Characterization and testing of monolithic RERTR fuel plates

Energy Technology Data Exchange (ETDEWEB)

Keiser, D.D.; Jue, J.F.; Burkes, D.E. [Idaho National Lab., Idaho Falls, ID (United States)

2007-07-01

Monolithic fuel plates are being developed as a LEU (low enrichment uranium) fuel for application in research reactors throughout the world. These fuel plates are comprised of a U-Mo alloy foil encased in aluminum alloy cladding. Three different fabrication techniques have been looked at for producing monolithic fuel plates: hot isostatic pressing (HIP), transient liquid phase bonding (TLPB), and friction stir welding (FSW). Of these three techniques, HIP and FSW are currently being emphasized. As part of the development of these fabrication techniques, fuel plates are characterized and tested to determine properties like hardness and the bond strength at the interface between the fuel and cladding. Testing of HIP-made samples indicates that the foil/cladding interaction behavior depends on the Mo content in the UMo foil, the measured hardness values are quite different for the fuel, cladding, and interaction zone phase and Ti, Zr and Nb are the most effective diffusion barriers. For FSW samples, there is a dependence of the bond strength at the foil/cladding interface on the type of tool that is employed for performing the actual FSW process. (authors)

Swelling Estimation of Multi-wire U-Mo Monolithic Fuel for HANARO Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Yoon-Sang; Ryu, Ho-Jin; Park, Jong-Man; Oh, Jong-Myeong; Kim, Chang-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-10-15

In order to use low-enriched uranium (LEU) instead of highly enriched uranium (HEU) for high performance research reactors, the reduced enrichment for research and test reactors (RERTR) program is developing high uranium density fuel such as U-Mo/Al dispersion fuel. U-Mo alloys have an excellent irradiation performance when compared to other uranium alloys or compounds. But the results from the post-irradiation examination of the U-Mo/Al dispersion fuels indicate that an interaction between the U-Mo alloy fuel and the Al matrix phases occurs readily during an irradiation and it is sensitively dependent on the temperature. In order to lessen these severe interactions, a concept of a multi-wire type fuel was proposed. The fuel configuration is that three to six U-Mo fuel wires (1.5 mm - 2 mm in diameter) are symmetrically arranged at the periphery side in the Al matrix. In this study temperature calculations and a swelling estimation of a multi-wire monolithic fuel were carried out. Also the results of a post irradiation analysis of this fuel will be introduced.

Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

Science.gov (United States)

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

2015-09-01

Low-enrichment (235U < 20 pct) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing consisted of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.

Effects of the shape of the foil corners on the irradiation performance of U10Mo alloy based monolithic mini-plates

Energy Technology Data Exchange (ETDEWEB)

Ozaltun, Hakan [Idaho National Laboratory; Medvedev, Pavel G [Idaho National Laboratory

2015-06-01

Monolithic plate-type fuel is a fuel form being developed for high performance research and test reactors to minimize the use of enriched material. These fuel elements are comprised of a high density, low enrichment, U-Mo alloy based fuel foil, sandwiched between Zirconium liners and encapsulated in Aluminum cladding. The use of a high density fuel in a foil form presents a number of fabrication and operational concerns, such as: foil centering, flatness of the foil, fuel thickness variation, geometrical tilting, foil corner shape etc. To benchmark this new design, effects of various geometrical and operational variables on irradiation performance have been evaluated. As a part of these series of sensitivity studies, the shape of the foil corners were studied. To understand the effects of the corner shapes of the foil on thermo-mechanical performance of the plates, a behavioral model was developed for a selected plate from RERTR-12 experiments (Plate L1P785). Both fabrication and irradiation processes were simulated. Once the thermo-mechanical behavior the plate is understood for the nominal case, the simulations were repeated for two additional corner shapes to observe the changes in temperature, displacement and stress-strain fields. The results from the fabrication simulations indicated that the foil corners do not alter the post-fabrication stress-strain magnitudes. Furthermore, the irradiation simulations revealed that post-fabrication stresses of the foil would be relieved very quickly in operation. While, foils with chamfered and filleted corners yielded stresses with comparable magnitudes, they are slightly lower in magnitudes, and provided a more favorable mechanical response compared with the foil with sharp corners.

Simulation of the irradiation-induced thermo-mechanical behaviors evolution in monolithic U–Mo/Zr fuel plates under a heterogeneous irradiation condition

Energy Technology Data Exchange (ETDEWEB)

Zhao, Yunmei; Gong, Xin; Ding, Shurong, E-mail: dsr1971@163.com

2015-04-15

Highlights: • The three-dimensional stress update algorithms in a co-rotational framework are developed for U–Mo and Zircalloy with the irradiation effects. • An effective method for three-dimensional modeling of the in-pile behaviors in heterogeneously irradiated monolithic fuel plates is established and validated. • The effects of the fission-induced creep effects in the U–Mo foil are investigated in detail. • A deformation phenomenon similar to the irradiation experimental results is obtained. - Abstract: For monolithic fuel plates with U–Mo foil and Zircalloy cladding, the three-dimensional large deformation incremental constitutive relations and stress update algorithms in the co-rotational coordinate framework are developed for the fuel and cladding with their respective irradiation effects involved. Three-dimensional finite element simulation of their in-pile thermo-mechanical coupling behaviors under a location-dependent irradiation condition is implemented via the validated user-defined subroutines UMATHT and UMAT in ABAQUS. Comparison of the simulation results for two cases with or without creep considered in the U–Mo foil indicates that with the irradiation creep included (1) considerable stress-relaxation appears in the U–Mo foil, and the mechanical interaction between fuel and cladding is weakened; (2) approximately identical thickness increments in the plate and fuel foil exist and become comparably larger; (3) plastic deformation in the cladding is significantly diminished.

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

International Nuclear Information System (INIS)

Keiser, D.D. Jr.; Yao, B.; Perez, E.; Sohn, Y.H.

2009-01-01

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.

Effects of irradiation on the interface between U-Mo and zirconium diffusion barrier

Science.gov (United States)

Jue, Jan-Fong; Keiser, Dennis D.; Miller, Brandon D.; Madden, James W.; Robinson, Adam B.; Rabin, Barry H.

2018-02-01

Irradiated fuel plates were characterized by microscopy that focused on the interface between U-Mo and Zr. Before irradiation, there were three major sub-layers identified in the U-Mo/Zr interface, namely, UZr2, Mo2Zr, and U with low Mo. The typical total thickness of this U-Mo/Zr interaction is 2-3 μm. The UZr2 sub-layer formed during fuel plate fabrication remains stable after irradiation, without large bubbles/porosity accumulation. However, this sub-layer becomes increasingly discontinuous as burnup increases. The low-Mo sub-layer exhibits numerous sub-micron bubbles/porosity at low burnup. Larger, interconnected porosity in this sub-layer was observed in a medium-burnup fuel specimen. However, at higher burnup, regions with the extra-large bubbles/porosity (i.e., larger than 5 μm) were observed in the U-Mo fuel foil at least 5 μm away from the original location of this sub-layer. The mechanism for the formation of the extra-large bubbles/porosity is still unclear at this time. In general, the U-Mo/Zr interface in monolithic U-Mo fuels is relatively stable after irradiation. No large detrimental defects, such as large interfacial bubbles or cracks/delamination, were observed in the fuel plates characterized.

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

Energy Technology Data Exchange (ETDEWEB)

Pasqualini, E.E. [Laboratorio de Nanotecnología Nuclear, Centro Atómico Constituyentes, Comisión Nacional de Energía Atómica, Av. General Paz 1499, B1650KNA, San Martín, Prov. Buenos Aires (Argentina); Robinson, A.B. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Wachs, D.M. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organisation, PMB 1, Menai, NSW, 2234 (Australia)

2016-10-15

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm{sup 3}, 3.8E+21 (peak).

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

Science.gov (United States)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

Energy Technology Data Exchange (ETDEWEB)

Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed. - Highlights: •Complementary fission gas release events are reported for U-Mo fuel with and without cladding. •Exothermic reaction between Zr diffusion layer and cladding influences fission gas release. •Mechanisms responsible for fission gas release are similar, but with varying timing and magnitude. •Behavior of samples is similar after 800 °C signaling the onset of superlattice destabilization.

Study on characteristics of U-Mo/Al-Si interaction layers of dispersion fuel plates

International Nuclear Information System (INIS)

Liu Lijian; Yin Changgeng; Chen Jiangang; Sun Changlong; Liu Yunming

2014-01-01

In this paper, we analyzed the characteristics of U-Mo/Al-Si interaction layers of dispersion fuel plates. The results show that the interaction layers (IL) are with irregular morphology and uneven thickness, and are mainly formed in the internal micro cracks of the dispersion fuel particles or at the interface between the particles and the substrates. The diffusion mechanism of U-Mo/Al-Si is the vacancy diffusion, Al and Si are migrating elements, and the diffusion reaction is that Al and Si diffuse to U-Mo alloy. Inside the interaction layers, the Al content keeps constant basically, but the Si content gradually increases with the substrate-fuel direction, and the maximum content of Si appears interaction layers near the U-Mo side. Adding about 5 wt% Si into Al matrix can restrain the diffusion reaction, and improve the performance of dispersion fuel plates finally. (authors)

The Microstructure of Multi-wire U-Mo Monolithic Fuel

Energy Technology Data Exchange (ETDEWEB)

Lee, Yoon Sang; Park, Eun Kee; Cho, Woo Hyoung; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-10-15

In order to use low-enriched uranium (LEU) instead of highly enriched uranium (HEU) for high performance research reactors, the reduced enrichment for research and test reactors (RERTR) program is developing high uranium density fuel such as U-Mo/Al dispersion fuel. U-Mo alloys have an excellent irradiation performance when compared to other uranium alloys or compounds. But the results from the post-irradiation examination of the U-Mo/Al dispersion fuels indicate that an interaction between the U-Mo alloy fuel and the Al matrix phases occurs readily during an irradiation and it is sensitively dependent on the temperature. In order to lessen these severe interactions, a concept of a multi-wire type fuel was proposed. The fuel configuration is that three to six U-Mo fuel wires (1.5 mm {approx} 2 mm in diameter) are symmetrically arranged at the periphery side in the Al matrix as shown. This multi-wire fuels showed very good fuel performance during the KOMO-3 irradiation test. At the KOMO-3 test, the specimen of the multi-wire fuels were U-7Mo/Al and U-7Mo-1Si/Al. In this study we investigate the microstructure change of the U-7Mo and U-7Mo-1Ti with some variation of annealing conditions. In addition to this, we want to check the effect of adding Ti element to U-7Mo on the gamma phase stability

Results of Microstructural Examinations of Irradiated LEU U-Mo Fuels

Energy Technology Data Exchange (ETDEWEB)

Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, ID 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organization (Australia)

2009-06-15

Introduction: The RERTR program is responsible for converting research reactors that use high-enriched uranium fuels to ones that use low-enriched uranium fuels [1]. As part of the development of LEU fuels, a variety of irradiation experiments are being conducted using the Advanced Test Reactor. Based on the results of initial fuel plate testing, adjustments have been made to the characteristics of fuel plates to improve the stability of the fuel microstructure. One improvement has been to add Si to the matrix of a dispersion fuel. This material is also being added at the fuel/cladding interface of a monolithic fuel. This paper will discuss the irradiation performance of these fuels, in terms of the stability of their microstructures during irradiation. Results and discussion: The post-irradiation examinations of fuel plates are performed at the Idaho National Laboratory. These examinations consist of visual examinations of fuel plates, gamma scanning, thickness measurements, oxide thickness measurements, and optical metallographic examinations of the fuel plate microstructures. Microstructural analysis is also performed using scanning electron microscopy. Overall, U-7Mo and U-10Mo alloy fuels have displayed the best irradiation performance, particularly, when a Si-containing Al alloy is used as the dispersion fuel matrix. The benefit of using this type of matrix is that the commonly observed fuel/cladding interaction that occurs during irradiation is reduced and the interaction layer that forms exhibit stable behavior during irradiation. Monolithic-type fuels, which consist of a U-Mo foil encased in Al alloy cladding, are also being developed. These types of fuels are also showing promise and will continue to be developed. One challenge with this type of fuel is in trying to maximize the bond strength at the foil/cladding interface. Fuel/cladding interactions can affect the quality of the boding at this interface. Si is being added to improve the characteristics

Modeling of high-density U-MO dispersion fuel plate performance

International Nuclear Information System (INIS)

Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Rest, J.; Snelgrove, J.L.

2002-01-01

Results from postirradiation examinations (PIE) of highly loaded U-Mo/Al dispersion fuel plates over the past several years have shown that the interaction between the metallic fuel particles and the matrix aluminum can be extensive, reducing the volume of the high-conductivity matrix phase and producing a significant volume of low-conductivity reaction-product phase. This phenomenon results in a significant decrease in fuel meat thermal conductivity during irradiation. PIE has further shown that the fuel-matrix interaction rate is a sensitive function of irradiation temperature. The interplay between fuel temperature and fuel-matrix interaction makes the development of a simple empirical correlation between the two difficult. For this reason a comprehensive thermal model has been developed to calculate temperatures throughout the fuel plate over its lifetime, taking into account the changing volume fractions of fuel, matrix and reaction-product phases within the fuel meat owing to fuel-matrix interaction; this thermal model has been incorporated into the dispersion fuel performance code designated PLATE. Other phenomena important to fuel thermal performance that are also treated in PLATE include: gas generation and swelling in the fuel and reaction-product phases, incorporation of matrix aluminum into solid solution with the unreacted metallic fuel particles, matrix extrusion resulting from fuel swelling, and cladding corrosion. The phenomena modeled also make possible a prediction of fuel plate swelling. This paper presents a description of the models and empirical correlations employed within PLATE as well as validation of code predictions against fuel performance data for U-Mo experimental fuel plates from the RERTR-3 irradiation test. (author)

Conceptual designs parameters for MURR LEU U-Mo fuel conversion design demonstration experiment. Revision 1

International Nuclear Information System (INIS)

Stillman, J.; Feldman, E.; Stevens, J.

2013-01-01

The design parameters for the conceptual design of a fuel assembly containing U-10Mo fuel foils with low-enriched uranium (LEU) for the University of Missouri Research Reactor (MURR) are described. The Design Demonstration Experiment (MURR-DDE) will use a prototypic MURR-LEU element manufactured according to the parameters specified here. Also provided are calculated performance parameters for the LEU element in the MURR, and a set of goals for the MURR-DDE related to those parameters. The conversion objectives are to develop a fuel element design that will ensure safe reactor operations, as well as maintaining existing performance. The element was designed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. A set of manufacturing assumptions were provided by the Fuel Development (FD) and Fuel Fabrication Capability (FFC) pillars of the GTRI Reduced Enrichment for Research and Test Reactors (RERTR) program to reliably manufacture the fuel plates. The proposed LEU fuel element has an overall design and exterior dimensions that are similar to those of the current highly-enriched uranium (HEU) fuel elements. There are 23 fuel plates in the LEU design. The overall thickness of each plate is 44 mil, except for the exterior plate that is furthest from the center flux trap (plate 23), which is 49 mil thick. The proposed LEU fuel plates have U-10Mo monolithic fuel foils with a 235U enrichment of 19.75% varying from 9 mil to 20 mil thick, and clad with Al-6061 aluminum. A thin layer of zirconium exists between the fuel foils and the aluminum as a diffusion barrier. The thinnest nominal combined zirconium and aluminum clad thickness on each side of the fuel plates is 12 mil. The LEU U-10Mo monolithic fuel is not yet qualified as driver fuel in research reactors, but is under intense development under the auspices of the GTRI FD and FFC programs.

Interaction Layer Characteristics in U-xMo Dispersion/Monolithic Fuels

International Nuclear Information System (INIS)

Porter, D.L.

2010-01-01

before irradiation of a fuel plate. It ensures that the IL contains beneficial phases, or prevents formation of some known to promote poor fuel performance. Significant progress has been made in determining the desired characteristics of the IL. (4) The use of a fuel with stable gamma phase appears to allow more predictable performance regarding both a beneficial pre-irradiation layer, and the fuel performance (low swelling) to high burnup. Destabilization of the gamma phase may create problems with IL breakaway growth. (5) A theory whereby prevention of the U6Mo4Al43 complex phase in interaction layers formed during fabrication may be a key to good irradiation performance. Si additions to the matrix allow for solubility of Mo in the desirable (U,Mo)(Al,Si)3 or perhaps (U,Mo)(Al,Si)4 phase, helping to prevent formation of the complex phase. Keeping alloy Mo content as low as possible may also help so long as decomposition does not occur in fabrication, forcing Mo into the interaction layer. This theory may explain a number of apparent anomalies observed in testing results. (6) More work is needed in order to prescribe the conditions to best produce a beneficial IL. Another necessity is a better understanding of any correlation between beneficial characteristics of the pre-fabricated IL and the irradiation conditions to which it will be subjected.

Interaction Layer Characteristics in U-xMo Dispersion/Monolithic Fuels

Energy Technology Data Exchange (ETDEWEB)

D. L. Porter

2010-11-01

before irradiation of a fuel plate. It ensures that the IL contains beneficial phases, or prevents formation of some known to promote poor fuel performance. Significant progress has been made in determining the desired characteristics of the IL. 4. The use of a fuel with stable gamma phase appears to allow more predictable performance regarding both a beneficial pre-irradiation layer, and the fuel performance (low swelling) to high burnup. Destabilization of the gamma phase may create problems with IL breakaway growth. 5. A theory whereby prevention of the U6Mo4Al43 complex phase in interaction layers formed during fabrication may be a key to good irradiation performance. Si additions to the matrix allow for solubility of Mo in the desirable (U,Mo)(Al,Si)3 or perhaps (U,Mo)(Al,Si)4 phase, helping to prevent formation of the complex phase. Keeping alloy Mo content as low as possible may also help so long as decomposition does not occur in fabrication, forcing Mo into the interaction layer. This theory may explain a number of apparent anomalies observed in testing results. 6. More work is needed in order to prescribe the conditions to best produce a beneficial IL. Another necessity is a better understanding of any correlation between beneficial characteristics of the pre-fabricated IL and the irradiation conditions to which it will be subjected.

Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

Science.gov (United States)

Jue, Jan-Fong; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

2014-05-01

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative is developing an advanced monolithic fuel to convert US high-performance research reactors to low-enriched uranium. Hot-isostatic-press (HIP) bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U-Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between the fuel meat, the cladding, and the diffusion barrier, as well as between the U-10Mo fuel meat and the Al-6061 cladding, were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are: line. Some of these attributes might be critical to the irradiation performance of monolithic U-10Mo nuclear fuel. There are several issues or concerns that warrant more detailed study, such as precipitation along the cladding-to-cladding bond line, chemical banding, uncovered fuel-zone edge, and the interaction layer between the U-Mo fuel meat and zirconium. Future post-irradiation examination results will focus, among other things, on identifying in-reactor failure mechanisms and, eventually, directing further fresh fuel characterization efforts.

Mechanical Calculations on U-Mo Dispersion fuel plates with MAIA

International Nuclear Information System (INIS)

Marelle, V.; Huet, F.; Lemoine, P.

2005-01-01

CEA has developed a 2D thermo-mechanical code, called MAIA, for modelling the behaviour of U-Mo dispersion fuel. MAIA uses a finite element method for the resolution of the thermal and mechanical problems. Physical models, issued of the DOE-ANL code PLATE, evaluate the fission products swelling and the volume fraction of the interaction between U-Mo and Al. They allow establishing strains in the meat imposed as loading for the mechanical calculation. MAIA has been validated on the irradiations IRIS 1 and RERTR-3 and a rather good agreement is obtained with post irradiation examinations. MAIA is used to calculate the last irradiation of the French UMo group, IRIS 2. MAIA predicts a maximum temperature of 112 deg. C and meat swelling of 16%. Mechanical calculations are finally performed to evaluate the sensitivity to some mechanical hypotheses such as constitutive laws and the way the meat swelling is applied. (author)

Coated U(Mo) Fuel: As-Fabricated Microstructures

Energy Technology Data Exchange (ETDEWEB)

Emmanuel Perez; Dennis D. Keiser, Jr.; Ann Leenaers; Sven Van den Berghe; Tom Wiencek

2014-04-01

As part of the development of low-enriched uranium fuels, fuel plates have recently been tested in the BR-2 reactor as part of the SELENIUM experiment. These fuel plates contained fuel particles with either Si or ZrN thin film coating (up to 1 µm thickness) around the U-7Mo fuel particles. In order to best understand irradiation performance, it is important to determine the starting microstructure that can be observed in as-fabricated fuel plates. To this end, detailed microstructural characterization was performed on ZrN and Si-coated U-7Mo powder in samples taken from AA6061-clad fuel plates fabricated at 500°C. Of interest was the condition of the thin film coatings after fabrication at a relatively high temperature. Both scanning electron microscopy and transmission electron microscopy were employed. The ZrN thin film coating was observed to consist of columns comprised of very fine ZrN grains. Relatively large amounts of porosity could be found in some areas of the thin film, along with an enrichment of oxygen around each of the the ZrN columns. In the case of the pure Si thin film coating sample, a (U,Mo,Al,Si) interaction layer was observed around the U-7Mo particles. Apparently, the Si reacted with the U-7Mo and Al matrix during fuel plate fabrication at 500°C to form this layer. The microstructure of the formed layer is very similar to those that form in U-7Mo versus Al-Si alloy diffusion couples annealed at higher temperatures and as-fabricated U-7Mo dispersion fuel plates with Al-Si alloy matrix fabricated at 500°C.

Feasibility Study for Electrical Discharge Machining of Large DU-Mo Castings

Energy Technology Data Exchange (ETDEWEB)

Hill, Mary Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). SIGMA Division; Dombrowski, David E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). SIGMA Division; Clarke, Kester Diederik [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). SIGMA Division; Forsyth, Robert Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). SIGMA Division; Aikin, Robert M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). SIGMA Division; Alexander, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). SIGMA Division; Tegtmeier, Eric Lee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). SIGMA Division; Robison, Jeffrey Curt [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). SIGMA Division; Beard, Timothy Vance [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). SIGMA Division; Edwards, Randall Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). SIGMA Division; Mauro, Michael Ernest [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). SIGMA Division; Scott, Jeffrey E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). SIGMA Division; Strandy, Matthew Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). SIGMA Division

2016-10-31

U-10 wt. % Mo (U-10Mo) alloys are being developed as low enrichment monolithic fuel for the CONVERT program. Optimization of processing for the monolithic fuel is being pursued with the use of electrical discharge machining (EDM) under CONVERT HPRR WBS 1.2.4.5 Optimization of Coupon Preparation. The process is applicable to manufacturing experimental fuel plate specimens for the Mini-Plate-1 (MP-1) irradiation campaign. The benefits of EDM are reduced machining costs, ability to achieve higher tolerances, stress-free, burr-free surfaces eliminating the need for milling, and the ability to machine complex shapes. Kerf losses are much smaller with EDM (tenths of mm) compared to conventional machining (mm). Reliable repeatability is achievable with EDM due to its computer-generated machining programs.

Feasibility Study for Electrical Discharge Machining of Large DU-Mo Castings

International Nuclear Information System (INIS)

Hill, Mary Ann; Dombrowski, David E.; Clarke, Kester Diederik; Forsyth, Robert Thomas; Aikin, Robert M.; Alexander, David John; Tegtmeier, Eric Lee; Robison, Jeffrey Curt; Beard, Timothy Vance; Edwards, Randall Lynn; Mauro, Michael Ernest; Scott, Jeffrey E.; Strandy, Matthew Thomas

2016-01-01

U-10 wt. % Mo (U-10Mo) alloys are being developed as low enrichment monolithic fuel for the CONVERT program. Optimization of processing for the monolithic fuel is being pursued with the use of electrical discharge machining (EDM) under CONVERT HPRR WBS 1.2.4.5 Optimization of Coupon Preparation. The process is applicable to manufacturing experimental fuel plate specimens for the Mini-Plate-1 (MP-1) irradiation campaign. The benefits of EDM are reduced machining costs, ability to achieve higher tolerances, stress-free, burr-free surfaces eliminating the need for milling, and the ability to machine complex shapes. Kerf losses are much smaller with EDM (tenths of mm) compared to conventional machining (mm). Reliable repeatability is achievable with EDM due to its computer-generated machining programs.

A modelling study of the inter-diffusion layer formation in U-Mo/Al dispersion fuel plates at high power

Energy Technology Data Exchange (ETDEWEB)

Ye, B.; Hofman, G. L.; Leenaers, A.; Bergeron, A.; Kuzminov, V.; Van den Berghe, S.; Kim, Y. S.; Wallin, H.

2018-02-01

Post irradiation examinations of full-size U-Mo/Al dispersion fuel plates fabricated with ZrN- or Sicoated U-Mo particles revealed that the reaction rate of irradiation-induced U-Mo-Al inter-diffusion, an important microstructural change impacting the performance of this type of fuel, is temperature and fission-rate dependent. In order to simulate the U-Mo/Al inter-diffusion layer (IL) growth behavior in full-size dispersion fuel plates, the existing IL growth correlation was modified with a temperaturedependent multiplication factor that transits around a threshold fission rate. In-pile irradiation data from four tests in the BR2 reactors, including FUTURE, E-FUTURE, SELEMIUM, and SELEMIUM-1a, were utilized to determine and validate the updated IL growth correlation. Irradiation behavior of the plates was simulated with the DART-2D computational code. The general agreement between the calculated and measured fuel meat swelling and constituent volume fractions as a function of fission density demonstrated the plausibility of the updated IL growth correlation. The simulation results also suggested the temperature dependence of the IL growth rate, similar to the temperature dependence of the intermixing rate in ion-irradiated bi-layer systems.

Thermal conductivity of fresh and irradiated U-Mo fuels

Science.gov (United States)

Huber, Tanja K.; Breitkreutz, Harald; Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Elgeti, Stefan; Reiter, Christian; Robinson, Adam. B.; Smith, Frances. N.; Wachs, Daniel. M.; Petry, Winfried

2018-05-01

The thermal conductivity of fresh and irradiated U-Mo dispersion and monolithic fuel has been investigated experimentally and compared to theoretical models. During in-pile irradiation, thermal conductivity of fresh dispersion fuel at a temperature of 150 °C decreased from 59 W/m·K to 18 W/m·K at a burn-up of 4.9·1021 f/cc and further to 9 W/m·K at a burn-up of 6.1·1021 f/cc. Fresh monolithic fuel has a considerably lower thermal conductivity of 15 W/m·K at a temperature of 150 °C and consequently its decrease during in-pile irradiation is less steep than for dispersion fuel. For a burn-up of 3.5·1021 f/cc of monolithic fuel, a thermal conductivity of 11 W/m·K at a temperature of 150 °C has been measured by Burkes et al. (2015). The difference of decrease for both fuels originates from effects in the matrix that occur during irradiation, like for dispersion fuel the gradual disappearance of the Al matrix with increased burn-up and the subsequent growth of an interaction layer (IDL) between the U-Mo fuel particle and Al matrix and subsequent matrix hardening. The growth of fission gas bubbles and the decomposition of the U-Mo crystal lattice also affect both dispersion and monolithic fuel.

Parametric study of fission-induced U-Mo fuel creep and structural analysis of fuel plates in view of implications for microstructure evolution

International Nuclear Information System (INIS)

Kim, Y.S.; Hofman, G.L.; Choo, Y.S.; Robinson, A.B.

2010-01-01

U-Mo fuel deformation during irradiation in U-Mo/Al dispersion plates is investigated by using the irradiation data from the RERTR-3 through -9 tests. The observation of fuel particle sintering during irradiation is also presented and its influence for fuel performance is discussed. Structural analysis was also performed to examine the relationship between the stress distribution in the plate and the location of matrix-pore formation in the plate. (author)

PENGARUH SERBUK U-Mo HASIL PROSES MEKANIK DAN HYDRIDE – DEHYDRIDE – GRINDING MILL TERHADAP KUALITAS PELAT ELEMEN BAKAR U-Mo/Al

Directory of Open Access Journals (Sweden)

Supardjo Supardjo

2015-07-01

serbuk dapat diperkecil.   INFLUENCE OF U-Mo POWDER BY MECHANICAL AND HYDRIDE - DEHYDRIDE - GRINDING MILL PROCESS RESULT OF U-Mo / Al FUEL PLATE QUALITY. Research of U-7Mo/Al fuel type plate is done in order to develop U3Si2/Al fuel to get a new fuel that has a higher uranium density, stable for use as fuel in the reactor and is easily done if the reprocessed. The scope of the research includes manufacture: U-7Mo alloy with smelting techniques, pulverizing U-7Mo to be filed and hydride–dehydride–grinding mill, U-7Mo/Al fuel core with the technique of compacting at a pressure of 20 bar, and U-7Mo/Al fuel plate with technique of hot rolling at a temperature of 425oC. The U-7Mo alloy results smelting process quite homogeneous, the density of 16.34 g/cm3 and is tenacious, then made powder by means of filed and hydride–dehydride–grinding mill. The U-7Mo powder shaped flat results miserly process, contaminants Fe is high enough, whereas powder process results hydride- dehydride-grinding mill, tend equiaxial with low contaminants. The second type of U-7Mo powder is used as a raw material for making U-7Mo/Al fuel core and U-7Mo/Al fuel plate with 7 gU/cm3 uranium density and obtained product with almost the same quality. The U-7Mo/Al fuel core test results measuring 25 x 15 x 3.15 ± 0.05 mm, there is no defect/crack, U-7Mo distribution in the matrix is quite homogeneous and there is no grouping/agglomeration U-7Mo dimension >1 mm. The U-7Mo/Al fuel plate outcome rolling with a final thickness of 1.45 mm, has a thickness of 0.60 mm and a mean meat cladding thickness of 0.4 mm, and there is one point of measurement of cladding with a thickness of 0.15 mm. By comparing the use of both types of U-7Mo powders the U-7Mo/Al fuel core and U-7Mo/Al fuel plate produced has almost the same quality. However, the use of U-7Mo powder results hydride– dehydride–grinding mill process is better because the workmanship is faster and impurities in the powders can be minimized.

High density fuels using dispersion and monolithic fuel

Energy Technology Data Exchange (ETDEWEB)

Gomes, Daniel S.; Silva, Antonio T.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br, E-mail: alfredo@ctmsp.mar.mil.br, E-mail: rafael.orm@gmail.com, E-mail: claudia.giovedi@ctmsp.mar.mil.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Departamento de Engenharia Naval e Oceânica

2017-07-01

Fuel plates used in high-performance research reactors need to be converted to low-enrichment uranium fuel; the fuel option based on a monolithic formulation requires alloys to contain 6 - 10 wt% Mo. In this case, the fuel plates are composed of the metallic alloy U-10Mo surrounded by a thin zirconium layer encapsulated in aluminum cladding. This study reviewed the physical properties of monolithic forms. The constraints produced during the manufacturing process were analyzed and compared to those of dispersed fuel. The bonding process used for dispersion fuels differs from the techniques applied to foil bonding used for pure alloys. The quality of monolithic plates depends on the fabrication method, which usually involves hot isostatic pressing and the thermal annealing effect of residual stress, which degrades the uranium cubic phase. The preservation of the metastable phase has considerable influence on fuel performance. The physical properties of the foil fuel under irradiation are superior to those of aluminum-dispersed fuels. The fuel meat, using zirconium as the diffusion barrier, prevents the interaction layer from becoming excessively thick. The problem with dispersed fuel is breakaway swelling with a medium fission rate. It has been observed that the fuel dispersed in aluminum was minimized in monolithic forms. The pure alloys exhibited a suitable response from a rate at least twice as much as the fission rate of dispersions. The foils can support fissile material concentration combined with a reduced swelling rate. (author)

High density fuels using dispersion and monolithic fuel

International Nuclear Information System (INIS)

Gomes, Daniel S.; Silva, Antonio T.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia; Universidade de São Paulo

2017-01-01

Fuel plates used in high-performance research reactors need to be converted to low-enrichment uranium fuel; the fuel option based on a monolithic formulation requires alloys to contain 6 - 10 wt% Mo. In this case, the fuel plates are composed of the metallic alloy U-10Mo surrounded by a thin zirconium layer encapsulated in aluminum cladding. This study reviewed the physical properties of monolithic forms. The constraints produced during the manufacturing process were analyzed and compared to those of dispersed fuel. The bonding process used for dispersion fuels differs from the techniques applied to foil bonding used for pure alloys. The quality of monolithic plates depends on the fabrication method, which usually involves hot isostatic pressing and the thermal annealing effect of residual stress, which degrades the uranium cubic phase. The preservation of the metastable phase has considerable influence on fuel performance. The physical properties of the foil fuel under irradiation are superior to those of aluminum-dispersed fuels. The fuel meat, using zirconium as the diffusion barrier, prevents the interaction layer from becoming excessively thick. The problem with dispersed fuel is breakaway swelling with a medium fission rate. It has been observed that the fuel dispersed in aluminum was minimized in monolithic forms. The pure alloys exhibited a suitable response from a rate at least twice as much as the fission rate of dispersions. The foils can support fissile material concentration combined with a reduced swelling rate. (author)

Irradiation experiment conceptual design parameters for MURR LEU U-Mo fuel conversion

International Nuclear Information System (INIS)

Stillman, J.; Feldman, E.; Stevens, J.; Wilson, E.

2013-03-01

This report contains the results of reactor design and performance calculations for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the nominal steady-state irradiation conditions of a key set of plates containing peak irradiation parameters found in MURR cores fueled with the LEU monolithic U-Mo alloy fuel with 10 wt% Mo.

Progress in development of low-enriched U-Mo dispersion fuels

International Nuclear Information System (INIS)

Hofman, G.L.; Snelgrove, J.L.; Hayes, S.L.; Meyer, M.K.

2002-01-01

Results from post irradiation examinations and analyses of U-Mo/Al dispersion mini plates are presented. Irradiation test RERTR-5 contained mini- fuel plates with fuel loadings of 6 and 8 g U cm -3 . The fuel material consisted of 6, 7 and 10 wt. % Mo-uranium-alloy powders in atomized and machined form. The swelling behavior of the various fuel types is analyzed, indicating athermal swelling of the U-Mo alloy and temperature-dependent swelling owing to U-Mo/Al interdiffusion. (author)

PLACA/DPLACA: a code to simulate the behavior of a monolithic/dispersed plate type fuel

International Nuclear Information System (INIS)

Denis, Alicia; Soba, Alejandro

2005-01-01

The PLACA code was originally built to simulate monolithic plate fuels contained in a metallic cladding, with a gap in between. The international program of high density fuels was recently oriented to the development of a plate-type fuel of a uranium rich alloy with a molybdenum content between 6 to 10 w %, without gap and with a Zircaloy cladding. To give account of these fuels, the DPLACA code was elaborated as a modification of the original code. The extension of the calculation tool to disperse fuels involves a detailed study of the properties and models (still in progress). Of special interest is the material formed by U Mo particles dispersed in an Al matrix. This material has appeared as a candidate fuel for high flux research reactors. However, the interaction layer that grows around the particles has a deleterious effect on the material performance in operation conditions and may represent a limit for its applicability. A number of recent experiments carried out on this material provide abundant information that allows testing of the numerical models. (author)

Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

Energy Technology Data Exchange (ETDEWEB)

Leenaers, A., E-mail: aleenaer@sckcen.be [Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Van den Berghe, S.; Koonen, E.; Kuzminov, V. [Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Detavernier, C. [Department of Solid State Sciences, Ghent University, Krijgslaan 281/S1, 9000 Ghent (Belgium)

2015-03-15

In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK• CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% {sup 235}U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL–matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium–Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)–matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

Science.gov (United States)

Leenaers, A.; Van den Berghe, S.; Koonen, E.; Kuzminov, V.; Detavernier, C.

2015-03-01

In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK•CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% 235U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL-matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)-matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

DEVELOPMENT OF HIGH-DENSITY U/AL DISPERSION PLATES FOR MO-99 PRODUCTION USING ATOMIZED URANIUM POWDER

OpenAIRE

RYU, HO JIN; KIM, CHANG KYU; SIM, MOONSOO; PARK, JONG MAN; LEE, JONG HYUN

2013-01-01

Uranium metal particle dispersion plates have been proposed as targets for Molybdenum-99 (Mo-99) production to improve the radioisotope production efficiency of conventional low enriched uranium targets. In this study, uranium powder was produced by centrifugal atomization, and miniature target plates containing uranium particles in an aluminum matrix with uranium densities up to 9 g-U/cm3 were fabricated. Additional heat treatment was applied to convert the uranium particles into UAlx compou...

Evaluation of plate type fuel options for small power reactors

International Nuclear Information System (INIS)

Andrzejewski, Claudio de Sa

2005-01-01

Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO 2 in stainless steel, of UO 2 in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

U-Mo fuels handbook. Version 1.0

International Nuclear Information System (INIS)

Rest, Jeffrey; Kim, Yeon Soo; Hofman, Gerard L.; Meyer, Mitchell K.; Hayes, Steven L.

2006-01-01

This handbook provides an overview of property data and fuel performance topics with an emphasis on data available for U-Mo alloys. These data often exist only in report format and have not been widely disseminated in the journal literature. For some topics there is more than one source of data, which are sometimes inconsistent. In this situation, the authors have attempted to select the best dataset to provide a standard for fuel designers and reactor operators. Following the section on unirradiated and irradiated materials properties for the monolithic U-Mo alloy, property data for cladding and matrix aluminum are presented. Property data for cladding aluminum are more widely available, and are not presented in great depth. Finally, some properties of (U-Mo)/Al dispersions are also included in this document. Where no data are available, best estimate correlations are provided. Best fits to the data are presented in order to facilitate use by fuel designers and reactor operators.

DEVELOPMENT OF HIGH-DENSITY U/AL DISPERSION PLATES FOR MO-99 PRODUCTION USING ATOMIZED URANIUM POWDER

Directory of Open Access Journals (Sweden)

HO JIN RYU

2013-12-01

Full Text Available Uranium metal particle dispersion plates have been proposed as targets for Molybdenum-99 (Mo-99 production to improve the radioisotope production efficiency of conventional low enriched uranium targets. In this study, uranium powder was produced by centrifugal atomization, and miniature target plates containing uranium particles in an aluminum matrix with uranium densities up to 9 g-U/cm3 were fabricated. Additional heat treatment was applied to convert the uranium particles into UAlx compounds by a chemical reaction of the uranium particles and aluminum matrix. Thus, these target plates can be treated with the same alkaline dissolution process that is used for conventional UAlx dispersion targets, while increasing the uranium density in the target plates

Interim Report on Mixing During the Casting of LEU-10Mo Plates in the Triple Plate Molds

Energy Technology Data Exchange (ETDEWEB)

Aikin, Jr., Robert M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-04-12

LEU-10%Mo castings are commonly produced by down blending unalloyed HEU with a DU-12.7%Mo master-alloy. This work uses process modeling to provide insight into the mixing of the unalloyed uranium and U-Mo master alloy during melting and mold filling of a triple plate casting. Two different sets of situations are considered: (1) mixing during mold filling from a compositionally stratified crucible and (2) convective mixing of a compositionally stratified crucible during mold heating. The mold filling simulations are performed on the original Y-12 triple plate mold and the horizontal triple plate mold.

Microstructural studies on chemical interactions in U-Mo with Al

International Nuclear Information System (INIS)

Martins, Ilson Carlos

2010-01-01

This research refers to the study of U-Mo alloy as an alternative material for producing nuclear fuel elements with high density of uranium, for research reactors of high performance. The international non-proliferation of nuclear weapons has enrichment level limited to 20% U 23 '5. U-Mo alloys with 6-10 wt% Mo can lead to a density up to 9 gU/cm 3 , inside the fuel core. The MTR fuel element plates are made from briquettes (U-Mo powder + Al) encapsulated in Al plates, then welded and rolled However, the U-Mo alloy is very reactive in the presence of Al. The reaction products of this interaction are undesirable from the standpoint of nuclear usage, since they cause a chemical interaction layer (IL) formed during thermal cycling and exposure to nuclear fission neutrons. As the IL has low thermal conductivity, they may cause structural failure in the fuel element during operation. The present study provides a new preparation technique for interdiffusion pairs made by hot rolling. The U-Mo alloy, in tablet format, is involved by matrix Al-plates, which is sealed and then hot rolled. This way to prepare the diffusion couples is an ideal condition to avoid the oxidation at the contact interface at U-Mo/Al. The hot rolling preparation also simulates the first reduction pass during MTR fuel plate manufacture. We chose to work with a Mo content of 10 wt% in U-Mo alloy to ensure greater phase formation, since this level favors a greater chemical stability in this phase. The Al alloy matrix was used as the AA1050 since it contains small impurity amounts. The interdiffusion couples U-10Mo/AA1050 were thermally treated in two temperature ranges (1500C and 5500C) and three soaking times (5h, 40h and 80h) to simulate the interdiffusion process and formation of chemical interaction layer. The analysis of the interaction layer U-10Mo/AA1050 was made by SEM/EDS and X-ray diffraction. It revealed a general trend of low interdiffusion of Al (about 8 atomic %) inside U-Mo. There was

A modelling study of the inter-diffusion layer formation in U-Mo/Al dispersion fuel plates at high power

Science.gov (United States)

Ye, B.; Hofman, G. L.; Leenaers, A.; Bergeron, A.; Kuzminov, V.; Van den Berghe, S.; Kim, Y. S.; Wallin, H.

2018-02-01

Post irradiation examinations of full-size U-Mo/Al dispersion fuel plates fabricated with ZrN- or Si- coated U-Mo particles revealed that the reaction rate of irradiation-induced U-Mo-Al inter-diffusion, an important microstructural change impacting the performance of this type of fuel, transited at a threshold temperature/fission rate. The existing inter-diffusion layer (IL) growth correlation, which does not describe the transition behavior of IL growth, was modified by applying a temperature-dependent multiplication factor that transits around a threshold fission rate. In-pile irradiation data from four tests in the BR2 reactors, including FUTURE, E-FUTURE, SELEMIUM, and SELEMIUM-1a, were utilized to determine and validate the updated IL growth correlation. Irradiation behavior of the plates was simulated with the DART-2D computational code. The general agreement between the calculated and measured fuel meat swelling and constituent volume fractions as a function of fission density demonstrated the plausibility of the updated IL growth correlation. The simulation results also suggested the temperature dependence of the IL growth rate, similar to the temperature dependence of the inter-mixing rate in ion-irradiated bi-layer systems.

EVALUATION OF U10MO FUEL PLATE IRRADIATION BEHAVIOR VIA NUMERICAL AND EXPERIMENTAL BENCHMARKING

Energy Technology Data Exchange (ETDEWEB)

Samuel J. Miller; Hakan Ozaltun

2012-11-01

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.

Numerical simulation research on rolling process of monolithic nuclear fuel plate

International Nuclear Information System (INIS)

Wan Jibo; Kong Xiangzhe; Ding Shurong; Xu Hongbin; Huo Yongzhong

2015-01-01

For the strain-rate-dependent constitutive relation of zircaloy cladding in UMo monolithic nuclear fuel plates, the three-dimensional stress updating algorithm was derived out, and the corresponding VUMAT subroutine to define its constitutive relation was developed and validated; the finite element model was built to simulate the frame rolling process of UMo monolithic nuclear fuel plates; with the explicit dynamic finite element method, the evolution rules of the deformation and contact pressure during the rolling process within the composite slab were obtained and analyzed. The research results indicate that it is convenient and efficient to define the strain-rate- dependent constitutive relations of materials with the user-defined material subroutine VUMAT; the rolling-induced contact pressure between the fuel meat and the covers varies with time, and the maximum pressure exits at the symmetric plane along the plate width direction. This study supplies a foundation and a computation method for optimizing the processing parameters to manufacture UMo monolithic nuclear fuel plates. (authors)

TEM investigation of irradiated U-7 weight percent Mo dispersion fuel

International Nuclear Information System (INIS)

Van den Berghe, S.

2009-01-01

In the FUTURE experiment, fuel plates containing U-7 weight percent Mo atomized powder were irradiated in the BR2 reactor. At a burn-up of approximately 33 percent 235 U (6.5 percent FIMA or 1.41 10 21 fissions/cm 3 meat), the fuel plates showed an important deformation and the irradiation was stopped. The plates were submitted to detailed PIE at the Laboratory for High and Medium level Activity. The results of these examinations were reported in the scientific report of last year and published in open literature. Since then, the microstructural aspects of the FUTURE fuel were studied in more detail using transmission electron microscopy (TEM), in an attempt to understand the nature of the interaction phase and the fission gas behavior in the atomized U(Mo) fuel. The FUTURE experiment is regarded as the definitive proof that the classical atomized U(Mo) dispersion fuel is not stable under irradiation, at least in the conditions required for normal operation of plate-type fuel. The main cause for the instability was identified to be the irradiation behavior of the U(Mo)-Al interaction phase which is formed between the U(Mo) particles and the pure aluminum matrix during irradiation. It is assumed to become amorphous under irradiation and as such cannot retain the fission gas in stable bubbles. As a consequence, gas filled voids are generated between the interaction layer and the matrix, resulting in fuel plate pillowing and failure. The objective of the TEM investigation was the confirmation of this assumption of the amorphisation of the interaction phase. A deeper understanding of the actual nature of this layer and the fission gas behaviour in these fuels in general can allow a more oriented search for a solution to the fuel failures

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

Science.gov (United States)

Keiser, Dennis D.; Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W.; Ross Finlay, M.; Moore, Glenn; Medvedev, Pavel; Meyer, Mitch

2017-05-01

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

Characterization of intergranular fission gas bubbles in U-Mo fuel

International Nuclear Information System (INIS)

Kim, Y. S.; Hofman, G.; Rest, J.; Shevlyakov, G. V.

2008-01-01

This report can be divided into two parts: the first part, which is composed of sections 1, 2, and 3, is devoted to report the analyses of fission gas bubbles; the second part, which is in section 4, is allocated to describe the mechanistic model development. Swelling data of irradiated U-Mo alloy typically show that the kinetics of fission gas bubbles is composed of two different rates: lower initially and higher later. The transition corresponds to a burnup of ∼0 at% U-235 (LEU) or a fission density of ∼3 x 10 21 fissions/cm 3 . Scanning electron microscopy (SEM) shows that gas bubbles appear only on the grain boundaries in the pretransition regime. At intermediate burnup where the transition begins, gas bubbles are observed to spread into the intragranular regions. At high burnup, they are uniformly distributed throughout fuel. In highly irradiated U-Mo alloy fuel large-scale gas bubbles form on some fuel particle peripheries. In some cases, these bubbles appear to be interconnected and occupy the interface region between fuel and the aluminum matrix for dispersion fuel, and fuel and cladding for monolithic fuel, respectively. This is a potential performance limit for U-Mo alloy fuel. Microscopic characterization of the evolution of fission gas bubbles is necessary to understand the underlying phenomena of the macroscopic behavior of fission gas swelling that can lead to a counter measure to potential performance limit. The microscopic characterization data, particularly in the pre-transition regime, can also be used in developing a mechanistic model that predicts fission gas bubble behavior as a function of burnup and helps identify critical physical properties for the future tests. Analyses of grain and grain boundary morphology were performed. Optical micrographs and scanning electron micrographs of irradiated fuel from RERTR-1, 2, 3 and 5 tests were used. Micrographic comparisons between as-fabricated and as-irradiated fuel revealed that the site of

Scanning electron microscopy analysis of fuel/matrix interaction layers in highly-irradiated U-Mo dispersion fuel plates with Al and Al-Si alloy matrices

Energy Technology Data Exchange (ETDEWEB)

Keiser, Dennis D. Jr; Jue, Jan Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adom B.; Medvedev, Pavel; Madden, James; Wachs, Dan; Meyer, Mitch [Nuclear Fuels and Materials Division, Idaho National Laboratory (United States)

2014-04-15

In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U-7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifically, samples from irradiated U-7Mo dispersion fuel elements with pure Al, Al-2Si and AA4043 (-4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U-7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U-7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al-Si matrices.

Numerical aspects of U-Mo core covered by Zry-4 miniplates co-rolling

International Nuclear Information System (INIS)

Picchetti, B.; Moscarda, M.V.; Taboada, H.

2013-01-01

The aim of this work is to support through adequate modeling the development of the co-rolling process of miniplates and plates starting with compacts including a monolithic U-Mo core with Zry-4 frame and cladding, Through relevant parameter identification and specific variables calculation a co rolling process model was set. The goal is to design a co-rolling optimal strategy related to the expected results through the use of such model. To that end the rolling process is depicted and some elements of strain stress theory on metals are employed. Plastic strain depends on deviator components of the stress tensor but no on the hydrostatic one. Metal sheet co-rolling is a plastic strain by planar compression at constant volume. During the co-rolling process the width constancy is assumed, being the piece of metal free to flow along its length. In this work the relationship between constitutive materials shield stresses U-Mo core and Zry-4 cladding under T= 650°C co-rolling is determined. This allows to modeling the reduction that exist in each co-rolling step for each one of phases present, which enables the design of a loop control lace optimizing the co rolling process. (author)

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

Energy Technology Data Exchange (ETDEWEB)

Keiser, Dennis D., E-mail: dennis.keiser@inl.gov [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States); Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States); Ross Finlay, M. [Australian Nuclear Science and Technology Organization, PMB 1, Menai, NSW 2234 (Australia); Moore, Glenn; Medvedev, Pavel; Meyer, Mitch [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States)

2017-05-15

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

Science.gov (United States)

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

2016-06-01

Low-enriched (U-235 RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

Progress in qualifying low-enriched U-Mo dispersion fuels

International Nuclear Information System (INIS)

Snelgrove, J.L.; Hofman, G.L.; Hayes, S.L.; Meyer, M.K.

2001-01-01

The U.S. Reduced Enrichment for Research and Test Reactors program is working to qualify dispersions of U-Mo alloys in aluminum with fuel-meat densities of 8 to 9 gU cm -3 . Post irradiation examinations of the small fuel plates irradiated in the Advanced Test Reactor during the high-temperature RERTR-3 tests are virtually complete, and analysis of the large quantity of data obtained is underway. We have observed that the swelling of the fuel plates is stable and modest and that the swelling is dominated by the temperature-dependent interaction of the U-Mo fuel and the aluminum matrix. In order to extract detailed information about the behavior of these fuels from the data, a complex fuel-plate thermal model is being developed to account for the effects of the changing fission rate and thermal conductivity of the fuel meat during irradiation. This paper summarizes the empirical results of the post irradiation examinations and the preliminary results of the model development. In addition, the schedule for irradiation of full-sized elements in the HFR-Petten is briefly discussed. (author)

Irradiation of diffusion couples U-Mo/Al. Thermal calculation

International Nuclear Information System (INIS)

Fortis, Ana M.; Mirandou, Monica; Denis, Alicia C.

2004-01-01

The development of new low enrichment fuel elements for research reactors has lead to obtaining a number of compounds and alloys where the decrease in the enrichment is compensated by a higher uranium density in the fuel material. This has been achieved in particular with the uranium silicides dispersed in an aluminum matrix, where uranium densities about 4.8 g/cm 3 have been reached. Among the diverse candidate alloys, those of U-Mo with molybdenum content in the range 6 to 10 w % can yield, upon dispersion, to uranium densities of about 8 g/cm 3 . The first irradiation experiments employing these alloys in fuel plates, either dispersed in Al or monolithic revealed certain phenomena which are worthy of further studies. Failures have been detected apparently due to the formation of reaction products between the fissile material and the aluminum matrix, which exhibit a poor irradiation behavior. An experiment was designed which final purpose is to irradiate diffusion couples U-Mo/Al in the RA-3 reactor and to analyze the interaction zone at the working temperatures of the fuel elements. A simple device was built consisting of two Al 6063 blocks which press the U-Mo sample in between, located in an Al capsule. The ensemble is placed in a tube, which can be filled with different gases and introduced in the reactor. For safety reasons temperature predictions are necessary before performing the experiment. To this end, the COSMOS code was used. As a preliminary step and in order to test to exactness of the numerical estimations, two irradiations were performed in the RA-1 reactor with He and N 2 as transference gases. The agreement between the measured and calculated temperatures was good, particularly in the case of He and, along with the numerical predictions for the RA-3 reactor, provides a reliable basis to proceed with the following steps. (author)

Evaluation of plate type fuel options for small power reactors; Avaliacao de alternativas de combustivel tipo placa para reatores de pequeno porte

Energy Technology Data Exchange (ETDEWEB)

Andrzejewski, Claudio de Sa

2005-07-01

Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO{sub 2} in stainless steel, of UO{sub 2} in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

Energy Technology Data Exchange (ETDEWEB)

Jian Gan; Brandon Miller; Dennis Keiser; Adam Robinson; James Madden; Pavel Medvedev; Daniel Wachs

2014-04-01

As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists of fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.

Transmission electron microscopy characterization of irradiated U-7Mo/Al-2Si dispersion fuel

International Nuclear Information System (INIS)

Gan, J.; Keiser, D.D.; Wachs, D.M.; Robinson, A.B.; Miller, B.D.; Allen, T.R.

