ML120030321

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12/7/2011 - Summary of Public Meeting with Exelon Regarding the Proposed Extended Power Uprate License Amendment Request for LaSalle, Units 1 and 2
ML120030321
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 01/19/2012
From: DiFrancesco N J
Plant Licensing Branch III
To:
Exelon Generation Co
DiFrancesco N, NRR/DORL/LPL3-2, 415-1115
Shared Package
ml120030332 List:
References
TAC ME7496, TAC ME7495
Download: ML120030321 (25)
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Text

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 January 19, 2012 LICENSEE: Exelon Generation Company, LLC FACILITY: LaSalle County Station, Units 1 and 2 SUBJECT: SUMMARY OF DECEMBER 7, 2011, PUBLIC MEETING WITH EXELON GENERATION COMPANY, LLC REGARDING THE PROPOSED EXTENDED POWER UPRA TE LICENSE AMENDMENT REQUEST FOR LASALLE COUNTY STATION, UNITS 1 AND 2 (TAC NOS. ME7495 AND ME7496) On December 7, 2011, a Category 1 public meeting was held between the U.S. Nuclear Regulatory Commission (NRC) and representatives of Exelon Generation Company, LLC (EGC, the licensee) at the NRC Headquarters, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland. The purpose of the meeting was to discuss a planned Extended Power Uprate (EPU) licensee amendment request (LAR) for LaSalle County Station, Units 1 and 2 (LaSalle). A list of attendees is provided as Enclosure 1. During the meeting, EGC informed the NRR staff of their plans to submit a LAR for a 12.5 percent increase in licensed thermal power for LaSalle in the 3rd Quarter of 2012. The meeting discussion focused on review standards for annulus pressurization loads and emergency core cooling system (ECCS) net positive suction head (NPSH) analysis. The licensee presented slides contained in Enclosure 2 and discussed a pre-application checklist enclosed in the public meeting notice (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 113200075). The meeting piloted the use of the Nuclear Energy Institute (NEI) developed pre-application checklist. The purpose of the pre-application meeting using the NEI checklist is for NRC staff and the licensees to reach common understanding of the regulatory criteria and standards to be applied in the review of significant licensing actions with a goal of enhancing the effectiveness and efficiency of the review process. A series of LaSalle EPU pre-application meetings are being planned for additional technical area topics. During each subsequent pre-application meeting, the NEI checklist will be used to support discussions. Additional information about the NEI checklist pilot process may be found in public meeting summary from November 2, 2011 (ADAMS Accession No. ML113210594). The following actions were agreed to between the NRC and the EGC prior to submission of the LaSalle EPU LAR. Summary of Follow-up Actions EGC and the NRC achieved an understanding of the methodology being proposed for use for determination of break Mass and Energy release and annulus pressurization calculations. The NRC acknowledged that the use of the Transient Reactor Analysis Code -G (TRACG) methodology was an acceptable approach. The TRACG evaluation for use in break Mass and Energy release and annulus pressurization calculations should also have a description identifying the differences in application of TRACG at LaSalle compared to its application in the economic simplified boiling-water reactor safety analysis and Grand Gulf's EPU application. EGC agreed to

