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| {{#Wiki_filter:.I Exeke n.m Exelon Generation Company, LLC LaSalle County Station 2601 North 21" Road Marseilles, IL 61341-9757 www.exeloncorp.corn Nuclear RA06-013 March 9, 2006 10 CFR 50.46 United States Nuclear Regulatory Commission Attention: | | {{#Wiki_filter:.I Exeke n.m Exelon Generation Company, LLC www.exeloncorp.corn LaSalle County Station Nuclear 2601 North 21" Road Marseilles, IL 61341-9757 RA06-013 March 9, 2006 10 CFR 50.46 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-1 1 and NPF-18 NRC Docket Nos. 50-373 and 50-374 |
| Document Control Desk Washington, D.C. 20555 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-1 1 and NPF-18 NRC Docket Nos. 50-373 and 50-374 | |
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| ==Subject:== | | ==Subject:== |
| Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report | | Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report |
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| ==Reference:== | | ==Reference:== |
| | Letter from D. J. Enright (Exelon Generation Company, LLC) to U. S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated March 9, 2005 In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," Exelon Generation Company, LLC, (EGC) submits the enclosed attachments to fulfill the 30-day and annual reporting requirements for LaSalle County Station (LSCS), Units 1 and 2. |
| | In the referenced letter, EGC reported the fuel peak cladding temperatures (PCTs) calculated based on an acceptable model to be 14000 F for General Electric (GE) fuel. There is no change in PCT for the GE Fuel for this reporting period. |
| | The referenced letter also provided the PCT of 1832TF for the Framatome Advanced Nuclear Power (FANP) fuel based on an acceptable model. Since the last evaluation, FANP ATRIUM-10 fuel has been re-introduced into Unit 1 and a new analysis was performed. Based on the new analysis, the PCT for FANP ATRIUM-10 fuel decreased to a value of 17290F. This is a change of over 500F from the last evaluation using a NRC approved acceptable model. For Unit 2, there were no changes for the FANP ATRIUM-9B fuel and the PCT remains at 18320F. |
| | Unit 1 and Unit 2 employ a mixed core design containing co-resident GE and FANP fuel. The Loss of Coolant Accident (LOCA) analyses of record for both GE and FANP fuel are within all of the acceptance criteria set forth in 10 CFR 50.46. |
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| Letter from D. J. Enright (Exelon Generation Company, LLC) to U. S. NRC,"Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated March 9, 2005 In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," Exelon Generation Company, LLC, (EGC) submits the enclosed attachments to fulfill the 30-day and annual reporting requirements for LaSalle County Station (LSCS), Units 1 and 2.In the referenced letter, EGC reported the fuel peak cladding temperatures (PCTs) calculated based on an acceptable model to be 1400 0 F for General Electric (GE) fuel. There is no change in PCT for the GE Fuel for this reporting period.The referenced letter also provided the PCT of 1832TF for the Framatome Advanced Nuclear Power (FANP) fuel based on an acceptable model. Since the last evaluation, FANP ATRIUM-10 fuel has been re-introduced into Unit 1 and a new analysis was performed.
| | U. S. Nuclear Regulatory Commission March 9, 2006 Page 2 Attachments 1, 2, and 3 provide PCT information for the limiting LOCA evaluations for LSCS, Units 1 and 2, including all assessments as of February 1, 2005. The assessment notes are contained in Attachment 4 and provide a detailed description for each change or error reported. |
| Based on the new analysis, the PCT for FANP ATRIUM-10 fuel decreased to a value of 1729 0 F. This is a change of over 50 0 F from the last evaluation using a NRC approved acceptable model. For Unit 2, there were no changes for the FANP ATRIUM-9B fuel and the PCT remains at 1832 0 F.Unit 1 and Unit 2 employ a mixed core design containing co-resident GE and FANP fuel. The Loss of Coolant Accident (LOCA) analyses of record for both GE and FANP fuel are within all of the acceptance criteria set forth in 10 CFR 50.46.
| | Should you have any questions concerning this letter, please contact Mr. Terrence W. Simpkin, Regulatory Assurance Manager, at (815) 415-2800. |
| U. S. Nuclear Regulatory Commission March 9, 2006 Page 2 Attachments 1, 2, and 3 provide PCT information for the limiting LOCA evaluations for LSCS, Units 1 and 2, including all assessments as of February 1, 2005. The assessment notes are contained in Attachment 4 and provide a detailed description for each change or error reported.Should you have any questions concerning this letter, please contact Mr. Terrence W. Simpkin, Regulatory Assurance Manager, at (815) 415-2800.Respectfully, I E Daniel J. Enright Plant Manager LaSalle County Station Attachments cc: Regional Administrator | | Respectfully, I E Daniel J. Enright Plant Manager LaSalle County Station Attachments cc: Regional Administrator - NRC Region IlIl NRC Senior Resident Inspector - LaSalle County Station |
| -NRC Region IlIl NRC Senior Resident Inspector | |
| -LaSalle County Station Attachment 1 LaSalle Units 1 and 2 10 CFR 50.46 Report (GE Fuel)PLANT NAME: ECCS EVALUATION MODEL: REPORT REVISION DATE: CURRENT OPERATING CYCLES: LaSalle Units I and 2 SAFER/GESTR LOCA February 1, 2006 LIC12* and L2Cl I ANALYSIS OF RECORD Evaluation Model Methodology: | |
| NEDE-23785-1 -PA, Rev. 1, "GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident (Volume III), SAFER/GESTR Application Methodology", October 1984.Calculation: "Project Task Report, Exelon LaSalle Unit I and 2 SAFER/GESTR Loss-of-Coolant Accident Analysis for GE 14 Fuel," GE report number GE-NE-0000-0022-8684-RI, dated December 2004.Fuel: GE14 Limiting Single Failure: Limiting Break Size and Location: Reference PCT: FIPCS Diesel Generator Double Ended Guillotine of Recirculation Pump Suction Piping 1400 0 F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS I OCFR50.46 report dated March 9, 2005 (Note 9) APCT = 0 :F Net PCT -_ 1400 OF B. CURRENT LOCA MODEL ASSESSMENTS None N/A Net PCT 1 1400 OF* Currently Unit I is in refueling.
