PLA-5865, Revision to Proposed Amendment No. 233 to Unit 2 License NPF-22: MCPR Safety Limits and Reference Changes

Revision to Proposed Amendment No. 233 to Unit 2 License NPF-22: MCPR Safety Limits and Reference Changes

ML050540551
Person / Time
Site:	Susquehanna
Issue date:	02/14/2005
From:	Mckinney B Susquehanna
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-5865, TAC MC4431
Download: ML050540551 (15)

Text

Britt T. McKlnney PPL Susquehanna, LLC , I I Vice President-Nuclear Site Operations 769 Salem Boulevard " I j a Berwick, PA 18603 41 Tel. 570.542.3149 Fax 570.542.1504

• 0o btmckinney¢Dpplweb.com FEB 1 4 2005 pl _

U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop OPI-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION REVISION TO PROPOSED AMENDMENT NO. 233 TO UNIT 2 LICENSE NPF-22: MCPR SAFETY LIMITS AND REFERENCE CHANGES PLA-5865 Docket No. 50-388

References:

I) PLA -5793, B. T. McKinney (PPL)to USNRC, "ProposedAmendmnetnt No. 233 to Unit 2 License NPF-22: MCPR Safety Limits and Reference Changes,"

dated September 08, 2004.

2) USNRC to B. L Shriver, "RequestforAdditionalInforimationz (RAI) - Regarding SSES 2 Minimum CriticalPower Ratio Safety Linz its and Reference Chanzges (TA C No. MC4431)," dated January24, 2005.

3) PLA-5860, B. T. McKinney (PPL) to USANRC, "Request for Additional InZfomniatiom Regarding ProposedAmendment No. 233 to Unit 2 License NPF-22: MCPR Safety Limits and Reference Changes," datedFebnary 1, 2005.

The purpose of this letter is to revise the PPL Susquehanna, LLC (PPL) amendment request submitted on September 8, 2004 in Reference 1. This revision to the amendment request is necessary to delete one of the references provided in Technical Specification (TS) Section 5.6.5 "Core Operating Limits Report (COLR)" as delineated in Reference 1.

It was recently identified that one of the references expected to be used to develop the core operating limits when Reference 1 was issued was not used.

The reference not used that is deleted by this revision is:

ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model," Advanced Nuclear Fuels Corporation.

This methodology describes an evaluation model methodology for licensing analyses of postulated LOCA's. The methodology was developed to comply with 10 CFR 50.46 and Appendix K criteria and is used to confirm NRC acceptance criteria are met. This methodology is not used to determine the Minimum Critical Power Ratio (MCPR).

A, 00 1

Document Control Desk PLA-5865 Instead, the following was used to perform the same analyses:

EMF-2361(P)(A), "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP.

This methodology is listed in the proposed TS Section 5.6.5b as delineated in Reference 1.

The "No Significant Hazards Consideration" provided in Reference 1 has been reviewed.

This revision does not impact the evaluation contained therein. to this letter contains markups of the pages provided in reference 1 reflecting the deletion of the ANF-91-048 (P)(A) methodology. provides updated camera ready pages reflecting the elimination of the ANF-91-048 (P)(A) document. Note that the camera ready pages provided includes the reference added by Amendment 192, which was approved subsequent to issuance of Reference 1.

Any questions regarding this request should be directed to Mr. Duane Filchner at (610) 774-7819.

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

Executed on: __ /__

B. T. McKinney Attachments: - Revised Reference 1 Pages - Revised Camera Ready Pages cc: NRC Region I Mr. A. J. Blamey, NRC Sr. Resident Inspector Mr. R. V. Guzman, NRC Project Manager Mr. R. Janati, DEP/BRP

Attachment l to PLA-5865 PLA-5793 Markup Reflecting Revision

PLA-5793 Markup of Insert Pages Reflecting the Revision

INSERT 1:

1. XN-NF-81-58(P)(A), 'RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company.

2. XN-NF-85-67(P)(A), 'Generic Mechanical Design for Exxon Nuclear Jet pump BWR Reload Fuel," Exxon Nuclear Company.

3. EMF-85-74(P)(A), URODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation.

4. ANF-89-98(P)(A), 'Generic Mechanical Design Criteria for BWR Fuel Designs,"

Advanced Nuclear Fuels Corporation.

5. XN-NF-80-1 9(P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors,"

Exxon Nuclear Company.

6. EMF-2158(P)(A), 'Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation.

\7 048(P)( ), "dvancedguclear Fuels Corpdration Meihodg y for B ng

\ ktrReactossaE EM BWRvaluation Mode!>Advane N u c~a6gu eI t Corporation{-/-

7 ,8 EMF-2361 (P)(A), 'EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP.

f A. EMF-2292(P)(A), -ATRIUMm-10: Appendix K Spray Heat Transfer Coefficients,"

Siemens Power Corporation.

pi6. XN-NF-84-1 05(P)(A), t XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," Exxon Nuclear Company.

,q ,4. ANF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors,'

Advanced Nuclear Fuels Corporation.