-provide a summary of the differences along with a justification for use of the methodology in the EPU application. Based on recent Advisory Committee on Reactor Safeguards comments associated with the use of American National Standards Institute and American Nuclear Society (ANSIIANS) 58.2-1988 for evaluating jet impingement loads, the NRC staff noted that the use of this standard as it applies to the proposed EPU may be inappropriate. The NRC staff also noted that ANSIIANS 58.2-1988 is a withdrawn standard. EGC and the NRC achieved an understanding regarding the proposed methodology used to determine the ECCS NPSH. EGC's approach and use of current guidance was considered to be acceptable. Agreement was achieved regarding the use of 21 percent uncertainty when performing EPU ECCS NPSH calculations. It was noted that the uncertainty value may be impacted by Boiling-Water Reactor Owner's Group (BWROG) evaluations currently underway for non-containment accident pressure plants. Also, there may be an elimination of the "Maximum Erosion Zone" requirement. EGC agreed to address any final results of the BWROG evaluation prior to submittal in the application. EGC and the NRC achieved an understanding that the EPU application would address mixed core analysis or provide justification for excluding. The meeting notice and agenda are available under ADAMS Accession No. ML 113200075. The public was invited to observe the meeting. No members of the public were in attendance. Public Meeting Feedback forms were not received. Please direct any inquiries to me at 301-415-1115, or Nicholas.DiFrancesco@nrc.gov. Sincerely, Nicholas DiFrancesco, Project Manager Plant licenSing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-373 and 50-374 Enclosures: List of Attendees Licensee Handout cc w/encl: Distribution via ListServ LIST OF DECEMBER 7,2011, PUBLIC MEETING WITH EXELON GENERATION COMPANY, REGARDING THE PROPOSED EXTENDED POWER UPRA TE LICENSE REQUEST FOR LASALLE COUNTY STATION, UNITS 1 AND NRC Jake Zimmerman Nicholas DiFrancesco Araceli T. Billoch Colon William Jessup John Huang Ogbonna Hopkins John Tsao Tom Alexion Sheldon Stuchell Greg Casto Rick Plasse Rick Stattel David Rahn TaiHuang Ahsan Sallman Tony Ulses Garry Armstrong Muhammad Razzaque Nageswara Karipineni Exelon Kenneth Ainger Kevin Borton John Rommel Harold Vinyard Tim Byam Vikram Shah Terry Simpkin Faramarz Pournia Stevie Du Pont GE-Hitachi Curt Robert Bruce Hagemeier Sara Ruddy Enclosure 1 Exelon Nuclear LaSalle County Station Pre-Application Meeting Extended Power Uprate December 7, 2011 F,nclosure 2 Exelon Kenneth Ainger -Project Management Director, EPU Kevin Borton -Power Uprate Licensing Manager John Rommel -Power Uprate Engineering Director Harold Vinyard -LaSalle Engineering Director Tim Byam -Power Uprate Lead Licensing Engineer Vikram Shah -Power Uprate Senior Engineering .Manager Terry Simpkin -LaSalle Regulatory Assurance Manager Faramarz Pournia -Power Uprate Project Manager Stevie Du Pont -Power Uprate Licensing Engineer Exelon0 Nuclear Agenda and Meeting Introduce LaSalle EPU Staff Describe NEI Pre-Submittal Meeting Pilot Present LaSalle Extended Power Uprate Schedule and Approach Describe Key Aspects of Technical Evaluations and Obtain Feedback Annulus Pressurization Loads ECCS NPSH Analysis Discuss Potential Topics for Future Meetings Nuclear NEI Pilot -Pre-submittal Purpose is to enhance License Amendment Request pre-submittal meetings Reach a common understanding on the regulatory criteria and standards to be applied during the NRC review of the proposed changes Identify potential application issues that can be addressed during the application conceptual phase that will reduce acceptance review time, requests for additional information, and application review time

  • Process Pilot Checklist is used to focus on applicable review criteria, codes, standards, justification required for use of a new analytical method, applicability of a precedent, or feasibility of a desired schedule in order to reach alignment with the NRC NRC meeting notice and meeting summary will docket the expectations and outcomes of the alignment in order to greatly reduce the risk and uncertainty associated with future application acceptance and NRC review