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| Expected Cycle 12 Startup -March 16, 2006 Attachment 2 LaSalle Unit 1 10 CFR 50.46 Report (FANP Fuel)PLANT NAME: ECCS EVALUATION MODEL: REPORT REVISION DATE: CURRENT OPERATING CYCLES: LaSalle Unit I EXEM BWR-2000 Evaluation Model February 1, 2006 LIC12*ANALYSIS OF RECORD Evaluation Model Methodology:
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| EMF-2361 (P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001.Calculation:
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| EMF-3230 (P) Revision 0, LaSalle Units I and 2 EXEM BWR-2000 LOCA Break Spectrum Analysis for ATRIUM -10 Fuel, November 2005.EMF-3231 (P) Revision 0, LaSalle Units I and 2 EXEM BWR-2000 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM -10 Fuel, November 2005.Fuel: ATRIUM- 10 Limiting Single Failure: Limiting Break Size and Location: Reference PCT: Low-pressure coolant injection Diesel Generator Double Ended Guillotine/0.8 discharge coefficient of Recirculation Pump Suction Piping 1729 OF MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS A FANP LOCA analysis was performed in November 2005 to support the re- APCT =0 0 F introduction of ATRIUM-10 for LIC12. This analysis addresses all errors and issues. In Cycle 12 there will be no ATRIUM-9B fuel in the Unit I Core Net PCT 1729 'F B. CURRENT LOCA MODEL ASSESSMENTS None N/A Net PCT 1729 F* Currently Unit I is in refueling.
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| Expected Cycle 12 startup -March 16, 2006.
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| Attachment 3 LaSalle Unit 2 10 CFR 50.46 Report (FANP Fuel)PLANT NAME: ECCS EVALUATION MODEL: REPORT REVISION DATE: CURRENT OPERATING CYCLE: LaSalle Unit 2 EXEM BWR Evaluation Model February 1, 2006 L2C 11 ANALYSIS OF RECORD Evaluation Model Methodology:
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| Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-9 1-048(P)(A), January 1993.BWR Jet Pump Model Revision for RELAX, ANF-91 -048(P)(A), Supplement I and Supplement 2, Siemens Power Corporation, October 1997.Calculation:
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| : 1. LaSalle LOCA-ECCS Analysis MAPLHGR Limits for ATRIUMŽ-9B Fuel, EMF-2175(P), March 1999.2. LOCA Break Spectrum Analysis for LaSalle Units I and 2, EMF-2174(P), March 1999.3. LaSalle Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTŽ-
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| 10 Fuel, EMF-2641(P), November 2001.4. LaSalle Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUMŽ"'-10 Fuel, EMF-2639(P), November 2001.Fuel: Limiting Fuel Limiting Single Failure: Limiting Break Size and Location: ATRIUMŽ -9B and ATRIUJMT-10 ATRJUMTM-9B HPCS Diesel Generator 1.1 ft 2 Recirculation Pump Discharge Side Line Break Reference PCT: MARGIN ALLOCATION 1807 0 F A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated May 7, 1999 (See Note 2) APCT= 0 'F 10 CFR 50.46 report dated February 9.2000 (See Note 3) APCT =18 'F 10 CFR 50.46 report dated June 12, 2000 (See Note 4) APCT= O 0 F 10 CFR 50.46 report dated June 8,2001 (See Note 5) APCT = O 0 F 10 CFR 50.46 report dated June 8,2002 (See Note 6) APCT = 2 'F 10 CFR 50.46 report dated June 9,2003 (See Note 7) APCT = 5 'F 10 CFR 50.46 report dated March 9,2004 (See Note 1) APCT = 0 'F 10 CFR 50.46 report dated March 9,2005 (See Note 8) APCT= O°F Net PCT 1832 OF CURRENT LOCA MODEL ASSESSMENTS No errors/issues for this reporting period APCT = O°F Net PCT 1832 "F B.
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| Attachment 4 LaSalle Units 1 and 2 10 CFR 50.46 Report (Assessment Notes)1. Prior LOCA model assessment for FANP fuel During the startup of LaSalle Unit I Cycle II several evaluations were performed for FANP LOCA analysis as reported in the Reference.
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| The net results of these evaluations were that there was a zero degree PCT impact. Additionally a problem was also identified by FANP pertaining with the transfer of RELAX coolant temperature data from PREHUXY to HUXY at the time of core spray. FANP determined that the impact of this problem on the limiting break spectrums results was zero degree. This was also reported in the Reference.
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| ==Reference:==
| | Attachment 1 LaSalle Units 1 and 2 10 CFR 50.46 Report (GE Fuel) |
| | PLANT NAME: LaSalle Units I and 2 ECCS EVALUATION MODEL: SAFER/GESTR LOCA REPORT REVISION DATE: February 1,2006 CURRENT OPERATING CYCLES: LIC12* and L2Cl I ANALYSIS OF RECORD Evaluation Model Methodology: NEDE-23785- 1-PA, Rev. 1, "GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident (Volume III), |
| | SAFER/GESTR Application Methodology", October 1984. |
| | Calculation: "Project Task Report, Exelon LaSalle Unit I and 2 SAFER/GESTR Loss-of-Coolant Accident Analysis for GE 14 Fuel," GE report number GE-NE-0000-0022-8684-RI, dated December 2004. |
| | Fuel: GE14 Limiting Single Failure: FIPCS Diesel Generator Limiting Break Size and Double Ended Guillotine of Recirculation Pump Suction Piping Location: |
| | Reference PCT: 14000 F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS IOCFR50.46 report dated March 9, 2005 (Note 9) APCT = 0 :F Net PCT -_ 1400 OF B. CURRENT LOCA MODEL ASSESSMENTS None N/A Net PCT 1 1400 OF |
| | * Currently Unit I is in refueling. Expected Cycle 12 Startup - March 16, 2006 |
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| Letter from Susan R Landahl (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated March 9,2004.]2. Prior LOCA Model Assessment for FANP fuel The May 1999 LOCA model assessment was a new analysis of record for Framatome (Formerly Siemens) due to the introduction of ATRILTM-9B fuel into the Unit 2 Cycle 8 core. Therefore, there is no PCT change. Analysis was performed for a core power of 3722 MWt that bounds the current uprated power of 3489 MWt.[
| | Attachment 2 LaSalle Unit 1 10 CFR 50.46 Report (FANP Fuel) |
| | PLANT NAME: LaSalle Unit I ECCS EVALUATION MODEL: EXEM BWR-2000 Evaluation Model REPORT REVISION DATE: February 1,2006 CURRENT OPERATING CYCLES: LIC12* |
| | ANALYSIS OF RECORD Evaluation Model Methodology: EMF-2361 (P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001. |
| | Calculation: EMF-3230 (P) Revision 0, LaSalle Units I and 2 EXEM BWR-2000 LOCA Break Spectrum Analysis for ATRIUM - 10 Fuel, November 2005. |
| | EMF-3231 (P) Revision 0, LaSalle Units I and 2 EXEM BWR-2000 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM - 10 Fuel, November 2005. |
