Pressurized Thermal Shock Technical Basis Reevaluation Project
October 12, 2000
Dr. William D. Travers
Executive Director for Operations
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555-0001
Dear Dr. Travers:
|SUBJECT: ||PRESSURIZED THERMAL SHOCK TECHNICAL BASIS REEVALUATION PROJECT |
During the 476th meeting of the Advisory Committee on Reactor Safeguards, October 5-7, 2000, we reviewed the status of activities associated with the Pressurized Thermal Shock (PTS) Technical Basis Reevaluation Project. During our 472nd meeting, May 11-13, 2000, we reviewed a draft Commission Paper concerning revising the technical basis for the PTS screening criterion. Our Subcommittee on Materials and Metallurgy met on September 21, 2000, to review draft reports associated with the Reevaluation Project. During our reviews we had the benefit of discussions with the NRC staff and of the documents referenced.
Recommendations and Conclusions
|1. ||The staff should examine the implications of using a large early release frequency (LERF) acceptance guideline based on an air-oxidation source term on the acceptance value for reactor pressure vessel failure frequency. |
|2. ||The reevaluation of the PTS rule is a well thought out project that integrates the results of thermal-hydraulic, probabilistic risk assessment, radiation damage, probabilistic fracture mechanics, and materials characterization studies. The work is still in progress, but it appears to be proceeding well. |
The PTS rule 10 CFR 50.61 was issued in 1983 as an "adequate protection" rule. A class of transient events had been identified in which a rapid cooldown of the reactor vessel coincides with high internal pressure. Such events produce high stresses and relatively low temperatures at the inner surface of the vessel. Because of the high stresses and the reduced fracture toughness near the inner surface, preexisting flaws can propagate through the wall and cause vessel failure. The probability of vessel failure is the product of the probability that a PTS initiating event will occur and the conditional probability of failure given an event. The staff established a screening criterion based on critical values for the reference temperature (RTNDT) that can be used to characterize the toughness of reactor vessels at temperatures associated with PTS events. The screening criterion was intended to ensure that the frequency of throughwall vessel failure was less than 5 X 10-6 events per reactor year. This throughwall vessel failure frequency value was also used as an acceptance value in Regulatory Guide 1.154, which provides the basis for a more detailed assessment of the probability of vessel failure if the PTS screening criterion is exceeded.
As part of the reevaluation of the PTS rule, the staff is reassessing the acceptance value for vessel failure because of the additional risk assessment experience that has been developed since the PTS rule was established in 1983. The staff described some of its considerations in determining an acceptance value in SECY-00-0140. This SECY paper discusses the choice of the acceptance value in terms of the core damage frequency and LERF acceptance guidelines used in Regulatory Guide 1.174. As we noted in our report on spent fuel pool accident risk dated April 13, 2000, the LERF acceptance guideline in Regulatory Guide 1.174 was based on the prompt fatalities associated with a steam-oxidation driven source term. In the cases of a spent fuel pool fire or a reactor pressure vessel failure due to PTS, core damage will occur in air rather than steam. The staff currently is considering the implications of air oxidation on the source term in its evaluation of spent fuel pool cooling. The staff also should consider the implications of an air-oxidation source term on the choice of an acceptance value for reactor pressure vessel failure frequency.
The staff's effort to update the Fracture Analysis of Vessels: Oak Ridge (FAVOR) probabilistic fracture mechanics code is nearing completion. The updated code is now capable of realistic descriptions of neutron fluence, material variability, flaw distributions, and fracture toughness. The staff has chosen not to update the code to include azimuthal variations in temperature based on the assumption that the thermal-hydraulic experiments at the APEX facility will confirm that the temperatures in the beltline region during a PTS event are nearly axisymmetric. Updating the FAVOR code is an important achievement of the PTS reevaluation project.
The staff reported on its continuing efforts to develop more realistic descriptions of flaw distributions in reactor pressure vessels, to characterize uncertainties in fracture toughness distributions, and to validate the thermal-hydraulic models for PTS events. This is work in progress and so it is premature for us to evaluate it. It appears, however, that the project integrates probabilistic risk assessment, probabilistic fracture mechanics, and thermal hydraulics well. We look forward to additional discussions with the staff on its efforts to provide a comprehensive estimate of the uncertainties in PTS calculations by integrating the treatment of uncertainties through a PRA-type analysis.
| ||Sincerely, |
|Dana A. Powers |
|1. ||U. S. Nuclear Regulatory Commission, Report Prepared with Oak Ridge National Laboratory, "An Updated Probabilistic Fracture Mechanics Methodology for Application to Pressurized Thermal Shock," issued in the Proceedings from the IAEA Specialists' Meeting, "Methodology and Supporting Research for the Pressurized Thermal Shock Evaluation," Rockville, MD, July 2000. |
|2. ||U. S. Nuclear Regulatory Commission, Office of Research, Draft, "Report on the Results of the Expert Judgment Process for the Generalized Flaw Size and Density Distribution for Domestic Reactor Pressure Vessels," received September 7, 2000. |
|3. ||E. D. Eason, and J. E. Wright, Modeling & Computing Services, Draft Report, "Updated Transition Temperature Shift Model," dated July 28, 2000. |
|4. ||U. S. Nuclear Regulatory Commission, report prepared by Oak Ridge National Laboratory, "Technical Basis for Statistical Models of Extended KIc and KIa Fracture Toughness Databases for RPV Steels," dated February 2000. |
|5. ||F. Li, et. al., University of Maryland, "KIc / KIa Uncertainty Characterization," dated June 23, 2000. |
|6. ||Report from Dana A. Powers, Chairman, ACRS, to Richard A. Meserve, Chairman, NRC, Subject: Draft Final Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants, dated April 13, 2000. |
|7. ||U.S. Nuclear Regulatory Commission, Regulatory Guide 1.154, "Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors," January 1987. |
|8. ||U. S. Nuclear Regulatory Commission, Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis," July 1998. |
|9. ||U.S. Nuclear Regulatory Commission, SECY-00-0140, memorandum from William D. Travers, Executive Director for Operations, NRC, for the Commissioners, Subject: Reevaluation of the Pressurized Thermal Shock Rule (10 CFR 50.61) Screening Criterion, dated June 23, 2000. |
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