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Manufacture of Fuel Plates for the Experimental Boiling Water Reactor

Description: From Abstract: "Discussed in the paper are the production of the uranium fuel alloy, the fabrication of the cladding components from Zircaloy-2 ingot, the assembly, welding, evacuation, and sealing of the cladding billets, the jacketing of the cladding billets with steel, and the roll bonding, heat treatment, stripping, and cleaning of the fuel plates. Evaluation of the fuel plates produced is also included."
Date: June 1957
Creator: Macherey, R. E.; Bean, C. H.; Carson, N. J., Jr. & Lindgren, J. R.
Partner: UNT Libraries Government Documents Department

Some Tooling for Manufacturing Research Reactor Fuel Plates

Description: This paper will discuss some of the tooling necessary to manufacture aluminum-based research reactor fuel plates. Most of this tooling is intended for use in a high-production facility. Some of the tools shown have manufactured more than 150,000 pieces. The only maintenance has been sharpening. With careful design, tools can be made to accommodate the manufacture of several different fuel elements, thus, reducing tooling costs and maintaining tools that the operators are trained to use. An important feature is to design the tools using materials with good lasting quality. Good tools can increase return on investment.
Date: October 3, 1999
Creator: Knight, R.W.
Partner: UNT Libraries Government Documents Department


Description: This article presents numerical simulation of dispersion fuel mini plates via fluid–thermal–structural interaction performed by commercial finite element solver COMSOL Multiphysics to identify initial mechanical response under actual operating conditions. Since fuel particles are dispersed in Aluminum matrix, and temperatures during the fabrication process reach to the melting temperature of the Aluminum matrix, stress/strain characteristics of the domain cannot be reproduced by using simplified models and assumptions. Therefore, fabrication induced stresses were considered and simulated via image based modeling techniques with the consideration of the high temperature material data. In order to identify the residuals over the U7Mo particles and the Aluminum matrix, a representative SEM image was employed to construct a microstructure based thermo-elasto-plastic FE model. Once residuals and plastic strains were identified in micro-scale, solution was used as initial condition for subsequent multiphysics simulations at the continuum level. Furthermore, since solid, thermal and fluid properties are temperature dependent and temperature field is a function of the velocity field of the coolant, coupled multiphysics simulations were considered. First, velocity and pressure fields of the coolant were computed via fluidstructural interaction. Computed solution for velocity fields were used to identify the temperature distribution on the coolant and on the fuel plate via fluid-thermal interaction. Finally, temperature fields and residual stresses were used to obtain the stress field of the plates via fluid-thermal-structural interaction.
Date: November 1, 2010
Creator: Ozaltun, Hakan; Shen, Herman & Madvedev, Pavel
Partner: UNT Libraries Government Documents Department

FY04 Inspection Results for Wet Uruguay Fuel in L-Basin

Description: The 2004 visual inspection of four Uruguay nuclear fuel assemblies stored in L-Basin was completed. This was the third inspection of this wet stored fuel since its arrival in the summer of 1998. Visual inspection photographs of the fuel from the previous and the recent inspections were compared and no evidence of significant corrosion was found on the individual fuel plate photographs. Fuel plates that showed areas of pitting in the cladding during the original receipt inspection were also identified during the 2004 inspection. However, a few pits were found on the non-fuel aluminum clamping plates that were not visible during the original and 2001 inspections.
Date: September 1, 2005
Partner: UNT Libraries Government Documents Department

Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

Description: This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at ORNL. Damage propagation is postulated to occur from thermal conduction between dmaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur beause of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A parametric study was done for several uncertain variables. The study included investigating effects of plate contact area, convective heat transfer coefficient, thermal conductivity on fuel swelling, and initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects of damage propagation. Results provide useful insights into how variouss uncertain parameters affect damage propagation.
Date: December 31, 1995
Creator: Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S. & Georgevich, V.
Partner: UNT Libraries Government Documents Department