| | Fuel: ATRIUM- 10 Limiting Single Failure: Low-pressure coolant injection Diesel Generator Limiting Break Size and Double Ended Guillotine/0.8 discharge coefficient of Recirculation Location: Pump Suction Piping Reference PCT: 1729 OF MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS A FANP LOCA analysis was performed in November 2005 to support the re- APCT =0 0F introduction of ATRIUM-10 for LIC12. This analysis addresses all errors and issues. In Cycle 12 there will be no ATRIUM-9B fuel in the Unit I Core Net PCT 1729 'F B. CURRENT LOCA MODEL ASSESSMENTS None N/A Net PCT 1729 F |
| | * Currently Unit I is in refueling. Expected Cycle 12 startup - March 16, 2006. |
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| ==Reference:== | | Attachment 3 LaSalle Unit 2 10 CFR 50.46 Report (FANP Fuel) |
| | PLANT NAME: LaSalle Unit 2 ECCS EVALUATION MODEL: EXEM BWR Evaluation Model REPORT REVISION DATE: February 1,2006 CURRENT OPERATING CYCLE: L2C 11 ANALYSIS OF RECORD Evaluation Model Methodology: Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-9 1-048(P)(A), January 1993. |
| | BWR Jet Pump Model Revision for RELAX, ANF-91 -048(P)(A), |
| | Supplement I and Supplement 2, Siemens Power Corporation, October 1997. |
| | Calculation: 1. LaSalle LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM'-9BFuel, EMF-2175(P), March 1999. |
| | : 2. LOCA Break Spectrum Analysis for LaSalle Units I and 2, EMF-2174(P), March 1999. |
| | : 3. LaSalle Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMT'-10 Fuel, EMF-2641(P), November 2001. |
| | : 4. LaSalle Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM'"'-10Fuel, EMF-2639(P), November 2001. |
| | Fuel: ATRIUM'-9B and ATRIUJMT-10 Limiting Fuel ATRJUMTM-9B Limiting Single Failure: HPCS Diesel Generator Limiting Break Size and Location: 1.1 ft2 Recirculation Pump Discharge Side Line Break Reference PCT: 1807 0F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated May 7, 1999 (See Note 2) APCT= 0 'F 10 CFR 50.46 report dated February 9.2000 (See Note 3) APCT =18 'F 10 CFR 50.46 report dated June 12, 2000 (See Note 4) APCT= O0 F 10 CFR 50.46 report dated June 8,2001 (See Note 5) APCT = O 0 F 10 CFR 50.46 report dated June 8,2002 (See Note 6) APCT = 2 'F 10 CFR 50.46 report dated June 9,2003 (See Note 7) APCT = 5 'F 10 CFR 50.46 report dated March 9,2004 (See Note 1) APCT = 0 'F 10 CFR 50.46 report dated March 9,2005 (See Note 8) APCT= O°F Net PCT 1832 OF B. CURRENT LOCA MODEL ASSESSMENTS No errors/issues for this reporting period APCT = O°F Net PCT 1832 "F |
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| Letter from J. A. Benjamin (ComEd) to U.S. NRC, "Report of Significant Change in Calculated Peak Cladding Temperature (PCT) -10CFR 50.46 Report," dated May 7, 1999.]3. Prior LOCA Model Assessment for FANP fuel The February 2000 50.46 report assessed the impact of errors in the LOCA evaluation model.[
| | Attachment 4 LaSalle Units 1 and 2 10 CFR 50.46 Report (Assessment Notes) |
| | : 1. Prior LOCA model assessment for FANP fuel During the startup of LaSalle Unit I Cycle II several evaluations were performed for FANP LOCA analysis as reported in the Reference. The net results of these evaluations were that there was a zero degree PCT impact. Additionally a problem was also identified by FANP pertaining with the transfer of RELAX coolant temperature data from PREHUXY to HUXY at the time of core spray. FANP determined that the impact of this problem on the limiting break spectrums results was zero degree. This was also reported in the Reference. |
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| ==Reference:== | | ==Reference:== |
| | Letter from Susan R Landahl (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated March 9,2004.] |
| | : 2. Prior LOCA Model Assessment for FANP fuel The May 1999 LOCA model assessment was a new analysis of record for Framatome (Formerly Siemens) due to the introduction of ATRILTM-9B fuel into the Unit 2 Cycle 8 core. Therefore, there is no PCT change. Analysis was performed for a core power of 3722 MWt that bounds the current uprated power of 3489 MWt. |
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| Letter from J; A. Benjamin (ComEd) to U.S. NRC, "Plant Specific ECCS Evaluation Changes -IOCFR 50.46 Report," dated February 9,2000.]4. Prior LOCA Model Assessment for FANP fuel The June 2000 10 CFR 50.46 report does not have any PCT assessment for ATRIUM-9B fuel.[ | | ==Reference:== |
| | Letter from J. A. Benjamin (ComEd) to U.S. NRC, "Report of Significant Change in Calculated Peak Cladding Temperature (PCT) - 10CFR 50.46 Report," dated May 7, 1999.] |
| | : 3. Prior LOCA Model Assessment for FANP fuel The February 2000 50.46 report assessed the impact of errors in the LOCA evaluation model. |
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| ==Reference:== | | ==Reference:== |
| | Letter from J; A. Benjamin (ComEd) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - IOCFR 50.46 Report," dated February 9,2000.] |
| | : 4. Prior LOCA Model Assessment for FANP fuel The June 2000 10 CFR 50.46 report does not have any PCT assessment for ATRIUM-9B fuel. |
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| Letter from C. G. Pardee (ComEd) to U.S. NRC, "Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated June 12, 2000.]5. Prior LOCA model assessment for FANP fuel The reference letter assessed impact of Unit 2 LPCS riser leakage, errors in FANP LOCA analysis model and Unit 2 Cycle 9 reload fuel.[ | | ==Reference:== |
| | Letter from C. G. Pardee (ComEd) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated June 12, 2000.] |
| | : 5. Prior LOCA model assessment for FANP fuel The reference letter assessed impact of Unit 2 LPCS riser leakage, errors in FANP LOCA analysis model and Unit 2 Cycle 9 reload fuel. |
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| ==Reference:== | | ==Reference:== |
| | Letter from M. A Schiavoni (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated June 8, 2001.1 |
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| Letter from M. A Schiavoni (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated June 8, 2001.1 Attachment 4 LaSalle Units 1 and 2 10 CFR 50.46 Report (Assessment Notes)6. Prior LOCA model assessment for FANP fuel The referenced letter assessed impact of errors in FANP LOCA analysis model, Unit 1 Cycle 10 reload fuel and ATRIUM-9B exposure extension.
