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| {{#Wiki_filter:Southern Nuclear Operating Company, Inc. Post Office Box 1295 Birmingham, Alabama 35201 -1295 Tel 205.992.5000 February 26, 2007 Docket Nos.: 50-348 50-364 COMPANY Energy to Serve Your WorldsM U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Ladies and Gentlemen: Pursuant to the reporting requirements of 10 CFR 50.46 (a)(3)(ii), Southern Nuclear Operating Company (SNC) is submitting the attached emergency core cooling system (ECCS) evaluation model significant change report for Farley Nuclear Plant Unit 2. Although the changes to the Unit 1 PCT do not meet the criteria for a significant change report, the evaluation results are included for completeness. | | {{#Wiki_filter:Southern Nuclear Operating Company, Inc. |
| This report serves as a 30 day Significant Change Report for small-break LOCA PCT for Unit 2. The change is significant due to the fact that the resulting calculated temperature changes by greater than 50 OF. As shown in Table 1, the small-break LOCA analysis PCT results for both units remain below the 10 CFR 50.46 limit of 2200 OF and therefore, no reanalysis is required. | | Post Office Box 1295 Birmingham, Alabama 35201-1295 Tel 205.992.5000 February 26, 2007 COMPANY Energy to Serve Your WorldsM Docket Nos.: 50-348 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Ladies and Gentlemen: |
| However, as a separate initiative, SNC will perform reanalysis of the small-break LOCA PCT and report the results in the 2007 10 CFR 50.46 ECCS Evaluation Model Annual Report. This letter contains no NRC commitments. | | Pursuant to the reporting requirements of 10 CFR 50.46 (a)(3)(ii), Southern Nuclear Operating Company (SNC) is submitting the attached emergency core cooling system (ECCS) evaluation model significant change report for Farley Nuclear Plant Unit 2. |
| If you have any questions, please advise. Sincerely, L!~Y~ B. J. Georg Manager, Nuclear Licensing | | Although the changes to the Unit 1 PCT do not meet the criteria for a significant change report, the evaluation results are included for completeness. |
| | This report serves as a 30 day Significant Change Report for small-break LOCA PCT for Unit 2. The change is significant due to the fact that the resulting calculated temperature changes by greater than 50 OF. As shown in Table 1, the small-break LOCA analysis PCT results for both units remain below the 10 CFR 50.46 limit of 2200 OF and therefore, no reanalysis is required. However, as a separate initiative, SNC will perform reanalysis of the small-break LOCA PCT and report the results in the 2007 10 CFR 50.46 ECCS Evaluation Model Annual Report. |
| | This letter contains no NRC commitments. If you have any questions, please advise. |
| | Sincerely, L!~Y~ |
| | B. J. Georg Manager, Nuclear Licensing |
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| |
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| ==Enclosure:== | | ==Enclosure:== |
| Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Significant Change Report U. S. Nuclear Regulatory Commission NL-07-0379 Page 2 cc: Southern Nuclear Operating Companv Mr. J. T. Gasser, Executive Vice President Mr. J. R. Johnson, Vice President - Farley Mr. D. H. Jones, Vice President - Engineering RTYPE: CFA04.054; LC# 14541 U. S. Nuclear Redatorv Commission Dr. W. D. Travers, Regional Administrator Ms. K. R. Cotton, NRR Project Manager - Farley Mr. E. L. Crowe, Senior Resident Inspector - Farley Enclosure Joseph M. | | Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Significant Change Report |
| Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Enclosure Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Significant Change Report BACKGROUND In accordance with 10 CFR 50.46(a)(3)(ii), this letter reports changes that have been made to the NOTRUMP Evaluation Model (EM) resulting in a change to the calculated small-break LOCA (SB LOCA) temperature of greater than 50 OF which meets the criteria for a Significant Change Report. DISCUSSION The following presents an assessment of the effect of modifications to the Westinghouse ECCS NOTRUMP Evaluation Model on the Farley SB LOCA analysis results. | | |
| Unit 2 implemented the Reactor Internals Upflow Conversion Program (Reference 2) in 2002, and as such a new PCT rack-up reflecting the new upflow configuration analysis is presented here for Unit 2. Small-Break LOCA Table 1 shows the SB LOCA PCT rack-ups for both Unit 1 and Unit 2. A. SB LOCA ECCS MODEL ANALYSIS-OF-RECORD The SB LOCA analyses for Farley Units 1 and 2 were examined to assess the effects of the above change to the Westinghouse SB LOCA ECCS Evaluation Model on PCT results. | | U. S. Nuclear Regulatory Commission NL-07-0379 Page 2 cc: Southern Nuclear Operating Companv Mr. J. T. Gasser, Executive Vice President Mr. J. R. Johnson, Vice President - Farley Mr. D. H. Jones, Vice President - Engineering RTYPE: CFA04.054; LC# 14541 U. S. Nuclear Redatorv Commission Dr. W. D. Travers, Regional Administrator Ms. K. R. Cotton, NRR Project Manager - Farley Mr. E. L. Crowe, Senior Resident Inspector - Farley |
| The SB LOCA ECCS analysis results were calculated using the NOTRUMP SB LOCA ECCS Evaluation Model (Reference 4). As noted earlier, the Unit 2 re-analysis reflects the Reactor Intemals Upflow Conversion implemented in 2002 (Reference 2). The Unit 1 and Unit 2 analyses assumed the following information important to the SB LOCA analyses: | | |
| o 17x 1 7 VANTAGE+ Fuel Assembly o Core Power | | Enclosure Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Significant Change Report |
| = 1.02 | | |
| * 2775 MWT o Upflow Configuration o FQ =2.50 o FAH = 1.70 For Farley Units 1 and 2, the limiting size break analysis-of-record for the VANTAGE+ fuel analysis is a 3-inch diameter break in the cold leg. The limiting PCT values determined for the Unit 1 and Unit 2 17x1 7 VANTAGE+ small-break are shown in Table 1. | | Enclosure Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Significant Change Report BACKGROUND In accordance with 10 CFR 50.46(a)(3)(ii), this letter reports changes that have been made to the NOTRUMP Evaluation Model (EM) resulting in a change to the calculated small-break LOCA (SB LOCA) temperature of greater than 50 OF which meets the criteria for a Significant Change Report. |
| NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 2 of 6 B. PRIOR SB LOCA ECCS MODEL ASSESSMENTS Prior 10 CFR 50.46 Assessments Reported as Significant The following SB LOCA 10 CFR 50.46 assessment was reported in March 2000 as significant. | | DISCUSSION The following presents an assessment of the effect of modifications to the Westinghouse ECCS NOTRUMP Evaluation Model on the Farley SB LOCA analysis results. |
| An overall PCT benefit of 62 OF for Unit 1 for the "Burst and BlockagetTime in Life" penalty resulted from the SPIKE computer code correlation revision. (Reference | | Unit 2 implemented the Reactor Internals Upflow Conversion Program (Reference 2) in 2002, and as such a new PCT rack-up reflecting the new upflow configuration analysis is presented here for Unit 2. |
| : 8) Prior 10 CFR 50.59 Assessments The following three plant change assessments were reported in the last submittal (Reference
| | Small-Break LOCA Table 1 shows the SB LOCA PCT rack-ups for both Unit 1 and Unit 2. |
| : 1) and occurred prior to 2001. The addition of permanent storage boxes in containment was evaluated and found not to cause a change to PCT (Reference 5).
