RA-15-008, Fourth 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)

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Fourth 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)
ML15058A378
Person / Time
Site: Oyster Creek
Issue date: 02/27/2015
From: Jim Barstow
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
EA-12-049, RA-15-008, RS-15-022
Download: ML15058A378 (24)
v • d • e


Text

Order No. EA-12-049 RS-15-022 RA-15-008 February 27, 2015 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NBC Docket No. 50-219

Subject:

Fourth 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, "Issuance of Order to Modify 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 O, dated August 2012

4. Exelon Generation Company, LLC's 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 25, 2012
5. Exelon Generation Company, LLC 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 (RS-13-023)
6. Exelon Generation Company, LLC 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 28, 2013 (RS-13-125)
7. Exelon Generation Company, LLC 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 049), dated February 28, 2014 (RS-14-013)
8. Exelon Generation Company, LLC Third 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 28, 2014 (RS-14-211)

U.S. Nuclear Regulatory Commission Integrated Plan Report to EA-12-049 February 27, 2015 Page 2

9. NRC letter to Exelon Generation Company, LLC, Oyster Creek Nuclear Generating Station - Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) (TAC No. MF0824), dated February 19, 2014 On March 12, 2012, the Nuclear Regulatory Commission ("NRC" or "Commission") issued an order (Reference 1) to Exelon Generation Company, LLC (EGC). Reference 1 was immediately effective and directs EGC 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 of Reference 1.

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 NEI 12-06, Revision O (Reference 3) with clarifications and exceptions identified in Reference 2. Reference 4 provided the EGC initial status report regarding mitigation strategies. Reference 5 provided the Oyster Creek Nuclear Generating Station overall integrated plan.

Reference 1 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. References 6, 7, and 8 provided the first, second, and third six-month status reports, respectively, pursuant to Section IV, Condition C.2, of Reference 1 for Oyster Creek Nuclear Generating Station. The purpose of this letter is to provide the fourth 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. The enclosed report 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. The enclosed report also addresses the NRC Interim Staff Evaluation Open and Confirmatory Items contained in Reference 9.

This letter contains no new regulatory commitments. If you have any questions regarding this report, please contact David P. Helker at 610-765-5525.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 27'h day of February 2015.

Respectfully submitted, James Barstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC

U.S. Nuclear Regulatory Commission Integrated Plan Report to EA-12-049 February 27, 2015 Page 3

Enclosure:

1. Oyster Creek Nuclear Generating Station Fourth 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 cc: Director, Office of Nuclear Reactor Regulation NRC Regional Administrator - Region I NRC Senior Resident Inspector - Oyster Creek Nuclear Generating Station NRC Project Manager, NRR - Oyster Creek Nuclear Generating Station Ms. Jessica A. Kratchman, NRR/JLD/PMB, NRC Mr. Jack R. Davis, NRR/DPR/MSD, NRC Mr. Eric E. Bowman, NRR/DPR/MSD, NRC Mr. Jeremy S. Bowen, NRR/DPR/MSD/MSPB, NRC Mr. Robert L. Dennig, NRR/DSS/SCVB, NRC Mr. John D. Hughey, NRR/JLD/JOMB, NRC Manager, Bureau of Nuclear Engineering - New Jersey Department of Environmental Protection Mayor of Lacey Township, Forked River, NJ

Enclosure Oyster Creek Nuclear Generating Station Fourth 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 (8 pages)

Oyster Creek Nuclear Generating Station Fourth Six Month Status Report for the Implementation of FLEX February, 2015 Enclosure Oyster Creek Nuclear Generating Fourth 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 Oyster Creek Nuclear Generating Station developed an Overall Integrated Plan (Reference 1),

documenting the diverse and flexible strategies (FLEX), in response to NRC Order Number EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Reference 2). This enclosure provides an update of milestone accomplishments since submittal of the last status report including any changes to the compliance method, schedule, or need for relief I relaxation and the basis, if any.

