Text

Second 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)
	ML14064A201
Person / Time
Site:	Cooper
Issue date:	02/26/2014
From:	Limpias O; Nebraska Public Power District (NPPD)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	EA-12-049, NLS2014019
	Download: ML14064A201 (61)
	v • d • e

H Nebraska Public Power District Always there when you need us NLS2014019 February 26, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Nebraska Public Power District's Second 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)

Cooper Nuclear Station, Docket No. 50-298, DPR-46

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. NRC Interim Staff Guidance JLD-ISG-2012-01, "Compliance with Order EA- 12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," Revision 0, dated August 29, 2012

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

4. NPPD Letter, "Initial 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 October 29, 2012

5. NPPD Letter "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

Dear Sir or Madam:

On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued an Order (Reference 1) to Nebraska Public Power District (NPPD). Reference I was immediately effective and directs NPPD to develop, implement, and maintain guidance and strategies to maintain or restore core cooling, containment, and spent fuel pool cooling capabilities in the event of a beyond-design-basis external event. Specific requirements are outlined in Attachment 2 of Reference 1.

COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com

NLS2014019 Page 2 of 3 Reference 1 required submission of an initial status report 60 days following issuance of the final interim staff guidance (Reference 2) and an overall integrated plan pursuant to Section IV, Condition C. Reference 2 endorses industry guidance document Nuclear Energy Institute 12-06, Revision 0 (Reference 3) with clarifications and exceptions identified in Reference 2. Reference 4 provided NPPD's initial status report for Cooper Nuclear Station (CNS) regarding mitigation strategies. Reference 5 provided CNS' Overall Integrated Plan for diverse and flexible coping strategies (FLEX).

Reference I requires submission of a status report at six-month intervals following submittal of the overall integrated plan. Reference 3 provides direction regarding the content of the status reports. The purpose of this letter is to provide the second six-month status report pursuant to Section IV, Condition C.2, of Reference 1, that delineates progress made in implementing the requirements of Reference 1. Attachment 1 provides an update of milestone accomplishments since the last status report, including any changes to the compliance method, schedule, or need for relief and the basis, if any. provides Revision 1 of CNS' Overall Integrated Plan for FLEX.

This letter contains no new regulatory commitments.

Should you have any questions regarding this report, please contact David Van Der Kamp, Licensing Manager, at (402) 825-2904.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on: Z- Z.-C. 4k-Since y, SA. Lidmpia//

L~mas Vice President - Nu lear and Chief Nuclear Officer

/bk Attachments: 1. Nebraska Public Power District's Second 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

2. Cooper Nuclear Station - Overall Integrated Plan for Diverse and Flexible Coping Strategies (FLEX), Revision I

NLS2014019 Page 3 of 3 cc: Regional Administrator, w/attachments USNRC - Region IV Director, w/attachments USNRC - Office of Nuclear Reactor Regulation Senior Resident Inspector, w/attachments USNRC - CNS Cooper Project Manager, w/attachments USNRC - NRR Project Directorate IV- 1 NPG Distribution, w/Attachment I only CNS Records, w/attachments

NLS2014019 Page 1 of 6 Attachment 1 Nebraska Public Power District's Second 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 Introduction Nebraska Public Power District (NPPD) developed an overall integrated plan for Cooper Nuclear Station (CNS) (Reference 1), documenting the diverse and flexible coping strategies (FLEX), in response to Reference 2. This attachment provides an update of milestone accomplishments since submittal of the first status report to the overall integrated plan (Reference 3), including any changes to the compliance method, schedule, or need for relief/relaxation and the basis, if any.

Milestone Accomplishments The following milestone(s) have been completed since the submittal of the first status report for the overall integrated plan, and are current as of February 15, 2014:

None.

Milestone Schedule Status The following table provides an update to Attachment 2 of the overall integrated plan. It provides the activity status of each item, and whether the expected completion date has changed.

The dates are planning dates subject to change as design and implementation details are developed.

The revised milestone target completion dates do not impact the Order implementation date.

Milestone Target Status Revised Target Completion Date Completion Date 60-day Status Update October 2012 Complete Submit Overall Integrated Plan February 2013 Complete 6-month Status Update August 2013 Complete Refine Strategies September 2013 Not Started June 2014 (post-NRC review)

Develop Strategies / Contract January 2014 Not Started August 2014 with Regional Response Center 6-month Status Update February 2014 Complete Regional Response Center TBD Not Started Operations

NLS2014019 Attachment I Page 2 of 6 Milestone Target Status Revised Target Completion Date Completion Date Develop Storage Plan May 2014 Not Started Purchase FLEX Equipment June 2014 Not Started Issue Maintenance Procedures August 2014 Not Started (for FLEX equipment) 6-month Status Update August 2014 Not Started Develop Training Plan October 2014 Not Started Implementation Outage 1 Fall 2014 Not Started Develop Online Mods and December 2014 Started Implementing Procedures 6-month Status Update February 2015 Not Started Walk-throughs or April 2015 Not Started Demonstrations Implement Training May 2015 Not Started 6-month Status Update August 2015 Not Started Develop Outage Mods and October 2015 Not Started Implementing Procedures Implement Online Mods and December 2015 Not Started Procedures 6-month Status Update February 2016 Not Started Perform Staffing Assessment May 2016 Not Started 6-month Status Update August 2016 Not Started Implementation Outage 2 Fall 2016 Not Started Implement Outage Mods and Fall 2016 Not Started Procedures Implement Training Updates Fall 2016 Not Started Submit Completion Report Fall 2016 Not Started Changes to Compliance Method There are no changes to the compliance method as documented in the overall integrated plan at this time.

NLS2014019 Attachment I Page 3 of 6 Need for Relief/Relaxation and Basis for the Relief/Relaxation NPPD expects to comply with the Order implementation date and no relief/relaxation is required at this time.

Open and Confirmatory Items in the Interim Staff Evaluation In Reference 4, the Nuclear Regulatory Commission (NRC) issued the Interim Staff Evaluation (ISE) for CNS relating to the overall integrated plan. The following table provides a summary of the confirmatory items documented in the ISE and the status of each item. There were no open items identified in the ISE.

ISE Confirmatory Items Item Number Description Status 3.1.1.2.A Confirm that the required debris removal Open equipment remains functional and deployable to clear obstructions from pathways between the FLEX storage locations and deployment locations, after the FLEX storage building locations are finalized.

3.1.1.4.A Confirm the location(s) of the staging area(s) for Open equipment from the RRC, and the licensee's plans for transportation from the RRC, staging, and on-site deployment is in accordance with the guidance in NEI 12-06, Sections 5.3.4, 6.3.4, 7.3.4, and 8.3.4, or provide an acceptable alternative to that guidance.

3.1.3.1 .A Confirm that when the FLEX equipment storage Open building locations are finalized, separation distance and axis of separation is reviewed to confirm that the building locations are consistent with the recommendations in NEI 12-06, Section 7.3.1.

3.1.4.2.A Confirm that obtaining makeup water from the Open Missouri River during an ELAP event adequately addresses NEI 12-06, Section 8.3.2, consideration 3.

3.2.1.1 .A Benchmarks must be identified and discussed Open which demonstrate that Modular Accident Analysis Program (MAAP) is an appropriate code for the simulation of an ELAP event at CNS, consistent with the NRC endorsement (ADAMS Accession No. ML13275A318) of the industry position paper on MAAP.

NLS2014019 Attachment I Page 4 of 6 ISE Confirmatory Items Item Number Description Status 3.2.1.1 .B The licensee should demonstrate that the collapsed Open reactor pressure vessel level remains above Top of Active Fuel and the reactor coolant system cool down rate is within technical specifications limits.

3.2.1.1 .C The licensee should demonstrate that MAAP is Open used in accordance with Sections 4.1, 4.2, 4.3, 4.4, and 4.5 of the June 2013 position paper (ADAMS Accession No. ML13190A201).

3.2.1.1 .D The licensee must identify, in using MAAP, the Open subset of key modeling parameters cited from Tables 4-1 through 4-6 of the "MAAP Application Guidance, Desktop Reference for Using MAAP Software, Revision 2" (Electric Power Research Institute Report 1020236, available at www.epri.com). This should include response at a plant-specific level regarding specific modeling options and parameter choices for key models that would be expected to substantially affect the ELAP analysis performed for CNS.