| | A. SB LOCA ECCS MODEL ANALYSIS-OF-RECORD The SB LOCA analyses for Farley Units 1 and 2 were examined to assess the effects of the above change to the Westinghouse SB LOCA ECCS Evaluation Model on PCT results. The SB LOCA ECCS analysis results were calculated using the NOTRUMP SB LOCA ECCS Evaluation Model (Reference 4). As noted earlier, the Unit 2 re-analysis reflects the Reactor Intemals Upflow Conversion implemented in 2002 (Reference 2). |
| The finalization of Replacement Steam Generator Data resulted in a 62 OF benefit for Unit 1 (Reference 7). Annular pellets were determined to have a 10 OF penalty for SB LOCA results for Unit I (Reference 6). Note that the Unit 2 result (in Table 1) is unaffected by these prior 50.59 plant changes. | | The Unit 1 and Unit 2 analyses assumed the following information important to the SB LOCA analyses: |
| | o 17x17 VANTAGE+ Fuel Assembly o Core Power = 1.02 |
| | * 2775 MWT o Upflow Configuration o FQ =2.50 o FAH = 1.70 For Farley Units 1 and 2, the limiting size break analysis-of-record for the VANTAGE+ fuel analysis is a 3-inch diameter break in the cold leg. The limiting PCT values determined for the Unit 1 and Unit 2 17x17 VANTAGE+ small-break are shown in Table 1. |
| | |
| | NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 2 of 6 B. PRIOR SB LOCA ECCS MODEL ASSESSMENTS Prior 10 CFR 50.46 Assessments Reported as Significant The following SB LOCA 10 CFR 50.46 assessment was reported in March 2000 as significant. |
| | An overall PCT benefit of 62 OF for Unit 1 for the "Burst and BlockagetTime in Life" penalty resulted from the SPIKE computer code correlation revision. (Reference 8) |
| | Prior 10 CFR 50.59 Assessments The following three plant change assessments were reported in the last submittal (Reference 1) and occurred prior to 2001. |
| | The addition of permanent storage boxes in containment was evaluated and found not to cause a change to PCT (Reference 5). |
| | The finalization of Replacement Steam Generator Data resulted in a 62 OF benefit for Unit 1 (Reference 7). |
| | Annular pellets were determined to have a 10 OF penalty for SB LOCA results for Unit I (Reference 6). |
| | Note that the Unit 2 result (in Table 1) is unaffected by these prior 50.59 plant changes. |
| The reason is that the Unit 2 Upflow Conversion implemented in 2002 required a SB LOCA re-analysis that included the above changes explicitly. | | The reason is that the Unit 2 Upflow Conversion implemented in 2002 required a SB LOCA re-analysis that included the above changes explicitly. |
| C. CURRENT SB LOCA ECCS MODEL ASSESSMENTS The following changes and errors were identified: | | C. CURRENT SB LOCA ECCS MODEL ASSESSMENTS The following changes and errors were identified: |
| Prior 10 CFR 50.46 Reported Assessments The following assessments were reported in the last PCT submittal (Reference 1). NOTRUMP Mixture Level TrackingIRegion Depletion Errors Several closely related errors have been discovered in how NOTRUMP deals with the stack mixture level transition across a node boundary in a stack of fluid nodes. | | Prior 10 CFR 50.46 Reported Assessments The following assessments were reported in the last PCT submittal (Reference 1). |
| As previously reported, the impact of this revision on the SB LOCA results has been determined to be a 13 OF penalty for Unit 1. In addition, the associated change in Burst and Blockage/Time in Life Components was an additional 12 OF penalty for Unit 1. Thus, the total change was a 25 OF penalty for Unit 1. This error does not impact Unit 2's NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 3 of 6 re-analysis result (see previously discussed Reactor Internals Upflow Conversion), since the re-analysis was performed with the corrected version of NOTRUMP. Current 10 CFR 50.46 PCT Assessments NOTRUMP-EM Refined Break Spectrum During the course of reviewing several extended power uprate and replacement steam generator SB LOCA analyses, the Nuclear Regulatory Commission (NRC) questioned the break spectrum analyzed in the NOTRUMP evaluation model (EM). The NRC was concerned that the resolution of the break spectrum used in the NOTRUMP EM (1.5,2, 3,4, and 6 inch cases) may not be fine enough to capture the worst break with regard to limiting peak clad temperature as per 10 CFR 50.46. That is, the plant could be SB LOCA limited with regard to overall LOCA results. In response to this, Westinghouse performed some preliminary work indicating that in some cases more limiting results could be obtained fiom non-integer break sizes; however, the magnitude of the impact was far less than that shown in preliminary work performed by the NRC. Based on this, Westinghouse performed evaluations to determine if Farley would maintain compliance with the 10 CFR 50.46 acceptance criteria when considering a refined SB LOCA break spectrum. | | NOTRUMP Mixture Level TrackingIRegion Depletion Errors Several closely related errors have been discovered in how NOTRUMP deals with the stack mixture level transition across a node boundary in a stack of fluid nodes. As previously reported, the impact of this revision on the SB LOCA results has been determined to be a 13 O F penalty for Unit 1. In addition, the associated change in Burst and Blockage/Time in Life Components was an additional 12 OF penalty for Unit 1. |
| It should be noted that use of a refined break spectrum is not an error, but a change, since evaluating only integer break sizes has been the standard practice since the initial licensing of NOTRUMP. The application of this refined break spectrum resulted in a 17 OF benefit for Unit 1 and a 74 OF benefit for Unit 2. CURRENT PLANNED PLANT CHANGE EVALUATIONS Starting with the 2001 annual report, the 10 CFR 50.59 Plant Change PCT values have been considered to be a part of the 50 OF error reporting section. | | Thus, the total change was a 25 OF penalty for Unit 1. This error does not impact Unit 2's |
| The 2005 annual report (Reference | | |
| : 1) was consistent with the change implemented in the 2001 annual report.
| | NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 3 of 6 re-analysis result (see previously discussed Reactor Internals Upflow Conversion), since the re-analysis was performed with the corrected version of NOTRUMP. |
| No applicable changes have been made since that report. Prior 10 CFR 50.59 Model Assessments None. Current Planned Plant Changes None. | | Current 10 CFR 50.46 PCT Assessments NOTRUMP-EM Refined Break Spectrum During the course of reviewing several extended power uprate and replacement steam generator SB LOCA analyses, the Nuclear Regulatory Commission (NRC)questioned the break spectrum analyzed in the NOTRUMP evaluation model (EM). The NRC was concerned that the resolution of the break spectrum used in the NOTRUMP EM (1.5,2, 3,4, and 6 inch cases) may not be fine enough to capture the worst break with regard to limiting peak clad temperature as per 10 CFR 50.46. That is, the plant could be SB LOCA limited with regard to overall LOCA results. |
| NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 4 of 6 E. TOTAL RESULTANT SB LOCA PCT As discussed above, the changes and errors in the Westinghouse SB LOCA ECCS Evaluation Model could affect the SB LOCA analysis results by altering the PCT. As shown in Table 1, the SB LOCA analysis PCT results for both units are below the 10 CFR 50.46 limit of 2200 OF. CONCLUSION As documented in the following table, the updated Farley SB LOCA analyses PCTs remain in compliance with 10 CFR 50.46(b)(l), specifically requiring that the PCT shall not exceed 2200 OF. As such, there is no need for reanalysis or taking any other actions in accordance with 10 CFR 50.46(a)(3)(ii) because compliance with 10 CFR 50.46(b)(l) has been maintained. | | In response to this, Westinghouse performed some preliminary work indicating that in some cases more limiting results could be obtained fiom non-integer break sizes; however, the magnitude of the impact was far less than that shown in preliminary work performed by the NRC. Based on this, Westinghouse performed evaluations to determine if Farley would maintain compliance with the 10 CFR 50.46 acceptance criteria when considering a refined SB LOCA break spectrum. It should be noted that use of a refined break spectrum is not an error, but a change, since evaluating only integer break sizes has been the standard practice since the initial licensing of NOTRUMP. |
| However, as a separate initiative, SNC will perform reanalysis of the SB LOCA PCT. | | The application of this refined break spectrum resulted in a 17 OF benefit for Unit 1 and a 74 OF benefit for Unit 2. |
| SNC will prepare a submittal to the NRC once this reanalysis is complete. | | CURRENT PLANNED PLANT CHANGE EVALUATIONS Starting with the 2001 annual report, the 10 CFR 50.59 Plant Change PCT values have been considered to be a part of the 50 OF error reporting section. The 2005 annual report (Reference 1) was consistent with the change implemented in the 2001 annual report. No applicable changes have been made since that report. |
| NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 5 of 6 REFERENCES 1. Letter from H. L. Surnner, Jr. to USNRC (NL-06-25 13), "Edwin I Hatch Nuclear Plant, Joseph M. Farley Nuclear Plant, Vogtle Electric Generating Plant 10 CFR 50.46 ECCS Evaluation Model Annual Reports for 2005," December 14,2006. 2. ALA-02-039, "Transmittal of Reactor Internals Upflow Conversion Program Engineering Report, J. M. Farley Nuclear Plant Unit 2," June 2002 (also see WCAP-15974, November 2002). 3. LTR-LIS-06-117, "10 CFR 50.46 Annual Notification and Reporting for 2005," March 6, 2006. 4. "Westinghouse Srnall-break ECCS Evaluation Model Using the NOTRUMP Code," WCAP- 10054-P-A (Proprietary), WCAP-1008 1 -A (Non-Proprietary), Lee, N., et. al, August 1985. 5. SECL-97-062. | | Prior 10 CFR 50.59 Model Assessments None. |
| Rev. 1, "Effects on LOCA PCT of Adding Permanent Storage Boxes and Lead Blankets Inside Containment," October 17, 1997. 6. WCAP-15098, "Joseph M. Farley Nuclear Plant Units 1 and 2 RSG Program NSSS Licensing Report," November 1998. 7. ALA-0 1 -0 1, "Southern Nuclear Operating Company, Joseph M. Farley Nuclear Plant Units 1 and 2, LBLOCA and SB LOCA Impacts Due to Final RSG Data for SGRP," February 11, 2000. 8. Letter from D. N. Morey to USNRC (NEL-00-0080), "Joseph M. | | Current Planned Plant Changes None. |
| Farley Nuclear Plant 10 CFR 50.46 Annual ECCS Evaluation Model Changes Report for 1999 and Significant Error Reports," March 29,2000. 9. LTR-LIS-07-69, Revision 1, "10 CFR 50.46 Report for NOTRLTMP-EM Refined Break Spectrum and Revised PCT Rackup Sheets for J. M. Farley Units 1 and 2," February 8,2007. | | |
| NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 6 of 6 TABLE 1 JOSEPH M. | | NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 4 of 6 E. TOTAL RESULTANT SB LOCA PCT As discussed above, the changes and errors in the Westinghouse SB LOCA ECCS Evaluation Model could affect the SB LOCA analysis results by altering the PCT. As shown in Table 1, the SB LOCA analysis PCT results for both units are below the 10 CFR 50.46 limit of 2200 OF. |
| FAFUEY NUCLEAR PLANT TOTAL RESULTANT SMALL-BREAK LOCA PCT (OF) A. SB LOCA ECCS MODEL ANALYSIS-OF-RECORD | | CONCLUSION As documented in the following table, the updated Farley SB LOCA analyses PCTs remain in compliance with 10 CFR 50.46(b)(l), specifically requiring that the PCT shall not exceed 2200 OF. As such, there is no need for reanalysis or taking any other actions in accordance with 10 CFR 50.46(a)(3)(ii) because compliance with 10 CFR 50.46(b)(l) has been maintained. |
| : 1. ECCS Analysis UNIT 1 UNIT 2 1883* 1868** 2. Burst and Blockage / Time in Life 137* 120** Total Analysis-of-Record 2020* 1988* B. PRIOR SB LOCA ECCS MODEL ASSESSMENTS 1. Prior 10 CFR 50.46 Assessments Reported as Significant | | However, as a separate initiative, SNC will perform reanalysis of the SB LOCA PCT. SNC will prepare a submittal to the NRC once this reanalysis is complete. |
| -62* 0 2. Prior 10 CFR 50.59 Assessments | | |
| : a. Addition of Permanent Storage Boxes in Containment 0* 0 b. Finalization of Replacement Steam Generator Data | | NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 5 of 6 REFERENCES |
| -62# 0 c. Annular Pellet Blanket 1 0* 0 Sum of Prior Assessments | | : 1. Letter from H. L. Surnner, Jr. to USNRC (NL-06-2513), "Edwin I Hatch Nuclear Plant, Joseph M. Farley Nuclear Plant, Vogtle Electric Generating Plant 10 CFR 50.46 ECCS Evaluation Model Annual Reports for 2005," December 14,2006. |
| -1 14* 0 C. CURRENT SB LOCA ECCS MODEL ASSESSMENTS | | : 2. ALA-02-039, "Transmittal of Reactor Internals Upflow Conversion Program Engineering Report, J. M. Farley Nuclear Plant Unit 2," June 2002 (also see WCAP-15974, November 2002). |
| : 1. NOTRUMP Mixture Level Tracking / Region Depl Errors 13* ** 2. Associated change in Burst and Blockage 12* ** 3. NOTRUMP-EM Refined Break Spectrum | | : 3. LTR-LIS-06-117, "10 CFR 50.46 Annual Notification and Reporting for 2005," March 6, 2006. |
| -17# -74## D. CURRENT PLANNED PLANT CHANGE EVALUATIONS | | : 4. "Westinghouse Srnall-break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (Proprietary), WCAP-1008 1-A (Non-Proprietary), Lee, N., et. al, August 1985. |
| : 1. None 0 0 E. TOTAL RESULTANT SB LOCA PCT Total - 1914" - 1914** The PCT values are rounded up to the next highest integer number to avoid reporting in decimal points. | | : 5. SECL-97-062. Rev. 1, "Effects on LOCA PCT of Adding Permanent Storage Boxes and Lead Blankets Inside Containment," October 17, 1997. |
| * See References 1 and 3 | | : 6. WCAP-15098, "Joseph M. Farley Nuclear Plant Units 1 and 2 RSG Program NSSS Licensing Report," November 1998. |
| ** The revised analysis-of-record reflects the Unit 2's conversion of downflow to upflow configuration (see References 1 and 2). # See Reference 7 ## See Reference 9}} | | : 7. ALA-0 1-0 1, "Southern Nuclear Operating Company, Joseph M. Farley Nuclear Plant Units 1 and 2, LBLOCA and SB LOCA Impacts Due to Final RSG Data for SGRP," February 11, 2000. |