2010-01-01

The plate-type dispersion fuels, with the atomized U(Mo) fuel particles dispersed in the Al or Al alloy matrix, are being developed for use in research and test reactors worldwide. It is found that the irradiation performance of a plate-type dispersion fuel depends on the radiation stability of the various phases in a fuel plate. Transmission electron microscopy was performed on a sample (peak fuel mid-plane temperature ∼109 deg. C and fission density ∼4.5 x 10 27 f m -3 ) taken from an irradiated U-7Mo dispersion fuel plate with Al-2Si alloy matrix to investigate the role of Si addition in the matrix on the radiation stability of the phase(s) in the U-7Mo fuel/matrix interaction layer. A similar interaction layer that forms in irradiated U-7Mo dispersion fuels with pure Al matrix has been found to exhibit poor irradiation stability, likely as a result of poor fission gas retention. The interaction layer for both U-7Mo/Al-2Si and U-7Mo/Al fuels is observed to be amorphous. However, unlike the latter, the amorphous layer for the former was found to effectively retain fission gases in areas with high Si concentration. When the Si concentration becomes relatively low, the fission gas bubbles agglomerate into fewer large pores. Within the U-7Mo fuel particles, a bubble superlattice ordered as fcc structure and oriented parallel to the bcc metal lattice was observed where the average bubble size and the superlattice constant are 3.5 nm and 11.5 nm, respectively. The estimated fission gas inventory in the bubble superlattice correlates well with the fission density in the fuel.

Characterization of the Microstructure of Irradiated U-Mo Dispersion Fuel with a Matrix that Contains Si

International Nuclear Information System (INIS)

Keiser, Jr. D.D.; Robinson, A.B.; Jue, J.F.; Medvedev, P.; Finlay, M.R.

2009-01-01

RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. Of particular interest is the irradiation performance of U-Mo dispersion fuels with Si added to the Al matrix. Si is added to improve the performance of U-Mo dispersion fuels. Microstructural examinations have been performed on fuel plates with Al-2Si matrix after irradiation to around 50% LEU burnup. Si-rich layers were observed in many areas around the various U-7Mo fuel particles. In one local area of one of the samples, where the Si-rich layer had developed into a layer devoid of Si, relatively large fission gas bubbles were observed in the interaction phase. There may be a connection between the growth of these bubbles and the amount of Si present in the interaction layer. Overall, it was found that having Si-rich layers around the fuel particles after fuel plate fabrication positively impacted the overall performance of the fuel plate

Mechanical properties examined by nanoindentation for selected phases relevant to the development of monolithic uranium-molybdenum metallic fuels

Energy Technology Data Exchange (ETDEWEB)

Newell, Ryan; Park, Youngjoo; Mehta, Abhishek [Department of Materials Science and Engineering, University of Central Florida, Orlando, FL, 32826 (United States); Keiser, Dennis [Nuclear Fuels and Materials Division, Idaho National Laboratory, Idaho Falls, ID, 83402 (United States); Sohn, Yongho, E-mail: Yongho.Sohn@ucf.edu [Department of Materials Science and Engineering, University of Central Florida, Orlando, FL, 32826 (United States)

2017-04-15

Nanomechanical properties, specifically the reduced modulus and hardness of several intermetallic and solid solution phases are reported to assist the development of the U-10 wt% Mo (U-10Mo) monolithic fuel system for research and test reactors. Findings from this study and reported values of mechanical properties provide data critical for understanding and predicting the structural behavior of the fuel system during fabrication and irradiation. The phases examined are products of interdiffusion and reaction between (1) the AA6061 cladding and the Zr diffusion barrier, namely (Al,Si){sub 3}Zr and Al{sub 3}Zr, (2) the U-10Mo fuel and the Zr diffusion barrier, namely UZr{sub 2}, Mo{sub 2}Zr, and α-U, and (3) the U (or U-10Mo) and Mo, namely a mixture gradient of α- and γ-phases. The UC inclusions observed within the fuel alloy were also examined. Only phases present in thick or continuous microstructure on cross-sectioned fuel plates and diffusion couples were investigated for reduced modulus and hardness. Concentration-dependence of room-temperature reduced modulus in U solid solution with 0–10 wt% Mo was semi-quantitatively modeled based on mixture of α- and γ-phases and solid solutioning within the γ-phase.

Mechanical properties examined by nanoindentation for selected phases relevant to the development of monolithic uranium-molybdenum metallic fuels

Science.gov (United States)

Newell, Ryan; Park, Youngjoo; Mehta, Abhishek; Keiser, Dennis; Sohn, Yongho

2017-04-01

Nanomechanical properties, specifically the reduced modulus and hardness of several intermetallic and solid solution phases are reported to assist the development of the U-10 wt% Mo (U-10Mo) monolithic fuel system for research and test reactors. Findings from this study and reported values of mechanical properties provide data critical for understanding and predicting the structural behavior of the fuel system during fabrication and irradiation. The phases examined are products of interdiffusion and reaction between (1) the AA6061 cladding and the Zr diffusion barrier, namely (Al,Si)3Zr and Al3Zr, (2) the U-10Mo fuel and the Zr diffusion barrier, namely UZr2, Mo2Zr, and α-U, and (3) the U (or U-10Mo) and Mo, namely a mixture gradient of α- and γ-phases. The UC inclusions observed within the fuel alloy were also examined. Only phases present in thick or continuous microstructure on cross-sectioned fuel plates and diffusion couples were investigated for reduced modulus and hardness. Concentration-dependence of room-temperature reduced modulus in U solid solution with 0-10 wt% Mo was semi-quantitatively modeled based on mixture of α- and γ-phases and solid solutioning within the γ-phase.

Elevated Temperature Tensile Tests on DU–10Mo Rolled Foils

Energy Technology Data Exchange (ETDEWEB)

Schulthess, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-09-01

Tensile mechanical properties for uranium-10 wt.% molybdenum (U–10Mo) foils are required to support modeling and qualification of new monolithic fuel plate designs. It is expected that depleted uranium-10 wt% Mo (DU–10Mo) mechanical behavior is representative of the low enriched U–10Mo to be used in the actual fuel plates, therefore DU-10Mo was studied to simplify material processing, handling, and testing requirements. In this report, tensile testing of DU-10Mo fuel foils prepared using four different thermomechanical processing treatments were conducted to assess the impact of foil fabrication history on resultant tensile properties.

Study of the feasibility of friction STIR welding applied to the fabrication of monolithic fuel elements

International Nuclear Information System (INIS)

Cabot, Pedro J.; Moglioni, A.; Mirandou, Marcela; Balart, Silvia N.

2004-01-01

The monolithic U-Mo fuel elements consist in a foil of a U-Mo alloy encased in Al. One of the techniques that is being tried to apply in their fabrication is Friction Stir Welding in the 'no contact at the interface' mode. The Laboratory of Welding at the National Atomic Energy Commission (Argentina) has a great experience in the conventional form of this technique so has started working on this new application. This paper describes the experiments performed to obtain the operative parameters. In the first experiments AA6061 T6 (Al) plates and sheets of AISI 316 (SS) were used to obtain the optimal operative parameters of the process. Welds were performed and evaluated for different operative variables such speed, angle and diameter of the tool and tool-interface gap keeping the rotation speed constant. Tensile test, pressure leak-proof test, bending test, non-destructive test and metallography were used to characterize the welds. Finally, SS and U-Mo foils were encased using the parameters selected from the first experiments. The samples prepared with U-Mo alloy will be used as diffusion couples and for the studies of interdiffusion under irradiation. (author)

Pore growth in U-Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Jeong, G.Y.; Sohn, D.-S. [Ulsan National Institute of Science and Technology, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan, 689-798 (Korea, Republic of); Jamison, L.M. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2016-09-15

U-Mo/Al dispersion fuel is currently under development in the DOE’s Material Management and Minimization program to convert HEU-fueled research reactors to LEU-fueled reactors. In some demanding conditions in high-power and high-performance reactors, large pores form in the interaction layers between the U-Mo fuel particles and the Al matrix, which pose a potential to cause fuel failure. In this study, comprehension of the formation and growth of these pores was explored. As a product, a model to predict pore growth and porosity increase was developed. The model includes three major topics: fission gas release from the U-Mo and the IL to the pores, stress evolution in the fuel meat, and the effect of amorphous IL growth. Well-characterized in-pile data from reduced-size plates were used to fit the model parameters. A data set from full-sized plates, independent and distinctively different from those used to fit the model parameters, was used to examine the accuracy of the model. The model showed fair agreement with the measured data. The model suggested that the growth of the IL has a critical effect on pore growth, as both its material properties and energetics are favorable to pore formation. Therefore, one area of the current effort, focused on suppressing IL growth, appears to be on the right track to improve the performance of this fuel.

Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation – Non-destructive analysis of the AFIP-1 fuel plates

Energy Technology Data Exchange (ETDEWEB)

Wachs, D.M., E-mail: daniel.wachs@inl.gov [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Robinson, A.B.; Rice, F.J. [Idaho National Laboratory, Characterization and Advanced PIE Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Kraft, N.C.; Taylor, S.C. [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Lillo, M. [Idaho National Laboratory, Nuclear Systems Design and Analysis Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Woolstenhulme, N.; Roth, G.A. [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2016-08-01

Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008–2009. The irradiation conditions were: ∼250 W/cm{sup 2} peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm{sup 3} peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

Interdiffusion among U-Mo-Zr and alloys of Al to 550oC

International Nuclear Information System (INIS)

Komar Varela, C.L; Arico, S.F; Gribaudo, L.M

2006-01-01

The international community, by means of the project 'Reduced Enrichment for Research and Test Reactors' is interested in the development of a new nuclear fuel of very high density of uranium and low enrichment (≤ 20% de U 235 ) for reactors of investigation and production of radioisotopes, that permit to reach greater neutron flows, with good capacity to be reprocessed One of these assemblies are the alloys of U with Mo contents between 7 and 10% in weight. In the fuels 'dispersed type plate' the particles of U-Mo are mixed with dust of aluminum and are co - laminated between two plates of an alloy of the same material. The existing contact among the particles permits the interdiffusion of the materials with the consequent apparition of new phases. Studies pursuit-irradiation have shown a badly behavior of these new phases. It is for this that is necessary to control the presence of these products of interaction. The aggregate of a third element to the alloys U - Mo has begun to be practiced with this purpose. In this work the modification of the start of the disorder of the phase γU in the alloy U-7%Mo-1%Zr was studied and the interdiffusion between pure aluminum and the same alloy to 550 o C. The results obtained are compared with other obtained for peers U-Mo/Al. The techniques of characterization utilized were: optical microscopy, analysis by diffraction of X-rays and microanalysis quantitative by microprobe electronic. It was observed that the aggregate of Zr refines the grain for a processing of homogenized in composition of Mo to 1000 o C and accelerates the start of the disorder of the phase γU to 550 o C. As for the zone of interaction, was found that the composed identifying do not they differ to them reported in the in peers U-Mo/Al. These are: (U,Mo)Al 4 y UAl 3 (AG)

Modelling of U-Mo/Al Dispersion fuel fission induced swelling and creep

Energy Technology Data Exchange (ETDEWEB)

Jeong, Gwan Yoon; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Argonne (United States)

2014-05-15

In a Dispersion fuel which U-Mo particles are dispersed in Al metal matrix, a similar phenomenon forming a bulge region was observed but it is difficult to quantify and construct a model for explaining creep and swelling because of its complex microstructure change during irradiation including interaction layer (IL) and porosity formation. In a Dispersion fuel meat, fission product induces fuel particles swelling and it has to be accommodated by the deformation of the Al matrix and newly formed IL during irradiation. Then, it is reasonable that stress from fuel swelling in the complex structure should be relaxed by local adjustments of particles, Al matrix, and IL. For analysis of U-Mo/Al Dispersion fuel creep, the creep of U-Mo particle, Al matrix, and IL should be considered. Moreover, not only fuel particle swelling and IL growth, but also fuel and Al matrix consumptions due to IL formation are accounted in terms of their volume fraction changes during irradiation. In this work, fuel particles, Al matrix and IL are treated in a way of homogenized constituents: Fuel particles, Al matrix and IL consist of an equivalent meat during irradiation. Meat volume swelling of two representative plates was measured: One (Plate A) was a pure Al matrix with 6g/cc uranium loading, the other (Plate B) a silicon added Al matrix with 8g/cc uranium loading. The meat swelling of calculated as a function of burnup. The meat swelling of calculation and measurement was compared and the creep rate coefficients for Al and IL were estimated by repetitions. Based on assumption that only the continuous phase of Al-IL combined matrix accommodated the stress from fuel particle swelling and it was allowed to have creep deformation, the homogenization modeling was performed. The meat swelling of two U-Mo/Al Dispersion fuel plates was modeled by using homogenization model.

Modelling of U-Mo/Al Dispersion fuel fission induced swelling and creep

International Nuclear Information System (INIS)

Jeong, Gwan Yoon; Sohn, Dong Seong; Kim, Yeon Soo

2014-01-01

In a Dispersion fuel which U-Mo particles are dispersed in Al metal matrix, a similar phenomenon forming a bulge region was observed but it is difficult to quantify and construct a model for explaining creep and swelling because of its complex microstructure change during irradiation including interaction layer (IL) and porosity formation. In a Dispersion fuel meat, fission product induces fuel particles swelling and it has to be accommodated by the deformation of the Al matrix and newly formed IL during irradiation. Then, it is reasonable that stress from fuel swelling in the complex structure should be relaxed by local adjustments of particles, Al matrix, and IL. For analysis of U-Mo/Al Dispersion fuel creep, the creep of U-Mo particle, Al matrix, and IL should be considered. Moreover, not only fuel particle swelling and IL growth, but also fuel and Al matrix consumptions due to IL formation are accounted in terms of their volume fraction changes during irradiation. In this work, fuel particles, Al matrix and IL are treated in a way of homogenized constituents: Fuel particles, Al matrix and IL consist of an equivalent meat during irradiation. Meat volume swelling of two representative plates was measured: One (Plate A) was a pure Al matrix with 6g/cc uranium loading, the other (Plate B) a silicon added Al matrix with 8g/cc uranium loading. The meat swelling of calculated as a function of burnup. The meat swelling of calculation and measurement was compared and the creep rate coefficients for Al and IL were estimated by repetitions. Based on assumption that only the continuous phase of Al-IL combined matrix accommodated the stress from fuel particle swelling and it was allowed to have creep deformation, the homogenization modeling was performed. The meat swelling of two U-Mo/Al Dispersion fuel plates was modeled by using homogenization model

Evidence for the presence of U-Mo-Al ternary compounds in the U-Mo/Al interaction layer grown by thermal annealing: a coupled micro X-ray diffraction and micro X-ray absorption spectroscopy study

International Nuclear Information System (INIS)

Palancher, H.; Martin, P.; Nassif, V.

2007-01-01

The systematic presence of the ternary phases U 6 Mo 4 Al 43 and UMo 2 Al 20 is reported in a U-Mo/Al interaction layer grown by thermal annealing. This work shows, therefore, the low Mo solubility in UAl 3 and UAl 4 binary phases; it contradicts the hypothesis of the formation of (U,Mo)Al 3 and (U,Mo)Al 4 solid solutions often admitted in the literature. Using μ-XAS (micro X-ray absorption spectroscopy) at the Mo K edge and μ-XRD (micro X-ray diffraction), the heterogeneity of the interaction layer obtained on a γ-U 0.85 Mo 0.15 /Al diffusion couple has been precisely investigated. The UMo 2 Al 20 phase has been identified at the closest location from the Al side. Moreover, μ-XRD mapping performed on an annealed fuel plate enabled the characterization of the four phases resulting from the γ-U 0.85 Mo 0.15 /Al and (U 2 Mo+α-U)/Al interactions. A strong correlation between the concentrations of UAl 4 and UMo 2 Al 20 and those of UAl 3 and U 6 Mo 4 Al 43 has been shown. (orig.)

Safety assessment of U–Mo fuel mini plates irradiated in HANARO reactor

International Nuclear Information System (INIS)

Jo, Daeseong; Kim, Haksung

2015-01-01

Highlights: • Neutronic and thermal-hydraulic analyses of U–Mo fuel irradiated in HANARO reactor. • A mock-up irradiation target was designed and tested to measure the flow rate. • During normal operation, boiling does not occur. • During limiting accidents, boiling occurs. However, fuel integrity is maintained. - Abstract: Neutronic and thermal hydraulic characteristics of U–Mo fuel mini plates irradiated in the HANARO reactor were analyzed for the safety assessment of these plates. A total of eight fuel plates were double-stacked; each stack contained three 8.0 gU/cc U–7Mo fuel plates and one 6.5 gU/cc U–7Mo fuel plate. The neutronic and thermal hydraulic analyses were carried out using the MCNP code and TMAP code, respectively. The core status used in the study was the equilibrium core, and four Control Absorber Rod (CAR) locations were considered: 350 mm, 450 mm, 550 mm, and 650 mm away from the bottom of the core. For the fuels in the lower stack, the maximum heat flux was found at the CAR located at 450 mm. For the fuels in the upper stack, the maximum heat flux was found at the CAR located at 650 mm. The axial power distributions for the upper and lower stacks were selected on the basis of thermal margin analyses. A mock-up irradiation target assembly was designed and tested at the out-of-pile test facility to measure the flow rate through the irradiation site, given that the maximum flow rate through the irradiation site at the HANARO reactor is limited to 12.7 kg/s. For conservative analyses, measurement and correlation uncertainties and engineering hot channel factors were considered. During normal operation, the minimum ONB temperature margins for the lower and upper stacks are 41.6 °C and 31.8 °C, respectively. This means that boiling does not occur. However, boiling occurs during the limiting accidents. Nevertheless, the fuel integrity is maintained since the minimum DNBR are 1.96 for the Reactivity Insertion Accident (RIA) and 2

Interdiffusion between U(Mo,Pt) or U(Mo,Zr) and Al or Al A356 alloy

International Nuclear Information System (INIS)

Komar Varela, C.; Mirandou, M.; Arico, S.; Balart, S.; Gribaudo, L.

2009-01-01

Solid state reactions in chemical diffusion couples U-7 wt.%Mo-0.9 wt.%Pt/Al at 580 deg. C and U-7 wt.%Mo-0.9 wt.%Pt/Al A356 alloy, U-7 wt.%Mo-1 wt.%Zr/Al and U-7 wt.%Mo-1 wt.%Zr/Al A356 alloy at 550 deg. C were characterized. Results were obtained from optical and scanning electron microscopy, electron probe microanalysis and X-ray diffraction. The UAl 3, UAl 4 and Al 20 Mo 2 U phases were identified in the interaction layers of γU(Mo,Pt)/Al and γU(Mo,Zr)/Al diffusion couples. Al 43 Mo 4 U 6 ternary compound was also identified in γU(Mo,Zr)/Al due to the decomposition of γU(Mo,Zr) phase. The U(Al,Si) 3 and U 3 Si 5 phases were identified in the interaction layers of γU(Mo,Pt)/Al A356 and γU(Mo,Zr)/Al A356 diffusion couples. These phases are formed due to the migration of Si to the interaction layer. In the diffusion couple U(Mo,Zr)/Al A356, Zr 5 Al 3 phase was also identified in the interaction layer. The use of synchrotron radiation at Brazilian Synchrotron Light Laboratory (LNLS, CNPq, Campinas, Brazil) was necessary to achieve a complete crystallographic characterization.

U.S. progress in the development of very high density low enrichment research reactor fuels

International Nuclear Information System (INIS)

Meyer, M. K.; Wachs, D. M.; Jue, J.-F.; Keiser, D. D.; Gan, J.; Rice, F.; Robinson, A.; Woolstenhulme, N. E.; Medvedev, P.; Hofman, G. L.; Kim, Y.-S.

2012-01-01

The effort to develop low-enriched fuels for high power research reactors began world-wide in 1996. Since that time, hundreds of fuel specimens have been tested to investigate the operational limits of many variations of U-Mo alloy dispersion and monolithic fuels. In the U.S., the fuel development program has focused on the development of monolithic fuel, and is currently transitioning from conducting research experiments to the demonstration of large scale, prototypic element assemblies. These larger scale, integral fuel performance demonstrations include the AFIP-7 test of full-sized, curved plates configured as an element, the RERTR-FE irradiation of hybrid fuel elements in the Advanced Test Reactor, reactor specific Design Demonstration Experiments, and a multi-element Base Fuel Demonstration. These tests are conducted alongside mini-plate tests designed to prove fuel stability over a wide range of operating conditions. Along with irradiation testing, work on collecting data on fuel plate mechanical integrity, thermal conductivity, fission product release, and microstructural stability is underway. (authors)

TEM characterization of irradiated U-7Mo/Mg dispersion fuel

Science.gov (United States)

Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J. F.; Robinson, A. B.; Madden, J.

2017-10-01

This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 1021 f/cm3, 7.4 × 1014 f/cm3/s and 123 °C, and 5.5 × 1021 f/cm3, 11.0 × 1014 f/cm3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al3Mg2 and Al12Mg17 along with precipitates of MgO, Mg2Si and FeAl5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.

Interdiffusion and reactions between U-Mo and Zr at 650 °C as a function of time

Science.gov (United States)

Park, Y.; Keiser, D. D.; Sohn, Y. H.

2015-01-01

Development of monolithic U-Mo alloy fuel (typically U-10 wt.%Mo) for the Reduced Enrichment for Research and Test Reactors (RERTR) program entails a use of Zr diffusion barrier to eliminate the interdiffusion-reactions between the fuel alloy and Al-alloy cladding. The application of Zr barrier to the U-Mo fuel system requires a co-rolling process that utilizes a soaking temperature of 650 °C, which represents the highest temperature the fuel system is exposed to during both fuel manufacturing and reactor application. Therefore, in this study, development of phase constituents, microstructure and diffusion kinetics of U-10 wt.%Mo and Zr was examined using solid-to-solid diffusion couples annealed at 650 °C for 240, 480 and 720 h. Phase constituents and microstructural development were analyzed by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Concentration profiles were mapped as diffusion paths on the isothermal ternary phase diagram. Within the diffusion zone, single-phase layers of β-Zr and β-U were observed along with a discontinuous layer of Mo2Zr between the β-Zr and β-U layers. In the vicinity of Mo2Zr phase, islands of α-Zr phases were also found. In addition, acicular α-Zr and U6Zr3Mo phases were observed within the γ-U(Mo) terminal alloy. Growth rate of the interdiffusion-reaction zone was determined to be 7.75 (± 5.84) × 10-16 m2/s at 650 °C, however with an assumption of a certain incubation period.

Reaction layer between U-7WT%Mo and Al alloys in chemical diffusion couples

International Nuclear Information System (INIS)

Mirandou, M.; Granovsky, M.; Ortiz, M.; Balart, S.; Arico, S.; Gribaudo, L.

2005-01-01

Several failures in U-Mo dispersion fuel plates like pillowing and large porosities have been reported during irradiation experiments. These failures have been assigned to the formation of a large (U-Mo)/Al interaction product under high operating conditions. The modification of the matrix by alloying Al to change the interaction layer and improve its irradiation behavior, has been proposed. This paper reports diffusion experiments performed between U-7wt%Mo and various Al alloys containing Mg and / or Si. By the use of Optical Microscopy, SEM and X-Ray diffraction, it was found that with a concentration of 5.2wt% or 7.1 wt%Si the interaction layer is constituted mainly by (U,Mo)(Si,Al) 3 and no (U,Mo)Al 4 is detected. As part of the studies of properties of the U-Mo alloys the time for isothermal transformation start at different temperatures of the γ phase is being evaluated for the present U-7wt%Mo alloy. These results are used to plan the future diffusion program that will include diffusion under irradiation at CNEA RA3 reactor. (author)

Origin and development of the new U-Mo nuclear fuel

International Nuclear Information System (INIS)

Boyard, M.; Languille, A.; Thomasson, J.; Hamy, J.M.

2002-01-01

Historically most research reactors have used highly enriched nuclear fuels (enrichment > 90 %). Since 1977 the non-proliferation policy has imposed to convert these reactors to far less enriched fuels (< 20 %). An international consensus has evolved towards a nuclear fuel with an enrichment factor of 19,75 %, this fuel is made of a powdered U-Mo alloy scattered in an aluminium die. The external dimensions and the cladding materials of the fuel plate are unchanged in order to minimize development and qualification costs. The U-Mo fuel is expected to maintain or even to increase the performance of reactors and to allow the processing of spent fuels in the same installations as those used for fuels issuing from power plants. Cea, Cogema, Cerca, Framatome, and Technicatome have shared their technical means, their know-how and their financial resources to develop this new nuclear fuel. 2006 is the contract date by which American authorities will stop repatriating the ancient spent fuel (uranium silicide) from research reactors so it is imperative to make available by this date a new nuclear fuel with a satisfactory end of cycle. This article also presents the French program of qualification of the U-Mo fuel. 2 series of irradiation have already been performed, one (Isis-1) in Osiris reactor (Saclay, France) and the second (Umus) in HFR (Petten, Netherlands). A clad failure has led to stop the Umus experiment. 2 new series of irradiation are scheduled to start in 2002. In a parallel way, in the framework of the design of the RJH (Jules Horowitz reactor) Cea will soon perform irradiation of U-Mo fuel plates in BR2 (Mol, Belgium). (A.C.)

Rupture of Al matrix in U-Mo/Al dispersion fuel by fission induced creep

Energy Technology Data Exchange (ETDEWEB)

Jeong, Gwan Yoon; Sohn, Dong Seong [UNIST, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Argonnge (United States); Lee, Kyu Hong [KAERI, Daejeon (Korea, Republic of)

2016-05-15

This phenomenon was found specifically in the dispersion fuel plate with Si addition in the Al matrix to suppress interaction layer (IL) formation between UMo and Al. It is known that the stresses induced by fission induced swelling in U-Mo fuel particles are relieved by creep deformation of the IL, surrounding the fuel particles, that has a much higher creep rate than the Al matrix. Thus, when IL growth is suppressed, the stress is instead exerted on the Al matrix. The observed rupture in the Al matrix is believed to be caused when the stress exceeded the rupture strength of the Al matrix. In this study, the possibility of creep rupture of the Al matrix between the neighboring U-Mo fuel particles was examined using the ABAQUS finite element analysis (FEA) tool. The predicted rupture time for a plate was much shorter than its irradiation life indicating a rupture during the irradiation. The higher stress leads Al matrix to early creep rupture in this plate for which the Al matrix with lower creep strain rate does not effectively relieve the stress caused by the swelling of the U-Mo fuel particles. For the other plate, no rupture was predicted for the given irradiation condition. The effect of creeping of the continuous phase on the state of stress is significant.

Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

Energy Technology Data Exchange (ETDEWEB)

Jue, Jan-Fong, E-mail: dennis.keiser@inl.gov; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

2014-05-01

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative is developing an advanced monolithic fuel to convert US high-performance research reactors to low-enriched uranium. Hot-isostatic-press (HIP) bonding was the single process down-selected to bond monolithic U–Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between the fuel meat, the cladding, and the diffusion barrier, as well as between the U–10Mo fuel meat and the Al-6061 cladding, were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are: • A typical Zr diffusion barrier with a thickness of 25 μm. • A transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 μm. • Chemical banding, in some areas more than 100 μm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13 wt.%. • Decomposed areas containing plate-shaped low-Mo phase. • A typical Zr/cladding interaction layer with a thickness of 1–2 μm. • A visible UZr{sub 2} bearing layer with a thickness of 1–2 μm. • Mo-rich precipitates (mainly Mo{sub 2}Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr{sub 2}-bearing layer and the U–Mo matrix. • No excessive interaction between cladding and the uncoated fuel edge. • Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O

Detailed measurements of local thickness changes for U-7Mo dispersion fuel plates with Al-3.5Si matrix after irradiation at different powers in the RERTR-9B experiment

Science.gov (United States)

Keiser, Dennis D.; Williams, Walter; Robinson, Adam; Wachs, Dan; Moore, Glenn; Crawford, Doug

2017-10-01

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. Swelling is an important irradiation behavior that needs to be well understood. Data from high resolution thickness measurements performed on U-7Mo dispersion fuel plates with Al-Si alloy matrices that were irradiated at high power is sparse. This paper reports the results of detailed thickness measurements performed on two dispersion fuel plates that were irradiated at relatively high power to high fission densities in the Advanced Test Reactor in the same RERTR-9B experiment. Both plates were irradiated to similar fission densities, but one was irradiated at a higher power than the other. The goal of this work is to identify any differences in the swelling behavior when fuel plates are irradiated at different powers to the same fission densities. Based on the results of detailed thickness measurments, more swelling occurs when a U-7Mo dispersion fuel with Al-3.5Si matrix is irradiated to a high fission density at high power compared to one irradiated at a lower power to high fission density.

Modeling Thermal and Stress Behavior of the Fuel-clad Interface in Monolithic Fuel Mini-plates

International Nuclear Information System (INIS)

Miller, Gregory K.; Medvedev, Pavel G.; Burkes, Douglas E.; Wachs, Daniel M.

2010-01-01

As part of the Global Threat Reduction Initiative, a fuel development and qualification program is in process with the objective of qualifying very high density low enriched uranium fuel that will enable the conversion of high performance research reactors with operational requirements beyond those supported with currently available low enriched uranium fuels. The high density of the fuel is achieved by replacing the fuel meat with a single monolithic low enriched uranium-molybdenum fuel foil. Doing so creates differences in the mechanical and structural characteristics of the fuel plate because of the planar interface created by the fuel foil and cladding. Furthermore, the monolithic fuel meat will dominate the structural properties of the fuel plate rather than the aluminum matrix, which is characteristic of dispersion fuel types. Understanding the integrity and behavior of the fuel-clad interface during irradiation is of great importance for qualification of the new fuel, but can be somewhat challenging to determine with a single technique. Efforts aimed at addressing this problem are underway within the fuel development and qualification program, comprised of modeling, as-fabricated plate characterization, and post-irradiation examination. An initial finite element analysis model has been developed to investigate worst-case scenarios for the basic monolithic fuel plate structure, using typical mini-plate irradiation conditions in the Advanced Test Reactor. Initial analysis shows that the stress normal to the fuel-clad interface dominates during irradiation, and that the presence of small, rounded delaminations at the interface is not of great concern. However, larger and/or fuel-clad delaminations with sharp corners can create areas of concern, as maximum principal cladding stress, strain, displacement, and peak fuel temperature are all significantly increased. Furthermore, stresses resulting from temperature gradients that cause the plate to bow or buckle in

Small-scale Specimen Testing of Monolithic U-Mo Fuel Foils

Energy Technology Data Exchange (ETDEWEB)

Ramprashad Prabhakaran; Douglas E. Burkes; James I. Cole; Indrajit Charit; Daniel M. Wachs

2008-10-01

The objective of this investigation is to develop a shear punch testing (SPT) procedure and standardize it to evaluate the mechanical properties of irradiated fuels in a hot-cell so that the tensile behavior can be predicted using small volumes of material and at greatly reduced irradiation costs. This is highly important in the development of low-enriched uranium fuels for nuclear research and test reactors. The load-displacement data obtained using SPT can be interpreted in terms of and correlated with uniaxial mechanical properties. In order to establish a correlation between SPT and tensile data, sub-size tensile and microhardness testing were performed on U-Mo alloys. In addition, efforts are ongoing to understand the effect of test parameters (such as specimen thickness, surface finish, punch-die clearance, crosshead velocity and carbon content) on the measured mechanical properties, in order to rationalize the technique, prior to employing it on a material of unknown strength.

Small-scale Specimen Testing of Monolithic U-Mo Fuel Foils

International Nuclear Information System (INIS)

Ramprashad Prabhakaran; Douglas E. Burkes; James I. Cole; Indrajit Charit; Daniel M. Wachs

2008-01-01

The objective of this investigation is to develop a shear punch testing (SPT) procedure and standardize it to evaluate the mechanical properties of irradiated fuels in a hot-cell so that the tensile behavior can be predicted using small volumes of material and at greatly reduced irradiation costs. This is highly important in the development of low-enriched uranium fuels for nuclear research and test reactors. The load-displacement data obtained using SPT can be interpreted in terms of and correlated with uniaxial mechanical properties. In order to establish a correlation between SPT and tensile data, sub-size tensile and microhardness testing were performed on U-Mo alloys. In addition, efforts are ongoing to understand the effect of test parameters (such as specimen thickness, surface finish, punch-die clearance, crosshead velocity and carbon content) on the measured mechanical properties, in order to rationalize the technique, prior to employing it on a material of unknown strength

Observation on the irradiation behavior of U-Mo alloy dispersion fuel

International Nuclear Information System (INIS)

Hofman, Gerard L.; Meyer, Mitchell K.; Park, Jong-Man

2000-01-01

Initial results from the postirradiation examination of high-density dispersion fuel test RERTR-3 are discussed. The U-Mo alloy fuels in this test were irradiated to 40% U-235 burnup at temperature ranging from 140 0 C to 240 0 C. Temperature has a significant effect on overall swelling of the test plates. The magnitude of the swelling appears acceptable and no unstable irradiation behavior is evident. (author)

Post-irradiation analysis of low enriched U-Mo/Al dispersions fuel miniplate tests, RERTR 4 and 5

International Nuclear Information System (INIS)

Hofman, G.L.; Finlay, M.R.; Kim, Y.S.

2005-01-01

Interpretation of the post irradiation data of U-Mo/Al dispersion fuel mini plates irradiated in the Advanced Test Reactor to a maximum U-235 burn up of 80% are presented. The analyses addresses fuel swelling and porosity formation as these fuel performance issues relate to fuel fabrication and irradiation parameters. Specifically, mechanisms involved in the formation of porosity observed in the U-Mo/Al interaction phase are discussed and, means of mitigating or eliminating this irradiation phenomenon are offered. (author)

Modeling RERTR experimental fuel plates using the PLATE code

International Nuclear Information System (INIS)

Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Snelgrove, J.L.; Brazener, R.A.

2003-01-01

Modeling results using the PLATE dispersion fuel performance code are presented for the U-Mo/Al experimental fuel plates from the RERTR-1, -2, -3 and -5 irradiation tests. Agreement of the calculations with experimental data obtained in post-irradiation examinations of these fuels, where available, is shown to be good. Use of the code to perform a series of parametric evaluations highlights the sensitivity of U-Mo dispersion fuel performance to fabrication variables, especially fuel particle shape and size distributions. (author)

Thermal and x-ray studies on Tl2U(MoO4)3 and Tl4U(MoO4)4

International Nuclear Information System (INIS)

Dahale, N.D.; Keskar, Meera; Kulkarni, N.K.; Singh Mudher, K.D.

2006-01-01

In the quaternary Tl-U(IV)-Mo-O system, two new compounds namely Tl 2 U(MoO 4 ) 3 and Tl 4 U(MoO 4 ) 4 were prepared and characterized by powder X-ray diffraction and thermal methods. These compounds were prepared by solid state reactions of Tl 2 MoO 4 , UMoO 5 and MoO 3 in the required stoichiometric ratio at 500 deg C in evacuated sealed quartz ampoule. The XRD data of Tl 2 U(MoO 4 ) 3 and Tl 4 U(MoO 4 ) 4 were indexed on orthorhombic cell. TG curves of Tl 2 U(MoO 4 ) 3 and Tl 4 U(MoO 4 ) 4 did not show any weight change up to 700 deg C in an inert atmosphere. During heating in an inert atmosphere, Tl 2 U(MoO 4 ) 3 and Tl 4 U(MoO 4 ) 4 showed endothermic Dta peaks due to melting of the compounds at 519 and 565 deg C, respectively. (author)

U-Mo fuel qualification program in HANARO

International Nuclear Information System (INIS)

Lee, K.H.; Lee, C.S.; Kim, H.R.; Kuk, I.H.; Kim, C.K.

2000-01-01

Atomized U-Mo fuel has shown good performance from the results of previous out-of-pile tests and post-irradiation examinations. A qualification program of rod type U-Mo fuel is in progress and the fuel will be irradiated in HANARO. 6 gU/cm 3 U-7Mo, U-8Mo and U-9Mo are considered in this program. The laboratory test results of porosity, mechanical property, thermal conductivity, and thermal compatibility test are discussed in this paper. In parallel with this qualification program, the feasibility study on the core conversion from the present U 3 Si fuel to U-Mo in HANARO will be initiated to provide technical bases for the policy making. Several options of core conversion for HANARO are proposed and each option will be addressed briefly in terms of the operation policy, fuel management, and licensing of HANARO. (author)

Development of a PVD-based manufacturing process of monolithic LEU irradiation targets for {sup 99}Mo production

Energy Technology Data Exchange (ETDEWEB)

Hollmer, Tobias

2015-08-03

{sup 99}Mo is the most important radioisotope in nuclear medicine. It is produced by fission of uranium in irradiation targets. The usage of cylindrical monolithic targets can ensure a safe supply of {sup 99}Mo and at the same reduce the amount of highly radioactive waste generated during production. To manufacture these targets, a novel PVD-based technique was developed. Both the feasibility and the high efficiency of this process were demonstrated in a prototype apparatus.

Annealing tests of in-pile irradiated oxide coated U-Mo/Al-Si dispersed nuclear fuel

Science.gov (United States)

Zweifel, T.; Valot, Ch.; Pontillon, Y.; Lamontagne, J.; Vermersch, A.; Barrallier, L.; Blay, T.; Petry, W.; Palancher, H.

2014-09-01

U-Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800 °C of in-pile irradiated U-Mo/Al-Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500 °C and 670 °C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500 °C FGs are released from IDL/matrix interfaces. The second peak at 670 °C representing the main release of FGs originates from the interaction between U-Mo and matrix in the vicinity of the cladding.

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

Energy Technology Data Exchange (ETDEWEB)

Kim, Y S; Hofman, G L [Nuclear Engineering Division

2011-06-01

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.

Study of the feasibility of friction STIR welding applied to the fabrication of monolithic fuel elements; Estudio para la aplicacion del proceso de soldadura por friccion-agitacion (FSW) a la fabricacion de elementos combustibles monoliticos

Energy Technology Data Exchange (ETDEWEB)

Cabot, Pedro J; Moglioni, A [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. ENDE; Mirandou, Marcela; Balart, Silvia N [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Materiales

2004-07-01

The monolithic U-Mo fuel elements consist in a foil of a U-Mo alloy encased in Al. One of the techniques that is being tried to apply in their fabrication is Friction Stir Welding in the 'no contact at the interface' mode. The Laboratory of Welding at the National Atomic Energy Commission (Argentina) has a great experience in the conventional form of this technique so has started working on this new application. This paper describes the experiments performed to obtain the operative parameters. In the first experiments AA6061 T6 (Al) plates and sheets of AISI 316 (SS) were used to obtain the optimal operative parameters of the process. Welds were performed and evaluated for different operative variables such speed, angle and diameter of the tool and tool-interface gap keeping the rotation speed constant. Tensile test, pressure leak-proof test, bending test, non-destructive test and metallography were used to characterize the welds. Finally, SS and U-Mo foils were encased using the parameters selected from the first experiments. The samples prepared with U-Mo alloy will be used as diffusion couples and for the studies of interdiffusion under irradiation. (author)

Characterization of a U-Mo alloy subjected to direct hydriding of the gamma phase

International Nuclear Information System (INIS)

Balart, Silvia N.; Bruzzoni, Pablo; Granovsky, Marta S.

2003-01-01

The Reduced Enrichment for Research and Test Reactors (RERTR) program has imposed the need to develop plate-type fuel elements based on high density uranium compounds, such as U-Mo alloys. One of the steps in the fabrication of the fuel elements is the pulverization of the fissile material. In the case of the U-Mo alloys, the pulverization can be accomplished through hydriding - dehydriding. Two alternative methods of the hydriding-dehydriding process, namely the selective hydriding in alpha phase (HS-alpha) and the massive hydriding in gamma phase (HM-gamma) are currently being studied at the Comision Nacional de Energia Atomica. The HM-gamma method was reproduced at laboratory scale starting from a U-7 wt % Mo alloy. The hydrided and dehydrided materials were characterized using metallographic techniques, scanning electron microscopy, energy dispersive X-ray analysis and X-ray diffraction. These results are compared with previous results of the HS-alpha method. (author)

Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

Directory of Open Access Journals (Sweden)

Jaroszewicz Janusz

2014-07-01

Full Text Available The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.

Irradiation performance of U-Mo-Ti and U-Mo-Zr dispersion fuels in Al-Si matrixes

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Hofman, G.L. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B.; Wachs, D.M. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Ryu, H.J.; Park, J.M.; Yang, J.H. [Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

2012-08-15

Performance of U-7 wt.%Mo with 1 wt.%Ti, 1 wt.%Zr or 2 wt.%Zr, dispersed in an Al-5 wt.%Si alloy matrix, was investigated through irradiation tests in the ATR at INL and HANARO at KAERI. Post-irradiation metallographic features show that the addition of Ti or Zr suppresses interaction layer growth between the U-Mo and the Al-5 wt.%Si matrix. However, higher fission gas swelling was observed in the fuel with Zr addition, while no discernable effect was found in the fuel with Ti addition as compared to U-Mo without the addition. Known to have a destabilizing effect on the {gamma}-phase U-Mo, Zr, either as alloy addition or fission product, is ascribed for the disadvantageous result. Considering its benign effect on fuel swelling, with slight disadvantage from neutron economy point of view, Ti may be a better choice for this purpose.

Effect of in-pile degradation of the meat thermal conductivity on the maximum temperature of the plate-type U-Mo dispersion fuels

International Nuclear Information System (INIS)

Medvedev, Pavel G.

2009-01-01

Effect of in-pile degradation of thermal conductivity on the maximum temperature of the plate-type research reactor fuels has been assessed using the steady-state heat conduction equation and assuming convection cooling. It was found that due to very low meat thickness, characteristic for this type of fuel, the effect of thermal conductivity degradation on the maximum fuel temperature is minor. For example, the fuel plate featuring 0.635 mm thick meat operating at heat flux of 600 W/cm2 would experience only a 20 C temperature rise if the meat thermal conductivity degrades from 0.8 W/cm-s to 0.3 W/cm-s. While degradation of meat thermal conductivity in dispersion-type U-Mo fuel can be very substantial due to formation of interaction layer between the particles and the matrix, and development of fission gas filled porosity, this simple analysis demonstrates that this phenomenon is unlikely to significantly affect the temperature-based safety margin of the fuel during normal operation.

Characterization of the interaction layer in diffusion couples U-Mo-Zr/Al and U-Mo-Zr/Al-A356 at 550 C degrees; Caracterizacion de la zona de interaccion en pares de difusion a 550 grados C U-Mo-Zr/Al y U-Mo-Zr/Al-A356

Energy Technology Data Exchange (ETDEWEB)

Komar Varela, Carolina; Arico, Sergio; Mirandou, Marcela; Balart, Silvia; Gribaudo, Luis [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Materiales; com, carolinakomar@gmail

2007-07-01

Out-of-pile diffusion experiments were performed between U-7 wt.% Mo-1 wt.% Zr and Al or Al A356 (7,1 wt.% Si) at 550 C degrees. In this work morphological characterization and phase identification on both interaction layers are presented. They were carried out by the use of different techniques: optical and scanning electron microscopy, X-ray diffraction and WDS microanalysis. In the interaction layer U-7 wt.% Mo-1 wt.% Zr/Al, the phases UAl{sub 3}, UAl{sub 4}, Al{sub 20}Mo{sub 2}U and Al{sub 43}Mo{sub 4}U{sub 6} were identified. Similar results in the interaction layer of the U-7 % Mo/Al at 580 C degrees were previously obtained. In the interaction layer U-7 wt.% Mo-1 wt.% Zr/Al A356, the phases U(Al,Si){sub 3} with 25 at.% Si and Si{sub 5}U{sub 3} were identified. This last phase, with a higher Si concentration, was identified with X-ray diffraction synchrotron radiation performed at the National Synchrotron Light Laboratory, Campinas, Brazil. (author) [Spanish] Se realizaron experiencias fuera de reactor en pares de difusion quimica U-7 % Mo-1 % Zr/Al y U-7 % Mo-1 % Zr/Al A356. En este trabajo se presentan los resultados de la caracterizacion morfologica e identificacion de fases presentes en la zona de interaccion que se forma al ser sometidos a un tratamiento isotermico de 1,5 h a 550 grados C. Las tecnicas utilizadas fueron: microscopia optica y electronica de barrido, difraccion de rayos X y microanalisis cuantitativo por sonda electronica. En la zona de interaccion correspondiente al par U-7 % Mo-1 % Zr/Al se identificaron las fases UAl{sub 3}, UAl{sub 4}, Al{sub 20}Mo{sub 2}U y Al{sub 43}Mo{sub 4}U{sub 6}. Estas cuatro fases fueron identificadas en pares U-7 % Mo/Al a 580 grados C en trabajos anteriores. En la zona de interaccion correspondiente al par U-7 % Mo-1 % Zr/Al A356 se identificaron las fases U(Al,Si){sub 3} (con una concentracion de 25 %at.Si) y Si{sub 5}U{sub 3}. Este compuesto rico en Si solo pudo ser identificado mediante la utilizacion de

U-Mo/Al-Si interaction: Influence of Si concentration

International Nuclear Information System (INIS)

Allenou, J.; Palancher, H.; Iltis, X.; Cornen, M.; Tougait, O.; Tucoulou, R.; Welcomme, E.; Martin, Ph.; Valot, C.; Charollais, F.; Anselmet, M.C.; Lemoine, P.

2010-01-01

Within the framework of the development of low enriched nuclear fuels for research reactors, U-Mo/Al is the most promising option that has however to be optimised. Indeed at the U-Mo/Al interfaces between U-Mo particles and the Al matrix, an interaction layer grows under irradiation inducing an unacceptable fuel swelling. Adding silicon in limited content into the Al matrix has clearly improved the in-pile fuel behaviour. This breakthrough is attributed to an U-Mo/Al-Si protective layer around U-Mo particles appeared during fuel manufacturing. In this work, the evolution of the microstructure and composition of this protective layer with increasing Si concentrations in the Al matrix has been investigated. Conclusions are based on the characterization at the micrometer scale (X-ray diffraction and energy dispersive spectroscopy) of U-Mo7/Al-Si diffusion couples obtained by thermal annealing at 450 deg. C. Two types of interaction layers have been evidenced depending on the Si content in the Al-Si alloy: the threshold value is found at about 5 wt.% but obviously evolves with temperature. It has been shown that for Si concentrations ranging from 2 to 10 wt.%, the U-Mo7/Al-Si interaction is bi-layered and the Si-rich part is located close to the Al-Si for low Si concentrations (below 5 wt.%) and close to the U-Mo for higher Si concentrations. For Si weight fraction in the Al alloy lower than 5 wt.%, the Si-rich sub-layer (close to Al-Si) consists of U(Al, Si) 3 + UMo 2 Al 20 , when the other sub-layer (close to U-Mo) is silicon free and made of UAl 3 and U 6 Mo 4 Al 43 . For Si weight concentrations above 5 wt.%, the Si-rich part becomes U 3 (Si, Al) 5 + U(Al, Si) 3 (close to U-Mo) and the other sub-layer (close to Al-Si) consists of U(Al, Si) 3 + UMo 2 Al 20 . On the basis of these results and of a literature survey, a scheme is proposed to explain the formation of different types of ILs between U-Mo and Al-Si alloys (i.e. different protective layers).

DENSITY-FUNCTIONAL STUDY OF U-Mo AND U-Zr ALLOYS

Energy Technology Data Exchange (ETDEWEB)

Landa, A; Soderlind, P; Turchi, P A

2010-11-01

Density-functional theory previously used to describe phase equilibria in U-Zr alloys [A. Landa, P. Soederlind, P.E.A. Turchi, J. Alloys Comp. 478 (2009) 103-110] is extended to investigate the ground-state properties of U-Mo solid solutions. We discuss how the heat of formation in both alloys correlates with the charge transfer between the alloy components, and how the specific behavior of the density of states in the vicinity of the Fermi level promotes the stabilization of the U{sub 2}Mo compound. Our calculations prove that, due to the existence of a single {gamma}-phase over the typical fuel operation temperatures, {gamma}-U-Mo alloys should indeed have much lower constituent redistribution than {gamma}-U-Zr alloys for which binodal decomposition causes a high degree of constituent redistribution.

Development and characterization of monolithic fuel miniplate alloy U-2.5Zr-7.5Nb, coated in zircaloy

International Nuclear Information System (INIS)

Machado, Geraldo Correa

2014-01-01

The autocthonal production of nuclear fuel in Brazil for test and research reactors is restricted to MTR (Material Test Reactor) fuel type dispersion plate, using U3Si2 alloy, coated and dispersed in aluminum, developed by IPEN-SP for use in IEA-R1 reactor. Moreover, the UO 2 fuel rod type for power reactors is manufactured by Rezende (RJ) with a German technology by INB under license. Currently, Brazil is performing two programs of developing reactors. Currently, Brazil is developing two reactors. One of them is the development, by CNEN, the Brazilian Multipurpose Reactor (RMB), for testing, research and radioisotope production. The other one is the development a power reactor for naval propulsion, conducted by the Brazilian Navy. This dissertation presents the development and characterization of monolithic fuel miniplate alloy U-2.5Zr-7.5Nb, coated in zircaloy (ZRY), on a laboratory scale. Due to its innovative features and properties, this fuel can be used as fuel in both test reactors, research and producing radioisotopes for power reactors as small and medium sizes. Thus, this high potential fuel can be used in domestic reactors currently under development. The development of monolithic fuel plate type is made using the technique called 'picture-frame' where a sandwich composed of a monolith alloy U-2.5Zr- 7.5Nb coupled to a frame and coated sheets of Zry is obtained. The alloy U-2.5Zr-7.5Nb was obtained by melting in an induction furnace and then was cast into rectangular ingots of graphite, thus achieving an ingot with approximate dimensions of 170 x 50 x 60 mm. The obtained ingot was hot rolled at 850 ºC, with a 50 % reduction in thickness, in order to refine the raw structure of fusion. Samples cut from the alloy U-2.5Zr-7.5Nb, with dimensions 20 x 20 x 6 mm were placed in frames and plates Zry and joined by TIG (Tungsten Inert Gas) under an atmosphere of argon, obtaining a set of 10 mm thick, 45 mm wide and 100 mm long. The sandwiches were hot rolled to

Characterization of an irradiated RERTR-7 fuel plate using transmission electron microscopy

International Nuclear Information System (INIS)

Gan, J.; Keiser, D.D. Jr.; Miller, B.D.; Robinson, A.B.; Medvedev, P.

2010-01-01

Transmission electron microscopy (TEM) has been used to characterize an irradiated fuel plate with Al-2Si matrix from the Reduced Enrichment Research and Test Reactor RERTR-7 experiment that was irradiated under moderate reactor conditions. The results of this work showed the presence of a bubble superlattice within the U-7Mo grains that accommodated fission gases (e.g., Xe). The presence of this structure helps the U-7Mo exhibit a stable swelling behaviour during irradiation. Furthermore, TEM analysis showed that the Si-rich interaction layers that develop around the fuel particles at the U-7Mo/matrix interface during fuel plate fabrication and irradiation become amorphous during irradiation. An important question that remains to be answered about the irradiation behaviour of U-Mo dispersion fuels is how do more aggressive irradiation conditions affect the behaviour of fission gases within the U-7Mo fuel particles and in the amorphous interaction layers on the microstructural scale that can be characterized using TEM? This paper will discuss the results of TEM analysis that was performed on a sample taken from an irradiated RERTR-7 fuel plate with Al-2Si matrix. This plate was exposed to more aggressive irradiation conditions than the RERTR-6 plate. The microstructural features present within the U-7Mo and the amorphous interaction layers will be discussed. The results of this analysis will be compared to what was observed in the earlier RERTR-6 fuel plate characterization. (author)

Characterization of interaction between U-Mo alloy and Al diffusion-couple

International Nuclear Information System (INIS)

Liu Yunming; Yin Changgeng; Sun Changlong; Chen Jiangang; Sun Xudong

2011-01-01

In this paper, the interaction behavior of U-Mo/Al was studied with the diffusion-couple method, and the couple was continuously jointed by hot-pressing with special device. Annealing experiments were accomplished in a vacuum hot-pressing furnace, and at 550∼570℃ for 5∼21 hours. The results show that the morphology and composition of interaction Layer depend on the interaction layer thickness. The content of U (Mo) and Al is mutational at the interface of U-Mo/interaction layer/Al. The layer close to U-Mo side is mainly composed of product (U, Mo)Al 3 , while the Al side is composed of (U, Mo)Al 4 and UMO 2 Al 20 . Diffusion process of U-Mo/Al is Al immigrating over the Al/U-Mo original interface into U-Mo side and reacting with U-Mo, subsequently the interaction layer is growing into Al. (authors)

Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

Science.gov (United States)

Keiser, Dennis D.; Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K.; Ross Finlay, M.