2. Milestone Accomplishments The Third 6 Month Update was submitted in August 2014.
3. Milestone Schedule Status The following 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 Schedule Revised Target Target Activity Activity Status Completion Completion Date Date Submit 60 Day Status Report October 2012 Complete Submit Overall Integrated Plan February 2013 Complete Contract with RRC Complete Submit 6 Month Updates:

Update 1 August 2013 Complete Update 2 February 2014 Complete Update 3 August 2014 Complete Complete with Update 4 February 2015 this submittal Update 5 August 2015 Not Started Update 6 February 2016 Not Started Update 7 August 2016 Not Started Page 1of8

Oyster Creek Nuclear Generating Station Fourth Six Month Status Report for the Implementation of FLEX February, 2015 Revised Target Target Activity Activity Status Completion Completion Date Date Submit Completion Report October 2016 Not Started Target Revised Target Activity Completion Activity Status Completion Date Date Modification Development Revised in Modification Development Started August 2013 August 2016 (All FLEX Phases) update Modification Implementatio n No Change October 2016 Not Started (All FLEX Phases)

Procedures:

Create Site-Specific Procedures October 2016 Started No Change Validate Procedures No Change October 2016 Not Started (NEI 12-06, Sect. 11.4.3)

Create Maintenance Procedures October 2016 Started No Change Perform Staffing Analysis June 2016 Not Started No Change Storage Plan and Construction October 2016 Started No Change FLEX Equipment Acquisition October 2016 Started No Change Training Completion October 2016 Not Started No Change Revised in Regional Response Center Operational December 2015 Started August 2014 update Unit 1 FLEX Implementatio n October 2016 Not Started No Change Full Site FLEX Implementatio n October 2016 Not Started No Change Page 2 of 8

Oyster Creek Nuclear Generating Station Fourth Six Month Status Report for the Implementation of FLEX February, 2015

4. Changes to Compliance Method None
5. Need for Relief/Relaxation and Basis for the Relief/Relaxation None
6. Open Items and Confirmatory items from Overall Integrated Plan and Interim Safety Evaluation The following tables provide a summary of the open and Confirmatory items documented in the Overall Integrated Plan or the Draft Safety Evaluation (SE) and the status of each item.

Section Overall Integrated Plan Open Items Status Reference Sequence of The times to complete actions in the Events Not Started events Timeline are based on operating judgment, (p. 10-12) conceptual designs, and current supporting analyses. The final timeline will be time validated once detailed designs are completed and procedures developed.

Sequence of Initial evaluations were used to determine the Started events fuel pool timelines. Formal calculations will be (p. 11-12) performed to validate this information during development of the spent fuel pool cooling strategy detailed design.

Identify how 1. Transportation routes will be developed 1. Started strategies will from the equipment storage area to the 2. Started be deployed in FLEX staging areas. 3. Not Started all modes 2. Identification of storage areas is an open (p. 13) item.

3. An administrative program will be developed to ensure pathways remain clear or compensatory actions will be implemented to ensure all strategies can be deployed during all modes of operation.

Identify how An administrative program for FLEX to Started the establish responsibilities, and testing &

programmatic maintenance requirements will be controls will implemented.

be met (p. 14)

Page 3 of 8

Oyster Creek Nuclear Generating Station Fourth Six Month Status Report for the Implementation of FLEX February, 2015 Maintain Complete an evaluation of the spent fuel pool Started Spent Fuel area for steam and condensation.

Pool Cooling (p.36)

Safety Evaluate the habitability conditions for the Started Functions Main Control Room and develop a strategy to Support maintain habitability.

(p. 44)

Safety Develop a procedure to prop open battery Tracked in Interim Safety Evaluation Functions room doors upon energizing the battery Confirmatory Items reference section Support chargers to prevent a buildup of hydrogen in 3.2.4.2.A.

(p. 44) the battery rooms.

Sequence of Issuance of BWROG document NEDC-33771P, Completed events (p. 10) "GEH Evaluation of FLEX Implementation Oyster Creek Station First Six Month Guidelines" on 01/31/2013 did not allow Update. (ML13240A267) sufficient time to perform the analysis of the deviations between Exelon's engineering analyses and the analyses contained in the BWROG document prior to commencing regulatory reviews of the Integrated Plan.

Baseline In response to NRC Order EA-12-049 and Not Started.

coping implementation of EPG Rev 3, containment capability venting will be part of the strategies. As part (p. 27) of the B.5.b response Oyster Creek incorporated Extensive Damage Mitigation Guidelines and developed procedure EDMG-SPX9 Manually Opening Containment Vent Valves in a B.5.b Event. This procedure is designed to allow operation of the Hardened Vents with no air supply, AC or DC power available.