| | : 8. Letter from D. N. Morey to USNRC (NEL-00-0080), "Joseph M. Farley Nuclear Plant 10 CFR 50.46 Annual ECCS Evaluation Model Changes Report for 1999 and Significant Error Reports," March 29,2000. |
| | : 9. LTR-LIS-07-69, Revision 1, "10 CFR 50.46 Report for NOTRLTMP-EM Refined Break Spectrum and Revised PCT Rackup Sheets for J. M. Farley Units 1 and 2," February 8,2007. |
| | |
| | NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 6 of 6 TABLE 1 JOSEPH M. FAFUEY NUCLEAR PLANT TOTAL RESULTANT SMALL-BREAK LOCA PCT (OF) |
| | A. SB LOCA ECCS MODEL ANALYSIS-OF-RECORD UNIT 1 UNIT 2 |
| | : 1. ECCS Analysis 1883* 1868** |
| | : 2. Burst and Blockage / Time in Life 137* 120** |
| | Total Analysis-of-Record 2020* 1988* |
| | B. PRIOR SB LOCA ECCS MODEL ASSESSMENTS |
| | : 1. Prior 10 CFR 50.46 Assessments Reported as Significant -62* 0 |
| | : 2. Prior 10 CFR 50.59 Assessments |
| | : a. Addition of Permanent Storage Boxes in Containment 0* 0 |
| | : b. Finalization of Replacement Steam Generator Data -62# 0 |
| | : c. Annular Pellet Blanket 10* 0 Sum of Prior Assessments -1 14* 0 C. CURRENT SB LOCA ECCS MODEL ASSESSMENTS |
| | : 1. NOTRUMP Mixture Level Tracking / Region Depl Errors 13* ** |
| | : 2. Associated change in Burst and Blockage 12* ** |
| | : 3. NOTRUMP-EM Refined Break Spectrum -17# -74## |
| | D. CURRENT PLANNED PLANT CHANGE EVALUATIONS |
| | : 1. None 0 0 E. TOTAL RESULTANT SB LOCA PCT Total - |
| | 1914" - |
| | 1914** |
| | The PCT values are rounded up to the next highest integer number to avoid reporting in decimal points. |
| | * See References 1 and 3 |
| | ** The revised analysis-of-record reflects the Unit 2's conversion of downflow to upflow configuration (see References 1 and 2). |
| | # See Reference 7 |
| | ## See Reference 9}} |
|
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Category:Letter type:NL
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Hatch Nuclear Plant ISFSI - Registration of Spent Fuel Cask Use2023-07-13013 July 2023 ISFSI and Edwin I. Hatch Nuclear Plant ISFSI - Registration of Spent Fuel Cask Use NL-23-0555, Request for Exemption from Physical Barrier Requirement2023-07-0707 July 2023 Request for Exemption from Physical Barrier Requirement NL-23-0506, to Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, Improved RAOC and CAOC Fq Surveillance Technical Specifications2023-07-0505 July 2023 to Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, Improved RAOC and CAOC Fq Surveillance Technical Specifications NL-23-0444, Quality Assurance Topical Report Submittal2023-06-15015 June 2023 Quality Assurance Topical Report Submittal NL-23-0457, ISFSI, and Edwin I. Hatch Nuclear Plant, ISFSI - Registration of Spent Fuel Cask Use2023-06-12012 June 2023 ISFSI, and Edwin I. Hatch Nuclear Plant, ISFSI - Registration of Spent Fuel Cask Use NL-23-0449, National Pollutant Discharge Elimination System (NPDES) Permit Renewal Application2023-06-0202 June 2023 National Pollutant Discharge Elimination System (NPDES) Permit Renewal Application NL-23-0383, SNC Response to Regulatory Issue Summary 2023-01:Preparation And.2023-05-19019 May 2023 SNC Response to Regulatory Issue Summary 2023-01:Preparation And. NL-23-0372, Units 1 & 2, Joseph M. Farley Nuclear Plant - Units 1 & 2, Vogtle Electric Generating Plant - Units 1 & 2, Annual Radiological Environmental Operating Reports for 20222023-05-10010 May 2023 Units 1 & 2, Joseph M. Farley Nuclear Plant - Units 1 & 2, Vogtle Electric Generating Plant - Units 1 & 2, Annual Radiological Environmental Operating Reports for 2022 NL-23-0337, Response to Request for Additional Information Related to License Amendment Request to Revise the Frequency of Surveillance Requirement 3.6.3.52023-05-0505 May 2023 Response to Request for Additional Information Related to License Amendment Request to Revise the Frequency of Surveillance Requirement 3.6.3.5 NL-23-0295, Reply to a Notice of Violation; EA-22-1012023-05-0101 May 2023 Reply to a Notice of Violation; EA-22-101 NL-23-0310, Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 20222023-04-25025 April 2023 Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 2022 NL-23-0019, GEN-ISI-ALT-2023-01, Request to Use Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI2023-04-12012 April 2023 GEN-ISI-ALT-2023-01, Request to Use Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI NL-23-0263, Southern Nuclear Company Submittal of Drug and Alcohol Testing Errors Identified 10 CFR 26.719(c) 30-Day Report2023-04-0505 April 2023 Southern Nuclear Company Submittal of Drug and Alcohol Testing Errors Identified 10 CFR 26.719(c) 30-Day Report NL-23-0208, Independent Spent Fuel Storage Installation ISFSI, Decommissioning Funding Plan Triennial Update2023-03-29029 March 2023 Independent Spent Fuel Storage Installation ISFSI, Decommissioning Funding Plan Triennial Update NL-23-0014, Southern Nuclear Operating Co Submittal of Report on Status of Decommissioning Funding2023-03-29029 March 2023 Southern Nuclear Operating Co Submittal of Report on Status of Decommissioning Funding NL-23-0213, Inservice Inspection Program Owner'S Activity Report for Outage 1R312023-03-21021 March 2023 Inservice Inspection Program Owner'S Activity Report for Outage 1R31 NL-23-0228, Nuclear Property Insurance Coverage as of April 1, 2023 and Licensee Guarantees of Payment of Deferred Premiums (10 CFR 140.21)2023-03-20020 March 2023 Nuclear Property Insurance Coverage as of April 1, 2023 and Licensee Guarantees of Payment of Deferred Premiums (10 CFR 140.21) NL-23-0080, Response to Request for Additional Information (RAI) Related to Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-05, Version 1.02023-02-0202 February 2023 Response to Request for Additional Information (RAI) Related to Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-05, Version 1.0 NL-23-0008, Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance Limiting.2023-01-17017 January 2023 Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance Limiting. NL-23-0011, Update to Supporting Documentation Regulatory Conference EA-22-101. Cover Letter Only2023-01-0505 January 2023 Update to Supporting Documentation Regulatory Conference EA-22-101. Cover Letter Only NL-22-0799, License Amendment Request to Revise the Frequency of Surveillance Requirement 3.6.3.52022-12-20020 December 2022 License Amendment Request to Revise the Frequency of Surveillance Requirement 3.6.3.5 NL-22-0897, Supplement to Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A,2022-12-0909 December 2022 Supplement to Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, NL-22-0887, Cycle 32 Core Operating Limits Report2022-11-21021 November 2022 Cycle 32 Core Operating Limits Report NL-22-0884, Response to NRC Inspection Report 05000348/2022440 and 05000364/2022440 EA-22-1012022-11-17017 November 2022 Response to NRC Inspection Report 05000348/2022440 and 05000364/2022440 EA-22-101 NL-20-0170, Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, Improved RAOC2022-10-14014 October 2022 Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, Improved RAOC NL-22-0756, Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-05, Version 1.02022-09-30030 September 2022 Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-05, Version 1.0 NL-22-0289, License Amendment Request to Revise Technical Specification 4.3 Fuel Storage to Correct Tabulated Values from the Associated Spent Fuel Pool (SFP) Criticality Analysis2022-09-21021 September 2022 License Amendment Request to Revise Technical Specification 4.3 Fuel Storage to Correct Tabulated Values from the Associated Spent Fuel Pool (SFP) Criticality Analysis NL-22-0608, Interim 10 CFR 21.21(a)(2) Report Regarding Framatome Supplied Siemens Medium Voltage Circuit Breakers2022-08-17017 August 2022 Interim 10 CFR 21.21(a)(2) Report Regarding Framatome Supplied Siemens Medium Voltage Circuit