Irradiation mixing of Al into U3Si

International Nuclear Information System (INIS)

Birtcher, R.C.; Ding, F.R.; Kestel, B.J.; Baldo, P.M.; Zaluzec, N.J.

1995-11-01

Thermal and irradiation induced intermixing of uranium silicide reactor fuels with the aluminum cladding is an important consideration in understanding their fission gas and fuel swelling behavior. The authors have used Rutherford backscattering to follow the behavior of an Al thin film on U 3 Si and U 3 Si 2 during 1.5 MeV Kr ion irradiation at temperatures of 30 and 350 C. After an initial dose during which no intermixing occurs, the Al mixes quickly into U 3 Si. The threshold dose is believed to be associated with an oxide layer between the Al and the uranium silicide. At 300 C and doses greater than threshold, rates of mixing and aluminide phase growth are extracted

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).

Simulation of accident-tolerant U3Si2 fuel using FRAPCON code

International Nuclear Information System (INIS)

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

2017-01-01

The research on accident-tolerant fuels (ATFs) increased after the Fukushima event. This benefited risk management in nuclear operations. In this investigation, the physical properties of the materials being developed for the ATF program were compared with those of the standard UO 2 - Zr fuel system. The research efforts in innovative fuel design include rigorous characterization of thermal, mechanical, and chemical assessment, with the objectives of making the burnup cycle longer, increasing power density, and improving safety performance. Fuels must reach a high uranium density - above that supported by UO 2 - and possess coating that exhibits better oxidation resistance than Zircaloy. The uranium density and thermal conductivity of ATFs, such as U 3 Si 2 , UN, and UC, is higher than that of UO 2 ; their combination with advanced cladding provides possible fuel - cladding options. An ideal combination of fuel and cladding must increase fuel performance in loss-of-coolant scenarios. The disadvantages of U 3 Si 2 , UN, and UC are their swelling rates, which are higher than that of UO 2 . The thermal conductivities of ATFs are approximately four times higher than that of UO2. To prevent the generation of hydrogen due to oxidation of zirconium-based alloys in contact with steam, cladding options, such as ferritic alloys, were studied. It was verified that FeCrAl alloys and SiC provide better response under severe conditions because of their thermophysical properties. The findings of this study indicate that U 3 Si 2 and the FeCrAl fuel cladding concept should replace UO 2 - Zr as the fuel system of choice. (author)

Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

International Nuclear Information System (INIS)

Zhang, Yongfeng

2016-01-01

U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significant progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.

Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

Energy Technology Data Exchange (ETDEWEB)

Zhang, Yongfeng [Idaho National Laboratory

2016-09-01

U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significant progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.

The technique for determination of surface contamination by uranium on U3Si2-Al plate-type fuel elements

International Nuclear Information System (INIS)

Li Shulan; He Fengqi; Wang Qingheng; Han Jingquan

1993-04-01

The NDT method for determining the surface contamination by uranium on U 3 Si 2 -Al plate-type fuel elements, the process of standard specimen preparation and the graduation curve are described. The measurement results of U 3 Si 2 -Al plate-type fuel elements show that the alpha counting method to measure the surface contamination by uranium on fuel plate is more reliable. The UB-1 type surface contamination meter, which was recently developed, has many advantages such as high sensitivity to determine the uranium pollution, short time in measuring, convenience for operation, and the minimum detectable amount of uranium is 5 x 10 -10 g/cm 2 . The measuring device is controlled by a microcomputer. Besides data acquisition and processing, it has functions of statistics, output data on terminal or to printer and alarm. The procedures of measurement are fully automatic. All of these will meet the measuring needs in batch process

State diagram of U-Al-Si as a basis for analysis of the processes in nuclear fuel compositions based on U(Al, Si)3 and U3Si compounds

International Nuclear Information System (INIS)

Chebotarev, N.T.; Konovalov, L.N.; Zhmak, V.A.; Chebotarev, Ya.N.

1996-01-01

Results of studies into the Al-UAl 3 -USi 3 -Si of the U-Al-Si ternary system are presented. It is established that phase equilibrium between the intermetallic compound U(Al, Si) 3 and the aluminium-silicon alloys may be presented in form of conodes on the isothermal cross-section of the state diagram. It is shown that the U(Al, Si) 3 intermetallic compound, containing up to 6.5 at.% silicon, interacts both with liquid and solid aluminium with the U(Al, Si) 4 phase formation [ru

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.