A y/. ANF-913(P)(A), 'COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation.

IZ . ANF-1358(P)(A), 'The Loss of Feedwater Heating Transient in Boiling Water Reactors," Advanced Nuclear Fuels Corporation.

,3 )4. EMF-2209(P)(A), USPCB Critical Power Correlation," Siemens Power Corporation.

(y . EMF-1 997(P)(A), UANFB-1 0 Critical Power Correlation", Siemens Power Corporation.

' /. EMF-CC-074(P)(A), "BWR Stability Analysis - Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation.

g> NE-092-001 A, "Licensing Topical Report for Power Uprate With Increased Core Flow," Pennsylvania Power & Light Company.

/- 1,E. Caldon, Inc., rTOPICAL REPORT: Improving Thermal Power Accuracy and Plant Safety while Increasing Operating Power Level Using the LEFTM4ITM System,"

Engineering Report - 80P.

/ ,. Caldon, Inc., 'Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFMNfrm or LEFM CheckPlus"' System," Engineering Report ER-1 60P.

PLA-5793 Camera Ready Pages Reflecting the Revision

PPL Rev.

Reporting Requirements 5.6 5.6 Reporting Requirements core thermal power level may not exceed the originally approved RTP of 3441 MWt, but the value of 3510 MWt (102% of 3441 MWt) remains the initial power level for the bounding licensing analysis.

Future revisions of approved analytical methods listed in this Technical Specification that are currently referenced to 102% of rated thermal power (3510 MWt) shall include reference that the licensed RTP is actually 3489 MWt.

The revisions shall document that the licensing analysis performed at 3510 MWt bounds operation at the RTP of 3489 MWt so long as the LEFM/ system is used as the feedwater flow measurement input into the core thermal power calculation.

The approved analytical methods are described in the following documents, the approved version(s) of which are specified in the COLR.

1. XN-NF-81-58(P)(A),'"RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company.

2. XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet pump BWR Reload Fuel," Exxon Nuclear Company.

3. EMF-85-74(P)(A), URODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation.

4. ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation.

5. XN-NF-80-1 9(P)(A), 'Exxon Nuclear Methodology for Boiling Water Reactors," Exxon Nuclear Company.

6. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation.

7:f2F51048 3(A), gAdva c~d Nuclearoels Corpo ilon Metodol y

\ frBoi in- ter Rea _5fsEXEM B R Evalua/tiModel,'A v

\g uclear Fels Corpefation. ,

j$. EMF-2361 (P)(A), -EXEM BWR-2000 ECCS Evaluation Model,"

Framatome ANP.

g ,. EMF-2292(P)(A), -ATRIUMW-10: Appendix K Spray Heat Transfer Coefficients," Siemens Power Corporation.

(continued)

SUSQUEHANNA - UNIT 2 TS / 5.0-22 Arpendment iV 1X9, 16, 1P

PPL Rev.

Reporting Requirements 5.6 V 5.6 Reporting Requirements q 1 XN-NF-84-1 05(P)(A), 'XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," Exxon Nuclear Company.

My. ANF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors," Advanced Nuclear Fuels Corporation.

/l <i. ANF-913(P)(A), LCOTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation.

Z IS. ANF-1358(P)(A), "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Advanced Nuclear Fuels Corporation.

I 14. EMF-2209(P)(A), -SPCB Critical Power Correlation," Siemens Power Corporation.

Iq 1,6. EMF-1 997(P)(A), uANFB-1 0 Critical Power Correlation", Siemens Power Corporation.

10 EMF-CC-074(P)(A), UBWR Stability Analysis - Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation.

Gt. NE-092-OO1A, "Licensing Topical Report for Power Uprate With Increased Core Flow," Pennsylvania Power & Light Company.

/ 34. Caldon, Inc., 'TOPICAL REPORT: Improving Thermal Power Accuracy and Plant Safety while Increasing Operating Power Level Using the LEFTM#M System," Engineering Report - 80P.

/ ,6A. Caldon, Inc., "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFMvm or LEFM CheckPlusm System," Engineering Report ER-160P.

(continued)

SUSQUEHANNA - UNIT 2 TS I 5.0-23 Agpend me nt 196 1,3, 1i,. X6

PLA-5793 List of References Reflecting the Revision

Attacfment '

BWR Approved Topical Reports for Susquehanna Page A-2 "J-IA. L_;TP"e C-A;;O;nE tl 0 0-f-r ----

Nuclear Plant Isuchnical ;Spedwcalsulon- GULUN autl>

Report Applicable LCO Methodology / Justification XN-NF-80-19(P)(A) Volumes 2, 2A, 2B and 2C, Exxon Nuclear 3.2.1 Describes an evaluation model methodology for licensing analyses of Methodology for Boiling Water Reactors: EXEM SWR ECCS postulated LOCAs in jet pump BWRs. The methodology was Evaluation Model, Exxon Nuclear Company, September 1982. developed to comply with 10 CFR 50.46 and Appendix K criteria to 10 CFR 50.

