Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor

• DOI: 10.13182/nt14-60
• Publication Date: 2015-05-01T15:23:59Z
• Authors: K. L. Davis; D. L. Knudson; J. L. Rempe; J. C. Crepeau; S. Solstad
• Keywords: Cladding (metalworking); Pressurized water reactor

New materials are being considered for fuel, cladding, and structures in next-generation existing nuclear reactors. Such can undergo significant dimensional physical changes during high-temperature irradiation. To accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions model development validation. provide data, programs such as the Advanced Test Reactor (ATR) National Scientific Users Facility (NSUF) have funded researchers at Idaho Laboratory (INL) High Temperature (HTTL) to develop several instrumented test rigs obtain real time from specimens irradiated well-controlled pressurized water reactor (PWR) coolant ATR. This technical note reports status of INL efforts evaluate prototype that rely on linear variable differential transformers (LVDTs) laboratory settings. Although similar LVDT-based been deployed lower-flux testing reactors (MTRs), this effort is unique because it relies robust LVDTs withstand higher temperatures fluxes than often found other MTR irradiations. Specifically, designed detecting length diameter ATR PWR loops. Once implemented, will users with capabilities sorely needed measurements, elongation caused by thermal expansion and/or creep loading, associated fuel cladding swelling, pellet-cladding interaction, crud buildup.