Simulation of accident-tolerant U{sub 3}Si{sub 2} fuel using FRAPCON code

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

The research on accident-tolerant fuels (ATFs) increased after the Fukushima event. This benefited risk management in nuclear operations. In this investigation, the physical properties of the materials being developed for the ATF program were compared with those of the standard UO{sub 2} - Zr fuel system. The research efforts in innovative fuel design include rigorous characterization of thermal, mechanical, and chemical assessment, with the objectives of making the burnup cycle longer, increasing power density, and improving safety performance. Fuels must reach a high uranium density - above that supported by UO{sub 2} - and possess coating that exhibits better oxidation resistance than Zircaloy. The uranium density and thermal conductivity of ATFs, such as U{sub 3}Si{sub 2}, UN, and UC, is higher than that of UO{sub 2}; their combination with advanced cladding provides possible fuel - cladding options. An ideal combination of fuel and cladding must increase fuel performance in loss-of-coolant scenarios. The disadvantages of U{sub 3}Si{sub 2}, UN, and UC are their swelling rates, which are higher than that of UO{sub 2}. The thermal conductivities of ATFs are approximately four times higher than that of UO2. To prevent the generation of hydrogen due to oxidation of zirconium-based alloys in contact with steam, cladding options, such as ferritic alloys, were studied. It was verified that FeCrAl alloys and SiC provide better response under severe conditions because of their thermophysical properties. The findings of this study indicate that U{sub 3}Si{sub 2} and the FeCrAl fuel cladding concept should replace UO{sub 2} - Zr as the fuel system of choice. (author)

Parametric study of the deformation of U3Si2-Al dispersion fuel plates

International Nuclear Information System (INIS)

Vieira, Edeval

2011-01-01

The Nuclear and Energy Research Institute - IPEN-CNEN/SP produces routinely the nuclear fuel necessary for operating its research reactor, IEA-R1. This fuel consists of fuel plates containing U 3 Si 2 -Al composites as the meat, which are fabricated by rolling. The rolling process currently deployed was developed with base on information obtained from literature, which were used as premises for defining the current manufacturing procedures, according to a methodology with essentially empirical character. Despite the current rolling process to be perfectly stable and highly reproducible, it is not well characterized and therefore is not fully known. The objective of this work is to characterize the rolling process for producing fuel plates, specifically the evolution of dimensional parameters of the fuel plate as a function of its deformation in the rolling process. Results are presented in terms of the evolution of the thickness of the fuel meat and cladding of the fuel plate along the deformation, as well as the terminals defects, microstructure and porosity of the fuel meat. (author)

The use of U3Si2 dispersed in aluminum in plate-type fuel elements for research and test reactors

International Nuclear Information System (INIS)

Snelgrove, J.L.; Domagala, R.F.; Hofman, G.L.; Wiencek, T.C.; Copeland, G.L.; Hobbs, R.W.; Senn, R.L.

1987-10-01

A high-density fuel based on U 3 Si 2 dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U 3 Si 2 fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U 3 Si 2 particle swelling rate is approximately the same as that of the commonly used UAl/sub x/ fuel particle. The presence of minor amounts of U 3 Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U 3 Si 2 -aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m 3 is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs

Analysis of transuranic isotopes in irradiated U3Si2-Al fuel by alpha spectrometry

International Nuclear Information System (INIS)

Dian Anggraini; Aslina B Ginting; Arif Nugroho

2011-01-01

Separation and analysis of transuranic isotopes (uranium and plutonium) in irradiated U 3 Si 2 -Al plate has been done. The analysis experiment includes sample preparation (i.e. cutting, dissolving, filtering, dilution), fission products separation from heavy elements, and analysis of transuranic isotopes content with alpha spectrometer. The separation of transuranic isotopes (U, Pu) was done by two methods, i.e. direct method and ion exchanger method with zeolite. Measurement of standard transuranic isotope (AMR 43) and standard U 3 O 8 was done in advance in order to determine percentage of 235 U recovery and detector efficiency. Recovery of 235 U isotope was obtained as much as 92,58%, which fulfills validation requirement, and the detector efficiency was 0.314. Based on the measured recovery and detector efficiency, the separation was done by direct electrodeposition method of 250 µL irradiated U 3 Si 2 -Al solution. The deposited sample was subsequently analyzed with alpha spectrometer. The separation with ion exchanger was done by mixing and shaking of 300 µL irradiated U 3 Si 2 -Al solution and 0.5 gram zeolite to separate the liquid phase from the solid phase. The liquid phase was electrodeposited and analyzed with alpha spectrometer. The analysis of transuranic isotopes (U, Pu) by both methods shows different results. Heavy element ( 238 U, 236 U, 234 U, 239 Pu) content obtained by direct method was 0.0525 g/g and 235 U= 0.0076 g/g, while the separation using zeolite ion exchanger resulted in Heavy element = 0.0253 g/g and 235 U = 0.0092 g/g. (author)