2012-06-01

The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

Energy Technology Data Exchange (ETDEWEB)

Keiser, Dennis D., E-mail: Dennis.Keiser@inl.gov [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Ross Finlay, M. [Australian Nuclear Science and Technology Organization, PMB 1, Menai, NSW 2234 (Australia)

2012-06-15

The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

First-principles study of the surface properties of U-Mo system

Energy Technology Data Exchange (ETDEWEB)

Mei, Zhi-Gang; Liang, Linyun; Yacout, Abdellatif M.

2018-02-01

U-Mo alloys are promising fuels for future high-performance research reactors with low enriched uranium. Surface properties, such as surface energy, are important inputs for mesoscale simulations (e.g., phase field method) of fission gas bubble behaviors in irradiated nuclear fuels. The lack of surface energies of U-Mo alloys prevents an accurate modeling of the morphology of gas bubbles and gas bubble-induced fuel swelling. To this end, we study the surface properties of U-Mo system, including bcc Mo, alpha-U, gamma-U, and gamma U-Mo alloys. All surfaces up to a maximum Miller index of three and two are calculated for cubic Mo and gamma-U and non-cubic alpha-U, respectively. The equilibrium crystal shapes of bcc Mo, alpha-U and gamma-U are constructed using the calculated surface energies. The dominant surface orientations and the area fraction of each facet are determined from the constructed equilibrium crystal shape. The disordered gamma U-Mo alloys are simulated using the Special Quasirandom Structure method. The (1 1 0) and (1 0 0) surface energies of gamma U-7Mo and U-10Mo alloys are predicted to lie between those of gamma-U and bcc Mo, following a linear combination of the two constituents' surface energies. To better compare with future measurements of surface energies, the area fraction weighted surface energies of alpha-U, gamma-U and gamma U-7Mo and U-10Mo alloys are also predicted. (C) 2017 Published by Elsevier B.V.

Installation for the Mo-99 production from fission products

International Nuclear Information System (INIS)

Marques, R.O.; Cristini, P.R.; Marziale, D.P.; Furnari, E.S.; Fernandez, H.O.

1988-01-01

The installation to produce Mo-99 from nuclear fission started going on August 12th 1985 in Ezeiza Atomic Center. The characteristics of the process, the emplacement of the power plant, target, and irradiation conditions are presented. The targets are plates with a nucleus of Al/U alloy, with U-235 enriched to 90 % covered by Al plates. Each plate consists of about 1.10 -3 Kg of U-235 and 13.10 -3 Kg of Al. The plates are irradiated with a 3.10 13 n cm -2 s -1 flux during five days in the RA-3 nucleus. The Mo-99 separation method, is presented, where it is foreseen te I-131 separation. An account of the treatment of solid, liquid and gaseous waste is provided. An equipment to transfer the filter precipitate was designed in order to recover the U. The installation to continue the U recovery process, to separate I-131 and Xe-133 and to incorporate a Mo-99 purification stage for sublimation is being extended. (M.E.L.) [es

Thermal compatibility of U-2wt.%Mo and U-10wt.%Mo fuel prepared by centrifugal atomization for high density research reactor fuels

International Nuclear Information System (INIS)

Kim Ki Hwan; Lee Don Bae; Kim Chang Kyu; Kuk Il Hyun; Hofman, G.E.

1997-01-01

Research on the intermetallic compounds of uranium was revived in 1978 with the decision by the international research reactor community to develop proliferation-resistant fuels. The reduction of 93% 235 U (HEU) to 20% 235 U (LEU) necessitates the use of higher U-loading fuels to accommodate the addition 238 U in the LEU fuels. While the vast majority of reactors can be satisfied with U 3 Si 2 -Al dispersion fuel, several high performance reactors require high loadings of up to 8-9 g U cm -3 . Consequently, in the renewed fuel development program of the Reduced Enrichment for Research and Test Reactors (RERTR) Program, attention has shifted to high density uranium alloys. Early irradiation experiments with uranium alloys showed promise of acceptable irradiation behavior, if these alloys can be maintained in their cubic γ-U crystal structure. It has been reported that high density atomized U-Mo powders prepared by rapid cooling have metastable isotropic γ-U phase saturated with molybdenum, and good γ-U phase stability, especially in U-10wt.%Mo alloy fuel. If the alloy has good thermal compatibility with aluminium, and this metastable gamma phase can be maintained during irradiation, U-Mo alloy would be a prime candidate for dispersion fuel for research reactors. In this paper, U-2w.%Mo and U-10w.%Mo alloy powder which have high density (above 15 g-U/cm 3 ), are prepared by centrifugal atomization. The U-Mo alloy fuel meats are made into rods extruding the atomized powders. The characteristics related to the thermal compatibility of U-2w.%Mo and U-10w.%Mo alloy fuel meat at 400 o C for time up to 2000 hours are examined. (author)

Investigation of powdering ductile gamma U-10 wt%Mo alloy for dispersion fuels

Energy Technology Data Exchange (ETDEWEB)

Leal Neto, R.M., E-mail: lealneto@ipen.br [Nuclear and Energy Research Institute, IPEN/CNEN-SP, São Paulo (Brazil); Rocha, C.J. [Nuclear and Energy Research Institute, IPEN/CNEN-SP, São Paulo (Brazil); Urano de Carvalho, E. [Nuclear and Energy Research Institute, IPEN/CNEN-SP, São Paulo (Brazil); Science and Technology Brazilian Institute, Innovating Nuclear Reactors (Brazil); Riella, H.G. [Science and Technology Brazilian Institute, Innovating Nuclear Reactors (Brazil); Chemical Engineering Department, Santa Catarina Federal University, Florianópolis (Brazil); Durazzo, M. [Nuclear and Energy Research Institute, IPEN/CNEN-SP, São Paulo (Brazil); Science and Technology Brazilian Institute, Innovating Nuclear Reactors (Brazil)

2014-02-01

This work forms part of the studies presently ongoing at Nuclear and Energy Research Institute – IPEN/CNEN-SP investigating the feasibility of powdering ductile U-10 wt%Mo alloy by hydriding–milling–dehydriding of the gamma phase (HMD). Hydriding was conducted at room temperature in a Sievert apparatus following heat treatment activation. Hydrided pieces were fragile enough to be hand milled to the desired particle size range. Hydrogen was removed by heating the samples under high vacuum. X-ray diffraction analysis of the hydrided material showed an amorphous-like pattern that is completely reversed following dehydriding. The hydrogen content of the hydrided samples corresponds to a trihydride, i.e. (U,Mo)H{sub 3}. SEM analysis of HMD powder particles revealed equiaxial powder particles together with some plate-like particles. A hypothesis for the amorphous hydride phase formation is suggested.

Effect of Silicon in U-10Mo Alloy

Energy Technology Data Exchange (ETDEWEB)

Kautz, Elizabeth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Devaraj, Arun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kovarik, Libor [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-08-31

This document details a method for evaluating the effect of silicon impurity content on U-10Mo alloys. Silicon concentration in U-10Mo alloys has been shown to impact the following: volume fraction of precipitate phases, effective density of the final alloy, and 235-U enrichment in the gamma-UMo matrix. This report presents a model for calculating these quantities as a function of Silicon concentration, which along with fuel foil characterization data, will serve as a reference for quality control of the U-10Mo final alloy Si content. Additionally, detailed characterization using scanning electron microscope imaging, transmission electron microscope diffraction, and atom probe tomography showed that Silicon impurities present in U-10Mo alloys form a Si-rich precipitate phase.

Interdiffusion studies on hot rolled U-10Mo/AA1050

Energy Technology Data Exchange (ETDEWEB)

Saliba-Silva, A.M.; Martins, I.C.; Carvalho, E.U.; Durazzo, M.; Riella, H.G. [Instituto de Pesquisas Energeticas e Nucleares (CCN/IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Combustivel Nuclear], e-mail: saliba@ipen.br

2010-07-01

The U-Mo alloys are investigated with the goal of becoming nuclear material to fabricate high-density fuel elements for high performance research reactors. This enrichment level suggests that the U-Mo alloys should be between 6 to 10wt%, which can give up to 9gU/cm{sup 3} as fuel density. Nevertheless, the U-Mo alloys are very reactive with Al. Interdiffusion reaction products are formed since nuclear fission promotes chemical interaction layer during operation, leading to potential structural failure. Present studies were made with treated hot rolled diffusion couples of U-10Mo inserted in Al (AA1050). The U-10Mo/AA1050 pairs were treated in two temperature (150 degree C and 550 degree C) with three soaking times (5h, 40h and 80h). From microstructure analyses, rapid diffusion of Al happened inside U-10Mo in the first heating at 540 degree C during 15 min, reaching 8 at%Al in a range of 200 {mu}m towards U-10Mo. Longer time (5, 40, 80h) at 550 degree C maintain this level of Al-content up to 1000 {mu}m inside U-10Mo. A minor depth ({approx}1 {mu}m) near the interdiffusion contact had higher Al-content, but not sufficient to form identifiable (U,Mo)Al{sub x} structures. Probably, residual elements reduced drastically the interdiffusion phenomena between U-10Mo and AA1050, maybe due to silicon presence. (author)

Characterization of the interaction layer in diffusion couples U-Mo-Zr/Al and U-Mo-Zr/Al-A356 at 550 C degrees

International Nuclear Information System (INIS)

Komar Varela, Carolina; Arico, Sergio; Mirandou, Marcela; Balart, Silvia; Gribaudo, Luis

2007-01-01

Out-of-pile diffusion experiments were performed between U-7 wt.% Mo-1 wt.% Zr and Al or Al A356 (7,1 wt.% Si) at 550 C degrees. In this work morphological characterization and phase identification on both interaction layers are presented. They were carried out by the use of different techniques: optical and scanning electron microscopy, X-ray diffraction and WDS microanalysis. In the interaction layer U-7 wt.% Mo-1 wt.% Zr/Al, the phases UAl 3 , UAl 4 , Al 20 Mo 2 U and Al 43 Mo 4 U 6 were identified. Similar results in the interaction layer of the U-7 % Mo/Al at 580 C degrees were previously obtained. In the interaction layer U-7 wt.% Mo-1 wt.% Zr/Al A356, the phases U(Al,Si) 3 with 25 at.% Si and Si 5 U 3 were identified. This last phase, with a higher Si concentration, was identified with X-ray diffraction synchrotron radiation performed at the National Synchrotron Light Laboratory, Campinas, Brazil. (author) [es

High temperature interdiffusion and phase equilibria in U-Mo

International Nuclear Information System (INIS)

Lundberg, L.B.

1988-01-01

Experimental data for interdiffusion and phase equilibria in the U-Mo system have been obtained over the temperature range 1400 to 1525 K as a fallout from compatibility experiments in which UO 2 was decomposed by lithium in closed molybdenum capsules. Composition-position, x-ray diffraction and microstructural data from the interdiffusion zones indicate that the intermediate phase U 2 Mo is found in this temperature range, contrary to the currently accepted equilibrium U-Mo phase diagram. The U-Mo interdiffusion data are in good agreement with published values. Inclusion of the U 2 Mo phase in a theoretical correlation of interdiffusion and phase equilibria data using Darken's equation indicate that high temperature interdiffusion of uranium and molybdenum follows the usual thermodynamic rules. Significant changes in the value of the thermodynamic based Darken factor near the U 2 Mo phase boundary on the high uranium side are indicated from both the new and published interdiffusion data. 9 refs., 10 figs., 3 tabs

Concept Feasibility Report for Using Co-Extrusion to Bond Metals to Complex Shapes of U-10Mo

Energy Technology Data Exchange (ETDEWEB)

Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Paxton, Dean M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Mark T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soulami, Ayoub [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2013-12-01

In support of the Convert Program of the U.S. Department of Energy’s National Nuclear Security Administration (DOE/NNSA) Global Threat Reduction Initiative (GTRI), Pacific Northwest National Laboratory (PNNL) has been investigating manufacturing processes for the uranium-10% molybdenum (U-10Mo) alloy plate fuel for the U.S. high-performance research reactors (USHPRR). This report documents the results of PNNL’s efforts to develop the extrusion process for this concept. The approach to the development of a co-extruded complex-shaped fuel has been described and an extrusion of DU-10Mo was made. The initial findings suggest that given the extrusion forces required for processing U-10Mo, the co-extrusion process can meet the production demands of the USHPRR fuel and may be a viable production method. The development activity is in the early stages and has just begun to identify technical challenges to address details such as dimensional tolerances and shape control. New extrusion dies and roll groove profiles have been developed and will be assessed by extrusion and rolling of U-10Mo during the next fiscal year. Progress on the development and demonstration of the co-extrusion process for flat and shaped fuel is reported in this document

Thermal behavior analysis of U-Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin; Park, Jong Mang; Lee, Yoon Sang; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2004-07-01

According to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program, low enriched uranium(LEU) fuel such as uranium silicide dispersion fuels are being used in research reactors. Because of a lower enrichment higher uranium density fuels are required for some high performance research reactors. Some uranium alloys with a high uranium density such as U-Mo alloys have been considered as one of the most promising candidates for a dispersion fuel due to the good irradiation performance. An international qualification program to replace the uranium silicide dispersion fuel with U-Mo dispersion fuel is being carried out under the RERTR program. Although U-Mo powders are conventionally supplied by the mechanical comminuting of as-cast U-Mo alloys, KAERI developed a centrifugal atomization method in order to simplify the preparation process and improve the properties. The centrifugally atomized powders have a rapidly solidified gamma uranium structure and a spherical shape. During the in-reactor operation of a dispersion fuel, interdiffusion or chemical reactions between the fuel particles and the matrix occurr. Intermetallic compounds in the form of UAlx are formed as a result of the diffusional reaction. Because the intermetallic compounds are less dense than the combined reactants, the volume of the fuel element increases after the reaction. In addition to the effect on the swelling performance, the reaction layers between the U-Mo and the Al matrix induces a degradation of the thermal properties of the U-Mo/Al dispersion fuels. It is important to investigate the thermal behavior of U-Mo/Al dispersion fuel according to reaction between the fuel particles and the matrix with the burnup and linear power. In this study, a finite element analysis was used for the calculation of the temperature distribution of the U-Mo/Al dispersion fuel with a burnup and linear power. Kinetics data of the reaction layers such as the growth

Thermal behavior analysis of U-Mo/Al dispersion fuel

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Mang; Lee, Yoon Sang; Kim, Chang Kyu

2004-01-01

According to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program, low enriched uranium(LEU) fuel such as uranium silicide dispersion fuels are being used in research reactors. Because of a lower enrichment higher uranium density fuels are required for some high performance research reactors. Some uranium alloys with a high uranium density such as U-Mo alloys have been considered as one of the most promising candidates for a dispersion fuel due to the good irradiation performance. An international qualification program to replace the uranium silicide dispersion fuel with U-Mo dispersion fuel is being carried out under the RERTR program. Although U-Mo powders are conventionally supplied by the mechanical comminuting of as-cast U-Mo alloys, KAERI developed a centrifugal atomization method in order to simplify the preparation process and improve the properties. The centrifugally atomized powders have a rapidly solidified gamma uranium structure and a spherical shape. During the in-reactor operation of a dispersion fuel, interdiffusion or chemical reactions between the fuel particles and the matrix occurr. Intermetallic compounds in the form of UAlx are formed as a result of the diffusional reaction. Because the intermetallic compounds are less dense than the combined reactants, the volume of the fuel element increases after the reaction. In addition to the effect on the swelling performance, the reaction layers between the U-Mo and the Al matrix induces a degradation of the thermal properties of the U-Mo/Al dispersion fuels. It is important to investigate the thermal behavior of U-Mo/Al dispersion fuel according to reaction between the fuel particles and the matrix with the burnup and linear power. In this study, a finite element analysis was used for the calculation of the temperature distribution of the U-Mo/Al dispersion fuel with a burnup and linear power. Kinetics data of the reaction layers such as the growth

Structure and thermal properties of as-fabricated U-7Mo/Mg and U-10Mo/Mg low-enriched uranium research reactor fuels

Energy Technology Data Exchange (ETDEWEB)

Kulakov, Mykola, E-mail: mykola.kulakov@cnl.ca [Fuel Development Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Saoudi, Mouna [Fuel Development Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Piro, Markus H.A. [Fuel and Fuel Channel Safety Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Donaberger, Ronald L. [Canadian Neutron Beam Centre, Chalk River, ON K0J 1J0 Canada (Canada)

2017-02-15

Aluminum-clad U-7Mo/Mg and U-10Mo/Mg pin-type mini-elements (with a core uranium loading of 4.5 gU/cm{sup 3}) have been fabricated at the Canadian Nuclear Laboratories for experimental tests and ultimately for use in research and test reactors. In this study, the microstructure and phase composition of unirradiated U-7Mo/Mg and U-10Mo/Mg fuel cores were analyzed using optical and scanning electron microscopy, and neutron powder diffraction. Thermal properties were characterized using a combination of experimental measurements and thermodynamic calculations. The thermal diffusivity was measured using the laser flash method. The temperature-dependent specific heat capacities were calculated based on the linear rule of mixture using the weight fraction of different crystalline phases and their specific heat capacity values taken from the literature. The thermal conductivity was then calculated using the measured thermal diffusivity, the measured density and the calculated specific heat capacity. The resulting thermal conductivity is practically identical for both types of fuel. The in-reactor temperatures were predicted using conjugate heat transfer simulations. - Highlights: • Neutron diffraction analysis shows that most of the γ-U(Mo) phase was retained in as-fabricated U-7Mo/Mg and U-10Mo/Mg fuel cores. • The experimental thermal conductivity of both types of fuel is practically identical. • Based on conjugate heat transfer simulations, under normal operating conditions, the in-reactor fuel centreline temperature is about 510 K.

Reaction layer in U-7WT%MO/Al diffusion couples

International Nuclear Information System (INIS)

Mirandou, M.I.; Balart, S.N.; Ortiz, M.; Granovsky, M.S.

2003-01-01

New results of the reaction layer characterization between γ (U-7wt%Mo) alloy and Al, in chemical diffusion couples, are presented. The analysis was performed using optical and scanning electron microscopy with EDAX and X-ray diffraction techniques. Besides the main components (U, Mo)Al 3 and (U, Mo)Al 4 , already reported, two ternary compounds of high Al content have been identified in the reaction layer when it grew in retained or decomposed γ (U, Mo) phase, respectively. The drastic consequence on the interdiffusion behavior due to the thermal instability of the retained γ (U, Mo) phase is discussed. (author)

Isothermal section of diagram of U-Mo-B and U-Re-B systems

International Nuclear Information System (INIS)

Val'ovka, I.P.; Kuz'ma, Yu.B.

1986-01-01

The methods of X-ray analysis are used to study the U-Mo-B and U-Re-B systems and to plot phase equilibrium diagrams at 1000 and 800 deg C, respectively. A formation of boride UMoB 4 (structure of the ThMoB 4 type) is confirmed in the U-Mo-B system and new compounds are found: U 2 MoB 6 (rhombic structure of the Y 2 ReB 6 type, a=0.9301(9), b=1.1434(11), c=0.3678(4) nm), ∼UMo 2 B 6 and ∼ UMo 4 B 4 with unknown structures. In the U-Re-B system besides previously known boride UReB 4 (the ThMoB 4 structure type), new ones are obtained: U 2 ReB 6 (Y 2 ReB 6 type, a=0.9373(9), b=1.1529(13), c0.3653(4) nm) and UReB 3 (hexagonal structure of the proper type, a=0.5083(1), c=0.5095(1) nm)

Triple Plate Mold Final Report: Optimization of the Mold Design and Casting Parameters for a Thin U-10mo Fuel Casting

Energy Technology Data Exchange (ETDEWEB)

Aikin, Jr., Robert M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-01-04

This work describes the experiments and modeling that have been performed to improve and try to optimize the simultaneous casting of three plates of U-10wt%Mo in a single coil vacuum induction melting (VIM) furnace. The plates of interest are 280 mm wide by 203 mm tall by 5 mm thick (11" x 8" x 0.2"). The initial mold design and processing parameters were supplied by Y-12. The mold and casting cavity were instrumented with a number of thermocouples, and the casting performed to determine the thermal history of the mold and casting. The resulting cast plates were radiographed and numerous defects identified. Metallography was performed to help identify the nature of the radiographically observed defects. This information was then used to validate a mold filling and solidification model of that casting. Based on the initial casting, good casting design practice, and process simulation of several design alternatives, a revised design was developed with the goal of minimizing casting defects such as porosity. The redesigned mold had a larger hot-top and had its long axis along the horizontal direction. These changes were to try to develop a strong thermal gradient conducive to good feeding and minimization of micro- and macroporosity in the cast plates. An instrumented casting was then performed with the revised mold design and a linear distributor. This design yielded cast plates with significantly less radiographically identified defects. Unfortunately, there was significant variation in plate weight and metal content in their hot-tops. Fluid flow simulations were then performed on this mold/distributor design. This helped identify the issue with this linear distributor design. Additional simulations were then performed on candidate distributor redesigns and a preferred distributor annular design was identified. This improved annular design was used to produce a third instrumented casting with favorable results. These refined designs and their radiographic

2011 Progress Report on HEU Minimization Activities in Argentina

Energy Technology Data Exchange (ETDEWEB)

Bonini, A.; Cristini, P.; Lio, L. De; Dell' Occhio, L.; Gil, D.; Gonzalez, A.G.; Gonzalez, R.; Varela, C. Komar; Lopez, M.; Novara, O.; Taboada, H. [Comision Nacional de Energia Atomica, Av. Del Libertador 8250 (1429) Buenos Aires (Argentina)

2011-07-01

After the core conversion of the RA-6 reactor finished in March 2008, an extension of the original CNEA-NNSA DoE contract was signed to enhance the final national HEU inventories minimization. Before this process, CNEA reserved a small inventory of HEU for R and D uses in fission chambers, neutronic probes and standards. This minimization comprises that all fresh and irradiated HEU remnant inventories coming from fuels and Mo99 irradiation targets fabrication and irradiated HEU-oxides retained in production filters and solutions will be recovered, down-blended into LEU and purified or dispose as waste whenever its recovery would not be advisable due to cost-benefit consideration. CNEA has a R and D program to develop the fabrication technology of both dispersed U-Mo (Al-Si matrix and Al cladding) and monolithic (Zry-4 cladding) miniplates to support the qualification activities of the RERTR program. Some monolithic 58% enrichment and LEU 8%Mo and U10%Mo miniplates and plates were and are being delivered to INL-DoE to be irradiated in the ATR reactor core. CNEA, a worldwide leader on LEU technology for fission radioisotope production is providing Brazil with 1/3 of the national requirements on Mo99 by weekly deliveries. Australia has started the fission radioisotope production through several batches by week, based on CNEA's LEU technology provided by INVAP SE. CNEA is also committed to improve the diffusion of LEU target and radiochemical technology for radioisotope production and target and process optimization. Future plans include: 1. Plans to recover and purify the LEU based inventories in Mo99 production filters, once the HEU to LEU campaign is over. 2. Fabrication and delivering to INL to be irradiated in the ATR core of U-8%Mo and U-10%Mo monolithic miniplates and development and fabrication of LEU very high density monolithic and dispersed U-Mo fuel plates with Zr cladding for the FUTURE-MONO experiment in the frame of the RERTR program. 3

Calculation of thermodynamic equilibrium between bcc disordered solid solutions U and Mo

International Nuclear Information System (INIS)

Alonso, Paula R.; Rubiolo, Gerardo H.

2003-01-01

There is actually an interest to develop a new fuel with higher density for research reactors. Fuel plates would be obtained by dispersion, a method that requires both a very dense fuel dispersant (>15.0 g U/cm 3 ) and a very high volume loading of the dispersant (>55%). Dispersants based in gamma (BCC) stabilized uranium alloys are being investigated, as they are able to reach uranium densities of 17.0 g U/cm 3 . Among them, we focus in U(Mo) bcc solid solutions with the addition of ternary elements to stabilize gamma phase. Transition metals, 4d and 5d, of groups VII and VIII are good candidates for the ternary alloy U - Mo - X. Their relative power to stabilize gamma phase seems to be in close relation with bonding energies between atoms in the alloy. A first approach to the calculation of these energies has been performed by the semi empiric method of Miedema where only bonds between pairs are considered, neglecting ternary and quaternary bonds. There is also a lack of information concerning solubilities of the ternary elements in the ternary cubic phase. In this work we aim to calculate bonding energies between atoms in the alloy using a cluster expansion of the formation energy (T=0 K) of a series of bcc ordered compounds in the systems U-Mo-X. Then the calculation of the equilibrium phase diagram by the Cluster Variation Method will be done (CVM). We show here the first part of the investigation devoted to calculation of phases equilibria in the U Mo system Formation energies of the ordered compounds were obtained by the first principles methods TB-LMTO-ASA and FP-LAPW. Another set of bonding energies was calculated in order to fit the known experimental diagram and new formation energies for the ordered compounds were derived from them. Discrepancies between both sets are discussed. (author)

Development of U-Mo/Al dispersion fuel for research reactors

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Man; Ryu, Ho Jin; Yang, Jae Ho; Jeong, Yong Jin; Lee, Yoon Sang [Korea Atomic Energy Research Inst., Research Reactor Fuel Development Division, Daejeon (Korea, Republic of)

2012-03-15

Currently, the KOMO-5 irradiation test for full size U-Mo/Al dispersion fuel rods has been underway since May 23, 2011. The purpose of the KOMO-5 test includes an investigation of the irradiation behaviors of silicide or nitride coated U-7Mo/Al(-Si) dispersion fuels and the effects of pre-formed interaction layers on U-Mo particles. It is expected that the irradiation test will be finished after attaining 60 at% U-235 burnup in May 2012, and the first PIE results of the KOMO-5 will be obtained in September 2012. In addition, an international cooperation program on the qualification of U-Mo dispersion fuels for small and medium size research reactors is going to be proposed in cooperation with the IAEA. Conversion from silicide fuel to U-Mo fuel will increase the cycle length with a smaller number of fuel assemblies and allow more flexible back-end options for spent fuel due to of the reprocessibility of U-Mo. (author)

Detecting the Extent of Eutectoid Transformation in U-10Mo

International Nuclear Information System (INIS)

Devaraj, Arun; Jana, Saumyadeep; McInnis, Colleen A.; Lombardo, Nicholas J.; Joshi, Vineet V.; Sweet, Lucas E.; Manandhar, Sandeep; Lavender, Curt A.

2016-01-01

During eutectoid transformation of U-10Mo alloy, uniform metastable ? UMo phase is expected to transform to a mixture of ?-U and ?'-U_2Mo phase. The presence of transformation products in final U-10Mo fuel, especially the ? phase is considered detrimental for fuel irradiation performance, so it is critical to accurately evaluate the extent of transformation in the final U-10Mo alloy. This phase transformation can cause a volume change that induces a density change in final alloy. To understand this density and volume change, we developed a theoretical model to calculate the volume expansion and resultant density change of U-10Mo alloy as a function of the extent of eutectoid transformation. Based on the theoretically calculated density change for 0 to 100% transformation, we conclude that an experimental density measurement system will be challenging to employ to reliably detect and quantify the extent of transformation. Subsequently, to assess the ability of various methods to detect the transformation in U-10Mo, we annealed U-10Mo alloy samples at 500°C for various times to achieve in low, medium, and high extent of transformation. After the heat treatment at 500°C, the samples were metallographically polished and subjected to optical microscopy and x-ray diffraction (XRD) methods. Based on our assessment, optical microscopy and image processing can be used to determine the transformed area fraction, which can then be correlated with the ? phase volume fraction measured by XRD analysis. XRD analysis of U-10Mo aged at 500°C detected only ? phase and no ?' was detected. To further validate the XRD results, atom probe tomography (APT) was used to understand the composition of transformed regions in U-10Mo alloys aged at 500°C for 10 hours. Based on the APT results, the lamellar transformation product was found to comprise ? phase with close to 0 at% Mo and ? phase with 28-32 at% Mo, and the Mo concentration was highest at the ?/? interface.

Effect of stress evolution on microstructural behavior in U-Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Jeong, G.Y. [Department of Nuclear Engineering, Ulsan National Institute of Science and Technology, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan 689-798 (Korea, Republic of); Kim, Yeon Soo; Jamison, L.M. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Lee, K.H. [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Sohn, Dong-Seong, E-mail: dssohn@unist.ac.kr [Department of Nuclear Engineering, Ulsan National Institute of Science and Technology, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan 689-798 (Korea, Republic of)

2017-04-15

U-Mo/Al dispersion fuel irradiated to high burnup at high power (high fission rate) exhibited microstructural changes including deformation of the fuel particles, pore growth, and rupture of the Al matrix. The driving force for these microstructural changes was meat swelling resulting from a combination of fuel particle swelling and interaction layer (IL) growth. In some cases, pore growth in the interaction layers also contributed to meat swelling. The main objective of this work was to determine the stress distribution within the fuel meat that caused these phenomena. A mechanical equilibrium between the stress generated by fuel meat swelling and the stress relieved by fission-induced creep in the meat constituents (U-Mo particles, Al matrix, and IL) was considered. Test plates with well-recorded fabrication data and irradiation conditions were used, and their post-irradiation examination (PIE) data was obtained. ABAQUS finite element analysis (FEA) was utilized to simulate the microstructural evolution of the plates. The simulation results allowed for the determination of effective stress and hydrostatic stress exerted on the meat constituents. The effects of fabrication and irradiation parameters on the stress distribution that drives microstructural evolutions, such as pore growth in the IL and Al matrix rupture, were investigated. - Highlights: •Post-irradiation data for irradiated miniplates were analyzed by using their optical microscopy images. •ABAQUS finite element analysis (FEA) package was utilized to simulate the microstructural evolution of the selected plates. •Stresses were assessed to analyze their effects on microstructural changes during irradiation.

Fuel Performance Modeling of U-Mo Dispersion Fuel: The thermal conductivity of the interaction layers of the irradiated U-Mo dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Mistarhi, Qusai M.; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

2016-05-15

U-Mo/Al dispersion fuel performed well at a low burn-up. However, higher burn-up and higher fission rate irradiation testing showed enhanced fuel meat swelling which was caused by high interaction layer growth and pore formation. The performance of the dispersion type fuel in the irradiation and un-irradiation environment is very important. During the fabrication of the dispersion type fuel an Interaction Layer (IL) is formed due to the inter-diffusion between the U-Mo fuel particles and the Al matrix which is an intermetallic compound (U,Mo)Alx. During irradiation, the IL becomes amorphous causing a further decrease in the thermal conductivity and an increase in the centerline temperature of the fuel meat. Several analytical models and numerical methods were developed to study the performance of the unirradiated U-Mo/Al dispersion fuel. Two analytical models were developed to study the performance of the irradiated U-Mo/Al dispersion fuel. In these models, the thermal conductivity of the IL was assumed to be constant. The properties of the irradiated U-Mo dispersion fuel have been investigated recently by Huber et al. The objective of this study is to develop a correlation for IL thermal conductivity during irradiation as a function of the temperature and fission density from the experimentally measured thermal conductivity of the irradiated U-Mo/Al dispersion fuel. The thermal conductivity of IL during irradiation was calculated from the experimentally measured data and a correlation was developed from the thermal conductivity of IL as a function of T and fission density.

Detecting the Extent of Eutectoid Transformation in U-10Mo

Energy Technology Data Exchange (ETDEWEB)

Devaraj, Arun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jana, Saumyadeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); McInnis, Colleen A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lombardo, Nicholas J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sweet, Lucas E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Manandhar, Sandeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-08-31

During eutectoid transformation of U-10Mo alloy, uniform metastable γ UMo phase is expected to transform to a mixture of α-U and γ’-U₂Mo phase. The presence of transformation products in final U-10Mo fuel, especially the α phase is considered detrimental for fuel irradiation performance, so it is critical to accurately evaluate the extent of transformation in the final U-10Mo alloy. This phase transformation can cause a volume change that induces a density change in final alloy. To understand this density and volume change, we developed a theoretical model to calculate the volume expansion and resultant density change of U-10Mo alloy as a function of the extent of eutectoid transformation. Based on the theoretically calculated density change for 0 to 100% transformation, we conclude that an experimental density measurement system will be challenging to employ to reliably detect and quantify the extent of transformation. Subsequently, to assess the ability of various methods to detect the transformation in U-10Mo, we annealed U-10Mo alloy samples at 500°C for various times to achieve in low, medium, and high extent of transformation. After the heat treatment at 500°C, the samples were metallographically polished and subjected to optical microscopy and x-ray diffraction (XRD) methods. Based on our assessment, optical microscopy and image processing can be used to determine the transformed area fraction, which can then be correlated with the α phase volume fraction measured by XRD analysis. XRD analysis of U-10Mo aged at 500°C detected only α phase and no γ’ was detected. To further validate the XRD results, atom probe tomography (APT) was used to understand the composition of transformed regions in U-10Mo alloys aged at 500°C for 10 hours. Based on the APT results, the lamellar transformation product was found to comprise α phase with close to 0 at% Mo and γ phase with 28–32 at% Mo, and the Mo concentration was highest at the �

The fracture toughness and DBTT of MoB particle-reinforced MoSi2 composites

International Nuclear Information System (INIS)

Xiong Zhi; Wang Gang; Jiang Wan

2005-01-01

The room temperature fracture toughness and the high temperature DBTT of MoB particle-reinforced MoSi 2 composites were investigated using Vickers indentation technique and MSP testing method, respectively. Modified small punch (MSP) test is a method for evaluation of mechanical properties using very small specimens, and it's appropriate for the determination of strength and DBTT. It was found that the approximate fracture toughness of the composite is 1.3 times that of monolithic MoSi 2 , and its DBTT is 100 C higher than that of monolithic MoSi 2 materials. Cracks deflection is a probable mechanism responsible for this behavior. (orig.)

Thermal expansion studies on UMoO5, UMoO6, Na2U(MoO4)3 and Na4U(MoO4)4

International Nuclear Information System (INIS)

Keskar, Meera; Dahale, N.D.; Krishnan, K.

2009-01-01

In the present work, thermal expansion behavior of lower valent sodium uranium molybdates, i.e., Na 2 U(MoO 4 ) 3 and Na 4 U(MoO 4 ) 4 were studied under vacuum in the temperature range of 298-873 K using high temperature X-ray diffractometry (HTXRD). Expansion behaviors of UMoO 5 and UMoO 6 were also studied in vacuum from 298 to 873 K and 773 K, respectively. UMoO 5 was synthesized by reacting UO 2 with MoO 3 in equi-molar proportion in evacuated sealed quartz ampoule at 1173 K for 14 h. Na 2 U(MoO 4 ) 3 and Na 4 U(MoO 4 ) 4 were prepared by reacting UMoO 5 and MoO 3 with 1 and 2 moles of Na 2 MoO 4 , respectively, at 873 K in evacuated sealed quartz ampoule. XRD data of UMoO 5 and UMoO 6 were indexed on orthorhombic and monoclinic systems, respectively, whereas, the data of Na 2 U(MoO 4 ) 3 and Na 4 U(MoO 4 ) 4 were indexed on tetragonal system. The lattice parameters and cell volume of all the four compounds, fit into polynomial expression with respect to temperature, showed positive thermal expansion (PTE) up to 873 K.

Performance evaluation of large U-Mo particle dispersed fuel irradiated in HANARO

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu

2008-01-01

U-Mo/Al dispersion fuel is being developed as advanced fuel for research reactors. Irradiation behavior of U-Mo/Al dispersion fuel has been studied to evaluate its fuel performance. One of the performance limiting factors is a chemical interaction between the U-Mo particle and the Al matrix because the thermal conductivity of fuel meat is decreased with the interaction layer growth. In order to overcome the interaction problem, large-sized U-Mo particles were fabricated by controlling the centrifugal atomization conditions. The fuel performance behavior of U-Mo/Al dispersion fuel was estimated by using empirical models formulated based on the microstructural analyses of the post-irradiation examination (PIE) on U-Mo/Al dispersion fuel irradiated in HANARO reactor. Temperature histories of U-Mo/Al dispersion fuel during irradiation tests were estimated by considering the effect of an interaction layer growth on the thermal conductivity of the fuel meat. When the fuel performances of the dispersion fuel rods containing U-Mo particles with various sizes were compared, fuel temperature was decreased as the average U-Mo particle size was increases. It was found that the dispersion of a larger U-Mo particle was effective for mitigating the thermal degradation which is associated with an interaction layer growth. (author)

Microstructural development from interdiffusion and reaction between U−Mo and AA6061 alloys annealed at 600° and 550 °C

Energy Technology Data Exchange (ETDEWEB)

Perez, E., E-mail: Emmanuel.Perez@inl.gov [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Keiser, D.D., E-mail: dennis.keiser@inl.gov [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Sohn, Y.H., E-mail: yongho.sohn@ucf.edu [Department of Materials Science and Engineering, University of Central Florida, 4000 Central Florida Blvd., Orlando, FL 32816 (United States)

2016-08-15

The U.S. Material Management and Minimization Reactor Conversion Program is developing low enrichment fuel systems encased in Al-alloy for use in research and test reactors. Monolithic fuel plates have local regions where the U−Mo fuel plate may come into contact with the Al-alloy 6061 (AA6061) cladding. This results in the development of interdiffusion zones with complex microstructures with multiple phases. In this study, the microstructural development of diffusion couples, U−7 wt%Mo, U−10 wt%Mo, and U−12 wt%Mo vs. AA6061, annealed at 600 °C for 24 h and at 550 °C for 1, 5, and 20 h, were analyzed by scanning electron microscopy with x-ray energy dispersive spectroscopy. The microstructural development and kinetics were compared to diffusion couples U−Mo vs. high purity Al and binary Al−Si alloys. The diffusion couples developed complex interaction regions where phase development was influenced by the alloying additions of the AA6061. - Highlights: • Diffusion couples of U−7Mo, U−10Mo, and U−12Mo vs. AA6061 were analyzed by SEM with XEDS. • The couples were annealed at 600 °C for 24 h and at 550 °C for 1, 5 and 20 h. • The interaction regions were more complex than those in diffusion couples of U−Mo vs. high purity Al and Al−Si alloys. • Analysis showed that the alloying additions of the AA6061 were present in the interaction regions. • Phase development was significantly influenced by the alloying additions of the AA6061.

2010 national progress report on R and D on LEU fuel and target technology in Argentina

International Nuclear Information System (INIS)

Balart, S.; Blaumann, H.; Cristini, P.; Gonzalez, A.G.; Gonzalez, R.; Hermida, J.D.; Lopez, M.; Mirandou, M.; Taboada, H.

2010-01-01

Since last RRFM meeting, CNEA has deployed several related tasks. The RA-6 MTR type reactor, converted its core from HEU to a new LEU silicide one is scaling up the power, according to a protocol requested by the national regulatory body, ARN. CNEA is deploying an intense R and D activity to fabricate both dispersed U-Mo (Al-Si matrix and Al cladding) and monolithic (Zry-4 cladding) miniplates to develop possible solutions to VHD dispersed and monolithic fuels technical problems. Some monolithic 58% enrichment U8%Mo and U10%Mo are being delivered to INL-DoE to be irradiated in ATR reactor core. A conscientious study on compound interphase formation in both cases is being carried out. CNEA, a worldwide leader on LEU technology for fission radioisotope production is providing Brazil with these radiopharmaceutical products and Egypt and Australia with the technology through INVAP SE. CNEA is also committed to improve the diffusion of LEU target and radiochemical technology for radioisotope production and target and process optimization. Future plans include: 1) Fabrication of a LEU dispersed U-Mo fuel prototype following the recommendations of the IAEA's Good Practices document, to be irradiated in a high flux reactor in the frame of the ARG/4/092 IAEA's Technical Cooperation project. 2) Development of LEU very high density monolithic and dispersed U-Mo fuel plates with Zry-4 or Al cladding as a part of the RERTR program. 3) Optimization of LEU target and radiochemical techniques for radioisotope production. (author)

Evaluation of Uranium-235 Measurement Techniques

Energy Technology Data Exchange (ETDEWEB)

Kaspar, Tiffany C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Dibert, Mark W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-05-23

Monolithic U-Mo fuel plates are rolled to final fuel element form from the original cast ingot, and thus any inhomogeneities in ²³⁵U distribution present in the cast ingot are maintained, and potentially exaggerated, in the final fuel foil. The tolerance for inhomogeneities in the ²³⁵U concentration in the final fuel element foil is very low. A near-real-time, nondestructive technique to evaluate the ²³⁵U distribution in the cast ingot is required in order to provide feedback to the casting process. Based on the technical analysis herein, gamma spectroscopy has been recommended to provide a near-real-time measure of the ²³⁵U distribution in U-Mo cast plates.

Stability Study of the RERTR Fuel Microstructure

Energy Technology Data Exchange (ETDEWEB)

Jian Gan; Dennis Keiser; Brandon Miller; Daniel Wachs

2014-04-01

The irradiation stability of the interaction phases at the interface of fuel and Al alloy matrix as well as the stability of the fission gas bubble superlattice is believed to be very important to the U-Mo fuel performance. In this paper the recent result from TEM characterization of Kr ion irradiated U-10Mo-5Zr alloy will be discussed. The focus will be on the phase stability of Mo2-Zr, a dominated second phase developed at the interface of U-10Mo and the Zr barrier in a monolithic fuel plate from fuel fabrication. The Kr ion irradiations were conducted at a temperature of 200 degrees C to an ion fluence of 2.0E+16 ions/cm2. To investigate the thermal stability of the fission gas bubble superlattice, a key microstructural feature in both irradiated dispersion U-7Mo fuel and monolithic U-10Mo fuel, a FIB-TEM sample of the irradiated U-10Mo fuel (3.53E+21 fission/cm3) was used for a TEM in-situ heating experiment. The preliminary result showed extraordinary thermal stability of the fission gas bubble superlattice. The implication of the TEM observation from these two experiments on the fuel microstructural evolution under irradiation will be discussed.

Influence of Fuel-Matrix Interaction on the Deformation of U-Mo Dispersion Fuel

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Chicago (United States)

2014-05-15

In order to predict the fuel plate failure leading to breakaway swelling in the meat, an understanding of the effects of the fuel-matrix interaction behavior on the deformation of fuel meat is necessary. However, the effects of IL formation on the development of breakaway swelling have not been studied thoroughly. A mechanism that explains large pore growth that leads to breakaway swelling has not been included in the existing fuel performance models. In this study, the effect of the fuel-matrix interaction on large interfacial porosity development at the IL-Al interface is analyzed using both mechanistic correlations and observations from the post-irradiation examination results of U-Mo Dispersion fuels. The effects of fuel-matrix interaction on the fuel performance of U-Mo/Al Dispersion fuel were investigated. Fuel-matrix interaction bears the causes for breakaway swelling that can lead to a fuel failure under a high-power irradiation condition. Fission gas atoms are released from U-Mo particles to the interaction layer via diffusion and recoil. The fission gases released from the U-Mo and produced in the ILs are further released to the IL-Al interface by diffusion in the IL and recoil. Large pore formation at the IL-Al interface is attributed to the active diffusion of fission gas atoms in the ILs and coalescence between the small bubbles there. A model calculation showed that IL growth increases the probability of forming a breakaway swelling condition. ILs are connected to each other and the Al matrix decreases as ILs grow. When more ILs are interconnected, breakaway swelling can occur when the effective stress from the fission gas pressure in the IL-Al interfacial pore becomes larger than the yield strength of the Al matrix.

In-pile test results of U-silicide or U-nitride coated U-7Mo particle dispersion fuel in Al

Science.gov (United States)

Kim, Yeon Soo; Park, J. M.; Lee, K. H.; Yoo, B. O.; Ryu, H. J.; Ye, B.

2014-11-01

U-silicide or U-nitride coated U-Mo particle dispersion fuel in Al (U-Mo/Al) was in-pile tested to examine the effectiveness of the coating as a diffusion barrier between the U-7Mo fuel kernels and Al matrix. This paper reports the PIE data and analyses focusing on the effectiveness of the coating in terms of interaction layer (IL) growth and general fuel performance. The U-silicide coating showed considerable success, but it also provided evidence for additional improvement for coating process. The U-nitride coated specimen showed largely inefficient results in reducing IL growth. From the test, important observations were also made that can be utilized to improve U-Mo/Al fuel performance. The heating process for coating turned out to be beneficial to suppress fuel swelling. The use of larger fuel particles confirmed favorable effects on fuel performance.

In-pile test results of U-silicide or U-nitride coated U-7Mo particle dispersion fuel in Al

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Park, J.M.; Lee, K.H.; Yoo, B.O. [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Ryu, H.J. [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Ye, B. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2014-11-15

U-silicide or U-nitride coated U-Mo particle dispersion fuel in Al (U-Mo/Al) was in-pile tested to examine the effectiveness of the coating as a diffusion barrier between the U-7Mo fuel kernels and Al matrix. This paper reports the PIE data and analyses focusing on the effectiveness of the coating in terms of interaction layer (IL) growth and general fuel performance. The U-silicide coating showed considerable success, but it also provided evidence for additional improvement for coating process. The U-nitride coated specimen showed largely inefficient results in reducing IL growth. From the test, important observations were also made that can be utilized to improve U-Mo/Al fuel performance. The heating process for coating turned out to be beneficial to suppress fuel swelling. The use of larger fuel particles confirmed favorable effects on fuel performance.

Characterization of the reaction layer in U-7wt%Mo/Al diffusion couples

Energy Technology Data Exchange (ETDEWEB)

Mirandou, M.I.; Balart, S.N.; Ortiz, M.; Granovsky, M.S. E-mail: granovsk@cnea.gov.ar

2003-11-15

The reaction layer in chemical diffusion couples U-7wt%Mo/Al was investigated using optical and scanning electron microscopy, electron probe microanalysis and X-ray diffraction (XRD) techniques. When the U-7wt%Mo alloy was previously homogenized and the {gamma}(U, Mo) phase was retained, the formation of (U, Mo)Al{sub 3} and (U, Mo)Al{sub 4} was observed at 580 deg. C. Also a very thin band was detected close to the Al side, the structure of the ternary compound Al{sub 20}UMo{sub 2} might be assigned to it. When the decomposition of the {gamma}(U, Mo) took place, a drastic change in the diffusion behavior was observed. In this case, XRD indicated the presence of phases with the structures of (U, Mo)Al{sub 3}, Al{sub 43}U{sub 6}Mo{sub 4}, {gamma}(U, Mo) and {alpha}(U) in the reaction layer.

Determination of elastic constants of fuels plates based on uranium by ultrasound testing

International Nuclear Information System (INIS)

Moreira Castro, Martin Ignacio

2015-01-01

Current nuclear reactors use as U-235 U-enriched compounds enriched with U-235, requiring U-alloys that increase the amount of atoms available for nuclear fission in a convenient way. This study was carried out on fuel plates manufactured in the Chilean Nuclear Energy Commission, whose cores are composed of a dispersed mixture Al-U_3Si_2 and Al-U_7Mo, with different densities of uranium, covered by a coating of Al6061. The objective was to characterize elastically and classify the fuel plates analyzed. Specifically, five Al-U_3Si_2 fuel plates with 1.7 gU/cm"3, eight A-U_3Si_2 with 3.4 gU/cm"3, five of A-l U_3Si_2 with 4.8 gU/cm"3 were successfully studied. The apparent elastic constants (Young and Shear modules, and Poisson coefficient) were determined in the area where the fuel is located (MEAT) by means of an ultrasound sampling technique, thus being able to characterize them and classify them according to their composition. The behavior of the elastic constants generally shows a tendency to decrease as the amount of U_3Si_2 particles dispersed in the MEAT zone of the fuel plates increases. In addition, the non-destructive test method used made it possible to detect several differences between the fuel plates analyzed, such as the amount of reduction in rolling, among others. Additionally, six experimental fuel miniplates were analyzed whose meat were formed by a dispersion of the Al-UMo type, specifically: two of Al-U_7Mo with 6.0 gU/cm"3, two of Al-U_7Mo with 7.0 gU/ cm"3 and two of Al-U_7Mo with 8.0 gU/cm"3. The response of the U-Mo fuel miniplates against this technique was not good, so several ideas were proposed to improve this situation

Solid state reactions of MoO3 and Na2MoO4 with (U.85,Ce.15)O2x

International Nuclear Information System (INIS)

Dahale, N.D.; Keskar, Meera; Singh Mudher, K.D.; Chawla, K.L.

1999-01-01

(U .85 ,Ce .15 )MoO 6-x was prepared by the solid state reactions of (U .85 ,Ce .15 )O 2±x with MoO 3 in air at 600 deg C. Solid state reactions of Na 2 MoO 4 with (U .85 ,Ce .15 )MoO 6.x up to 550 deg C in air led to the formation of Na 2 (U .85 ,Ce .15 )Mo 2 O 10-x and Na 2 (U .85 , Ce .15 ) 2 Mo 3 O 16-x . These compounds were characterised by x-ray and thermal methods. The x-ray powder data of (U .85 , Ce .15 ) MoO 6-x were indexed on monoclinic system whereas, data of Na 2 (U .85 ,Ce .15 ) Mo 2 O 10-x and Na 2 (U .85 ,Ce .15 ) 2 Mo 3 O 16-x were indexed on orthorhombic and monoclinic system respectively. (author)

Improving 6061-Al Grain Growth and Penetration across HIP-Bonded Clad Interfaces in Monolithic Fuel Plates: Initial Studies

Energy Technology Data Exchange (ETDEWEB)

Hackenberg, Robert E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McCabe, Rodney J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Montalvo, Joel D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Clarke, Kester D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dvornak, Matthew J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Edwards, Randall L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Crapps, Justin M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Trujillo, R. Ralph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aikin, Beverly [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vargas, Victor D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hollis, Kendall J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lienert, Thomas J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Forsyth, Robert T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Harada, Kiichi L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2013-05-06

Grain penetration across aluminum-aluminum cladding interfaces in research reactor fuel plates is desirable and was obtained by a legacy roll-bonding process, which attained 20-80% grain penetration. Significant grain penetration in monolithic fuel plates produced by Hot Isostatic Press (HIP) fabrication processing is equally desirable but has yet to be attained. The goal of this study was to modify the 6061-Al in such a way as to promote a much greater extent of crossinterface grain penetration in monolithic fuel plates fabricated by the HIP process. This study documents the outcomes of several strategies attempted to attain this goal. The grain response was characterized using light optical microscopy (LOM) electron backscatter diffraction (EBSD) as a function of these prospective process modifications done to the aluminum prior to the HIP cycle. The strategies included (1) adding macroscopic gaps in the sandwiches to enhance Al flow, (2) adding engineering asperities to enhance Al flow, (3) adding stored energy (cold work), and (4) alternative cleaning and coating. Additionally, two aqueous cleaning methods were compared as baseline control conditions. The results of the preliminary scoping studies in all the categories are presented. In general, none of these approaches were able to obtain >10% grain penetration. Recommended future work includes further development of macroscopic grooving, transferred-arc cleaning, and combinations of these with one another and with other processes.

Preparation of U-Si/U-Me (Me = Fe, Ni, Mn) aluminum-dispersion plate-type fuel (miniplates) for capsule irradiation

International Nuclear Information System (INIS)

Ugajin, Mitsuhiro; Itoh, Akinori; Akabori, Mitsuo

1993-06-01

Details of equipment installed, method adopted and final products were described on the preparation of uranium silicides and other fuels for capsule irradiation. Main emphasis was placed on the preparation of laboratory-scale aluminum-dispersion plate-type fuel (miniplates) loaded to the first and second JMTR silicide capsules. Fuels contained in the capsules are as follows: (A) uranium-silicide base alloys U 3 Si 2 , Mo- added U 3 Si 2 , U 3 Si 2 +U 3 Si, U 3 Si 2 +USi, U 3 Si, U 3 (Si 0.8 Ge 0.2 ), U 3 (Si 0.6 Ge 0.4 ) (B) U 6 Me-type alloys with higher uranium density U 6 Mn, U 6 Ni, U 6 (Fe 0.4 Ni 0.6 ), U 6 (Fe 0.6 Mn 0.4 ) The powder-metallurgical picture-frame method was adopted and laboratory-scale technique was established for the preparation of miniplates. As a result of inspection for capsule irradiation, miniplates were prepared to meet the requirements of specification. (author)

Performance of Nb protective diffusion coating on U-Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Ji-Hyeon; Sohn, Dong-Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, Sunghwan; Nam, Ji Min; Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

To achieve this aim, it is necessary to increase the volume fraction of fuel particles inside the meat. However, the technical limit is reached at approximately 55 vol.% of fuel particles in the aluminum matrix. As a solution, an uranium compound with an higher uranium density than existing U3Si2 fuel has to be selected. Also alloying the uranium must stabilize γ-phase of uranium at room temperature because adequate properties of the γ -phase of uranium showed a good irradiation behavior in the past. Hence, U-Mo alloys were selected as the best candidates. The formation of interaction phase is a critical problem to apply U-Mo alloys to the high performance research reactor. Different means have been proposed to reduce the interaction between U-Mo fuel and Al matrix. There are three means. : 1. Addition of a diffusion limiting element to the matrix 2. Insertion of a diffusion barrier at the interface between the U-Mo and the Al 3. Alloying of the U-Mo with a third element Here we present the effect of Nb coating as diffusion barrier on formation of interaction layers between UMo powders and Al matrix. We present the effect of Nb coating on formation of interaction layers between U-Mo powders and Al matrix. Centrifugally atomized U-7 wt.% Mo powders were used, and Nb was coated on the surface of U-7 wt.% Mo by sputtering. Subsequently, the Nb-coated U-7 wt.% Mo powders were mixed with pure Al powders, and were made into compacts. The compacts were annealed at 550 .deg. C for 1, 3, 5 hours, respectively, and the result showed that the Nb coating on U-7 wt.% Mo effectively suppressed the growth of interaction layers between U-7 wt.% Mo and Al matrix.

