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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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DOE OSTI2011

AFIP-6 Characterization Summary Report

The AFIP-6 (ATR Full-size-plate In center flux trap Position) Characterization Summary Report outlines the fresh fuel characterization efforts performed during the AFIP-6 experiment. The AFIP-6 experiment was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of Advanced Test Reactor (ATR) fuel plates (45-inches long). The AFIP-6 test was the first test with plates that were swaged into the rails of the assembly. This test served to examine the effects of a plate in a swaged condition with longer fuel zones (22.5-inches long), that were centered in the plate. AFIP-6 test plates employed a zirconium interlayer that was co-rolled with the fuel foil. Previous mini-plate and AFIP irradiation experiments performed in ATR have demonstrated the stable behavior of the interface between the U-Mo fuel and the zirconium interlayer.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

AFIP-6 Mark II Irradiation Summary Report

Due to the incompletion of the A dvanced Test Reactor (ATR) F ull size plate I n center flux trap P osition (AFIP) experiment AFIP-6, a second irradiation (named AFIP-6 Mark II) of identical plates was performed. AFIP-6 Mark II was designed to evaluate high-power large-scale performance of monolithic uranium-molybdenum (U-Mo) fuels. The following report summarizes the life of the AFIP-6 MKII 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↗
DOE OSTI2012

AFIP-6 MKII First Cycle Report

The first fuel plate frame assembly of the AFIP-6 MKII experiment was irradiated as planned from December, 2011 through February, 2012 in the center flux trap of the Advanced Test Reactor during cycle 151A. Following irradiation in this cycle and while reconfiguring the experiment in the ATR canal, a non-fueled component (the bottom plate) of the first fuel plate frame assembly became separated from the rail sides. There is no evidence that the fueled region of the fuel plate frame assembly was compromised by this incident or the irradiation conditions. The separation of this component was determined to have been caused by flow induced vibrations, where vortex shedding frequencies were resonant with a natural frequency of the bottom plate component. This gave way to amplification, fracture, and separation from the assembly. Although parallel flow induced vibrations were analyzed, vortex shedding flow induced vibrations was an unfamiliar failure mode that was difficult to identify. Both the once-irradiated first fuel plate and un-irradiated second fuel plate frame assemblies were planned for irradiation in the subsequent cycle 151B. The AFIP-6 MKII experiment was excluded from irradiation in cycle 151B because non-trivial design modifications would be needed to mitigate this type of incident during the second irradiation cycle. All items of the experiment hardware were accounted for and cycle 151B occurred with a non-fueled AFIP backup assembly in the center flux trap. Options for completion of the AFIP-6 MKII experiment campaign are presented and future preventative actions are recommended.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2012

Recent Accomplishments in the Irradiation Testing of Engineering-Scale Monolithic Fuel Specimens

The US fuel development team is focused on qualification and demonstration of the uranium-molybdenum monolithic fuel including irradiation testing of engineering-scale specimens. The team has recently accomplished the successful irradiation of the first monolithic multi-plate fuel element assembly within the AFIP-7 campaign. The AFIP-6 MKII campaign, while somewhat truncated by hardware challenges, exhibited successful irradiation of a large-scale monolithic specimen under extreme irradiation conditions. The channel gap and ultrasonic data are presented for AFIP-7 and AFIP-6 MKII, respectively. Finally, design concepts are summarized for future irradiations such as the base fuel demonstration and design demonstration experiment campaigns.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

A Statistical Approach to Safety Margin Evaluation of Irradiation Experiments in the ATR

The current safety basis of the INL Advanced Test Reactor requires that irradiation experiments meet minimum safety margins to critical heat flux and flow instability during anticipated reactor fault conditions. Compliance to these requirements may be demonstrated by calculating safety margins using safety factors that account for uncertainties, or using a statistical approach to quantify the uncertainty in the safety margins. Safety margins obtained from the first approach are conservative since the individual events leading to the cumulative safety factor are assumed to occur simultaneously at their maximum uncertainty. The statistical approach is less conservative in the case where the uncertainties are not too large, and may reduce the conservatism resulting from an overabundance of safety factors. This report describes the procedure for safety margin evaluation using a statistical approach and identifies the important thermal-hydraulic variables to consider in an uncertainty analysis. In particular, safety margins to critical heat flux and flow instability are evaluated for three recent high power fueled experiments (MP-1, FSP-1 and AFIP-6) using RELAP5 coupled to RAVEN to perform statistical analysis and uncertainty quantification. The results demonstrate adequate safety margins to critical heat flux and flow instability during an anticipated loss of commercial power to the reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Four-point Bend Testing of Irradiated Monolithic U-10Mo Fuel

This paper presents results of recently completed studies aimed at characterizing the mechanical properties of irradiated U-10Mo fuel in support of monolithic base fuel qualification. Mechanical properties were evaluated in four-point bending. Specimens were taken from fuel plates irradiated in the RERTR-12 and AFIP-6 Mk. II irradiation campaigns, and tests were conducted in the Hot Fuel Examination Facility (HFEF) at Idaho National Laboratory (INL). The monolithic fuel plates consist of a U-10Mo fuel meat covered with a Zr diffusion barrier layer fabricated by co-rolling, clad in 6061 Al using a hot isostatic press (HIP) bonding process. Specimens exhibited nominal (fresh) fuel meat thickness ranging from 0.25 mm to 0.64 mm, and fuel plate average burnup ranged from approximately 0.4 x 1021 fissions/cm 3 to 6.0 x 1021 fissions/cm 3 . After sectioning the fuel plates, the 6061 Al cladding was removed by dissolution in concentrated NaOH. Pre- and post-dissolution dimensional inspections were conducted on test specimens to facilitate accurate analysis of bend test results. Four-point bend testing was conducted on the HFEF Remote Load Frame at a crosshead speed of 0.1 mm/min using custom-designed test fixtures and calibrated load cells. All specimens exhibited substantially linear elastic behavior and failed in a brittle manner. The influence of burnup on the observed slope of the stress-strain curve and the calculated fracture strength is discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