Analysis of hydrogen, carbon, sulfur and volatile compounds in (U3Si2 - Al) nuclear fuel

International Nuclear Information System (INIS)

Moura, Sergio C.; Redigolo, Marcelo M.; Amaral, Priscila O.; Leao, Claudio; Oliveira, Glaucia A.C. de; Bustillos, Oscar V.

2015-01-01

Uranium silicide U 3 Si 2 is used as nuclear fuel in the research nuclear reactor IEA-R1 at IPEN/CNEN, Sao Paulo, Brazil. The U 3 Si 2 is dispersed in aluminum reaching high densities of uranium in the nucleus of the fuel, up to 4.8 gU cm -3 . This nuclear fuel must comply with a quality control, which includes analysis of hydrogen, carbon and sulfur for the U 3 Si 2 and volatile compound for the aluminum. Hydrogen, carbon and sulfur are analyzed by the method of Radio Frequency gas extraction combustion coupled with Infrared detector. Volatile compounds are analyzed by the method of heated gas extraction coupled with gravimetric measurement. These methods are recommended by the American Society for Testing Materials (ASTM) for nuclear materials. The average carbon and sulfur measurements are 30 μg g -1 and 3 μg g -1 , respectively, and 40 μg g -1 for volatile compounds. The hydrogen analyzer is a TCHEN 600 LECO, carbon and sulfur analyzer is a CS 244 LECO and the volatile compounds analyzer is a home-made apparatus that use a resistant furnace, a gas pipe measurement and a glove-box with controlled atmosphere where an analytical balance has been installed, this analyzer was made at IPEN laboratory. (author)

Effect of the fabrication process on fatigue performance of U3Si2 fuel plate with sandwich structure

International Nuclear Information System (INIS)

Wang Xishu; Li Shuangshou; Wang Qingyuan; Xu Yong

2005-01-01

U 3 Si 2 -Al fuel plate is one of the dispersion fuel structure materials recently developed and widely used in research reactors. The mechanical properties of this structural material, especially the fatigue performance, are strongly dependent on its fabrication process. To investigate the effects of these processing technologies, the fatigue tests for the different specimens were carried out. The S-N curves indicate that the fabrication processing technologies of U 3 Si 2 fuel plate, such as the addition of U 3 Si 2 particles into aluminum powder to form the fuel meat, holding and rolling the processes of meat and cladding of 6061-Al alloy, plays an important role in improving the mechanical properties and fatigue performance of this fuel plate. In addition, some factors that influence the crack initiation and propagation are summarized based on the fatigue images that are in situ observations with SEM. The critical criterion for fatigue damage is proposed based on the fatigue data of the structural material, which were obtained at the different conditions

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)

Comparison of thermal compatibility between atomized and comminuted U3Si dispersion fuels

International Nuclear Information System (INIS)

Ryu, Woo-Seog; Park, Jong-Man; Kim, Chang-Kyu; Kuk, II-Hyun

1997-01-01

Thermal compatibility of atomized U 3 Si dispersion fuels were evaluated up to 2600 hours in the temperature range from 250 to 500 degrees C, and compared with that of comminuted U 3 Si. Atomized U 3 Si showed better performance in terms of volume expansion of fuel meats. The reaction zone of U 3 Si and Al occurred along the grain boundaries and deformation bands in U 3 Si particles. Pores around fuel particles appeared at high temperature or after long-term annealing tests to remain diffusion paths over the trench of the pores. The constraint effects of cladding on fuel rod suppressed the fuel meat, and reduced the volume expansion

Preparation of High-Density Uranium-Silicide U3Sl2-Uss: Effects of Preirradiation Heat Treatment on As-Cast Ingot Fuel Plates

International Nuclear Information System (INIS)