Finite-element model to predict roll-separation force and defects during rolling of U-10Mo alloys

Science.gov (United States)

Soulami, Ayoub; Burkes, Douglas E.; Joshi, Vineet V.; Lavender, Curt A.; Paxton, Dean

2017-10-01

A major goal of the Convert Program of the U.S. Department of Energy's National Nuclear Security Administration (DOE/NNSA) is to enable high-performance research reactors to operate with low-enriched uranium rather than the high-enriched uranium currently used. To this end, uranium alloyed with 10 wt% molybdenum (U-10Mo) represents an ideal candidate because of its stable gamma phase, low neutron caption cross section, acceptable swelling response, and predictable irradiation behavior. However, because of the complexities of the fuel design and the need for rolled monolithic U-10Mo foils, new developments in processing and fabrication are necessary. This study used a finite-element code, LS-DYNA, as a predictive tool to optimize the rolling process. Simulations of the hot rolling of U-10Mo coupons encapsulated in low-carbon steel were conducted following two different schedules. Model predictions of the roll-separation force and roll pack thicknesses at different stages of the rolling process were compared with experimental measurements. The study reported here discussed various attributes of the rolled coupons revealed by the model (e.g., waviness and thickness non-uniformity like dog-boning). To investigate the influence of the cladding material on these rolling defects, other cases were simulated: hot rolling with alternative can materials, namely, 304 stainless steel and Zircaloy-2, and bare-rolling. Simulation results demonstrated that reducing the mismatch in strength between the coupon and can material improves the quality of the rolled sheet. Bare-rolling simulation results showed a defect-free rolled coupon. The finite-element model developed and presented in this study can be used to conduct parametric studies of several process parameters (e.g., rolling speed, roll diameter, can material, and reduction).

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

Science.gov (United States)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

2018-04-01

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature- and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS). The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code.

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

Energy Technology Data Exchange (ETDEWEB)

Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade; Weiss, Aaron; Howard, Trevor

2017-06-01

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.

Comparison of U-Pu-Mo, U-Pu-Nb, U-Pu-Ti and U-Pu-Zr alloys; Comparaison des alliages U-Pu-Mo, U-Pu-Nb, U-Pu-Ti, U-Pu-Zr

Energy Technology Data Exchange (ETDEWEB)

Boucher, R; Barthelemy, P [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-01

The data concerning the U-Pu, U-Pu-Mo and U-Pu-Nb are recalled. The results obtained with U-Pu-Ti and U-Pu-Zr alloys containing 15-20 per cent Pu and 10 wt. per cent ternary element are reported. The transformation temperatures, the expansion coefficients, the nature of phases, the thermal cycling behaviour have been determined. A list of the principal properties of these different alloys is presented and the possibilities of their use as fast reactor's fuel element are considered. The U-Pu-Ti alloys seem to be quite promising: easiness of fabrication, large thermal stability, excellent behaviour in air, small quantity of zeta phase, temperature of solidus superior to 1100 deg. C. (authors) [French] On rappelle brievement les connaissances acquises sur les alliages U-Pu, U-Pu-Mo et U-Pu-Nb. On presente les resultats obtenus avec les alliages U-Pu-Ti et U-Pu-Zr pour des teneurs de 15 a 20 pour cent de plutonium et 10 pour cent en poids d'element ternaire. On a determine les temperatures de transformation, les coefficients de dilatation, la nature des phases, la conductibilite thermique a 20 deg. C, la tenue au cyclage thermique et diverses autres proprietes. Un tableau resume les principales proprietes des divers alliages. On considere les possibilites d'emploi de ces alliages comme combustibles de reacteur rapide. Les alliages U-Pu-Ti paraissent particulierement interessants: facilite d'elaboration, stabilite thermique etendue, tenue dans l'air excellente, faible quantite de la phase U-Pu zeta, temperature de fusion commencante superieure a 1100 deg. C. (auteurs)

The Effect of Uncertainties on the Operating Temperature of U-Mo/Al Dispersion Fuel

Energy Technology Data Exchange (ETDEWEB)

Sweidana, Faris B.; Mistarihia, Qusai M.; Ryu Ho Jin [KAIST, Daejeon (Korea, Republic of); Yim, Jeong Sik [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this study, uncertainty and combined uncertainty studies have been carried out to evaluate the uncertainty of the parameters affecting the operational temperature of U-Mo/Al fuel. The uncertainties related to the thermal conductivity of fuel meat, which consists of the effects of thermal diffusivity, density and specific heat capacity, the interaction layer (IL) that forms between the dispersed fuel and the matrix, fuel plate dimensions, heat flux, heat transfer coefficient and the outer cladding temperature were considered. As the development of low-enriched uranium (LEU) fuels has been pursued for research reactors to replace the use of highly-enriched uranium (HEU) for the improvement of proliferation resistance of fuels and fuel cycle, U-Mo particles dispersed in an Al matrix (UMo/Al) is a promising fuel for conversion of the research reactors that currently use HEU fuels to LEUfueled reactors due to its high density and good irradiation stability. Several models have been developed for the estimation of the thermal conductivity of U–Mo fuel, mainly based on the best fit of the very few measured data without providing uncertainty ranges. The purpose of this study is to provide a reasonable estimation of the upper bounds and lower bounds of fuel temperatures with burnup through the evaluation of the uncertainties in the thermal conductivity of irradiated U-Mo/Al dispersion fuel. The combined uncertainty study using RSS method evaluated the effect of applying all the uncertainty values of all the parameters on the operational temperature of U-Mo/Al fuel. The overall influence on the value of the operational temperature is 16.58 .deg. C at the beginning of life and it increases as the burnup increases to reach 18.74 .deg. C at a fuel meat fission density of 3.50E+21 fission/cm{sup 3}. Further studies are needed to evaluate the behavior more accurately by including other parameters uncertainties such as the interaction layer thermal conductivity.

Mo-99 production by fission and future projections

International Nuclear Information System (INIS)

Carranza, E.C.; Novello, A.; Bronca, M.; Cestau, D.; Bavaro, R.; Centurion, R.; Bravo, C.; Bronca, P.; Gualda, E.; Fraguas, F.; Giomi, A.; Ivaldi, L.

2012-01-01

Description of the I-131 and Mo-99 production process: The process starts with the irradiation of uranium-aluminum mini plates in the RA-3, Argentinean Reactor No.3, Ezeiza Atomic Center. In a nuclear reactor there is a constant flow of neutrons and when a neutron with proper energy impacts on a nucleus of U-235, it is absorbed at the same time generate an unstable configuration nuclear. For this reason, the nucleus formed is fission, getting two different atoms. Approximately 6% of the fissions produce Mo-99 and 3% produce I-131; the percentage remaining corresponds to formation of atoms without interest for use in medicine. In conclusion, the objective of the process developed in the Fission Plant, is starting from uranium mini plates, separate the Mo-99 and I-131 generated, the remaining elements formed. - Evolution of Mo-99 Production in the last 10 years: The Fission Mo-99 Plant Production begins routine production of Mo-99 in 1985, using targets made of uranium enriched at 90% U-235. In the 1990s, global concern regarding the use of highly enriched uranium, due to non-proliferation issues, caused the interruption of supply of nuclear material (HEU enriched at 90% of U-235). Following this, Argentina developed target based on low-enriched uranium (less than 20% U-235), becoming in 2002 the first country in the world to produce Mo-99 with LEU targets. From 2002 to date, the activity produced of Mo-99 has been tripled annually (author)

An Effort to Improve U Foil Fabrication Technology of Roll-casting for Fission Mo Target

Energy Technology Data Exchange (ETDEWEB)

Kim, Chang Kyu; Woo, Yun Myeong; Kim, Ki Hwan; Oh, Jong Myeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Sim, Moon Soo [Chungnam University, Green Energy Technology, Daejeon (Korea, Republic of)

2010-10-15

Mo-99 isotope has been produced mainly by extracting fission products of {sup 235}U. The targets for irradiating in reactor have used as stainless tube coated with highly enriched UO{sub 2} at the inside surface and highly enriched UAlx plate cladded with aluminum. In connection with non-proliferation policy the RERTR program developed a new process of Mo-99 using low enriched uranium (LEU) instead of highly enriched uranium (HEU). LEU should be put about five times more quantity than HEU because the {sup 235}U contents of LEU and HEU are 20% and higher than 90%, respectively. Accordingly pure uranium metal foil target was adopted as a promising target material due to high uranium density. ANL and BATAN developed a Cintichem process using uranium metal foil target of 130 {mu}m in thickness jointly and the RERTR program is trying to disseminate the new process world-widely. However, uranium foil is made by lots of times rolling work on uranium plate, which is laborious and tedious. In order to avoid this difficulty KAERI developed a new process of making foil directly from uranium melt by roll casting. This process is very much simple, productive, and cost-effective. But the outside surface of foil is generally very rough. A typical transverse cross section had a minimum thickness of 65 {mu}m and a maximum thickness of 205 {mu}m. This roughness could affect (1) target fabrication, where the U foil, or the Ni foil might be damaged during drawing, and (2) irradiation behavior, where gaps between the target walls and the U metal might affect cooling of the target

Low-enriched uranium-molybdenum fuel plate development

International Nuclear Information System (INIS)

Wiencek, T.C.; Prokofiev, I.G.

2000-01-01

To examine the fabricability of low-enriched uranium-molybdenum powders, full-size 450 x 60 x 0.5-mm (17.7 x 2.4 x 0.020-in.) fuel zone test plates loaded to 6 g U/cm 3 were produced. U-10 wt.% Mo powders produced by two methods, centrifugal atomization and grinding, were tested. These powders were supplied at no cost to Argonne National Laboratory by the Korean Atomic Energy Research Institute and Atomic Energy of Canada Limited, respectively. Fuel homogeneity indicated that both of the powders produced acceptable fuel plates. Operator skill during loading of the powder into the compacting die and fuel powder morphology were found to be important when striving to achieve homogeneous fuel distribution. Smaller, 94 x 22 x 0.6-mm (3.7 x 0.87 x 0.025-in.) fuel zone, test plates were fabricated using U-10 wt.% Mo foil disks instead of a conventional powder metallurgy compact. Two fuel plates of this type are currently undergoing irradiation in the RERTR-4 high-density fuel experiment in the Advanced Test Reactor. (author)

Kinetics of the U-1% Mo alloy transformation during continual cooling; Kinetika transformacije legura U-1% Mo pri kontinuiranom hladjenju

Energy Technology Data Exchange (ETDEWEB)

Mihajlovic, A; Djuric, B; Tepavac, P [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-11-15

Study of continuous cooling of the U-1% Mo alloy is significant if it could be used as fuel in the nuclear reactor. Previous studies were dealing with relatively low cooling rate up to 3 deg C/s{sup 1}, which produced alpha + gamma structure. This task was devoted to testing the U-1% Mo alloy properties at higher cooling rates in order to discover whether bainite reaction and favourable alpha grain could be achieved under certain conditions.

Study of relationships between microstructures and service properties, of U(Mo) fissile alloys particles

International Nuclear Information System (INIS)

Champion, G.

2013-01-01

This thesis enters in the Material and Testing Reactors (MTRs) framework where the necessity to use a Low- Enriched Uranium (LEU) fuel has led to the development of a dense fissile material based on U(Mo) alloys. The designed fuel is a composite material, made of dispersed U(Mo) particles embedded in an Al based matrix. Post- Irradiation Examinations of these LEU fuel plates showed that the irradiation behaviour of the fuel is not fit for purpose yet. This is mainly due to the growth of an interaction layer between the fuel and the matrix and to the bad gas retention efficiency of the fuel particles. This thesis had for purpose the development of several solutions in order to modify and/or decrease or even inhibit the fuel/matrix interaction and to increase the gas retention capacities of the fuel. In order to achieve so, two solutions have been tested during this thesis, (i) optimization of the U(Mo) alloy intrinsic microstructural properties and (ii) modification of the fuel meat/matrix interface, through the deposition of a layer acting as a 'diffusion barrier'. Concerning the first axis of study, a characterization campaign of the reference powders has been performed, as a first step, in order to identify the key parameters for the development of products showing an 'optimized' microstructure. Two novel products have then been developed: one based on a combined process associating 'atomization + grinding' and another, which consists in a magnesiothermy process. These products were subjected to characterization: X-Ray and neutron diffraction, electron backscattered diffraction and transmission electron microscopy have been performed in particular. We managed to show that these powders can be an advantage concerning the issue with the gas retention capacities of the fuel. Concerning the growth of the interaction layer, a third product has been developed: an U(Mo) atomized powder, coated with an alumina layer. We managed to show that a thickness between 100 and

Corrosion report for the U-Mo fuel concept

Energy Technology Data Exchange (ETDEWEB)

Henager, Charles H. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Bennett, Wendy D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Doherty, Ann L. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Fuller, E. S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hardy, John S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

2014-08-28

The Fuel Cycle Research and Development (FCRD) program of the Office of Nuclear Energy (NE) has implemented a program to develop a Uranium-Molybdenum (U-Mo) metal fuel for Light Water Reactors (LWR)s. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties, which includes high thermal conductivity for less stored heat energy. With sufficient development, it may be able to provide the Light Water industry with a melt-resistant accident tolerant fuel with improved safety response. However, the corrosion of this fuel in reactor water environments needs to be further explored and optimized by additional alloying. The Pacific Northwest National Laboratory has been tasked with performing ex-reactor corrosion testing to characterize the performance of U-Mo fuel. This report documents the results of the effort to characterize and develop the U-Mo metal fuel concept for LWRs with regard to corrosion testing. The results of a simple screening test in buffered water at 30°C using surface alloyed U-10Mo is documented and discussed. The screening test was used to guide the selection of several potential alloy improvements that were found and are recommended for further testing in autoclaves to simulate PWR water conditions more closely.

Development of dispersion U(Mo)/Al–Si miniplates fabricated at 500 °C with Al 6061 as cladding

Energy Technology Data Exchange (ETDEWEB)

Mirandou, M.I., E-mail: mirandou@cnea.gov.ar [Gerencia Materiales-GAEN-CNEA, Avda. Gral. Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina); Aricó, S.F. [Gerencia Materiales-GAEN-CNEA, Avda. Gral. Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina); Instituto Sabato UNSAM-CNEA, Avda. Gral. Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina); Balart, S.N. [Gerencia Materiales-GAEN-CNEA, Avda. Gral. Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina); Fabro, J.O. [Departamento ECRI, Gerencia de Ciclo del Combustible Nuclear, CNEA, Avda. Gral. Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina)

2015-02-15

In the frame of U(Mo) dispersion fuel elements qualification, Si additions to Al matrix arose as a promising solution to the unacceptable failures found when pure Al is used. Analysis of as-fabricated fuel plates made with Al–Si matrices demonstrated that good irradiation behavior is correlated with the formation during fabrication of a Si-containing interaction layer around the U(Mo) particles. Thus, the analysis of the influence of fabrication parameters becomes important. Studies on Al–Si dispersion miniplates fabricated in CNEA, Argentina, have been initiated to determine how to obtain the better interaction layer characteristics with the lesser modifications to the fabrication process and the smaller amount of Si in the matrix. In this work results for miniplates made of atomized U–7 wt%Mo particles dispersed in Al–2 wt%Si and Al–4 wt%Si matrices, obtained by mixing pure Al and Si powders, and Al 6061 as cladding are presented. Interaction layer grown during fabrication process (500 °C) consists of Si-containing phases being U(Al, Si){sub 3} its principal component. Its uniformity is not satisfactory due to the formation of an oxide layer.

Update on Fresh Fuel Characterization of U-Mo Alloys

International Nuclear Information System (INIS)

Burkes, D.E.; Wachs, D.M.; Keiser, D.D.; Okuniewski, M.A.; Jue, J.F.; Rice, F.J.; Prabhakaran, R.

2009-01-01

The need to provide more accurate property information on U-Mo fuel alloys to operators, modellers, researchers, fabricators, and government increases as success of the GTRI Reactor Convert program continues. This presentation provides an update on fresh fuel characterization activities that have occurred at the INL since the RERTR 2008 conference in Washington, D.C. The update is particularly focused on properties recently obtained and on the development progress of new measurement techniques. Furthermore, areas where useful and necessary information is still lacking is discussed. The update deals with mechanical, physical, and microstructural properties for both integrated and separate effects. Appropriate discussion of fabrication characteristics, impurities, thermodynamic response, and effects on the topic areas are provided, along with a background on the characterization techniques used and developed to obtain the information. Efforts to measure similar characteristics on irradiated fuel plates are discussed.

Fuel performance analysis for the HAMP-1 mini plate test

Energy Technology Data Exchange (ETDEWEB)

Cho, Byoung Jin; Tahka, Y. W.; Yim, J. S.; Lee, B. H. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

U-7wt%Mo/Al- 5wt%Si dispersion fuel with 8gU/cm{sup 3} is chosen to achieve more efficiency and higher performance than the conventional U{sub 3}Si{sub 2} fuel. As part of the fuel qualification program for the KiJang research reactor (KJRR), three irradiation tests with mini-plates are on the way at the High-flux Advanced Neutron Application Reactor (HANARO). The first test among three HANARO Mini-Plate Irradiation tests (HAMP-1, 2, 3) has completed. PLATE code has been initially developed to analyze the thermal performance of high density U-Mo/Al dispersion fuel plates during irradiation [1]. We upgraded the PLATE code with the latest irradiation results which were implemented by corrosion, thermal conductivity and swelling model. Fuel performance analysis for HAMP-1 was conducted with updated PLATE. This paper presents results of performance evaluation of the HAMP-1. Maximum fuel temperature was obtained 136 .deg., which is far below the preset limit of 200 .deg. for the irradiation test. The meat swelling and corrosion thickness was also confirmed that the developed fuel would behave as anticipated.

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

2018-04-01

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature-and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS).The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code. (c) 2018 Elsevier B.V. All rights reserved.

Thermophysical properties of heat-treated U-7Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Cho, Tae Won; Kim, Yeon Soo; Park, Jong Man; Lee, Kyu Hong; Kim, Sunghwan; Lee, Chong Tak; Yang, Jae Ho; Oh, Jang Soo; Sohn, Dong-Seong

2018-04-01

In this study, the effects of interaction layer (IL) on thermophysical properties of U-7Mo/Al dispersion fuel were examined. Microstructural analyses revealed that ILs were formed uniformly on U-Mo particles during heating of U-7Mo/Al samples. The IL volume fraction was measured by applying image analysis methods. The uranium loadings of the samples were calculated based on the measured meat densities at 298 K. The density of the IL was estimated by using the measured density and IL volume fraction. Thermal diffusivity and heat capacity of the samples after the heat treatment were measured as a function of temperature and volume fractions of U-Mo and IL. The thermal conductivity of IL-formed U-7Mo/Al was derived by using the measured thermal diffusivity, heat capacity, and density. The thermal conductivity obtained in the present study was lower than that predicted by the modified Hashin–Shtrikman model due to the theoretical model’s inability to consider the thermal resistance at interfaces between the meat constituents.

Thermophysical properties of heat-treated U-7Mo/Al dispersion fuel

Science.gov (United States)

Cho, Tae Won; Kim, Yeon Soo; Park, Jong Man; Lee, Kyu Hong; Kim, Sunghwan; Lee, Chong Tak; Yang, Jae Ho; Oh, Jang Soo; Sohn, Dong-Seong

2018-04-01

In this study, the effects of interaction layer (IL) on thermophysical properties of U-7Mo/Al dispersion fuel were examined. Microstructural analyses revealed that ILs were formed uniformly on U-Mo particles during heating of U-7Mo/Al samples. The IL volume fraction was measured by applying image analysis methods. The uranium loadings of the samples were calculated based on the measured meat densities at 298 K. The density of the IL was estimated by using the measured density and IL volume fraction. Thermal diffusivity and heat capacity of the samples after the heat treatment were measured as a function of temperature and volume fractions of U-Mo and IL. The thermal conductivity of IL-formed U-7Mo/Al was derived by using the measured thermal diffusivity, heat capacity, and density. The thermal conductivity obtained in the present study was lower than that predicted by the modified Hashin-Shtrikman model due to the theoretical model's inability to consider the thermal resistance at interfaces between the meat constituents.

Improvement of Silicide Coating Method as Diffusion Barrier for U-Mo Dispersion Fuel

Energy Technology Data Exchange (ETDEWEB)

Nam, Ji Min; Kim, Sunghwan; Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The excessive interaction between the U-Mo alloys and their surrounding Al matrix lead to excessive local swelling called 'pillowing'. For this reason, KAERI suggested several remedies such as alloying U-Mo with Ti, or Al matrix with Si. In addition, silicide or nitride coatings on the surface of U-Mo particles have also been proposed to hinder the growth of the interaction layer. In this study, centrifugally atomized U-Mo-Ti alloy powders were coated with silicide layers. The coating process was improved when compared to the previous coating in terms of the ball milling and heat treatment conditions. Subsequently, silicide coated U-Mo-Ti powders and pure aluminum powders were mixed and made into a compact for the annealing test. The compacts were annealed at 550 .deg. C for 2hr, and characterized using scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDS). 1. Uniform, homogeneous, thickness controllable silicide layers were successfully coated on the surface of U-7wt%Mo-1wt%Ti powders. 2. U{sub 3}Si, U{sub 3}Si{sub 2} silicide layers formed on the surface of U-7wt%Mo-1wt%Ti powders, and were identified by XRD and EDS analyses.

The irradiation behavior of atomized U-Mo alloy fuels at high temperature

Science.gov (United States)

Park, Jong-Man; Kim, Ki-Hwan; Kim, Chang-Kyu; Meyer, M. K.; Hofman, G. L.; Strain, R. V.

2001-04-01

Post-irradiation examinations of atomized U-10Mo, U-6Mo, and U-6Mo-1.7Os dispersion fuels from the RERTR-3 experiment irradiated in the Advanced Test Reactor (ATR) were carried out in order to investigate the fuel behavior of high uranium loading (8 gU/cc) at a high temperature (higher than 200°C). It was observed after about 40 at% BU that the U-Mo alloy fuels at a high temperature showed similar irradiation bubble morphologies compared to those at a lower temperature found in the RERTR-1 irradiation result, but there was a thick reaction layer with the aluminum matrix which was found to be greatly affected by the irradiation temperature and to a lesser degree by the fuel composition. In addition, the chemical analysis for the irradiated U-Mo fuels using the Electron Probe Micro Analysis (EPMA) method were conducted to investigate the compositional changes during the formation of the reaction product.

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

Energy Technology Data Exchange (ETDEWEB)

Collette, R. [Colorado School of Mines, Nuclear Science and Engineering Program, 1500 Illinois St, Golden, CO 80401 (United States); King, J., E-mail: kingjc@mines.edu [Colorado School of Mines, Nuclear Science and Engineering Program, 1500 Illinois St, Golden, CO 80401 (United States); Buesch, C. [Oregon State University, 1500 SW Jefferson St., Corvallis, OR 97331 (United States); Keiser, D.D.; Williams, W.; Miller, B.D.; Schulthess, J. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2016-07-15

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends when comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. The results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program. - Highlights: • Automated image processing is used to extract fission gas bubble data from irradiated U−Mo fuel samples. • Verification and validation tests are performed to ensure the algorithm's accuracy. • Fission bubble parameters are predictably difficult to compare across samples of varying compositions. • The 2-D results suggest the need for more homogenized fuel sampling in future studies. • The results also demonstrate the value of 3-D reconstruction techniques.

Phase transformation of metastable cubic γ-phase in U-Mo alloys

International Nuclear Information System (INIS)

Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Dey, G.K.; Kamath, H.S.

2010-01-01

Over the past decade considerable efforts have been put by many fuel designers to develop low enriched uranium (LEU 235 ) base U-Mo alloy as a potential fuel for core conversion of existing research and test reactors which are running on high enriched uranium (HEU > 85%U 235 ) fuel and also for the upcoming new reactors. U-Mo alloy with minimum 8 wt% molybdenum shows excellent metastability with cubic γ-phase in cast condition. However, it is important to characterize the decomposition behaviour of metastable cubic γ-uranium in its equilibrium products for in reactor fuel performance point of view. The present paper describes the phase transformation behaviour of cubic γ-uranium phase in U-Mo alloys with three different molybdenum compositions (i.e. 8 wt%, 9 wt% and 10 wt%). U-Mo alloys were prepared in an induction melting furnace and characterized by X-ray diffraction (XRD) method for phase determination. Microstructures were developed for samples in as cast condition. The alloys were hot rolled in cubic γ-phase to break the cast structure and then they were aged at 500 o C for 68 h and 240 h, so that metastable cubic γ-uranium will undergo eutectoid decomposition to form equilibrium phases of orthorhombic α-uranium and body centered tetragonal U 2 Mo intermetallic compound. U-Mo alloy samples with different ageing history were then characterized by XRD for phase and development of microstructure.

Characterization of hierarchical α-MoO3 plates toward resistive heating synthesis: electrochemical activity of α-MoO3/Pt modified electrode toward methanol oxidation at neutral pH

Science.gov (United States)

Filippo, Emanuela; Baldassarre, Francesca; Tepore, Marco; Guascito, Maria Rachele; Chirizzi, Daniela; Tepore, Antonio

2017-05-01

The growth of MoO3 hierarchical plates was obtained by direct resistive heating of molybdenum foils at ambient pressure in the absence of any catalysts and templates. Plates synthesized after 60 min resistive heating typically grow in an single-crystalline orthorhombic structure that develop preferentially in the [001] direction, and are characterized by high resolution transmission electron microscopy, selected area diffraction pattern and Raman-scattering measurements. They are about 100-200 nm in thickness and a few tens of micrometers in length. As heating time proceeds to 80 min, plates of α-MoO3 form a branched structure. A more attentive look shows that primary plates formed at until 60 min could serve as substrates for the subsequent growth of secondary belts. Moreover, a full electrochemical characterization of α-MoO3 plates on platinum electrodes was done by cyclic voltammetric experiments, at pH 7 in phosphate buffer, to probe the activity of the proposed composite material as anode to methanol electro-oxidation. Reported results indicate that Pt MoO3 modified electrodes are appropriate to develop new an amperometric non-enzymatic sensor for methanol as well as to make anodes suitable to be used in direct methanol fuel cells working at neutral pH.

Effects of Silicide Coating on the Interdiffusion between U-7Mo and Al

Energy Technology Data Exchange (ETDEWEB)

Nam, Ji Min; Kim, Ji Hyun; Kim, Sunghwan; Lee, Kyu Hong; Park, Jong Man; Jeong, Yong Jin; Kim, Ki Nam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The excessive interaction between the U-Mo alloys and their surrounding Al matrix lead to and excessive local swelling called 'pillowing'. For this reason, KAERI suggested several remedies such as alloying U-Mo with Ti, or Al matrix with Si. In addition, silicide, or nitride coatings on the surface of U-Mo particles have also been proposed to hinder the growth of interaction layer. In this study, centrifugally atomized U-7Mo alloy powders were coated with silicide layers at varying T (T = 900 and 1000 .deg. C) for 30 min, respectively. U-Mo alloy powder was blended with Si powders and subsequently heat-treated to form uranium-silicide coating layers on the surface of U-Mo alloy particles. For an annealing test, silicide-coated U-Mo alloy powders were made into a compact, and Al powders were used as a matrix. From EDS results, transformed uranium aluminide intermetallic compounds were mainly U(Al,Si)3. U(Al,Si)3 phase left the silicide coating layer behind, and formed inside of U-7Mo particles, as shown in Fig. 3(a) and (b). In the case of sample B, Al could not penetrate the silicide coating layer and the coating layers were remained constant, as shown in Fig. 3(c) and (d). From the results, we made a comparison between the compacts of sample A and B, and it was shown that Al can easily diffuse into unreacted Si and U{sub 3}Si{sub 5} mixed layer while U{sub 3}Si{sub 2} acted as a good diffusion barrier at 550 .deg. C though those layers had the same thickness.

Effects of Silicide Coating on the Interdiffusion between U-7Mo and Al

International Nuclear Information System (INIS)

Nam, Ji Min; Kim, Ji Hyun; Kim, Sunghwan; Lee, Kyu Hong; Park, Jong Man; Jeong, Yong Jin; Kim, Ki Nam

2015-01-01

The excessive interaction between the U-Mo alloys and their surrounding Al matrix lead to and excessive local swelling called 'pillowing'. For this reason, KAERI suggested several remedies such as alloying U-Mo with Ti, or Al matrix with Si. In addition, silicide, or nitride coatings on the surface of U-Mo particles have also been proposed to hinder the growth of interaction layer. In this study, centrifugally atomized U-7Mo alloy powders were coated with silicide layers at varying T (T = 900 and 1000 .deg. C) for 30 min, respectively. U-Mo alloy powder was blended with Si powders and subsequently heat-treated to form uranium-silicide coating layers on the surface of U-Mo alloy particles. For an annealing test, silicide-coated U-Mo alloy powders were made into a compact, and Al powders were used as a matrix. From EDS results, transformed uranium aluminide intermetallic compounds were mainly U(Al,Si)3. U(Al,Si)3 phase left the silicide coating layer behind, and formed inside of U-7Mo particles, as shown in Fig. 3(a) and (b). In the case of sample B, Al could not penetrate the silicide coating layer and the coating layers were remained constant, as shown in Fig. 3(c) and (d). From the results, we made a comparison between the compacts of sample A and B, and it was shown that Al can easily diffuse into unreacted Si and U 3 Si 5 mixed layer while U 3 Si 2 acted as a good diffusion barrier at 550 .deg. C though those layers had the same thickness

A plate-on-plate sandwiched Z-scheme heterojunction photocatalyst: BiOBr-Bi{sub 2}MoO{sub 6} with enhanced photocatalytic performance

Energy Technology Data Exchange (ETDEWEB)

Wang, Shengyao [College of Science, Huazhong Agricultural University, Wuhan 430070 (China); Key Laboratory of Environment Correlative Dietology, Ministry of Education, Wuhan 430070 (China); Yang, Xianglong; Zhang, Xuehao; Ding, Xing; Yang, Zixin [College of Science, Huazhong Agricultural University, Wuhan 430070 (China); Dai, Ke [College of Resources and Environment, Huazhong Agricultural University, Wuhan 430070 (China); Chen, Hao, E-mail: hchenhao@mail.hzau.edu.cn [College of Science, Huazhong Agricultural University, Wuhan 430070 (China); Key Laboratory of Environment Correlative Dietology, Ministry of Education, Wuhan 430070 (China)

2017-01-01

Highlights: • A visible light heterojunction photocatalyst of BiOBr-Bi{sub 2}MoO{sub 6} was simply synthesized. • Carriers transferred efficiently in sandwiched layers causing an enhance activity. • A possible direct Z-scheme charge transfer mechanism of BiOBr-Bi2MoO6 is proposed. - Abstract: In this study, a direct Z-scheme heterojunction BiOBr-Bi{sub 2}MoO{sub 6} with greatly enhanced visible light photocatalytic performance was fabricated via a two-step coprecipitation method. It was indicated that a plate-on-plate heterojunctions be present between BiOBr and Bi{sub 2}MoO{sub 6} through different characterization techniques including X-ray powder diffraction (XRD), scanning electron microscopy (SEM), transmission electron microscopy (TEM), UV–vis diffuse reflectance spectroscopy (DRS) and photoelectrochemical measurements. The crystal structure and morphology analysis revealed that the heterointerface in BiOBr-Bi{sub 2}MoO{sub 6} occurred mainly on the (001) facets of BiOBr and (001) facets of Bi{sub 2}MoO{sub 6}. The photocatalytic activity of the BiOBr-Bi{sub 2}MoO{sub 6} was investigated by degradation of RhB and about 66.7% total organic carbon (TOC) could be removed. Ciprofloxacin (CIP) was employed to rule out the photosensitization. It was implied that the higher activity of BiOBr-Bi{sub 2}MoO{sub 6} could be attribute to the strong redox ability in the Z-scheme system, which was subsequently confirmed by photoluminescence spectroscopy (PL) and active spices trapping experiments. This study provides a promising platform for Z-scheme heterojunction constructing and also sheds light on highly efficient visible-light-driven photocatalysts designing.

CHARACTERIZATION OF MONOLITHIC FUEL FOIL PROPERTIES AND BOND STRENGTH

International Nuclear Information System (INIS)

D E Burkes; D D Keiser; D M Wachs; J S Larson; M D Chapple

2007-01-01

Understanding fuel foil mechanical properties, and fuel/cladding bond quality and strength in monolithic plates is an important area of investigation and quantification. Specifically, what constitutes an acceptable monolithic fuel--cladding bond, how are the properties of the bond measured and determined, and what is the impact of fabrication process or change in parameters on the level of bonding? Currently, non-bond areas are quantified employing ultrasonic determinations that are challenging to interpret and understand in terms of irradiation impact. Thus, determining mechanical properties of the fuel foil and what constitutes fuel/cladding non-bonds is essential to successful qualification of monolithic fuel plates. Capabilities and tests related to determination of these properties have been implemented at the INL and are discussed, along with preliminary results

A Study on Silicide Coatings as Diffusion barrier for U-7Mo Dispersion Fuel

Energy Technology Data Exchange (ETDEWEB)

Won, Ju Jin; Kim, Sung Hwan; Lee, Kyu Hong; Jeong, Yong Jin; Kim, Ki Nam; Park, Jong Man; Lee, Chong Tak [KAERI, Daejeon (Korea, Republic of)

2016-05-15

Gamma phase U-Mo alloys are regarded as one of the promising candidates for advanced research reactor fuel when it comes to the irradiation performance. However, it has been reported that interaction layer formation between the UMo alloys and Al matrix degrades the irradiation performance of U-Mo dispersion fuel. The excessive interaction between the U-Mo alloys and their surrounding Al matrix lead to excessive local swelling called 'pillowing'. For this reason, KAERI suggested several remedies such as alloying U-Mo with Al matrix with Si. In addition, silicide or nitride coatings on the surface of U-Mo particles have also been proposed to hinder the growth of the interaction layer. In this study, centrifugally atomized U-7Mo alloy powders were coated with silicide layers at 900 .deg. C for 1hr. U-Mo alloy powder was mixed with MoSi{sub 2}, Si and ZrSi{sub 2} powders and subsequently heat-treated to form uranium-silicide coating layers on the surface of U-Mo alloy particles. Silicide coated U-Mo powders and characterized using scanning electron microscopy (SEM), energy dispersive x-ray spectroscopy (EDS) and X-ray diffractometer (XRD). The ZrSi{sub 2} coating layers has a thickness of about 1∼ 2μm. The surface of a silicide coated particle was very rough and silicide powder attached to the surface of the coating layer. 3. The XRD analysis of the coating layers showed that, they consisted of compounds such as U3Si{sub 2}, USi{sub 2}.

Structural instability and ground state of the U_2Mo compound

International Nuclear Information System (INIS)

Losada, E.L.; Garcés, J.E.

2015-01-01

This work reports on the structural instability at T = 0 °K of the U_2Mo compound in the C11_b structure under the distortion related to the C_6_6 elastic constant. The electronic properties of U_2Mo such as density of states (DOS), bands and Fermi surface (FS) are studied to understand the source of the instability. The C11_b structure can be interpreted as formed by parallel linear chains along the z-directions each one composed of successive U–Mo–U blocks. Hybridization due to electronic interactions inside the U–Mo–U blocks is slightly modified under the D_6 distortion. The change in distance between chains modifies the U–U interaction and produces a split of f-states. The distorted structure is stabilized by a decrease in energy of the hybridized states, mainly between d-Mo and f-U states, together with the f-band split. Consequently, an induced Peierls distortion is produced in U_2Mo due to the D_6 distortion. It is important to note that the results of this work indicate that the structure of the ground state of the U_2Mo compound is not the assumed C11_b structure. It is suggested for the ground state a structure with hexagonal symmetry (P6 #168), ∼0.1 mRy below the energy of the recently proposed Pmmn structure. - Highlights: • Structural instability of the C11b compound due to the D6 deformation. • Induced Peierls distortion due to the D6 deformation. • Distorted structure is stabilized by hybridization and split of f-Uranium state. • P6 (#168) suggested ground state for the U_2Mo compound.

Reaction layer growth and reaction heat of U-Mo/Al dispersion fuels using centrifugally atomized powders

International Nuclear Information System (INIS)

Ryu, Ho Jin; Han, Young Soo; Park, Jong Man; Park, Soon Dal; Kim, Chang Kyu

2003-01-01

The growth behavior of reaction layers and heat generation during the reaction between U-Mo powders and the Al matrix in U-Mo/Al dispersion fuels were investigated. Annealing of 10 vol.% U-10Mo/Al dispersion fuels at temperatures from 500 to 550 deg. C was carried out for 10 min to 36 h to measure the growth rate and the activation energy for the growth of reaction layers. The concentration profiles of reaction layers between the U-10Mo vs. Al diffusion couples were measured and the integrated interdiffusion coefficients were calculated for the U and Al in the reaction layers. Heat generation of U-Mo/Al dispersion fuels with 10-50 vol.% of U-Mo fuel during the thermal cycle from room temperature to 700 deg. C was measured employing the differential scanning calorimetry. Exothermic heat from the reaction between U-Mo and the Al matrix is the largest when the volume fraction of U-Mo fuel is about 30 vol.%. The unreacted fraction in the U-Mo powders increases as the volume fraction of U-Mo fuel increases from 30 to 50 vol.%

A survey of the mechanical properties of uranium alloys U-5Mo-3Nb wt.% and U-3Mo-3Nb wt.%

Energy Technology Data Exchange (ETDEWEB)

Dupont, G.

1969-04-15

In a continuing program on the development of soft and ductile uranium alloys for armament applications, two compositions were studied. These gamma extruded uranium alloys were U-5Mo-3Nb wt.% and U-3Mo-3Nb wt.%. This study was carried out to determine the influence of tempering heat treatments associated with extrusion on the ductility of these uranium alloys. The mechanical properties of both alloys were measured in the extruded condition, in the extruded and annealed condition and in the quenched and tempered condition. A maximum elongation of 13.7% in tension with a low amount of work hardening was obtained for the U-3Mo-3Nb wt.% alloy after 1 1/2 hours anneal at 1200 deg F (650 deg C) followed by a rapid cooling in water at 70 deg F (21 deg C). A maximum elongation of 17.3% with a large amount of work hardening was obtained for alloy U-5Mo-3Nb wt.% after vacuum annealing, normalizing, gamma phase solubilizing at 1500 deg F (815 deg C) and quenching in water at 700 deg F (210 deg C). The maximum ductility achieved in these two alloys by our approaches is low compared with the ductility of Armco Iron employed for the same applications in the field of ballistics.

An alternative route for the preparation of the medical isotope 99Mo from the 238U(γ, f) and 100Mo(γ, n) reactions

International Nuclear Information System (INIS)

Naik, H.; Goswami, A.; Suryanarayana, S.V.; Jagadeesan, K.C.; Thakare, S.V.; Joshi, P.V.; Nimje, V.T.; Mittal, K.C.; Venugopal, V.; Kailas, S.

2013-01-01

The radionuclide 99 Mo, which has a half-life of 65.94 h was produced from 238 U(γ, f) and 100 Mo(γ, n) reactions using a 10 MeV electron linac at EBC, Kharghar Navi-Mumbai, India. This has been investigated since the daughter product 99m Tc is very important from a medical point of view and can be produced in a generator from the parent 99 Mo. The activity of 99 Mo was analyzed by a γ-ray spectrometric technique using a HPGe detector. From the detected γ-rays activity of 140.5 and 739.8 keV, the amount of 99 Mo produced was determined. For comparison, the amount of 99 Mo from 238 U(γ, f) and 100 Mo(γ, n) reactions was also estimated using the experimental photon flux from 197 Au(γ, n) 196 Au reaction. The amount of 99 Mo from the detected γ-lines is in agreement with the estimated value for 238 U(γ, f) and 100 Mo(γ, n) reactions. The production of 99 Mo activity from 238 U(γ, f) and 100 Mo(γ, n) reactions is a relevant and novel approach, which provides alternative routes to 235,238 U(n, f) and 98 Mo(n, γ) reactions, circumventing the need for a reactor. The viability and practicality of the 99 Mo production from the 238 U(γ, f) and 100 Mo(γ, n) reactions alternative to 235,238 U(n, f) and 98 Mo(n, γ) reactions has been emphasize. An estimate has been also arrived based on the experimental data of present work to fulfill the requirement of DOE. (author)

Irradiation behavior of the interaction product of U-Mo fuel particle dispersion in an Al matrix

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Hofman, G.L. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2012-06-15

Highlights: Black-Right-Pointing-Pointer We in-pile tested U-Mo dispersion in Al matrix. Black-Right-Pointing-Pointer We observed interaction layer growth between U-Mo and Al and pore formation there. Black-Right-Pointing-Pointer Pores degrades thermal conductivity and structural integrity of the fueled zone. Black-Right-Pointing-Pointer The amorphous behavior of interaction layers is thought to be the main reason for unstable large pore growth. Black-Right-Pointing-Pointer A mechanism for pore formation and possible remedy to prevent it are proposed. - Abstract: Irradiation performance of U-Mo fuel particles dispersed in Al matrix is stable in terms of fuel swelling and is suitable for the conversion of research and test reactors from highly enriched uranium (HEU) to low enriched uranium (LEU). However, tests of the fuel at high temperatures and high burnups revealed obstacles caused by the interaction layers forming between the fuel particle and matrix. In some cases, fission gas filled pores grow and interconnect in the interdiffusion layer resulting in fuel plate failure. Postirradiation observations are made to examine the behavior of the interdiffusion layers. The interdiffusion layers show a fluid-like behavior characteristic of amorphous materials. In the amorphous interdiffusion layers, fission gas diffusivity is high and the material viscosity is low so that the fission gas pores readily form and grow. Based on the observations, a pore formation mechanism is proposed and potential remedies to suppress the pore growth are also introduced.

U-8 wt %Mo and 7 wt %Mo alloys powder obtained by an hydride-de hydride process; Obtencion de polvo de aleaciones U-8% Mo y U-7% Mo (en peso) mediante hidruracion

Energy Technology Data Exchange (ETDEWEB)

Balart, Silvia N; Bruzzoni, Pablo; Granovsky, Marta S; Gribaudo, Luis M.J.; Hermida, Jorge D; Ovejero, Jose; Rubiolo, Gerardo H; Vicente, Eduardo E [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Materiales

2000-07-01

Uranium-molybdenum alloys are been tested as a component in high-density LEU dispersion fuels with very good performances. These alloys need to be transformed to powder due to the manufacturing requirements of the fuels. One method to convert ductile alloys into powder is the hydride-de hydride process, which takes advantage of the ability of the U-{alpha} phase to transform to UH{sub 3}: a brittle and relatively low-density compound. U-Mo alloys around 7 and 8 wt % Mo were melted and heat treated at different temperature ranges in order to partially convert {gamma} -phase to {alpha} -phase. Subsequent hydriding transforms this {alpha} -phase to UH{sub 3}. The volume change associated to the hydride formation embrittled the material which ends up in a powdered alloy. Results of the optical metallography, scanning electron microscopy, X-ray diffraction during different steps of the process are shown. (author)

Interdiffusion between U-Zr-Mo and stainless steel cladding

International Nuclear Information System (INIS)

Hwang, J. Y.; Lee, B. S.; Lee, J. T.; Kang, Y. H.

1998-01-01

Interdiffusion investigations were carried out at 700 deg C for 200 hours for the diffusion couples assembled with the U-Zr-Mo ternary fuel versus austenitic stainless steel D9 and the U-Zr-Mo ternary fuel versus martensitic stainless steel HT9 respectively to investigate the fuel-cladding compatibility. SEM-EDS analysis was utilized to determine the composition and the penetration depths of the reaction layers. In the case of Fuel/D9 couple, (Fe, Cr, Ni) of the cladding elements formed the precipitates with the Zr, Mo and diminished the U concentration upto 800μ length from the fuel side. Composition of the precipitates was varied with the penetrated elements. In Fuel/HT9 couple, reaction layer was smaller than that of D9 couples and was less affected by cladding elements. The eutectic reaction appeared partially in the Fuel/HT9 diffusion couple

Vortex distribution in small star-shaped Mo{sub 80}Ge{sub 20} plate

Energy Technology Data Exchange (ETDEWEB)

Vu, The Dang, E-mail: vu-dang@pe.osakafu-u.ac.jp [Department of Physics and Electronics, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Department of Physics and Electronics, University of Sciences, Vietnam National University HCMC (Viet Nam); Matsumoto, Hitoshi; Miyoshi, Hiroki [Department of Physics and Electronics, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Huy, Ho Thanh [Department of Physics and Electronics, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Department of Physics and Electronics, University of Sciences, Vietnam National University HCMC (Viet Nam); Shishido, Hiroaki [Department of Physics and Electronics, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Institute for Nanofabrication Research, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Kato, Masaru [Institute for Nanofabrication Research, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Department of Mathematical Science, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Ishida, Takekazu [Department of Physics and Electronics, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Institute for Nanofabrication Research, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan)

2017-02-15

Highlights: • We found the general feature of vortex configuration in small star-shaped Mo{sub 80}Ge{sub 20} plates such as the appearance of symmetric line, the rule of shell filling and the existence of a magic number in both theoretical predictions and experimental results. • We found that the vortex distribution in a concave decagon tends to adapt to one of the five symmetric axes of the star-shaped plate expected in confining vortices in a restricted sample geometry. • The numerical results of Ginzburg–Landau equation confirmed that the filling rules for a vortex configuration and the existence of a magic number for small star-shaped plates are in good agreement with experiment results. - Abstract: We investigated vortex states in small star-shaped Mo{sub 80}Ge{sub 20} plates both theoretically and experimentally. The numerical calculations of the Ginzburg–Landau equation have been carried out with the aid of the finite element method, which is convenient to treat an arbitrarily shaped superconductor. The experimental results were observed by using a scanning SQUID microscope. Through systematic measurements, we figured out how vortices form symmetric configuration with increasing the magnetic field. The vortex distribution tends to adapt to one of five mirror symmetric lines when vortices were located at the five triangular horns of a star-shaped plate. The crystalline homogeneity of a sample was confirmed by the X-ray diffraction and the superconducting properties so that vortices are easily able to move for accommodating vortices in the geometric symmetry of the star-shaped plate. The experimental vortex configurations obtained for a star-shaped plate are in good agreement with theoretical predictions from the nonlinear Ginzburg–Landau equation.

Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

Science.gov (United States)

Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

2015-12-01

U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

Manufacturing and investigation of U-Mo LEU fuel granules by hydride-dehydride processing

International Nuclear Information System (INIS)

Stetskiy, Y.A.; Trifonov, Y.I.; Mitrofanov, A.V.; Samarin, V.I.

2002-01-01

Investigations of hydride-dehydride processing for comminution of U-Mo alloys with Mo content in the range 1.9/9.2% have been performed. Some regularities of the process as a function of Mo content have been determined as well as some parameters elaborated. Hydride-dehydride processing has been shown to provide necessary phase and chemical compositions of U-Mo fuel granules to be used in disperse fuel elements for research reactors. Pin type disperse mini-fuel elements for irradiation tests in the loop of 'MIR' reactor (Dmitrovgrad) have been fabricated using U-Mo LEU fuel granules obtained by hydride-dehydride processing. Irradiation tests of these mini-fuel elements loaded to 4 g U tot /cm 3 are planned to start by the end of this year. (author)

Neutronic feasibility studies using U-Mo dispersion fuel (9 Wt % Mo, 5.0 gU/cm3) for LEU conversion of the MARIA (Poland), IR-8 (Russia), and WWR-SM (Uzbekistan) research reactors

International Nuclear Information System (INIS)

Bretscher, M.M.; Deen, J.R.; Hanan, N.A.; Matos, E.

2000-01-01

U-Mo alloys dispersed in an Al matrix offer the potential for high-density uranium fuels needed for the LEU conversion of many research reactors. On-going fuel qualification tests by the US RERTR Program show good irradiation properties of U-Mo alloy dispersion fuel containing 7-10 weight percent molybdenum. For the neutronic studies in this paper the alloy was assumed to contain 9 wt % Mo (U-9Mo) with a uranium density in the fuel meat of 5.00 gU/cm 3 which corresponds to 32.5 volume % U-9Mo. Fuels containing U-9Mo have been used in Russian reactors since the 1950's. For the three research reactors analyzed here, LEU fuel element thicknesses are the same as those for the Russian-fabricated HEU reference fuel elements. Relative to the reference fuels containing 80-90% enriched uranium, LEU U-9Mo Al-dispersion fuel with 5.00 gU/cm 3 doubles the cycle length of the MARIA reactor and increases the IR-8 cycle length by about 11%. For the WWR-SM reactor, the cycle length, and thus the number of fuel assemblies used per year, is nearly unchanged. To match the cycle length of the 36% enriched fuel currently used in the WWR-SM reactor will require a uranium density in the LEU U-9Mo Al-dispersion fuel of about 5.4 gU/cm 3 . The 5.00 gU/cm 3 LEU fuel causes thermal neutron fluxes in water holes near the edge of the core to decrease by (6-8)% for all three reactors. (author)

Development and characterization of monolithic fuel miniplate alloy U-2.5Zr-7.5Nb, coated in zircaloy; Desenvolvimento e caracterizacao do combustivel nuclear tipo placa monolitico da liga U-2,5Zr-7,5Nb revestido em zircaloy

Energy Technology Data Exchange (ETDEWEB)

Machado, Geraldo Correa

2014-06-01

The autocthonal production of nuclear fuel in Brazil for test and research reactors is restricted to MTR (Material Test Reactor) fuel type dispersion plate, using U3Si2 alloy, coated and dispersed in aluminum, developed by IPEN-SP for use in IEA-R1 reactor. Moreover, the UO{sub 2} fuel rod type for power reactors is manufactured by Rezende (RJ) with a German technology by INB under license. Currently, Brazil is performing two programs of developing reactors. Currently, Brazil is developing two reactors. One of them is the development, by CNEN, the Brazilian Multipurpose Reactor (RMB), for testing, research and radioisotope production. The other one is the development a power reactor for naval propulsion, conducted by the Brazilian Navy. This dissertation presents the development and characterization of monolithic fuel miniplate alloy U-2.5Zr-7.5Nb, coated in zircaloy (ZRY), on a laboratory scale. Due to its innovative features and properties, this fuel can be used as fuel in both test reactors, research and producing radioisotopes for power reactors as small and medium sizes. Thus, this high potential fuel can be used in domestic reactors currently under development. The development of monolithic fuel plate type is made using the technique called 'picture-frame' where a sandwich composed of a monolith alloy U-2.5Zr- 7.5Nb coupled to a frame and coated sheets of Zry is obtained. The alloy U-2.5Zr-7.5Nb was obtained by melting in an induction furnace and then was cast into rectangular ingots of graphite, thus achieving an ingot with approximate dimensions of 170 x 50 x 60 mm. The obtained ingot was hot rolled at 850 ºC, with a 50 % reduction in thickness, in order to refine the raw structure of fusion. Samples cut from the alloy U-2.5Zr-7.5Nb, with dimensions 20 x 20 x 6 mm were placed in frames and plates Zry and joined by TIG (Tungsten Inert Gas) under an atmosphere of argon, obtaining a set of 10 mm thick, 45 mm wide and 100 mm long. The sandwiches were

Structural instability and ground state of the U{sub 2}Mo compound

Energy Technology Data Exchange (ETDEWEB)

Losada, E.L., E-mail: losada@cab.cnea.gov.ar [SIM" 3, Centro Atómico Bariloche, Comisión Nacional de Energía Atómica (Argentina); Garcés, J.E. [Gerencia de Investigación y Aplicaciones Nucleares, Comisión Nacional de Energía Atómica (Argentina)

2015-11-15

This work reports on the structural instability at T = 0 °K of the U{sub 2}Mo compound in the C11{sub b} structure under the distortion related to the C{sub 66} elastic constant. The electronic properties of U{sub 2}Mo such as density of states (DOS), bands and Fermi surface (FS) are studied to understand the source of the instability. The C11{sub b} structure can be interpreted as formed by parallel linear chains along the z-directions each one composed of successive U–Mo–U blocks. Hybridization due to electronic interactions inside the U–Mo–U blocks is slightly modified under the D{sub 6} distortion. The change in distance between chains modifies the U–U interaction and produces a split of f-states. The distorted structure is stabilized by a decrease in energy of the hybridized states, mainly between d-Mo and f-U states, together with the f-band split. Consequently, an induced Peierls distortion is produced in U{sub 2}Mo due to the D{sub 6} distortion. It is important to note that the results of this work indicate that the structure of the ground state of the U{sub 2}Mo compound is not the assumed C11{sub b} structure. It is suggested for the ground state a structure with hexagonal symmetry (P6 #168), ∼0.1 mRy below the energy of the recently proposed Pmmn structure. - Highlights: • Structural instability of the C11b compound due to the D6 deformation. • Induced Peierls distortion due to the D6 deformation. • Distorted structure is stabilized by hybridization and split of f-Uranium state. • P6 (#168) suggested ground state for the U{sub 2}Mo compound.