Breakers NL-22-0552, Inservice Inspection Program Owner'S Activity Report for Outage 2R282022-07-28028 July 2022 Inservice Inspection Program Owner'S Activity Report for Outage 2R28 NL-22-0510, Plants Units 1 and 2, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20212022-07-14014 July 2022 Plants Units 1 and 2, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2021 NL-22-0223, License Amendment to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance Limiting Condition for Operation (LCO) Value2022-06-30030 June 2022 License Amendment to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance Limiting Condition for Operation (LCO) Value NL-22-0361, Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification 5.5.17, Containment Leakage Rate Testing Program to Increase Calculated Peak.2022-06-20020 June 2022 Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification 5.5.17, Containment Leakage Rate Testing Program to Increase Calculated Peak. NL-22-0362, Southern Nuclear Operating Company, Inc. - Foreign Ownership, Control, or Influence Five-Year Renewal Filing and NRC Facility Clearance Update2022-06-0101 June 2022 Southern Nuclear Operating Company, Inc. - Foreign Ownership, Control, or Influence Five-Year Renewal Filing and NRC Facility Clearance Update NL-22-0388, Cycle 29 Core Operating Limits Report2022-05-31031 May 2022 Cycle 29 Core Operating Limits Report NL-22-0344, Annual Radiological Environmental Operating Reports for 20212022-05-10010 May 2022 Annual Radiological Environmental Operating Reports for 2021 NL-22-0340, Response to Request for Additional Information Related to Request for License Amendment to Relocate Augmented Piping Inspection Program Details from Technical2022-05-10010 May 2022 Response to Request for Additional Information Related to Request for License Amendment to Relocate Augmented Piping Inspection Program Details from Technical NL-22-0240, SNC Response to Regulatory Issue Summary 2022-01: Preparation and Scheduling of Operator Licensing Examinations2022-03-31031 March 2022 SNC Response to Regulatory Issue Summary 2022-01: Preparation and Scheduling of Operator Licensing Examinations 2024-01-11
[Table view] Category:Report
MONTHYEARNL-22-0510, Plants Units 1 and 2, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20212022-07-14014 July 2022 Plants Units 1 and 2, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2021 ML21214A3182021-08-0202 August 2021 301 Comments NL-20-0713, Request for One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Pandemic2020-09-21021 September 2020 Request for One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Pandemic NL-20-0034, Inservice Inspection Program Owner'S Activity Report for Outage 1R292020-01-17017 January 2020 Inservice Inspection Program Owner'S Activity Report for Outage 1R29 NL-19-0674, Proposed Alternative to Use ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 12019-09-30030 September 2019 Proposed Alternative to Use ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 1 NL-19-0795, Cameron Non-Proprietary Engineering Reports; ER-1180NP, Rev. 1, ER-1181NP, Rev. 1; ER-1182NP, Rev. 1; and ER-1183NP, Rev. 1.2019-05-31031 May 2019 Cameron Non-Proprietary Engineering Reports; ER-1180NP, Rev. 1, ER-1181NP, Rev. 1; ER-1182NP, Rev. 1; and ER-1183NP, Rev. 1. NL-17-1713, Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure2018-04-0606 April 2018 Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure NL-17-2061, Special Report 2017-002-00, Nonfunctional Radiation Monitor R-60B2017-12-14014 December 2017 Special Report 2017-002-00, Nonfunctional Radiation Monitor R-60B NL-17-1401, Post Accident Monitoring Report for 8 Train RVLIS2017-08-10010 August 2017 Post Accident Monitoring Report for 8 Train RVLIS NL-16-2541, Notification of 10 CFR 26.719(c) 30-Day Report for False Positive Drug Test Results2016-11-29029 November 2016 Notification of 10 CFR 26.719(c) 30-Day Report for False Positive Drug Test Results NL-16-0810, Submittal of Post Accident Monitoring Report2016-06-0909 June 2016 Submittal of Post Accident Monitoring Report NL-16-0683, Special Report No. 2016-002-00, Alternate AC Source Out of Service2016-06-0202 June 2016 Special Report No. 2016-002-00, Alternate AC Source Out of Service NL-16-0751, Special Report 2016-001-00, Nonfunctional Radiation Monitor R-60B and R-60C2016-05-23023 May 2016 Special Report 2016-001-00, Nonfunctional Radiation Monitor R-60B and R-60C ML16103A5722016-04-21021 April 2016 Transmittal of Final Farley Nuclear Plant Unit 2, Accident Sequence Precursor Analysis ML16103A5732016-04-21021 April 2016 Final ASP Analysis (LER 364-14-002) ML15253A4102016-03-11011 March 2016 Enclosure 2 Screening Analysis Report NL-15-2276, Submittal of 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20142015-12-23023 December 2015 Submittal of 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2014 NL-15-1489, 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-12082015-08-17017 August 2015 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-1208 ML15226A2292015-07-31031 July 2015 PWR Vessel Internals Program Plan for Aging Management of Reactor Internals NL-15-1507, J.M. Farley Nuclear Plant, Unit 1, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals2015-07-31031 July 2015 J.M. Farley Nuclear Plant, Unit 1, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals ML15182A1752015-06-26026 June 2015 Completion of Required Action by NRC Order EA-12-051 Reliable Spent Fuel Pool Level Instrumentation ML15106A3102015-04-17017 April 2015 April 22, 2015, Meeting with Southern Nuclear Company, Draft Minimum Shift Staffing Analysis NL-15-0418, Revision 6, Pressure Temperature Limits Reports2015-03-0505 March 2015 Revision 6, Pressure Temperature Limits Reports NL-15-0080, Special Report 2015-001-002015-01-26026 January 2015 Special Report 2015-001-00 NL-14-1899, Vogle, Units 1 and 2 - 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20132014-12-0808 December 2014 Vogle, Units 1 and 2 - 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2013 NL-14-1109, Third Six-Month Status Report of the Implementation of the Requirements of the Commission Order with Regard to Reliable Spent Fuel Pool Instrumentation (EA-12-051)2014-08-26026 August 2014 Third Six-Month Status Report of the Implementation of the Requirements of the Commission Order with Regard to Reliable Spent Fuel Pool Instrumentation (EA-12-051) ML14128A0832014-06-0303 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 ML14101A1192014-04-16016 April 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident ML14098A4752014-04-16016 April 2014 Unit 1 - Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident NL-14-0342, Units 1 and 2, Seismic Hazard and Screening Report for CEUS Sites2014-03-31031 March 2014 Units 1 and 2, Seismic Hazard and Screening Report for CEUS Sites NL-14-0249, Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 501 of 562 Through 542 of 5622014-02-20020 February 2014 Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 501 of 562 Through 542 of 562 NL-14-0249, Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 392 of 562 Through 429 of 5622014-02-20020 February 2014 Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 392 of 562 Through 429 of 562 NL-14-0249, Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 543 of 562 Through Attachment 62014-02-20020 February 2014 Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 543 of 562 Through Attachment 6 NL-14-0249, Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 430 of 562 Through 462 of 5622014-02-20020 February 2014 Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 430 of 562 Through 462 of 562 NL-14-0249, Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 7 Page 57 of 217 Through 107 of 2172014-02-20020 February 2014 Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 7 Page 57 of 217 Through 107 of 217 NL-14-0249, Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 351 of 