3.2.1.2.A Confirm that the analysis for a long-duration Open ELAP event shows that the reactor recirculation pump seal leakage value does not exceed the value used in analysis of a 4-hour station blackout event.

3.2.1.3.A Confirm that the method for transferring water Open from the hotwells to the ECSTs, including flow path, valves, pumps, and related equipment, when developed, is reliable.

3.2.1.3.B Confirm that the RCIC room heatup evaluation Open and RCIC room flooding time evaluation are completed with acceptable results.

3.2.1.3.C Confirm that the licensee's staffing assessment is Open completed and it shows that proposed actions from the FLEX strategies can be completed within the specified time constraints.

3.2.1.4.A Confirm that the Phase 2 FLEX equipment Open performance criterion, when developed, supports the licensee's mitigation strategies.

NLS2014019 Attachment I Page 5 of 6 3.2.2.1 .A Confirm that modifications to the reactor building Open roof hatch provide the ability to maintain adequate SFP area ventilation.

3.2.3.A Confirm that CNS's containment venting strategy Open is finalized and that the strategy supports both containment pressure protection and proposed RCIC and Phase 2 FLEX pump operation.

3.2.3.B With regard to maintaining containment, the Open implementation of BWROG Emergency Procedure Guideline/Severe Accident Guideline, Revision 3, including any associated plant-specific evaluations, must be completed in accordance with the provisions of NRC letter dated January 9, 2014 (ADAMS Accession No. ML13358A206).

3.2.4.2.A Confirm that fan sizing evaluations support Open adequate ventilation in the main control room, in the RCIC room, and in other applicable plant areas.

3.2.4.6.A Confirm that analyses addressing heat up in areas Open that might have personnel habitability considerations conform to the guidance in NEI 12-06, Section 3.2.2, Guideline 11, or provide an acceptable alternative to that guidance.

3.2.4.7.A Confirm that the design provisions, as well as Open operational and protection requirements for the new on-site well/water treatment equipment used for Phase 2 water sources adequately support CNS's proposed ELAP strategies.

3.2.4.8.A Confirm that adequate electrical interaction and Open isolation considerations are adequately addressed.

3.2.4.8.B Confirm that the sizing of the portable FLEX Open diesel generators adequately supports CNS's ELAP mitigation strategy.

3.2.4.8.C Provide single-line diagrams showing the proposed Open connections of Phase 2 and Phase 3 electrical equipment and showing protection information (e.g., breaker, relay, or fuse) and rating for the equipment used when available.

3.2.4.10.A Confirm that the minimum dc voltage and dc load Open profile for the ELAP have been determined, the minimum dc bus voltage and the associated load profile supports CNS's proposed ELAP mitigation strategy.

NLS2014019 Page 6 of 6 As discussed with the NRC during the audit process, the overall integrated plan was revised to correct minor discrepancies identified between the Sequence of Events Timeline (Attachment 1A) and the Sequence of Events Time Constraints discussion (Pages 5 and 6). Additionally, the Milestone Schedule (Attachment 2) was updated. to this letter provides Revision I of CNS' Overall Integrated Plan for FLEX.

References The following references support the status update to the overall integrated plan described in this attachment:

1. NPPD Letter, "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

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. NPPD Letter, "Cooper Nuclear Station's First 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 049)," dated August 27, 2013

4. NRC Letter, "Cooper Nuclear Station - Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) (TAC No.

MF0972)," dated February 11, 2014

NLS2014019 Page 1 of 52 Attachment 2 Cooper Nuclear Station -

Overall Integrated Plan for Diverse and Flexible Coping Strategies (FLEX)

Revision 1

CNS Overall Integrated Plan for Diverse and Flexible Coping Strategies (FLEX)

Revision 1 (February 15, 2014)

General Integrated Plan Elements (BWR)

Determine Applicable Input the hazards applicable to the site; seismic, external Extreme External Hazard. flood, high winds, snow, ice, cold, high temps.

Describe how NEI 12-06 sections 5 - 9 were applied and Ref: NEI 12-06 section 4.0-9.0 the basisfor why the plant screenedout for certain hazards.

JLD-ISG-2012-01 section 1.0 Seismic Hazard Assessment:

Per the Updated Safety Analysis Report (USAR) (Reference 1) Section 11-5, the seismic criteria for Cooper Nuclear Station (CNS) include two design basis earthquake spectra:

Operating Basis Earthquake at 0.1g and Safe Shutdown Earthquake (SSE) at 0.2g. Per Reference 2, all sites will consider the seismic hazard.

External Floodinq Hazard Assessment:

Per the USAR (Reference 1) Section 11-4, the design basis flood is a value of 903.0 Mean Sea Level (MSL) for the Probable Maximum Flood (PMF). The general ground elevation surrounding CNS Class I Structures is elevated 13 feet above the natural floodplain to 903 feet MSL. The finished floor elevation of all Class I Structures is placed at elevation 903.5 feet MSL, or 1/2 feet above the PMF event. These structures were designed for a hydraulic load equivalent to a groundwater elevation of 903 feet. The station site grade level of 903 feet MSL has been raised 13 feet above the natural grade level of 890 feet MSL, in order to bring final grade one foot above the existing 902 feet MSL levee constructed by the Corps of Engineers. This levee was raised above its original design level and presently has a three foot minimum free board over the 1952 flood of record (899 feet MSL). Flooding of the station is considered to be extremely unlikely due to the combination of upstream Missouri River flood control and the high final site grade. With respect to the 1,000 year, 10,000 year, and 1,000,000 year (PMF) floods, these water levels will provide 3-12 feet, 1-1/2 feet, and 6 inches of freeboard respectively below the 903'6" grade floor elevation of the principle structures.

Per Reference 2, the site is considered a "dry" site, i.e., the plant is built above the design basis flood level and the external flooding hazard need not be considered.

High Wind Hazard Assessment:

Per the USAR (Reference 1) Section 11-3, the design wind pressure for the station and structures is 30 lbs per square foot which is the equivalent of sustained winds up to 100 mph. Station structures have been designed to withstand this wind velocity in accordance with ASCE Paper 3269. Additionally, per USAR (Reference 1) Section XlI-2, Class I structures are designed to the following:

Tornado design criteria

1. A tangential velocity of 300 mph.

2. A transverse velocity of 60 mph.

3. A pressure drop of 3 psi occurring over a 3 second time interval.

Tornado generated missile criteria

A 35 foot long utility pole with a 14 inch butt with an impact velocity of 200 mph.

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A one-ton missile such as compact-type automobile with an impact velocity of 100 mph and a contact area of 25 square feet.

" A 2 inch extra heavy pipe, 12 feet long.

Any other missile resulting from failure of a structure or component or one which has potential of being lifted from storage or working areas at the site.

The CNS site is located in an area characterized by the Nuclear Regulatory Commission (NRC) as having tornado design wind speeds greater than 130 mph (Reference 2, figure 7-

2) and as such, the tornado hazard will be considered.

Extreme Cold Hazard Assessment:

Per the USAR (Reference 1) Section 11-3, the design low outside temperature is -5°F dry bulb which will only be exceeded 1% of the time during the winter. The CNS site is located within the region characterized by the National Oceanic and Atmospheric Administration as having a 3-day snowfall of up to 18" (Reference 2, figure 8-1) and would need to consider snow removal in the deployment of the FLEX strategy. The CNS site is also located within the region characterized by Electric Power Research Institute (EPRI) as ice severity level 4 (Reference 2, figure 8-2). As such, the CNS site is subject to severe damage to power lines and/or existence of large amounts of ice. Per Reference 2, cold temperatures, snow, and ice will be considered.

Extreme High Temperature Hazard Assessment:

Per the USAR (Reference 1) Section 11-3, the CNS design high outside temperature is 97°F dry bulb (79 0 F wet bulb). Based on historical records, this temperature is only expected to be exceeded 1% of the time during the summer. Per Reference 2, all sites will address high temperatures.

==

Conclusion:==

The hazards applicable to CNS to consider on the deployment of the strategies to meet the baseline coping capability are seismic, high winds, extreme cold, snow, ice, and high temperature.

References:

1. Cooper Nuclear Station Updated Safety Analysis Report (USAR)

2. NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, Revision 0, August 2012 Key Site assumptions to Provide key assumptions associatedwith implementation of implement NEI 12-06 FLEX Strategies:

strategies.