XRD and neutron diffraction analyses of heat treated U-Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Nam, Ji Min; Kim, Woo Jeong; Ryu, Ho Jin; Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

High density U Mo alloys are regarded as promising candidates for advanced research reactor fuel because they have shown stable irradiation performance when compared to other uranium alloys and compounds. However, interaction layer formation between the U Mo alloys and Al matrix degrades the irradiation performance of U Mo dispersion fuel. Therefore, addition of Ti in U Mo alloys, addition of Si in Al matrix and silicide or nitride coating on the surface of U Mo particles have been proposed in order to inhibit the interaction layer growth. In order to analyze the mechanisms of interaction layer growth inhibition by adding Ti in U Mo alloys or Si in Al matrix, accurate phase characterization of the interaction layers is required. While previous studies using X ray diffraction have been reported, laboratory X ray diffraction method has limitations such as low resolution and small measurement volume. Neutron diffraction method can be a more accurate analysis when compared with X ray diffraction method due to the large penetration depth of neutron. In this study, X ray diffraction and neutron diffraction experiments have been performed by using the laboratory X ray diffractometer and high resolution powder diffractometer (HRPD) of the HANARO research reactor in KAERI.

Study on microstructure change of Uranium nitride coated U-7wt%Mo powder by heat treatment

Energy Technology Data Exchange (ETDEWEB)

Cho, Woo Hyoung; Park, Jae Soon; Lee, Hae In; Kim, Woo Jeong; Yang, Jae Ho; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-10-15

Uranium-molybdenum alloy particle dispersion fuel in an aluminum matrix with a high uranium density has been developed for a high performance research reactor in the RERTR program. In order to retard the fuel-matrix interaction in U-Mo/Al dispersion fuel in which the U-Mo fuel particles were dispersed in Al matrix, nitride layer coated U-Mo fuel particle has been designed and techniques to fabricate nitride-layer coated U-7wt%Mo particles have been developed in our lab. In this study, uranium nitride coated U-Mo particle has heat treatment for several times and degree. And we suggested for interaction layer remedy in U-Mo dispersion fuel. We investigate effect of heat treatment interaction layer evolution on uranium nitride coated U-Mo powder. The EDS and XRD analysis to investigate the phase evolution in uranium nitride coated layer is also a part of the present work

Theoretical Model for Volume Fraction of UC, 235U Enrichment, and Effective Density of Final U 10Mo Alloy

Energy Technology Data Exchange (ETDEWEB)

Devaraj, Arun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); Prabhakaran, Ramprashad [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); Hu, Shenyang Y. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); McGarrah, Eric J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL)

2016-04-12

The purpose of this document is to provide a theoretical framework for (1) estimating uranium carbide (UC) volume fraction in a final alloy of uranium with 10 weight percent molybdenum (U-10Mo) as a function of final alloy carbon concentration, and (2) estimating effective ²³⁵U enrichment in the U-10Mo matrix after accounting for loss of ²³⁵U in forming UC. This report will also serve as a theoretical baseline for effective density of as-cast low-enriched U-10Mo alloy. Therefore, this report will serve as the baseline for quality control of final alloy carbon content

US Progress on Property Characterization to Support LEU U-10 Mo Monolithic Fuel Development

Energy Technology Data Exchange (ETDEWEB)

Cole, James Irvin [Idaho National Laboratory; Rabin, Barry H [Idaho National Laboratory; Smith, James Arthur [Idaho National Laboratory; Scott, Clark Landon [Idaho National Laboratory; Benefiel, Bradley Curtis [Idaho National Laboratory; Larsen, Eric David [Idaho National Laboratory; Lind, Robert Paul [Idaho National Laboratory; Sell, David Alan [Idaho National Laboratory

2016-03-01

The US High Performance Research Reactor program is pursuing development and qualification of a new high density monolithic LEU fuel to facilitate conversion of five higher power research reactors located in the US (ATR, HFIR, NBSR, MIT and MURR). In order to support fabrication development and fuel performance evaluations, new testing capabilities are being developed to evaluate the properties of fuel specimens. Residual stress and fuel-cladding bond strength are two characteristics related to fuel performance that are being investigated. In this overview, new measurement capabilities being developed to assess these characteristics in both fresh and irradiated fuel are described. Progress on fresh fuel testing is summarized and on-going hot-cell implementation efforts to support future PIE campaigns are detailed. It is anticipated that benchmarking of as-fabricated fuel characteristics will be critical to establishing technical bases for specifications that optimize fuel fabrication and ensure acceptable in-reactor fuel performance.

A novel ionic liquid monolithic column and its separation properties in capillary electrochromatography

International Nuclear Information System (INIS)

Wang Yu; Deng Qiliang; Fang Guozhen; Pan Mingfei; Yu Yang; Wang Shuo

2012-01-01

Highlights: ► ILs as functional monomer for capillary monolithic column. ► Separation of alkylbenzenes, thiourea analogues, and amino acids. ► The column generate a stable reversed EOF from pH 2.0 to 12.0. ► The column efficiency of 147,000 plates m −1 was obtained for thiourea. - Abstract: A novel ionic liquid (IL) monolithic capillary column was successfully prepared by thermal free radical copolymerization of IL (1-vinyl-3-octylimidazolium chloride, ViOcIm + Cl − ) together with lauryl methacrylate (LMA) as the binary functional monomers and ethylene dimethacrylate (EDMA) as the cross-linker in binary porogen. The proportion of monomers, porogens and cross-linker in the polymerization mixture was optimized in detail. The resulting IL-monolithic column could not only generate a stable reversed electroosmotic flow (EOF) in a wide pH range (2.0–12.0), but also effectively eliminate the wall adsorption of the basic analytes. The obtained IL-monolithic columns were examined by scanning electron microscopy (SEM) and Fourier transform infrared (FT-IR). These results indicated that the IL-monolithic capillary column possessed good pore properties, mechanical stability and permeability. The column performance was also evaluated by separating different kinds of compounds, such as alkylbenzenes, thiourea and its analogues, and amino acids. The lowest plate height of ∼6.8 μm was obtained, which corresponded to column efficiency (theoretical plates, N) of ∼147,000 plates m −1 for thiourea. ILs, as a new type of functional monomer, present a promising option in the fabrication of the organic polymer-based monolithic columns in CEC.

Analyses of Interaction Phases of U Mo Dispersion Fuel by Synchrotron X ray Diffraction

Energy Technology Data Exchange (ETDEWEB)

Kim, Woo Jeong; Nam, Ji Min; Ryu, Ho Jin; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Herve, Palancher; Charollais, Francois [Saint Paul Lez Durance Cedex, Rhone (France); Bonnin, Anne; Honkimaeki, Veijo [Grenoble Cedex, Grenoble (France); Patrick Lemoined [Gif sur Yvette, Paris (France)

2012-10-15

Gamma phase U Mo alloys are one of the promising candidates to be used as advanced high uranium density fuel for high power research reactors due to their excellent irradiation performance. However, formation of interaction layers between the U Mo particles and Al matrix degrades the irradiation performance of U Mo dispersion fuel. One of the remedies to the interaction problem is a Si addition to the Al matrix. Recent irradiation tests have shown that the use of Al (2{approx}5wt%)Si matrices retarded the growth of interaction layers effectively during irradiation. Recently, KAERI has proposed silicide or nitride coated U Mo fuel for the minimization of the interaction layer growth. The silicide or nitride coatings are expected to act as interdiffusion barriers and their out of pile tests showed the improved diffusion barrier performances of the silicide and nitride layers. In order to characterize constituent phases in the coated layers on U Mo particles and the interaction layers of coated U Mo particle dispersed fuel, synchrotron X ray diffraction experiments have been performed at the ESRF (European Synchrotron Radiation Facility), France as a KAERI CEA cooperation program.

Comparison of U-Pu-Mo, U-Pu-Nb, U-Pu-Ti and U-Pu-Zr alloys

International Nuclear Information System (INIS)

Boucher, R.; Barthelemy, P.

1964-01-01

The data concerning the U-Pu, U-Pu-Mo and U-Pu-Nb are recalled. The results obtained with U-Pu-Ti and U-Pu-Zr alloys containing 15-20 per cent Pu and 10 wt. per cent ternary element are reported. The transformation temperatures, the expansion coefficients, the nature of phases, the thermal cycling behaviour have been determined. A list of the principal properties of these different alloys is presented and the possibilities of their use as fast reactor's fuel element are considered. The U-Pu-Ti alloys seem to be quite promising: easiness of fabrication, large thermal stability, excellent behaviour in air, small quantity of zeta phase, temperature of solidus superior to 1100 deg. C. (authors) [fr

Code structure for U-Mo fuel performance analysis in high performance research reactor

Energy Technology Data Exchange (ETDEWEB)

Jeong, Gwan Yoon; Cho, Tae Won; Lee, Chul Min; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

A performance analysis modeling applicable to research reactor fuel is being developed with available models describing fuel performance phenomena observed from in-pile tests. We established the calculation algorithm and scheme to best predict fuel performance using radio-thermo-mechanically coupled system to consider fuel swelling, interaction layer growth, pore formation in the fuel meat, and creep fuel deformation and mass relocation, etc. In this paper, we present a general structure of the performance analysis code for typical research reactor fuel and advanced features such as a model to predict fuel failure induced by combination of breakaway swelling and pore growth in the fuel meat. Thermo-mechanical code dedicated to the modeling of U-Mo dispersion fuel plates is being under development in Korea to satisfy a demand for advanced performance analysis and safe assessment of the plates. The major physical phenomena during irradiation are considered in the code such that interaction layer formation by fuel-matrix interdiffusion, fission induced swelling of fuel particle, mass relocation by fission induced stress, and pore formation at the interface between the reaction product and Al matrix.

Report on the Synchrotron Characterization of U-Mo and U-Zr Alloys and the Modeling Results

Energy Technology Data Exchange (ETDEWEB)

Okuniewski, Maria A. [Purdue Univ., West Lafayette, IN (United States); Ganapathy, Varsha [Purdue Univ., West Lafayette, IN (United States); Hamilton, Brenden [Purdue Univ., West Lafayette, IN (United States); Cassutt, Paul [Purdue Univ., West Lafayette, IN (United States); Zhang, Fan [Purdue Univ., West Lafayette, IN (United States); Velaquez, Daniel [Illinois Inst. of Technology, Chicago, IL (United States); Seibert, Rachel [Illinois Inst. of Technology, Chicago, IL (United States); Terry, Jeff [Illinois Inst. of Technology, Chicago, IL (United States); Sprouster, David [Brookhaven National Lab. (BNL), Upton, NY (United States); Ecker, Lynne [Brookhaven National Lab. (BNL), Upton, NY (United States); Elbakhshwan, Mohamed [Brookhaven National Lab. (BNL), Upton, NY (United States)

2016-09-01

ABSTRACT Uranium-molybdenum (U-Mo) and uranium-zirconium (U-Zr) are two promising fuel candidates for nuclear transmutation reactors which burn long-lived minor actinides and fission products within fast spectrum reactors. The objectives of this research are centered on understanding the early stages of fuel performance through the examination of the irradiation induced microstructural changes in U-Zr and U-Mo alloys subjected to low neutron fluences. Specimens that were analyzed include those that were previously irradiated in the Advanced Test Reactor at INL. This most recent work has focused on a sub-set of the irradiated specimens, specifically U-Zr and U-Mo alloys that were irradiated to 0.01 dpa at temperatures ranging from (150-800oC). These specimens were analyzed with two types of synchrotron techniques, including X-ray absorption fine structure and X-ray diffraction. These techniques provide non-destructive microstructural analysis, including phase identification and quantitation, lattice parameters, crystallite sizes, as well as bonding, structure, and chemistry. Preliminary research has shown changes in the phase fractions, crystallite sizes, and lattice parameters as a function of irradiation and temperature. Future data analyses will continue to explore these microstructural changes.

Vortex distribution in amorphous Mo{sub 80}Ge{sub 20} plates with artificial pinning center

Energy Technology Data Exchange (ETDEWEB)

Matsumoto, Hitoshi [Department of Physics and Electronics, Osaka Prefecture University, 1-1 Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan); Huy, Ho Thanh [Department of Physics and Electronics, Osaka Prefecture University, 1-1 Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan); Department of Physics and Electronics, University of Sciences, Vietnam National University HCMC, 227 Nguyen Van Cu, District 5, HoChiMinh City (Viet Nam); Miyoshi, Hiroki; Okamoto, Takuto; Dang, Vu The [Department of Physics and Electronics, Osaka Prefecture University, 1-1 Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan); Kato, Masaru [Institute for Nanofabrication Research, Osaka Prefecture University, 1-1 Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan); Department of Mathematical Science, Osaka Prefecture University1-1 Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan); Ishida, Takekazu, E-mail: ishida@center.osakafu-u.ac.jp [Department of Physics and Electronics, Osaka Prefecture University, 1-1 Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan); Institute for Nanofabrication Research, Osaka Prefecture University, 1-1 Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan)

2016-11-15

Highlights: • We reveal that the vortex distribution in small amorphous Mo{sub 80}Ge{sub 20} superconducting starshaped plate by using a scanning SQUID microscope. • We find that vortex configuration evolves systematically when the applied magnetic field is changed at the several different fields. • We fabricate an artificial dip by Ar ion milling in a mesoscopic plate, and find this works as a pinning center by comparing the vortex behavior in a sample without pins. - Abstract: Vortices in superconductor give rise to a rich variety of phenomena because they interact with shielding currents, temperature gradients, sample defects, boundaries, and other neighboring vortices. It would be very important to understand particular features of vortex states in a downsized system. Our study focuses on vortex states in small star-shaped Mo{sub 80}Ge{sub 20} plates with and without an artificial pin at the plate center. Vortex states are greatly influenced by the sample geometry, the temperature and the magnetic field, and they can be occasionally exotic compared to the bulk case. We use the amorphous Mo{sub 80}Ge{sub 20} films due to the nature of weak pinning in studying vortex configurations. We applied scanning superconducting quantum interference device (SQUID) microscopy because it enables us to see vortex states directly and it is the most sensitive instrument for mapping tiny local current flows or magnetic moments without damaging the sample. We interpreted that vortex configurations had essentially the nature of mirror reflection symmetry in both cases with an artificial pin and without an artificial pin and pinned cases while the influence of disorder was seen in our observation on the specimen without an artificial pin.

Study on characterization of interaction layer between U-10wt%Mo alloy and LT24Al

International Nuclear Information System (INIS)

Chen Jiangang; Yin Changgeng; Sun Changlong; Pang Xiaoxuan; Liu Yunming

2009-01-01

The characterization of interaction layer(IL) between U-10wt%Mo alloy and LT24 Al was studied in detail in this paper. Sandwich structured U-Mo/LT24 Al diffusion couples were hot pressed at different temperature and pressure for different time. Then they were analyzed by Optical Microscope (OM) and Scanning Electron Microscope (SEM) to observe the width of the IL. The distribution of the diffusion elements and the phases in the IL were determined by Energy Dispersive Spectroscopy (EDS) and X Ray Diffraction (XRD). Analysis results are as follows: the diffusion manner was reaction diffusion, and diffusion direction mainly was that Al atoms diffused to U-Mo alloy; diffusion mechanism was vacancy diffusion and growth kinetics showed reaction was controlled by the diffusion speed; the IL containing single phase was constituted mainly by (U, Mo) Al 3 ; the IL containing two phases or more was constituted mainly by (U, Mo) Al 3 and (U, Mo) Al 4 and Al 20 Mo 2 U; and Si impurity in the LT24 Al was easy to enrich in the IL which showed Si added to Al could play positive role on improve compatibility between U-Mo and Al. (authors)

Recent observations at the post-irradiation examination of low-enriched U-Mo miniplates irradiated to high burn-up

International Nuclear Information System (INIS)

Hofman, G.L.; Kim, Y.S.; Finlay, M.R.; Snelgrove, J.L.; Hayes, S.L.; Meyer, M.K.; Clark, C.R.

2003-01-01

High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL Alpha-Gamma Hot Cell Facility. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm -3 atomized and machined uranium alloy powders containing 6.5 wt% to 10 wt% molybdenum. In addition, two miniplates contained solid U-10wt%Mo foils. Recent results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from a companion test performed to 50% burnup, RERTR-5, are presented. (author)

Modeling solute segregation during the solidification of γ-phase U-Mo alloys

Energy Technology Data Exchange (ETDEWEB)

Steiner, M.A., E-mail: mas4cw@virginia.edu [University of Virginia, Material Science and Engineering, 395 McCormick Rd, Charlottesville, VA 22904 (United States); Garlea, E. [Y-12 National Security Complex, Oak Ridge, TN 37831 (United States); Agnew, S.R. [University of Virginia, Material Science and Engineering, 395 McCormick Rd, Charlottesville, VA 22904 (United States)

2016-06-15

Using first principles calculations, it is demonstrated that solute segregation during U-Mo solidification can be modeled using the classic Brody-Fleming limited diffusion framework. The necessary supporting equations specific to the U-Mo alloy, along with careful verification of the assumptions underpinning the Brody-Fleming model are developed, allowing for concentration profile predictions as a function of alloy composition and cooling rate. The resulting model is compared to experimental solute concentration profiles, showing excellent agreement. Combined with complementary modeling of dendritic feature sizes, the solute segregation model can be used to predict the complete microstructural state of individual U-Mo volume elements based upon cooling rates, informing ideal processing routes.

Titanium plate supported MoS{sub 2} nanosheet arrays for supercapacitor application

Energy Technology Data Exchange (ETDEWEB)

Wang, Lina; Ma, Ying [State Key Laboratory for Oxo Synthesis & Selective Oxidation, and National Engineering Research Center for Fine Petrochemical Intermediates, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Yang, Min [State Key Laboratory for Oxo Synthesis & Selective Oxidation, and National Engineering Research Center for Fine Petrochemical Intermediates, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Qi, Yanxing, E-mail: qiyx@lzb.ac.cn [State Key Laboratory for Oxo Synthesis & Selective Oxidation, and National Engineering Research Center for Fine Petrochemical Intermediates, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China)

2017-02-28

A promising new concept is to apply binder-free supercapacitor electrode by directly growing active materials on current collectors. However, there are many challenges to be solved, such as fabrication of well quality electronic contact and good mechanical stability films through a simple and feasible method. In this study, MoS{sub 2} nanosheet arrays supported on titanium plate has been synthesized by a hydrothermal method without other additives, surface active agents and toxic reagents. As the supercapacitor electrode, a good capacitance of 133 F g{sup −1} is attained at a discharge current density of 1 A g{sup −1}. The specific energy density is 11.11 Wh kg{sup −1} at a power density of 0.53 kW kg{sup −1}. Moreover, the electrode shows an excellent cyclic stability. The loss of capacity is only 7% even after 1000 cycles. In addition, the formation mechanism is proposed. The facile method of fabricating MoS{sub 2} nanosheet arrays on titanium plate affords an green and effective way to prepare other metal sulfides for the application in electrochemical capacitors.

Alternative Crucibles for U-Mo Microwave Melting

Energy Technology Data Exchange (ETDEWEB)

Kirby, Brent W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-03-31

The crucibles used currently for microwave melting of U-Mo alloy at the Y-12 Complex contain silicon carbide (SiC) in a mullite (3Al₂O₃-2SiO₂) matrix with an erbia coating in contact with the melt. Due to observed silicon contamination, Pacific Northwest National Laboratory has investigated alternative crucible materials that are susceptible to microwave radiation and are chemically compatible with molten U-Mo at 1400 1500C. Recommended crucibles for further testing are: 1) high-purity alumina (Al₂O₃); 2) yttria-stabilized zirconia (ZrO₂); 3) a composite of alumina and yttria-stabilized zirconia; 4) aluminum nitride (AlN). Only AlN does not require an erbia coating. The recommended secondary susceptor, for heating at low temperature, is SiC in a “picket fence” arrangement.

Cross section TEM characterization of high-energy-Xe-irradiated U-Mo

Energy Technology Data Exchange (ETDEWEB)

Ye, B., E-mail: bye@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Ave. Lemont, IL 60439 (United States); Jamison, L.; Miao, Y. [Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Ave. Lemont, IL 60439 (United States); Bhattacharya, S. [Department of Materials Science and Engineering, Northwestern University, 2220 Campus Dr. Evanston, IL 60208 (United States); Hofman, G.L.; Yacout, A.M. [Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Ave. Lemont, IL 60439 (United States)

2017-05-15

U-Mo alloys irradiated with 84 MeV Xe ions to various doses were characterized with transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) techniques. The TEM thin foils were prepared perpendicular to the irradiated surface to allow a direct observation of the entire region modified by ions. Therefore, depth-selective microstructural information was revealed. Varied irradiation-induced phenomena such as gas bubble formation, phase reversal, and recrystallization were observed at different ion penetration depths in U-Mo. - Highlights: •Three distinct zones were observed along the ion traveling direction in U-7Mo irradiated with 84 MeV Xe ions at 350 °C. •The α-U particles within the Xe-implanted region were reverted to γ-U phase by irradiation. •High-density random intra-granular bubbles in a size of 4–5 nm were found in the irradiated region, coexisting with large inter-granular bubbles. •The high lattice stresses built up during the irradiation-induced phase reversal is probably the driving force for the small grain formation at cell boundaries.

Reduced interaction layer growth of U-Mo dispersion in Al-Si

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Park, Jong Man; Ryu, Ho Jin; Jung, Yang Hong [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong, Daejeon 305-353 (Korea, Republic of); Hofman, G.L. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2012-11-15

Development of high U-density U-Mo fuel particle dispersion in Al is needed to convert high power research and test reactors from HEU to LEU. Interaction layer growth between U-Mo and Al poses a challenge to this goal. The KOMO-4 test was designed at KAERI and irradiated in the HANARO reactor to {approx}50% burnup of initial 19.75% U-235 enrichment at {approx}200 Degree-Sign C. The main objective of the test was to examine the effect of the Si content in the matrix up to 8 wt.%. U-Mo/Al-Si dispersion samples with a Si addition in the range 0-8 wt.% in the matrix were tested. A sample with pre-irradiation Si-containing interaction layers (ILs) was also tested. As the Si content in the matrix increases, the IL growth was progressively reduced. Contrary to the thermodynamics prediction and out-of-pile observations, however, Si accumulation in the ILs occurred near the IL-matrix interface with only a slight increase in concentration. The effect of the pre-formed ILs was insignificant in reducing IL growth.

Neutronic and thermal-hydraulic analysis of devices for irradiation of LEU targets type of UALx-Al and U-Ni to production of 99Mo in reactor IEA-R1 and RMB

International Nuclear Information System (INIS)

Domingos, Douglas Borges

2014-01-01

In this work neutronic and thermal-hydraulic analyses were made to compare three types of targets (UAl 2 -Al, U-Ni cylindrical and U-Ni plate) used for the production of 99 Mo by fission of 235 U. Some experiments were conducted to validate the neutronic and thermal-hydraulics methodologies used in this work. For the neutronic calculations the computational programs NJOY99.0, AMPX-II and HAMMERTECHNION were used to generate the cross sections. SCALE 6.0 and CITATION computational programs were used for three-dimensional calculations of the reactor cores, fuel burning and the production of 99 Mo. The computational programs MTRCR-IEAR1 and ANSYS CFX were used to calculate the thermal and hydraulic parameters of the irradiation devices and for comparing them to limits and design criteria. First were performed neutronic and thermal-hydraulic analyzes for the reactor IEA-R1 with the targets of UAl 2 -Al (10 mini plates). Analyses have shown that the total activity obtained for 99 Mo on the mini plates does not meet the demand of Brazilian hospitals (450 Ci/week) and that no limit of thermo-hydraulic design is overtaken. Next, the same calculations were performed for the three target types in Multipurpose Brazilian Reactor (MBR). The neutronic analyzes demonstrated that the three targets meet the demand of Brazilian hospitals. The thermal hydraulic analysis shows that a minimum speed of 7 m/s for the target UAl 2 -Al, 8 m/s for the cylindrical target U-Ni and 9 m/s for the target U-Ni plate will be necessary in the irradiation device to not exceed the design limits. Were performed experiments using a test bench for validate the methodologies for the thermal-hydraulic calculation. The experiments performed to validate the neutronic calculations were made in the reactor IPEN/MB-01. All experiments were simulated with the methodologies described above and the results compared. The simulations results showed good agreement with experimental results. (author)

A novel ionic liquid monolithic column and its separation properties in capillary electrochromatography.

Science.gov (United States)

Wang, Yu; Deng, Qi-Liang; Fang, Guo-Zhen; Pan, Ming-Fei; Yu, Yang; Wang, Shuo

2012-01-27

A novel ionic liquid (IL) monolithic capillary column was successfully prepared by thermal free radical copolymerization of IL (1-vinyl-3-octylimidazolium chloride, ViOcIm(+)Cl(-)) together with lauryl methacrylate (LMA) as the binary functional monomers and ethylene dimethacrylate (EDMA) as the cross-linker in binary porogen. The proportion of monomers, porogens and cross-linker in the polymerization mixture was optimized in detail. The resulting IL-monolithic column could not only generate a stable reversed electroosmotic flow (EOF) in a wide pH range (2.0-12.0), but also effectively eliminate the wall adsorption of the basic analytes. The obtained IL-monolithic columns were examined by scanning electron microscopy (SEM) and Fourier transform infrared (FT-IR). These results indicated that the IL-monolithic capillary column possessed good pore properties, mechanical stability and permeability. The column performance was also evaluated by separating different kinds of compounds, such as alkylbenzenes, thiourea and its analogues, and amino acids. The lowest plate height of ~6.8 μm was obtained, which corresponded to column efficiency (theoretical plates, N) of ~147,000 plates m(-1) for thiourea. ILs, as a new type of functional monomer, present a promising option in the fabrication of the organic polymer-based monolithic columns in CEC. Copyright © 2011 Elsevier B.V. All rights reserved.

Neutronic and thermal-hydraulic analysis of devices for irradiation of LEU targets type of UAL{sub x}-Al and U-Ni to production of {sup 99}Mo in reactor IEA-R1 and RMB; Analises neutronicas e termo-hidraulica de dispositivos para irradiacao de alvos tipo LEU de UAL{sub x}-Al e U-Ni para producao de {sup 99}Mo nos reatores IEA-R1 e RMB

Energy Technology Data Exchange (ETDEWEB)

Domingos, Douglas Borges

2014-07-01

In this work neutronic and thermal-hydraulic analyses were made to compare three types of targets (UAl{sub 2}-Al, U-Ni cylindrical and U-Ni plate) used for the production of {sup 99}Mo by fission of {sup 235}U. Some experiments were conducted to validate the neutronic and thermal-hydraulics methodologies used in this work. For the neutronic calculations the computational programs NJOY99.0, AMPX-II and HAMMERTECHNION were used to generate the cross sections. SCALE 6.0 and CITATION computational programs were used for three-dimensional calculations of the reactor cores, fuel burning and the production of {sup 99}Mo. The computational programs MTRCR-IEAR1 and ANSYS CFX were used to calculate the thermal and hydraulic parameters of the irradiation devices and for comparing them to limits and design criteria. First were performed neutronic and thermal-hydraulic analyzes for the reactor IEA-R1 with the targets of UAl{sub 2}-Al (10 mini plates). Analyses have shown that the total activity obtained for {sup 99}Mo on the mini plates does not meet the demand of Brazilian hospitals (450 Ci/week) and that no limit of thermo-hydraulic design is overtaken. Next, the same calculations were performed for the three target types in Multipurpose Brazilian Reactor (MBR). The neutronic analyzes demonstrated that the three targets meet the demand of Brazilian hospitals. The thermal hydraulic analysis shows that a minimum speed of 7 m/s for the target UAl{sub 2}-Al, 8 m/s for the cylindrical target U-Ni and 9 m/s for the target U-Ni plate will be necessary in the irradiation device to not exceed the design limits. Were performed experiments using a test bench for validate the methodologies for the thermal-hydraulic calculation. The experiments performed to validate the neutronic calculations were made in the reactor IPEN/MB-01. All experiments were simulated with the methodologies described above and the results compared. The simulations results showed good agreement with experimental

Miniplates irradiation in the ATR (Idaho, USA)

International Nuclear Information System (INIS)

Pasqualini, Enrique E.

2007-01-01

High density U Mo alloys are promising for its utilization in the reconversion of HEU fuels to LEU for research nuclear reactors. Ought to the thermomechanical properties of the alloy U Mo and its interaction with aluminium it is necessary to develop new technologies and fabrication procedures to qualify this material as a nuclear fuel. In this work a review is made about the evolution of the idea and PIE experiments of monolithic LEU U 7 Mo fuel with Zr-4 cladding. The irradiation took place in the frame of international qualification efforts of dispersed and monolithic U Mo fuels. Dispersed and monolithic fuels, elaborated and in intermediate steps of development, are discussed. (author) [es

Analysis of Neutron Flux Distribution in Rsg-Gas Reactor With U-Mo Fuels

Directory of Open Access Journals (Sweden)

Taswanda Taryo

2004-01-01

Full Text Available The use of U-Mo fuels in research reactors seems to be promising and, recently, world researchers have carried out these such activities actively. The National Nuclear Energy Agency (BATAN which owns RSG-GAS reactor available in Serpong Research Center for Atomic Energy should anticipate this trend. It is, therefore, this research work on the use of U-Mo fuels in RSG-GAS reactor should be carried out. The work was focused on the analysis of neutron flux distribution in the RSG-GAS reactor using different content of molybdenum in U-Mo fuels. To begin with, RSG-GAS reactor core model was developed and simulated into X, Y and Z dimensions. Cross section of materials based on the developed cells of standard and control fuels was then generated using WIMS-D5-B. The criticality calculations were finally carried out applying BATAN-2DIFF code. The results showed that the neutron flux distribution obtained in U-Mo-fuel-based RSG-GAS core is very similar to those achieved in the 300-gram sillicide-fuel-based RSG-GAS reactor core. Indeed, the utilization of the U-Mo RSG-GAS core can be very similar to that of the high-density sillicide reactor core and even could be better in the future.

Analysis of Mo99 production irradiating 20% U targets

International Nuclear Information System (INIS)

Calabrese, C. Ruben; Grant, Carlos R.; Marajofsky, Andres; Parkansky, David G.

1999-01-01

At present time, the National Atomic Energy Commission is producing about 800 Ci of Mo99 per week irradiating 90% enriched uranium-aluminum alloy plate targets in the RA-3 reactor, a 5 MW. Mtr type one. In order to change to 20% enriched uranium, and to increase the production to about 3000 Ci per week some configurations were studied with rod and plate geometry with uranium (20% enriched) -aluminum targets. The first case was the irradiation of a plate target element in the normal reactor configuration. Results showed a good efficiency, but both reactivity value and power density were too high. An element with rods was also analyzed, but results showed a poor efficiency, too much aluminum involved in the process, although a low reactivity and an acceptable rod power density. Finally, a solution consisting of plate elements with a Zircaloy cladding was adopted, which has shown not only a good efficiency, but it is also acceptable from the viewpoint of safety, heat transference criteria and feasibility

An investigation on the irradiation behavior of atomized U-Mo/Al dispersion rod fuels

International Nuclear Information System (INIS)

Park, J.M.; Ryu, H.J.; Lee, Y.S.; Lee, D.B.; Oh, S.J.; Yoo, B.O.; Jung, Y.H.; Sohn, D.S.; Kim, C.K.

2005-01-01

The second irradiation fuel experiment, KOMO-2, for the qualification test of atomized U-Mo dispersion rod fuels with U-loadings of 4-4.5 gU/cc at KAERI was finished after an irradiation up to 70 at% U 235 peak burn-up and subjected to the IMEF (Irradiation material Examination Facility) for a post-irradiation analysis in order to understand the fuel irradiation performance of the U-Mo dispersion fuel. Current results for PIE of KOMO-2 revealed that the U-Mo/Al dispersion fuel rods exhibited a sound performance without any break-away swelling, but most of the fuel rods irradiated at a high linear power showed an extensive formation of the interaction phase between the U-Mo particle and the Al matrix. In this paper, the analysis of the PIE results, which focused on the diffusion related microstructures obtained from the optical and EPMA (Electron Probe Micro Analysis) observations, will be presented in detail. And a thermal modeling will be carried out to calculate the temperature of the fuel rod during an irradiation. (author)

ON-GOING STATUS OF KJRR FUEL (U-7MO) QUALIFICATION

Energy Technology Data Exchange (ETDEWEB)

Yim, J. S.; Tahk, Y. W.; Oh, J. Y.; Kim, H. J.; Kong, E. H.; Lee, B. H.; Park, J. M.; Jeong, Y. J.; Lee, K. H.; Kim, S. H.; Lee, C. T.; Beasley, A. A.; Choi, Y. J.; Crawford, D. S.; Nielsen, J. W.; Woolstenhulme, N. E.

2017-03-01

In order to cope with global shortage of Mo-99 supplies and with growing demand of neutron transmutation doping, KJRR construction plan has been launched since April 2012 to provide self-sufficiency of domestic RI demand, and to extend Si doping capacity for power device market growth. Through comprehensive surveillance of the fuels in-reactor behavior, KAERI has selected the fuel meat of U-7%Mo dispersion in an aluminum matrix with 5wt%Si for the KJRR fuel. As part of the efforts for fuel licensing and qualification of the KJRR fuel, an LTA irradiation test at the ATR started from November 2015 was successfully completed by reaching at 219 EFPD in the end of February 2017. Together with the results of HAMP-1 already completed irradiation and PIE, the successful irradiation of the LTA also demonstrates the fuel integrity under more rigorous conditions than the KJRR operation conditions. This paper updates the current status of the KJRR U7Mo (8 g-U/cm3) LTA irradiation and PIE plan up to date as of February 2017.

A study of HANARO core conversion using high density U-Mo fuel

International Nuclear Information System (INIS)

Lee, K.H.; Lee, C.S.; Lee, B.C.; Park, S.J.; Kim, H.; Kim, C.K.

2002-01-01

Currently, HANARO is using 3.15gU/cc U3Si/Al as a driver fuel. HANARO has seven vertical irradiation holes in the core region. Three of them including a central trap are located in the inner region of the core and mainly being used for material irradiation tests. Four of them are located in the reflector tank but cooled by primary coolant. They are used for fuel irradiation tests or radioisotope development tests. For minimum core modification using high density U-Mo fuels, no dimension change is assumed in the current fuel rods and the cladding thickness remains the same in this study. The high density U-Mo fuel will have up to about twice the linear uranium loading of a current HANARO driver fuel. Using this high density fuel 8 fuel sites can be replaced with irradiation sites. Three kinds of conceptual cores are considered using 5 gU/cc U-7Mo/Al and 16 gU/cc U-7Mo. The increase of the linear heat generation rate due to the decrease of total fuel length can be overcome by more uniform radial and axial power distribution using different uranium densities and different fuel meat diameters are introduced into those cores. The new core has 4.54 times larger surface-to-volume ratio than the reference core. The core uranium loading, linear heat generation rate, excess reactivity, and control rod worth as well as the neutron spectra are analysed for each core. (author)

Postirradiation examination of high-U-loaded, low-enriched U3O8, UAl2, and U3Si test fuel plates

International Nuclear Information System (INIS)

Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofman, G.L.; Snelgrove, J.L.

1985-01-01

The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded, low-enriched U 3 O 8 , UAl 2 and U 3 Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examined, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the RERTR Program. Postirradiation examination of these plates showed satisfactory performance for the oxides, aluminides and silicides (except for the highest-loaded U 3 Si plate) with the only indication of detrimental behavior being the slight bowing of some plates at about 80% burnup

Postirradiation examination of high-U-loaded, low-enriched U3O8, UAl2, and U3Si test fuel plates

International Nuclear Information System (INIS)

Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofman, G.L.; Snelgrove, J.L.

1985-01-01

The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded, low-enriched U 3 O 8 , UAl 2 and U 3 Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examined, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the REM Program. Postirradiation examination of these plates showed satisfactory performance for the oxides, aluminides and silicides (except for the highest-loaded U 3 Si plate) with the only indication of detrimental behavior being the slight bowing of some plates at about 80% burnup. (author)

Postirradiation examination of high-U-loaded low-enriched U3O8, UAl2, and U3Si test fuel plates

International Nuclear Information System (INIS)

Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofmann, G.; Snelgrove, J.L.

1984-01-01

The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded low-enriched U 3 O 8 , UAl 2 and U 3 Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examinated, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the RERTR Program. Postirradiation examination of these plates showed satisfactory poerformance for the oxides, aluminides and silicides (except for the highest-loaded U 3 Si plate) with the only indication of detrimental behavior during the slight bowing of some plates at about 80% burnup

Electronic properties of γ-U and superconductivity of U–Mo alloys

International Nuclear Information System (INIS)

Tkach, I.; Kim-Ngan, N.-T.H.; Warren, A.; Scott, T.; Gonçalves, A.P.; Havela, L.

2014-01-01

Highlights: • The bcc phase of uranium was stabilized to low temperature in U–Mo alloys. • Ultrafast cooling was utilized. • Negative coefficient dρ/dT indicates very strong disorder. • The alloys are superconducting with T c ≈ 2.1 K. • They exhibit high critical field exceeding 5 T. - Abstract: Fundamental electronic properties of γ-Uranium were determined using Mo doping combined with ultrafast (splat) cooling, which allowed stabilization of the bcc structure to low temperatures. The Sommerfeld coefficient γ e is enhanced to 16 mJ/mol K 2 from 11 mJ/mol K 2 for α-U. Magnetic susceptibility remains weak and T-independent, ≈5 × 10 −8 m 3 /mol. The Mo-doped γ-U exhibits a conventional BCS superconductivity with T c ≈ 2.1 K and critical field exceeding 5 T for 15 at.% Mo. This type of superconductivity is qualitatively different from the one found for pure U splat, which has T c higher than 1 K but the weak specific heat anomaly proves that it is not real bulk effect

Electroplating of Ni-Mo Coating on Stainless Steel for Application in Proton Exchange Membrane Fuel Cell Bipolar Plate

Directory of Open Access Journals (Sweden)

H. Rashtchi

2018-03-01

Full Text Available Stainless steel bipolar plates are preferred choice for use in Proton Exchange Membrane Fuel Cells (PEMFCs. However, regarding the working temperature of 80 °C and corrosive and acidic environment of PEMFC, it is necessary to apply conductive protective coatings resistant to corrosion on metallic bipolar plate surfaces to enhance its chemical stability and performance. In the present study, by applying Ni-Mo and Ni-Mo-P alloy coatings via electroplating technique, corrosion resistance was improved, oxid layers formation on substrates which led to increased electrical conductivity of the surface was reduced and consequently bipolar plates fuction was enhanced. Evaluation tests included microstructural and phase characterizations for evaluating coating components; cyclic voltammetry test for electrochemical behavior investigations; wettability test for measuring hydrophobicity characterizations of the coatings surfaces; interfacial contact resistance measurements of the coatings for evaluating the composition of applied coatings; and polarization tests of fuel cells for evaluating bipolar plates function in working conditions. Finally, the results showed that the above-mentioned coatings considerably decreased the corrosion and electrical resistance of the stainless steel.

Analyses on the U-Mo/Al Chemical Interaction and the Effects of Diffusion Barrier Coatings

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin; Kim, Woo Jeong; Cho, Woo Hyung; Jeong, Yong Jin; Lee, Yoon Sang; Park, Jong Man; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

While many HEU-fueled research reactors have been converted by adopting LEU U{sub 3}Si{sub 2} fuel in harmony with the Reduced Enrichment for Research and Test Reactors (RERTR) program, some high performance research reactors still need the development of advanced fuels with higher uranium densities. Currently, gamma-phase U-Mo alloys are considered promising candidates to be used as high uranium density fuel for the high performance reactors. For the production of UMo alloy powder, the centrifugal atomization technology developed by KAERI has been considered the most promising way because of high yield production and excellent powder quality when compared with other possible methods such as grinding, machining or hydriding-dehydriding. However, severe pore formation associated with an extensive interaction between the U-Mo and Al matrix, although the irradiation performance of U-Mo itself showed most stable, delay the fuel qualification of UMo fuel for high performance research reactors. Because the reaction products, i.e. uranium aluminides (UAlx), is less dense than the mixed reactants, the volume of the fuel meat increases after formation of interaction layer(IL). In addition to the impact on the swelling performance, the reaction layers between the U-Mo and Al matrix induces a degradation of the thermal conductivities of the U-Mo/Al dispersion fuels. The chemical interaction between the U-Mo and Al matrix are analyzed in this study to find remedies to reduce the growth of the interaction layers during irradiation. In addition, various coating technologies for the formation of diffusion barriers on U-Mo particles are proposed as a result of the analyses

Monolith electroplating process

Science.gov (United States)

Agarrwal, Rajev R.

2001-01-01

An electroplating process for preparing a monolith metal layer over a polycrystalline base metal and the plated monolith product. A monolith layer has a variable thickness of one crystal. The process is typically carried in molten salts electrolytes, such as the halide salts under an inert atmosphere at an elevated temperature, and over deposition time periods and film thickness sufficient to sinter and recrystallize completely the nucleating metal particles into one single crystal or crystals having very large grains. In the process, a close-packed film of submicron particle (20) is formed on a suitable substrate at an elevated temperature. The temperature has the significance of annealing particles as they are formed, and substrates on which the particles can populate are desirable. As the packed bed thickens, the submicron particles develop necks (21) and as they merge into each other shrinkage (22) occurs. Then as micropores also close (23) by surface tension, metal density is reached and the film consists of unstable metal grain (24) that at high enough temperature recrystallize (25) and recrystallized grains grow into an annealed single crystal over the electroplating time span. While cadmium was used in the experimental work, other soft metals may be used.

DART-TM: A thermomechanical version of DART for LEU VHD dispersed and monolithic fuel analysis

International Nuclear Information System (INIS)

Saliba, Roberto; Taboada, Horacio; Moscarda, Ma.Virginia; Rest, Jeff

2003-01-01

A collaboration agreement between ANL/USDOE and CNEA Argentina, in the area of Low Enriched Uranium Advanced Fuels has been in place since October 16, 1997 under the 'Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy'. An annex concerning DART code optimization has been operative since February 8, 1999. Previously, as a part of this annex a visual thermal FASTDART version was developed that includes mechanistic models for the calculation of the fission-gas-bubble and fuel particle size distribution, reaction layer thickness, and meat thermal conductivity. FASTDART was presented at the last RERTR Meeting that included validation against RERTR 3 irradiation data. The thermal FASTDART version was assessed as an adequate tool for modeling the behavior of LEU U-Mo dispersed fuels under irradiation against PIE RERTR irradiation data. During this past year the development of a 3-D thermo-mechanical version of the code for modeling the irradiation behavior of LEU U-Mo monolithic and dispersion fuel was initiated. Some preliminary results of this work will be shown during RERTR-2003 meeting. (author)

Coupled Mo-U abundances and isotopes in a small marine euxinic basin: Constraints on processes in euxinic basins

Science.gov (United States)

Bura-Nakić, Elvira; Andersen, Morten B.; Archer, Corey; de Souza, Gregory F.; Marguš, Marija; Vance, Derek

2018-02-01

Sedimentary molybdenum (Mo) and uranium (U) abundances, as well as their isotope systematics, are used to reconstruct the evolution of the oxygenation state of the surface Earth from the geological record. Their utility in this endeavour must be underpinned by a thorough understanding of their behaviour in modern settings. In this study, Mo-U concentrations and their isotope compositions were measured in the water column, sinking particles, sediments and pore waters of the marine euxinic Lake Rogoznica (Adriatic Sea, Croatia) over a two year period, with the aim of shedding light on the specific processes that control Mo-U accumulation and isotope fractionations in anoxic sediment. Lake Rogoznica is a 15 m deep stratified sea-lake that is anoxic and euxinic at depth. The deep euxinic part of the lake generally shows Mo depletions consistent with near-quantitative Mo removal and uptake into sediments, with Mo isotope compositions close to the oceanic composition. The data also, however, show evidence for periodic additions of isotopically light Mo to the lake waters, possibly released from authigenic precipitates formed in the upper oxic layer and subsequently processed through the euxinic layer. The data also show evidence for a small isotopic offset (∼0.3‰ on 98Mo/95Mo) between particulate and dissolved Mo, even at highest sulfide concentrations, suggesting minor Mo isotope fractionation during uptake into euxinic sediments. Uranium concentrations decrease towards the bottom of the lake, where it also becomes isotopically lighter. The U systematics in the lake show clear evidence for a dominant U removal mechanism via diffusion into, and precipitation in, euxinic sediments, though the diffusion profile is mixed away under conditions of increased density stratification between an upper oxic and lower anoxic layer. The U diffusion-driven precipitation is best described with an effective 238U/235U fractionation of +0.6‰, in line with other studied euxinic

U-10Mo Sample Preparation and Examination using Optical and Scanning Electron Microscopy

Energy Technology Data Exchange (ETDEWEB)

Prabhakaran, Ramprashad [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rhodes, Mark A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schemer-Kohrn, Alan L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Guzman, Anthony D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-10-01

The purpose of this document is to provide guidelines to prepare specimens of uranium alloyed with 10 weight percent molybdenum (U-10Mo) for optical metallography and scanning electron microscopy. This document also provides instructions to set up an optical microscope and a scanning electron microscope to analyze U-10Mo specimens and to obtain the required information.

U-10Mo Sample Preparation and Examination using Optical and Scanning Electron Microscopy

Energy Technology Data Exchange (ETDEWEB)

Prabhakaran, Ramprashad [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rhodes, Mark A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schemer-Kohrn, Alan L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Guzman, Anthony D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-03-30

The purpose of this document is to provide guidelines to prepare specimens of uranium alloyed with 10 weight percent molybdenum (U-10Mo) for optical metallography and scanning electron microscopy. This document also provides instructions to set up an optical microscope and a scanning electron microscope to analyze U-10Mo specimens and to obtain the required information.

Nitride Coating Effect on Oxidation Behavior of Centrifugally Atomized U-Mo Fuel

Energy Technology Data Exchange (ETDEWEB)

Jeong, Yong Jin; Cho, Woo Hyoung; Park, Jong Man; Lee, Yoon Sang; Yang, Jae Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-10-15

Uranium metal and uranium compounds are being used as nuclear fuel materials and generally known as pyrophoric materials. Nowadays the importance of nuclear fuel about safety is being emphasized due to the vigorous exchanges and co-operations among the international community. According to the reduced enrichment for research and test reactors (RERTR) program, the international research reactor community has decided to use low-enriched uranium instead of high-enriched uranium. As a part of the RERTR program, KAERI has developed centrifugally atomized U-Mo alloys as a promising candidate of research reactor fuel. Kang et al. studied the oxidation behavior of centrifugally atomized U-10wt% Mo alloy and it showed better oxidation resistance than uranium. In this study, the oxidation behavior of nitride coated U-7wt% Mo alloy is investigated to enhance the safety against pyrophoricity

POST-IRRADIATION ANALYSES OF U-MO DISPERSION FUEL RODS OF KOMO TESTS AT HANARO

Directory of Open Access Journals (Sweden)

H.J. RYU

2013-12-01

Full Text Available Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4–5 g-U/cm3 were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr, additional protective coatings (silicide or nitride, and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

LEU fuel development at CERCA. Status as of October 1997. Preliminary developments of MTR plates with UMo fuel

International Nuclear Information System (INIS)

Durand, J.P.; Lavastre, Y.; Grasse, M.

1997-01-01

UMo fuels are considered by the RERTR programme because of their higher density as compared to U 3 Si 2 . This paper is focused on the preliminary results about the manufacture feasibility of Uranium/Molybdenum fuel plates carried out by CERCA. A special procedure of casting and heat treatment has been developed in order to get an homogeneous gamma phase of UMo alloy Although U-5%Mo allows to reach densities up to 9.9 U/cm3 with the advanced process developed by CERCA for the high loaded plates, it is not a good candidate on the thermal stability point of view. U-9%Mo alloy seems to gather all the criteria for a good fuel alloy but it is a little less effective on the Uranium density point of view as compared to U-5%Mo alloy. In any case, the preliminary feasibility results are very much encouraging because UMo alloys seem to be compatible with the Aluminium matrix when taking special care while manufacturing. A good compromise could be an intermediate percentage of Molybdenum or the addition of metal traces in order to thermally stabilise 5%Mo. (author)

Powder fabrication of U-Mo alloys for nuclear dispersion fuels

Energy Technology Data Exchange (ETDEWEB)

Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes, E-mail: mdurazzo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

Powder fabrication of U-Mo alloys for nuclear dispersion fuels

International Nuclear Information System (INIS)

Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes

2011-01-01

For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

Development of multilayer perceptron networks for isothermal time temperature transformation prediction of U-Mo-X alloys

Energy Technology Data Exchange (ETDEWEB)

Johns, Jesse M., E-mail: jesse.johns@pnnl.gov; Burkes, Douglas, E-mail: douglas.burkes@pnnl.gov

2017-07-15

In this work, a multilayered perceptron (MLP) network is used to develop predictive isothermal time-temperature-transformation (TTT) models covering a range of U-Mo binary and ternary alloys. The selected ternary alloys for model development are U-Mo-Ru, U-Mo-Nb, U-Mo-Zr, U-Mo-Cr, and U-Mo-Re. These model's ability to predict 'novel' U-Mo alloys is shown quite well despite the discrepancies between literature sources for similar alloys which likely arise from different thermal-mechanical processing conditions. These models are developed with the primary purpose of informing experimental decisions. Additional experimental insight is necessary in order to reduce the number of experiments required to isolate ideal alloys. These models allow test planners to evaluate areas of experimental interest; once initial tests are conducted, the model can be updated and further improve follow-on testing decisions. The model also improves analysis capabilities by reducing the number of data points necessary from any particular test. For example, if one or two isotherms are measured during a test, the model can construct the rest of the TTT curve over a wide range of temperature and time. This modeling capability reduces the cost of experiments while also improving the value of the results from the tests. The reduced costs could result in improved material characterization and therefore improved fundamental understanding of TTT dynamics. As additional understanding of phenomena driving TTTs is acquired, this type of MLP model can be used to populate unknowns (such as material impurity and other thermal mechanical properties) from past literature sources.

INFLUENCE OF FUEL-MATRIX INTERACTION ON THE BREAKAWAY SWELLING OF U-MO DISPERSION FUEL IN AL

OpenAIRE

HO JIN RYU; YEON SOO KIM

2014-01-01

In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model prediction...

Evolution of microstructure of U-Mo alloys in as cast and sintered forms

International Nuclear Information System (INIS)

Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Kamath, H.S.; Dey, G.K.

2009-01-01

Over the years U 3 Si 2 compound dispersed in aluminium matrix has been successfully used as potential Low Enriched Uranium (LEU 235 ) base dispersion fuel in new research and test reactors and also for converting High Enriched Uranium (HEU > 85% U 235 ) cores to LEU in most of the existing research and test reactors. The maximum density achievable with U 3 Si 2 -AI dispersion fuel is around 4.8 g U cm -3 . To achieve a uranium density of 8.0 to 9.0 g U cm -3 in dispersion fuel with aluminium as matrix material, it is required to use γ-stabilized uranium metal powders. At Metallic Fuels Division, R and D efforts are on to develop these high density uranium alloys. Molybdenum plays a crucial role in metastabilising the γ-phase of uranium at room temperature which is very much evident when we see the microstructures of different U-Mo alloys with varying molybdenum concentration as solute atom. The paper describes the role of molybdenum in imparting metastability in U-Mo alloys from their microstructures in as cast and sintered forms. The paper also covers the role of tailored microstructure in U-Mo alloy for the purpose of hydriding and dehydriding treatment to generate alloy powders. (author)

Incomplete deep inelastic processes in 100Mo + 100Mo and 120Sn + 120Sn at 18 and 24 MeV/u

International Nuclear Information System (INIS)

Petrovici, M.

1989-12-01

Experimental evidence on inclomplete deep inelastic process in 100 Mo + 100 Mo at 18.67 MeV/u, 23.75 MeV/u and in 120 Sn + 120 Sn at 18.34 MeV/u are presented. Such a mechanism is responsible for strong deviations observed at these incident energies in σ 2 Z -TKEL/l g (for two-body) and P 3 /(P 2 + P 3 )-TKEL (for three-body) systematics. Calculations which predict the number of preequilibrium emitted nucleons and the corresponding excitation energy per nucleon that remains in the dinuclear system could explain the observed discrepancies. (author)

Influence of fuel-matrix interaction on the breakaway swelling of U-Mo dispersion fuel in Al

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Yeon Soo [Nuclear Engineering Division, Argonne National Laboratory, Arogonne (United States)

2014-04-15

In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model predictions, advantageous fuel design parameters are recommended to prevent breakaway swelling.

