Thermal-aged surveillance materials consisting of SA-533, Grade B, Class 1 plate material; SA-508, Class 2 forging material; and 2 Mn-Mo-Ni/Linde 80 weld metals were removed from two commercial reactor pressure vessels. The material from the first reactor vessel received a thermal exposure of approximately 103,000 hours at 282°C (540°F), while the material from the second reactor vessel received a thermal exposure of approximately 93,000 hours at 282°C (540°F). Tensile and 1/2T compact fracture toughness specimens were fabricated from these materials and tested. In addition, to examine the effects of annealing, selected thermal-aged and unaged specimens were annealed at 454°C (850°F) and tested. Varying responses in the fracture toughness properties were observed for all materials after exposure to the thermal-aging temperature. The base metal plate had an observed decrease in J-values after its respective aging exposure, while no significant difference in the J-values were observed for the Linde 80 weld metals. No significant difference was seen in the J-data for the aged/annealed materials, but because of the small number of test specimens available, no conclusion could be determined for the response to annealing.
Thermal-aged surveillance material was removed from two commercial reactor pressure vessels. The material from the first reactor vessel received a thermal exposure of approximately 103,000 hours at 282°C (540°F), while the material from the second reactor vessel received a thermal exposure of approximately 93,000 hours at 282°C (540°F). The surveillance material removed consisted of SA-508, Class 2 and SA-533, Grade B, Class 1, base metals; Mn-Mo-Ni/Linde 80 flux weld metals; and SA-533, Grade B, Class 1, correlation monitor material. Charpy V-notch impact specimens were fabricated from these materials and tested. In addition, selected thermal-aged specimens were annealed at 343°C (650°F) and 454°C (850°F) and impact tested. Unaged material was tested to establish baseline data for evaluating the thermal-aged material and thermal-aged/annealed material. Small variations in the impact properties were observed for all materials. Small increases in transition temperature were observed for the forging base metal and weld metal B surveillance materials, while the upper-shelf energies exhibited small decreases. The 454°C (850°F) anneal restored the upper-shelf energy for the forging base metal and weld metal B surveillance materials.
Rupture characteristics of internally heated Zircaloy-4 nuclear fuel cladding were investigated to develop rupture data that would provide a basis to bench-mark mathematical modeling of clad rupture. Single- and five-tube configurations of pressurized water reactor cladding were ruptured at constant pressures and heating rates with simulated reactor boundary conditions. The results showed that the α + β material property transformation region had a strong effect on the material ductility, and the clad rupture temperature varied linearly with hoop stress in the α + β region. Axial distribution of multitube ruptures was not coplanar. The rupture positions on all specimens were distributed normally about the point of maximum temperature. This paper emphasizes test techniques as well as the rupture characteristics of Zircaloy-4 clad.
The effects of temperature during irradiation on the tensile and impact properties of two low-alloy steels were studied. Tension and impact specimens of 1.25Cr-0.5Mo and 2.25Cr-1Mo steels were irradiated in a range of 2.8 x 1018 to 2.0 x 1020 nvt (greater than 1 Mev) and over a temperature range from room temperature to 1700 F. Results from postirradiation examinations conducted at room temperature indicate that the radiation effects on tensile properties are reduced at irradition temperatures above 725 F and at irradiation temperatures between 900 and 1100 F there is no apparent irradiation damage. For specimens irradiated above approximately 1100 F the tensile and yield strength increased again and strength levels equal to the unirradiated values were not observed until the steel was heated above the A3 point (1630 F). The second increase in tensile and yield strength may be associated with microstructural changes. Results from subsize impact specimens showed that irradiated data equivalent to the unirradiated impact values were not recovered until temperatures in the stress relief range were reached. At the higher temperatures the shift in impact properties appears to be more sensitive to temperature than to radiation dose.
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