Flood and seismic re-evaluationspursuant to the 10 CFR 50.54(W letter of March 12, 2012 are not Ref: NEI 12-06 section 3.2.1 completed and therefore not assumed in this submittal.

As the re-evaluationsare completed, appropriateissues will be entered into the corrective action system and Page 2 of 51

CNS Overall Integrated Plan for Diverse and Flexible Coping Strategies (FLEX)

Revision 1 (February 15, 2014) addressedon a schedule commensurate with other licensing bases changes.

" Exceptions for the site securityplan or other (license/sitespecific) requirements of 10CFR may be required.

" Deployment resources are assumed to begin arrivingat hour 6 andfully staffed by 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Certain Technical Specifications cannot be complied with during FLEX implementation.

Key assumptions associated with implementation of FLEX Strateqies:

" Flood and seismic re-evaluations pursuant to the 10 CFR 50.54(f) letter of March 12, 2012, are not complete and therefore not assumed in this submittal. As the re-evaluations are completed, appropriate issues will be entered into the corrective action system and addressed on a schedule commensurate with other licensing bases changes.

Following conditions exist for the baseline case:

Seismically designed DC banks are available.

Seismically designed AC and DC distribution available.

Plant initial response is the same as Station Blackout (SBO).

Best estimate analysis and decay heat is used to establish operator time and action.

No single failure of Structure, System or Component assumed.

The water in the Emergency Condensate Storage Tanks (ECST) is available for use.

The ECSTs are located in a seismically robust structure.

The water in the main condenser hotwells is available for use. CNS is an Alternate Source Term plant, and as such, the hotwells have been shown to survive a design basis earthquake (Reference 1).

The design hardened connection is protected against external event or redundant locations.

Implementation strategies and roads are assessed for hazards impact.

All Phase II components are stored at site and available after the event they were designed to be protected against.

" Additional staff resources are expected to arrive beginning at 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the site will be fully staffed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the event.

Open items where CNS does not have clear .quidance to complete an action related to this submittal:

Primary and secondary storage locations have not been selected yet; once locations are finalized implementation routes will be defined.

The location and sizing of the new raw water well has not been determined.

The method of transferring water from the hotwell to the ECSTs has not been determined.

" The staging area for the Regional Response Center (RRC) equipment has not been determined.

" Exceptions for the site security plan or other (license/site specific) requirements of 10 CFR may be required. Coordination of site FLEX integrated Plan with security personnel is required.

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CNS Overall Integrated Plan for Diverse and Flexible Coping Strategies (FLEX)

Revision 1 (February 15, 2014) 10 CFR 50.54(xW and/or 10 CFR 73.55(D):

This plan defines strategies capable of mitigating a simultaneous loss of all AC power and loss of normal access to the ultimate heat sink resulting from a beyond-design-basis event by providing adequate capability to maintain or restore core cooling, containment, and Spent Fuel Pool (SFP) cooling capabilities at all units on a site. Though specific strategies are being developed, due to the inability to anticipate all possible scenarios, the strategies are also diverse and flexible to encompass a wide range of possible conditions. These pre-planned strategies developed to protect the public health and safety will be incorporated into the unit emergency procedures and guidelines in accordance with established change processes, and their impact to the design basis capabilities of the unit evaluated under 10 CFR 50.59. The plant Technical Specifications contain the limiting conditions for normal unit operations to ensure that design safety features are available to respond to a design basis accident and direct the required actions to be taken when the limiting conditions are not met. The result of the beyond-design-basis event may place the plant in a condition where it cannot comply with certain Technical Specifications and/or with its Security Plan, and, as such, may warrant invocation of 10 CFR 50.54(x) and/or 10 CFR 73.55(p)

(Reference 2).

References:

1. Engineering Evaluation 01-147, Summary of Main Steam Isolation Valve (MSIV)

Leakage Pathway to the Condenser Seismic Qualification

2. Task Interface Agreement 2004-04, "Acceptability of Proceduralized Departures from Technical Specifications (TSs) Requirements at the Surry Power Station," (TAC Nos.

MC4331 and MC4332)," dated September 12, 2006 (Accession No. ML060590273)

Extent to which the guidance, Include a descriptionof any alternativesto the guidance, JLD-ISG-2012-01 and NEI 12- andprovide a milestone schedule ofplanned action.

06, are being followed.

Identify any deviations to JLD-ISG-2012-01 and NEI 12-06.

Ref: JLD-ISG-2012-01 NEI 12-06 13.1 Full conformance with JLD-ISG-2012-01 and NEI 12-06 is expected.

Provide a sequence of events Strategies that have a time constraintto be successful and identify any time should be identified with a technical basis and a constraint required for success justificationprovided that the time can reasonablybe met including the technical basis (for example, a walk through of deployment).

for the time constraint.

Ref: NEI 12-06 section 3.2.1.7 Describe in detail in this section the technical basisfor the Page 4 of 51

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Revision 1 (February 15, 2014)

JLD-ISG-2012-01 section 2.1 time constraintidentified on the sequence of events timeline Attachment ]A.

See attachedsequence of events timeline (Attachment ]A).

Technical Basis Support information, see attachedNSSS Significant Reference Analysis Deviation Table (Attachment 1B).

Discussion of time constraints identified in Attachment 1A table:

t-2: 11 - 13 minutes, Operators will secure High Pressure Coolant Injection (HPCI) after one cycle or 10 minutes of operation to limit HPCl room heatup (Reference 1).

t-3:, 30 minutes, Operators will open the DC Switchgear (SWGR) room doors and Control Room panel doors for panel without open backs to allow cooling (Reference 2)

t-4: 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Operations will designate the event as an Extended Loss of AC Power (ELAP). The one hour is based on a reasonable amount of time to diagnose the event. Notification of the RRC is expected to take place here; however, the exact timing will be determined during the development of the site's playbook.

t-5: 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Secure Main Turbine Emergency Lube Oil Pump (MTELOP). Action is required to support 250 VDC Division II Battery duration. The actions to secure the MTELOP consist of verifying that the Main Turbine has stopped rotating locally and then placing the control switch for the MTELOP in pull-to-lock in the Control Room.

Per References 11 through 14, the batteries have the following average loads:

125 VDC Division I 131 Amperes

125 VDC Division II 120 Amperes

250 VDC Division I 166 Amperes

250 VDC Division II 144 Amperes Per Reference 3, the battery must have at least 90% of capacity at all times. Per Reference 15, given the average loads, and assuming an end-of-discharge value of 1.85 volts-per-cell, the batteries are estimated as being capable of supplying all required loads for at least the following time periods:

125 VDC Division I 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

125 VDC Division II 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months

250 VDC Division I 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

250 VDC Division II 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months

" t-6: 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , Begin aligning Severe Accident Management Guideline (SAMG) Diesel Generator (DG) to battery chargers. Per Reference 2, 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months is required to place the SAMG DG in service. Personnel will begin arriving on-site at 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The shortest battery life is 250 VDC Division I at 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months .

t-7: 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , Begin alignment to transfer hotwell water to the ECSTs. When ECST inventory is near depletion, they will need to be re-filled to continue to allow RCIC to inject with cool water. The ECSTs have capacity for - 6 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . When personnel begin arriving on site, it will be necessary to align and commence filling the ECSTs. The procedure/method for transferring water from the hotwell has not yet been developed. CNS will develop a method and associated procedure to accomplish this.

" t-8: 6 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , Transfer Reactor Core Isolation Cooling (RCIC) suction to the torus.

When ESCT level no longer supports continued RCIC operation, Operators will transfer the suction to the torus.

t-9: 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , as Torus temperature rises to near the Heat Capacity Temperature Limit Page 5 of 51

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(HCTL), Operators will lower Reactor Pressure Vessel (RPV) pressure as low as possible while still maintaining RCIC in service per the Emergency Operating Procedures (EOP).

t-10: 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , Commence makeup to ECSTs from hotwell. At 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months into the event the makeup/leakage flow rate to the RPV will be 120 gpm (Reference 4), therefore makeup to the ECST will need to be at least 120 gpm.

" t-1 1: 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , Begin torus venting to maintain containment parameters within limits.

Per NEDC 33771 P (Reference 5), containment venting is assumed to start at 4 or 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . In conjunction with NRC Order EA-12-050 (Reference 6), CNS is upgrading the existing 10" Hard Pipe Vent (HPV) to a 12" Reliable Hardened Containment Vent (RHCV). With a 12" RHCV, drywell and wetwell parameters are maintained within limits.

