Engineering topics

Perez, D M

Publications and source records attributed to Perez, D M.

4 records

DOE OSTI2012

TEMPERATURE DEPENDANT BEHAVIOUR OBSERVED IN THE AFIP-6 IRRADIATION TEST

The AFIP-6 test assembly was irradiated for one cycle in the Advanced Test Reactor at Idaho National Laboratory. The experiment was designed to test two monolithic fuel plates at power and burn-ups which bounded the operating conditions of both ATR and HFIR driver fuel. Both plates contained a solid U-Mo fuel foil with a zirconium diffusion barrier between 6061-aluminum cladding plates bonded by hot isostatic pressing. The experiment was designed with an orifice to restrict the coolant flow in order to obtain prototypic coolant temperature conditions. While these coolant temperatures were obtained, the reduced flow resulted in a sufficiently low heat transfer coefficient that failure of the fuel plates occurred. The increased fuel temperature led to significant variations in the fission gas retention behaviour of the U-Mo fuel. These variations in performance are outlined herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2011

RERTR-9 Summary Report

The RERTR-9 experiment was designed to test the effect of modified fuel/clad interfaces in monolithic fuel plates and to demonstrate that the addition of Si to the matrix material in dispersion plates continued to be effective at high loading (~8.5 g U/cc). Several monolithic fuel plates were fabricated by Hot Isostatic Pressing (HIP) and Friction Bonding (FB) with thin layers of Si inserted and by HIP with a Zr diffusion barrier between the fuel and cladding. Si was applied to the interface by thermal spray of Al Si mixtures and by the insertion of thin Si-rich Al alloy foil between the fuel/clad interface. The dispersion fuel plates were fabricated by semi-standard rolling techniques (the reduction by rolling was lowered to limit fabrication defects). Matrix materials consisted of Al-Si alloys and mixtures with various levels of Si. The following report summarizes the life of the RERTR-9A/B experiment through end of irradiation, including as-run neutronic analysis, thermal analysis and hydraulic testing results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2011

RERTR-10 Irradiation Summary Report

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-10 was designed to further test the effectiveness of modified fuel/clad interfaces in monolithic fuel plates. The experiment was conducted in two campaigns: RERTR-10A and RERTR-10B. The fuel plates tested in RERTR-10A were all fabricated by Hot Isostatic Pressing (HIP) and were designed to evaluate the effect of various Si levels in the interlayer and the thickness of the Zr interlayer (0.001”) using 0.010” and 0.020” nominal foil thicknesses. The fuel plates in RERTR-10B were fabricated by Friction Bonding (FB) with two different thickness Si layers and Nb and Zr diffusion barriers.1 The following report summarizes the life of the RERTR-10A/B experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2011

AFIP-1 Irradiation Summary Report

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-1 was designed to demonstrate the performance of second-generation dispersion fuels at a prototypic scale with a length of 21.5 inches (54.6 cm), width of 2.25 inches (5.75 cm) and a thickness of 0.050 inch (0.13 cm). The experiment was fabricated using commercially standard practices at BWX Technology, Inc. (BWXT). The U-7Mo fuel particles were supplied by the Korean Atomic Energy Research Institute (KAERI) using equipment intended for commercial supply. Two fuel plates were tested that incorporated two different matrix compositions, Al-2Si and Al-4043.1 The following report summarizes the life of the AFIP-1 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