Suripto, A; Yuwono

1998-01-01

Heat treatment experiments upon U 3 Si 2 - U ss ingot have been cam e d out to obtain free uranium particle size improvement which is required to enhance the U-Al inter-diffusion reaction in the fuel plate meat. . Heat treatment experiments upon fuel plates containing dispersion of U 3 Si 2 - U ss in Al matrix have also been carried out to study the effect of temperature and treatment duration on the extent of inter-diffusion reaction between free uranium particle and aluminium matrix in the fuel plate meat. Both the experiments indicate that a drastic size improvement has occurred with the U 3 Si 2 as well as free uranium particles upon heat treatment at controlled temperature between the U 3 Si 2 peritectic and peritectoid temperatures and that the inter-diffusion reaction between free uranium and Al matrix occurs quite significantly at temperatures higher than that ordinarily used in the fabrication procedure

Post-irradiation examination of Al-61 wt% U3Si fuel rods from the NRU reactor

International Nuclear Information System (INIS)

Sears, D.F.; Wang, N.

1997-01-01

This paper describes the post-irradiation examination of 4 intact low enrichment uranium (LEU) fuel rods from the national research universal (NRU) reactor at the Chalk River Laboratories of AECL. The rods were irradiated during the period 1993 through 1995, under typical driver fuel operating conditions in NRU, i.e., nominal D 2 O coolant inlet temperature 37E C, inlet pressure 654 kPa and mass flow 12.4 L/s. Irradiation exposures ranged from 147 to 251 full-power days, corresponding to 40 to 84 atom % 235 U burnup. The maximum rod power was ∼2 MW, with element linear power ratings up to 68 kW/m. Post-irradiation examinations, conducted in 1997, focused on optical metallography to measure cladding oxide thickness and fuel core and cladding microstructural examinations. The cladding oxide was approximately 24 : m thick at the mid-plane of fuel rods irradiated to 251 full-power days, with small areas up to 34 : m thick on the fins. The cladding retained significant ductility after irradiation, and its microstructure appeared unchanged. Fuel core diametral increases were small (up to 4%) and within the range previously observed on Al-61 wt% U 3 Si fuel irradiated in the NRU reactor. (author)

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.

In-reactor behaviour of centrifugally atomized U3Si dispersion fuel irradiated at high temperature in HANARO

International Nuclear Information System (INIS)

Kim, Ki Hwan; Park, Jong Man; Yoo, Byeong Ok; Park, Dae Kyu; Lee, Choong Sung; Kim, Chang Kyu

2002-01-01

The irradiation test on full-size U 3 Si dispersion fuel elements, prepared by centrifugal atomization and conventional comminution method, has been performed up to about 77 at.% U-235 in maximum burn-up at CT hole position having the highest power condition in the HANARO reactor, in order to examine the irradiation performance of the atomized U 3 Si for the driver fuels of HANARO. The in-reactor interaction of the atomized U 3 Si dispersion fuel meats is generally assumed to be acceptable with the range of 5-15 μm in average thickness. The atomized spherical particles have more uniform and thinner reaction layer than the comminuted irregular particles. The U 3 Si particles have relatively fine and uniform size distribution of fission gas bubbles, irrespective of the powdering method. The bubble population in the atomized particles appears to be finer and more homogeneous with the characteristics of narrower bubble size distribution than that of the comminuted fuel. The atomized U 3 Si dispersion fuel elements exhibit sound swelling behaviours of 5 % in ΔV/V m even at ∼77 at.% U-235 burn-up, which meets with the safety criterion of the fuel rod, 20vol.% for HANARO. The atomized U3Si dispersion fuel elements show smaller swelling than the comminuted fuel elements

Comparison of thermal compatibility between atomized and comminuted U{sub 3}Si dispersion fuels

Energy Technology Data Exchange (ETDEWEB)

Ryu, Woo-Seog; Park, Jong-Man; Kim, Chang-Kyu; Kuk, II-Hyun [Korea Atomic Research Institute, Taejon (Korea, Republic of)

1997-08-01

Thermal compatibility of atomized U{sub 3}Si dispersion fuels were evaluated up to 2600 hours in the temperature range from 250 to 500{degrees}C, and compared with that of comminuted U{sub 3}Si. Atomized U{sub 3}Si showed better performance in terms of volume expansion of fuel meats. The reaction zone of U{sub 3}Si and Al occurred along the grain boundaries and deformation bands in U{sub 3}Si particles. Pores around fuel particles appeared at high temperature or after long-term annealing tests to remain diffusion paths over the trench of the pores. The constraint effects of cladding on fuel rod suppressed the fuel meat, and reduced the volume expansion.