562 Through 391 of 5622014-02-20020 February 2014 Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 351 of 562 Through 391 of 562 NL-14-0249, Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 313 of 562 Through 350 of 5622014-02-20020 February 2014 Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 313 of 562 Through 350 of 562 ML14051A7342014-02-20020 February 2014 Updated Seismic Recommendation 2.3 Walkdown Report NL-14-0249, Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 463 of 562 Through 500 of 5622014-02-20020 February 2014 Joseph M. Farley, Unit 1, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 463 of 562 Through 500 of 562 NL-14-0249, Enclosure 1, Updated Seismic Recommendation 2.3 Walkdown Report, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 22014-02-20020 February 2014 Enclosure 1, Updated Seismic Recommendation 2.3 Walkdown Report, Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 2 ML14071A0622014-02-20020 February 2014 Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 1 of 562 Through 46 of 562 ML14071A0632014-02-20020 February 2014 Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 7 Page 107 of 217 Through Attachment 8 Through End ML14071A0642014-02-20020 February 2014 Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 47 of 562 Through 90 of 562 ML14071A0652014-02-20020 February 2014 Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 91 of 562 Through 125 of 562 ML14071A0662014-02-20020 February 2014 Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 126 of 562 Through 160 of 562 ML14071A0682014-02-20020 February 2014 Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 197 of 562 Through 233 of 562 ML14071A0692014-02-20020 February 2014 Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 234 of 562 Through 272 of 562 ML14071A0702014-02-20020 February 2014 Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 273 of 562 Through 312 of 562 ML14071A0712014-02-20020 February 2014 Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 313 of 562 Through 350 of 562 ML14071A0762014-02-20020 February 2014 Enclosure 2, Updated Farley Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Force Recommendation 2.3: Seismic Through Attachment 3 Page 501 of 562 Through 542 of 562 2022-07-14
[Table view] Category:Miscellaneous
MONTHYEARML21214A3182021-08-0202 August 2021 301 Comments NL-20-0034, Inservice Inspection Program Owner'S Activity Report for Outage 1R292020-01-17017 January 2020 Inservice Inspection Program Owner'S Activity Report for Outage 1R29 NL-17-1713, Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure2018-04-0606 April 2018 Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure NL-17-2061, Special Report 2017-002-00, Nonfunctional Radiation Monitor R-60B2017-12-14014 December 2017 Special Report 2017-002-00, Nonfunctional Radiation Monitor R-60B NL-16-2541, Notification of 10 CFR 26.719(c) 30-Day Report for False Positive Drug Test Results2016-11-29029 November 2016 Notification of 10 CFR 26.719(c) 30-Day Report for False Positive Drug Test Results NL-16-0810, Submittal of Post Accident Monitoring Report2016-06-0909 June 2016 Submittal of Post Accident Monitoring Report NL-16-0683, Special Report No. 2016-002-00, Alternate AC Source Out of Service2016-06-0202 June 2016 Special Report No. 2016-002-00, Alternate AC Source Out of Service NL-16-0751, Special Report 2016-001-00, Nonfunctional Radiation Monitor R-60B and R-60C2016-05-23023 May 2016 Special Report 2016-001-00, Nonfunctional Radiation Monitor R-60B and R-60C ML15253A4102016-03-11011 March 2016 Enclosure 2 Screening Analysis Report NL-15-2276, Submittal of 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20142015-12-23023 December 2015 Submittal of 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2014 NL-15-1489, 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-12082015-08-17017 August 2015 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-1208 ML15182A1752015-06-26026 June 2015 Completion of Required Action by NRC Order EA-12-051 Reliable Spent Fuel Pool Level Instrumentation ML15106A3102015-04-17017 April 2015 April 22, 2015, Meeting with Southern Nuclear Company, Draft Minimum Shift Staffing Analysis NL-15-0418, Revision 6, Pressure Temperature Limits Reports2015-03-0505 March 2015 Revision 6, Pressure Temperature Limits Reports NL-15-0080, Special Report 2015-001-002015-01-26026 January 2015 Special Report 2015-001-00 NL-14-1899, Vogle, Units 1 and 2 - 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20132014-12-0808 December 2014 Vogle, Units 1 and 2 - 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2013 NL-14-1109, Third Six-Month Status Report of the Implementation of the Requirements of the Commission Order with Regard to Reliable Spent Fuel Pool Instrumentation (EA-12-051)2014-08-26026 August 2014 Third Six-Month Status Report of the Implementation of the Requirements of the Commission Order with Regard to Reliable Spent Fuel Pool Instrumentation (EA-12-051) ML14128A0832014-06-0303 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 ML14101A1192014-04-16016 April 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident NL-14-0145, Proposed Alternative for the Fourth Lnterval Lnservice Inspection (FNP-ISI-ALT-14, Version 1.0)2014-02-18018 February 2014 Proposed Alternative for the Fourth Lnterval Lnservice Inspection (FNP-ISI-ALT-14, Version 1.0) NL-13-1936, 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B2013-09-30030 September 2013 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B NL-13-0954, Proposed Path to Closure of Generic Safety Issue-191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance.2013-05-16016 May 2013 Proposed Path to Closure of Generic Safety Issue-191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance. NL-13-0171, Overall Integrated Plan in Response to March 12,2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation.2013-02-27027 February 2013 Overall Integrated Plan in Response to March 12,2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation. ML13004A2622012-11-26026 November 2012 SNCF164-RPT-02, Ver. 1.0, Farley Unit 2 Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 4, Page 64 of 143 Through Page 129 of 143 ML12355A7872012-11-26026 November 2012 SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3: Seismic Walkdown Checklists, Page 374 of 561 - Page 415 of 561 ML12355A7862012-11-26026 November 2012 SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3: Seismic Walkdown Checklists, Page 331 of 561 - Page 373 of 561 ML12355A7852012-11-26026 November 2012 SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3: Seismic Walkdown Checklists, Page 284 of 561 - Page 330 of 561 ML12355A7842012-11-26026 November 2012 SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3: Seismic Walkdown Checklists, Page 234 of 561 - Page 283 of 561 ML12355A7832012-11-26026 November 2012 SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3: Seismic Walkdown Checklists, Page 184 of 561 - Page 233 of 561 ML12355A7822012-11-26026 November 2012 SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3: Seismic Walkdown Checklists, Page 141 of 561 - Page 183 of 561 ML12355A7812012-11-26026 November 2012 SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3: Seismic Walkdown Checklists, Page 104 of 561 - Page 140 of 561 ML12355A7802012-11-26026 November 2012 SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3: Seismic Walkdown Checklists, Page 59 of 561 - Page 103 of 561 NL-12-2266, SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Cover - Attachment 3: Seismic Walkdown Checklists, Page 58 of 5612012-11-26026 November 2012 SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Cover - Attachment 3: Seismic Walkdown Checklists, Page 58 of 561 ML13004A2632012-11-26026 November 2012 SNCF164-RPT-02, Ver. 1.0, Farley Unit 2 Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 4, Page 130 of 143 Through End NL-12-2266, Joseph M. Farley, Unit 1, SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 4: Unit 1-Area Walk-by Checklists, Page 71 of 161 - Pag2012-11-26026 November 2012 Joseph M. Farley, Unit 1, SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 4: Unit 1-Area Walk-by Checklists, Page 71 of 161 - Page NL-12-2266, Joseph M. Farley, Unit 1, SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 4: Unit 1-Area Walk-by Checklists, Page 1 of 161 - Page2012-11-26026 November 2012 Joseph M. Farley, Unit 1, SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 4: Unit 1-Area Walk-by Checklists, Page 1 of 161 - Page 7 NL-12-2266, Joseph M. Farley, Unit 1, SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3: Seismic Walkdown Checklists, Page 59 of 561 - Page 12012-11-26026 November 2012 Joseph M. Farley, Unit 1, SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3: Seismic Walkdown Checklists, Page 59 of 561 - Page 103 NL-12-2266, Joseph M. Farley, Unit 1, SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Cover - Attachment 3: Seismic Walkdown Checklists, Page 58 of 5612012-11-26026 November 2012 Joseph M. Farley, Unit 1, SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Cover - Attachment 3: Seismic Walkdown Checklists, Page 58 of 561 ML12355A7892012-11-26026 November 2012 SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3: Seismic Walkdown Checklists, Page 445 of 561 - Page 482 of 561 ML13004A2612012-11-26026 November 2012 SNCF164-RPT-02, Ver. 1.0, Farley Unit 2 Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3, Page 431 of 467 Through Attachment 4, Page 63 of 143 ML13004A2602012-11-26026 November 2012 SNCF164-RPT-02, Ver. 1.0, Farley Unit 2 Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3, Page 373 of 467 Through Attachment 3, Page 431 of 467 ML13004A2582012-11-26026 November 2012 SNCF164-RPT-02, Ver. 1.0, Farley Unit 2 Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3, Page 307 of 467 Through Attachment 3, Page 372 of 467 ML13004A2572012-11-26026 November 2012 SNCF164-RPT-02, Ver. 1.0, Farley Unit 2 Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3, Page 232 of 467 Through Attachment 3, Page 306 of 467 ML13004A2562012-11-26026 November 2012 SNCF164-RPT-02, Ver. 1.0, Farley Unit 2 Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3, Page 177 of 467 Through Attachment 3, Page 231 of 467 ML13004A2552012-11-26026 November 2012 SNCF164-RPT-02, Ver. 1.0, Farley Unit 2 Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3, Page 123 of 467 Through Attachment 3, Page 176 of 467 ML13004A2542012-11-26026 November 2012 SNCF164-RPT-02, Ver. 1.0, Farley Unit 2 Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 3, Page 58 of 467 Through Attachment 3, Page 122 of 467 ML13004A2532012-11-26026 November 2012 SNCF164-RPT-02, Ver. 1.0, Farley Unit 2 Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 1, Page 1 of 26 Through Attachment 3, Page 57 of 467 NL-12-2267, SNCF164-RPT-02, Ver. 1.0, Farley Unit 2 Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Cover Through Page 46 of 462012-11-26026 November 2012 SNCF164-RPT-02, Ver. 1.0, Farley Unit 2 Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Cover Through Page 46 of 46 ML12355A7952012-11-26026 November 2012 SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 4: Unit 1-Area Walk-by Checklists, Page 118 of 161 - End ML12355A7942012-11-26026 November 2012 SNCF164-RPT-01, Version 1.0, Seismic Walkdown Report, Rer SNC432467 for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Attachment 4: Unit 1-Area Walk-by Checklists, Page 71 of 161 - Page 117 of 161 2021-08-02
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Southern Nuclear Operating Company, Inc.
Post Office Box 1295 Birmingham, Alabama 35201-1295 Tel 205.992.5000 February 26, 2007 COMPANY Energy to Serve Your WorldsM Docket Nos.: 50-348 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Ladies and Gentlemen:
Pursuant to the reporting requirements of 10 CFR 50.46 (a)(3)(ii), Southern Nuclear Operating Company (SNC) is submitting the attached emergency core cooling system (ECCS) evaluation model significant change report for Farley Nuclear Plant Unit 2.
Although the changes to the Unit 1 PCT do not meet the criteria for a significant change report, the evaluation results are included for completeness.
This report serves as a 30 day Significant Change Report for small-break LOCA PCT for Unit 2. The change is significant due to the fact that the resulting calculated temperature changes by greater than 50 OF. As shown in Table 1, the small-break LOCA analysis PCT results for both units remain below the 10 CFR 50.46 limit of 2200 OF and therefore, no reanalysis is required. However, as a separate initiative, SNC will perform reanalysis of the small-break LOCA PCT and report the results in the 2007 10 CFR 50.46 ECCS Evaluation Model Annual Report.
This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely, L!~Y~
B. J. Georg Manager, Nuclear Licensing
Enclosure:
Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Significant Change Report
U. S. Nuclear Regulatory Commission NL-07-0379 Page 2 cc: Southern Nuclear Operating Companv Mr. J. T. Gasser, Executive Vice President Mr. J. R. Johnson, Vice President - Farley Mr. D. H. Jones, Vice President - Engineering RTYPE: CFA04.054; LC# 14541 U. S. Nuclear Redatorv Commission Dr. W. D. Travers, Regional Administrator Ms. K. R. Cotton, NRR Project Manager - Farley Mr. E. L. Crowe, Senior Resident Inspector - Farley
Enclosure Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Significant Change Report
Enclosure Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Significant Change Report BACKGROUND In accordance with 10 CFR 50.46(a)(3)(ii), this letter reports changes that have been made to the NOTRUMP Evaluation Model (EM) resulting in a change to the calculated small-break LOCA (SB LOCA) temperature of greater than 50 OF which meets the criteria for a Significant Change Report.
DISCUSSION The following presents an assessment of the effect of modifications to the Westinghouse ECCS NOTRUMP Evaluation Model on the Farley SB LOCA analysis results.
Unit 2 implemented the Reactor Internals Upflow Conversion Program (Reference 2) in 2002, and as such a new PCT rack-up reflecting the new upflow configuration analysis is presented here for Unit 2.
Small-Break LOCA Table 1 shows the SB LOCA PCT rack-ups for both Unit 1 and Unit 2.
A. SB LOCA ECCS MODEL ANALYSIS-OF-RECORD The SB LOCA analyses for Farley Units 1 and 2 were examined to assess the effects of the above change to the Westinghouse SB LOCA ECCS Evaluation Model on PCT results. The SB LOCA ECCS analysis results were calculated using the NOTRUMP SB LOCA ECCS Evaluation Model (Reference 4). As noted earlier, the Unit 2 re-analysis reflects the Reactor Intemals Upflow Conversion implemented in 2002 (Reference 2).
The Unit 1 and Unit 2 analyses assumed the following information important to the SB LOCA analyses:
o 17x17 VANTAGE+ Fuel Assembly o Core Power = 1.02
- 2775 MWT o Upflow Configuration o FQ =2.50 o FAH = 1.70 For Farley Units 1 and 2, the limiting size break analysis-of-record for the VANTAGE+ fuel analysis is a 3-inch diameter break in the cold leg. The limiting PCT values determined for the Unit 1 and Unit 2 17x17 VANTAGE+ small-break are shown in Table 1.
NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 2 of 6 B. PRIOR SB LOCA ECCS MODEL ASSESSMENTS Prior 10 CFR 50.46 Assessments Reported as Significant The following SB LOCA 10 CFR 50.46 assessment was reported in March 2000 as significant.
An overall PCT benefit of 62 OF for Unit 1 for the "Burst and BlockagetTime in Life" penalty resulted from the SPIKE computer code correlation revision. (Reference 8)
Prior 10 CFR 50.59 Assessments The following three plant change assessments were reported in the last submittal (Reference 1) and occurred prior to 2001.
The addition of permanent storage boxes in containment was evaluated and found not to cause a change to PCT (Reference 5).