" t-1 2: 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months , Place the SAMG DG in service. At 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months batteries will start to reach the end of their capabilities. Placing the SAMG DG in service will preserve the Division I Batteries.

" t-1 3: 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months , Realign RCIC back to the ECSTs. The available volume to transfer to the ECSTs from the hotwell is 80,625 gallons (Reference 4).

" t-14: 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months , Align RCIC suction to the torus to allow continued RCIC operation.

The minimum useable water available to RCIC from the ECSTs and hotwell is 178,369 gallons (Reference 4). Using the methodology in NUMARC 87-00 (Reference 8) the total required condensate for an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months SBO is 117,676 gallons and for a 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months SBO it is 215,326 gallons. A straight interpolation to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months shows 166,501 gallons required (((215,326 - 117,676)/2) + 117, 676 = 166, 501) and interpolating that to 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months shows that 190,914 gallons are required (((215,326 -

166,501)/2) + 166,501 = 190,914). Between 12 and 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months RCIC suction will need to be transferred back to the torus. This timing is not critical as the torus has sufficient capacity for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months without injection from the ECSTs (Reference 4).

Operators will monitor ECST level and transfer when required per Procedure 5.3ALT-STRATEGY (Reference 10). CNS Calculation NEDC 89-1886 (Reference 9) verified the condensate inventory for the original 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SBO coping analysis and demonstrated that the NUMARC methodology is a bounding method for all reactors and is conservative. Included in Reference 9 is a RPV leakage rate of 66 gpm which consists of the Technical Specifications maximum total leakage of 30 gpm and 18 gpm for each Reactor Recirculation Pump.

t-1 5:18 hours, Begin aligning raw water sources for makeup. Initially torus water level will rise due to injection from the ECSTs. When containment venting is required, torus water level will steady out and remain constant until RCIC suction is transferred to the torus, at which time torus level will begin to lower. Per Reference 5, approximately 3 feet of level in the suppression pool is vaporized during a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period through containment venting. The primary raw water source will be a well installed at the site. This new well will be sized for the makeup requirements 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after shutdown and be hardened against the appropriate external hazards.

The backup raw water source will be the Missouri River. The use of portable strainer/filters/demineralizers will be procured and used to assure the river water is acceptable for use in the reactor as an emergency makeup source.

" t-16: 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , Begin makeup to the ECSTs to maintain RCIC availability while Phase 3 resources are being placed into service.

Technical Basis Support information:

1. On behalf of the Boiling Water Reactor Owners Group (BWROG), GE-Hitachi (GEH)

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CNS Overall Integrated Plan for Diverse and Flexible Coping Strategies (FLEX)

Revision 1 (February 15, 2014) developed a document (NEDC-33771 P, Revision 1 (Reference 5)) to supplement the guidance in NEI 12-06 by providing additional BWR-specific information regarding the individual plant response to the ELAP and loss of Ultimate Heat Sink (UHS) events. The document includes identification of the generic event scenario and expected plant response, the associated analytical bases, and recommended actions for performance of a site-specific gap analysis. In the document, GEH utilized the NRC accepted SUPERHEX (SHEX) computer code methodology for BWR's long term containment analysis for the ELAP analysis. As part of this document, a generic BWR 4/Mark I containment nuclear steam supply system (NSSS) evaluation was performed. The BWR 4/Mark I containment analysis is applicable to CNS (a BWR 4, Mark I plant) coping strategy because it supplements the guidance in NEI 12-06 by providing BWR-specific information regarding plant response for core cooling, containment integrity, and SFP cooling. The guidance provided in Reference 5 was utilized as appropriate to develop coping strategies and for prediction of the plant's response.

2. CNS containment integrity for Phases 1 through 3 was evaluated by use of computer code MAAP 4.05 (Reference 7).

References:

1. NLS9100631, Response to Recommendations on Station Blackout, 10CFR50.63

2. Procedure 5.3SBO, Station Blackout

3. Technical Specification Surveillance Requirement 3.8.4.8

4. CNS-PSA-007, Cooper PRA Deterministic Calculations Notebook, Revision 1

5. NEDC 33771 P, GEH Evaluation of FLEX Implementation Guidelines, Revision 1

6. NRC EA-1 2-050, Order to Modify Licenses with Regard to Reliable Hardened Containment Vents

7. NEDC 13-004, CNS Evaluation of Diverse and Flexible Coping Strategies for Extended Loss of AC Power

8. NUMARC 87-00, Revision 1, Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors

9. Calculation NEDC 89-1886, CNS Station Blackout (SBO) Condensate Inventory

10. Procedure 5.3ALT-STRATEGY, Alternate Core Cooling Mitigating Strategies

11. NEDC 87-131A, 250VDC Division 1 Load and Voltage Study, Revision 13

12. NEDC 87-131 B, 250VDC Division 2 Load and Voltage Study, Revision 12

13. NEDC 87-131C, 125VDC Division 1 Load and Voltage Study, Revision 14

14. NEDC 87-131D, 125VDC Division 2 Load and Voltage Study, Revision 13

15. CNS Vendor Manual 1188, 125 & 250 Volt Batteries and Chargers

** * *' **Y.,;; ':.: .-:: ,".*:; :**'*, - ** "" .!* ** ,k . .** *:* o; :** ,,'. ; *.**:.* *: :,. ** ** ' '.*

Identify how strategies will be Describehow the strategieswill be deployed in all modes.

deployed in all modes.

Ref. NEI 12-06 section 13.1.6 Once storage locations have been determined, deployment routes will be designated to transport the FLEX equipment to the staging areas. An administrative program will be developed to keep the routes and staging areas clear during all modes of operation.

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CNS Overall Integrated Plan for Diverse and Flexible Coping Strategies (FLEX)

Revision 1 (February 15, 2014)

Provide a milestone schedule. The dates specifically requiredby the order are obligated This schedule should include: or committed dates. Other dates are planned dates subject

" Modifications timeline to change. Updates will be provided in the periodic (six o Phase I Modifications month) status reports. See attached milestone schedule o Phase 2 Modifications Attachment 2.

o Phase 3 Modifications

Procedure guidance development complete o Strategies o Maintenance

Storage plan (reasonable protection)

" Staffing analysis completion

" FLEX equipment acquisition timeline

Training completion for the strategies

" Regional Response Centers operational Ref: NEI 12-06 section 13.1 See attached milestone schedule Attachment 2.

Identify how the Provide a descriptionof the programmaticcontrols programmatic controls will be equipment protection,storage and deployment and met. equipment quality. See section 11 in NEI 12-06. Storage of equipment, 11.3, will be documented in later sections of this Ref: NEI 12-06 section 11 template and need not be included in this section.

JLD-ISG-2012-01 section 6.0 See section 6.0 ofJLD-ISG-2012-01.

CNS will implement an administrative program. A program owner will be assigned with responsibility for configuration control, maintenance, and testing. The equipment for ELAP will be dedicated and will have unique identification number. CNS FLEX equipment will be categorized as Quality Augmented (QA). The QA will be based on selected Appendix B similar to Appendix K and SBO guidance. Standard industry preventive maintenance (PM) will be established for all components and testing procedures will be developed and frequencies established based on type of equipment and considerations made within EPRI guidelines. CNS will assess the addition of program description into USAR and Technical Requirements Manual.

Describe training plan. List trainingplansfor affected organizationsor describe the planfor trainingdevelopment.

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Revision 1 (February 15, 2014)

New training of general station staff and Emergency Preparedness will be performed in 2016, prior to the design implementation. Simulator and license operator training will not be impacted. The Systematic Approach to Training will be utilized to implement this training.

Describe Regional Response Discussion in this section may include the following Center plan. information and will be further developed as the Regional Response Center development is completed.

" Site-specific RRC plan

" Identification of the primary and secondary RRC sites

" Identification of any alternateequipment sites (i.e.

another nearby site with compatible equipment that can be deployed)

" Describe how delivery to the site is acceptable

" Describe how all requirements in NEI 12-06 are identified The industry is establishing two RRCs to support utilities during beyond design basis events. The RRCs will hold five sets of equipment, four of which will be able to be fully deployed when requested, the fifth set will have equipment in a maintenance cycle. Equipment will be moved from an RRC to a local Assemble Area, established by the Strategic Alliance for FLEX Emergency Response (SAFER) team and the utility. Communications will be established between the nuclear site and the SAFER team and required equipment moved to the site as needed. First arriving equipment, as established during development of the nuclear site's playbook, will be delivered to the site within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from the initial request.

