Sk Iskander

Review of Pressurized-Water-Reactor-Related Thermal Shock Studies

Flaw behavior trends associated with pressurized thermal shock (PTS) loading of pressurized water reactor (PWR) pressure vessels have been under investigation at Oak Ridge National Laboratory (ORNL) for approximately twelve years. During that time, eight thermal shock experiments with thick-walled steel cylinders were conducted as a part of the investigations. These experiments demonstrated, in good agreement with linear elastic fracture mechanics (LEFM), crack initiation and arrest, a series of initiation/arrest events with deep penetration of the wall, long crack jumps without significant dynamic effects at arrest, arrest in a rising stress-intensity-factor (K I ) field, extensive surface extension of an initially short and shallow flaw, and warm prestressing with K I ≤ 0. This information was used in the development of a fracture mechanics model that is being used extensively in the evaluation of the PTS issue.

Effects of 50/degree/C surveillance and test reactor irradiations on ferritic pressure vessel steel embrittlement

The results of surveillance tests on the High-Flux Isotope Reactor (HFIR) pressure vessel at the Oak Ridge National Laboratory revealed that a greater than expected embrittlement had taken place after about 17.5 effective full-power years of operation and an operational assessment program was undertaken to fully evaluate the vessel condition and recommend conditions under which operation could be resumed. A research program was undertaken that included irradiating specimens in the Oak Ridge Research Reactor. Specimens of the A212 grade B vessel shell material were included, along with specimens from a nozzle qualification weld and a submerged-arc weld fabricated at ORNL to reproduce the vessel seam weld. The results of the surveillance program and the materials research program performed in support of the evaluation of the HFIR pressure vessel are presented and show the welds to be more radiation resistant than the A212B. Results of irradiated tensile and annealing experiments are described as well as a discussion of mechanisms which may be responsible for enhanced hardening at low damage rates. 20 refs., 22 figs., 5 tabs.

Heavy-section steel irradiation program summary☆

Abstract In light-water-cooled nuclear power plants, a steel reactor pressure vessel (RPV) forms the primary containment boundary for the nuclear fuel assembly. Since a failure of the RPV carries the potential of major contamination release and severe accident, it is imperative to safe reactor operation to understand and be able to accurately predict realistic failure models of the vessel material. A major known risk arises from irradiation-induced degradation of the RPV fracture resistance which occurs as a result of normally occurring neutron flux during the service life of the reactor. Better understanding of the irradiation degradation of the vessel material will provide better input to improve the engineering models vital for prediction of RPV integrity and safe service life. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water RPVs. The program includes the direct continuation of irradiation studies previously conducted within the Heavy-Section Steel Technology Program augmented by enhanced examinations of the accompanying microstructural changes. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties including fracture toughness ( K 1c and J 1c ), crack-arrest toughness ( K 1a ), ductile tearing resistance (d J /da), Charpy V-notch impact energy, dropweight nil-ductility temperature (NDT), and tensile properties. Models based on observations of radiation-induced microstructural changes using field ion and high-resolution transmission electron microscopy provide a firmer basis for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI Program are highcopper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs. In addition, a limited effort Will focus on stainless steel weld overlay cladding, typical of that used on the inner surface of RPVs, since its postirradiation fracture properties have the potential for strongly affecting the extension of small surface flaws during overcooling transients. Results from the HSSI studies will be integrated to aid in resolving major regulatory issues facing the U.S. Nuclear Regulatory Commission which involve RPV irradiation embrittlement such as pressurized-thermal shock, operating pressure-temperature limits, low-temperature overpressurization, and the specialized problems associated with low upper-shelf welds. Taken together the results of these studies also provide guidance and bases for evaluating both the aging behavior and the potential for plant life extension of light-water RPVs.

Effects of Annealing Time on the Recovery of Charpy V-Notch Properties of Irradiated High-Ccopper Weld Metal

One of the options to mitigate the effects of irradiation on reactor pressure vessels is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. An important issue to be resolved is the effect on the toughness properties of reirradiating a vessel that has been annealed. This paper describes the annealing response of irradiated high-copper submerged-arc weld HSSI 73W. For this study, the weld has been annealed at 454 C (850 F) for lengths of time varying between 1 and 14 days. The Charpy V-notch 41-J (30-ft-lb) transition temperature (TT{sub 41J}) almost fully recovered for the longest period studied, but recovered to a lesser degree for the shorter periods. No significant recovery of the TT{sub 41J} was observed for a 7-day anneal at 343 C (650 F). At 454 C for the durations studied, the values of the upper-shelf impact energy of irradiated and annealed weld metal exceeded the values in the unirradiated condition. Similar behavior was observed after aging the unirradiated weld metal at 460 and 490 C for 1 week.