| | Attachment 4 LaSalle Units 1 and 2 10 CFR 50.46 Report (Assessment Notes) |
| | : 6. Prior LOCA model assessment for FANP fuel The referenced letter assessed impact of errors in FANP LOCA analysis model, Unit 1 Cycle 10 reload fuel and ATRIUM-9B exposure extension. |
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| ==Reference:== | | ==Reference:== |
| | | Letter from M. A Schiavoni (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated June 8, 2002.1 |
| Letter from M. A Schiavoni (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated June 8, 2002.1 7. Prior LOCA model assessment for FANP fuel The June 2003 50.46 report assessed the impact of errors in the LOCA evaluation, Unit 2 jet pump leakage, Unit 2 Cycle 10 reload Fuel and the Unit 1 mid-cycle reload.[ | | : 7. Prior LOCA model assessment for FANP fuel The June 2003 50.46 report assessed the impact of errors in the LOCA evaluation, Unit 2 jet pump leakage, Unit 2 Cycle 10 reload Fuel and the Unit 1 mid-cycle reload. |
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| ==Reference:== | | ==Reference:== |
| | | Letter from Susan R Landahl (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated June 9,2003.] |
| Letter from Susan R Landahl (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated June 9,2003.]8. Prior LOCA model assessment for FANP fuel The March 2005 10 CFR 50.46 report does not have any PCT assessment. | | : 8. Prior LOCA model assessment for FANP fuel The March 2005 10 CFR 50.46 report does not have any PCT assessment. |
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| ==Reference:== | | ==Reference:== |
| | | Letter from Daniel J. Enright (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated March 9, 2005.] |
| Letter from Daniel J. Enright (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated March 9, 2005.]9. Prior LOCA model assessment for GE fuel A GE LOCA analysis was performed in December 2004 to support the introduction of GE 14 for L2C 11. This analysis bounds both LaSalle Units and addressed all errors and issues. This was reported in the Reference. | | : 9. Prior LOCA model assessment for GE fuel A GE LOCA analysis was performed in December 2004 to support the introduction of GE 14 for L2C 11. This analysis bounds both LaSalle Units and addressed all errors and issues. This was reported in the Reference. |
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| ==Reference:== | | ==Reference:== |
| | | Letter from Daniel J. Enright (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated March 9, 2005.]}} |
| Letter from Daniel J. Enright (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated March 9, 2005.]}} | |
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Category:Letter type:RA
MONTHYEARRA24-008, Post-Outage Inservice Inspection (ISI) Summary Report2024-06-20020 June 2024 Post-Outage Inservice Inspection (ISI) Summary Report RA24-023, Response to NRC LaSalle County Station, Units 1 and 2, Integrated Inspection Report 05000373/2024001 and 05000374/20240012024-06-12012 June 2024 Response to NRC LaSalle County Station, Units 1 and 2, Integrated Inspection Report 05000373/2024001 and 05000374/2024001 RA24-011, 2023 Annual Radiological Environmental Operating Report2024-05-0303 May 2024 2023 Annual Radiological Environmental Operating Report RA24-009, 2023 Annual Radioactive Effluent Release Report2024-04-26026 April 2024 2023 Annual Radioactive Effluent Release Report RA24-010, Annual Dose Report for 20232024-04-24024 April 2024 Annual Dose Report for 2023 RA24-006, Cycle 21 Core Operating Limits Reports2024-03-11011 March 2024 Cycle 21 Core Operating Limits Reports RA23-072, Regulatory Commitment Change Summary Report2023-12-15015 December 2023 Regulatory Commitment Change Summary Report RA23-057, Registration of Use of Cask to Store Spent Fuel2023-08-11011 August 2023 Registration of Use of Cask to Store Spent Fuel RA23-055, Registration of Use of Cask to Store Spent Fuel2023-08-11011 August 2023 Registration of Use of Cask to Store Spent Fuel RA23-052, Registration of Use of Cask to Store Spent Fuel2023-08-0404 August 2023 Registration of Use of Cask to Store Spent Fuel RA23-051, Registration of Use of Cask to Store Spent Fuel2023-07-21021 July 2023 Registration of Use of Cask to Store Spent Fuel RA23-048, Registration of Use of Cask to Store Spent Fuel2023-07-14014 July 2023 Registration of Use of Cask to Store Spent Fuel RA23-045, Completion of License Renewal Activities Prior to Entering the Period of Extended Operation2023-06-30030 June 2023 Completion of License Renewal Activities Prior to Entering the Period of Extended Operation RA23-047, Registration of Use of Cask to Store Spent Fuel2023-06-30030 June 2023 Registration of Use of Cask to Store Spent Fuel RA23-018, Post-Outage Inservice Inspection (ISI) Summary Report2023-06-27027 June 2023 Post-Outage Inservice Inspection (ISI) Summary Report RA23-046, Registration of Use of Cask to Store Spent Fuel2023-06-23023 June 2023 Registration of Use of Cask to Store Spent Fuel RA23-042, Registration of Use of Cask to Store Spent Fuel2023-06-16016 June 2023 Registration of Use of Cask to Store Spent Fuel RA23-029, 2022 Annual Radiological Environmental Operating Report2023-05-12012 May 2023 2022 Annual Radiological Environmental Operating Report RA23-028, Submittal of 2022 Annual Radioactive Effluent Release Report2023-04-28028 April 2023 Submittal of 2022 Annual Radioactive Effluent Release Report RA23-027, Annual Dose Report for 20222023-04-21021 April 2023 Annual Dose Report for 2022 RA23-025, Post Accident Monitoring Report2023-04-0707 April 2023 Post Accident Monitoring Report RA23-013, Submittal of Triennial Chlorine Survey Report2023-03-30030 March 2023 Submittal of Triennial Chlorine Survey Report RA23-019, Cycle 20 Core Operating Limits Reports2023-03-21021 March 2023 Cycle 20 Core Operating Limits Reports RA22-066, 150 Day Submittal Cover Letter2022-12-0202 December 2022 150 Day Submittal Cover Letter RA22-060, Regulatory Commitment Change Summary Report2022-11-30030 November 2022 Regulatory Commitment Change Summary Report RA22-051, Submittal of the Snubber Program Plan for the Fourth 10-Year Interval2022-11-16016 November 2022 Submittal of the Snubber Program Plan for the Fourth 10-Year Interval RA22-041, Unit 2 Cycle 19 Core Operating Limits Report Revision2022-10-0505 October 2022 Unit 2 Cycle 19 Core Operating Limits Report Revision RA22-029, Cycle 20 Startup Test Report Summary2022-06-10010 June 2022 Cycle 20 Startup Test Report Summary RA22-021, Post-Outage 90-Day Inservice Inspection (ISI) Summary Report2022-06-0808 June 2022 Post-Outage 90-Day Inservice Inspection (ISI) Summary Report RA22-020, Annual Radiological Environmental Operating Report2022-05-13013 May 2022 Annual Radiological Environmental Operating Report RA22-018, Annual Dose Report for 20212022-04-22022 April 2022 Annual Dose Report for 2021 RA22-010, Cycle 20 Core Operating Limits Reports2022-03-21021 March 2022 Cycle 20 Core Operating Limits Reports RA21-059, Regulatory