Nebraska Public Power District (NPPD) has entered into a contract with Pooled Equipment Management Company to obtain, maintain and deliver the equipment specified by NPPD to the designated staging area within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Notes: None.

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Maintain Core Cooling Determine Baseline coping capability with installed coping' modifications not including FLEX modifications, utilizing methods described in Table 3-1 of NEI 12-06:

RCIC/HPCI/IC

Depressurize RPV for injection with portable injection source a Sustained water source BWR Installed Equipment Phase 1:

Provide a general descriptionof the coping strategies using installed equipment including modificationsthat are proposed to maintain core cooling. Identify methods (RCIC/HPCI/IC) and strategy(ies) utilized to achieve this coping time.

Power Operation, Startup, and Hot Shutdown:

The primary strategy for core cooling is to supply high quality water via RCIC with suction from the ECSTs. Two 50,000-gallon ECSTs are installed for the exclusive use of the RCIC and HPCI systems (Reference 4, Section VI-5.2.7). An additional 80,625 gallons is available in the Main Condenser Hotwells (Reference 6). The hotwells have been evaluated to be available after a seismic event (Reference 7). The combined capacity of the ECSTs, hotwells, and suppression pool is sufficient to support RPV makeup for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months without external makeup sources.

At the initiation of the Beyond-Design-Basis External Event (BDBEE), MSIVs automatically close, feedwater is lost, and Safety Relief Valves (SRVs) automatically cycle to control pressure, causing reactor water level to decrease. When reactor water level reaches -42 inches, HPCI and RCIC automatically start with suction from the ECST (Reference 1, Technical Specification (TS), Table 3.3.5.1-1, Function 3A and TS Table 3.3.5.2-1, Function 1) and operate to inject makeup water to the reactor vessel. This injection recovers the reactor level to the high level setpoint of 54" (Reference 1, Tables 3.3.5.1-1 (Function 3c) and 3.3.5.2-1 (Function 2)). The SRVs will be used to control reactor pressure between approximately 800-1000 psig per EOP-1A RPV Control (Reference 2).

In a typical SBO event (and in the ELAP event), RCIC is able to provide make-up and maintain RPV level. HPCI is secured after one cycle or 10 minutes of operation (Reference 3).

After determination that Emergency Diesel Generators cannot be restarted, the operating crew determines the event is a beyond-design-basis event and anticipate a loss of power for an extended time period at approximately 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months into the event. RCIC is maintained feeding the reactor vessel with suction from the ECST. The RCIC trip signals and isolation signals that could possibly prevent RCIC operation when needed during the ELAP will be overridden in accordance with EOPs (Reference 2). Additionally, the Automatic Depressurization System (ADS) will be either placed in 'inhibit' or closely monitored to prevent automatic initiation of ADS. This is necessary to ensure reactor pressure is not reduced to a pressure which would prevent operation of RCIC.

' Coping modifications consist of modifications installed to increase initial coping time, i.e. generators to preserve vital instruments or increase operating time on battery powered equipment.

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Revision 1 (February 15, 2014)

As stated above, the primary method of reactor pressure control is by operation of the SRVs. Operator control of reactor pressure using SRVs require DC control power and pneumatic pressure (supplied by station batteries and the drywell pneumatics system (Reference 4, Section IV-4.6) to open. For Phase 1, the power for the SRVs is supplied by the station batteries. At event initiation the normal pneumatic supply is lost due to loss of power; however, each SRV is provided an accumulator which contains enough pneumatic pressure to operate each valve for multiple open/close cycles. The accumulators for the 2 SRVs associated with Low-Low SET (LLS) are sized to allow 14 cycles of each valve (Reference 4, Section IV-4.6). A plant specific drywell analysis was used to model the effects of a SBO event of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (Reference 5). The calculation models the operator taking manual action to disable LLS and initiate a manual cooldown of the RPV per EOPs (Reference 2). Twenty-seven individual SRV actuations in LLS mode and manual mode combined are required to maintain pressure control during the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SBO event. For the ELAP, after 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the 6 SRVs associated with ADS will be used to control pressure.

The accumulators for the ADS valves are sized for 5 cycles and are credited for operation for approximately 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months after a SSE (Reference 4, Section IV-4.6). The combined capacity of the accumulators is sufficient to allow operation of the SRVs until Phase 2 when pneumatics will be restored with FLEX equipment.

The torus continues to heat up due to RCIC exhaust and SRV cycling. During the time that torus temperature is increasing operators reduce reactor pressure to a pressure range (200 to 400 psig) which provides margin to the Unsafe Region of the Heat Capacity Temperature Limit (HCTL) curve (Reference 2). When the torus temperature reaches the Unsafe Region of the HCTL, RPV emergency depressurization is required (Reference 2).

In accordance with Emergency Procedure Guidelines and per BWROG guidance, EOPs have been revised to allow termination of RPV emergency depressurization at a pressure that will allow continued RCIC operation, because steam driven RCIC is the sole means of core cooling.

Based on experience derived from Fukushima, the RCIC system can run at a much higher lube oil temperature and suction source temperature than that originally assumed for the operation of RCIC. Additionally, the BWROG is developing a RCIC study (Reference 8) which will allow operation of RCIC at a lube oil temperature of >230°F; CNS will take the necessary actions to allow operation at elevated temperatures. Regarding Net Positive Suction Head (NPSH) for RCIC, CNS will perform a site-specific calculation to determine the available NPSH at elevated temperatures. Alternatively, if the analysis determines that adequate NPSH is not available, CNS will revise its strategy and use the water in the suppression pool first, preserving the ECST water into Phase 2 when makeup sources are available with FLEX equipment.

Cold Shutdown and Refueling:

The overall strategy for core cooling for Cold Shutdown and Refueling (Modes 4 and 5, respectively) are generally similar to those for Power Operation, Startup, and Hot Shutdown.

If an ELAP occurs during Cold Shutdown, water in the vessel will heatup. When temperature reaches 212°F, (Hot Shutdown) the vessel will begin to pressurize. During the pressure rise RCIC can be returned to service with suction from the ECST to provide injection flow. When pressure rises to the SRV setpoints, pressure will be controlled by SRVs. The primary and alternate strategies for Cold Shutdown are the same as those for Page 11 of 51

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Revision 1 (February 15, 2014)

Power Operation, Startup, and Hot Shutdown as discussed above for core cooling. While the drywell airlock may be open, the amount of steaming will have limited impact on the Reactor Building until Phase 2 manpower is available to shut the airlock and well before Phase 2 actions are required in the Reactor Building airlock area.

During refueling, many variables exist which impact the ability to cool the core. In the event of an ELAP during Refueling, there are no installed plant systems to cool the core available. To accommodate the activities of vessel disassembly and refueling, water levels in the reactor vessel and the reactor cavity are often changed. The most limiting condition is the case in which the reactor head is removed and water level in the vessel is at or below the reactor vessel flange. If an ELAP/LUHS occurs during this condition then, depending on the time after shutdown, boiling in the core may occur quite rapidly.

Deploying and implementation of portable FLEX pumps to supply injection flow must commence immediately from the time of the event. This rapid response is plausible because more personnel are on site during outages to provide the necessary resources.

Guidance will be provided to ensure that sufficient area is available for deployment and that haul paths remain accessible without interference from outage equipment during refueling outages.

References:

1. Cooper Nuclear Station Technical Specifications & Bases

2. Procedure 5.8 Emergency Operating Procedures (EOPs)

3. NLS9100631, Response to Recommendation on Station Blackout, 10CFR50.63

4. Cooper Nuclear Station Updated Safety Analysis Report

5. NEDC 91-261, Station Blackout with RCIC and RR Seal Leak, Revision 3

6. CNS-PSA-007, Cooper PRA Deterministic Calculations Notebook, Revision 1

7. Engineering Evaluation 01-147, Summary of Main Steam Isolation Valve (MSIV)

Leakage Pathway to the Condenser Seismic Qualification, Revision 2

8. 0000-0155-1545-Draft A, RCIC Pump and Turbine Durability Evaluation - Pinch Point Study Details:

Provide a brief Confirm that procedure/guidanceexists or will be developed to description of support implementation.