Modeling and analysis of thermal-hydraulic response of uranium- aluminum reactor fuel plates under transient heatup conditions

Description: A 3-D model to predict the thermal behavior of ANS (Advanced Neutron Source) fuel miniplates has been developed. Possibility of explosive boiling was considered, and it was concluded that the heating rates (existant in NSRR tests) are not large enough for this to occur. However, transient boiling effects were pronounced. Because of the complexity of transient pool boiling and the unavailability of experimental data for the situations studied, an approximation was made that predicted the data very well within the uncertainties present. If pool boiling from the miniplates had been assumed to be steady during the heating pulse, the experimental data would have been greatly overestimated. This shows the importance of considering the transient nature of heat transfer in analysis of reactivity excursion accidents. An additional contribution of this work is that it provided data on highly subcooled steady nucleate boiling from the cooling portion of the thermocouple traces.
Date: December 31, 1995
Creator: Navarro-Valenti, S.; Kim, S.H.; Georgevich, V.; Taleyarkhan, R.P.; Fuketa, T.; Soyama, Kk. et al.
Partner: UNT Libraries Government Documents Department


Description: This paper discusses how candidate fuel plates for RERTR Fuel Development experiments are examined and tested for acceptance prior to reactor insertion. These tests include destructive and nondestructive examinations (DE and NDE). The DE includes blister annealing for dispersion fuel plates, bend testing of adjacent cladding, and microscopic examination of archive fuel plates. The NDE includes Ultrasonic (UT) scanning and radiography. UT tests include an ultrasonic scan for areas of “debonds” and a high frequency ultrasonic scan to determine the "minimum cladding" over the fuel. Radiography inspections include identifying fuel outside of the maximum fuel zone and measurements and calculations for fuel density. Details of each test are provided and acceptance criteria are defined. These tests help to provide a high level of confidence the fuel plate will perform in the reactor without a breach in the cladding.
Date: October 1, 2008
Creator: Wight, J.M.; Moore, G.A. & Taylor, S.C.
Partner: UNT Libraries Government Documents Department

Improved performance of U-Mo dispersion fuel by Si addition in Al matrix.

Description: The purpose of this report is to collect in one publication and fit together work fragments presented in many conferences in the multi-year time span starting 2002 to the present dealing with the problem of large pore formation in U-Mo/Al dispersion fuel plates first observed in 2002. Hence, this report summarizes the excerpts from papers and reports on how we interpreted the relevant results from out-of-pile and in-pile tests and how this problem was dealt with. This report also provides a refined view to explain in detail and in a quantitative manner the underlying mechanism of the role of silicon in improving the irradiation performance of U-Mo/Al.
Date: June 1, 2011
Creator: Kim, Y. S. & Hofman, G. L.
Partner: UNT Libraries Government Documents Department

AFIP-6 Fabrication Summary Report

Description: The AFIP-6 (ATR Full-size plate In center flux trap Position) experiment was designed to evaluate the performance of monolithic fuels at a scale prototypic of research reactor fuel plates. Two qualified fueled plates were fabricated for the AFIP-6 experiment; to be irradiated in the INL Advanced Test Reactor (ATR). This report provides details of the fuel fabrication efforts, including material selection, fabrication processes, and fuel plate qualification.
Date: September 1, 2011
Creator: Moore, Glenn A. & Marshall, M. Craig
Partner: UNT Libraries Government Documents Department