Convert EDMG-SPX9 Manually Opening Containment Vent Valves to FSG procedure.

Section Interim Safety Evaluation Open Items Status Reference None None NA Page 4 of 8

Oyster Creek Nuclear Generating Station Fourth Six Month Status Report for the Implementation of FLEX February, 2015 Section Interim Safety Evaluation Confirmatory Items Status Reference 3.1.1.3.A Confirm that the licensee develops a reference Started source describing what actions should be taken if instruments were lost due to a seismic event.

3.1.1.4.A Confirm the location of the off-site staging area(s) Not Started and acceptability of the access routes considering the seismic, flooding, high wind and snow, ice and extreme cold hazard.

3.1.2.2.A Confirm that if temporary flood barriers are used, Complete they are stored such that they can be easily Procedure OP-OC-108-109-1001, deployed. Severe Weather Preparation T&RM for Oyster Creek was revised to credit sandbags as the temporary flood barriers for the Near Term Task Force (NTIF) reevaluated results.

3.1.3.1.A Verify that the separation of the planned outdoor Started storage areas is sufficient to preclude damage of both sets of FLEX equipment.

3.1.3.1.B Confirm qualified storage locations for the Started hurricane and extreme snow and icing hazards are identified.

3.1.3.2.A Confirm that the licensee's evaluation of water Started quality and resulting action are sufficient to preclude blockage of flow to the core or SFP.

3.2.1.1.A Confirm that benchmarks are identified and Complete discussed that demonstrate that MAAP is an Modular Accident Analysis Program appropriate code for the simulation of an HAP (MAAP) Justification 11385-467 is event at your facility. accepted by the site and entered into The site's Electronic Document Management System (EDMS) as OC-MISC-012.

(See Attachment 1)

Page 5 of 8

Oyster Creek Nuclear Generating Station Fourth Six Month Status Report for the Implementation of FLEX February, 2015 3.2.1.1.B Confirm that the collapsed level remains above Complete Top of Active Fuel {TAF) and the cool down rate MAAP Justification 11385-467 is remains within technical specifications limits. accepted by the site and entered into EDMS as OC-MISC-012.

{See Attachment 1) 3.2.1.1.C Confirm that MAAP is used in accordance with Complete Sections 4.1, 4.2, 4.3, 4.4, and 4.5 of the June 2013 MAAP Justification 11385-467 is position paper. accepted by the site and entered into EDMS as OC-MISC-012.

(See Attachment 1) 3.2.1.1.D Confirm that the licensee identifies and justifies Complete the subset of key modeling parameters cited from MAAP Justification 11385-467 is 11 Tables 4-1 through 4-6 of the MAAP Application accepted by the site and entered into Guidance, Desktop Reference for Using MAAP EDMS as OC-MISC-012.

Software, Revision 2" (Electric Power Research Institute Report 1020236). This should include {See Attachment 1) 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 that licensee's plant. Although some suggested key phenomena are identified below, other parameters considered important in the simulation of the ELAP event by the vendor I licensee should also be included.

a. Nodalization
b. General two-phase flow modeling
c. Modeling of heat transfer and losses
d. Choked flow
e. Vent line pressure losses
f. Decay heat {fission products I actinides I etc.)

Page 6 of 8

Oyster Creek Nuclear Generating Station Fourth Six Month Status Report for the Implementation of FLEX February, 2015 3.2.1.1.E Confirm that the specific MAAP analysis case that Complete was used to validate the timing of mitigating MAAP Justification 11385-467 is strategies in the Integrated Plan is identified and accepted by the site and entered into available on the ePortal for NRC staff to view.

EDMS as OC-MISC-012.

Alternately, a comparable level of information may be included in the supplemental response. In (See Attachment 1) either case, the analysis should include a plot of the collapsed vessel level to confirm that TAF is not reached (the elevation of the TAF should be provided) and a plot of the temperature cool down to confirm that the cool down is within technical specifications limits.

3.2.1.3.A The SOE final timeline will be time validated once Not Started detailed designs are completed and procedures are developed. The licensee should provide the results for NRC staff review.