Procedures /

Strategies /

Guidelines.

Procedure 5.3SBO and the EOPs currently provide guidance to implement Phase 1.

Identify List modifications.

modifications NPPD will evaluate the following potential modifications to address this phase:

Upgrade ESCT level indicator piping to Seismic Class 1.

" Upgrade power supply and indicator for torus level.

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Revision 1 (February 15, 2014)

Key Reactor List instrumentationcreditedfor this coping evaluation.

Parameters Parameter Instrument Indicator Location RPV Level Narrow Range RFC-LI-94A, B and C (0 to 60") Control Room RPV Level Narrow Range NBI-LIS 83A and B (0 to 60") Local rack NBI-LIS-101A, B, C and D (0 to 60") Local rack RPV Level Wide Range NBI-LI-185B (-155" to + 60") Local ASDR RPV Level Wide Range NBI-LIS-57A and B (-150" to +60") Local rack NBI-LIS-58A and B (-150" to +60") Local rack NBI-LIS-72A, B, C and D (-150" to Local rack

+60")

RPV Level Fuel Zone NBI-LI-1 91 B (-320" to +60") Local ASDR RPV Level Fuel Zone NBI-LITS-73A and B -260" to +40") Local rack RPV Pressure RFC-PI-90A, B and C (0 to 1200 psig) Control Room RPV Pressure NBI-PIS-60A and B (0 to 1500 psig) Local rack NBI-PIS-52B and D (0 to 500 psig) Local rack NBI-PI-61 (0 to 1500 psig) Local rack Drywell Pressure PC-PI-513 (0 to 2 psig), PC-PI-2104AG Local rack (0 to 100 psig)

Torus Pressure PC-PI-20 (0 to 2 psig), PC-PI-2104BG Local rack (0 to 100 psig)

Drywell Temperature PC-TI-505A, B, C, D and E (50 to Control Room 600F)

Drywell Temperature PC-TE-505A,B, C, D and E, with use of Control Room M&TE (50 to 600F)

Torus Temperature PC-TE-1A thru H and PC-TE-2A thru H Local, Cable with use of M&TE (0 to 250F) Spreading Room Torus Temperature PC-TI-2A, C, E and G (0 to 250F) Local ASDR Torus Level PC-LI-1 10 (-4" to +6" H20) Local ASDR ECST Level CM-LI-1 681B (0 to 16' H20) Local ASDR ECST Level HPCI-PI- 17A, convert pressure to Local rack level Spent Fuel Pool Level TBD, NRC Order EA-12-051 Spent Fuel Pool Level Visual Local at SFP Spent Fuel Pool Temperature M&TE Local at SFP ASDR - Aux Shutdown Room TAF = -158" Notes: None.

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Maintain Core Cooling BWR Portable Equipment Phase 2:

Provide a general descriptionof the coping strategies using on-siteportable equipment including modifications that are proposed to maintain core cooling. Identify methods (RCIC/HPCI/IC)and strategy(ies) utilized to achieve this coping time.

Primary Strategy:

During Phase 2, as in Phase 1, reactor core cooling is maintained using RCIC in automatic mode (i.e., with operators controlling the RCIC flow controller) with suction from the ECST.

When ECST grade water becomes depleted, the ECST will be refilled from a yet to be installed on-site well. The well will be hardened, sized for the required makeup capacity and powered by a portable FLEX generator. The water will be pumped to the ECST connection point through hose. Alternatively, makeup for the ECST will be provided from the Missouri River through a series of portable strainers/filters/demineralizers to ensure sufficient water quality. The torus water level will be influenced both by the addition of water from the SRV discharge as well as evaporation through the RHCV which is placed in service at about 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months into the event to maintain containment parameters below design limits (see Maintain Containment response for additional discussion of the RHCV).

During Phase 2, reactor pressure is controlled by operation of SRVs as described in Phase 1 with the use of a portable FLEX pneumatic source.

The 125 VDC Division I batteries are available for approximately 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months without recharging.

Connecting the SAMG DG to Motor Control Center (MCC) LX or TX provides the ability to power Battery Chargers 125 VDC 'C' and 250 VDC 'C' which charge the 125 VDC Division I battery, the 250 VDC Division 1 battery and supply DC loads. The SAMG 480 VAC, 175 kW DG will be connected at approximately 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months and is sized to power two 125/250 VDC Battery Chargers and the DC Busses. Permanently installed cables will be installed from a point near the MCC to the exterior of the Control Building. The deployment area of the SAMG 480 VAC DG powering the 480 V MCC will be located near the Turbine Building (Reference 3). Cables from the generators are run to a connection point on the exterior of the Control Building.

Alternate Strategy:

Providing defense in depth for RCIC, a FLEX pump deployed at the river can provide RPV injection via the normal Residual Heat Removal Service Water (RHRSW) crosstie to the Residual Heat Removal (RHR) injection flow path. Piping will be installed to a connection point to the exterior of the Control Building. RPV pressure will be reduced to below the shutoff head of the FLEX pump after which the RHRSW to RHR flow path is established by opening valves SW-V-1 18 or 119, and SW-V-1 20 (Reference 4) to allow flow into the 24" line to the Low Pressure Coolant Injection valves, RHR-MO-27A and RHR-MO-25A (RHR-MO-27B and RHR-MO-25B for loop B). This provides the flow path into the RPV (References 1 and 2). The FLEX pump will supply water to the connection point via hose and strainer/filter/demineralizer to assure water quality.

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Revision 1 (February 15, 2014)

As an alternate strategy to power the battery chargers from their MCCs, connection points are planned to be installed on each of the 480V Busses. This will allow connecting power cables from a larger FLEX 480 VAC DG directly to the bus.

For Cold Shutdown and Refueling strategies see the discussion in the Phase 1 section above.

References:

1. B&R Drawing 2036, Reactor Building Service Water System

2. B&R Drawing 2040, RHR System

3. Procedure 2.2.100, SAMG Diesel Generator System

4. B&R Drawing 2006, Control Building SW System Details:

Provide a brief Confirm thatprocedure/guidanceexists or will be developed to description of support implementation.

Procedures /

Strategies /

Guidelines.

CNS will utilize the industry developed guidance from the Owners Groups, EPRI, and NEI Task team to develop site specific procedures or guidelines to address the criteria in NEI 12-

06. These procedures and/or guidelines will support the existing symptom based command and control strategies in the current EOPs.

Identify List modifications.

modifications NPPD will evaluate the following potential modifications to address this phase:

An external connection point on the exterior of the Control Building for connection of FLEX equipment.

RHRSW tie-in to external connection point.

New well for supplying makeup water.

Hotwell to ECST transfer.

Torus Wide Range Level Indication power supply.

'C' Chargers tie-in to external connection point.

RCIC mods to support running at high torus temp.

SRV pneumatic tie-in to external connection point.

480 FLEX Generator to 480 F and G Bus tie-ins.

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Revision 1 (February 15, 2014)

Key Reactor List instrumentationcredited or recoveredfor this coping evaluation.

Parameters Parameter Instrument Indicator Location RPV Level Narrow Range RFC-LI-94A, B and C (0 to 60") Control Room RPV Level Narrow Range NBI-LIS 83A and B (0 to 60") Local rack NBI-LIS-101A, B, C and D (0 to 60") Local rack RPV Level Wide Range NBI-LIS-57A and B (-150" to +60") Local rack NBI-LIS-58A and B (-150" to +60") Local rack NBI-LIS-72A, B, C and D (-150" to Local rack

+60")

RPV Level Fuel Zone NBI-LITS-73A and B -260" to +40") Local rack RPV Pressure RFC-PI-90A, B and C (0 to 1200 psig) Control Room RPV Pressure NBI-PIS-60A and B (0 to 1500 psig) Local rack NBI-PIS-52B and D (0 to 500 psig) Local rack NBI-PI-61 (0 to 1500 psig) Local rack Drywell Pressure PC-PI-513 (0 to 2 psig), PC-PI-2104AG Local rack (0 to 100 psig)

Torus Pressure PC-PI-20 (0 to 2 psig), PC-PI-2104BG Local rack (0 to 100 psig)

Drywell Temperature PC-TE-505A,B, C, D and E, with use of Control Room M&TE (50 to 600F)

Torus Temperature PC-TE-1A thru H and PC-TE-2A thru H Local, Cable with use of M&TE (0 to 250F) Spreading Room ECST Level HPCI-PI-1 17A, convert pressure to Local rack level Spent Fuel Pool Level TBD, NRC Order EA-12-051 Spent Fuel Pool Level Visual Local at SFP Spent Fuel Pool Temperature M&TE Local at SFP Containment/Torus Wide PC-LRPR-1AM1B (0 to 100ft) and (0 to Control Room Range Level 30ft)

TAF = -158" Storage / Protection of Equipment:

Describe storage / protection plan or schedule to determine storage requirements.

