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{{#Wiki_filter:Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) | {{#Wiki_filter:Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) | ||
: 4. The Rod Block Monitor Upscale Instrumentation Setpoint for the Rod Block Monitor-Upscale Function Allowable Value for Specification 3.3.2.1.5. The OPRM setpoints for the trip function for SR 3.3.1.3.3. | : 4. The Rod Block Monitor Upscale Instrumentation Setpoint for the Rod Block Monitor-Upscale Function Allowable Value for Specification 3.3.2.1. | ||
: 5. The OPRM setpoints for the trip function for SR 3.3.1.3.3. | |||
: b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: | : b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: | ||
: 1. XN-NF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors." 2. ANF-913(P)(A), "COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analysis." 3. ANF-CC-33(P)(A), "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50, Appendix K Heatup Option." 4. XN-NF-80-19(P)(A), "Advanced Nuclear Fuel Methodology for Boiling Water Reactors." 5. XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel." 6. EMF-CC-074(P)(A), Volume 4 -"BWR Stability Analysis: Assessment of STAIF with input from MICROBURN-B2." 7. XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model." 8. XN-NF-84-105(P)(A), "XCOBRA-T: | : 1. XN-NF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors." | ||
A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis." (continued) | : 2. ANF-913(P)(A), "COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analysis." | ||
LaSalle 1 and 2 5.6-3 Amendment No181 168 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) | : 3. ANF-CC-33(P)(A), "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50, Appendix K Heatup Option." | ||
: 9. EMF-2209(P)(A), "SPCB Critical Power Correlation." 10. ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs." 11. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel." 12. NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods." 13. EMF-85-74(P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model." 14. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: | : 4. XN-NF-80-19(P)(A), "Advanced Nuclear Fuel Methodology for Boiling Water Reactors." | ||
Evaluation and Validation of CASMO-4/MICROBURN-B2." 15. NEDC-33106P, "GEXL97 Correlation for Atrium-i0 Fuel." 16. EMF-2245(P)(A), "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel." 17. EMF-2361(P)(A), "EXEM BWR-2000 ECCS Evaluation Model." 18. NEDO-32465-A, "BWR Owners' Group Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications," August 1996.19. ANF-1358(P)(A), "The Loss of Feedwater Heating Transient in Boiling Water Reactors." (continued) | : 5. XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel." | ||
LaSalle 1 and 2 5.6-4 Amendment No.1 | : 6. EMF-CC-074(P)(A), Volume 4 - "BWR Stability Analysis: | ||
Assessment of STAIF with input from MICROBURN-B2." | |||
: 7. XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model." | |||
: 8. XN-NF-84-105(P)(A), "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis." | |||
(continued) | |||
LaSalle 1 and 2 5.6-3 Amendment No181 168 | |||
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) | |||
: 9. EMF-2209(P)(A), "SPCB Critical Power Correlation." | |||
: 10. ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs." | |||
: 11. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel." | |||
: 12. NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods." | |||
: 13. EMF-85-74(P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model." | |||
: 14. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2." | |||
: 15. NEDC-33106P, "GEXL97 Correlation for Atrium-i0 Fuel." | |||
: 16. EMF-2245(P)(A), "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel." | |||
: 17. EMF-2361(P)(A), "EXEM BWR-2000 ECCS Evaluation Model." | |||
: 18. NEDO-32465-A, "BWR Owners' Group Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications," August 1996. | |||
: 19. ANF-1358(P)(A), "The Loss of Feedwater Heating Transient in Boiling Water Reactors." | |||
(continued) | |||
LaSalle 1 and 2 5.6-4 Amendment No.1 81 / 1 68}} |
Latest revision as of 10:19, 23 November 2019
ML070520035 | |
Person / Time | |
---|---|
Site: | LaSalle ![]() |
Issue date: | 02/15/2007 |
From: | NRC/NRR/ADRO/DORL/LPLIII-2 |
To: | |
Sands S,NRR/DORL, 415-3154 | |
Shared Package | |
ml062220400 | List: |
References | |
TAC MD1356, TAC MD1357 | |
Download: ML070520035 (2) | |
Text
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 4. The Rod Block Monitor Upscale Instrumentation Setpoint for the Rod Block Monitor-Upscale Function Allowable Value for Specification 3.3.2.1.
- 5. The OPRM setpoints for the trip function for SR 3.3.1.3.3.
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1. XN-NF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors."
- 2. ANF-913(P)(A), "COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analysis."
- 3. ANF-CC-33(P)(A), "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50, Appendix K Heatup Option."
- 4. XN-NF-80-19(P)(A), "Advanced Nuclear Fuel Methodology for Boiling Water Reactors."
- 5. XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel."
- 6. EMF-CC-074(P)(A), Volume 4 - "BWR Stability Analysis:
Assessment of STAIF with input from MICROBURN-B2."
- 7. XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model."
- 8. XN-NF-84-105(P)(A), "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis."
(continued)
LaSalle 1 and 2 5.6-3 Amendment No181 168
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 9. EMF-2209(P)(A), "SPCB Critical Power Correlation."
- 10. ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs."
- 11. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel."
- 12. NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods."
- 13. EMF-85-74(P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model."
- 14. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2."
- 15. NEDC-33106P, "GEXL97 Correlation for Atrium-i0 Fuel."
- 16. EMF-2245(P)(A), "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel."
- 17. EMF-2361(P)(A), "EXEM BWR-2000 ECCS Evaluation Model."
- 18. NEDO-32465-A, "BWR Owners' Group Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications," August 1996.
(continued)
LaSalle 1 and 2 5.6-4 Amendment No.1 81 / 1 68