Examination of U3Si2-Al fuel elements from the Oak Ridge Research Reactor

International Nuclear Information System (INIS)

Copeland, G.L.; Snelgrove, J.L.; Hofman, G.L.

1986-01-01

The results of postirradiation examination of low-enriched U 3 Si 2 fuel elements from the Oak Ridge Research Reactor are presented. The elements replaced standard high-enriched elements and were handled routinely except that the burnup of half the elements was extended beyond normal limits up to about 98% peak. The elements were manufactured by commercial fuel suppliers. The performance was completely satisfactory for all the elements

Unirradiated characteristics of U-Si alloys as dispersed-phase fuels

International Nuclear Information System (INIS)

Domagala, R.F.; Wiencek, T.C.

1987-06-01

To satisfy the power demands of many research reactors, a new LEU fuel with a high density and U content was needed. Any fuel must be compatible with Al and its alloys so that it may be fabricable as a dispersed-phase in Al alloy and Al matrix plate-type elements following, as nearly as possible, established commercial manufacturing techniques. U-Si and U-Si-Al alloys at or near the composition of U 3 Si were immediately attractive because of work documented by the Canadians. 8 refs., 2 figs

Low silicon U(Al,Si)3 stabilization by Zr addition

International Nuclear Information System (INIS)

Pizarro, L.M.; Alonso, P.R.; Rubiolo, G.H.

2009-01-01

Previous knowledge states that (U,Zr)Al 3 and U(Al,Si) 3 phases with Zr and Si content higher than 6 at.% (7.7 wt%) and 4 at.% (1.4 wt%), respectively, does not partially transform to UAl 4 at 600 o C. In this work, four alloys within the quaternary system U-Al-Si-Zr were made with a fixed nominal 0.18 at.% (0.1 wt%) Si content in order to assess the synergetic effect of both Zr and Si alloying elements to the thermodynamic stability of the (U,Zr)(Al,Si) 3 phase. Heat treatments at 600 deg. C were undertaken and samples were analyzed by means of XRD, EPMA and EDS techniques. A remarkable conclusion is that addition of 0.3 at.% Si in the (U,Zr)(Al,Si) 3 phase reduces in 2.7 at.% the necessary Zr content to inhibit its transformation to U(Al,Si) 4 .

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)

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.

Irradiation behaviour of U-Si-Al with peripheral voidage

International Nuclear Information System (INIS)

Fehrenbach, P.J.; Feraday, M.A.; Cotnam, K.D.; Morel, P.A.

1977-12-01

Peripheral voidage in the form of longitudinal slots on the surface of the fuel is effective in accommodating the irradiation induced swelling of U-Si-Al rods clad in cold-worked Zr-2.5 wt% Nb. Diametral increases in elements operated at steady powers between 50 and 80 kW/m to a burnup of 240 MW.h/kg U were less than 1.25%. Similar to U 3 Si elements, lower sheath strength or higher temperatures resulted in greater diametral increases. An oxide or graphite interlayer between the fuel and sheath was shown to be effective in preventing the development of a fuel/sheath bond. The ability of U-Si-Al elements to successfully survive a power increase from 35 to 70 kW/m after appreciable burnup at the lower power has also been demonstrated. (author)

A Prediction Study on Oxidation of Aluminum Alloy Cladding of U{sub 3}Si{sub 2}-Al Fuel Plate

Energy Technology Data Exchange (ETDEWEB)

Tahk, Y.W.; Lee, B.H.; Oh, J.Y.; Park, J.H.; Yim, J.S. [Research Reactor Design and Engineering Div., Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong, Daejeon 305-353 (Korea, Republic of)

2011-07-01

U{sub 3}Si{sub 2}-Al dispersion fuel with aluminum alloy cladding will be used for the Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding undergoes corrosion at slow rates under operational status. This causes thinning of the cladding walls and impairs heat transfer to the coolant. Predictions of the aluminum oxide thickness of the fuel cladding and the maximum temperature difference across the oxide film are needed for reliability evaluation based on the design criteria and limits which prohibit spallation of oxide film. In this work, several oxide thickness prediction models were compared with the measured data of in-pile test results from RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model were performed for JRTR fuel. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, fresh fuel is discharged after 900 effective full power days (EFPD), which is too long a span to predict oxidation properly without an elaborate model. The latest model developed by Kim et al. is in good agreement with the recent in-pile test data as well as with the out-of-pile test data available in the literature, and is one of the best predictors for the oxidation of aluminum alloy cladding in various operating condition. Accordingly, this model was chosen for estimating the oxide film thickness. Through the preliminarily evaluation, water pH level is to be controlled lower than 6.2 for the conservativeness in the case of including the effect of anticipated operational occurrences and the spent fuel residence time in the storage rack after discharging. (author)

