Evaluation of the potential safety significance of shallow flaws in reactor pressure vessels during normal plant cooldowns
A report published by Oak Ridge National Laboratories in February 2016, "The Effect of Shallow Inside-Surface-Breaking Flaws on the Probability of Brittle Fracture of Reactors Subjected to Postulated and Actual Operational Cool-Down Transients: A Status Report," analyzed reactor pressure vessels (RPVs) during routine plant cooldowns from operating temperature and pressure to cold shutdown. These plant cooldowns were analyzed using the probabilistic fracture mechanics (PFM) code FAVOR to determine the Conditional Probability of vessel Failure (CPF) for inner diameter (ID) small surface-breaking flaws (SSBFs) with various depths. This ORNL 2016 analysis showed that the Conditional Probability of crack Initiation (CPI) and CPF for shallow, circumferential, ID surface-breaking flaws that extend just into the ferritic steel vessel wall may have calculated values significantly greater than those for the ¼ thickness reference flaw in the ASME Code.
To determine whether the higher CPF values shown in the ORNL 2016 study could result in estimated Through-Wall Cracking Frequency (TWCF) values higher than the safety goal of 10-6 per year, additional sensitivity studies were performed as described in two interim reports (ML21260A245 and ML21260A246) and a final summary report (ML21260A246), "Summary of Investigations into Addressing the Shallow Surface-Breaking Flaw Issue" (see links below). Nuclear power plants typically cool down for refueling outages approximately once every 18-24 months depending on the plant fuel cycle design. Plants may also cool down to cold shutdown during an operating cycle for required maintenance during a cycle. Therefore, the frequency of a normal plant cooldown is conservatively assumed in this analysis as once per year. These reports evaluated plant cooldowns at the maximum pressure-temperature (P-T) limit curves. Plants normally maintain margin to these limit curves to avoid exceeding them. Therefore, the occurrence of these cooldowns at the maximum allowed limits is hypothetical, with estimated frequencies of 6 x 10-6 per year for PWRs and 10-7 per year for BWRs per Table 1 of Enclosure 6 in the NRC Branch Technical Position (BTP) 5-3 closure memorandum.
As discussed in Section 1 of the summary report, cooldowns from plant operating temperature to 70°F ambient temperature at a constant cooldown rate (CDR) of 50°F to 100°F per hour along the P-T limit curve may result in a CPF above 10-6 per year. Several potential changes in FAVOR modeling assumptions were investigated to assess the sensitivity of the FAVOR results to these changes. The changes in modeling assumptions evaluated in Sections 2 and 3 of this report include (1) Stress-Free Temperature, (2) Coefficient of Thermal Expansion, (3) Warm Pre-Stress and (4) Shop Hydro Testing. No changes in FAVOR plant modeling assumptions were identified that systematically reduce CPF to less than 10-6 per year for a constant CDR above 50°F per hour.
As discussed in Appendix A of the summary report, 42 actual plant cooldown histories were obtained from 17 different PWRs. These cooldown histories were analyzed to determine the pressure and temperature histories corresponding to various percentiles of the pressure and temperature history distributions based on the population considered. The 42 plant cooldowns appeared to represent successfully completed cooldowns, and were treated as a representative sample. As discussed in Section 4.1, the CPF values for internal SSBF flaws based on FAVOR analyses of the actual PWR plant cooldowns shown in Appendix A are significantly less than 10-6 per year. The studies documented in the shallow flaw reports demonstrate that, when considering realistic transient frequencies, the probability of brittle fracture leading to a TWCF is far below the safety goal of 10-6 per year.
https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML21260A248
Smith, M., T. Dickson, A. Dyszel (NUMARK Associates, Inc.); P. Raynaud (U.S. Nuclear Regulatory Commission), Summary of Investigations into Addressing the Shallow Surface-Breaking Flaw Issue, Technical Letter Report TLR-RES/DE/REB-2-21-13, December 2021.
Smith, M., T. Dickson, A. Dyszel (NUMARK Associates, Inc.); P. Raynaud (U.S. Nuclear Regulatory Commission), Assessment of Reactor Pressure Vessel Inside Diameter Shallow Surface Breaking Flaws, First Progress Report: October 2018, Technical Letter Report TLR-RES/DE/REB-2-21-13, Enclosure 1, December 2021.
Smith, M., T. Dickson, A. Dyszel (NUMARK Associates, Inc.); P. Raynaud (U.S. Nuclear Regulatory Commission), Assessment of Reactor Pressure Vessel Inside Diameter Shallow Surface Breaking Flaws, Second Progress Report: July 2019, Technical Letter Report TLR-RES/DE/REB-2-21-13, Enclosure 2, December 2021.