Seismic I List how equipment is protectedor schedule to protect.

Structures to provide protection of the FLEX equipment will be constructed to meet the requirements identified in NEI 12-06 Section 11. The schedule to construct the structures is still to be determined.

CNS procedures and programs are being developed to address storage structure requirements, haul path requirements, and FLEX equipment requirements relative to the hazards applicable to CNS.

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Flooding List how equipment is protectedor schedule to protect.

Note: if stored below current flood level, then ensure procedures exist to move equipment prior to exceeding flood level Not applicable per NEI 12-06 as outlined within the first section of this Integrated Plan.

Severe Storms with List how equipment is protectedor schedule to protect.

High Winds Structures to provide protection of the FLEX equipment will be constructed to meet the requirements identified in NEI 12-06 Section 11. The schedule to construct the structures is still to be determined.

CNS procedures and programs are being developed to address storage structure requirements, haul path requirements, and FLEX equipment requirements relative to the hazards applicable to CNS.

Snow, Ice, and List how equipment is protectedor schedule to protect.

Extreme Cold Structures to provide protection of the FLEX equipment will be constructed to meet the requirements identified in NEI 12-06 Section 11. The schedule to construct the structures is still to be determined.

CNS procedures and programs are being developed to address storage structure requirements, haul path requirements, and FLEX equipment requirements relative to the hazards applicable to CNS.

High Temperatures List how equipment is protectedor schedule to protect.

Structures to provide protection of the FLEX equipment will be constructed to meet the requirements identified in NEI 12-06 Section 11. The schedule to construct the structures is still to be determined.

CNS procedures and programs are being developed to address storage structure requirements, haul path requirements, and FLEX equipment requirements relative to the hazards applicable to CNS.

Deployment Conceptual Modification (Attachment 3 contains Conceptual Sketches)

Strategy Modifications Protection of connections Identify Strategy Identify modifications. Identify how the connection is including how the protected.

equipment will be deployed to the point of use.

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Tentative storage See above potential

An external connection locations have been modifications regarding point structure for FLEX proposed. Once these connection points. equipment is planned to be are finalized, clear constructed on the exterior deployment paths will be of the Control Building.

identified and controlled This will house all of the by administrative external connections to procedures to allow plant systems with the connection of the FLEX exception of the larger equipment to the FLEX 480 VAC generator.

external connection This structure will be point. designed to withstand the applicable hazards.

The connection points for the FLEX 480 VAC generator will be inside the existing Class I Critical SWGR Rooms. Pre-staged cables will be used to connect the generator to the connections.

New FLEX piping will be installed to meet necessary seismic requirements.

Notes: None.

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CNS Overall Integrated Plan for Diverse and Flexible Coping Strategies (FLEX)

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Maintain Core Cooling BWR Portable Equipment Phase 3:

Providea general description of the coping strategies usingphase 3 equipment including modifications that are proposed to maintain core cooling. Identify methods (RCIC/HPCI/IC)and strategy(ies) utilized to achieve this coping time.

For Phase 3, the reactor core cooling strategy is to place one loop of RHR into the Shutdown Cooling (SDC) mode. This will be accomplished by powering up a Division I or II RHR pump from the Class 1E emergency F or G 4160 VAC bus utilizing a 4160 VAC RRC FLEX portable diesel generator (i.e., from the RRC) and supplying the RHR Heat Exchanger with river water with a large portable RRC FLEX pump (i.e., from the RRC) from the Missouri River via the RHRSW piping external connection point.

The 4160 VAC RRC FLEX diesel generator will be capable of carrying approximately 3250 kW load which is sufficient to carry all of the loads on 4160 VAC bus F or G necessary to support the Phase 3 FLEX strategies which includes an RHR pump and its support equipment (i.e., Motor Operator Valves, jockey pump, room coolers, etc.). In order to prevent pipe damage due to water hammer, the Reactor Building Aux Condensate Pump will be repowered to allow proper venting prior to RHR shutdown cooling operation. The primary strategy is provided by RHR A(B) and the secondary strategy is provided by RHR B(A).

Alternative Strategqy:

Alternate means of core cooling can be provided by connecting to and using the opposite division of RHR and RHRSW as that used for the primary function.

An alternate means of providing power to the RHR pumps for SDC operation is to run cable from the 4160 VAC RRC DG directly to the component by connecting either at the switchgear end of the component's power cable or locally at the pump end of the power cable.

Details:

Provide a brief Confirm thatprocedure/guidanceexists or will be developed to description of support implementation.

Procedures /

Strategies /

Guidelines.

CNS will utilize the industry developed guidance from the Owners Groups, EPRI, and NEI Task team to develop site specific procedures or guidelines to address the criteria in NEI 12-06. These procedures and/or guidelines will support the existing symptom based command and control strategies in the current EOPs.

Identify List modifications.

modifications Page 19 of 51

CNS Overall Integrated Plan for Diverse and Flexible Coping Strategies (FLEX)

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NPPD will evaluate the following potential modifications to address this phase:

Modification for connection of 4160 VAC RRC DG to the F or G 4160 VAC bus is made by modifying a spare breaker in the warehouse to be able to accept the cable connections from the RRC DG such that in Phase 3 that breaker can be inserted into a spare slot or non-required breaker to connect the DG to the bus. (Primary Strategy)

Modification to allow re-powering the Reactor Building Aux Condensate Pump.

Key Reactor List instrumentationcredited or recoveredfor this coping evaluation.

Parameters Parameter Instrument Indicator Location RPV Level Narrow Range RFC-LI-94A, B and C (0 to 60") Control Room RPV Level Narrow Range NBI-LIS 83A and B (0 to 60") Local rack NBI-LIS-101A, B, C and D (0 to 60") Local rack RPV Level Wide Range NBI-LI-185B (-155" to + 60") Local ASDR RPV Level Wide Range NBI-LIS-57A and B (-150" to +60") Local rack NBI-LIS-58A and B (-150" to +60") Local rack NBI-LIS-72A, B, C and D (-150" to Local rack

+60")

RPV Level Fuel Zone NBI-LI-191B (-320" to +60") Local ASDR RPV Level Fuel Zone NBI-LITS-73A and B -260" to +40") Local rack RPV Pressure RFC-PI-90A, B and C (0 to 1200 psig) Control Room RPV Pressure NBI-PIS-60A and B (0 to 1500 psig) Local rack NBI-PIS-52B and D (0 to 500 psig) Local rack NBI-PI-61 (0 to 1500 psig) Local rack Drywell Pressure PC-PI-513 (0 to 2 psig), PC-PI-2104AG Local rack (0 to 100 psig)

Torus Pressure PC-PI-20 (0 to 2 psig), PC-PI-.2104BG Local rack (0 to 100 psig)

Drywell Temperature PC-TI-505A, B, C, D and E (50 to Control Room 600F)

Drywell Temperature PC-TE-505A,B, C, D and E, with use of Control Room M&TE (50 to 600F)

Torus Temperature PC-TE-1A thru H and PC-TE-2A thru H Local, Cable with use of M&TE (0 to 250F) Spreading Room Torus Temperature PC-TI-2A, C, E and G (0 to 250F) Local ASDR Torus Level PC-LI-1 10 (-4" to +6" H20) Local ASDR ECST Level CM-LI-1 681B (0 to 16' H20) Local ASDR ECST Level HPCI-PI- 17A, convert pressure to Local rack level Spent Fuel Pool Level TBD, NRC Order EA-12-051 Spent Fuel Pool Level Visual Local at SFP Spent Fuel Pool Temperature M&TE Local at SFP ASDR - Aux Shutdown Room TAF = -158" Page 20 of 51

CNS Overall Integrated Plan for Diverse and Flexible Coping Strategies (FLEX)

Revision 1 (February 15, 2014)

Deployment Conceptual Modification (Attachment 3 contains Conceptual Sketches)

Strategy Modifications Protection of connections Identify Strategy Identify modifications Identify how the connection is including how the protected equipment will be deployed to the point of use Phase 3 equipment will See above potential See Phase 2 for the external be provided by the modifications regarding connection point for RHRSW.