3.2.4.2.A The licensee stated that battery room ventilation Not Started to address high/low temperatures and prevention of hydrogen buildup will be addressed through procedure changes and that the proposed methods of ventilation, open doors and fans, will be confirmed during the detailed design process.

3.2.4.4.A The NRC staff has reviewed the licensee Started communications assessment (ADAMS Accession Nos. ML12306A199 and ML13056A135) in response to the March 12, 2012 50.54(f) request for information letter for OCNGS and, as documented in the staff analysis (ADAMS Accession No. ML13114A067) has determined that the assessment for communications is reasonable, and the analyzed existing systems, proposed enhancements, and interim measures will help to ensure that communications are maintained.

Verification of required upgrades has been identified as a confirmatory item.

3.2.4.8.A Confirm the procedures to isolate the vital USS's Not Started from the generator.

3.2.4.8.A Ensure that the diesel generator is equipped with Started overload protection in the generator skid.

3.2.4.8.B Confirm/review technical basis and/or Started calculations provided as basis for the generator sizing.

Page 7 of 8

Oyster Creek Nuclear Generating Station Fourth Six Month Status Report for the Implementation of FLEX February, 2015 3.2.4.10.A Confirm completion of analysis to determine Started battery coping time with no actions and with battery load shed.

3.4.A NEI 12-06, Section 12.2 lists minimum capabilities Not Started for offsite resources for which each Licensee should establish the availability. Confirm implementatio n of Guidelines 2 through 10 in NEI 12-06, Section 12.2.

7. Potential Draft Safety Evaluation Impacts There are no potential impacts to the Draft Safety Evaluation identified at this time.
8. References The following references support the updates to the Overall Integrated Plan described in this enclosure.
1. Oyster Creek Nuclear Generating Station'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.
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.
9. Attachment OC-MISC-012, Rev 000, 20140408, Use of MAAP In Support of Flex Implementatio n (12 pages)

Page 8 of 8 OC MISC-012, Rev 000, 20140408, Use of MAAP in RM DOCUMENTATION NO. OC-MISC-012 REV: 0 PAGE NO. 1 STATION: Oyster Creek UNIT(S) AFFECTED: Unit 1 TITLE: Use of MAAP in Support of FLEX Implementation

SUMMARY

(Include UREs incorporated):

MAAP 4.0.5 calculations (OC-MISC-010-R1) were performed to estimate the containment pressure and temperature response to a variety of extended Station Blackout (SBO) events. The NRC has requested that some additional information be provided relating to the use of MAAP for FLEX analysis as part of the periodic update to the plants response to EA-12-049. The attached information is being provided to include in the next update to EA-12-049.

[ ] Review required after periodic Update

[ X ] Internal RM Documentation [ ] External RM Documentation Electronic Calculation Data Files: N/A Method of Review: [ X] Detailed [ ] Alternate [ ] Review of External Document This RM documentation supersedes: NIA in its entirety.

Alex H. Duvall Prepared by: -------*--

Print I OPx 13~ Sign I 'I /1 /1~

Date Reviewed by: Gary W. Ha~ner Print I

~ I 4-L1!1f 1

Bate Approved by: Edward T. Burns I rm~ I 1-/r?/l'f Print Sign Date

Attachment I OC MISC-012, Rev 000. 20140408. Use of MAAP in T AHLE OF CONTENTS Section Page 1.0 PURPOSE & SCOPE .............................................................................................. 3 2.0 REQUESTED INFORMATION ON THE USE OF MAAP ....................................... 3 OC-MISC-012-RO.doc 2

Attachment I OC MISC-012, Rev 000, 201 Use of MAAP in to EA-12-049 1.0 PURPOSE & SCOPE The purpose of the included information is to respond to NRC questions relating to the use of MAAP in support of the plant's response to EA-12-049. The MAAP analysis is documented separately in OC-MISC-010-R I.

2.0 REQUESTED INFORMATIO N ON THE USE OF MAAP In response to the letter of October 3, 2013 from Jack Davis (NRR) to Joe Pollock (NEI), the following responses have been developed regarding the use of the Modular Accident Analysis Program (MAAP) for estimating accident progression timing in support of the Overall Integrated Plan for Oyster Creek.