  • LaSalle specific checklist focus
  • Verify methodology and approach used for analyses of Annulus Pressurization Load
  • Verify current NRC expectations and approach regarding ECCS NPSH calculations Nuclear LaSalle Original License Unit 1 licensed 19821 Unit 2 licensed 1983 Original Licensed Thermal Power (OL TP) of 3323 MWt per unit LaSalle Previous Uprates Stretch Power Uprate of 50/0 in 2000 to 3489 MWt MUR Uprate of 1.6% in 2010 to 3546 MWt Current Licensed Thermal Power (CL TP) of 3546 MWt EPU Projected Power Uprate level of 3988 MWt (increase -12.5% of current licensed power or 120% of original licensed power) Nuclear Schedule NRC Communication Schedule
  • Submit LAR:
  • Unit 2 Implementation:
  • Unit 1 Implementation: Target February Target April Target June Target 3rd QTR Target 1st QTR 2014 (20 2nd QTR 2015 (Outage 1st QTR 2016 (Outage L 1 Steam Dryer Evaluation could impact above Nuclear EPU LAR LAR will meet criteria in NRC RS-001, "Review Standard for Extended Power Uprates" Evaluations supporting the LAR were performed using Constant Pressure Power Uprate (CPPU) Licensing Topical Report (NEDC 33004P-A) (commonly called CLTR) Fuel related evaluations were performed to the guidance in NRC-approved NEDC-32424P-A (commonly called EL TR1) Safety issues identified in EL TR1 that should be addressed in a plant-specific EPU license amendment request are addressed in the LaSalle Specific Power Uprate Safety Analysis Report (PUSAR) (NEDC-33603P) For generically evaluated issues -the PUSAR references the NRC-approved generic evaluations in either ELTR1 or EL TR2 (NEDC-32523P-A) No Submittals Linked to Proposed EPU Submittal Incorporated Past RAls Nuclear EPU LAR Exelon's submittal will include
  • Cover Letter and Amendment Request Attachments Description/Evaluation of Proposed Changes including No Significant Hazards Consideration Markup of Operating License and Technical Specifications Markup of Technical Specifications Bases and Technical Requirements Manual (Information Only) Power Uprate Safety Analysis Report (PUSAR) (non-proprietary, proprietary, and affidavit) Regulatory Commitments Supplemental Environmental Report List of Modifications EPU Startup Test Plan Grid Stability Study PRA Report Flow Induced Vibration (FIV) Piping and Component Evaluation Instrument Setpoint Calculations Affected by EPU Steam Dryer Evaluation (High Cycle Fatigue Report) (non-proprietary, proprietary, and affidavit) Nuclear

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Annulus Pressurization (AP) Loads Purpose Verify methodology and approach used for analyses of Pressurization Perform the EPU break mass and energy release (M&E) and pressurization calculations for the annulus pressurization using the GEH TRACG computer code TRACG based M&E release methodology for the AP loads will address GEH corrective action of the Safety Communication SC09-01 Methodology Current CLTP Methodology NEDO-24548, Annulus Pressurization Load Adequacy Evaluation Computer Code: RELAP41 MOD 3 EPU Methodology NEDE-32176P, Revision 4, TRACG Model Description NEDE-32177P, Revision 3, TRACG Qualification NEDE-33083P-A, TRACG Application for ESBWR, October 2005 NEDE-33440P, Revision 2, TRACG ESBWR Safety Analysis -Additional Information, March 2010 Computer Code: TRACG V.04 Same computer code was used in the Grand Gulf EPU Submittal to address AP loads Annulus Pressurization (AP) Analysis Overview High line com "",.,','.,.. ".-,--.----..-..-.. ".... ,..,'.. 15 10t----/ u: i '" 10E.()3 1.0E.Q2 1.OE-01 tOE+OO 1.0E+01 'i r;' '\ ;; ,,:,.. Tlmo (Me; e..., .,,""" ,**Mass I energy Annulus pressure release vs. time vs. 'on :: n "I.--..;, *X I": :: I ..'