Commitment Change Summary Report2021-11-30030 November 2021 Regulatory Commitment Change Summary Report RA21-050, Completion of License Renewal Activities Prior to Entering the Period of Extended Operation2021-11-0909 November 2021 Completion of License Renewal Activities Prior to Entering the Period of Extended Operation RA21-051, County Station Physical Security Plan (Revision 18)2021-09-29029 September 2021 County Station Physical Security Plan (Revision 18) RA21-052, Registration of Use of Cask Store Spent Fuel2021-09-17017 September 2021 Registration of Use of Cask Store Spent Fuel RA21-046, Registration of Use of Cask to Store Spent Fuel2021-08-0505 August 2021 Registration of Use of Cask to Store Spent Fuel RA21-043, Cycle 19 Startup Test Report Summary2021-07-15015 July 2021 Cycle 19 Startup Test Report Summary RA21-031, 2020 Annual Radiological Environmental Operating Report2021-05-13013 May 2021 2020 Annual Radiological Environmental Operating Report RA21-029, Unit 2 - 2020 Annual Radioactive Effluent Release Report2021-04-29029 April 2021 Unit 2 - 2020 Annual Radioactive Effluent Release Report RA21-026, Annual Dose Report for 20202021-04-22022 April 2021 Annual Dose Report for 2020 RA21-016, Netco Rack Insert 10-year Surveillance Report2021-04-15015 April 2021 Netco Rack Insert 10-year Surveillance Report RA21-017, Post-Outage 90-Day Inservice Inspection (ISI) Summary Report2021-04-0909 April 2021 Post-Outage 90-Day Inservice Inspection (ISI) Summary Report RA21-019, Unit 2 Cycle 19 Core Operating Limits Reports2021-04-0707 April 2021 Unit 2 Cycle 19 Core Operating Limits Reports RA20-058, Licensee Post Exam Submittal Letter2020-12-0202 December 2020 Licensee Post Exam Submittal Letter RA20-044, Unit 2 - Unit 1 Cycle 19 Core Operating Limits Report2020-09-10010 September 2020 Unit 2 - Unit 1 Cycle 19 Core Operating Limits Report RA20-026, 2019 Annual Radiological Environmental Operating Report2020-05-13013 May 2020 2019 Annual Radiological Environmental Operating Report RA20-023, Annual Dose Report for 20192020-04-28028 April 2020 Annual Dose Report for 2019 RA20-016, Corrected Drawing for One-Line Electrical Power Supply for Containment Vent System Final Integrated Plan2020-03-24024 March 2020 Corrected Drawing for One-Line Electrical Power Supply for Containment Vent System Final Integrated Plan RA20-015, Triennial Chlorine Survey Report2020-03-20020 March 2020 Triennial Chlorine Survey Report 2024-06-20
[Table view] Category:Report
MONTHYEARRA24-008, Post-Outage Inservice Inspection (ISI) Summary Report2024-06-20020 June 2024 Post-Outage Inservice Inspection (ISI) Summary Report RS-24-014, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2024-02-22022 February 2024 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors ML23181A1522023-06-30030 June 2023 Attachment 2 - LaSalle County Station Units 1 and 2, Pressure and Temperature Limits Report (PTLR) Up to 54 Effective Full-Power Years (Efpy), Revision 3 RS-23-066, Containment Post-Tensioning System Inservice Inspection Basis for Proposed Extension of Examination Interval Technical Report2023-05-0303 May 2023 Containment Post-Tensioning System Inservice Inspection Basis for Proposed Extension of Examination Interval Technical Report RA22-060, Regulatory Commitment Change Summary Report2022-11-30030 November 2022 Regulatory Commitment Change Summary Report ML22332A4552022-11-10010 November 2022 Attachment 7 - BWRVIP-135, Revision 4: BWR Vessel Internals Project Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations (Non-Proprietary Version) ML22332A4542022-11-10010 November 2022 Attachment 6 - LaSalle County Generating Station Units 1 and 2 Pressure and Temperature Limits Report (PTLR) for 54 Effective Full-Power Years (EFPY) (Non-Proprietary) NMP1L3469, Constellation Energy Company, LLC, Request for Use of Honeywell Mururoa V4F1 R Supplied Air Suits2022-06-30030 June 2022 Constellation Energy Company, LLC, Request for Use of Honeywell Mururoa V4F1 R Supplied Air Suits RS-22-018, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors for LaSalle County Station2022-02-22022 February 2022 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors for LaSalle County Station RS-21-115, Supplemental Information for License Amendment Request Regarding New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies, with Proposed Changes to Technical Specifications 4.3.1 and 5.6.52021-11-0404 November 2021 Supplemental Information for License Amendment Request Regarding New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies, with Proposed Changes to Technical Specifications 4.3.1 and 5.6.5 ML21265A5582021-06-30030 June 2021 Attachment 5: Appendix C, NEI 12-16 Criticality Analysis Checklist RA21-016, Netco Rack Insert 10-year Surveillance Report2021-04-15015 April 2021 Netco Rack Insert 10-year Surveillance Report RS-21-001, Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping2021-01-0404 January 2021 Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping ML20303A1752020-10-23023 October 2020 Proposed Relief Request from Section XI Repair/Replacement Documentation for Bolting Replacement of Pressure Retaining Bolting RS-20-133, Response to Request for Additional Information Re License Amendment Request to Renewed Facility Operating Licenses to Adopt 10 CFR 50.692020-10-16016 October 2020 Response to Request for Additional Information Re License Amendment Request to Renewed Facility Operating Licenses to Adopt 10 CFR 50.69 ML20085F8612020-03-24024 March 2020 Corrected Drawing for One-Line Electric Power Supply for Lasalle Hardened Containment Vent System (HCVS) Final Integrated Plan (FIP) RA20-015, Triennial Chlorine Survey Report2020-03-20020 March 2020 Triennial Chlorine Survey Report RS-20-024, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2020-03-0909 March 2020 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors ML19228A0232019-08-15015 August 2019 Proposed Alternative to Utilize Code Case N-879 RA19-005, Regulatory Commitment Change Summary Report2019-01-14014 January 2019 Regulatory Commitment Change Summary Report RA18-110, Submittal of 2017 Regulatory Commitment Change Summary Report2018-12-27027 December 2018 Submittal of 2017 Regulatory Commitment Change Summary Report RS-18-131, Report of Full Compliance with Phase 1 and Phase 2 of June 6, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions2018-12-14014 December 2018 Report of Full Compliance with Phase 1 and Phase 2 of June 6, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions ML18024A2772018-08-16016 August 2018 Lessons Learned from Us NRC Response to the February 2017 LaSalle County Station Unit 2 High Pressure Core Spray Injection Valve Failure RA17-072, Post Accident Monitoring Report2017-08-0202 August 2017 Post Accident Monitoring Report RA17-069, Post Accident Monitoring Report2017-07-21021 July 2017 Post Accident Monitoring Report RS-17-025, Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Focused Evaluation Summary Submittal2017-03-0808 March 2017 Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Focused Evaluation Summary Submittal ML16355A4182017-01-11011 January 2017 Flood Hazard Mitigation Strategies Assessment RA16-092, Regulatory Commitment Change Summary Report2016-12-27027 December 2016 Regulatory Commitment Change Summary Report RS-16-182, Mitigating Strategies Flood Hazard Assessment (Msfha) Submittal2016-10-28028 October 2016 Mitigating Strategies Flood Hazard Assessment (Msfha) Submittal ML16305A2962016-10-27027 October 2016 0000-0163-8881-R0-NP, Revision 0, Exelon Nuclear LaSalle County Generating Station Units 1 & 2 Pool Swell Response. ML16167A5022016-09-0606 September 2016 UFSAR, Rev 22, Chap 09 Figures Part 12 of 14 - Redacted ML16167A4462016-09-0606 September 2016 MF7633 and MF7634, UFSAR, Las Fpr Rev._Redacted RS-16-125, Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2016-08-31031 August 2016 Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident ML16224A7862016-08-26026 August 2016 Notice of Availability of the Final Plant-Specific Supplement 57 to the Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding LaSalle County Station, Units 1 and 2 (TAC Nos. MF5567 and MF5568) ML16200A1562016-07-18018 July 2016 Report on the Safety Aspects of the License Renewal Application of the LaSalle County Station, Units 1 and 2 RS-15-300, Phase 1 (Updated) and Phase 2 Overall Integrated Plan in Response to June 6, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident..2015-12-16016 December 2015 Phase 1 (Updated) and Phase 2 Overall Integrated Plan in Response to June 6, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident.. RA15-075, Transmittal of the 2014 Regulatory Commitment Change Summary Report2015-11-20020 November 2015 Transmittal of the 2014 Regulatory Commitment Change Summary Report ML15107A1542015-04-21021 April 2015 Final Accident Sequence Precursor Analysis for the April 17, 2013, Dual Unit Loss of Offsite Power RS-15-004, County Station Introduction of Lead Use Assemblies2015-01-20020 January 2015 County Station Introduction of Lead Use Assemblies ML14352A1902014-12-18018 December 2014 7491-318563-HAO-1, Rev. 2, LaSalle Requested Documents RS-14-277, Proposed Alternative to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 12014-09-24024 September 2014 Proposed Alternative to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1 ML14177A8592014-06-30030 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima DAI-ICHI Nuclear Power Plant Accident ML14176A9612014-06-24024 June 2014 Submittal of Non-Proprietary BWROG Technical Product, BWROGTP-11-006 - ECCS Containment Walkdown Procedure, Rev 1 (January 2011), as Formally Requested During the Public Meeting Held on April 30, 2014 RS-14-054, LaSalle, Unit 1, Supplemental Information Supporting License Amendment Request to Revise Reactor Coolant System (RCS) Pressure and Temperature (Pit) Curves2014-02-26026 February 2014 LaSalle, Unit 1, Supplemental Information Supporting License Amendment Request to Revise Reactor Coolant System (RCS) Pressure and Temperature (Pit) Curves ML14057A5492014-02-26026 February 2014 Supplemental Information Supporting License Amendment Request to Revise Reactor Coolant System (RCS) Pressure and Temperature (Pit) Curves ML14030A2232014-02-21021 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14042A3252014-02-19019 February 2014 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for LaSalle County Station, Units 1 and 2, TAC Nos.: MF1121 and MF1122 ML13358A3652013-12-20020 December 2013 Attachment 5: LaSalle County Station, Unit 1, Pressure/Temperature Limits Report ML13344A9822013-12-0909 December 2013 2012 Regulatory Commitment Change Summary Report ML13282A3502013-10-0404 October 2013 Final Report, Assessment of Wind Speeds Over the LaSalle County Station Ultimate Heat Sink. 2024-06-20
[Table view] Category:Miscellaneous
MONTHYEARRA24-008, Post-Outage Inservice Inspection (ISI) Summary Report2024-06-20020 June 2024 Post-Outage Inservice Inspection (ISI) Summary Report RS-24-014, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2024-02-22022 February 2024 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors RA22-060, Regulatory Commitment Change Summary Report2022-11-30030 November 2022 Regulatory Commitment Change Summary Report ML21265A5582021-06-30030 June 2021 Attachment 5: Appendix C, NEI 12-16 Criticality Analysis Checklist RA21-016, Netco Rack Insert 10-year Surveillance Report2021-04-15015 April 2021 Netco Rack Insert 10-year Surveillance Report ML20303A1752020-10-23023 October 2020 Proposed Relief Request from Section XI Repair/Replacement Documentation for Bolting Replacement of Pressure Retaining Bolting RA20-015, Triennial Chlorine Survey Report2020-03-20020 March 2020 Triennial Chlorine Survey Report RS-20-024, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2020-03-0909 March 2020 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors RA19-005, Regulatory Commitment Change Summary Report2019-01-14014 January 2019 Regulatory Commitment Change Summary Report RA18-110, Submittal of 2017 Regulatory Commitment Change Summary Report2018-12-27027 December 2018 Submittal of 2017 Regulatory Commitment Change Summary Report RA17-072, Post Accident Monitoring Report2017-08-0202 August 2017 Post Accident Monitoring Report RA17-069, Post Accident Monitoring Report2017-07-21021 July 2017 Post Accident Monitoring Report RS-17-025, Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Focused Evaluation Summary Submittal2017-03-0808 March 2017 Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Focused Evaluation Summary Submittal ML16355A4182017-01-11011 January 2017 Flood Hazard Mitigation Strategies Assessment RA16-092, Regulatory Commitment Change Summary Report2016-12-27027 December 2016 Regulatory Commitment Change Summary Report RS-16-182, Mitigating Strategies Flood Hazard Assessment (Msfha) Submittal2016-10-28028 October 2016 Mitigating Strategies Flood Hazard Assessment (Msfha) Submittal ML16305A2962016-10-27027 October 2016 0000-0163-8881-R0-NP, Revision 0, Exelon Nuclear LaSalle County Generating Station Units 1 & 2 Pool Swell Response. RS-16-125, Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2016-08-31031 August 2016 Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident ML16224A7862016-08-26026 August 2016 Notice of Availability of the Final Plant-Specific Supplement 57 to the Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding LaSalle County Station, Units 1 and 2 (TAC Nos. MF5567 and MF5568) ML16200A1562016-07-18018 July 2016 Report on the Safety Aspects of the License Renewal Application of the LaSalle County Station, Units 1 and 2 RS-15-300, Phase 1 (Updated) and Phase 2 Overall Integrated Plan in Response to June 6, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident..2015-12-16016 December 2015 Phase 1 (Updated) and Phase 2 Overall Integrated Plan in Response to June 6, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident.. RA15-075, Transmittal of the 2014 Regulatory Commitment Change Summary Report2015-11-20020 November 