Effect of the Zr elements with thermal properties changes of U-7Mo-xZr/Al dispersion fuel

International Nuclear Information System (INIS)

Supardjo; Agoeng Kadarjono; Boybul; Aslina Br Ginting

2016-01-01

Thermal properties data of nuclear fuel is required as input data to predict material properties change phenomenon during the fabrication process and irradiated in a nuclear reactor. Study the influence of Zr element in the U-7Mo-xZr/Al (x = 1%, 2% and 3%) fuel dispersion to changes in the thermal properties at various temperatures have been stiffened. Thermal analysis includes determining the melting temperature, enthalpy, and phase changes made using Differential Thermal Analysis (DTA) in the temperature range between 30 °C up to 1400 °C, while the heat capacity of U-7Mo-xZr alloy and U-7Mo-xZr/Al dispersion fuel using Differential Scanning Calorimeter (DSC) at room temperature up to 450 °C. Thermal analyst data DTA shows that Zr levels of all three compositions showed a similar phenomenon. At temperatures between 565.60 °C - 584.98 °C change becomes α + δ to α + γ phase and at 649.22 °C – 650.13 °C happen smelting Al matrix Occur followed by a reaction between Al matrix with U-7Mo-xZr on 670.38 °C - 673.38 °C form U (Al, Mo)x Zr. Furthermore a phase change α + β becomes β + γ Occurs at temperatures 762.08 °C - 776.33 °C and diffusion between the matrix by U-7Mo-xZr/Al on 853.55 °C - 875.20 °C. Every phenomenon that Occurs, enthalpy posed a relative stable. Consolidation of uranium Occur in 1052.42 °C - 1104.99 °C and decomposition reaction of U (Al, Mo)x and U (Al, Zr)_x becomes (UAl_4, UAl_3, UAl_2), U-Mo, and UZr on 1328,34 °C - 1332,06 °C , The existence of Zr in U-Mo alloy increases the heat capacity of the U-7Mo-xZr/Al, dispersion fuel and the higher heat capacity of Zr levels increased due to interactions between the atoms of Zr with Al matrix so that the heat absorbed by the fuel increase. (author)

U-Mo Alloy Powder Obtained Through Selective Hydriding. Particle Size Control

International Nuclear Information System (INIS)

Balart, S.N.; Bruzzoni, P.; Granovsky, M.S.

2002-01-01

Hydride-dehydride methods to obtain U-Mo alloy powder for high-density fuel elements have been successfully tested by different authors. One of these methods is the selective hydriding of the α phase (HSα). In the HSα method, a key step is the partial decomposition of the γ phase (retained by quenching) to α phase and an enriched γ phase or U 2 Mo. This transformation starts mainly at grain boundaries. Subsequent hydrogenation of this material leads to selective hydriding of the α phase, embrittlement and intergranular fracture. According to this picture, the particle size of the final product should be related to the γ grain size of the starting alloy. The feasibility of controlling the particle size of the product by changing the γ grain size of the starting alloy is currently investigated. In this work an U-7 wt% Mo alloy was subjected to various heat treatments in order to obtain different grain sizes. The results on the powder particle size distribution after applying the HSα method to these samples show that there is a strong correlation between the original γ grain size and the particle size distribution of the powder. (author)

Development of low enrichment technologies for high density fuels and for isotope production targets

International Nuclear Information System (INIS)

Taboada, Horacio; Gonzalez, Alfredo G.

2005-01-01

Since more than twenty years ago, CNEA has carried out RERTR activities. Main goals are to convert the RA 6 reactor core from HEU to LEU, to get a comprehensive understanding of U-Mo/Al compounds phase formation in dispersed and monolithic fuels, to develop possible solutions to VHD dispersed and monolithic fuels technical problems, and to optimize techniques to recover U from silicide scrap samples. The future plans include: 1) Completion the RA 6 reactor conversion to LEU; 2) Qualification by irradiation of the promising solutions found for the high density fuels; 3) Irradiation of mini plates and full scale fuel assemblies at the RA 3 reactor and at higher flux and temperature reactors; 4) Optimization of LEU target and radiochemical techniques for radioisotope production. (author) [es

Crystallographic study of Si and ZrN coated U–Mo atomised particles and of their interaction with al under thermal annealing

International Nuclear Information System (INIS)

Zweifel, T.; Palancher, H.; Leenaers, A.; Bonnin, A.; Honkimaki, V.; Tucoulou, R.; Van Den Berghe, S.; Jungwirth, R.; Charollais, F.; Petry, W.

2013-01-01

A new type of high density fuel is needed for the conversion of research and test reactors from high to lower enriched uranium. The most promising one is a dispersion of atomized uranium-molybdenum (U–Mo) particles in an Al matrix. However, during in-pile irradiation the growth of an interaction layer between the U–Mo and the Al matrix strongly limits the fuel’s performance. To improve the in-pile behaviour, the U–Mo particles can be coated with protective layers. The SELENIUM (Surface Engineering of Low ENrIched Uranium–Molybdenum) fuel development project consists of the production, irradiation and post-irradiation examination of 2 flat, full-size dispersion fuel plates containing respectively Si and ZrN coated U–Mo atomized powder dispersed in a pure Al matrix. In this paper X-ray diffraction analyses of the Si and ZrN layers after deposition, fuel plate manufacturing and thermal annealing are reported. It was found for the U–Mo particles coated with ZrN (thickness 1 μm), that the layer is crystalline, and exhibits lower density than the theoretical one. Fuel plate manufacturing does not strongly influence these crystallographic features. For the U–Mo particles coated with Si (thickness 0.6 μm), the measurements of the as received material suggest an amorphous state of the deposited layer. Fuel plate manufacturing strongly modifies its composition: Si reacts with the U–Mo particles and the Al matrix to grow U(Al, Si) 3 and U 3 Si 5 phases. Finally both coatings have shown excellent performances under thermal treatment by limiting drastically the U–Mo/Al interdiffusion

Fuel performance of rod-type research reactor fuel using a centrifugally atomized U-Mo powder

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Man; Lee, Yoon Sang; Kim, Chang Kyu

2009-01-01

A low enriched uranium nuclear fuel for research reactors has been developed in order to replace a highly enriched uranium fuel according to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program. In KAERI, a rod-type U 3 Si dispersion fuel has been developed for a localization of the HANARO fuel and a U 3 Si/Al dispersion fuel of 3.15 gU/cc has been used at HANARO as a driver fuel since 2005. Although uranium silicide dispersion fuels such as U 3 Si 2 /Al and U 3 Si/Al are being used widely, high uranium density dispersion fuels (8-9 g/cm 3 ) are required for some high performance research reactors. U-Mo alloys have been considered as one of the most promising uranium alloys for a dispersion fuel due to their good irradiation performance. An international qualification program on U-Mo fuel to replace a uranium silicide dispersion fuel with a U-Mo dispersion fuel has been carried out

Phase development in a U-7 wt.% Mo vs. Al-7 wt.% Ge diffusion couple

Science.gov (United States)

Perez, E.; Keiser, D. D.; Sohn, Y. H.

2013-10-01

Fuel development for the Reduced Enrichment for Research and Test Reactors (RERTR) program has demonstrated that U-Mo alloys in contact with Al develop interaction regions with phases that have poor irradiation behavior. The addition of Si to the Al has been considered with positive results. In this study, compositional modification is considered by replacing Si with Ge to determine the effect on the phase development in the system. The microstructural and phase development of a diffusion couple of U-7 wt.% Mo in contact with Al-7 wt.% Ge was examined by transmission electron microscopy, scanning electron microscopy and energy dispersive spectroscopy. The interdiffusion zone developed a microstructure that included the cubic-UGe3 phase and amorphous phases. The UGe3 phase was observed with and without Mo and Al solid solution developing a (U,Mo)(Al,Ge)3 phase.

Modeling of the behavior under fuel dispersed irradiation of U-Mo with aluminum matrix from the thermal point of view and its interrelationship with the interdiffusion phase fuel / matrix

International Nuclear Information System (INIS)

Moscarda, Maria V.; Taboada, Horacio H.; Rest, J.

2009-01-01

Results from postirradiation examinations of U-Mo / Al dispersion fuels plates denotes a strong interrelation and feedback between the fuel-matrix interaction and the fuel temperature, bringing undesired consequences on the total swelling and behavior under irradiation. The present work approaches this problem, modeling the profile of temperatures moment by moment to be able to evaluate the increase of this interaction. The Fast Dart program is used, optimized version of program Dart, developed by Dr. J. Rest in collaboration with Dr. H. Taboada. A subroutine of thermal calculation was implemented in this code, which allowed to calculate the evolution of the interaction between the fuel and the matrix. The results of simulations are compared with the results of postirradiation examinations realized by the Reduced Enrichment for Research and Test Reactors International Program. In particular, a good adjustment in the calculation of the depth of interdiffusion U-Mo/Al is observed, demonstrating a right estimation of the profile of temperatures on the fuel plate. It is considered necessary the inclusion of a model that describes the phases that form in the zone of interaction, denoting its thermal dependency and effects due to the radiation damage. (author)

A model for recovery of scrap monolithic uranium molybdenum fuel by electrorefining

Science.gov (United States)

Van Kleeck, Melissa A.

The goal of the Reduced Enrichment for Research and Test Reactors program (RERTR) is toreduce enrichment at research and test reactors, thereby decreasing proliferation risk at these facilities. A new fuel to accomplish this goal is being manufactured experimentally at the Y12 National Security Complex. This new fuel will require its own waste management procedure,namely for the recovery of scrap from its manufacture. The new fuel is a monolithic uraniummolybdenum alloy clad in zirconium. Feasibility tests were conducted in the Planar Electrode Electrorefiner using scrap U-8Mo fuel alloy. These tests proved that a uranium product could be recovered free of molybdenum from this scrap fuel by electrorefining. Tests were also conducted using U-10Mo Zr clad fuel, which confirmed that product could be recovered from a clad version of this scrap fuel at an engineering scale, though analytical results are pending for the behavior of Zr in the electrorefiner. A model was constructed for the simulation of electrorefining the scrap material produced in the manufacture of this fuel. The model was implemented on two platforms, Microsoft Excel and MatLab. Correlations, used in the model, were developed experimentally, describing area specific resistance behavior at each electrode. Experiments validating the model were conducted using scrap of U-10Mo Zr clad fuel in the Planar Electrode Electrorefiner. The results of model simulations on both platforms were compared to experimental results for the same fuel, salt and electrorefiner compositions and dimensions for two trials. In general, the model demonstrated behavior similar to experimental data but additional refinements are needed to improve its accuracy. These refinements consist of a function for surface area at anode and cathode based on charge passed. Several approximations were made in the model concerning areas of electrodes which should be replaced by a more accurate function describing these areas.

U-10Mo Sample Preparation and Examination using Optical and Scanning Electron Microscopy. Rev. 1

International Nuclear Information System (INIS)

Prabhakaran, Ramprashad; Joshi, Vineet V.; Rhodes, Mark A.; Schemer-Kohrn, Alan L.; Guzman, Anthony D.; Lavender, Curt A.

2016-01-01

The purpose of this document is to provide guidelines to prepare specimens of uranium alloyed with 10 weight percent molybdenum (U-10Mo) for optical metallography and scanning electron microscopy. This document also provides instructions to set up an optical microscope and a scanning electron microscope to analyze U-10Mo specimens and to obtain the required information.

Porous polymer monoliths functionalized through copolymerization of a C60 fullerene-containing methacrylate monomer for highly efficient separations of small molecules

KAUST Repository

Chambers, Stuart D.

2011-12-15

Monolithic poly(glycidyl methacrylate-co-ethylene dimethacrylate) and poly(butyl methacrylate-co-ethylene dimethacrylate) capillary columns, which incorporate the new monomer [6,6]-phenyl-C 61-butyric acid 2-hydroxyethyl methacrylate ester, have been prepared and their chromatographic performance have been tested for the separation of small molecules in the reversed phase. While addition of the C60-fullerene monomer to the glycidyl methacrylate-based monolith enhanced column efficiency 18-fold, to 85 000 plates/m at a linear velocity of 0.46 mm/s and a retention factor of 2.6, when compared to the parent monolith, the use of butyl methacrylate together with the carbon nanostructured monomer afforded monolithic columns with an efficiency for benzene exceeding 110 000 plates/m at a linear velocity of 0.32 mm/s and a retention factor of 4.2. This high efficiency is unprecedented for separations using porous polymer monoliths operating in an isocratic mode. Optimization of the chromatographic parameters affords near baseline separation of 6 alkylbenzenes in 3 min with an efficiency of 64 000 plates/m. The presence of 1 wt % or more of water in the polymerization mixture has a large effect on both the formation and reproducibility of the monoliths. Other factors such as nitrogen exposure, polymerization conditions, capillary filling method, and sonication parameters were all found to be important in producing highly efficient and reproducible monoliths. © 2011 American Chemical Society.

Status of the back-end optional advanced research reactor fuel development in Korea

International Nuclear Information System (INIS)

Kim, Chang-Kyu; Lee, Yoon-Sang; Lee, Don-Bae; Oh, Seuk-Jin; Kim, Ki-Hwan; Chae, Hee-Taek; Park, Jong-Man; Sohn Dong-Seong

2003-01-01

U-Mo fuel development has been carried out for a reactor upgrade of HANARO and the back-end option in Korea. The 2nd irradiation test of the U-Mo dispersion rod fuels is underway in HANARO in order to find the optimum uranium loading density and to investigate the applicability of the monolithic U-Mo ring fuel as well as other parameters such as particle size and cladding surface-treatment. The optical observation using an immersion camera showed that the cladding surfaces of the two U 3 Si and U-Mo fuels with a high power rate changed in to the darker color, which is not as severe as those of the driving fuels in HANARO. Presumably it would be acceptable. The other fuels were observed as maintaining their initial good conditions. In connection with monolithic U-Mo fuel development, some achievements such as preliminary U-Mo tube production by a continuous casting process and a successful U-Mo foil production using a roll casting process have been obtained. In addition, some investigation on the surface-treatment of multilayer coating and Zr sputtering coating has showed the possibility of eliminating the problem of a temperature rise due to the corrosion layer formation having quite a low conductivity. The next irradiation test will aim mainly at the qualification of the U-Mo dispersion fuel for HANARO around the end of next year. In the 3rd irradiation fuel bundle, some fuels related to the basic investigation tests for the monolithic U-Mo fuel and surface-treatment for anticorrosion will be loaded. (author)

Modeling a failure criterion for U-Mo/Al dispersion fuel

Science.gov (United States)

Oh, Jae-Yong; Kim, Yeon Soo; Tahk, Young-Wook; Kim, Hyun-Jung; Kong, Eui-Hyun; Yim, Jeong-Sik

2016-05-01

The breakaway swelling in U-Mo/Al dispersion fuel is known to be caused by large pore formation enhanced by interaction layer (IL) growth between fuel particles and Al matrix. In this study, a critical IL thickness was defined as a criterion for the formation of a large pore in U-Mo/Al dispersion fuel. Specifically, the critical IL thickness is given when two neighboring fuel particles come into contact with each other in the developed IL. The model was verified using the irradiation data from the RERTR tests and KOMO-4 test. The model application to full-sized sample irradiations such as IRISs, FUTURE, E-FUTURE, and AFIP-1 tests resulted in conservative predictions. The parametric study revealed that the fuel particle size and the homogeneity of the fuel particle distribution are influential for fuel performance.

Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

Science.gov (United States)

Chang, G. S.; Lillo, M. A.

2009-08-01

The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y

Development of U-Mo Research Reactor Fuel for Next Generation

International Nuclear Information System (INIS)

Park, Jong Man; Lee, Y. S.; Yang, J. H.; Ryu, H. J.; Kim, C. K.; Chae, H. T.; Seo, C. G.

2010-08-01

- Exportation of centrifugal atomized U-Mo powder - Completion of post irradiation examination for KOMO-3 irradiated fuel rods. - Select the dispersion fuel rod candidates for KOMO-4 irradiation test. - Irradiation test to solve the problems of interaction layer formation (KOMO-4) - Set the post irradiation examination of KOMO-4 irradiated fuel rods. - Development and characterization of innovative high U density fuel rods - Obtain and analyze foreign new irradiation test D

Decomposition of the metastable phase γU in U-7% and U-7% Mo-0.9% Pt

International Nuclear Information System (INIS)

Arico, Sergio F.; Gribaudo, Luis M.

2004-01-01

The 'Reduced Enrichment for Research and Test Reactors' is an international project for the development of a nuclear fuel with high density in uranium capable to get a great neutron flux with good capacity for being reprocessed. One of the candidates is a fuel containing U-Mo alloy powder, as bcc metastable phase γ, dispersed in Al powder. In order to know the influence of Pt as a stabilizing element two U-7 wt.% Mo alloys are studied, one of them with 0.9 wt.% Pt. They were fabricated in an arc furnace and both homogenized in composition during 2 h at 1000 C degrees. Then, isothermal treatments at 480, 430 and 350 C degrees were performed at times between 1 and 177 h. The decomposition of the γ phase was studied by metallography and X-ray diffraction analysis. Adding Pt, the start of the decomposition of the γ phase is delayed, but the initial grain size of the alloys is an important variable which has also to be considered. (author) [es

U-8 wt %Mo and 7 wt %Mo alloys powder obtained by an hydride-de hydride process

International Nuclear Information System (INIS)

Balart, Silvia N.; Bruzzoni, Pablo; Granovsky, Marta S.; Gribaudo, Luis M. J.; Hermida, Jorge D.; Ovejero, Jose; Rubiolo, Gerardo H.; Vicente, Eduardo E.

2000-01-01

Uranium-molybdenum alloys are been tested as a component in high-density LEU dispersion fuels with very good performances. These alloys need to be transformed to powder due to the manufacturing requirements of the fuels. One method to convert ductile alloys into powder is the hydride-de hydride process, which takes advantage of the ability of the U-α phase to transform to UH 3 : a brittle and relatively low-density compound. U-Mo alloys around 7 and 8 wt % Mo were melted and heat treated at different temperature ranges in order to partially convert γ -phase to α -phase. Subsequent hydriding transforms this α -phase to UH 3 . The volume change associated to the hydride formation embrittled the material which ends up in a powdered alloy. Results of the optical metallography, scanning electron microscopy, X-ray diffraction during different steps of the process are shown. (author)

Corrosion resistance of zinc-nickel plated U-O.75 Ti

International Nuclear Information System (INIS)

Dini, J.W.; Johnson, H.R.

1979-09-01

As part of a program for the US Army directed at improving the corrosion performance of U-0.75 Ti, specimens were coated with Zn-10 Ni alloy electroplate and then subjected to various corrosion tests. This work revealed that the Zn-Ni coatings provided good protection for U-0.75 Ti in salt fog and in non-sealed moist-nitrogen systems. In sealed, moist-nitrogen environments the Zn-Ni coatings deteriorated quickly and provided no protection. Some plating with Zn alone, using some of the new non-cyanide plating solutions, was also attempted, but the results were inconsistent

Atomistic simulations of thermodynamic properties of Xe gas bubbles in U10Mo fuels

Energy Technology Data Exchange (ETDEWEB)

Hu, Shenyang, E-mail: shenyang.hu@pnnl.gov; Setyawan, Wahyu; Joshi, Vineet V.; Lavender, Curt A.

2017-07-15

Xe gas bubble superlattice formation is observed in irradiated uranium–10 wt% molybdenum (U10Mo) fuels. However, the thermodynamic properties of the bubbles (the relationship among bubble size, equilibrium Xe concentration, and bubble pressure) and the mechanisms of bubble superlattice formation are not well known. In this work, the molecular dynamics (MD) method is used to study these properties and mechanisms. The results provide important inputs for quantitative mesoscale models of gas bubble evolution and fuel performance. In the MD simulations, the embedded-atom method (EAM) potential of U10Mo-Xe [1] is employed. Initial gas bubbles with a low Xe concentration (underpressured) are generated in a body-centered cubic (bcc) U10Mo single crystal. Then Xe atoms are sequentially added into the bubbles one by one, and the evolution of pressure and dislocation emission around the bubbles is analyzed. The relationship between pressure, equilibrium Xe concentration, and radius of the bubbles is established. It was found that an overpressured gas bubble emits partial dislocations with a Burgers vector along the <111> direction and a slip plane of (11-2). Meanwhile, dislocation loop punch out was not observed. The overpressured bubble also induces an anisotropic stress field. A tensile stress was found along <110> directions around the bubble, favoring the nucleation and formation of a face-centered cubic bubble superlattice in bcc U10Mo fuels.

Atomistic simulations of thermodynamic properties of Xe gas bubbles in U10Mo fuels

Science.gov (United States)

Hu, Shenyang; Setyawan, Wahyu; Joshi, Vineet V.; Lavender, Curt A.

2017-07-01

Xe gas bubble superlattice formation is observed in irradiated uranium-10 wt% molybdenum (U10Mo) fuels. However, the thermodynamic properties of the bubbles (the relationship among bubble size, equilibrium Xe concentration, and bubble pressure) and the mechanisms of bubble superlattice formation are not well known. In this work, the molecular dynamics (MD) method is used to study these properties and mechanisms. The results provide important inputs for quantitative mesoscale models of gas bubble evolution and fuel performance. In the MD simulations, the embedded-atom method (EAM) potential of U10Mo-Xe [1] is employed. Initial gas bubbles with a low Xe concentration (underpressured) are generated in a body-centered cubic (bcc) U10Mo single crystal. Then Xe atoms are sequentially added into the bubbles one by one, and the evolution of pressure and dislocation emission around the bubbles is analyzed. The relationship between pressure, equilibrium Xe concentration, and radius of the bubbles is established. It was found that an overpressured gas bubble emits partial dislocations with a Burgers vector along the direction and a slip plane of (11-2). Meanwhile, dislocation loop punch out was not observed. The overpressured bubble also induces an anisotropic stress field. A tensile stress was found along directions around the bubble, favoring the nucleation and formation of a face-centered cubic bubble superlattice in bcc U10Mo fuels.

Microporous spongy chitosan monoliths doped with graphene oxide as highly effective adsorbent for methyl orange and copper nitrate (Cu(NO3)2) ions.

Science.gov (United States)

Wang, Ying; Liu, Xu; Wang, Hongfang; Xia, Guangmei; Huang, Wei; Song, Rui

2014-02-15

In the current study, microporous spongy chitosan monoliths doped with small amount of graphene oxide (CSGO monoliths) with high porosity (96-98%), extraordinary high water absorption (more than 2000%) and low density (0.0436-0.0607 g cm(-3)) were prepared by the freeze-drying method and used as adsorbents for anionic dyes methyl orange (MO) and Cu(2+) ions. The adsorption behavior of the CSGO monoliths and influencing factors such as pH value, graphene oxide (GO) content, concentration of pollutants as well as adsorption kinetics were studied. Specifically, the saturated adsorption capacity for MO is 567.07 mg g(-1), the highest comparing with other publication results, and it is 53.69 mg g(-1) for Cu(2+) ions. Since they are biodegradable, non-toxic, efficient, low-cost and easy to prepare, we believe that these microporous spongy CSGO monoliths will be the promising candidates for water purification. Copyright © 2013 Elsevier Inc. All rights reserved.

Microstructure and mechanical properties of friction stir welded 18Cr–2Mo ferritic stainless steel thick plate

International Nuclear Information System (INIS)

Han, Jian; Li, Huijun; Zhu, Zhixiong; Barbaro, Frank; Jiang, Laizhu; Xu, Haigang; Ma, Li

2014-01-01

Highlights: • We focus on friction stir welding of 18Cr–2Mo ferritic stainless steel thick plate. • We produce high-quality joints with special tool and optimised welding parameters. • We compare microstructure and mechanical properties of steel and joint. • Friction stir welding is a method that can maintain the properties of joint. - Abstract: In this study, microstructure and mechanical properties of a friction stir welded 18Cr–2Mo ferritic stainless steel thick plate were investigated. The 5.4 mm thick plates with excellent properties were welded at a constant rotational speed and a changeable welding speed using a composite tool featuring a chosen volume fraction of cubic boron nitride (cBN) in a W–Re matrix. The high-quality welds were successfully produced with optimised welding parameters, and studied by means of optical microscopy (OM), scanning electron microscopy (SEM), electron back-scattered diffraction (EBSD) and standard hardness and impact toughness testing. The results show that microstructure and mechanical properties of the joints are affected greatly, which is mainly related to the remarkably fine-grained microstructure of equiaxed ferrite that is observed in the friction stir welded joint. Meanwhile, the ratios of low-angle grain boundary in the stir zone regions significantly increase, and the texture turns strong. Compared with the base material, mechanical properties of the joint are maintained in a comparatively high level

Superconductivity in U-T alloys (T = Mo, Pt, Pd, Nb, Zr stabilized in the cubic γ-U structure by splat-cooling technique

Directory of Open Access Journals (Sweden)

N.-T.H. Kim-Ngan

2016-06-01

Full Text Available We succeed to retain the high-temperature (cubic γ-U phase down to low temperatures in U-T alloys with less required T alloying concentration (T = Mo, Pt, Pd, Nb, Zr by means of splat-cooling technique with a cooling rate better than 106 K/s. All splat-cooled U-T alloys become superconducting with the critical temperature Tc in the range of 0.61 K–2.11 K. U-15 at.% Mo splat consisting of the γ-U phase with an ideal bcc A2 structure is a BCS superconductor having the highest critical temperature (2.11 K.

The Effect of Rolling As-Cast and Homogenized U-10Mo Samples on the Microstructure Development and Recovery Curves

Energy Technology Data Exchange (ETDEWEB)

Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Paxton, Dean M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-07-30

Over the past several years Pacific Northwest National Laboratory (PNNL) has been actively involved in supporting the U.S. Department of Energy National Nuclear Security Administration Office of Material Management and Minimization (formerly Global Threat Reduction Initiative). The U.S. High- Power Research Reactor (USHPRR) project is developing alternatives to existing highly enriched uranium alloy fuel to reduce the proliferation threat. One option for a high-density metal fuel is uranium alloyed with 10 wt% molybdenum (U-10Mo). Forming the U-10Mo fuel plates/foils via rolling is an effective technique and is actively being pursued as part of the baseline manufacturing process. The processing of these fuel plates requires systematic investigation/understanding of the pre- and post-rolling microstructure, end-state mechanical properties, residual stresses, and defects, their effect on the mill during processing, and eventually, their in-reactor performance. In the work documented herein, studies were conducted to determine the effect of cold and hot rolling the as-cast and homogenized U-10Mo on its microstructure and hardness. The samples were homogenized at 900°C for 48 h, then later annealed for several durations and temperatures to investigate the effect on the material’s microstructure and hardness. The rolling of the as-cast plate, both hot and cold, was observed to form a molybdenum-rich and -lean banded structure. The cold rolling was ineffective, and in some cases exacerbated the as-cast defects. The grains elongated along the rolling direction and formed a pancake shape, while the carbides fractured perpendicularly to the rolling direction and left porosity between fractured particles of UC. The subsequent annealing of these samples at sub-eutectoid temperatures led to rapid precipitation of the ' lamellar phase, mainly in the molybdenum-lean regions. Annealing the samples above the eutectoid temperature did not refine the grain size or the banded

Thermo-mechanical treatment of the Cr-Mo constructional steel plates with Nb, Ti and B additions

International Nuclear Information System (INIS)

Adamczyk, J.; Opiela, M.

2002-01-01

Results of investigations of the influence of parameters of thermomechanical treatment, carried out by rolling with controlled recrystallization, on the microstructure and mechanical properties of Cr-Mo constructional steel with Nb, Ti and B microadditions, destined for the manufacturing of weldable heavy plates, are presented. These plates show a yield point of over 960 MPa after heat treatment. Two variants of thermomechanical treatment were worked out, based on the obtained results of investigations, when rolling a plate 40 mm thick in several passes to a plate 15 mm thick in a temperature range from 1100 to 900 o C. It was found that the lack of complete recrystallization of the austenite in the first rolling variant, leads to localization of plastic deformation in form of shear bands. There exists a segregation of MC-type carbides and alloying elements in these bands, causing a distinctive reduction of the crack resistance of the steel, as also a disadvantageous anisotropy of plastic properties of plate after tempering. For plates rolled under the same conditions, using a retention shield, a nearly three times higher impact energy in - 40 o C was obtained, as also only a slight anisotropy of plastic properties, saving the required mechanical properties. (author)

Chromatographic assessment of two hybrid monoliths prepared via epoxy-amine ring-opening polymerization and methacrylate-based free radical polymerization using methacrylate epoxy cyclosiloxane as functional monomer.

Science.gov (United States)

Wang, Hongwei; Ou, Junjie; Lin, Hui; Liu, Zhongshan; Huang, Guang; Dong, Jing; Zou, Hanfa

2014-11-07

Two kinds of hybrid monolithic columns were prepared by using methacrylate epoxy cyclosiloxane (epoxy-MA) as functional monomer, containing three epoxy moieties and one methacrylate group. One column was in situ fabricated by ring-opening polymerization of epoxy-MA and 1,10-diaminodecane (DAD) using a porogenic system consisting of isopropanol (IPA), H2O and ethanol at 65°C for 12h. The other was prepared by free radical polymerization of epoxy-MA and ethylene dimethacrylate (EDMA) using 1-propanol and 1,4-butanediol as the porogenic solvents at 60°C for 12h. Two hybrid monoliths were investigated on the morphology and chromatographic assessment. Although two kinds of monolithic columns were prepared with epoxy-MA, their morphologies looked rather different. It could be found that the epoxy-MA-DAD monolith possessed higher column efficiencies (25,000-34,000plates/m) for the separation of alkylbenzenes than the epoxy-MA-EDMA monolith (12,000-13,000plates/m) in reversed-phase nano-liquid chromatography (nano-LC). Depending on the remaining epoxy or methacrylate groups on the surface of two pristine monoliths, the epoxy-MA-EDMA monolith could be easily modified with 1-octadecylamine (ODA) via ring-opening reaction, while the epoxy-MA-DAD monolith could be modified with stearyl methacrylate (SMA) via free radical reaction. The chromatographic performance for the separation of alkylbenzenes on SMA-modified epoxy-MA-DAD monolith was remarkably improved (42,000-54,000 plates/m) when compared with that on pristine epoxy-MA-DAD monolith, while it was not obviously enhanced on ODA-modified epoxy-MA-EDMA monolith when compared with that on pristine epoxy-MA-EDMA monolith. The enhancement of the column efficiency of epoxy-MA-DAD monolith after modification might be ascribed to the decreased mass-transfer resistence. The two kinds of hybrid monoliths were also applied for separations of six phenols and seven basic compounds in nano-LC. Copyright © 2014 Elsevier B.V. All

A facile route to large-scale synthesis MoO2 and MoO3 as electrode materials for high-performance supercapacitors

International Nuclear Information System (INIS)

Xuan, H.C.; Du, Y.W.; Zhang, Y.Q.; Xu, Y.K.; Li, H.; Han, P.D.; Wang, D.H.

2016-01-01

MoO 3 and MoO 2 materials have been successfully synthesized by thermal decomposition of ammonium paramolybdate in air and a sealed quartz tube, respectively. The microstructure of as-synthesized MoO 3 is composed of irregular lamellar plates with a plate thickness around 100 nm and MoO 2 has the larger grain size with lamellar plates connected with each other. A maximum specific capacitance of 318 F/g at 0.5 A/g is obtained for MoO 2 prepared in a closed environment. On the other hand, the sample MoO 3 exhibits excellent rate capacity with specific capacitances of 218, 209, 196, 188, 176, and 160 F/g at current densities of 0.5, 1, 2, 3, 4, and 5 A/g, respectively. These results pave the way to consider MoO 3 and MoO 2 as prospective materials for energy-storage applications. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Fiber-based monolithic columns for liquid chromatography.

Science.gov (United States)

Ladisch, Michael; Zhang, Leyu

2016-10-01

Fiber-based monoliths for use in liquid chromatographic separations are defined by columns packed with aligned fibers, woven matrices, or contiguous fiber structures capable of achieving rapid separations of proteins, macromolecules, and low molecular weight components. A common denominator and motivating driver for this approach, first initiated 25 years ago, was reducing the cost of bioseparations in a manner that also reduced residence time of retained components while achieving a high ratio of mass to momentum transfer. This type of medium, when packed into a liquid chromatography column, minimized the fraction of stagnant liquid and resulted in a constant plate height for non-adsorbing species. The uncoupling of dispersion from eluent flow rate enabled the surface chemistry of the stationary phase to be considered separately from fluid transport phenomena and pointed to new ways to apply chemistry for the engineering of rapid bioseparations. This paper addresses developments and current research on fiber-based monoliths and explains how the various forms of this type of chromatographic stationary phase have potential to provide new tools for analytical and preparative scale separations. The different stationary phases are discussed, and a model that captures the observed constant plate height as a function of mobile phase velocity is reviewed. Methods that enable hydrodynamically stable fiber columns to be packed and operated over a range of mobile phase flow rates, together with the development of new fiber chemistries, are shown to provide columns that extend the versatility of liquid chromatography using monoliths, particularly at the preparative scale. Graphical Abstract Schematic representation of a sample mixture being separated by a rolled-stationary phase column, resulting separated peaks shown in the chromatogram.

Examinations of the irradiation behaviour of U3Si2 test fuel plates with low enrichment

International Nuclear Information System (INIS)

Muellauer, J.

1989-01-01

Five low-enriched (19.7% 235 U), high-density (4.7 gU/cm/ 3 ) U 3 Si 2 -test fuel plates (miniplates) with different fine grain contents have been qualified under irradiation. During the course of irradiation up to burnup of 63% 235 U depletion, no released fractions of gaseous or solid fission products from the fuel plate to the rig coolant were detected. The measured swelling rate of the fuel zone (meat) is less than 0.45% ΔV/10 20 fissions/cm 3 the blister-threshold temperature of the fuel plates is above 520 0 C. The favourable irradiation behavior of the U 3 Si 2 fuel plates was not influenced by using higher amounts of fine grained particles (40% [de

Thermodynamic stabilities of MO2+x(s) (M = U, Np, Pu and Am), pourbaix diagrams

International Nuclear Information System (INIS)

Vitorge, Pierre; Faure, Marie-Helene; Vercouter, Thomas; Capdevila, Helene; Maillard, Serge

2002-01-01

The experimental solubilities of the hydrated amorphous freshly precipitated M(OH) z (am) and MO 2 (OH) z (am) compounds are often used as an upper limit for the safety assessments of deep waste repositories, since these compounds slowly transform to less soluble ones, as typically M(OH) 4 (am) to MO 2 (cr). Solubility (vs. redox potential) at pH=8, and E-pH predominance diagrams are plotted in aqueous solutions at 25degC by using thermodynamic data recently selected by the NEA-TDB review, or estimated by using classical chemical analogies for the non-redox reactions. The solubilities and relative stabilities are also calculated for the MO 2+x (s) crystalline compounds of known stabilities: U 4 O 9 (s), U 3 O 7 (s), U 3 O 8 (s) and Np 2 O 5 (s) where 2+x = 2.25, 2.33, 2.67 and 2.5 respectively. The stabilities of the other MO 2+x (s) compounds are estimated by analogy: M 4 O 9 (s) (M=U, Np, Pu), M 3 O 7 (s) and M 3 O 8 (s) (M=U, Pu), and M 2 O 5 (s) (M=Np, Am) are predicted to be more stable (i.e. less soluble), than the amorphous hydroxides. However their precipitation have never been observed at room temperature possibly for kinetic reasons or difficulties in interpreting solubility experiments. (author)

Post-Irradiation Non-Destructive Analyses of the AFIP-7 Experiment

Science.gov (United States)

Williams, W. J.; Robinson, A. B.; Rabin, B. H.

2017-12-01

This article reports the results and interpretation of post-irradiation non-destructive examinations performed on four curved full-size fuel plates that comprise the AFIP-7 experiment. These fuel plates, having a U-10 wt.%Mo monolithic design, were irradiated under moderate operating conditions in the Advanced Test Reactor to assess fuel performance for geometries that are prototypic of research reactor fuel assemblies. Non-destructive examinations include visual examination, neutron radiography, profilometry, and precision gamma scanning. This article evaluates the qualitative and quantitative data taken for each plate, compares corresponding data sets, and presents the results of swelling analyses. These characterization results demonstrate that the fuel meets established irradiation performance requirements for mechanical integrity, geometric stability, and stable and predictable behavior.

Scaling up the production capacity of U-Mo powder by HMD process

International Nuclear Information System (INIS)

Pasqualini, E.E.; Lopez, M.; Helzel Garcia, L.J.; Echenique, P.; Adelfang, P.

2002-01-01

The recent discovery that uranium alloys in metastable gamma phase can be hydrided at low temperatures and pressures have allowed developing the method of commuting bulk materials by milling the hydride to desired size and then dehydriding the powder. This process is called HMD (hydriding-milling-dehydriding) and needs an initial step of hydrogen incorporation to allow the alloy to be hydrided. This four step process has been conveniently set up for the production of U-7Mo powder for its use in nuclear fuels. Low equipment investment and low man power are needed for this achievement. The process is being analyzed in its scaling up for one kilogram batches and a 50 kilogram per year production capacity of U-Mo powder. (author)

Irradiation performance of uranium-molybdenum alloy dispersion fuels; Desempenho sob irradiacao de elementos combustiveis do tipo U-Mo

Energy Technology Data Exchange (ETDEWEB)

Almeida, Cirila Tacconi de

2005-07-01

The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm{sup 3} were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm{sup 3} showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

Post irradiation examinations on UMo full-sized plates - IRIS2 experiment

International Nuclear Information System (INIS)

Huet, F.; Noirot, J.; Marelle, V.; Dubois, S.; Boulcourt, P.; Sacristan, P.; Naury, S.; Lemoine, P.

2005-01-01

IRIS2 irradiation was the last irradiation of 4 full sized plates launched by CEA for the French UMo group to test in which operating conditions the coarse porosity forms in the UMo/Al interaction product. IRIS2 consists in four plates with high uranium loading and U-7wt%Mo atomised powder irradiated up to 60 days at OSIRIS reactor in IRIS device at a peak power of 238 W.cm -2 . The results show that in the tested conditions pillowing of the plate started from a fission density over 2.10 21 fission.cm -3 . Moreover, they show that the fission products and impurities have a key-role in the origin of the excessive plate swelling. (author)

Cholesterol-imprinted macroporous monoliths: Preparation and characterization.

Science.gov (United States)

Stepanova, Mariia А; Kinziabulatova, Lilia R; Nikitina, Anna A; Korzhikova-Vlakh, Evgenia G; Tennikova, Tatiana B

2017-11-01

The development of sorbents for selective binding of cholesterol, which is a risk factor for cardiovascular disease, has a great importance for analytical science and medicine. In this work, two series of macroporous cholesterol-imprinted monolithic sorbents differing in the composition of functional monomers (methacrylic acid, butyl methacrylate, 2-hydroxyethyl methacrylate and ethylene dimethacrylate), amount of a template (4, 6 and 8 mol%) used for molecular imprinting, as well as mean pore size were synthesized by in situ free-radical process in stainless steel housing of 50 mm × 4.6 mm i.d. All prepared materials were characterized regarding to their hydrodynamic permeability and porous properties, as well as examined by BET and SEM methods. Imprinting factors, apparent dynamic dissociation constants, the maximum binding capacity, the number of theoretical plates and the height equivalent to a theoretical palate of MIP monoliths at different mobile phase flow rates were determined. The separation of a mixture of structural analogues, namely, cholesterol and prednisolone, was demonstrated. Additionally, the possibility of using the developed monoliths for cholesterol solid-phase extraction from simulated biological solution was shown. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Strength of normal sections of NPP composite monolithic constructions with ribbed reinforced panels

International Nuclear Information System (INIS)

Klyashitskij, V.I.; Kirillov, A.P.

1980-01-01

Strength characteristics and recommendations on designing composite-monolytic structures of NPP with ribbed reinforced panels are considered. Ribbed reinforced panel consists of a system of cross ribs joined with a comparatively thin (25 mm thick) plate. The investigations were carried on using models representing columns symmetrically reinforced with reinforced panels with a low percent of reinforcing. The monolithic structures consisting of ribbed reinforced panels and cast concrete for making monoliths as well as monolithic having analogous strength characteristics of extended and compressed zones have similar strengths. It is shown that calculation of supporting power of composite-monolithic structures is performed according to techniques developed for monolithic structures. Necessity of structural transverse fittings no longer arises in case of corresponding calculational substitution of stability of compressed parts of fittings. Supporting power of a structure decreases not more than by 10% in the presence of cracks in the reinforced panels of the compressed zone. Application of composite-monolithic structures during the construction of the Kursk, Smolensk and Chernobylskaya NPPs permitted to decrease labour content and reduce periods of accomplishment of these works which saves over 6 million roubles

PENGARUH TEMPERATUR DAN IRADIASI TERHADAP INTERDIFUSI PARTIKEL BAHAN BAKAR JENIS U−7Mo/Al

Directory of Open Access Journals (Sweden)

Maman Kartaman Ajiriyanto

2016-06-01

Mg2 plate produced (U,Mo Alx compound on the interfaces. It was evidenced by interdiffusion reaction analysis used DTA that showed that U−7Mo / Al alloy at 500 °C had good heat compactibility, but at temperatures upper than 550 °C it had been phase changed from a + d to a + g phase. The heating in DTA furnace up to 679.14 °C produced U(Al,Mox meta stable phase and then interdiffusion process with uranium molten formed layer interaction that formed UAlx compound agglomerates (UAl4, UAl3 and UAl2. Agglomerates was formed from the heating process which was similar to agglomerates that caused by irradiation. U−7Mo / Al Fuel alloy that had 58% burn up had been interdiffusion between U−7Mo with Al matrix produced U(Al,Mox metastable phase that turned into (U, Mo Al7 layer, UMo2Al20 precipitates, (UMoAl3−Al and formed a boundary or UAlx (UAl4, UAl3 and UAl2 agglomerates.The results of microstructure analysis used SEM and interdiffusion reactions used DTA was supported by the analysis of micro hardness used Vickers Hardness. The results of hardness analysis that was done to AlMg cladding and U−7Mo alloy (before and after heating and diffusion couple of U−7Mo / Al samples with AlMg2 plate after heating at 550 °C were respectively 64.62 and 340.45 HV (before heating and 52.34; 303.16 and 497.34 HV (after heating. Diffusion couple U−7Mo/Al with AlMg2 plate samples had the highest hardness value. This hardness difference showed that the interdiffusion test used diffusion couple produced a new compound (U, Mo Alx in interface zone that had different character, but the formation of interaction layer is not expected in the fuel U−Mo / Al dispersion because micro hardness and density of (U, Mo Alx compound’s layer was lower than the average density of U−7Mo/ Al alloy. Keywords: U−7Mo/Al, diffusion couple, interaction layer, microstructure, DTA and micro hardness.

Characterization of U-Mo Foils for AFIP-7

Energy Technology Data Exchange (ETDEWEB)

Edwards, Danny J.; Ermi, Ruby M.; Schemer-Kohrn, Alan L.; Overman, Nicole R.; Henager, Charles H.; Burkes, Douglas; Senor, David J.

2012-11-07

Twelve AFIP in-process foil samples, fabricated by either Y-12 or LANL, were shipped from LANL to PNNL for potential characterization using optical and scanning electron microscopy techniques. Of these twelve, nine different conditions were examined to one degree or another using both techniques. For this report a complete description of the results are provided for one archive foil from each source of material, and one unirradiated piece of a foil of each source that was irradiated in the Advanced Test Reactor. Additional data from two other LANL conditions are summarized in very brief form in an appendix. The characterization revealed that all four characterized conditions contained a cold worked microstructure to different degrees. The Y-12 foils exhibited a higher degree of cold working compared to the LANL foils, as evidenced by the highly elongated and obscure U-Mo grain structure present in each foil. The longitudinal orientations for both of the Y-12 foils possesses a highly laminar appearance with such a distorted grain structure that it was very difficult to even offer a range of grain sizes. The U-Mo grain structure of the LANL foils, by comparison, consisted of a more easily discernible grain structure with a mix of equiaxed and elongated grains. Both materials have an inhomogenous grain structure in that all of the characterized foils possess abnormally coarse grains.

Microstructural study on gamma phase stability in U-9 wt% Mo alloy system

International Nuclear Information System (INIS)

Saify, M.T.; Jha, S.K.; Hussain, M.M.; Singh, R.P.; Neogy, S.; Srivastava, D.; Dey, G.K.

2009-01-01

Uranium exists in three polymorphic forms viz., orthorhombic α phase - stable up to 667 deg C, tetragonal β phase - stable between 667 deg C and 771 deg C and bcc γ phase - stable above 771 deg C. When alloying of uranium is done, the alloying additions alter the temperature ranges over which the α, β and γ phases are stable. In addition, they frequently retard the rates at which phase transformations occur. As a result, a number of metastable phases can be obtained in uranium alloys. It has been well known among reactor designers that a pure uranium metal is not suitable for power reactor fuel mainly because of (i) phase changes occurring at lower temperatures and (ii) poor irradiation behavior of α phase. γ phase uranium alloys containing small amount of another metal to stabilize the γ-U solid solution provides good prospects in this respect. U-Mo alloy is one of the prospective materials for low enrichment uranium fuel with high U loading because a solid solution of Mo in the γ-U phase possesses acceptable irradiation and mechanical properties and is formed over a wide range of Mo concentration. In the present work vacuum induction melted and cast U-9 wt% Mo alloy was subjected to different thermo mechanical processing to investigate the stability of the γ phase. The as cast alloy was rolled at 550 deg C and then homogenized at 1000 deg C in the γ phase field for 24 hours followed by (i) water quenching and (ii) furnace cooling to generate two different starting conditions. Two of the water-quenched samples were aged at 500 deg C for 5 days and 14 days and one as-rolled sample was aged at 500 deg C for 5 days. The as-cast, as-rolled, homogenized and aged samples were subjected to optical microscopy and X-ray Diffraction (XRD) investigations. All the samples were also subjected to microhardness measurements. The as cast sample contained predominantly the gamma phase along with inclusions. After homogenizing the alloy at 1000 deg C and quenching in

Carprofen-imprinted monolith prepared by reversible addition-fragmentation chain transfer polymerization in room temperature ionic liquids.

Science.gov (United States)

Ban, Lu; Han, Xu; Wang, Xian-Hua; Huang, Yan-Ping; Liu, Zhao-Sheng

2013-10-01

To obtain fast separation, ionic liquids were used as porogens first in combination with reversible addition-fragmentation chain transfer (RAFT) polymerization to prepare a new type of molecularly imprinted polymer (MIP) monolith. The imprinted monolithic column was synthesized using a mixture of carprofen (template), 4-vinylpyridine, ethylene glycol dimethacrylate, [BMIM]BF4, and chain transfer agent (CTA). Some polymerization factors, such as template-monomer molar ratio, the degree of crosslinking, the composition of the porogen, and the content of CTA, on the column efficiency and imprinting effect of the resulting MIP monolith were systematically investigated. Affinity screening of structurally similar compounds with the template can be achieved in 200 s on the MIP monolith due to high column efficiency (up to 12,070 plates/m) and good column permeability. Recognition mechanism of the imprinted monolith was also investigated.

Characterization of fuel miniplates fabricated with U(Mo) particles dispersed in Al-Si matrices

International Nuclear Information System (INIS)

Arico, S F; Mirandou, M I; Balart, S N; Fabro, J O

2012-01-01

In 2011 ECRI facility (Depto. ECRI, GCCN, CNEA) restarted the development for the fabrication of dispersion miniplates fuel elements in Al-Si matrix. This miniplates are fabricated with atomized U-7wt%Mo particles dispersed in a matrix formed by a mixture of pure Al and pure Si powders. The first results for an Al-4wt%Si matrix were presented at the AATN 2011 Annual Meeting. In this work, new results from the microstructural characterization of the meat in Al- 2wt%Si and pure Al miniplates are presented and compared with the previous ones. It is the intention to study the influence of the fabrication parameters as well as different Si concentration in the matrix, on the formation and characteristics of the interaction layer formed between the particles and the matrix at the end of the fabrication process. According to the results presented in this work an improvement can be observed on miniplates with Al-Si matrix respect to the one with pure Al. On the miniplates with Al- Si matrix, almost 100 % of the U(Mo) particles presented, at least in some fraction of its surface, an interaction layer composed by phases that contain Si. Moreover its morphological characteristics are independent of the crystallographic state of the U(Mo) particles. However, the oxide layer formed on the U(Mo) during the hot rolling acts as a barrier to the formation of the interaction layer. As a consequence, it is then mandatory to introduce some changes on the fabrication parameters to avoid, or at least minimize, this oxide layer (author)

Modeling the homogenization kinetics of as-cast U-10wt% Mo alloys

Energy Technology Data Exchange (ETDEWEB)

Xu, Zhijie, E-mail: zhijie.xu@pnnl.gov [Computational Mathematics Group, Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Joshi, Vineet [Energy Processes & Materials Division, Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Hu, Shenyang [Reactor Materials & Mechanical Design, Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Paxton, Dean [Nuclear Engineering and Analysis Group, Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Lavender, Curt [Energy Processes & Materials Division, Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Burkes, Douglas [Nuclear Engineering and Analysis Group, Pacific Northwest National Laboratory, Richland, WA 99352 (United States)

2016-04-01

Low-enriched U-22at% Mo (U–10Mo) alloy has been considered as an alternative material to replace the highly enriched fuels in research reactors. For the U–10Mo to work effectively and replace the existing fuel material, a thorough understanding of the microstructure development from as-cast to the final formed structure is required. The as-cast microstructure typically resembles an inhomogeneous microstructure with regions containing molybdenum-rich and -lean regions, which may affect the processing and possibly the in-reactor performance. This as-cast structure must be homogenized by thermal treatment to produce a uniform Mo distribution. The development of a modeling capability will improve the understanding of the effect of initial microstructures on the Mo homogenization kinetics. In the current work, we investigated the effect of as-cast microstructure on the homogenization kinetics. The kinetics of the homogenization was modeled based on a rigorous algorithm that relates the line scan data of Mo concentration to the gray scale in energy dispersive spectroscopy images, which was used to generate a reconstructed Mo concentration map. The map was then used as realistic microstructure input for physics-based homogenization models, where the entire homogenization kinetics can be simulated and validated against the available experiment data at different homogenization times and temperatures.

In situ polymerization of monolith based on poly(Triallyl Isocyanurate-co-trimethylolpropane triacrylate) and its application in high-performance liquid chromatography.

Science.gov (United States)

Zhong, Jing; Bai, Ligai; Qin, Junxiao; Wang, Jiafei; Hao, Mengbei; Yang, Gengliang

2015-04-01

A novel organic monolithic stationary phase was prepared for high-performance liquid chromatography (HPLC) by in situ copolymerization. In which, triallyl isocyanurate (TAIC) and trimethylolpropane triacrylate (TMPTA) in a binary porogenic solvent consisting of polyethylene glycol 200 and 1, 2-propanediol were used. The resultant monoliths with different column properties (e.g., morphology and pressure) were optimized by adjusting the ratio of TMPTA/TAIC and the composition of porogenic solvent. The resulting poly(TAIC-co-TMPTA) monolith showed a relatively homogeneous structure, good permeability and mechanical stability. The chemical group of the monolith was assayed by the infrared spectra method, the morphology of monolithic material was studied by scanning electron microscopy and the pore size distribution was determined by a mercury porosimeter. A series of small molecules were used to evaluate the column performance in terms of hydrophobic mode. At an optimized flow rate of 1.0 mL min(-1), the theoretical plate number of analyte was >15,000 plates m(-1). These applications demonstrated that the monoliths could be successfully used as the stationary phase in conjunction with HPLC to separate small molecules from the mixture. © The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

Energy Technology Data Exchange (ETDEWEB)

Cummins, Dustin Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hollis, Kendall Jon [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Brown, Donald William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dombrowski, David E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-10-18

This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffraction data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results, optimizations to

Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

International Nuclear Information System (INIS)

Cummins, Dustin Ray; Vogel, Sven C.; Hollis, Kendall Jon; Brown, Donald William; Dombrowski, David E.