The finalization of Replacement Steam Generator Data resulted in a 62 OF benefit for Unit 1 (Reference 7).
Annular pellets were determined to have a 10 OF penalty for SB LOCA results for Unit I (Reference 6).
Note that the Unit 2 result (in Table 1) is unaffected by these prior 50.59 plant changes.
The reason is that the Unit 2 Upflow Conversion implemented in 2002 required a SB LOCA re-analysis that included the above changes explicitly.
C. CURRENT SB LOCA ECCS MODEL ASSESSMENTS The following changes and errors were identified:
Prior 10 CFR 50.46 Reported Assessments The following assessments were reported in the last PCT submittal (Reference 1).
NOTRUMP Mixture Level TrackingIRegion Depletion Errors Several closely related errors have been discovered in how NOTRUMP deals with the stack mixture level transition across a node boundary in a stack of fluid nodes. As previously reported, the impact of this revision on the SB LOCA results has been determined to be a 13 O F penalty for Unit 1. In addition, the associated change in Burst and Blockage/Time in Life Components was an additional 12 OF penalty for Unit 1.
Thus, the total change was a 25 OF penalty for Unit 1. This error does not impact Unit 2's
NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 3 of 6 re-analysis result (see previously discussed Reactor Internals Upflow Conversion), since the re-analysis was performed with the corrected version of NOTRUMP.
Current 10 CFR 50.46 PCT Assessments NOTRUMP-EM Refined Break Spectrum During the course of reviewing several extended power uprate and replacement steam generator SB LOCA analyses, the Nuclear Regulatory Commission (NRC)questioned the break spectrum analyzed in the NOTRUMP evaluation model (EM). The NRC was concerned that the resolution of the break spectrum used in the NOTRUMP EM (1.5,2, 3,4, and 6 inch cases) may not be fine enough to capture the worst break with regard to limiting peak clad temperature as per 10 CFR 50.46. That is, the plant could be SB LOCA limited with regard to overall LOCA results.
In response to this, Westinghouse performed some preliminary work indicating that in some cases more limiting results could be obtained fiom non-integer break sizes; however, the magnitude of the impact was far less than that shown in preliminary work performed by the NRC. Based on this, Westinghouse performed evaluations to determine if Farley would maintain compliance with the 10 CFR 50.46 acceptance criteria when considering a refined SB LOCA break spectrum. It should be noted that use of a refined break spectrum is not an error, but a change, since evaluating only integer break sizes has been the standard practice since the initial licensing of NOTRUMP.
The application of this refined break spectrum resulted in a 17 OF benefit for Unit 1 and a 74 OF benefit for Unit 2.
CURRENT PLANNED PLANT CHANGE EVALUATIONS Starting with the 2001 annual report, the 10 CFR 50.59 Plant Change PCT values have been considered to be a part of the 50 OF error reporting section. The 2005 annual report (Reference 1) was consistent with the change implemented in the 2001 annual report. No applicable changes have been made since that report.
Prior 10 CFR 50.59 Model Assessments None.
Current Planned Plant Changes None.
NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 4 of 6 E. TOTAL RESULTANT SB LOCA PCT As discussed above, the changes and errors in the Westinghouse SB LOCA ECCS Evaluation Model could affect the SB LOCA analysis results by altering the PCT. As shown in Table 1, the SB LOCA analysis PCT results for both units are below the 10 CFR 50.46 limit of 2200 OF.
CONCLUSION As documented in the following table, the updated Farley SB LOCA analyses PCTs remain in compliance with 10 CFR 50.46(b)(l), specifically requiring that the PCT shall not exceed 2200 OF. As such, there is no need for reanalysis or taking any other actions in accordance with 10 CFR 50.46(a)(3)(ii) because compliance with 10 CFR 50.46(b)(l) has been maintained.
However, as a separate initiative, SNC will perform reanalysis of the SB LOCA PCT. SNC will prepare a submittal to the NRC once this reanalysis is complete.
NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 5 of 6 REFERENCES
- 1. Letter from H. L. Surnner, Jr. to USNRC (NL-06-2513), "Edwin I Hatch Nuclear Plant, Joseph M. Farley Nuclear Plant, Vogtle Electric Generating Plant 10 CFR 50.46 ECCS Evaluation Model Annual Reports for 2005," December 14,2006.
- 2. ALA-02-039, "Transmittal of Reactor Internals Upflow Conversion Program Engineering Report, J. M. Farley Nuclear Plant Unit 2," June 2002 (also see WCAP-15974, November 2002).
- 3. LTR-LIS-06-117, "10 CFR 50.46 Annual Notification and Reporting for 2005," March 6, 2006.
- 4. "Westinghouse Srnall-break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (Proprietary), WCAP-1008 1-A (Non-Proprietary), Lee, N., et. al, August 1985.
- 5. SECL-97-062. Rev. 1, "Effects on LOCA PCT of Adding Permanent Storage Boxes and Lead Blankets Inside Containment," October 17, 1997.
- 6. WCAP-15098, "Joseph M. Farley Nuclear Plant Units 1 and 2 RSG Program NSSS Licensing Report," November 1998.
- 7. ALA-0 1-0 1, "Southern Nuclear Operating Company, Joseph M. Farley Nuclear Plant Units 1 and 2, LBLOCA and SB LOCA Impacts Due to Final RSG Data for SGRP," February 11, 2000.
- 8. Letter from D. N. Morey to USNRC (NEL-00-0080), "Joseph M. Farley Nuclear Plant 10 CFR 50.46 Annual ECCS Evaluation Model Changes Report for 1999 and Significant Error Reports," March 29,2000.
- 9. LTR-LIS-07-69, Revision 1, "10 CFR 50.46 Report for NOTRLTMP-EM Refined Break Spectrum and Revised PCT Rackup Sheets for J. M. Farley Units 1 and 2," February 8,2007.
NL-07-0379, Enclosure 10 CFR 50.46 ECCS Evaluation Model Significant Change Report Page 6 of 6 TABLE 1 JOSEPH M. FAFUEY NUCLEAR PLANT TOTAL RESULTANT SMALL-BREAK LOCA PCT (OF)
A. SB LOCA ECCS MODEL ANALYSIS-OF-RECORD UNIT 1 UNIT 2
- 1. ECCS Analysis 1883* 1868**
- 2. Burst and Blockage / Time in Life 137* 120**
Total Analysis-of-Record 2020* 1988*
B. PRIOR SB LOCA ECCS MODEL ASSESSMENTS
- 1. Prior 10 CFR 50.46 Assessments Reported as Significant -62* 0
- 2. Prior 10 CFR 50.59 Assessments
- a. Addition of Permanent Storage Boxes in Containment 0* 0
- b. Finalization of Replacement Steam Generator Data -62# 0
- c. Annular Pellet Blanket 10* 0 Sum of Prior Assessments -1 14* 0 C. CURRENT SB LOCA ECCS MODEL ASSESSMENTS
- 1. NOTRUMP Mixture Level Tracking / Region Depl Errors 13* **
- 2. Associated change in Burst and Blockage 12* **
- 3. NOTRUMP-EM Refined Break Spectrum -17# -74##
D. CURRENT PLANNED PLANT CHANGE EVALUATIONS
- 1. None 0 0 E. TOTAL RESULTANT SB LOCA PCT Total -
1914" -
1914**
The PCT values are rounded up to the next highest integer number to avoid reporting in decimal points.
- The revised analysis-of-record reflects the Unit 2's conversion of downflow to upflow configuration (see References 1 and 2).
- See Reference 7
- See Reference 9