. :! .. " " Displacements, accelerations, forces, stresses, moments and response spectra Piping and component loads, stress, fatigue, and accelerations Annulus Pressurization Loads Analysis Methods 'cltP***Method Mass and Energy Generic NEDO-TRACG 04 Note 1 24648 Annulus Pressurization RELAP4IMOD 3 TRACG 04 Note 2 Jet Loads ANSI 176 ANSI/ANS 68.2-Same method, Old 1988 standard superseded. Pipe Whip PDA PDA Same Dynamic Structural SAP4G07 SAP4G07 Same Note 1 -The TRACG 04 allows the calculation of mass and energy (M&E) release rates to include the physical attributes of the piping system for both rated and off-rated conditions. TRACG eliminates unphysical and artificially imposed jumps in mass and energy. Providing estimates of M&E at off-rated and rated conditions addresses the concerns identified in GEH Safety Communication SC 09-01, Annulus Pressurization Loads Evaluation, dated June 8, 2009 Note 2 -The use of the TRACG 04 vessel component together with a fine mesh model (336 nodes) of LaSalle annulus provides a more detailed annulus pressurization response than the analysis of which uses a coarse node (35 node) RELAP 4/MOD 3 Nuclear Annulus Pressurization (AP) Loads Summary

  • The application of TRACG for both the mass and energy release analysis and the annulus pressurization analysis is appropriate to provide a response frequency used in all downstream load analyses
  • SC 09-01 will be addressed by analyzing the pipe breaks considered in the LaSalle design and licensing basis at various rated and off-rated operating conditions with bounding conditions being used in the downstream analysis Confirm NRC agreement that the above approach to calculation of AP Loads is acceptable Exelon Nuclear

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ECCS NPSH Purpose: Verify methodology and Exelon's approach to perform NPSH analysis for a non-CAP credit plant is in accordance with NRC's expectations, and draft regulatory guidance Methodology SECY 11-0014, Use of Containment Accident Pressure in Analyzing Emergency Core Cooling System and Containment Heat Removal System Pump Performance in Postulated Accidents Event-specific analyses determine containment -Suppression Pool Event-specific NPSHa determined for each applicable event/pump NPSHa compared to NPSHreff for each applicable pump Time in Maximum Erosion Zone Nuclear ECCS NPSH Key Assumptions No Containment Accident Pressure Deterministic analysis with conservative inputs for DBA-LOCA, ASDC and Small Break LOCA. Nominal inputs for non-design basis events (SBO, Appendix R, ATWS) analyses Vendor supplied NPSHr 3% curves 21 % uncertainty for NPSHreff for DBA-LOCA, ASDC and Small Break LOCA events 0% uncertainty for non-design basis events (ATWS, SBO, Appendix R) Assumes minimum Suppression Pool Inventory (level) for all events All events are evaluated at 102% of EPU power Nuclear ECCS NPSH Preliminary Event Pump NPSHa (Feet) NPSHreff (Feet) Margin (Feet) OSA-LOCA RHR 18.8 16.9 1.9 OSA-LOCA HPCS 19.7 6.1 13.6 OSA-LOCA LPCS 19.0 2.4 16.6 ASOC RHR 17.8 16.9 0.9 ASOC HPCS 17.5 6.1 11.4 ASOC LPCS 18.0 2.4 15.6 SSO RHR 19.2 14.0 5.2 SSO RCIC Analysis in Progress SBO HPCS 18.9 5.0 13.9 ATWS RHR 32.4 14.0 18.4 App R RHR 25.9 14.0 11.9 AppR HPCS 25.7 5.0 20.7 AppR LPCS 26.1 2.0 24.1 NPSHreft values consistent with draft guidance (NPSHreff =NPSHr3% + uncertainties) including 21 % uncertainty for Design Basis Events and 0% uncertainty for non-Design Basis Events No modifications are required to achieve the above results Exelon Nuclear ECCS NPSH Summary Exelon's approach to perform NPSH analysis is in accordance with NRC expectations and draft guidance Demonstrates that adequate positive margin exist for ECCS/RCIC pumps Demonstrates that ECCS/RCIC pumps will perform their safety functions Confirm NRC agreement that the above approach to determine ECCS NPSH is acceptable Nuclear Future EPU Meeting Topics and