2015 Transmittal of the 2014 Regulatory Commitment Change Summary Report ML14352A1902014-12-18018 December 2014 7491-318563-HAO-1, Rev. 2, LaSalle Requested Documents ML14177A8592014-06-30030 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima DAI-ICHI Nuclear Power Plant Accident ML13344A9822013-12-0909 December 2013 2012 Regulatory Commitment Change Summary Report IR 05000456/20132022013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000289/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000272/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee ML13008A2192013-01-31031 January 2013 U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000254/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000277/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000219/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000352/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000373/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000237/20132022013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000461/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000454/20132022013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee ML12353A2172012-11-0101 November 2012 12Q0108.50-R-002, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the LaSalle County Station, Unit 2, Part 9 of 16 ML12353A2232012-11-0101 November 2012 12Q0108.50-R-002, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the LaSalle County Station, Unit 2, Part 15 of 16 ML12353A2162012-11-0101 November 2012 12Q0108.50-R-002, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the LaSalle County Station, Unit 2, Part 8 of 16 ML12353A2152012-11-0101 November 2012 12Q0108.50-R-002, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the LaSalle County Station, Unit 2, Part 7 of 16 ML12353A2132012-11-0101 November 2012 12Q0108.50-R-002, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the LaSalle County Station, Unit 2, Part 6 of 16 ML12353A2112012-11-0101 November 2012 12Q0108.50-R-002, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the LaSalle County Station, Unit 2, Part 5 of 16 ML12353A2102012-11-0101 November 2012 12Q0108.50-R-002, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the LaSalle County Station, Unit 2, Part 4 of 16 ML12353A2092012-11-0101 November 2012 12Q0108.50-R-002, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the LaSalle County Station, Unit 2, Part 3 of 16 ML12353A2082012-11-0101 November 2012 12Q0108.50-R-002, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the LaSalle County Station, Unit 2, Part 2 of 16 ML12353A2062012-11-0101 November 2012 12Q0108.50-R-001, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the LaSalle County Station, Unit 1, Part 15 of 15 ML12353A2052012-11-0101 November 2012 12Q0108.50-R-001, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the LaSalle County Station, Unit 1, Part 14 of 15 ML12353A2202012-11-0101 November 2012 12Q0108.50-R-002, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the LaSalle County Station, Unit 2, Part 12 of 16 ML12353A2212012-11-0101 November 2012 12Q0108.50-R-002, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the LaSalle County Station, Unit 2, Part 13 of 16 2024-06-20
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.I Exeke n.m Exelon Generation Company, LLC www.exeloncorp.corn LaSalle County Station Nuclear 2601 North 21" Road Marseilles, IL 61341-9757 RA06-013 March 9, 2006 10 CFR 50.46 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-1 1 and NPF-18 NRC Docket Nos. 50-373 and 50-374
Subject:
Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report
Reference:
Letter from D. J. Enright (Exelon Generation Company, LLC) to U. S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated March 9, 2005 In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," Exelon Generation Company, LLC, (EGC) submits the enclosed attachments to fulfill the 30-day and annual reporting requirements for LaSalle County Station (LSCS), Units 1 and 2.
In the referenced letter, EGC reported the fuel peak cladding temperatures (PCTs) calculated based on an acceptable model to be 14000 F for General Electric (GE) fuel. There is no change in PCT for the GE Fuel for this reporting period.
The referenced letter also provided the PCT of 1832TF for the Framatome Advanced Nuclear Power (FANP) fuel based on an acceptable model. Since the last evaluation, FANP ATRIUM-10 fuel has been re-introduced into Unit 1 and a new analysis was performed. Based on the new analysis, the PCT for FANP ATRIUM-10 fuel decreased to a value of 17290F. This is a change of over 500F from the last evaluation using a NRC approved acceptable model. For Unit 2, there were no changes for the FANP ATRIUM-9B fuel and the PCT remains at 18320F.
Unit 1 and Unit 2 employ a mixed core design containing co-resident GE and FANP fuel. The Loss of Coolant Accident (LOCA) analyses of record for both GE and FANP fuel are within all of the acceptance criteria set forth in 10 CFR 50.46.
U. S. Nuclear Regulatory Commission March 9, 2006 Page 2 Attachments 1, 2, and 3 provide PCT information for the limiting LOCA evaluations for LSCS, Units 1 and 2, including all assessments as of February 1, 2005. The assessment notes are contained in Attachment 4 and provide a detailed description for each change or error reported.
Should you have any questions concerning this letter, please contact Mr. Terrence W. Simpkin, Regulatory Assurance Manager, at (815) 415-2800.
Respectfully, I E Daniel J. Enright Plant Manager LaSalle County Station Attachments cc: Regional Administrator - NRC Region IlIl NRC Senior Resident Inspector - LaSalle County Station
Attachment 1 LaSalle Units 1 and 2 10 CFR 50.46 Report (GE Fuel)
PLANT NAME: LaSalle Units I and 2 ECCS EVALUATION MODEL: SAFER/GESTR LOCA REPORT REVISION DATE: February 1,2006 CURRENT OPERATING CYCLES: LIC12* and L2Cl I ANALYSIS OF RECORD Evaluation Model Methodology: NEDE-23785- 1-PA, Rev. 1, "GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident (Volume III),
SAFER/GESTR Application Methodology", October 1984.
Calculation: "Project Task Report, Exelon LaSalle Unit I and 2 SAFER/GESTR Loss-of-Coolant Accident Analysis for GE 14 Fuel," GE report number GE-NE-0000-0022-8684-RI, dated December 2004.
Fuel: GE14 Limiting Single Failure: FIPCS Diesel Generator Limiting Break Size and Double Ended Guillotine of Recirculation Pump Suction Piping Location:
Reference PCT: 14000 F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS IOCFR50.46 report dated March 9, 2005 (Note 9) APCT = 0 :F Net PCT -_ 1400 OF B. CURRENT LOCA MODEL ASSESSMENTS None N/A Net PCT 1 1400 OF
- Currently Unit I is in refueling. Expected Cycle 12 Startup - March 16, 2006
Attachment 2 LaSalle Unit 1 10 CFR 50.46 Report (FANP Fuel)
PLANT NAME: LaSalle Unit I ECCS EVALUATION MODEL: EXEM BWR-2000 Evaluation Model REPORT REVISION DATE: February 1,2006 CURRENT OPERATING CYCLES: LIC12*
ANALYSIS OF RECORD Evaluation Model Methodology: EMF-2361 (P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001.