(1) From the June 2013 position paper, benchmarks must be identified and discussed which demonstrate that MAAP4 is an appropriate code for the simulation of an ELAP event at your facility.

Response to item 1:

Generic response provided by EPRI Technical Report 3002002749, "Technical Basis for Establishing Success Timelines in Extended Loss of AC Power Scenarios in Boiling Water Reactors Using MAAP4,"

A Guide to MAAP Thermal-Hydraulic Models".

(2) The collapsed level must remain above Top of Active Fuel (TAF) and the cool down rate must be within technical specification limits.

Response to item 2:

Attachment IA of the Oyster Creek Integrated Plan (Feb. 2013) states that the operators would commence a cooldown of the RPV at I 0 min at a rate of 50°F/hr which is within the technical specifications limit of 100 °F/hr. The following plot of the RPV pressure from the MAAP analysis confirms this cooldown rate for the supporting MAAP calculation.

OC-MTSC-012-RO.doc 3

Attachment I OC MISC-01 Rev 000. 20140408. Use of MAAP in 1200 U5 1000

-en Cl..

Cl.. 800

-a:w CL.

600 en w

en a: 400 CL.

>CL.

a: 200 0

0 2 4 6 8 10 TIME, HOURS MAAP Calculation of RPV Pressure During RPV Depressurization For the representative MAAP run (Case 6), the collapsed RPV water level inside the shroud remains above or near Top of Active Fuel (T AF) for the duration of the analysis. The plot below shows that the lowest RPV level, calculated by MAAP, was at approximately O" relative to instrument zero which is 353" above vessel zero. TAF is located at instrument zero. As shown in the following plot, the collapsed RPV water level in the shroud briefly drops below TAF. The collapsed water level remains near T AF before rapidly increasing to more than 170" above T AF for the remainder of the scenario.

OC-MISC-012-RO.doc 4

Attachment I OC MISC-OJ Use of MAAP in 250 WO J: a:

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-50 0 20 40 60 80 TIME, HOURS MAAP Calculation of Collapsed RPV Water Level Inside the Shroud (TAF = 0")

As shown in the following plot, the collapsed RPV water level inside the shroud remains at least 25" above T AF for the duration of the analysis. While the collapsed water level in the shroud may briefly drop below T AF, the boiled-up water level in the core remains adequately above T AF during the scenario.

Since the core remains covered by the boiled-up water level and >95% covered by the collapsed level, core damage does not occur. Additionally, since the collapsed water level is calculated as dropping to and hovering near T AF for only a few minutes, conditions do not exist for the fuel to rapidly heat up to harmful temperatures.

OC-MISC-012-RO.doc 5 OC MISC-OJ 2, Rev 000, 20140408, Use of MAAP in Support of OC Response to EA-12-049 250 0

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0 20 40 60 80 HOURS MAAP Calculation of Boiled-UP RPV Water Level Inside the Shroud (3) MAAP4 must be used in accordance with Sections 4. 1, 4.2, 4.3, 4.4, and 4.5 of the June 2013 position paper.

Response to item 3:

MAAP analysis performed for Oyster Creek was can"ied out in accordance with Sections 4.1, 4.2, 4.3, 4.4, and 4.5 of the June 2013 position paper, EPRI Technical Report 3002001785, "Use of Modular Accident Analysis Program (MAAP) in Support of Post-Fukushima Applications". Preparation and Review of the MAAP analysis is conducted under engineering training certification guide ENANRM08.

OC-MISC-01 2-RO.doc 6 OC MISC-012, Rev Use ofMAAP in (4) In using MAAP4, the licensee must identify and justify the subset of key modeling parameters cited from Tables 4-1 through 4-6 of the "MAAP4 Application Guidance, Desktop Reference for Using MAAP4 Software, Revision 2" (Electric Power Research Institute Report 1020236). This should include response at a plant-specific level regarding specific coding options and parameter choices for key models that would be expected to substantially affect the ELAP analysis performed for that licensee's plant. Although some suggested key phenomena are identified below, other parameters considered important in the simulation of the ELAP event by the vendor I licensee should also be included.

a. Nodalization
b. General two-phase flow modeling
c. Modeling of heat transfer and losses
d. Choked.flow
e. Vent line pressure losses f Decay heat (fission products I actinides I etc.)