2016-01-01

This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffraction data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results, optimizations to

Creep-fatigue evaluation method for weld joint of Mod.9Cr-1Mo steel Part II: Plate bending test and proposal of a simplified evaluation method

Energy Technology Data Exchange (ETDEWEB)

Ando, Masanori, E-mail: ando.masanori@jaea.go.jp; Takaya, Shigeru, E-mail: takaya.shigeru@jaea.go.jp

2016-12-15

Highlights: • Creep-fatigue evaluation method for weld joint of Mod.9Cr-1Mo steel is proposed. • A simplified evaluation method is also proposed for the codification. • Both proposed evaluation method was validated by the plate bending test. • For codification, the local stress and strain behavior was analyzed. - Abstract: In the present study, to develop an evaluation procedure and design rules for Mod.9Cr-1Mo steel weld joints, a method for evaluating the creep-fatigue life of Mod.9Cr-1Mo steel weld joints was proposed based on finite element analysis (FEA) and a series of cyclic plate bending tests of longitudinal and horizontal seamed plates. The strain concentration and redistribution behaviors were evaluated and the failure cycles were estimated using FEA by considering the test conditions and metallurgical discontinuities in the weld joints. Inelastic FEA models consisting of the base metal, heat-affected zone and weld metal were employed to estimate the elastic follow-up behavior caused by the metallurgical discontinuities. The elastic follow-up factors determined by comparing the elastic and inelastic FEA results were determined to be less than 1.5. Based on the estimated elastic follow-up factors obtained via inelastic FEA, a simplified technique using elastic FEA was proposed for evaluating the creep-fatigue life in Mod.9Cr-1Mo steel weld joints. The creep-fatigue life obtained using the plate bending test was compared to those estimated from the results of inelastic FEA and by a simplified evaluation method.

Creep-fatigue evaluation method for weld joint of Mod.9Cr-1Mo steel Part II: Plate bending test and proposal of a simplified evaluation method

International Nuclear Information System (INIS)

Ando, Masanori; Takaya, Shigeru

2016-01-01

Highlights: • Creep-fatigue evaluation method for weld joint of Mod.9Cr-1Mo steel is proposed. • A simplified evaluation method is also proposed for the codification. • Both proposed evaluation method was validated by the plate bending test. • For codification, the local stress and strain behavior was analyzed. - Abstract: In the present study, to develop an evaluation procedure and design rules for Mod.9Cr-1Mo steel weld joints, a method for evaluating the creep-fatigue life of Mod.9Cr-1Mo steel weld joints was proposed based on finite element analysis (FEA) and a series of cyclic plate bending tests of longitudinal and horizontal seamed plates. The strain concentration and redistribution behaviors were evaluated and the failure cycles were estimated using FEA by considering the test conditions and metallurgical discontinuities in the weld joints. Inelastic FEA models consisting of the base metal, heat-affected zone and weld metal were employed to estimate the elastic follow-up behavior caused by the metallurgical discontinuities. The elastic follow-up factors determined by comparing the elastic and inelastic FEA results were determined to be less than 1.5. Based on the estimated elastic follow-up factors obtained via inelastic FEA, a simplified technique using elastic FEA was proposed for evaluating the creep-fatigue life in Mod.9Cr-1Mo steel weld joints. The creep-fatigue life obtained using the plate bending test was compared to those estimated from the results of inelastic FEA and by a simplified evaluation method.

Prediction of U-Mo dispersion nuclear fuels with Al-Si alloy using artificial neural network

International Nuclear Information System (INIS)

Susmikanti, Mike; Sulistyo, Jos

2014-01-01

Dispersion nuclear fuels, consisting of U-Mo particles dispersed in an Al-Si matrix, are being developed as fuel for research reactors. The equilibrium relationship for a mixture component can be expressed in the phase diagram. It is important to analyze whether a mixture component is in equilibrium phase or another phase. The purpose of this research it is needed to built the model of the phase diagram, so the mixture component is in the stable or melting condition. Artificial neural network (ANN) is a modeling tool for processes involving multivariable non-linear relationships. The objective of the present work is to develop code based on artificial neural network models of system equilibrium relationship of U-Mo in Al-Si matrix. This model can be used for prediction of type of resulting mixture, and whether the point is on the equilibrium phase or in another phase region. The equilibrium model data for prediction and modeling generated from experimentally data. The artificial neural network with resilient backpropagation method was chosen to predict the dispersion of nuclear fuels U-Mo in Al-Si matrix. This developed code was built with some function in MATLAB. For simulations using ANN, the Levenberg-Marquardt method was also used for optimization. The artificial neural network is able to predict the equilibrium phase or in the phase region. The develop code based on artificial neural network models was built, for analyze equilibrium relationship of U-Mo in Al-Si matrix

Status and progress of the RERTR program in the year 2004

International Nuclear Information System (INIS)

Travelli, A.

2005-01-01

The overall status of the RERTR program at the time of the last RERTR meeting is reviewed and the progress achieved since that meeting is described. In the fuel area, unexpected failures of LEU U-Mo dispersion plates and tubes under irradiation testing have prompted a revision of the plans to qualify these fuels. While potential solutions to the difficulties with U-Mo dispersion fuels are being explored in collaboration with our international partners, greater emphasis has been placed on accelerating development of monolithic LEU U-Mo fuel. The feasibility of converting several Russian-designed research reactors to LEU fuels has been addressed, and progress has been made in the development of LEU based 99 Mo production processes. The Russian RERTR program has made significant advances. A very important event of 2004 was the US DOE establishment of the Global Threat Reduction Initiative (GTRI). This new program accelerates and combines under the same US DOE management several programs, including RERTR, which aim to secure, remove, or dispose of, nuclear and other radioactive materials throughout the world that are vulnerable to theft by terrorists. (author)

Preparation of organic monolithic columns in polytetrafluoroethylene tubes for reversed-phase liquid chromatography

International Nuclear Information System (INIS)

Catalá-Icardo, M.; Torres-Cartas, S.; Meseguer-Lloret, S.; Gómez-Benito, C.; Carrasco-Correa, E.; Simó-Alfonso, E.F.; Ramis-Ramos, G.; Herrero-Martínez, J.M.

2017-01-01

In this work, a method for the preparation and anchoring of polymeric monoliths in a polytetrafluoroethylene (PTFE) tubing as a column housing for microbore HPLC is described. In order to assure a covalent attachment of the monolith to the inner wall of the PTFE tube, a two-step procedure was developed. Two surface etching reagents, a commercial sodium naphthalene solution (Fluoroetch"®), or mixtures of H_2O_2 and H_2SO_4, were tried and compared. Then, the obtained hydroxyl groups on the PTFE surface were modified by methacryloylation. Attenuated total reflectance Fourier-transform infrared (ATR-FTIR) spectroscopy and scanning electron microscopy (SEM) confirmed the successful modification of the tubing wall and the stable anchorage of monolith to the wall, respectively. Special emphasis was also put on the reduction of the unwanted effects of shrinking of monolith during polymerization, by using an external proper mold and by selecting the adequate monomers in order to increase the flexibility of the polymer. Poly(glycidyl methacrylate-co-divinylbenzene) monoliths were in situ synthesized by thermal polymerization within the confines of surface-vinylized PTFE tubes. The modified PTFE tubing tightly held the monolith, and the monolithic column exhibited good pressure resistance up to 20 MPa. The column performance was also evaluated via the isocratic separation of a series of alkylbenzenes in the reversed-phase mode. The optimized monolithic columns gave plate heights ranged between 70 and 80 μm. The resulting monoliths were also satisfactorily applied to the separation of proteins. - Highlights: • Successful surface etching of PTFE inner wall tubing was done. • The modified PTFE support was next methacryloylated with GMA. • Organic polymeric monolith was in situ prepared in the functionalized PTFE tube. • The monolithic columns gave suitable pressure resistance and separation of proteins.

Preparation of organic monolithic columns in polytetrafluoroethylene tubes for reversed-phase liquid chromatography

Energy Technology Data Exchange (ETDEWEB)

Catalá-Icardo, M., E-mail: mocaic@qim.upv.es [Research Institute for Integrated Management of Coastal Areas, Universitat Politècnica de València, Paranimf 1, 46730, Grao de Gandía, Valencia (Spain); Torres-Cartas, S.; Meseguer-Lloret, S.; Gómez-Benito, C. [Research Institute for Integrated Management of Coastal Areas, Universitat Politècnica de València, Paranimf 1, 46730, Grao de Gandía, Valencia (Spain); Carrasco-Correa, E.; Simó-Alfonso, E.F.; Ramis-Ramos, G. [Department of Analytical Chemistry, Universitat de València, Dr. Moliner 50, 46100, Burjassot, Valencia (Spain); Herrero-Martínez, J.M., E-mail: jmherrer@uv.es [Department of Analytical Chemistry, Universitat de València, Dr. Moliner 50, 46100, Burjassot, Valencia (Spain)

2017-04-01

In this work, a method for the preparation and anchoring of polymeric monoliths in a polytetrafluoroethylene (PTFE) tubing as a column housing for microbore HPLC is described. In order to assure a covalent attachment of the monolith to the inner wall of the PTFE tube, a two-step procedure was developed. Two surface etching reagents, a commercial sodium naphthalene solution (Fluoroetch{sup ®}), or mixtures of H{sub 2}O{sub 2} and H{sub 2}SO{sub 4}, were tried and compared. Then, the obtained hydroxyl groups on the PTFE surface were modified by methacryloylation. Attenuated total reflectance Fourier-transform infrared (ATR-FTIR) spectroscopy and scanning electron microscopy (SEM) confirmed the successful modification of the tubing wall and the stable anchorage of monolith to the wall, respectively. Special emphasis was also put on the reduction of the unwanted effects of shrinking of monolith during polymerization, by using an external proper mold and by selecting the adequate monomers in order to increase the flexibility of the polymer. Poly(glycidyl methacrylate-co-divinylbenzene) monoliths were in situ synthesized by thermal polymerization within the confines of surface-vinylized PTFE tubes. The modified PTFE tubing tightly held the monolith, and the monolithic column exhibited good pressure resistance up to 20 MPa. The column performance was also evaluated via the isocratic separation of a series of alkylbenzenes in the reversed-phase mode. The optimized monolithic columns gave plate heights ranged between 70 and 80 μm. The resulting monoliths were also satisfactorily applied to the separation of proteins. - Highlights: • Successful surface etching of PTFE inner wall tubing was done. • The modified PTFE support was next methacryloylated with GMA. • Organic polymeric monolith was in situ prepared in the functionalized PTFE tube. • The monolithic columns gave suitable pressure resistance and separation of proteins.

Grain boundary engineering to control the discontinuous precipitation in multicomponent U10Mo alloy

Energy Technology Data Exchange (ETDEWEB)

Devaraj, Arun; Kovarik, Libor; Kautz, Elizabeth; Arey, Bruce; Jana, Saumyadeep; Lavender, Curt; Joshi, Vineet

2018-06-01

Grain boundaries in metallic alloys often play a crucial role, not only in determining the mechanical properties or thermal stability of alloys, but also in dictating the phase transformation kinetics during thermomechanical processing. We demonstrate that locally stabilized structure and compositional segregation at grain boundaries—“grain boundary complexions”—in a complex multicomponent alloy can be modified to influence the kinetics of cellular transformation during subsequent thermomechanical processing. Using aberration-corrected scanning transmission electron microscopy and atom probe tomography analysis of a metallic nuclear fuel highly relevant to worldwide nuclear non-proliferation efforts —uranium-10 wt% molybdenum (U-10Mo) alloy, new evidence for the existence of grain boundary complexion is provided. We then modified the concentration of impurities dissolved in Υ-UMo grain interiors and/or segregated to Υ-UMo grain boundaries by changing the homogenization treatment, and these effects were used used to retard the kinetics of cellular transformation during subsequent sub-eutectoid annealing in this U-10-Mo alloy during sub-eutectoid annealing. Thus, this work provided insights on tailoring the final microstructure of the U-10Mo alloy, which can potentially improve the irradiation performance of this important class of alloy fuels.

Preparation and evaluation of poly(alkyl methacrylate-co-methacrylic acid-co-ethylene dimethacrylate) monolithic columns for separating polar small molecules by capillary liquid chromatography

Energy Technology Data Exchange (ETDEWEB)

Lin, Shu-Ling; Wu, Yu-Ru; Lin, Tzuen-Yeuan; Fuh, Ming-Ren, E-mail: msfuh@scu.edu.tw

2015-04-29

Highlights: • Methacrylic acid (MAA) was used to increase hydrophilicity of polymeric monoliths. • Fast separation of phenol derivatives was achieved in 5 min using MAA-incorporated column. • Separations of aflatoxins and phenicol antibiotics were achieved using MAA-incorporated column. - Abstract: In this study, methacrylic acid (MAA) was incorporated with alkyl methacrylates to increase the hydrophilicity of the synthesized ethylene dimethacrylate-based (EDMA-based) monoliths for separating polar small molecules by capillary LC analysis. Different alkyl methacrylate–MAA ratios were investigated to prepare a series of 30% alkyl methacrylate–MAA–EDMA monoliths in fused-silica capillaries (250-μm i.d.). The porosity, permeability, and column efficiency of the synthesized MAA-incorporated monolithic columns were characterized. A mixture of phenol derivatives is employed to evaluate the applicability of using the prepared monolithic columns for separating small molecules. Fast separation of six phenol derivatives was achieved in 5 min with gradient elution using the selected poly(lauryl methacrylate-co-MAA-co-EDMA) monolithic column. In addition, the effect of acetonitrile content in mobile phase on retention factor and plate height as well as the plate height-flow velocity curves were also investigated to further examine the performance of the selected poly(lauryl methacrylate-co-MAA-co-EDMA) monolithic column. Moreover, the applicability of prepared polymer-based monolithic column for potential food safety applications was also demonstrated by analyzing five aflatoxins and three phenicol antibiotics using the selected poly(lauryl methacrylate-co-MAA-co-EDMA) monolithic column.

Towards cleaner methods for the production of Mo-99 using refractory ceramics and its relevance to actinide partitioning and transmutation

Energy Technology Data Exchange (ETDEWEB)

Luca, V.; Dos Santos, L.; Vaccaro, J. [Comision Nacional de Energia Atomica, Centro Atomico Constituyentes, Av. General Paz 1499, 1650 San Martin, Buenos Aires (Argentina)

2016-07-01

Mo-99 is the most utilized isotope in nuclear medicine accounting for over 30 million medical diagnostic procedures annually worldwide. The process for the production of Mo-99 through fission of U-235 normally involves the irradiation of UAl{sub x} dispersion plate fuel in a research reactor, the subsequent dissolution of the fuel plate, the selective separation of the Mo-99 from all of the other fission products and possibly also the recovery of U-235 for future reuse. Compared to the amount of product recovered, copious radioactive waste is generated during the Mo-99 production process. Gaseous wastes are produced at the head-end during the plate dissolution and several liquid wastes are produced during the recovering of Mo-99 using solid extractants, typically polymeric ion exchange resins, which themselves constitute an additional waste stream. It would be extremely advantageous to devise a new process that generates little or no waste. We have been working on a new strategy for the production of fission Mo-99 that involves replacing the dispersion plate targets that are used in the traditional process with inert or active matrix fuel particles that do not need to be dissolved. In one embodiment of the strategy the preparation of new highly porous ZrC{sub x} and graphite-ZrC{sub x} composite target kernels are used that are prepared through polymer templating. The surface properties of these porous materials have been studied and are such that they can be easily loaded with uranium, or for that matter, with any other actinide. In our work we are exploring the possibility of selectively extracting the Mo-99 from the irradiated target kernels by either solution or gas-phase methods and then easily recover the uranium. The fission product-containing kernels can be oxidized in air to generate ZrO{sub 2} that can act as a stable host material either alone or as part of a multiphase ceramic matrix or possibly even as an actinide transmutation host. At the conceptual

A Monolithic CMOS Magnetic Hall Sensor with High Sensitivity and Linearity Characteristics.

Science.gov (United States)

Huang, Haiyun; Wang, Dejun; Xu, Yue

2015-10-27

This paper presents a fully integrated linear Hall sensor by means of 0.8 μm high voltage complementary metal-oxide semiconductor (CMOS) technology. This monolithic Hall sensor chip features a highly sensitive horizontal switched Hall plate and an efficient signal conditioner using dynamic offset cancellation technique. An improved cross-like Hall plate achieves high magnetic sensitivity and low offset. A new spinning current modulator stabilizes the quiescent output voltage and improves the reliability of the signal conditioner. The tested results show that at the 5 V supply voltage, the maximum Hall output voltage of the monolithic Hall sensor microsystem, is up to ±2.1 V and the linearity of Hall output voltage is higher than 99% in the magnetic flux density range from ±5 mT to ±175 mT. The output equivalent residual offset is 0.48 mT and the static power consumption is 20 mW.

A Monolithic CMOS Magnetic Hall Sensor with High Sensitivity and Linearity Characteristics

Directory of Open Access Journals (Sweden)

Haiyun Huang

2015-10-01

Full Text Available This paper presents a fully integrated linear Hall sensor by means of 0.8 μm high voltage complementary metal-oxide semiconductor (CMOS technology. This monolithic Hall sensor chip features a highly sensitive horizontal switched Hall plate and an efficient signal conditioner using dynamic offset cancellation technique. An improved cross-like Hall plate achieves high magnetic sensitivity and low offset. A new spinning current modulator stabilizes the quiescent output voltage and improves the reliability of the signal conditioner. The tested results show that at the 5 V supply voltage, the maximum Hall output voltage of the monolithic Hall sensor microsystem, is up to ±2.1 V and the linearity of Hall output voltage is higher than 99% in the magnetic flux density range from ±5 mT to ±175 mT. The output equivalent residual offset is 0.48 mT and the static power consumption is 20 mW.

A facile route to large-scale synthesis MoO{sub 2} and MoO{sub 3} as electrode materials for high-performance supercapacitors

Energy Technology Data Exchange (ETDEWEB)

Xuan, H.C.; Du, Y.W. [College of Materials Science and Engineering, Taiyuan University of Technology, Taiyuan, 030024 (China); Laboratory of Solid State Microstructures, Nanjing University, Nanjing, 210093 (China); Zhang, Y.Q.; Xu, Y.K.; Li, H.; Han, P.D. [College of Materials Science and Engineering, Taiyuan University of Technology, Taiyuan, 030024 (China); Wang, D.H. [Laboratory of Solid State Microstructures, Nanjing University, Nanjing, 210093 (China)

2016-09-15

MoO{sub 3} and MoO{sub 2} materials have been successfully synthesized by thermal decomposition of ammonium paramolybdate in air and a sealed quartz tube, respectively. The microstructure of as-synthesized MoO{sub 3} is composed of irregular lamellar plates with a plate thickness around 100 nm and MoO{sub 2} has the larger grain size with lamellar plates connected with each other. A maximum specific capacitance of 318 F/g at 0.5 A/g is obtained for MoO{sub 2} prepared in a closed environment. On the other hand, the sample MoO{sub 3} exhibits excellent rate capacity with specific capacitances of 218, 209, 196, 188, 176, and 160 F/g at current densities of 0.5, 1, 2, 3, 4, and 5 A/g, respectively. These results pave the way to consider MoO{sub 3} and MoO{sub 2} as prospective materials for energy-storage applications. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Structure and properties of GMA surfaced armour plates

OpenAIRE

A. Klimpel; K. Luksa; M. Burda

2010-01-01

Purpose: In the combat vehicles many materials can be used for the armour. Application of the monolithic armour plates in light combat vehicles is limited by the high armour weigh. Introduction of the layered armour plates is a way to limit the vehicle weight. In the paper test results of graded and nanostructural GMA surfaced armour plates are presented.Design/methodology/approach: Metallographic structure, chemical composition and hardness of surfaced layers were investigated in order to ex...

Full-sized plates irradiation with high UMo fuel loading. Final results of IRIS 1 experiment

International Nuclear Information System (INIS)

Huet, F.; Marelle, V.; Noirot, J.; Sacristan, P.; Lemoine, P.

2003-01-01

As a part of the French UMo Group qualification program, IRIS 1 experiment contained full-sized plates with high uranium loading in the meat of 8 g.cm -3 . The fuel particles consisted of 7 and 9 wt% Mo-uranium alloys ground powders. The plate were irradiated at OSIRIS reactor in IRIS device up to 67.5% peak burnup within the range of 136 W.cm - '2 for the heat flux and 72 deg. C for the cladding temperature. After each reactor cycle the plates thickness were measured. The results show no swelling behaviour differences versus burnup between UMo7 and UMo9 plates. The maximum plate swelling for peak burnup location remains lower than 6%. The wide set of PIE has shown that, within the studied irradiation conditions, the interaction product have a global formulation of '(U-Mo)Al -7 ' and that there is no aluminium dissolution in UMo particles. IRIS1 experiment, as the first step of the UMo fuel qualification for research reactor, has established the good behaviour of UMo7 and UMo9 high uranium loading full-sized plate within the tested conditions. (author)

Fabrication of high-uranium-loaded U{sub 3}O{sub 8}-Al developmental fuel plates

Energy Technology Data Exchange (ETDEWEB)

Copeland, G L; Martin, M M [Oak Ridge National Laboratory, TN (United States)

1983-08-01

A common plate-type fuel for Research and Test Reactors (RERTR) is U{sub 3}0{sub 8} dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the {sup 235}U enrichment from above 90% to below 20% for these fuels to lessen the risk of diversion of the uranium for non-peaceful uses. Thus, the uranium content of the fuel plates has to be increased to maintain the performance of the reactors. This paper describes work at ORNL to determine the maximal uranium loading for these fuels that can be fabricated with commercially proven materials and techniques and that can be expected to perform satisfactorily in service. We fabricated developmental fuel plates with cores containing from 60 to 100 wt U{sub 3}0{sub 8} in aluminum encapsulated in 6061 aluminum alloy and evaluated them for aspects of fabricability, nondestructive testing, and expected performance. We recommend 75 wt U{sub 3}0{sub 8}-Al 3.1 Mg U/m{sup 3}) as the highest loading in the initial irradiation test. This upper limit is based on a qualitative assessment of the mechanical integrity of the core made by using current fabrication techniques and materials. As the oxide loading is increased beyond this point, planar areas and extensive stringers of oxide and voids develop, which leave little strength in the thickness direction. Fuel plates may then blister over these areas as fission gases collect during irradiation. Current size plates are easily fabricable to the 75 wt % U{sub 3}0{sub 8}-Al core loading by current fabrication techniques. Dogboning is a potential problem at this loading for some applications; however, this can be easily solved by using tapered compact ends. Current nondestructive radiography and transmission x-ray scanning are applicable to the highly loaded plates. Ultrasonic testing for non-bonds is marginal because of the abrupt change in conductance at the cladding-core interface. Plate thickness can be increased if desired; we fabricated 75 wt % plates with

Effects of Particle Size and Shape on U-Mo/Al Thermal Conductivity

Energy Technology Data Exchange (ETDEWEB)

Cho, Tae-Won; Sohn, Dong-Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-10-15

The thermal conductivity of atomized U-Mo/Al dispersion fuels was measured only by Lee et al. by laser-flash and differential scanning calorimetry (DSC) methods. For the U-Mo particles, they are deformed during manufacturing process such as hot rolling and during irradiation by the creep deformation. Fricke developed a model for the effective thermal conductivity of a dilute suspension of randomly oriented spheroidal particles. In general, the thermal conductivity of composite increase when the particle shape is not sphere. This model is also based on continuum theory which assumes both temperature and heat flux are continuous across the interface. Kapitza, however, showed that there is a discontinuity in temperature across the interface at metal/liquid helium interface. In general, the discontinuity is from the thermal resistance at the interface. If the thermal resistance has a significant impact on the thermal conductivity, particle size is one of the essential parameter for determining the effective thermal conductivity of composite materials. Every, et al modified Bruggeman model to consider the interfacial thermal resistance. The U-Mo/Al dispersion fuel thermal conductivity calculation can be improved by considering the anisotropic effects and interface thermal resistances. There have been various works to analyze the thermal conductivity through Finite Element Method (FEM). Coulson developed a realistic FEM model to calculate the effective thermal conductivity of the fuel meat. This FEM model does not consider the anisotropic effects and interface thermal resistances. Therefore, these effects can be evaluated by comparing the FEM calculated effective thermal conductivity with measured data. In this work, the FEM analysis was done and the anisotropic effects and interface thermal resistances was estimated. From this results, the particle shape and size effects will be discussed. Many thermal conductivity models for the particle dispersed composites have been

Effects of Particle Size and Shape on U-Mo/Al Thermal Conductivity

International Nuclear Information System (INIS)

Cho, Tae-Won; Sohn, Dong-Seong

2014-01-01

The thermal conductivity of atomized U-Mo/Al dispersion fuels was measured only by Lee et al. by laser-flash and differential scanning calorimetry (DSC) methods. For the U-Mo particles, they are deformed during manufacturing process such as hot rolling and during irradiation by the creep deformation. Fricke developed a model for the effective thermal conductivity of a dilute suspension of randomly oriented spheroidal particles. In general, the thermal conductivity of composite increase when the particle shape is not sphere. This model is also based on continuum theory which assumes both temperature and heat flux are continuous across the interface. Kapitza, however, showed that there is a discontinuity in temperature across the interface at metal/liquid helium interface. In general, the discontinuity is from the thermal resistance at the interface. If the thermal resistance has a significant impact on the thermal conductivity, particle size is one of the essential parameter for determining the effective thermal conductivity of composite materials. Every, et al modified Bruggeman model to consider the interfacial thermal resistance. The U-Mo/Al dispersion fuel thermal conductivity calculation can be improved by considering the anisotropic effects and interface thermal resistances. There have been various works to analyze the thermal conductivity through Finite Element Method (FEM). Coulson developed a realistic FEM model to calculate the effective thermal conductivity of the fuel meat. This FEM model does not consider the anisotropic effects and interface thermal resistances. Therefore, these effects can be evaluated by comparing the FEM calculated effective thermal conductivity with measured data. In this work, the FEM analysis was done and the anisotropic effects and interface thermal resistances was estimated. From this results, the particle shape and size effects will be discussed. Many thermal conductivity models for the particle dispersed composites have been

Monolithic fuel injector and related manufacturing method

Science.gov (United States)

Ziminsky, Willy Steve [Greenville, SC; Johnson, Thomas Edward [Greenville, SC; Lacy, Benjamin [Greenville, SC; York, William David [Greenville, SC; Stevenson, Christian Xavier [Greenville, SC

2012-05-22

A monolithic fuel injection head for a fuel nozzle includes a substantially hollow vesicle body formed with an upstream end face, a downstream end face and a peripheral wall extending therebetween, an internal baffle plate extending radially outwardly from a downstream end of the bore, terminating short of the peripheral wall, thereby defining upstream and downstream fuel plenums in the vesicle body, in fluid communication by way of a radial gap between the baffle plate and the peripheral wall. A plurality of integral pre-mix tubes extend axially through the upstream and downstream fuel plenums in the vesicle body and through the baffle plate, with at least one fuel injection hole extending between each of the pre-mix tubes and the upstream fuel plenum, thereby enabling fuel in the upstream plenum to be injected into the plurality of pre-mix tubes. The fuel injection head is formed by direct metal laser sintering.

Interdiffusion between U-Mo alloys and Al

International Nuclear Information System (INIS)

Mirandou, M.I.; Balart, S.N.; Ortiz, M.; Granovsky, M.S.; Hofman, G.L.

2002-01-01

During the fabrication and/or irradiation of the dispersion fuel elements, the fuel particles react with the surrounding Al matrix. This reaction results in the formation of a zone consisting of intermetallic compounds. The low thermal conductivity of these compounds has a major effect on the fuel temperature as well as on the swelling of the fuel. Interdiffusion between U-Mo/Al is being investigated using chemical diffusion couples. In this paper the first results obtained with optical and scanning electron microscope, electron microprobe and X-Ray diffraction are presented. Investigation of the effect on the formation of the interdiffusion zone of small additions of Mg to Al is the primary purpose of this study. (author)

Comparing monolithic and fused core HPLC columns for fast chromatographic analysis of fat-soluble vitamins.

Science.gov (United States)

Kurdi, Said El; Muaileq, Dina Abu; Alhazmi, Hassan A; Bratty, Mohammed Al; Deeb, Sami El

2017-06-27

HPLC stationary phases of monolithic and fused core type can be used to achieve fast chromatographic separation as an alternative to UPLC. In this study, monolithic and fused core stationary phases are compared for fast separation of four fat-soluble vitamins. Three new methods on the first and second generation monolithic silica RP-18e columns and a fused core pentafluoro-phenyl propyl column were developed. Application of three fused core columns offered comparable separations of retinyl palmitate, DL-α-tocopheryl acetate, cholecalciferol and menadione in terms of elution speed and separation efficiency. Separation was achieved in approx. 5 min with good resolution (Rs > 5) and precision (RSD ≤ 0.6 %). Monolithic columns showed, however, a higher number of theoretical plates, better precision and lower column backpressure than the fused core column. The three developed methods were successfully applied to separate and quantitate fat-soluble vitamins in commercial products.

Using fractional extraction method to separate Mo from U in high concentration solution

International Nuclear Information System (INIS)

Zhao Pinzhi; Cheng Guangrong; Ma Xiuhua

1996-01-01

The author presents investigation on separating Mo from U in acid high concentration lixivium with fractional extraction of secondary amine (7203) and D2EHPA and preparing qualified products of ammonium molybdate and sodium diuranate

Monoliths in Bioprocess Technology

Directory of Open Access Journals (Sweden)

Vignesh Rajamanickam

2015-04-01

Full Text Available Monolithic columns are a special type of chromatography column, which can be used for the purification of different biomolecules. They have become popular due to their high mass transfer properties and short purification times. Several articles have already discussed monolith manufacturing, as well as monolith characteristics. In contrast, this review focuses on the applied aspect of monoliths and discusses the most relevant biomolecules that can be successfully purified by them. We describe success stories for viruses, nucleic acids and proteins and compare them to conventional purification methods. Furthermore, the advantages of monolithic columns over particle-based resins, as well as the limitations of monoliths are discussed. With a compilation of commercially available monolithic columns, this review aims at serving as a ‘yellow pages’ for bioprocess engineers who face the challenge of purifying a certain biomolecule using monoliths.

Turbomachine combustor nozzle including a monolithic nozzle component and method of forming the same

Science.gov (United States)

Stoia, Lucas John; Melton, Patrick Benedict; Johnson, Thomas Edward; Stevenson, Christian Xavier; Vanselow, John Drake; Westmoreland, James Harold

2016-02-23

A turbomachine combustor nozzle includes a monolithic nozzle component having a plate element and a plurality of nozzle elements. Each of the plurality of nozzle elements includes a first end extending from the plate element to a second end. The plate element and plurality of nozzle elements are formed as a unitary component. A plate member is joined with the nozzle component. The plate member includes an outer edge that defines first and second surfaces and a plurality of openings extending between the first and second surfaces. The plurality of openings are configured and disposed to register with and receive the second end of corresponding ones of the plurality of nozzle elements.

Highly crosslinked polymeric monoliths for reversed-phase capillary liquid chromatography of small molecules.

Science.gov (United States)

Liu, Kun; Tolley, H Dennis; Lee, Milton L

2012-03-02

Seven crosslinking monomers, i.e., 1,3-butanediol dimethacrylate (1,3-BDDMA), 1,4-butanediol dimethacrylate (1,4-BDDMA), neopentyl glycol dimethacrylate (NPGDMA), 1,5-pentanediol dimethacrylate (1,5-PDDMA), 1,6-hexanediol dimethacrylate (1,6-HDDMA), 1,10-decanediol dimethacrylate (1,10-DDDMA), and 1,12-dodecanediol dimethacrylate (1,12-DoDDMA), were used to synthesize highly cross-linked monolithic capillary columns for reversed-phase liquid chromatography (RPLC) of small molecules. Dodecanol and methanol were chosen as "good" and "poor" porogenic solvents, respectively, for these monoliths, and were investigated in detail to provide insight into the selection of porogen concentration using 1,12-DoDDMA. Isocratic elution of alkylbenzenes at a flow rate of 300 nL/min was conducted for all of the monoliths. Gradient elution of alkylbenzenes and alkylparabens provided high resolution separations. Optimized monoliths synthesized from all seven crosslinking monomers showed high permeability. Several of the monoliths demonstrated column efficiencies in excess of 50,000 plates/m. Monoliths with longer alkyl-bridging chains showed very little shrinking or swelling in solvents of different polarities. Column preparation was highly reproducible; the relative standard deviation (RSD) values (n=3) for run-to-run and column-to-column were less than 0.25% and 1.20%, respectively, based on retention times of alkylbenzenes. Copyright © 2012 Elsevier B.V. All rights reserved.

Process variables in the obtention of U-Mo powder by the hydriding-milling-dehydriding method (HMD process)

International Nuclear Information System (INIS)

Pasqualini, Enrique E.; Helzel Garcia, Javier; Lopez, Marisol

2003-01-01

In the next few years nuclear fuels based on uranium oxides, aluminides and silicides for MTR reactors will be replaced by the high density alloy uranium- 7% (w/w) molybdenum (U-7 Mo). Actually there is only one commercial supplier of this raw material that has to be provided as powder containing 20% enriched uranium ( 235 U). In the Nuclear Fuels Department of the National Atomic Energy Commission (CNEA) at Buenos Aires was developed an alternative way of producing U-7 Mo powder in a production scale. Meanwhile CNEA is participating in the International Program (RERTR) for final qualification of this nuclear material. This new method of production consists in the hydriding of the alloy, milling the hydride to final size and dehydriding the powder. These results were achieved because a special technique was discovered for the massive hydriding of the U-7 Mo alloy. The production method is simple, requires conventional equipment and low investment. Argentine can have important comparative advantages for its production and exportation. A scale production plant is being planed. (author)

Study on HANARO core conversion using U-Mo fuel

International Nuclear Information System (INIS)

Lee, K.H.; Lee, C.S.; Seo, C.G.; Park, S.J.; Kim, H.; Kim, C.K.

2002-01-01

Two types of fuel rods with different fuel meat diameter and uranium density are considered for HANARO core conversion with high density U-Mo fuel. Arranging standard fuels of 5.0 g U/cc and 6.35 mm in diameter at the inner ring of an assembly and reduced fuels of 4.3 g U/cc and 5.49 mm in diameter at the outer ring of an assembly flattens the assembly power distribution and avoids the increase of linear heat generation rate due to using higher uranium density and less number of fuel rods. The maximum linear heat generation rate is similar with the current reference core and four fuel sites at the outer core in the reflector tank is converted to the irradiation sites to suit more demand on fuel tests and radioisotope production at outer core sites. This new core has 32% longer fuel cycle than the current reference core. (author)

Comparing monolithic and fused core HPLC columns for fast chromatographic analysis of fat-soluble vitamins

Directory of Open Access Journals (Sweden)

Kurdi Said El

2017-06-01

Full Text Available HPLC stationary phases of monolithic and fused core type can be used to achieve fast chromatographic separation as an alternative to UPLC. In this study, monolithic and fused core stationary phases are compared for fast separation of four fat-soluble vitamins. Three new methods on the first and second generation monolithic silica RP-18e columns and a fused core pentafluoro-phenyl propyl column were developed. Application of three fused core columns offered comparable separations of retinyl palmitate, DL-α-tocopheryl acetate, cholecalciferol and menadione in terms of elution speed and separation efficiency. Separation was achieved in approx. 5 min with good resolution (Rs > 5 and precision (RSD ≤ 0.6 %. Monolithic columns showed, however, a higher number of theoretical plates, better precision and lower column backpressure than the fused core column. The three developed methods were successfully applied to separate and quantitate fat-soluble vitamins in commercial products.

Modeling of interaction layer growth between U-Mo particles and an Al matrix

International Nuclear Information System (INIS)

Kim, Yeon Soo; Horman, G. L.; Ryu, Ho Jin; Park, Jong Man; Robinson, A. B.; Wachs, D. M.

2013-01-01

Interaction layer growth between U-Mo alloy fuel particles and Al in a dispersion fuel is a concern due to the volume expansion and other unfavorable irradiation behavior of the interaction product. To reduce interaction layer (IL) growth, a small amount of Si is added to the Al. As a result, IL growth is affected by the Si content in the Al matrix. In order to predict IL growth during fabrication and irradiation, empirical models were developed. For IL growth prediction during fabrication and any follow-on heating process before irradiation, out-of-pile heating test data were used to develop kinetic correlations. Two out-of-pile correlations, one for the pure Al matrix and the other for the Al matrix with Si addition, respectively, were developed, which are Arrhenius equations that include temperature and time. For IL growth predictions during irradiation, the out-of-pile correlations were modified to include a fission-rate term to consider fission enhanced diffusion, and multiplication factors to incorporate the Si addition effect and the effect of the Mo content. The in-pile correlation is applicable for a pure Al matrix and an Al matrix with the Si content up to 8 wt%, for fuel temperatures up to 200 .deg. C, and for Mo content in the range of 6 - 10wt%. In order to cover these ranges, in-pile data were included in modeling from various tests, such as the US RERTR-4, -5, -6, -7 and -9 tests and Korea's KOMO-4 test, that were designed to systematically examine the effects of the fission rate, temperature, Si content in Al matrix, and Mo content in U-Mo particles. A model converting the IL thickness to the IL volume fraction in the meat was also developed

MODELING OF INTERACTION LAYER GROWTH BETWEEN U-Mo PARTICLES AND AN Al MATRIX

Directory of Open Access Journals (Sweden)

YEON SOO KIM

2013-12-01

Full Text Available Interaction layer growth between U-Mo alloy fuel particles and Al in a dispersion fuel is a concern due to the volume expansion and other unfavorable irradiation behavior of the interaction product. To reduce interaction layer (IL growth, a small amount of Si is added to the Al. As a result, IL growth is affected by the Si content in the Al matrix. In order to predict IL growth during fabrication and irradiation, empirical models were developed. For IL growth prediction during fabrication and any follow-on heating process before irradiation, out-of-pile heating test data were used to develop kinetic correlations. Two out-of-pile correlations, one for the pure Al matrix and the other for the Al matrix with Si addition, respectively, were developed, which are Arrhenius equations that include temperature and time. For IL growth predictions during irradiation, the out-of-pile correlations were modified to include a fission-rate term to consider fission enhanced diffusion, and multiplication factors to incorporate the Si addition effect and the effect of the Mo content. The in-pile correlation is applicable for a pure Al matrix and an Al matrix with the Si content up to 8 wt%, for fuel temperatures up to 200 °C, and for Mo content in the range of 6 – 10wt%. In order to cover these ranges, in-pile data were included in modeling from various tests, such as the US RERTR-4, -5, -6, -7 and -9 tests and Korea's KOMO-4 test, that were designed to systematically examine the effects of the fission rate, temperature, Si content in Al matrix, and Mo content in U-Mo particles. A model converting the IL thickness to the IL volume fraction in the meat was also developed.

Fission gas behaviour and interdiffusion layer growth in in-pile and out-of-pile irradiated U-Mo/Al nuclear fuels

International Nuclear Information System (INIS)

Zweifel, Tobias

2014-01-01

Worldwide, research and test reactors are to convert their fuel from highly towards lower enriched uranium, among them the FRM II. One prospective fuel is an alloy of uranium and molybdenum (abbr. U-Mo). Test irradiations showed an insufficient irradiation behavior of this new fuel due to the growth of an interdiffusion layer (abbr. IDL) between the U-Mo fuel and the surrounding Al matrix. Furthermore, this layer accumulates fission gases. In this work, heavy ion irradiated U-Mo/Al layer systems were studied and compared to in-reactor irradiated fuel to study the fission gas dynamics. It is demonstrated that the gas behavior is identical for both in-reactor and out-of-reactor approaches.

Enhanced photoresponse of monolayer molybdenum disulfide (MoS2) based on microcavity structure

Science.gov (United States)

Lu, Yanan; Yang, Guofeng; Wang, Fuxue; Lu, Naiyan

2018-05-01

There is an increasing interest in using monolayer molybdenum disulfide (MoS2) for optoelectronic devices because of its inherent direct band gap characteristics. However, the weak absorption of monolayer MoS2 restricts its applications, novel concepts need to be developed to address the weakness. In this work, monolayer MoS2 monolithically integrates with plane microcavity structure, which is formed by the top and bottom chirped distributed Bragg reflector (DBR), is demonstrated to improve the absorption of MoS2. The optical absorption is 17-fold enhanced, reaching values over 70% at work wavelength. Moreover, the monolayer MoS2-based photodetector device with microcavity presents a significantly increased photoresponse, demonstrating its promising prospects in MoS2-based optoelectronic devices.

Studies on the influence of processing methods on the corrosion characteristics of electroslag-tape-plating with NiMo16Cr16Ti. Untersuchungen zum Einfluss der Prozessfuehrung auf die Korrosionseigenschaften beim Elektroschlacke-Bandplattieren mit NiMo16Cr16Ti

Energy Technology Data Exchange (ETDEWEB)

Blum, J

1991-01-01

The aim of the project was to demonstrate an economic possibility of using NiMo16Cr16Ti reliably as a plating material with simple to use methods. For this, the electroslag and submerged arc welding with tape electrodes as a coating technology are compared with each other. As the research showed, it is possible to coat large surface area, thick walled components economically with NiMo16Cr16Ti using electroslag (RES) tape plating. The choice of powder is important for the hot crack freedom and the precipitation poverty and thus the corrosion resistance of the plating. The silicon content was proved to be of importance in the plating. With the right choice of powder, the second layer already possesses the endurance of laminated materials against intercrystalline corrosion in coating tests as well as against pitting in a 10% FeCl{sub 3} solution. The coating capability lies at 0.5 m{sup 2}/h. - In addition the use of video thermography for on-line surveillance of seam on RES plates is documented. The evaluation of the pictures makes possible the targeting of the external magnets and thus the influencing of the melting bath flux. (orig./RHM) With 80 figs., 12 tabs.

Investigation of the Cause of Low Blister Threshold Temperatures in the RERTR-12 and AFIP-4 Experiments

Energy Technology Data Exchange (ETDEWEB)

Mitchell K Meyer

2012-06-01

Blister–threshold testing of fuel plates is a standard method through which the safety margin for operation of plate-type in research and test reactors is assessed. The blister-threshold temperature is indicative of the ability of fuel to operate at high temperatures for short periods of time (transient conditions) without failure. This method of testing was applied to the newly developed U-Mo monolithic fuel system. Blister annealing studies on the U-Mo monolithic fuel plates began in 2007, with the Reduced Enrichment for Research and Test Reactors (RERTR)-6 experiment, and they have continued as the U-Mo fuel system has evolved through the research and development process. Blister anneal threshold temperatures from early irradiation experiments (RERTR-6 through RERTR-10) ranged from 400 to 500°C. These temperatures were projected to be acceptable for NRC-licensed research reactors and the high-power Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) based on current safety-analysis reports (SARs). Initial blister testing results from the RERTR-12 experiment capsules X1 and X2 showed a decrease in the blister-threshold temperatures. Blister threshold temperatures from this experiment ranged from 300 to 400°C. Selected plates from the AFIP-4 experiment, which was fabricated using a process similar to that used to fabricate the RERTR-12 experiment, also underwent blister testing to determine whether results would be similar. The measured blister-threshold temperatures from the AFIP-4 plates fell within the same blister-threshold temperature range measured in the RERTR-12 plates. Investigation of the cause of this decrease in bister threshold temperature is being conducted under the guidance of Idaho National Laboratory PLN-4155, “Analysis of Low Blister Threshold Temperatures in the RERTR-12 and AFIP-4 Experiments,” and is driven by hypotheses. The main focus of the investigation is in the following areas: 1. Fabrication variables 2. Pre

A Very High Uranium Density Fission Mo Target Suitable for LEU Using atomization Technology

Energy Technology Data Exchange (ETDEWEB)

Kim, C. K.; Kim, K. H.; Lee, Y. S.; Ryu, H. J.; Woo, Y. M.; Jang, S. J.; Park, J. M.; Choi, S. J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

Currently HEU minimization efforts in fission Mo production are underway in connection with the global threat reduction policy. In order to convert HEU to LEU for the fission Mo target, higher uranium density material could be applied. The uranium aluminide targets used world widely for commercial {sup 99}Mo production are limited to 3.0 g-U/cc in uranium density of the target meat. A consideration of high uranium density using the uranium metal particles dispersion plate target is taken into account. The irradiation burnup of the fission Mo target are as low as 8 at.% and the irradiation period is shorter than 7 days. Pure uranium material has higher thermal conductivity than uranium compounds or alloys. It is considered that the degradation by irradiation would be almost negligible. In this study, using the computer code of the PLATE developed by ANL the irradiation behavior was estimated. Some considerations were taken into account to improve the irradiation performance further. It has been known that some alloying elements of Si, Cr, Fe, and Mo are beneficial for reducing the swelling by grain refinement. In the RERTR program recently the interaction problem could be solved by adding a small amount of Si to the aluminum matrix phase. The fabrication process and the separation process for the proposed atomized uranium particles dispersion target were reviewed

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

Energy Technology Data Exchange (ETDEWEB)

Drera, Saleem S., E-mail: saleem.drera@gmail.com [Mechanical Engineering, Colorado School of Mines, Golden, CO 80401 (United States); Hofman, Gerard L. [Argonne National Laboratory, Chicago, IL 60439 (United States); Kee, Robert J. [Mechanical Engineering, Colorado School of Mines, Golden, CO 80401 (United States); King, Jeffrey C. [Metallurgical and Materials Engineering, Colorado School of Mines, Golden, CO 80401 (United States)

2014-10-15

Highlights: • This article presents a cellular automata (CA) algorithm to synthesize the growth of intermetallic interaction layers in U–Mo/Al dispersion fuel. • The method utilizes a 3D representation of the fuel, which is discretized into separate voxels that can change identy based on derived CA rules. • The CA model is compared to ILT measurements for RERTR experimental data. • The primary objective of the model is to synthesize three-dimensional microstructures that can be used in subsequent thermal and mechanical modeling. • The CA model can be used for predictive analysis. For example, it can be used to study the dependence of temperature on interaction layer growth. - Abstract: Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium–molybdenum (U–Mo) particles within an aluminum matrix. Fresh U–Mo particles typically range between 10 and 100 μm in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction–diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates.

Ground state structure of U2Mo: static and lattice dynamics study

International Nuclear Information System (INIS)

Mukherjee, D.; Sahoo, B.D.; Joshi, K.D.; Kaushik, T.C.

2016-01-01

According to experimental reports, the ground state stable structure of U 2 Mo is tetragonal. However, various theoretical studies performed in past do not get tetragonal phase as the stable structure at ambient conditions. Therefore, the ground state structure of U 2 Mo is still unresolved. In an attempt to understand the ground state properties of this system, we have carried out first principle electronic band structure calculations. The structural stability analysis carried out using evolutionary structure search algorithm in conjunction with ab-inito method shows that a hexagonal structure (space group P6/mmm) is the lowest enthalpy structure at ambient condition and remains stable upto 200 GPa. The elastic and lattice dynamical stability further supports the stability of this phase at ambient condition. Further, using the 0 K calculations in conjunction with finite temperature corrections, we have derived the isotherm and shock adiabat (Hugoniot) of this material. Various equilibrium properties such as ambient pressure volume, bulk modulus, pressure derivative of bulk modulus etc. are derived from equation of state. (author)

Thermal Characteristic Of AIMg2 Cladding And Fuel Plates Of U3Si2-Al With Various Uranium Loading

International Nuclear Information System (INIS)

Aslina, Br. G.; Suparjo; Aggraini, D.; Hasbullah, N.

1998-01-01

Thermal characteristic analyzed in this paper included linear expansion value, coefficient expansion, and enthalpy of cladding material fuel core and fuel plate of U 3 Si 2 -AI. Before analyzing, the fresh cladding of AIMg2 (without treatment) and the rolled AIMg2 were annealed at temperature of 425 o C for 1 hour, and the fuel plates of U 3 Si 2 -AI was prepared for various uranium loading of 0.9 - 3.6 - 4.2 - 4.8 and 5.2 g/cm 3 . Linear expansion nominal value and expansion coefficient were analyzed by using Dilatometer whereas enthalpy determination used Differential Thermal Analysis (DTA). The linear expansion and expansion coefficient analysis was performed to study the dimension cladding and of fuel plates during their stay in the reactor core, whereas determination of enthalpy was carried out to estimate the energy absorbed and released by fuel meat of U 3 Si 2 -AI to the cooling water through AlMg2 as a cladding. The result showed that the linear expansion and expansion coefficient of fresh AIMg2 cladding, rolled AIMg2 and fuel plates of U 3 Si 2 -AI are increased with the increase of temperature as well as the increase of uranium loading. The enthalpy measure showed that the enthalpy of fresh AIMg2 is smaller than that of rolled AIMg2 but melting temperature of fresh AIMg2 is greater than that of rolled AIMg2. The enthalpy of fuel plates and meat of U 3 Si 2 -AI is less than that of plates of U 3 Si 2 -AI. The enthalpy of fuel platers and meat of U 3 Si 2 -AI decrease with the increase of uranium loading. It is concluded that the fuel meat more reactive than fuel plates of U 3 Si 2 -AI

Powder production of U-Mo alloy, HMD process (Hydriding- Milling- Dehydriding)

Energy Technology Data Exchange (ETDEWEB)

Pasqualini, E. E.; Garcia, J.H.; Lopez, M.; Cabanillas, E.; Adelfang, P. [Dept. Combustibles Nucleares. Comision Nacional de Energia Atomica, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina)

2002-07-01

Uranium-molybdenum (U-Mo) alloys can be hydrided massively in metastable {gamma} (gamma) phase. The brittle hydride can be milled and dehydrided to acquire the desired size distributions needed for dispersion nuclear fuels. The developments of the different steps of this process called hydriding-milling- dehydriding (HMD Process) are described. Powder production scales for industrial fabrication is easily achieved with conventional equipment, small man-power and low investment. (author)

Powder production of U-Mo alloy, HMD process (Hydriding- Milling- Dehydriding)

International Nuclear Information System (INIS)

Pasqualini, E. E.; Garcia, J.H.; Lopez, M.; Cabanillas, E.; Adelfang, P.

2002-01-01

Uranium-molybdenum (U-Mo) alloys can be hydrided massively in metastable γ (gamma) phase. The brittle hydride can be milled and dehydrided to acquire the desired size distributions needed for dispersion nuclear fuels. The developments of the different steps of this process called hydriding-milling- dehydriding (HMD Process) are described. Powder production scales for industrial fabrication is easily achieved with conventional equipment, small man-power and low investment. (author)

The Experiment Production And Examination Of The U3Si2-AI Mini plates For Irradiation Test

International Nuclear Information System (INIS)

Supardjo; Boybul; Yowono; Susworo; Permana, S.

1998-01-01

The fuel plates containing U 3 Si 2 -AI dispersion fuel having respective loading of 3.55; 4.20; and 4.80 g/cm 3 were prepared by dispersing certain amount of U 3 Si 2 powder in the AI powder as matrix. The weight ratio of U 3 Si 2 and AI at different loading was chosen based on the 19.23 cm 3 volume basis fuel core calculation. Each fuel mixture was pressed into a fuel core having dimension of 100.20 x 60.35 x 3.15 +- (0.05) mm, which was then cut into mini fuel core having dimension of 16 x 8 x 3.15 +- (0.05) mm. The mini plates were prepared by picture and frame technique using AIMg2 as cladding material. The mini plates have been tested for blister, homogeneity, white spots, surface defects and their cladding thickness, revealing that out of 74 mini plates, they are ten (10) mini plates that have to be rejected due to blisters and white spots, thus of 64 mini plates can be further fabricated as samples for irradiation test

Preparation and evaluation of 400μm I.D. polymer-based hydrophilic interaction chromatography monolithic columns with high column efficiency.

Science.gov (United States)

Liu, Chusheng; Li, Haibin; Wang, Qiqin; Crommen, Jacques; Zhou, Haibo; Jiang, Zhengjin

2017-08-04

The quest for higher column efficiency is one of the major research areas in polymer-based monolithic column fabrication. In this research, two novel polymer-based HILIC monolithic columns with 400μm I.D.×800μm O.D. were prepared based on the thermally initiated co-polymerization of N,N-dimethyl-N-(3-methacrylamidopropyl)-N-(3-sulfopropyl) ammonium betaine (SPP) and ethylene glycol dimethacrylate (EDMA) or N,N'-methylenebisacrylamide (MBA). In order to obtain a satisfactory performance in terms of column permeability, mechanical stability, efficiency and selectivity, the polymerization parameters were systematically optimized. Column efficiencies as high as 142, 000 plates/m and 120, 000 plates/m were observed for the analysis of neutral compounds at 0.6mm/s on the poly(SPP-co-MBA) and poly(SPP-co-EDMA) monoliths, respectively. Furthermore, the Van Deemter plots for thiourea on the two monoliths were compared with that on a commercial silica based ZIC-HILIC column (3.5μm, 200Å, 150mm×300μm I.D.) using ACN/H 2 O (90/10, v/v) as the mobile phase at room temperature. It was noticeable that the Van Deemter curves for both monoliths, particularly the poly(SPP-co-MBA) monolith, are significantly flatter than that obtained for the ZIC-HILIC column, which indicates that in spite of their larger internal diameters, they yield better overall efficiency, with less peak dispersion, across a much wider range of usable linear velocities. A clearly better separation performance was also observed for nucleobases, nucleosides, nucleotides and small peptides on the poly(SPP-co-MBA) monolith compared to the ZIC-HILIC column. It is particularly worth mentioning that these 400μm I.D. polymer-based HILIC monolithic columns exhibit enhanced mechanical strength owing to the thicker capillary wall of the fused-silica capillaries. Copyright © 2017 Elsevier B.V. All rights reserved.