Calculation: EMF-3230 (P) Revision 0, LaSalle Units I and 2 EXEM BWR-2000 LOCA Break Spectrum Analysis for ATRIUM - 10 Fuel, November 2005.
EMF-3231 (P) Revision 0, LaSalle Units I and 2 EXEM BWR-2000 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM - 10 Fuel, November 2005.
Fuel: ATRIUM- 10 Limiting Single Failure: Low-pressure coolant injection Diesel Generator Limiting Break Size and Double Ended Guillotine/0.8 discharge coefficient of Recirculation Location: Pump Suction Piping Reference PCT: 1729 OF MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS A FANP LOCA analysis was performed in November 2005 to support the re- APCT =0 0F introduction of ATRIUM-10 for LIC12. This analysis addresses all errors and issues. In Cycle 12 there will be no ATRIUM-9B fuel in the Unit I Core Net PCT 1729 'F B. CURRENT LOCA MODEL ASSESSMENTS None N/A Net PCT 1729 F
- Currently Unit I is in refueling. Expected Cycle 12 startup - March 16, 2006.
Attachment 3 LaSalle Unit 2 10 CFR 50.46 Report (FANP Fuel)
PLANT NAME: LaSalle Unit 2 ECCS EVALUATION MODEL: EXEM BWR Evaluation Model REPORT REVISION DATE: February 1,2006 CURRENT OPERATING CYCLE: L2C 11 ANALYSIS OF RECORD Evaluation Model Methodology: Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-9 1-048(P)(A), January 1993.
BWR Jet Pump Model Revision for RELAX, ANF-91 -048(P)(A),
Supplement I and Supplement 2, Siemens Power Corporation, October 1997.
Calculation: 1. LaSalle LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM'-9BFuel, EMF-2175(P), March 1999.
- 2. LOCA Break Spectrum Analysis for LaSalle Units I and 2, EMF-2174(P), March 1999.
- 3. LaSalle Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMT'-10 Fuel, EMF-2641(P), November 2001.
- 4. LaSalle Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM'"'-10Fuel, EMF-2639(P), November 2001.
Fuel: ATRIUM'-9B and ATRIUJMT-10 Limiting Fuel ATRJUMTM-9B Limiting Single Failure: HPCS Diesel Generator Limiting Break Size and Location: 1.1 ft2 Recirculation Pump Discharge Side Line Break Reference PCT: 1807 0F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated May 7, 1999 (See Note 2) APCT= 0 'F 10 CFR 50.46 report dated February 9.2000 (See Note 3) APCT =18 'F 10 CFR 50.46 report dated June 12, 2000 (See Note 4) APCT= O0 F 10 CFR 50.46 report dated June 8,2001 (See Note 5) APCT = O 0 F 10 CFR 50.46 report dated June 8,2002 (See Note 6) APCT = 2 'F 10 CFR 50.46 report dated June 9,2003 (See Note 7) APCT = 5 'F 10 CFR 50.46 report dated March 9,2004 (See Note 1) APCT = 0 'F 10 CFR 50.46 report dated March 9,2005 (See Note 8) APCT= O°F Net PCT 1832 OF B. CURRENT LOCA MODEL ASSESSMENTS No errors/issues for this reporting period APCT = O°F Net PCT 1832 "F
Attachment 4 LaSalle Units 1 and 2 10 CFR 50.46 Report (Assessment Notes)
- 1. Prior LOCA model assessment for FANP fuel During the startup of LaSalle Unit I Cycle II several evaluations were performed for FANP LOCA analysis as reported in the Reference. The net results of these evaluations were that there was a zero degree PCT impact. Additionally a problem was also identified by FANP pertaining with the transfer of RELAX coolant temperature data from PREHUXY to HUXY at the time of core spray. FANP determined that the impact of this problem on the limiting break spectrums results was zero degree. This was also reported in the Reference.
[
Reference:
Letter from Susan R Landahl (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated March 9,2004.]
- 2. Prior LOCA Model Assessment for FANP fuel The May 1999 LOCA model assessment was a new analysis of record for Framatome (Formerly Siemens) due to the introduction of ATRILTM-9B fuel into the Unit 2 Cycle 8 core. Therefore, there is no PCT change. Analysis was performed for a core power of 3722 MWt that bounds the current uprated power of 3489 MWt.
[
Reference:
Letter from J. A. Benjamin (ComEd) to U.S. NRC, "Report of Significant Change in Calculated Peak Cladding Temperature (PCT) - 10CFR 50.46 Report," dated May 7, 1999.]
- 3. Prior LOCA Model Assessment for FANP fuel The February 2000 50.46 report assessed the impact of errors in the LOCA evaluation model.
[
Reference:
Letter from J; A. Benjamin (ComEd) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - IOCFR 50.46 Report," dated February 9,2000.]
- 4. Prior LOCA Model Assessment for FANP fuel The June 2000 10 CFR 50.46 report does not have any PCT assessment for ATRIUM-9B fuel.
[
Reference:
Letter from C. G. Pardee (ComEd) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated June 12, 2000.]
- 5. Prior LOCA model assessment for FANP fuel The reference letter assessed impact of Unit 2 LPCS riser leakage, errors in FANP LOCA analysis model and Unit 2 Cycle 9 reload fuel.
[
Reference:
Letter from M. A Schiavoni (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated June 8, 2001.1
Attachment 4 LaSalle Units 1 and 2 10 CFR 50.46 Report (Assessment Notes)
- 6. Prior LOCA model assessment for FANP fuel The referenced letter assessed impact of errors in FANP LOCA analysis model, Unit 1 Cycle 10 reload fuel and ATRIUM-9B exposure extension.
[
Reference:
Letter from M. A Schiavoni (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated June 8, 2002.1
- 7. Prior LOCA model assessment for FANP fuel The June 2003 50.46 report assessed the impact of errors in the LOCA evaluation, Unit 2 jet pump leakage, Unit 2 Cycle 10 reload Fuel and the Unit 1 mid-cycle reload.
[
Reference:
Letter from Susan R Landahl (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated June 9,2003.]
- 8. Prior LOCA model assessment for FANP fuel The March 2005 10 CFR 50.46 report does not have any PCT assessment.
[
Reference:
Letter from Daniel J. Enright (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated March 9, 2005.]
- 9. Prior LOCA model assessment for GE fuel A GE LOCA analysis was performed in December 2004 to support the introduction of GE 14 for L2C 11. This analysis bounds both LaSalle Units and addressed all errors and issues. This was reported in the Reference.
[
Reference:
Letter from Daniel J. Enright (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated March 9, 2005.]