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O'Brien, Robert C

Publications and source records attributed to O'Brien, Robert C.

5 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Preliminary Analysis of a Nuclear Grade Sandwich Composite for Transportable Micro-Reactor Shielding

Transportable microreactors pose a unique shielding challenge compared with conventional reactors. Many vendors are seeking to transport a microreactor to a site, operate it for a number of years, and then transport it to a secondary site. This involves ensuring the reactor has sufficient shielding during operations and post-irradiation transportation to ensure the safety of those around it, all while ensuring it can be transported via traditional means. This work examines the use of a nuclear grade sandwich composite (NGSC) which seeks to combine the reactor pressure vessel and biological shielding into a single unit. Through this work, it was determined that the NGSC was sufficient to adequately reduce neutron and gamma current exiting the structure to warrant further investigation as a potential shielding option for microreactors.

42 ENGINEERING↗
DOE OSTI2022

Sirius Irradiation Experiments and Post-Irradiation Examinations for Nuclear Thermal Propulsion

Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high temperature gas-cooled fission reactor to provide thermal energy to heat the coolant which also serves as the propellant. The fuel systems of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, has been irradiated in Idaho National Laboratory’s (INL) Transient Reactor Test Facility (TREAT) enabling evaluation of these materials under representative nuclear heating rates (~95 K/s) and peak temperatures (~2527 K). These tests named Sirius-1 (UN-W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced cracks in the fuel specimen and spalling of surface material. Uranium soot was found on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and edge features producing a laminar microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Options for Subscale Maturation of Advanced Reactor Technologies Testing for Nuclear Thermal Propulsion

Several options could be implemented to establish an irradiation testing capability suitable for investigation of the performance of multiple nuclear thermal propulsion fuel elements at prototypic conditions. The prototypic conditions of interest are based on the current needs of the National Aeronautics and Space Administration’s Space Nuclear Power Program. The results of such testing are also intended to reduce the risks currently seen for any future subscale or full-scale ground testing of an engine-reactor system. The optimal solution is dependent upon several factors such as performance, cost, availability, schedule, technology readiness level, and plans for future testing in the SNP Program. Three options, based on different combinations of these factors, are considered in this report.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2019

Irradiation damage and IASCC of printed 316L for use as fuel cladding

Additive manufacturing is an area of increasing interest to the nuclear industry. It offers a number of benefits over traditional manufacturing methods, such as the ability to create complex geometries without the need for welds. In addition to the potential for creating nuclear structural components, additive manufacturing may be used to create novel fuel/cladding designs. However, printed steel behaves differently than traditional wrought material. A number of industries are already studying the effects of the printed microstructures on mechanical properties. For cladding material and some structural nuclear components, however, the nuclear industry must also understand the material response to irradiation. This work examines the irradiation damage and the irradiation assisted stress corrosion cracking behavior of both proton irradiated (1 and 5 dpa) printed 316L steel and wrought 316L steel that were irradiated side by side at the Michigan Ion Beam Laboratory. As orientation with respect to the build direction can affect the mechanical properties of printed metal, specimens were taken from multiple orientations within the printed steel. For the IASCC testing, specimens were strained to 4% in simulated BWR NWC water (288°C, 0.2 µS/cm). TEM, SEM and X-ray microCT were used to examine coupons of the material as well as tensile bars used in the IASCC testing. This paper will show the key findings from this work, namely the IASCC response of printed metal (as compared to wrought), as well as the effects of the printed microstructure (such as the pre-existing voids) on the irradiation damage.

36 MATERIALS SCIENCE↗