Margin to Onset of Nucleate boiling and Flow Instability Studies for Preliminary MITR Design-Demonstration Element Thermal-Hydraulics
This study covers the onset of flow instability (OFI) preliminary results obtained from leveraging correlations, in addition to the preliminary thermal hydraulics results such as pressure, flow velocity, temperature, and oxide layer over the design demonstration experiment (DDE) for the Massachusetts Institute of Technology Reactor (MITR). Current computational fluid dynamics (CFD) models in fluid structure […]
Accelerated irradiation testing and post-irradiation characterization: U.S.-based capabilities for advanced nuclear systems and radioisotope production
Irradiation experiments and post-irradiation examinations (PIE)—together referred to as “Irradiation Testing” IRT—are prerequisites for nuclear fuel and material qualification for the deployment of new and advanced reactors, as well as radioisotope production, thereby ensuring regulatory compliance. Qualified research and test reactors (RTRs) and testing facilities are essential to enable IRT to verify performance and safety […]
Fluid structural interaction simulation of the MP-1 plate performance irradiated in the advanced test reactor
Due to the high neutron fluxes they generate and increased heat transfer performance, plate type fuels are used in the U.S. high-performance research reactors. During irradiation, a significant amount of fission energy (i.e., ~200 MeV per atom) is released by the U-235 chain reaction that is carried away by the coolant. Through thermal hydraulic analysis, […]
Thermal-Hydraulic Analysis of the AFIP-7 Irradiation Test in the Advanced Test Reactor—Oxide Growth Prediction and Correlation
Knowing the thickness of the oxide layer on the surface of aluminum fuel cladding is vitally important for predicting fuel temperature due to the low thermal conductivity of the oxide layer. Several correlation models for predicting oxide growth can be found in the literature. In previous research, the correlations were combined with heat transfer simulations […]
Thermal Hydraulic Analysis of the AFIP-7 Irradiation Test in the Advanced Test Reactor – Model Correlation and Performance Evaluation
As a system level code, RELAP5 is widely used in the nuclear field for the reactor hydraulic analysis. In the component/experimental level, it is frequently employed as well. This paper demonstrates that the RELAP5 flow simulation of the AFIP-7 irradiation experiment which has a complicated 3D geometry/flow path deviates by ~28% from the flow evaluation […]