Description: This article analyzes dimensional changes due to irradiation of monolithic plate-type nuclear fuel and compares results with finite element analysis of the plates during fabrication and irradiation. Monolithic fuel plates tested in the Advanced Test Reactor (ATR) at Idaho National Lab (INL) are being used to benchmark proposed fuel performance for several high power research reactors. Post-irradiation metallographic images of plates sectioned at the midpoint were analyzed to determine dimensional changes of the fuel and the cladding response. A constitutive model of the fabrication process and irradiation behavior of the tested plates was developed using the general purpose commercial finite element analysis package, Abaqus. Using calculated burn-up profiles of irradiated plates to model the power distribution and including irradiation behaviors such as swelling and irradiation enhanced creep, model simulations allow analysis of plate parameters that are either impossible or infeasible in an experimental setting. The development and progression of fabrication induced stress concentrations at the plate edges was of primary interest, as these locations have a unique stress profile during irradiation. Additionally, comparison between 2D and 3D models was performed to optimize analysis methodology. In particular, the ability of 2D and 3D models account for out of plane stresses which result in 3-dimensional creep behavior that is a product of these components. Results show that assumptions made in 2D models for the out-of-plane stresses and strains cannot capture the 3-dimensional physics accurately and thus 2D approximations are not computationally accurate. Stress-strain fields are dependent on plate geometry and irradiation conditions, thus, if stress based criteria is used to predict plate behavior (as opposed to material impurities, fine micro-structural defects, or sharp power gradients), unique 3D finite element formulation for each plate is required.
Date: November 1, 2012
Creator: Miller, Samuel J. & Ozaltun, Hakan
Partner: UNT Libraries Government Documents Department

AFIP-2 Fabrication Summary Report

Description: The Advanced Test Reactor (ATR) Full-size Plate In Center Flux Trap Position (AFIP)-2 experiment was designed to evaluate the performance of monolithic fuels at a scale prototypic of research reactor fuel plates. Two qualified fueled plates were fabricated for the AFIP 2 experiment to be irradiated in the Idaho National Laboratory ATR. This report provides details of the fuel fabrication efforts, including material selection, fabrication processes, and fuel plate qualification.
Date: February 1, 2010
Creator: Moore, Glenn
Partner: UNT Libraries Government Documents Department

SEM and TEM Characterization of As-Fabricated U-7Mo Disperson Fuel Plates

Description: The starting microstructure of a dispersion fuel plate can have a dramatic impact on the overall performance of the plate during irradiation. To improve the understanding of the as-fabricated microstructures of dispersion fuel plates, SEM and TEM analysis have been performed on RERTR-9A archive fuel plates, which went through an additional hot isostatic procsssing (HIP) step during fabrication. The fuel plates had depleted U-7Mo fuel particles dispersed in either Al-2Si or 4043 Al alloy matrix. For the characterized samples, it was observed that a large fraction of the ?-phase U-7Mo alloy particles had decomposed during fabrication, and in areas near the fuel/matrix interface where the transformation products were present significant fuel/matrix interaction had occurred. Relatively thin Si-rich interaction layers were also observed around the U-7Mo particles. In the thick interaction layers, (U)(Al,Si)3 and U6Mo4Al43 were identified, and in the thin interaction layers U(Al,Si)3, U3Si3Al2, U3Si5, and USi1.88-type phases were observed. The U3Si3Al2 phase contained some Mo. Based on the results of this work, exposure of dispersion fuel plates to relatively high temperatures during fabrication impacts the overall microstructure, particularly the nature of the interaction layers around the fuel particles. The time and temperature of fabrication should be carefully controlled in order to produce the most uniform Si-rich layers around the U-7Mo particles.
Date: November 1, 2009
Creator: D. D. Keiser, Jr.; Yao, B.; Perez, E. & Sohn, Y. H.
Partner: UNT Libraries Government Documents Department

SEM Characterization of an Irradiated Dispersion Fuel Plate with U-10Mo Particles and 6061 Al Matrix

