05000335/FIN-2015004-05: Difference between revisions
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| identified by = NRC | | identified by = NRC | ||
| Inspection procedure = IP 71152 | | Inspection procedure = IP 71152 | ||
| Inspector = A Goldau, A Sengupta, B Collins, D Bacon, J Reyes, J Rivera | | Inspector = A Goldau, A Sengupta, B Collins, D Bacon, J Reyes, J Rivera-Ortiz, L Suggs, P Capehart, T Morrissey | ||
| CCA = N/A for ROP | | CCA = N/A for ROP | ||
| INPO aspect = | | INPO aspect = | ||
| description = An NRC-identified, Non-cited Violation of 10 CFR Appendix B, Criterion III, Design Control, was identified for the failure to verify the adequacy of the Unit 1 and Unit 2 replacement steam generators (RSGs) design with respect to the requirements in the American Society of Mechanical Engineers Boiler Pressure Vessel Code (ASME Code), Section III, Article NB-3000, for the primary stress and fatigue analyses of the pressure-retaining tube-to-tubesheet welds. The licensee entered the issue in the corrective action program, and performed the required analyses for the Unit 1 and Unit 2 RSGs to demonstrate that the design met the ASME Code requirements. The inspectors used the guidance in NRC Inspector Manual Chapter (IMC) 0612, Appendix B, Issue Screening, and determined that the performance deficiency was more-than-minor because it was associated with the design control attribute of the Initiating Events Cornerstone, and adversely affected the cornerstone objective. Specifically, the failure to verify that the required stress and fatigue analyses were performed in accordance with the ASME Code did not support the objective of limiting the likelihood of primary-to-secondary leakage events that could upset plant stability and challenge critical safety functions during shutdown, as well as power operations. The inspectors evaluated this finding using NRC IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, Exhibit 1 Initiating Events Screening Questions. The finding screened as Green because the stress calculations demonstrated that there was no degraded steam generator (SG) tube condition where one tube could not sustain three times the differential pressure across a tube during normal full power, and none of the SGs violated the accident leakage performance criterion. Additionally, the stress calculations demonstrated that the finding did not result in a condition that exceeded the reactor coolant system leak rate for a small loss of coolant accident (LOCA), or affected other systems used to mitigate a LOCA resulting in a total loss of their function (e.g., Interfacing System LOCA). The inspectors determined that no cross-cutting aspect was associated with this finding because the performance deficiency occurred more than 3 years ago, and it was not reflective of present performance. | | description = An NRC-identified, Non-cited Violation of 10 CFR Appendix B, Criterion III, Design Control, was identified for the failure to verify the adequacy of the Unit 1 and Unit 2 replacement steam generators (RSGs) design with respect to the requirements in the American Society of Mechanical Engineers Boiler Pressure Vessel Code (ASME Code), Section III, Article NB-3000, for the primary stress and fatigue analyses of the pressure-retaining tube-to-tubesheet welds. The licensee entered the issue in the corrective action program, and performed the required analyses for the Unit 1 and Unit 2 RSGs to demonstrate that the design met the ASME Code requirements. The inspectors used the guidance in NRC Inspector Manual Chapter (IMC) 0612, Appendix B, Issue Screening, and determined that the performance deficiency was more-than-minor because it was associated with the design control attribute of the Initiating Events Cornerstone, and adversely affected the cornerstone objective. Specifically, the failure to verify that the required stress and fatigue analyses were performed in accordance with the ASME Code did not support the objective of limiting the likelihood of primary-to-secondary leakage events that could upset plant stability and challenge critical safety functions during shutdown, as well as power operations. The inspectors evaluated this finding using NRC IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, Exhibit 1 Initiating Events Screening Questions. The finding screened as Green because the stress calculations demonstrated that there was no degraded steam generator (SG) tube condition where one tube could not sustain three times the differential pressure across a tube during normal full power, and none of the SGs violated the accident leakage performance criterion. Additionally, the stress calculations demonstrated that the finding did not result in a condition that exceeded the reactor coolant system leak rate for a small loss of coolant accident (LOCA), or affected other systems used to mitigate a LOCA resulting in a total loss of their function (e.g., Interfacing System LOCA). The inspectors determined that no cross-cutting aspect was associated with this finding because the performance deficiency occurred more than 3 years ago, and it was not reflective of present performance. | ||
}} | }} | ||
Latest revision as of 20:53, 20 February 2018
| Site: | Saint Lucie  |
|---|---|
| Report | IR 05000335/2015004 Section 4OA2 |
| Date counted | Dec 31, 2015 (2015Q4) |
| Type: | NCV: Green |
| cornerstone | Initiating Events |
| Identified by: | NRC identified |
| Inspection Procedure: | IP 71152 |
| Inspectors (proximate) | A Goldau A Sengupta B Collins D Bacon J Reyes J Rivera-Ortiz L Suggs P Capehart T Morrissey |
| Violation of: | 10 CFR 50 Appendix B Criterion III, Design Control |
| INPO aspect | |
| ' | |