TY - GEN
T1 - Adaptation of crack growth detection techniques to u.s. Material test reactors
AU - Palmer, Joseph
AU - Teysseyre, Sebastien P.
AU - Davis, Kurt L.
AU - Kohse, Gordon
AU - Ostrovsky, Yakov
AU - Carpenter, David M.
AU - Rempe, Joy L.
PY - 2016
Y1 - 2016
N2 - Idaho National Laboratory and the Massachusetts Institute of Technology Nuclear Reactor Laboratory researchers have recently completed a project to develop a test rig capable of evaluating the potential for irradiation- Assisted stress corrosion cracking in the high-flux conditions possible in U.S. material test reactors. Adapting techniques pioneered at the Halden Boiling Water Reactor, the project included designing and testing the loader mechanism, testing individual test rig components and electronics support equipment, and autoclave testing of the rig design. Technical challenges involved developing robust connections to the specimen for the applied current and voltage measurements, selecting appropriate ceramic insulating materials that can endure light-water reactor environments, coping with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, primarily associated with fast-neutron damage that changes specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the Massachusetts Institute of Technology Research Reactor. The test rig and associated support equipment were used to apply loads to a representative compact tensile specimen and measure crack growth using the direct current potential drop method. Although the test period was limited to approximately 70 days and the accumulated neutron dose was relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8,000 times (more than typical for a long-term irradiation- Assisted stress corrosion cracking test), which was sufficient to propagate a crack of over 2 mm. Post-irradiation measurements of the specimen confirmed a crack length equivalent to that inferred by the direct current potential drop signals. Measurements also demonstrated the necessary infrastructure for handling the irradiated sample rig and satisfactory performance of a low-cost system for crack measurement in the hot cell.
AB - Idaho National Laboratory and the Massachusetts Institute of Technology Nuclear Reactor Laboratory researchers have recently completed a project to develop a test rig capable of evaluating the potential for irradiation- Assisted stress corrosion cracking in the high-flux conditions possible in U.S. material test reactors. Adapting techniques pioneered at the Halden Boiling Water Reactor, the project included designing and testing the loader mechanism, testing individual test rig components and electronics support equipment, and autoclave testing of the rig design. Technical challenges involved developing robust connections to the specimen for the applied current and voltage measurements, selecting appropriate ceramic insulating materials that can endure light-water reactor environments, coping with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, primarily associated with fast-neutron damage that changes specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the Massachusetts Institute of Technology Research Reactor. The test rig and associated support equipment were used to apply loads to a representative compact tensile specimen and measure crack growth using the direct current potential drop method. Although the test period was limited to approximately 70 days and the accumulated neutron dose was relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8,000 times (more than typical for a long-term irradiation- Assisted stress corrosion cracking test), which was sufficient to propagate a crack of over 2 mm. Post-irradiation measurements of the specimen confirmed a crack length equivalent to that inferred by the direct current potential drop signals. Measurements also demonstrated the necessary infrastructure for handling the irradiated sample rig and satisfactory performance of a low-cost system for crack measurement in the hot cell.
UR - https://www.scopus.com/pages/publications/84986277929
M3 - Conference contribution
AN - SCOPUS:84986277929
T3 - International Congress on Advances in Nuclear Power Plants, ICAPP 2016
SP - 930
EP - 938
BT - International Congress on Advances in Nuclear Power Plants, ICAPP 2016
PB - American Nuclear Society
T2 - 2016 International Congress on Advances in Nuclear Power Plants, ICAPP 2016
Y2 - 17 April 2016 through 20 April 2016
ER -