Description: It has been observed that during irradiation of a dispersion fuel plate, fuel/matrix interactions can impact the overall fuel plate performance. To continue the investigation of the irradiation performance of Si-rich fuel/matrix interaction layers, RERTR-6 fuel plate V1R010 (U- 10Mo/6061 Al) was characterized using scanning electron microscopy. This fuel plate was of particular interest because of its similarities to fuel plate R1R010, which had U-7Mo particles dispersed in 6061 Al. Both fuel plates were irradiated as part of the RERTR-6 experiment and saw very similar irradiation conditions. R1R010 was characterized in another study and was observed to form relatively uniform Si-rich layers during fabrication that remained stable during irradiation. Since U-10Mo does not interact as much with 6061 Al at high temperatures to form layers, it was of interest to characterize a fuel plate with these particles since it would allow for a comparison of fuel plates with different amounts of preirradiation interaction zone formation, which were then exposed to similar irradiation conditions. This paper demonstrates how the lower amount of interaction layer development in V1R010 during fabrication appears to impact the overall performance of the fuel plate, such that it does not behave as well as R1R010 in terms of interaction layer stability. Additionally, the results of this study are applied to improve the understanding of fuel/cladding interactions in monolithic fuel plates that consist of U-10Mo foils encased in 6061 Al cladding.
Date: November 1, 2009
Creator: Keiser, D. D.; Jue, J. F.; Robinson, A. B.; Medvedev, P. G. & Finlay, M. R.
Partner: UNT Libraries Government Documents Department

AFIP-4 Fabrication Summary Report

Description: The AFIP-4 (ATR Full –size-plate In center flux trap Position) experiment was designed to evaluate the performance of monolithic fuels at a scale prototypic of research reactor fuel plates. Twelve qualified fueled plates were fabricated for the AFIP-4 experiment; to be irradiated in the INL Advanced Test Reactor (ATR). This report provides details of the fuel fabrication efforts; including material selection, fabrication processes, and fuel plate qualification.
Date: February 1, 2010
Creator: Moore, Glenn A.
Partner: UNT Libraries Government Documents Department

The potential pyrophoricity of BMI-SPEC and aluminum plate spent fuels retrieved from underwater storage

Description: Physical/chemical factors in U metal and hydride combustion, particularly pyrophoricity in ambient environment, were evaluated for BMI-SPEC and UAl{sub x} plate fuels. Some metal fuels may be highly reactive (spontaneously igniting in air) due to high specific surface area, high decay heat, or a high U hydride content from corrosion during underwater storage. However, for the BMI-SPEC and the aluminum plate fuels, this reactivity is too low to present a realistic threat of uncontrolled spontaneous combustion at ambient conditions. While residual U hydride is expected in these corroded fuels, the hydride levels are expected to be too low and the configuration too unfavorable to ignite the fuel meat when the fuels are retrieved from the basin and dried. Furthermore the composition and microstructure of the UAl{sub x} fuels further mitigate that risk.
Date: August 1996
Creator: Ebner, M. A.
Partner: UNT Libraries Government Documents Department

Proposed subcritical measurements for fresh and spent highly enriched plate type fuel assemblies

Description: A collaborative experimental research program has been established between industry and university partners to evaluate the subcritical behavior of fresh and spent highly enriched fuel assemblies at the University of Missouri Research Reactor (MURR). This proposed program will involve a series of subcritical measurements using the Oak Ridge National Laboratory (ORNL) developed {sup 252}Cf source-driven noise technique. Measurements evaluating the subcritical behavior of simple arrays of fresh MURR assemblies will be performed for evaluating the spectral effects of materials typically found in shipping casks such as lead, steel, aluminum, and boron. Also, measurements will be performed on spent assemblies to characterize physics parameters which may be useful in determining the subcritical behavior of fuels for reactivity credit of actinide burnup and fission product poisoning.
Date: September 1, 1997
Creator: Zino, J.F.; Williamson, T.G. & Mihalczo, J.T.
Partner: UNT Libraries Government Documents Department

