components that would be susceptible to overpressurization under ELAP conditions, clarify their locations, and provide justification that the seal leakage rate would remain in an acceptable range if the affected piping or components were to rupture.

Serial No. 14-393C Docket No. 50-423 Order EA-1 2-049 Attachment 2 Page 58 of 59 DNC Response:

a) PWROC-14008-P, R2, "No. 1 Seal Flow Rate for Westinghouse Reactor Coolant Pumps following Loss of All AC Power, Task 1: Documentation of Plant Configurations" (September 2014) defines MPS3 as a "Category 1" plant relative to PWROC-14015-P.

PWROC-14015-P, RO, "No. 1 Seal Flow Rate for Westinghouse Reactor Coolant Pumps following Loss of All AC Power, Task 2: Determine Seal Flow Rates" (June 2014) defines a Category 1 plant as having a maximum 951 psia operating pressure at the pump/seal connection and 659 psia downstream of the seal return line flow elements.

NRC Information Notice 2003-19, "Unanalyzed Condition of Reactor Coolant Pump Seal Leak-off Line during Postulated Fire Shutdown or Station Blackout" was based upon a MPS3 LER that identified an overpressure condition in the RCP # 1 seal leak-off return lines. Subsequent to this information notice, the MPS3 RCP #1 seal leak-off lines have been modified so that all components are capable of withstanding the operating pressure predicted during a SBO event which includes an extended loss of seal cooling transient (

Reference:

Calculation ENG-04062C3 and MMOD DM3-00-0008-04). The SBO transient is initially indistinguishable from the ELAP transient.

Therefore, all seal return line components are capable of withstanding the operating pressures predicted during an ELAP event.

Per Calculation ENG-04062C3, the MPS3 RCP #1 seal leak-off line reanalysis work used a 2045 psig at 518*F peak operating condition upstream of the flow elements which envelopes the PWROC-14015-P results. Also, a 2045 psig at 424°F operating condition is used to assess piping downstream of the flow elements. These operating conditions envelope the PWROC-14015-P operating pressure results.

MMOD DM3-00-0008-04 replaced seal return line valves with valves that are suitable for the extended loss of all seal cooling transient. These valves remain suitable in light of the PWROC-14015-P operating pressure results.

In summary, we conclude that all RCP # 1 seal return line components remain suitable for the extended loss of all seal cooling transient that occurs during an ELAP event and given the PWROC-14015-P results.

b) During the first 45-minutes, the SBO and ELAP scenarios are basically indistinguishable. In the SBO coping analysis, DNC credits operator action outside the control room to isolate the RCP # 1 seal leak-off return line quickly (e.g., 14-minutes) via 3CHS*MV8100 closure to ensure hot water is not transported to the vicinity of the charging pump suction or to the Volume Control Tank (which would complicate the

Serial No. 14-393C Docket No. 50-423 Order EA-12-049 Attachment 2 Page 59 of 59 charging pump restart evolution after the SBO diesel generator becomes available at about 1-hour into a SBO event).

This seal return line isolation is accomplished via EOP 35 ECA-0.0, "Loss of All AC Power", Step 6 consistent with the current SBO coping scenario analysis. This credited operator action is not needed to protect piping inside or outside containment from over-pressurization. However, relief valve discharges outside containment (which prevent over-pressurization) would be a coping strategy issue, if 3CHS*MV8100 was not isolated by operator action because it could create Auxiliary Building (AB) flooding or an AB harsh environment.

In summary, an SBO or ELAP event are initially indistinguishable and 3CHS*MV8100 isolation is required in our SBO coping analysis. Therefore, this action occurs during the ELAP scenario via 3CHS*MV8100 local closure. This operator action is associated with ensuring an uncomplicated charging pump restart evolution, if an EDG or SBO diesel generator subsequently becomes available to support charging pump operation.

c) Seal leak-off line over-pressurization is precluded by RCP #1 seal leak-off line modifications made circa 2003/2004 in response to NRC Information Notice 2003-19 and the associated LER.

d) Seal leak-off line over-pressurization is precluded by RCP #1 seal leak-off line modifications made circa 2003/2004 in response to NRC Information Notice 2003-19 and the associated LER.

The documents referenced in the above response have previously been provided to the NRC staff and are available for their review.

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