  • Follow-up EPU Meetings
  • Proposed Topics -Steam Dryer Strategy and FIV Analysis -Impact and Changes to Human Factors -Impact on Primary Containment internal pressure (P a) -Ultimate Heat Sink Analysis -Setpoint Calculations -Alternate Source Term Analysis
  • Next meeting target February 2012 Meeting
  • Pilot Alignment and Outcome
  • Discussion
  • Checklist Mark-up
  • ASDC -Alternate Shut Down Cooling
  • CAP -Containment Accident Pressure
  • CLTP -Current Licensed Thermal Power
  • DBA -Design Basis Accident
  • ECCS -Emergency Core Cooling System
  • ESBWR -Economic Simplified Boiling Water Reactor
  • FIV -Flow Induced Vibration
  • GEH -General Electric -Hitachi
  • LAR -License Amendment Request
  • LOCA -Loss of Coolant Accident
  • L TR -Licensing Topical Report
  • MWt -Mega Watts thermal
  • NEI-Nuclear Energy Institute
  • NPSH -Net Positive Suction Head
  • NPSHa -Net Positive Suction Head available
  • NPSHr -Net Positive Suction Head required
  • NPSHreff -Effective Net Positive Suction Head required
  • PRA -Probabilistic Risk Assessment
  • PUR -Power Uprate
  • RAI -Request for Additional Information
  • RHR -Residual Heat Removal
  • SBO -Station Black Out
  • SC -GEH Safety Communication
  • SECY -Commission Papers (Written issues papers the NRC staff submits to the Commission to inform them about policy, rulemaking, and adjudicatory matters) Nuclear provide a summary of the differences along with a justification for use of the methodology in the EPU application. Based on recent Advisory Committee on Reactor Safeguards comments associated with the use of American National Standards Institute and American Nuclear Society (ANSI/ANS) 58.2-1988 for evaluating jet impingement loads, the NRC staff noted that the use of this standard as it applies to the proposed EPU may be inappropriate. The NRC staff also noted that ANSI/ANS 58.2-1988 is a withdrawn standard. EGC and the NRC achieved an understanding regarding the proposed methodology used to determine the ECCS NPSH. EGC's approach and use of current guidance was considered to be acceptable. Agreement was achieved regarding the use of 21 percent uncertainty when performing EPU ECCS NPSH calculations. It was noted that the uncertainty value may be impacted by Boiling-Water Reactor Owner's Group (BWROG) evaluations currently underway for non-containment accident pressure plants. Also, there may be an elimination of the "Maximum Erosion Zone" requirement. EGC agreed to address any final results of the BWROG evaluation prior to submittal in the application. EGC and the NRC achieved an understanding that the EPU application would address mixed core analysis or provide justification for excluding. The meeting notice and agenda are available under ADAMS Accession No. ML 113200075. The public was invited to observe the meeting. No members of the public were in.attendance. Public Meeting Feedback forms were not received. Please direct any inquiries to me at 301-415-1115, or Nicholas.DiFrancesco@nrc.gov. Sincerely, IRA! Nicholas DiFrancesco, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-373 and 50-374 Enclosures: List of Attendees Licensee Handout cc w/encl: Distribution via ListServ DISTRIBUTION: PUBLIC RldsNrrDorlLpl3-2 Resource RidsRgn3MailCenter Resource LPL3-2 R/F RidsNrrPMLasalle Resource ASnyder. EDO Region III RidsNrrAdro Resource RidsOgcMailCenter Resource RidsNrrDori Resource RidsOpaMaii Resource 000 M . SPackage Accession No. Ml120030332 Meeting Notice ML 1132 75 eetlng ummary ML 120030321 NRC-001 OFFICE DORULPL3-2 PM DORULPL3-2/LA DORULPL3-2/BC DORULPL3-2/PM NAME NDiFrancesco SRohrer(BTully for) JZimmerman NDiFrancesco DATE 1/17112 1117112 1/18/12 1/19/12 OFFICIAL RECORD COPY
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