Response to item 4:

a. The reactor vessel nodalization is fixed by the MAAP code and cannot be altered by the user, with the exception of the detailed core nodalization. The Oyster Creek MAAP 4.0.5 parameter file divides the core region into 5 equal volume radial regions (See NCHAN) and 13 axial regions (See NAXNOD).

The axial nodalization represents 8 equal-sized fueled nodes (see NROWS), 1 unfueled node at the top (see NNFT), and 2 unfueled nodes at the bottom (see NNFB). The figure below, taken from the MAAP Users Manual, illustrates the vessel nodalization scheme.

OC-MISC-0 l 2-RO.doc 7

Attachment I OC MlSC-012, Rev 000, 201 Use of MAAP in i:::I Heat Sink 7 Lower Head Containment nodalization is defined by the user. The standard nodalization scheme is used in the Oyster Creek MAAP 4.0.5 parameter file and represents the following individual compartments:

1. Reactor pedestal region
2. Drywell
3. Drywell vents to torus
4. Torus (Wetwell)

OC-MISC-012-RO.doc 8

Attachment I OC M ISC-0 I Rev 000, 20140408, Use of MAAP in to EA-12-049 The figure below illustrates the Oyster Creek containment nodali1,ation along with an identification of containment flow junctions.

1 Pedestal Door to Drywell 2 Drywell to Vents/Downcomers 3 Vents/Downcomers to Suppression Pool Vesse Vent 4 Drywell 4 Vent/Failure 5,6,7,8 Suppression Leakage Chamber to Vent Vacuum Breakers 9 Suppression Chamber Vent/Failure Shell Failure 33 10 Drywell Leakage 33 Drywell Shell 2 Failure Vent/Failure 5,6,7,8 9 3

General two-phase flow from the reactor vessel is described in the EPRI Technical Report 3002002749.

In the case of the scenario outlined in the integrated plan, flow can ex_it the RPV via the open SRV(s) and from the assumed recirculation pump seal leakage. Flow from the SRV(s) will be single-phase steam and flow from the recirc pump seal or other RPV leakage will be single-phase liquid due to the location of the break low in the RPV with RPV level maintained above T AF. Upon exiting the RPV, the seal leakage will flash a portion of the flow to steam based on saturated conditions in the drywell, creating a steam source and a liquid water source to the drywell. As described in the EPRI Technical Repo1i 3002002749, "Technical Basis for Establishing Success Timelines in Extended Loss of AC Power Scenarios in Boiling Water Reactors Using MAAP4 A Guide to MAAP Thermal-Hydraulic Models", there are two MAAP parameters that can influence the two-phase level in the RPV- FCO (void concentration factor) and OC-MISC-012-RO.doc 9

Attachment I OC MISC-OJ Rev 000, 201 FCHTUR (churn-turbulent critical velocity coefficient). The following table confirms that the parameter values match the recommended values as outlined in the EPRI Technical Report 3002002749.

VALUE USED IN THE OYSTER CREEK MAAP EPRI PARAMETER NAME ANALYSIS RECOMMENDED VALUE FCO 1.5248 1.5248 FCHTUR l.53 l.53

b. Modeling of heat transfer and losses from the RPV are described in the EPRI Technical Report 3002002749. The MAAP parameters that control these processes, as defined in the EPRI report, are provided below with the values selected to represent Oyster Creek.

VALUE USED IN THE OYSTER CREEK MAAP PARAMETER NAME ANALYSIS COMMENT QCO noHlmi-insulation heat 3.4122E6 BTU/hr Plant specific value based on transfer from RPV during drywell heat removal to coolers normal operation. during normal operation. Typical values range between 1-2 MW (3.4E6 to 6.8E6 BTU/hr).

FINPLT - number of plates in 8.0 Plant-specific value reflective insulation XTINS average reflective 0.333 ft Plant-specific value insulation thickness OC-MISC-012-RO.doc 10

Attachment I OC MISC-OJ Use of MAAP in to EA-12-049 At the request or the NRC, the following information, as used in the MAAP analysis, is provided.