Covalent attachment of polymeric monolith to polyether ether ketone (PEEK) tubing.

Science.gov (United States)

Lv, Chunguang; Heiter, Jaana; Haljasorg, Tõiv; Leito, Ivo

2016-08-17

A new method of reproducible preparation of vinylic polymeric monolithic columns with a key step of covalently anchoring the monolith to PEEK surface is described. In order to chemically attach the polymer monolith to the tube wall, methacrylate functional groups were introduced onto PEEK surface by a three-step procedure, including surface etching, surface reduction and surface methacryloylation. The chemical state of the modified tubing surface was characterized by attenuated total reflectance infrared (ATR-IR) spectroscopy. It was found that the etching step is the key to successfully modifying the PEEK tubing surface. Poly(styrene-co-divinylbenzene) monoliths were in situ synthesized by thermally initiated free radical copolymerization within the confines of surface-vinylized PEEK tubings of dimensions close to ones conventionally used in HPLC and UHPLC (1.6 mm internal diameter, 10.0-12.5 cm length). Adhesion test was done by measuring the operating pressure drop, which the prepared stationary phases can withstand. Good pressure resistance, up to 140 bar/10 cm (flow rate 0.5 mL min(-1), acetonitrile as a mobile phase), indicates strong bonding of monolith to the tubing wall. The monolithic material was proven to have a permeability of 1.7 × 10 (-14) m(2), applying acetonitrile-water 70:30 (v/v) as a mobile phase. The column performance was reproducible from column to column and was evaluated via the isocratic separation of a series of alkylbenzenes in the reversed-phase mode (acetonitrile-water 70:30, v/v). The numbers of plates per meter at optimal flow rate were found to be between 26 000 and 32 000 for the different analytes. Copyright © 2016 Elsevier B.V. All rights reserved.

Investigation of the microstructure influence in the thermo-physical properties of U-Mo alloys through the laser flash method

Energy Technology Data Exchange (ETDEWEB)

Pedrosa, Tercio A.; Alves, Fabio F.; Kelmer, Paula F.; Santos, Ana Maria M.; Camarano, Denise das M.; Ferraz, Wilmar B., E-mail: tap@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2013-07-01

The U-Mo alloys are the most investigated and promising nuclear fuel material to be used in research and test reactors, according to the premises of the RERTR program, whose objective is to minimize the threats of nuclear weapons proliferation through the conversion of the nuclear fuels of research and test reactors form a high enrichment grade, HEU (235U>90%, to a low enrichment grade, LEU ({sup 235}U<20%). The high density of the U-Mo alloys associated with its ability to keep the gamma phase metastable at room temperature are the main advantages of these alloys, with Mo contents of 5, 7 and 10 wt% were induction melted and ageing heat treated at 300 and 500 deg C for 72, 120 and 240 h. Microstructural characterization was carried out in the as-cast and aged conditions through XRD and OM techniques. The laser Flash Method at environmental temperature was employed to investigate the variation of the thermal diffusivity as a function of the microstructure obtained in the as-cast and aged conditions. (author)

Investigation of the microstructure influence in the thermo-physical properties of U-Mo alloys through the laser flash method

International Nuclear Information System (INIS)

Pedrosa, Tercio A.; Alves, Fabio F.; Kelmer, Paula F.; Santos, Ana Maria M.; Camarano, Denise das M.; Ferraz, Wilmar B.

2013-01-01

The U-Mo alloys are the most investigated and promising nuclear fuel material to be used in research and test reactors, according to the premises of the RERTR program, whose objective is to minimize the threats of nuclear weapons proliferation through the conversion of the nuclear fuels of research and test reactors form a high enrichment grade, HEU (235U>90%, to a low enrichment grade, LEU ( 235 U<20%). The high density of the U-Mo alloys associated with its ability to keep the gamma phase metastable at room temperature are the main advantages of these alloys, with Mo contents of 5, 7 and 10 wt% were induction melted and ageing heat treated at 300 and 500 deg C for 72, 120 and 240 h. Microstructural characterization was carried out in the as-cast and aged conditions through XRD and OM techniques. The laser Flash Method at environmental temperature was employed to investigate the variation of the thermal diffusivity as a function of the microstructure obtained in the as-cast and aged conditions. (author)

Operation of plant to produce Mo-99 from fission products

International Nuclear Information System (INIS)

Marques, R.O.; Cristini, P.R.; Marziale, D.P.; Furnari, E.S.; Fernandez, H.O.

1987-01-01

As it is well known, the production of Mo-99/Tc-99m generators has an outstanding place in radioisotope programs of the Argentine National Atomic Energy Commission. The basic raw material is Mo-99 from fission of U-235. In 1985 the production plant of this radionuclide began to operate, according to an adaptation of the method that was developed in Kernforschungszentrum Karlsruhe. The present work describes the target irradiation conditions in the reactor RA-3 (mini plates of U/Al alloy with 90% enriched uranium), the flow diagram and the operative conditions of the production process. The containment, filtration and removal conditions of the generated fission gases and the disposal of liquid and solid wastes are also analyzed. On the basis of the experience achieved in the development of more than twenty production processes, process efficiency is analyzed, taking into account the theoretical evaluation resulting from the application of the computer program 'Origin'(ORML) to the conditions of our case. The purity characteristics of the final product are reported (Zr-95 0,1 ppm; Nb-95 1 ppm; Ru-103 20 ppm; I-131 10 ppm) as well as the chemical characteristics that make it suitable to be used in the production of Mo-99/I c-99m generators. (Author)

Preparation of polymer monolithic column functionalized by arsonic acid groups for mixed-mode capillary liquid chromatography.

Science.gov (United States)

Qin, Zhang-Na; Yu, Qiong-Wei; Wang, Ren-Qi; Feng, Yu-Qi

2018-04-27

A mixed-mode polymer monolithic column functionalized by arsonic acid groups was prepared by single-step in situ copolymerization of monomers p-methacryloylaminophenylarsonic acid (p-MAPHA) and pentaerythritol triacrylate (PETA). The prepared poly(p-MAPHA-co-PETA) monolithic column has a homogeneous monolithic structure with good permeability and mechanical stability. Zeta potential measurements reveal that the monolithic stationary phase holds a negative surface charge when the mobile phase resides in the pH range of 3.0-8.0. The retention mechanisms of prepared monolithic column are explored by the separation of selected polycyclic aromatic hydrocarbons (PAHs), nucleosides, and three basic compounds. The results indicate that the column functions in three different separation modes associated with reversed-phase chromatography based on hydrophobic interaction, hydrophilic interaction chromatography, and cation-exchange chromatography. The column efficiency of prepared monolithic column is estimated to be 70,000 and 76,000 theoretical plates/m for thiourea and naphthalene, respectively, at a linear flow velocity of 0.85 mm/s using acetonitrile/H 2 O (85/15, v/v) as the mobile phase. Furthermore, an analysis of the retention factors obtained for the PAHs indicates that the prepared monolithic column exhibits good reproducibility with relative standard deviations of 2.9%, 4.0%, and 4.7% based on run-to-run injections, column-to-column preparation, and batch-to-batch preparation, respectively. Finally, we investigate the separation performance of the proposed monolithic column for select phenols, sulfonamides, nucleobases and nucleosides. Copyright © 2018 Elsevier B.V. All rights reserved.

Star counts in M15 on U, B and V plates

Energy Technology Data Exchange (ETDEWEB)

Calvani, M [Padua Univ. (Italy). Ist. di Astronomia; Nobili, L [Padua Univ. (Italy). Ist. di Fisica; Turolla, R [Scuola Internazionale Superiore di Studi Avanzati, Trieste (Italy)

1980-11-01

We present new counts of stars in M15, using plates in B, V and U. We are able to explore relatively close to the central parts of the cluster (0.1 pc) and we derive the best fitting parameters for the star distribution.

Selenium fuel: Surface engineering of U(Mo) particles to optimise fuel performance

International Nuclear Information System (INIS)

Van den Berghe, S.; Leenaers, A.; Detavernier, C.

2010-01-01

Recent developments on the stabilisation of U(Mo) in-pile behaviour in plate-type fuel have focussed almost exclusively on the addition of Si to the Al matrix of the fuel. This has now culminated in a qualification effort in the form of the European LEONIDAS initiative for which irradiations will start in 2010. In this framework, many discussions have been held on the Si content of the matrix needed for stabilisation of the interaction phase and the requirement for the formation of Si-rich layers around the particles during the fabrication steps. However, it is clear that the Si needs to be incorporated in the interaction phase for it to be effective, for which the currently proposed methods depend on a diffusion mechanism, which is difficult to control. This has lead to the concept of a Si coated particle as a more efficient way of incorporating the Si in the fuel by putting it immediately where it will be required : at the fuel-matrix interface. As part of the SELENIUM (Surface Engineered Low ENrIched Uranium-Molybdenum fuel) project, SCK CEN has built a sputter coater for PVD magnetron sputter coating of particles in collaboration with the University of Ghent. The coater is equipped with three 3 inch magnetron sputter heads, allowing deposition of 3 different elements or a single element at high deposition speed. The particles are slowly rotated in a drum to produce homogeneous layer thicknesses. (author)

Fission induced swelling of U–Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Jeong, G.Y. [Ulsan National Institute of Science and Technology, 50 UNIST-gil, Eonyang-eup, Uljoo-gun, Ulsan 689-798 (Korea, Republic of); Park, J.M. [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2015-10-15

Fission-induced swelling of U–Mo/Al dispersion fuel meat was measured using microscopy images obtained from post-irradiation examination. The data of reduced-size plate-type test samples and rod-type test samples were employed for this work. A model to predict the meat swelling of U–Mo/Al dispersion fuel was developed. This model is composed of several submodels including a model for interaction layer (IL) growth between U–Mo and Al matrix, a model for IL thickness to IL volume conversion, a correlation for the fission-induced swelling of U–Mo alloy particles, a correlation for the fission-induced swelling of IL, and models of U–Mo and Al consumption by IL growth. The model was validated using full-size plate data that were not included in the model development.

Advanced Gasification Mercury/Trace Metal Control with Monolith Traps

Energy Technology Data Exchange (ETDEWEB)

Musich, Mark; Swanson, Michael; Dunham, Grant; Stanislowski, Joshua

2010-10-05

Two Corning monoliths and a non-carbon-based material have been identified as potential additives for mercury capture in syngas at temperatures above 400°F and pressure of 600 psig. A new Corning monolith formulation, GR-F1-2189, described as an active sample appeared to be the best monolith tested to date. The Corning SR Liquid monolith concept continues to be a strong candidate for mercury capture. Both monolith types allowed mercury reduction to below 5-μg/m{sup 3} (~5 ppb), a current U.S. Department of Energy (DOE) goal for trace metal control. Preparation methods for formulating the SR Liquid monolith impacted the ability of the monolith to capture mercury. The Energy & Environmental Research Center (EERC)-prepared Noncarbon Sorbents 1 and 2 appeared to offer potential for sustained and significant reduction of mercury concentration in the simulated fuel gas. The Noncarbon Sorbent 1 allowed sustained mercury reduction to below 5-μg/m{sup 3} (~5 ppb). The non-carbon-based sorbent appeared to offer the potential for regeneration, that is, desorption of mercury by temperature swing (using nitrogen and steam at temperatures above where adsorption takes place). A Corning cordierite monolith treated with a Group IB metal offered limited potential as a mercury sorbent. However, a Corning carbon-based monolith containing prereduced metallic species similar to those found on the noncarbon sorbents did not exhibit significant or sustained mercury reduction. EERC sorbents prepared with Group IB and IIB selenide appeared to have some promise for mercury capture. Unfortunately, these sorbents also released Se, as was evidenced by the measurement of H2Se in the effluent gas. All sorbents tested with arsine or hydrogen selenide, including Corning monoliths and the Group IB and IIB metal-based materials, showed an ability to capture arsine or hydrogen selenide at 400°F and 600 psig. Based on current testing, the noncarbon metal-based sorbents appear to be the most

ADVANCED GASIFICATION MERCURY/TRACE METAL CONTROL WITH MONOLITH TRAPS

Energy Technology Data Exchange (ETDEWEB)

Mark A. Musich; Michael L. Swanson; Grant E. Dunham; Joshua J. Stanislowski

2010-07-31

Two Corning monoliths and a non-carbon-based material have been identified as potential additives for mercury capture in syngas at temperatures above 400°F and pressure of 600 psig. A new Corning monolith formulation, GR-F1-2189, described as an active sample appeared to be the best monolith tested to date. The Corning SR Liquid monolith concept continues to be a strong candidate for mercury capture. Both monolith types allowed mercury reduction to below 5-μg/m3 (~5 ppb), a current U.S. Department of Energy (DOE) goal for trace metal control. Preparation methods for formulating the SR Liquid monolith impacted the ability of the monolith to capture mercury. The Energy & Environmental Research Center (EERC)-prepared Noncarbon Sorbents 1 and 2 appeared to offer potential for sustained and significant reduction of mercury concentration in the simulated fuel gas. The Noncarbon Sorbent 1 allowed sustained mercury reduction to below 5-μg/m3 (~5 ppb). The non-carbon-based sorbent appeared to offer the potential for regeneration, that is, desorption of mercury by temperature swing (using nitrogen and steam at temperatures above where adsorption takes place). A Corning cordierite monolith treated with a Group IB metal offered limited potential as a mercury sorbent. However, a Corning carbon-based monolith containing prereduced metallic species similar to those found on the noncarbon sorbents did not exhibit significant or sustained mercury reduction. EERC sorbents prepared with Group IB and IIB selenide appeared to have some promise for mercury capture. Unfortunately, these sorbents also released Se, as was evidenced by the measurement of H2Se in the effluent gas. All sorbents tested with arsine or hydrogen selenide, including Corning monoliths and the Group IB and IIB metal-based materials, showed an ability to capture arsine or hydrogen selenide at 400°F and 600 psig. Based on current testing, the noncarbon metal-based sorbents appear to be the most effective arsine

Optimization of parameters in the simulation of the interdiffusion layer growth in Al-U couples

International Nuclear Information System (INIS)

Kniznik, Laura; Alonso, Paula R.; Gargano, Pablo H.; Rubiolo, Gerardo H.

2009-01-01

U-Mo alloy dispersed in aluminum is considered as a high U density fuel for research reactors. In and out of pile experiments showed a reaction layer in U-Mo/Al interphase with formation of intermetallics compounds: Al 2 U, Al 3 U and Al 4 U. Under irradiation, porosities originate an unacceptable swelling of the fuel plate. The kinetics of growth of the intermetallic compounds in the U-Mo/Al interphase is treated in the Al 3 U/Al couple as a planar moving boundary problem due to diffusion of Al and U atoms in the direction perpendicular to the interphase surface. Using data from literature, we built a thermodynamic database to be read by the Thermocalc code to calculate phase equilibria. The diffusion problem was carried out by the DICTRA simulation package which articulates data evaluated by Thermocalc with a mobility database. In a previous work we built preliminary databases, for both free energy and mobilities. In the present work, we adjust the parameters from experimental thermodynamic equilibria and concentration profiles existing in literature, and we simulate satisfactorily the growth of the Al 4 U phase. (author)

In situ synthesis of metal-organic frameworks in a porous polymer monolith as the stationary phase for capillary liquid chromatography.

Science.gov (United States)

Yang, Shengchao; Ye, Fanggui; Zhang, Cong; Shen, Shufen; Zhao, Shulin

2015-04-21

In this study, HKUST-1 was synthesized in situ on the porous polymer monolith as the stationary phase for capillary liquid chromatography (cLC). The unique carboxyl functionalized poly(methacrylic acid-co-ethylene dimethacrylate) (poly(MAA-co-EDMA)) monolith was used as a support to directly grow HKUST-1 by a controlled layer-by-layer self-assembly strategy. X-ray diffraction, scanning electron microscopy, energy dispersive X-ray spectrometry, and Fourier transform infrared spectroscopy of the resulting HKUST-1-poly(MAA-co-EDMA) monoliths indicated that HKUST-1 was successfully grafted onto the pore surface of the poly(MAA-co-EDMA) monolith. The column performance of HKUST-1-poly(MAA-co-EDMA) monoliths for the separation of various small molecules, such as benzenediols, xylenes, ethylbenzenes, and styrenes, was evaluated. The chromatographic performance was found to improve with increasing HKUST-1 density, and the column efficiencies and resolutions of HKUST-1-poly(MAA-co-EDMA) monoliths were 18 320-19 890 plates m(-1) and 1.62-6.42, respectively, for benzenediols. The HKUST-1-poly(MAA-co-EDMA) monolith displayed enhanced resolution for the separation of positional isomers when compared to the traditional C18 and HKUST-1 incorporated polymer monoliths. Hydrophobic, π-π, and hydrogen bonding interactions within the HKUST-1-poly(MAA-co-EDMA) monolith were observed in the separation of small molecules. The results showed that the HKUST-1-poly(MAA-co-EDMA) monoliths are promising stationary phases for cLC.

Spark plasma sintering and microstructural analysis of pure and Mo doped U3Si2 pellets

Science.gov (United States)

Lopes, Denise Adorno; Benarosch, Anna; Middleburgh, Simon; Johnson, Kyle D.

2017-12-01

U3Si2 has been considered as an alternative fuel for Light Water Reactors (LWRs) within the Accident Tolerant Fuels (ATF) initiative, begun after the Fukushima-Daiichi Nuclear accidents. Its main advantages are high thermal conductivity and high heavy metal density. Despite these benefits, U3Si2 presents an anisotropic crystallographic structure and low solubility of fission products, which can result in undesirable effects under irradiation conditions. In this paper, spark plasma sintering (SPS) of U3Si2 pellets is studied, with evaluation of the resulting microstructure. Additionally, exploiting the short sintering time in SPS, a molybdenum doped pellet was produced to investigate the early stages of the Mo-U3Si2 interaction, and analyze how this fission product is accommodated in the fuel matrix. The results show that pellets of U3Si2 with high density (>95% TD) can be obtained with SPS in the temperature range of 1200°C-1300 °C. Moreover, the short time employed in this technique was found to generate a unique microstructure for this fuel, composed mainly of closed nano-pores (uranium with small quantities of dissolved Si and Mo at the front of the reaction.

A review of microstructural analysis on U3Si2-Al plate-type fuel

International Nuclear Information System (INIS)

Ti Zhongxin; Guo Yibai

1995-12-01

The microstructure of U 3 Si 2 -Al plate-type fuel, that is the microstructure of fuel particles, compatibility of the fuel particles and Al matrix, fuel particles distribution, dogbone area morphology, clad and meat thickness, bone quality of clad/frame and clad/fuel core, and the effect of these factors on products quality were comprehensively investigated and analyzed by means of optical microscopy (OM), scanning electron microscopy (SEM), X-ray diffraction (XRD), energy dispersive X-ray spectrometry (EDX), image processing technique, etc.. The main results are as following: U-7.7%Si alloy contains two phases: primary U 3 Si 2 and small amount of USi (about 12%), free-uranium was not detected in fuel particles; the dogbone area is the key factor affecting fuel plate quality (1 ref., 16 figs., 4 tabs.)

Metallographic analysis of irradiated U3Si2/Al fuel element plate of 2.96 gU/cm3 density

International Nuclear Information System (INIS)

Maman Kartaman Ajiriyanto; Aslina Br Ginting; Junaedi

2018-01-01

Metallographic analysis of U 3 Si 2 /Al fuel element plate has been performed in hot cell. The purpose of metallographic analysis is to study changes in PEB U 3 Si 2 /Al microstructure and AlMg 2 cladding thickness after irradiation in reactor until burn up of 56 %. The fuel element plate of irradiated U 3 Si 2 /Al was cut in top, middle and bottom positions with each size around 5 x 5 x 1.37 mm. Metallographic preparation starts from sample cutting using cutting machine with low speed and sample mounting, grinding and polishing in hot cell 104–105. Sample mounting was done by using resin for more than 10 hours followed by grinding with sand papers up to grit size of 2400 and polishing with diamond paste of size 3 to 1 micron at a rotational speed of 150 rpm for 5 minutes. Microstructure observation was performed with optical microscope in hot cell 107 at 200 times magnification. Microstructure examination reveals U 3 Si 2 particles with inverse forms and sizes, Al matrix and AlMg 2 cladding were spread along the U 3 Si 2 /Al side. Microstructure observation of irradiated U 3 Si 2 /Al has not shown good result because only topography observation of U 3 Si 2 /Al meat, Al matrix and AlMg 2 cladding can be done due to limited capability of the optical microscope in hot cell, where maximum magnification can be attained only at 200 times so that the phenomenon of interaction layer and small gas bubble can not be observed. However, U 3 Si 2 /Al microstructure of 56 % burnup, if compared to the microstructure of U 3 Si 2 /Al fuel element plate of 60 % burnup from previous researcher, shows interaction between U 3 Si 2 meat with Al matrix and the existence of layers with a thickness about 5 up to 20 microns. Meanwhile, the observed thickness of AlMg 2 cladding is greater than 0.25 mm, which indicates that irradiation does not significantly change the thickness of AlMg 2 cladding so that the overall irradiated U 3 Si 2 -Al still has good integrity and stability. (author)

Facile preparation of organic-silica hybrid monolith for capillary hydrophilic liquid chromatography based on "thiol-ene" click chemistry.

Science.gov (United States)

Chen, Ming-Luan; Zhang, Jun; Zhang, Zheng; Yuan, Bi-Feng; Yu, Qiong-Wei; Feng, Yu-Qi

2013-04-05

In this work, a one-step approach to facile preparation of organic-inorganic hybrid monoliths was successfully developed. After vinyl-end organic monomers and azobisisobutyronitrile (AIBN) were mixed with hydrolyzed tetramethoxysilane (TMOS) and 3-mercaptopropyltrimethoxysilane (MPTMS), the homogeneous mixture was introduced into a fused-silica capillary for simultaneous polycondensation and "thiol-ene" click reaction to form the organic-silica hybrid monoliths. By employing this strategy, two types of organic-silica hybrid monoliths with positively charged quaternary ammonium and amide groups were prepared, respectively. The functional groups were successfully introduced onto the monoliths during the sol-gel process with "thiol-ene" click reaction, which was demonstrated by ζ-potential assessment, energy dispersive X-ray spectroscopy (EDX), and Fourier transform infrared (FT-IR) spectroscopy. The porous structure of the prepared monolithic columns was examined by scanning electron microscopy (SEM), nitrogen adsorption-desorption measurement, and mercury intrusion porosimetry. These results indicate the prepared organic-silica hybrid monoliths possess homogeneous column bed, large specific surface area, good mechanical stability, and excellent permeability. The prepared monolithic columns were then applied for anion-exchange/hydrophilic interaction liquid chromatography. Different types of analytes, including benzoic acids, inorganic ions, nucleosides, and nucleotides, were well separated with high column efficiency around 80,000-130,000 plates/m. Taken together, we present a facile and universal strategy to prepare organic-silica hybrid monoliths with a variety of organic monomers using one-step approach. Copyright © 2013 Elsevier B.V. All rights reserved.

Application of laser ablation inductivly coupled plasma mass spectrometry for characterization of U-7Mo/Al-55i dispersion fuels

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeong Mook; Park, Jai Il; Youn, Young Sang; Ha, Yeong Keong; Kim, Jong Yun [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-04-15

This technical note demonstrates the feasibility of using laser ablation inductively coupled plasma mass spectrometry for the characterization of U–7Mo/Al–5Si dispersion fuel. Our measurements show 5.0% Relative Standard Deviation (RSD) for the reproducibility of measured {sup 98}Mo/{sup 238}U ratios in fuel particles from spot analysis, and 3.4% RSD for {sup 98}Mo/{sup 238}U ratios in a NIST-SRM 612 glass standard. Line scanning allows for the distinction of U–7Mo fuel particles from the Al–5Si matrix. Each mass spectrum peak indicates the presence of U–7Mo fuel particles, and the time width of each peak corresponds to the size of that fuel particle. The size of the fuel particles is estimated from the time width of the mass spectrum peak for {sup 98}Mo by considering the scan rate used during the line scan. This preliminary application clearly demonstrates that laser ablation inductively coupled plasma mass spectrometry can directly identify isotope ratios and sizes of the fuel particles in U–Mo/Al dispersion fuel. Once optimized further, this instrument will be a powerful tool for investigating irradiated dispersion fuels in terms of fission product distributions in fuel matrices, and the changes in fuel particle size or shape after irradiation.

Incorporation of ionic liquid into porous polymer monoliths to enhance the separation of small molecules in reversed-phase high-performance liquid chromatography.

Science.gov (United States)

Wang, Jiafei; Bai, Ligai; Wei, Zhen; Qin, Junxiao; Ma, Yamin; Liu, Haiyan

2015-06-01

An ionic liquid was incorporated into the porous polymer monoliths to afford stationary phases with enhanced chromatographic performance for small molecules in reversed-phase high-performance liquid chromatography. The effect of the ionic liquid in the polymerization mixture on the performance of the monoliths was studied in detail. While monoliths without ionic liquid exhibited poor resolution and low efficiency, the addition of ionic liquid to the polymerization mixture provides highly increased resolution and high efficiency. The chromatographic performances of the monoliths were demonstrated by the separations of various small molecules including aromatic hydrocarbons, isomers, and homologues using a binary polar mobile phase. The present column efficiency reached 27 000 plates/m, which showed that the ionic liquid monoliths are alternative stationary phases in the separation of small molecules by high-performance liquid chromatography. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Możliwości leczenia kamicy moczowodowej u dzieci

Directory of Open Access Journals (Sweden)

Ewa Straż-Żebrowska

2010-09-01

Full Text Available Kamicą układu moczowego nazywamy schorzenie, w którym wytwarzane są złogi w obrębie nerek lub dróg moczowych. Obecnie obserwuje się wzrost częstości występowania kamicy. Szacuje się, iż częstość występowania kamicy układu moczowego wynosi 0,5-5% w grupie dorosłych i 0,1-5% u dzieci. Istotnym problemem jest wykrywanie złogów u małych dzieci, również u niemowląt. Biorąc pod uwagę coraz częstsze wykrywanie kamicy układu moczowego, jej nawrotowość i powikłania, ważne jest wypracowanie schematów leczniczych skutecznych i jednocześnie bezpiecznych, nawet przy konieczności ich kilkukrotnego powtarzania. Postępowanie zachowawcze jest skuteczne w 80%, w przypadkach złogów o wielkości do 4 mm. Powszechnie stosowane do lat 80. XX wieku leczenie operacyjne związane z nacięciem miąższu nerki obecnie w większości przypadków zastępowane jest metodami mniej inwazyjnymi, takimi jak ESWL, PCNL, URSL, a wskazania do leczenia chirurgicznego kamicy zostały znacznie ograniczone. W niniejszej pracy przedstawiono przypadek pacjentki z kamicą układu moczowego leczonej kilkakrotnie metodą ESWL bez efektu, a następnie z uwagi na obecność złogów w moczowodzie za pomocą URSL. Jednym z powikłań po zabiegu ESWL może być „droga kamieni”. Po wykonaniu zabiegu ESWL pacjent wymaga monitorowania poprzez wykonywanie badań USG, co pozwala na szybkie uwidocznienie powikłania w postaci „drogi kamieni” i wdrożenie dodatkowego postępowania ułatwiającego wydalenie złogów. „Droga kamieni” może spowodować utrudnienie odpływu moczu, co może być przyczyną dolegliwości bólowych oraz różnego stopnia poszerzeń moczowodów i układów kielichowo-miedniczkowych. Często wymaga interwencji zabiegowej w trybie pilnym. W ciągu ostatnich dwudziestu lat nastą- piły w Polsce znaczne zmiany w sposobach leczenia kamicy moczowodowej. Rutynowo stosuje się URSL i ESWL, do rzadkości natomiast należy zak

Computational and experimental analysis of causes for local deformation of research reactor U-Mo fuel pin claddings in case of high burn-ups

International Nuclear Information System (INIS)

Popov, V.V.; Khmelevsky, M.Ya.; Lukichev, V.A.; Golosov, O.A.

2005-01-01

Post-reactor investigations of (U-Mo) fuel pins irradiated in the IVV-2M reactor have allowed to determine: the change in a fuel pin volume; the dimensions and the kind of the local deformation of fuel pin claddings; the amount of gases released under the cladding from the fuel composition, the thickness and appearance of the interaction layer of between the (U-Mo) particles and aluminium as a matrix material. The computational analysis of the stressed-strained state of fuel pins has shown that the major contribution to the increase of the fuel pin volume is made by the fuel swelling caused by the solid products of fission being formed in the process of operation. The emergence of the (U-Mo) fuel-aluminium matrix interaction layers around the (U-Mo) particles results in formation and evolution of lamination cavities inside the fuel composition under the joint action of the pressure of process gases and gaseous fission products. In case of high burn-up a local bulge of a fuel pin cladding is being formed in the fuel lamination area caused by the pressure of gases in the presence of creep in the fuel pin cladding material. The computational results relating to the local strain in a research reactor (U-Mo) fuel pin are in a good accordance with the results of the post-reactor investigations. (author)

Corrosion of MTR type fuel plates containing U3O8-Al cermet cores

International Nuclear Information System (INIS)

Durazzo, M.

1985-01-01

The fuel plate samples containing U 3 O 8 -Al cermet cores with concentrations from 10 to 90% of U 3 O 8 weight were fabricated. Samples with 58% of U 3 O 8 eight were fabricated using compacts with densities from 75 to 95% of theoretical density. The influences of U 3 O 8 concentration and porosity of compacted core on porosity and uniformity of core thickness are discussed. The U 3 O 8 -Al cores were submitted to corrosion tests and exposed to deionized water at temperatures of 30, 50, 70 and 90 0 C by cladding deffect produced artificially. The results shown that core corrosion is accompanied by hydrogen release. The total volum of released hydrogen and the time interval to observe the initiation of hydrogen releasing (incubation time) are depending on core pososity and absolute temperature. A mechanism for U 3 O 8 -Al core corrosion process is proposed and discussed. The cladding of fuel plate samples was submitted to corrosion tests under similar conditons of the IAE-R1 reactor operating at 2, 5 and 10 MW. (Author) [pt

Deconvolution of trace element (As, Cr, Mo, Th, U) sources and pathways to surface waters of a gold mining-influenced watershed.

Science.gov (United States)

Grosbois, C; Schäfer, J; Bril, H; Blanc, G; Bossy, A

2009-03-01

The Upper Isle River (SW France) drains the second most productive gold-mining district of France. A high resolution survey during one hydrological year of As, Cl(-), Cr, Fe, Mn, Mo, SO(4)(2-), Th and U dissolved concentrations in surface water aimed to better understand pathways of trace element export to the river system downstream from the mining district. Dissolved concentrations of As (up to 35000 ng/L) and Mo (up to 292 ng/L) were about 3-fold higher than the regional dissolved background and showed a negative logarithmic relation with discharge. Dissolved concentrations of Cr (up to 483 ng/L), Th (up to 48 ng/L) and U (up to 184 ng/L) increased with discharge. Geochemical relationships between molar ratios in surface water, geochemical background as well as rain- and groundwater data were combined. The contrasting behavior of distinct element groups was explained by a scenario involving three seasonal components: (i) The high flow component is poorly concentrated in As and Mo but highly concentrated in Cr, Th, U. This has been attributed to diffuse sources such as water-soil interactions, atmospheric inputs, bedrock and bed sediment weathering. Although this component probably also includes a contribution by weathering of sulfide veins, this signal is masked by dilution. (ii) One low flow component presents high SO(4)(2-), Fe, As and Mo and moderate Cr, Th and U concentrations. This component has been attributed to point sources such as mine gallery effluents, mining waste weathering and groundwater inputs from natural and/or mining-induced sulfide oxidation in the ore deposit. (iii) A second low flow component showing high As plus Mo concentrations associated with very low SO(4)(2-), Fe, Cr, Th and U concentrations, probably reflects trace element scavenging by ferric oxyhydroxide formation in the adjacent aquifer. This is supported by the decrease of Fe, Cr, Th and U in surface waters. Flux estimates suggest contrasting element-specific impacts on annual

Development of integrated cask body and base plate

International Nuclear Information System (INIS)

Sasaki, T.; Koyama, Y.; Yoshida, T.; Wada, T.

2015-01-01

The average of occupancy of stored spent-fuel in the nuclear power plants have reached 70 percent and it is anticipated that the demand of metal casks for the storage and transportation of spent-fuel rise after resuming the operations. The main part of metal cask consists of main body, neutron shield and external cylinder. We have developed the manufacturing technology of Integrated Cask Body and Base Plate by integrating Cask Body and Base Plate as monolithic forging with the goal of cost reduction, manufacturing period shortening and further reliability improvement. Here, we report the manufacturing technology, code compliance and obtained properties of Integrated Cask body and Base Plate. (author)

Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

Energy Technology Data Exchange (ETDEWEB)

Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

2013-12-15

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

International Nuclear Information System (INIS)

Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

2013-01-01

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60% 235 U; the mini-rods were irradiated to an average burnup of ∼ 85% 235 U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%

Preparation of 235U target by electrodeposition

International Nuclear Information System (INIS)

Chen Qiping; Zhong Wenbin; Li Yougen

2004-12-01

A target for the production of fission 99 Mo in a nuclear reactor is composed of an enclosed, cylindrical vessel. Preferable vessel is comprised of stainless steel, having a thin, continuous, uniform layer of 235 U integrally bonded to its inner walls. Two processes are introduced for electrodepositing uranium on to the inner walls of the vessel. One processes is electrodepositing UO 2 from UO 2 (NO 3 ) 2 -(NH 4 ) 2 CO 4 ·H 2 O solution; the other is electrodepositing pure uranium metal from molten salt. Its plating efficiency and plating quantity from a molten bath is higher than UO 2 from the aqueous system. (authors)

Soft zone formation in dissimilar welds between two Cr-Mo steels

International Nuclear Information System (INIS)

Albert, S.K.; Gill, T.P.S.; Tyagi, A.K.; Mannan, S.L.; Rodriguez, P.; Kulkarni, S.D.

1997-01-01

Two dissimilar weldments between 9Cr-1Mo and 2.25Cr-1Mo ferritic steels have been characterized for their microstructural stability during various postweld heat treatments (PWHTs). The samples for the investigation were extracted from bead-on-plate weldments made by depositing 2.25Cr-1Mo weld metal on 9Cr-1Mo base plate and vice versa. Subsequent application of PWHT resulted in the formation of a soft zone in the low Cr ferritic steel weld or base plate. A carbide-rich hard zone, adjoining the soft zone, was also detected in the high Cr side of the weldment. Unmixed zones in the weld metal provided additional soft and hard zones in the weld metals. The migration of carbon from low-Cr steel to high-Cr steel, driven by the carbon activity gradient, has been shown to be responsible for the formation of soft and hard zones. A carbon activity diagram for 2.25Cr-1Mo/9Cr-1Mo weldments has been proposed to aid in the selection of welding consumables for reducing or preventing the soft zone formation

Fabrication of a novel hemin-based monolithic column and its application in separation of protein from complex bio-matrix.

Science.gov (United States)

Jiang, Xiaoya; Zhang, Doudou; Li, Xueying; Wang, Xixi; Bai, Ligai; Liu, Haiyan; Yan, Hongyuan

2017-05-10

A novel polymer-based monolithic column was prepared via redox initiation system within the confines of a stainless steel column with 4.6mm i.d. In the processes, hemin and lauryl methacrylate were used as co-monomers; ethylene dimethacrylate as crosslinking agent; n-butyl alcohol, ethanediol, and N, N-dimethylformamide as tri-porogens; benzoyl peroxide and N, N-dimethyl aniline as redox initiation system. The resulting polymer-based monolithic columns were characterized by scanning electron microscopy, nitrogen adsorption-desorption instrument, and mercury intrusion porosimeter, respectively. The results illustrated that the improved monolith had relative uniform porous structure, good permeability, and low back pressure. Aromatic compounds were used to test the chromatographic behavior of the monolith, resulting in highest column efficiency of 19 880 plates per meter with reversed-phase mechanism. Furthermore, the homemade monolith was used as the stationary phase of high performance liquid chromatography to separate proteins from complex bio-matrix, including human plasma, egg white, and snailase. The results showed that the monolithic column occupied good separation ability with these complex bio-samples. Excellent specific character of the homemade hemin-based monolith was that it could simultaneously remove high-abundance proteins (including human serum albumin, immunoglobulin G, and human fibrinogen) from human plasma and separate other proteins to different fractions. Copyright © 2017 Elsevier B.V. All rights reserved.

Incorporation of metal-organic framework HKUST-1 into porous polymer monolithic capillary columns to enhance the chromatographic separation of small molecules.

Science.gov (United States)

Yang, Shengchao; Ye, Fanggui; Lv, Qinghui; Zhang, Cong; Shen, Shufen; Zhao, Shulin

2014-09-19

Metal-organic framework (MOF) HKUST-1 nanoparticles have been incorporated into poly(glycidyl methacrylate-co-ethylene dimethacrylate) (HKUST-1-poly(GMA-co-EDMA)) monoliths to afford stationary phases with enhanced chromatographic performance of small molecules in the reversed phase capillary liquid chromatography. The effect of HKUST-1 nanoparticles in the polymerization mixture on the performance of the monolithic column was explored in detail. While the bare poly(GMA-co-EDMA) monolith exhibited poor resolution (RsHKUST-1 nanoparticles to the polymerization mixture provide high increased resolution (Rs≥1.3) and high efficiency ranged from 16,300 to 44,300plates/m. Chromatographic performance of HKUST-1-poly(GMA-co-EDMA) monolith was demonstrated by separation of various analytes including polycyclic aromatic hydrocarbons, ethylbenzene and styrene, phenols and aromatic acids using a binary polar mobile phase (CH3CN/H2O). The HKUST-1-poly(GMA-co-EDMA) monolith displayed enhanced hydrophobic and π-π interaction characteristics in the reversed phase separation of test analytes compared to the bare poly(GMA-co-EDMA) monolith. The experiment results showed that HKUST-1-poly(GMA-co-EDMA) monoliths are an alternative to enhance the chromatographic separation of small molecules. Copyright © 2014 Elsevier B.V. All rights reserved.

COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

Directory of Open Access Journals (Sweden)

ALEKSEY. L. IZHUTOV

2013-12-01

The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; the mini-rods were irradiated to an average burnup of ∼ 85%235U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

Fast preparation of hybrid monolithic columns via photo-initiated thiol-yne polymerization for capillary liquid chromatography.

Science.gov (United States)

Ma, Shujuan; Zhang, Haiyang; Li, Ya; Li, Yanan; Zhang, Na; Ou, Junjie; Ye, Mingliang; Wei, Yinmao

2018-02-23

Although several approaches have been developed to fabricate hybrid monoliths, it would still take a few hours to finish the formation of monoliths. Herein, photo-initiated thiol-yne polymerization was first adopted to in situ fabricate hybrid monoliths within the confines of UV-transparent fused-silica capillary. A silicon-containing diyne (1,3-diethynyltetramethyl-disiloxane, DYDS) was copolymerized with three multithiols, 1,6-hexanedithiol, trimethylolpropane tris(3-mercaptopropionate) and pentaerythriol tetrakis(3-mercaptopropionate), by using a binary porogenic system of diethylene glycol diethyl ether (DEGDE)/poly(ethylene glycol) (PEG200) within 10 min. Several characterizations of three hybrid monoliths (assigned as I, II and III, respectively) were performed. The results showed that these hybrid monoliths possessed bicontinuous porous structure, which was remarkably different from that via typical free-radical polymerization. The highest column efficiency of 76,000 plates per meter for butylbenzene was obtained on the column I in reversed-phase liquid chromatography (RPLC). It was observed that the efficiencies for strong-retained butylbenzene were almost close to those of weak-retained benzene, indicating a retention-independent efficient performance of small molecules on hybrid column I. The surface area of this hybrid monolith was very small in the dry state (less than 10.0 m 2 /g), and the chromatographic behavior of hybrid monolithic columns would be possibly explained by radical-mediated step-growth process of thiol-yne polymerization. Finally, the column I was applied for separation of BSA tryptic digest by cLC-MS/MS, indicating satisfactory separation ability for complicated samples. Copyright © 2018 Elsevier B.V. All rights reserved.

Solid State Characterizations of Long-Term Leached Cast Stone Monoliths

Energy Technology Data Exchange (ETDEWEB)

Asmussen, Robert M.; Pearce, Carolyn I.; Parker, Kent E.; Miller, Brian W.; Lee, Brady D.; Buck, Edgar C.; Washton, Nancy M.; Bowden, Mark E.; Lawter, Amanda R.; McElroy, Erin M.; Serne, R Jeffrey

2016-09-30

This report describes the results from the solid phase characterization of six Cast Stone monoliths from the extended leach tests recently reported on (Serne et al. 2016),that were selected for characterization using multiple state-of-the-art approaches. The Cast Stone samples investigated were leached for > 590 d in the EPA Method 1315 test then archived for > 390 d in their final leachate. After reporting the long term leach behavior of the monoliths (containing radioactive ⁹⁹Tc and stable ¹²⁷I spikes and for original Westsik et al. 2013 fabricated monoliths, ²³⁸U), it was suggested that physical changes to the waste forms and a depleting inventory of contaminants of potential concern may mean that effective diffusivity calculations past 63 d should not be used to accurately represent long-term waste form behavior. These novel investigations, in both length of leaching time and application of solid state techniques, provide an initial arsenal of techniques which can be utilized to perform such Cast Stone solid phase characterization work, which in turn can support upcoming performance assessment maintenance. The work was performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to characterize several properties of the long- term leached Cast Stone monolith samples.

SEM in situ MiniCantilever Beam Bending of U-10Mo/Zr/Al Fuel Elements

Energy Technology Data Exchange (ETDEWEB)

Mook, William [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Baldwin, Jon K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez, Ricardo M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mara, Nathan A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2014-06-16

In this work, the fracture behavior of Al/Zr and Zr/dU-10Mo interfaces was measured via the minicantilever bend technique. The energy dissipation rates were found to be approximately 3.7-5 mj/mm² and 5.9 mj/mm² for each interface, respectively. It was found that in order to test the Zr/U-10Mo interface, location of the hinge of the cantilever was a key parameter. While this test could be adapted to hot cell use through careful alignment fixturing and measurement of crack lengths with an optical microscope (as opposed to SEM, which was used here out of convenience), machining of the cantilevers via MiniMill in such a way as to locate the interfaces at the cantilever hinge, as well as proper placement of a femtosecond laser notch will continue to be key challenges in a hot cell environment.

The genesis of Kurišková U-Mo ore deposit

International Nuclear Information System (INIS)

Demko, R.; Biroň, A.; Novotný, L.; Bartalský, B.

2014-01-01

The U-Mo ores of the known uranium deposit Kurišková located in the Huta volcano-sedimentary complex (HVC) of lower Permian age belongs to the Petrova Hora Formation of the North-Gemeric tectonic unit (Western Carpathians). The HVC is built up by volcanic rocks of bimodal basalt-rhyolite association, intercalated with sandstones, mudstones and claystones. Based on the sedimentary facies reconstruction, it is supposed paleoenvironment of seasonally flooded shallow lakes of continental fluvial plain with transition to estuaries and shallow marine facies of continental shelf in the upper part of HVC.

Study of the residual porosity in fuel plate cores based on U3O8 - Al dispersions

International Nuclear Information System (INIS)

Durazzo, M.

2005-01-01

The residual porosity in the meat of nuclear dispersion fuel plates, the fabrication voids, explains the corrosion behaviour of the meats when exposed to the water used as coolant and moderator of MTR type research reactors. The fabrication voids also explain variations in irradiation performance of many fuel dispersion for nuclear reactors. To obtain improved corrosion and irradiation performance, we must understand the fabrication factors that control the amount of void volume in fuel plate meats. The purpose of this study was to investigate the void content of aluminum-base dispersion-type U 3 O 8 -Al fuel plates depending on the characteristics of the starting fuel dispersion used to produce the fuel meat, which is fabricated by pressing. The void content depends on the U 3 O 8 concentration. For a particular U 3 O 8 content, the rolling process establishes a constant void concentration, which is called equilibrium porosity. The equilibrium quantity of voids is insensitive to the initial density of the fuel compact. (author)

Periodic imidazolium-bridged hybrid monolith for high-efficiency capillary liquid chromatography with enhanced selectivity.

Science.gov (United States)

Qiao, Xiaoqiang; Zhang, Niu; Han, Manman; Li, Xueyun; Qin, Xinying; Shen, Shigang

2017-03-01

A novel periodic imidazolium-bridged hybrid monolithic column was developed. With diene imidazolium ionic liquid 1-allyl-3-vinylimidazolium bromide as both cross-linker and organic functionalized reagent, a new periodic imidazolium-bridged hybrid monolithic column was facilely prepared in capillary with homogeneously distributed cationic imidazolium by a one-step free-radical polymerization with polyhedral oligomeric silsesquioxane methacryl substituted. The successful preparation of the new column was verified by Fourier transform infrared spectroscopy, scanning electron microscopy, elemental analysis, and surface area analysis. Most interestingly, the bonded amount of 1-allyl-3-vinylimidazolium bromide of the new column is three times higher than that of the conventional imidazolium-embedded hybrid monolithic column and the specific surface area of the column reached 478 m 2 /g. The new column exhibited high stability, excellent separation efficiency, and enhanced separation selectivity. The column efficiency reached 151 000 plates/m for alkylbenzenes. Furthermore, the new column was successfully used for separation of highly polar nucleosides and nucleic acid bases with pure water as mobile phase and even bovine serum albumin tryptic digest. All these results demonstrate the periodic imidazolium-bridged hybrid monolithic column is a good separation media and can be used for chromatographic separation of small molecules and complex biological samples with high efficiency. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Fabrication of high-uranium-loaded U/sub 3/O/sub 8/-Al developmental fuel plates

Energy Technology Data Exchange (ETDEWEB)

Copeland, G.L.; Martin, M.M.

1980-12-01

A common plate-type fuel for research and test reactors is U/sub 3/O/sub 8/ dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the /sup 235/U enrichment from above 90% to below 20% for these fuels to lessen the risk of diversion of the uranium for nonpeaceful uses. Thus, the uranium content of the fuel plates has to be increased to maintain the performance of the reactors. This paper describes work at ORNL to determine the maximal uranium loading for these fuels that can be fabricated with commercially proven materials and techniques and that can be expected to perform satisfactorily in service.

Fission-induced recrystallization effect on intergranular bubble-driven swelling in U-Mo fuel

Energy Technology Data Exchange (ETDEWEB)

Liang, Linyun; Mei, Zhi-Gang; Yacout, Abdellatif M.

2017-10-01

We have developed a mesoscale phase-field model for studying the effect of recrystallization on the gas-bubble-driven swelling in irradiated U-Mo alloy fuel. The model can simulate the microstructural evolution of the intergranular gas bubbles on the grain boundaries as well as the recrystallization process. Our simulation results show that the intergranular gas-bubble-induced fuel swelling exhibits two stages: slow swelling kinetics before recrystallization and rapid swelling kinetics with recrystallization. We observe that the recrystallization can significantly expedite the formation and growth of gas bubbles at high fission densities. The reason is that the recrystallization process increases the nucleation probability of gas bubbles and reduces the diffusion time of fission gases from grain interior to grain boundaries by increasing the grain boundary area and decreasing the diffusion distance. The simulated gas bubble shape, size distribution, and density on the grain boundaries are consistent with experimental measurements. We investigate the effect of the recrystallization on the gas-bubble-driven fuel swelling in UMo through varying the initial grain size and grain aspect ratio. We conclude that the initial microstructure of fuel, such as grain size and grain aspect ratio, can be used to effectively control the recrystallization and therefore reduce the swelling in U-Mo fuel.

Flaw behavior in mechanically loaded clad plates

International Nuclear Information System (INIS)

Iskander, S.K.; Robinson, G.C.; Oland, C.B.

1989-01-01

A small crack near the inner surface of clad nuclear reactor pressure vessels is an important consideration in the safety assessment of the structural integrity of the vessel. Four-point bend tests on large plate specimens, conforming to ASTM specification for pressure vessel plates, alloy steels, quenched and tempered, Mn-Mo and Mn-Mo-Ni (A533) grade B six clad and two unclad with stainless steels 308, 309 and 312 weld wires, were performed to determine the effect of cladding upon the propagation of small surface cracks subjected to stress states. Results indicated that the tough surface layer composed of cladding and/or heat-affected zone has enhanced the load-bearing capacity of plates under conditions where unclad plates have ruptured. The results are interpreted in terms of fracture mechanics. The behavior of flaws in clad reactor pressure vessels is examined in the light of the test results. 11 refs., 8 figs., 2 tabs

Characterization of fission gas bubbles in irradiated U-10Mo fuel

Energy Technology Data Exchange (ETDEWEB)

Casella, Andrew M.; Burkes, Douglas E.; MacFarlan, Paul J.; Buck, Edgar C.

2017-09-01

Irradiated U-10Mo fuel samples were prepared with traditional mechanical potting and polishing methods with in a hot cell. They were then removed and imaged with an SEM located outside of a hot cell. The images were then processed with basic imaging techniques from 3 separate software packages. The results were compared and a baseline method for characterization of fission gas bubbles in the samples is proposed. It is hoped that through adoption of or comparison to this baseline method that sample characterization can be somewhat standardized across the field of post irradiated examination of metal fuels.

Fibrous monolithic ceramics

International Nuclear Information System (INIS)

Kovar, D.; King, B.H.; Trice, R.W.; Halloran, J.W.

1997-01-01

Fibrous monolithic ceramics are an example of a laminate in which a controlled, three-dimensional structure has been introduced on a submillimeter scale. This unique structure allows this all-ceramic material to fail in a nonbrittle manner. Materials have been fabricated and tested with a variety of architectures. The influence on mechanical properties at room temperature and at high temperature of the structure of the constituent phases and the architecture in which they are arranged are discussed. The elastic properties of these materials can be effectively predicted using existing models. These models also can be extended to predict the strength of fibrous monoliths with an arbitrary orientation and architecture. However, the mechanisms that govern the energy absorption capacity of fibrous monoliths are unique, and experimental results do not follow existing models. Energy dissipation occurs through two dominant mechanisms--delamination of the weak interphases and then frictional sliding after cracking occurs. The properties of the constituent phases that maximize energy absorption are discussed. In this article, the authors examine the structure of Si 3 N 4 -BN fibrous monoliths from the submillimeter scale of the crack-deflecting cell-cell boundary features to the nanometer scale of the BN cell boundaries

Ground state of the U{sub 2}Mo compound: Physical properties of the Ω-phase

Energy Technology Data Exchange (ETDEWEB)

Losada, E.L. [SIM3, Centro Atómico Bariloche, Comisión Nacional de Energía Atómica (Argentina); Garcés, J.E., E-mail: garces@cab.cnea.gov.ar [GIA, Centro Atómico Bariloche, Comisión Nacional de Energía Atómica (Argentina)

2016-10-15

Using ab initio calculations, unexpected structural instability was recently found in the ground state of the U{sub 2} Mo compound. Instead of the unstable I4/mmm and the Pmmn structures, in this work the P6/mmm (#191) space group, usually called Ω-phase, is proposed as the fundamental state. Total energy calculations using Wien2k code slightly favoured the last structure. Electronic and elastic properties are studied in this work in order to characterize the physical properties of this new phase. The stability of the Ω-phase is studied by means of its elastic constants calculation and phonon dispersion spectrum. Analysis of isotropic indices shows that the new phase is a ductile material with a minimal degree of anisotropy, suggesting that U{sub 2} Mo in the P6/mmm structure is an elastic isotropic material. Analysis of charge density, density of electronic states (DOS) and the character of the bands revealed a high level of hybridization between d-molybdenum electronic states and d- and f-uranium ones.

The Recovery of Uranium From The Rejected Fuel Plate Dispersion Type of U3O8-Al and U3Si2Al by NaOH

International Nuclear Information System (INIS)

Widodo, G; Aji, D

1998-01-01

The recovery of uranium from the rejected fuel plate dispersion type of U 3 O 8 -AI And U 3 Si 2 -AI with a dissolution has been performed.Each of 5 fragment of fuel plate dispersion of U 3 O 8 -AI or U 3 Si 2 Al of 1x4 cm size was put in the distilled glass content of 250 ml NaOH solution whit The concentration variation 10,15,20,25,and 30%,and than was heated at temperature of 102 o C and was stirred constantly by magnetic stirred.Uranium in the form of U 3 O 8 or U 3 Si 2 was separated by filtration and Either residu and filtrate was analyzed by potentiometry using modified Devies Gray method. From the experiment data it was found in the residu that presentation of uranium was 83.99-84.05% and 84.67-86.556% while in filtrate it was found 53.90 ppm and 69.3 ppm

Interdiffusion between U-Mo alloys and Al or Al alloys at 340 deg. C. Irradiation plan