ANF-9 048(P)(A), Acvanced NucleaFuels Corporation 3.2.1 escribes an upgaded evaluation odel met pdology for icensi Met dology for Bgilfng Water Reaptors EXEM BW // analyses of pVsfulated LOCAs iet pump WRs. The pfethod ogy Edluation ModrAdvanced Nulear Fuels Corpo abion, was develo 6d to comply wI10 CFR .46 and A ;>endix criteria to anuary 1993' _ 10 CFR EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS 3.2.1 Describes an upgraded evaluation model methodology for licensing Evaluation Model, Framatome ANP, May 2001. analyses of postulated LOCAs in jet pump BWRs. The methodology was developed to comply with 10 CFR 50.46 and Appendix K criteria to 10 CFR 50.

EMF-2292(P)(A) Revision 0, ATRIUMTm-10: Appendix K Spray 3.2.1 Provides measured cladding temperatures from spray heat transfer Heat Transfer Coefficients, Siemens Power Corporation, tests to justify the use of Appendix K coefficients for ATRIUM-1 0 fuel September 2000. LOCA analyses.

XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear 3.2.2 Provides overall methodology for determining a MCPR operating limit.

Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.

XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 3.2.2 Provides acapability to perform analyses of transient heat transfer and 2, XCOBRA-T: A Computer Code for BWR Transient behavior in BWR assemblies.

Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.

ANF-524(P)(A) Revision 2 and Supplements 1and 2,ANF 3.2.2 Provides a methodology for the determination of thermal margins, Critical Power Methodology for Boiling Water Reactors, specifically the MCPR safety limit.

Advanced Nuclear Fuels Corporation, November 1990.

ANF-913(P)(A) Volume 1 Revision I and Volume 1 3.2.2 Provides a computer program for analyzing BWR system transients.

Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.

1_

I.

.&4M

Attachment 2 to PLA-5865 Revised Unit 2 Technical Specification Changes (Camera Ready)

PPL Rev. 2 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued) core thermal power level may not exceed the originally approved RTP of 3441 MWt, but the value of 3510 MWt (102% of 3441 MWt) remains the initial power level for the bounding licensing analysis.

Future revisions of approved analytical methods listed in this Technical Specification that are currently referenced to 102% of rated thermal power (3510 MWt) shall include reference that the licensed RTP is actually 3489 MWt.

The revisions shall document that the licensing analysis performed at 3510 MWt bounds operation at the RTP of 3489 MWt so long as the LEFMvT system is used as the feedwater flow measurement input into the core thermal power calculation.

The approved analytical methods are described in the following documents, the approved version(s) of which are specified in the COLR.

1. XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company.

2. XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet pump BWR Reload Fuel,' Exxon Nuclear Company.

3. EMF-85-74(P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation.

4. ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation.

5. XN-NF-80-19(P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors,' Exxon Nuclear Company.

6. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation.

7. EMF-2361 (P)(A), "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP.

8. EMF-2292(P)(A), "ATRIUM T ^'-10: Appendix K Spray Heat Transfer Coefficients," Siemens Power Corporation.

9. XN-NF-84-105(P)(A), "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," Exxon Nuclear Company.

(continued)

SUSQUEHANNA - UNIT 2 TS I 5.0-22 A endment 10 179,1 6,1 84

PPL Rev. 2 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

10. ANF-524(P)(A), TANF Critical Power Methodology for Boiling Water l Reactors," Advanced Nuclear Fuels Corporation.

11. ANF-913(P)(A), "COTRANSA2: A Computer Program for Boiling Water l Reactor Transient Analyses," Advanced Nuclear Fuels Corporation.

12. ANF-1358(P)(A), "The Loss of Feedwater Heating Transient in Boiling l Water Reactors," Advanced Nuclear Fuels Corporation.

13. EMF-2209(P)(A), "SPCB Critical Power Correlation," Siemens Power Corporation.

14. EMF-1997(P)(A), UANFB-10 Critical Power Correlation", Siemens Power l Corporation.

15. EMF-CC-074(P)(A), "BWR Stability Analysis - Assessment of STAIF with l Input from MICROBURN-B2," Siemens Power Corporation.

16. NE-092-001 A, "Licensing Topical Report for Power Uprate With Increased l Core Flow," Pennsylvania Power & Light Company.

17. Caldon, Inc., "TOPICAL REPORT: Improving Thermal Power Accuracy l and Plant Safety while Increasing Operating Power Level Using the LEFTMVT^' System," Engineering Report - 80P.

18. Caldon, Inc., "Supplement to Topical Report ER-80P: Basis for a Power l Uprate with the LEFM4T^' or LEFM CheckPlust '1 System," Engineering Report ER-1 60P.

19. NEDO-32465-A, "BWROG Reactor Core Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications."

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midoycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

SUSQUEHANNA - UNIT 2 TS / 5.0-23 Amendme t 11 10,4,192 192

PPL Rev. 2 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 EDG Failures Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days.

Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4.

5.6.7 PAM Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation,' a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

SUSQUEHANNA - UNIT 2 TS / 5.0-23a Amendmet 1 10,10, 19,2 192