RRC which is to be connection points. The,4160 RCC FLEX DG located in Memphis, connection will be in the TN. Equipment existing Class I Critical SWGR transported to the site Rooms.

will be either immediately staged at the point of use location (pumps and generators) or at a staging area yet to be determined.

Notes: None.

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Maintain Containment Determine Baseline coping capability with installed coping 2 modifications not including FLEX modifications, utilizing methods described in Table 3-1 of NEI 12-06:

Containment Venting or Alternate Heat Removal

Hydrogen Igniters (Mark III containments only)

BWR Installed Equipment Phase 1:

Providea general description of the coping strategies using installed equipment including modifications that are proposed to maintain core cooling. Identify methods (containment vent or alternative/Hydrogen Igniters) and strategy(ies) utilized to achieve this coping time.

During Phase 1, containment integrity is maintained by normal design features of the containment, such as the containment isolation valves and RHCV. In accordance with NEI 12-06 (Reference 1), the containment is assumed to be isolated following the event. As the torus heats up and the water begins to boil, the containment will begin to heat up and pressurize. Additionally, the torus level rises due to the transfer of inventory from the ECST to the torus (via RCIC and SRVs). According to BWROG analysis (Reference 2) containment parameters can be controlled within design limits by utilization of the RHCV and venting the containment. In this case, the RHCV is used as implemented per EA-12-050, Reliable Hardened Containment Vents (Reference 4) with control from the Main Control Room (MCR). CNS has performed a Probabilistic Risk Assessment level MAAP analysis to validate the strategy in the BWROG analysis (Reference 3) and will perform additional calculations to establish the exact timing and duration of containment venting.

The containment design pressure is 62 psig (Reference 5, Section V-2.3). Containment pressure limits are not expected to be reached during the event as indicated by MAAP analysis (Reference 3), because the RHCV will be opened prior to exceeding any containment pressure limits.

Phase 1 (i.e., the use of permanently installed plant equipment/features) of containment integrity will be maintained throughout the duration of the event; no non-permanently installed equipment will be required to maintain containment integrity. Therefore, there is no defined end time for the Phase 1 coping period for maintaining containment integrity. An alternative strategy for containment during Phase 1 is not provided, because containment integrity is maintained by the plant's design features.

References:

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

2. NEDC 33771 P, GEH Evaluation of FLEX Implementation Guidelines, Revision 1

3. NEDC 13-004, CNS Evaluation of Diverse and Flexible Coping Strategies for Extended Loss of AC Power

4. Interim Staff Guidance JLD-ISG-2012-02, Compliance with Order EA-12-050, Reliable Hardened Containment Vents, Revision 0

5. Cooper Nuclear Station Updated Safety Analysis Report 2 Coping modifications consist of modifications installed to increase initial coping time, i.e. generators to preserve vital instruments or increase operating time on battery powered equipment.

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Revision 1 (February 15, 2014)

Details:

Provide a brief Confirm thatprocedure/guidanceexists or will be developed to description of support implementation.

Procedures /

Strategies /

Guidelines.

CNS will utilize the industry developed guidance from the Owners Groups, EPRI, and NEI Task team to develop site specific procedures or guidelines to address the criteria in NEI 12-

06. These procedures and/or guidelines will support the existing symptom based command and control strategies in the current EOPs.

Identify modifications List modifications.

Hardened Containment Vent System (i.e., Reliable Hardened Vent) is currently installed but will be enhanced in accordance with NRC Order EA-12-050, Issuance of Order to Modify Licenses with Regard to Reliable Hardened Containment Vents.

Key Containment List instrumentation creditedfor this coping evaluation.

Parameters Parameter Instrument Indicator Location Drywell Pressure PC-PI-513 (0 to 2 psig), PC-PI- Local rack 2104AG (0 to 100 psig)

Torus Pressure PC-PI-20 (0 to 2 psig), PC-PI- Local rack 2104BG (0 to 100 psig)

Drywell Temperature PC-TI-505A, B, C, D and E (50 to Control Room 600F)

Drywell Temperature PC-TE-505A,B, C, D and E, with use Control Room of M&TE (50 to 600F)

Torus Temperature PC-TE-1A thru H and PC-TE-2A Local, Cable thru H with use of M&TE (0 to 250F) Spreading Room Torus Temperature PC-TI-2A, C, E and G (0 to 250F) Local ASDR Torus Level PC-LI-1 10 (-4" to +6" H20) Local ASDR Hard Pipe Vent Open Indication PC-RIL-SPV32 Control Room Hard Pipe Vent Closed PC-GIL-SPV32 Control Room Indication PC 233MV Open Indication PC-RIL-MV233 Control Room PC 233 MV Closed Indication PC-GIL-MV233 Control Room PC 237AV Open Indication PC-RIL-SPV237 Control Room PC 237AV Closed Indication PC-RIL-SPV237 Control Room ASDR - Aux Shutdown Room Notes: None.

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CNS Overall Integrated Plan for Diverse and Flexible Coping Strategies (FLEX)

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Maintain Containment BWR Portable Equipment Phase 2:

Provide a general description of the coping strategies using on-site portable equipment including modifications that are proposed to maintain core cooling. Identify methods (containment vent or alternative /Hydrogen Igniters) and strategy(ies) utilized to achieve this coping time.

Containment integrity is maintained by permanently installed equipment. Portable FLEX diesel generators will be employed, as discussed in Phase 2 Core Cooling section, to charge the station batteries and maintain DC bus voltage. See Phase 1 description for discussion of containment integrity applicable throughout the event.

Details:

Provide a brief Confirm that procedure/guidanceexists or will be developed to description of support implementation.

Procedures /

Strategies /

Guidelines.

CNS will utilize the industry developed guidance from the Owners Groups, EPRI, and NEI Task team to develop site specific procedures or guidelines to address the criteria in NEI 12-

06. These procedures and/or guidelines will support the existing symptom based command and control strategies in the current EOPs.

Identify List modifications.

modifications See modifications listed in Phase 2 discussion.

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Revision 1 (February 15, 2014)

Key Containment List instrumentationcredited or recoveredfor this coping evaluation.

Parameters Parameter Instrument Indicator Location Drywell Pressure PC-PI-513 (0 to 2 psig), PC-PI-2104AG Local rack (0 to 100 psig)

Torus Pressure PC-PI-20 (0 to 2 psig), PC-PI--2104BG Local rack (0 to 100 psig)

Drywell Temperature PC-TI-505A, B, C, D and E (50 to Control Room 600F)

Drywell Temperature PC-TE-505A,B, C, D and E, with use of Control Room M&TE (50 to 600F)

Torus Temperature PC-TE-1A thru H and PC-TE-2A thru H Local, Cable with use of M&TE (0 to 250F) Spreading Room Hard Pipe Vent Open PC-RIL-SPV32 Control Room Indication Hard Pipe Vent Closed PC-GIL-SPV32 Control Room Indication PC 233MV Open Indication PC-RIL-MV233 Control Room PC 233 MV Closed Indication PC-GIL-MV233 Control Room PC 237AV Open Indication PC-RIL-SPV237 Control Room PC 237AV Closed Indication PC-RIL-SPV237 Control Room Storage / Protection of Equipment:

Describe storage / protection plan or schedule to determine storage requirements Seismi List how equipment is protectedor schedule to protect.

Structures to provide protection of the FLEX equipment will be constructed to meet the requirements identified in NEI 12-06 Section 11. The schedule to construct the structures is still to be determined.

CNS procedures and programs are being developed to address storage structure requirements, haul path requirements, and FLEX equipment requirements relative to the hazards applicable to CNS.

Flooding List how equipment is protected or schedule to protect.

Note: if stored below current flood level, then ensure procedures exist to move equipment prior to exceeding flood level.

Not applicable per NEI 12-06 as outlined within the first section of this Integrated Plan.

Severe Storms with List how equipment is protectedor schedule to protect.

High Winds Page 25 of 51