Fabrication and microstructural analysis of UN-U_3Si_2 composites for accident tolerant fuel applications

International Nuclear Information System (INIS)

Johnson, Kyle D.; Raftery, Alicia M.; Lopes, Denise Adorno; Wallenius, Janne

2016-01-01

In this study, U_3Si_2 was synthesized via the use of arc-melting and mixed with UN powders, which together were sintered using the SPS method. The study revealed a number of interesting conclusions regarding the stability of the system – namely the formation of a probable but as yet unidentified ternary phase coupled with the reduction of the stoichiometry in the nitride phase – as well as some insights into the mechanics of the sintering process itself. By milling the silicide powders and reducing its particle size ratio compared to UN, it was possible to form a high density UN-U_3Si_2 composite, with desirable microstructural characteristics for accident tolerant fuel applications. - Highlights: • U_3Si_2 fabricated from elemental uranium and silicon through arc melting. • Homogeneity of the silicides assessed through densitometry, XRD, SEM and EDS, chemical etching and optical microscopy. • UN powder fabricated using hydriding-nitriding method. • No phase transformations detected when sintering using silicide particle sizes less than UN particle size. • High density composite (98%TD) fabricated with silicide grain coating using spark plasma sintering at 1450 °C.

Status of LEU fuel development and conversion of NRU

International Nuclear Information System (INIS)

Sears, D.F.; Herbert, L.N.; Vaillancourt, K.D.

1989-11-01

The status of the low-enrichment uranium (LEU) fuel development and NRU conversion program at Chalk River Nuclear Laboratories is reviewed. Construction of a new fuel fabrication facility is essentially completed and installation of LEW fuel manufacturing equipment has begun. The irradiation of 31 prototype Al-61 wt% U 3 Si dispersion fuel rods, approximately one third of a full NRU core, is continuing without incident. Recent post-irradiation examination of spent fuel rods revealed that the prototype LEU fuel achieved the design burnup (80 at%) in excellent condition, confirming that the Al-U 3 Si 2 dispersion fuel to complement out Al-U 3 Si capability. Three full-size NRU rods containing Al-U 3 Si 2 dispersion fuel have been fabricated for a qualification irradiation in NRU. Post-irradiation examinations of mini-elements containing Al-U 3 Si 2 fuel revealed that the U 3 Si 2 behaved similarly to U 3 Si 2 fuel revealed that the U 3 Si 2 particles and the aluminum matrix, and fission gas bubbles up to 10 μm in diameter, could be seen in the particles after 60 at% and 80 at% burnup. The mini-elements contained a variety of silicide particle sizes; however, no significant swelling dependence on particle size distribution was observed

CERCA's 25 years experience in U3Si2 fuel manufacturing

International Nuclear Information System (INIS)

Durand, JP.; Duban, B.; Lavastre, Y.; Perthuis, S. de

2003-01-01

This paper documents the experience gained at CERCA in manufacturing, testing, and inspecting U 3 Si 2 fuel elements for various Material Test Reactors (MTR) since the beginning of the RERTR Program in 1978, up to now. It emphasises how the company controls the product to insure compliance with the fuel-related safety parameters. Finally, those statements are considered in the UMo fuel production perspective. (author)

Application of non-destructive methods for qualification of the U3O8-Al and U3Si2-Al dispersion fuels in the IEA-R1 Reactor

International Nuclear Information System (INIS)

Silva, Jose Eduardo Rosa da

2011-01-01

IPEN/CNEN-SP manufactures fuels to be used in its nuclear research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil doesn't have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds, internationally tested and qualified to be used in research reactors, and has gotten experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans to increase the uranium density of these fuels. The objective of this thesis work was to study and to propose a set of non-destructive methods to qualify the dispersions fuels U 3 O 8 -Al e U 3 Si 2 -Al with high uranium density produced at IPEN/CNEN-SP. For that, the irradiation resources in the IEA-R1, and the application of non-destructive methods in the reactor pool available in the Institution were considered. The proposal is to specify, manufacture and irradiate fuel mini plates in IEA-R1 at the maximum densities, qualified internationally, and to monitor their general conditions during the period of irradiation, using non-destructive methods in the reactor pool. In addition to the non-destructive visual inspection and sipping methods, already used at the Institution, the infrastructure for dimensional sub-aquatic testing to evaluate the swelling of irradiated fuel mini plates was completed. The analyses of the results will provide means to assess and decide whether or not to continue with the irradiation of mini plates, until the desired burnup for the irradiation tests at IEA-R1 are reached. (author)