VALUE USED IN THE PARAMETER OYSTER CREEK MAAP PARAMETER DEFINITION NAMEINMAAP ANALYSIS Power level, MWth QCRO 1930 MWth Initial CST water volume, qal VCSTO (fn 499,963 qal Initial CST water temperature, F HCST (enthalpy) 90°F Initial suppression pool water Calculated from input 5,435,000 mass, lbm Initial suppression pool water XWRBO(i), where i is node 12.4 ft level, ft number for wetwell Initial suppression pool water TWRBO(i), where i is node 90.2°F temperature, F number for wetwell Drywell free volume, ft" VOLRB(i), where I is node 146,844 ft" number for drywell Wetwell free volume, ft" VOLRB(i) - volume of 209,975 ft" suppression pool water from initial pool mass Containment vent pressure, Refer to MAAP analysis 49.7 psia psi a document RCIC max flow rate, gpm WVRCIC N/A (no RCIC system at OC)

Max FLEX pump flow rate, gpm Refer to MAAP analysis 600 gpm document (@160 psiq in RPV)

Lowest set SRV flow rate, lb/hr Derived from SRV area, ASRV 602,900 lb/hr Lowest set SRV pressure, psia PS ET RV 1079.7 psi Recirc pump seal leakage, gpm Value that was used to define 35 gpm LOCA area, ALOCA Total leakage used in the Value that was used to define 35 gpm transient, aom LOCA area, ALOCA

c. Choked flow from the SRV and the recirculation pump seal leakage is discussed in the EPRI Technical Report 30020027 49. The parameters identified that impact the flow calculation are listed below with input values identified.

VALUE USED IN THE OYSTER CREEK MAAP EPRI PARAMETER NAME ANALYSIS RECOMMENDED VALUE ASRV effective flow area for 0.0733 n- (based on rated flow at Plant-specific value relief valve pressure)

ALOCA seal leakage area 5.75E-4 ft" (35 gpm at normal Plant-specific value conditions)

FCDBRK - discharge coefficient 0.75 0.75 for seal leakage OC-MISC-012-RO.doc I1

Use of MAAP in to EA-12-049

d. Vent line pressure loss can be represented in two ways. The actual piping flow area can be input along with a discharge coefficient (FCDJ). An alternative method would be to calculate the effective flow are given the estimated piping losses, and input a loss coefficient of 1.0. For the Oyster Creek analysis, the vent area is input based on a 8" diameter pipe and a discharge coefficient of 0.75 was selected.
e. The decay heat calculation in MAAP is discussed in the EPRI Technical Report 30020027 49. Input parameters used to compute the decay heat are identified in the EPRI report and are listed in the following table along with their values used in the Oyster Creek analysis.

VALUE USED IN THE OYSTER CREEK MAAP EPRI PARAMETER NAME ANALYSIS RECOMMENDED VALUE FENRCH normal fuel 0.0338 Plant-specific value enrichment EXPO - averaqe exposure 25,453 MW-dav/ton Plant-specific value FCR - total capture rate of U- 0.324 Plant-specific value 238 I total absorption rate FFAF - total absorption rate I 2.37 Plant-specific value total fission rate FQFR1 - traction of fission 0.476 Plant-specific value power due to U-235 and PU-241 FQFR2 - traction of fission 0.437 Plant-specific value power due to PU-239 FQFR3 - fraction of fission 0.087 Plant-specific value power due to U-238 TIRRAD - average effective 8,333.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Plant-specific value irradiation time for entire core (5) The specific MAAP4 analysis case that was used to validate the timing of mitigating strategies in the integrated plan must be identified and should be available on the ePortal for NRC staff to view. Alternately, a comparable level of information may be included in the supplemental response. In either case, the analysis should include a plot of the collapsed vessel level to confirm that TAF is not reached (the elevation of the TAF should be provided) and a plot of the temperature cool down to confirm that the cool down is within tech spec limits.

Response to item 5:

The MAAP analysis performed in suppo1t of the Oyster Creek Integrated Plan is documented in calculation OC-MISC-010 Rev. 1 and is available on the ePortal. Case 6 was the specific MAAP run selected to represent the scenario as described in Attachment lA of the integrated plan.

OC-MISC-Ol 2-RO.doc 12

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