RERTR fuel fabrication glovebox and facility development at ANL-W

Description: In order to support fuel plate production and physical metallurgy studies at ANL-W for the RERTR program, extensive facility modifications and equipment installation are underway. The particulate nature of the uranium alloys used in the fuel plate production requires glovebox isolation for several of the processing steps. A small glovebox was installed to meet the short-term powder processing needs of the project. A larger glovebox has been designed to handle the expanding needs of the project. In addition, a rolling mill and furnace were installed to allow hot rolling of the fuel plates. An arc-melting furnace will provide feedstock for powder production and metallurgy studies on uranium alloys. Future plans include the potential installation of a gas atomizer to aid in powder production.
Date: October 1, 1997
Creator: Clark, C.R.; Hansen, P.A.; Lawrence, J.D. & Meyer, M.K.
Partner: UNT Libraries Government Documents Department

Flow excursion time scales in the advanced neutron source reactor

Description: Flow excursion transients give rise to a key thermal limit for the proposed Advanced Neutron Source (ANS) reactor because its core involves many parallel flow channels with a common pressure drop. Since one can envision certain accident scenarios in which the thermal limits set by flow excursion correlations might be exceeded for brief intervals, a key objective is to determine how long a flow excursion would take to bring about a system failure that could lead to fuel damage. The anticipated time scale for flow excursions has been examined by subdividing the process into its component phenomena: bubble nucleation and growth, deceleration of the resulting two-phase flow, and finally overcoming thermal inertia to heat up the reactor fuel plates. Models were developed to estimate the time required for each individual stage. Accident scenarios involving sudden reduction in core flow or core exit pressure have been examined, and the models compared with RELAP5 output for the ANS geometry. For a high-performance reactor like the ANS, flow excursion time scales were predicted to be in the millisecond range, so that even very brief transients might lead to fuel damage. These results should prove useful whenever one must determine the time involved in any portion of a flow excursion transient.
Date: April 1995
Creator: Sulfredge, C. D.
Partner: UNT Libraries Government Documents Department

Design and fabrication of high density uranium dispersion fuels.

Description: Twelve different uranium alloys and compounds with uranium densities greater than 13.8 g/cc were fabricated into fuel plates. Sixty-four experimental fuel plates, referred to as microplates, with overall dimensions of 76.2 mm x 22.2 mm x 1.3 mm and elliptical fuel zone of nominal dimensions of 51 mm x 9.5 mm, began irradiation in the Advanced Test Reactor on August 23, 1997. The fuel test matrix consists of machined or comminuted (compositions are in weight%) U-10Mo,U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6 Ru, 10Mo-0.05Sn, U{sub 2}Mo and U{sub 3}Si{sub 2}(as a control). The low enriched ({sup 235}U < 20%) fuel materials were cast, powdered, mixed with aluminum dispersant at a volume ratio of 1:3, compacted and hot rolled to form the microplates. Spherical atomized powders of two fuels, U-10Mo and U{sub 3}Si{sub 2}, were utilized to make microplates and included in the irradiation test as well. The experimental design and fabrication steps employed in the selection and production of the fueled microplates is discussed.
Date: November 5, 1997
Creator: Clark, C. R.; McGann, D. J.; Meyer, M. K.; Trybus, C. L. & Wiencek, T. C.
Partner: UNT Libraries Government Documents Department


Description: Activities in a ptogram concerned with development of plasma-jet spray- coating techniques suitable for production of clad ceramic fuel plates are described. Experiments on application of zirconia coatings are also described. A survey of UO/sub 2/ powder was conducted to determine its suitability for plasma spraying. Also conditions were established for spraying fused and milled UO/sub 2/. The effects of process variables on coating and deposition characteristics were found to correlate. Densities of UO/sub 2/ coatings of 75 to 80% were achieved. (J.R.D.)
Date: October 31, 1962
Creator: Weare, N.E.
Partner: UNT Libraries Government Documents Department


Description: 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.
Date: March 1, 2012
Creator: Robinson, A. B.; Wachs, D. M.; Medvedev, P.; Miller, S.J.; Rice, F. J.; Meyer, M. K. et al.
Partner: UNT Libraries Government Documents Department