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

Atomization of U3Si2/U3Si for research reactor fuel

International Nuclear Information System (INIS)

Kuk, Il Hiun

2004-01-01

Instead of comminuting, U 3 Si 2 /U 3 Si powders are produced by atomizating directly from the molten alloys. Many benefits are introduced by applying the atomization technique: reduction of the process, homogeneous alloy composition within a particle and between particles, increase of the thermal conductivity and decrease of the chemical reactivity with aluminium due to particle's spherical shape. (author)

Prospect of Uranium Silicide fuel element with hypostoichiometric (Si ≤3.7%)

International Nuclear Information System (INIS)

Suripto, A.; Sardjono; Martoyo

1996-01-01

An attempt to obtain high uranium-loading in silicide dispersion fuel element using the fabrication technology applicable nowadays can reach Uranium-loading slightly above 5 gU/cm 3 . It is difficult to achieve a higher uranium-loading than that because of fabricability constraints. To overcome those difficulties, the use of uranium silicide U 3 Si based is considered. The excess of U is obtained by synthesising U 3 Si 2 in Si-hypostoichiometric stage, without applying heat treatment to the ingot as it can generate undesired U 3 Si. The U U will react with the matrix to form U al x compound, that its pressure is tolerable. This experiment is to consider possibilities of employing the U 3 Si 2 as nuclear fuel element which have been performed by synthesising U 3 Si 2 -U with the composition of 3.7 % weigh and 3 % weigh U. The ingot was obtained and converted into powder form which then was fabricated into experimental plate nuclear fuel element. The interaction between free U and Al-matrix during heat-treatment is the rolling phase of the fuel element was observed. The study of the next phase will be conducted later

COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

OpenAIRE

ALEKSEY. L. IZHUTOV; VALERIY. V. IAKOVLEV; ANDREY. E. NOVOSELOV; VLADIMIR. A. STARKOV; ALEKSEY. A. SHELDYAKOV; VALERIY. YU. SHISHIN; VLADIMIR. M. KOSENKOV; ALEKSANDR. V. VATULIN; IRINA. V. DOBRIKOVA; VLADIMIR. B. SUPRUN; GENNADIY. V. KULAKOV

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%235U; th...

Uncertainty measurement evaluation of WDXRF and EDXRF techniques for the Si and U{sub total} determination in U{sub 3}Si{sub 2} used as nuclear fuel material

Energy Technology Data Exchange (ETDEWEB)

Scapin, Marcos A.; Salvador, Vera L.R.; Cotrim, Marycel E.B.; Pires, Maria Ap. F.; Sato, Ivone M., E-mail: mascapin@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) Sao Paulo, SP (Brazil). Centro de Quimica e Meio Ambiente

2009-07-01

Uranium silicide (U{sub 3}Si{sub 2}), 20% {sup 235}U enriched powder, is an intermetallic compound used as nuclear fuel material; that is the state-of-the-art among nuclear fuel materials used in modern research reactors. It is produced by IPEN and used as nuclear fuel of the IEA-R1 reactor (IPEN/CNEN, Sao Paulo, Brazil); U{sub 3}Si{sub 2} has 92.3%wtU and 7.7%wtSi. The qualification of this material requires chemical and physical tests such as Si and U{sub total} content, isotope ratio, impurities, density, specific surface area and particle size determination. The Si and U{sub total} determination were made by gravimetric and volumetric procedures at the Environment Chemistry Center (CQMA-IPEN/CNEN). Usually, these classical methods require a long time for analyses and are expensive. The objective of this study was to establish a fast and efficient analytical method to meet ISO/IEC 17025:2005 requirements in the Si and U{sub total} determination. The X-ray fluorescence techniques (XRF) were chosen to allow a direct and non-destructive testing, what is a principal advantage faced to other instrumental techniques, since previous chemical treatments are not necessary. In this study, the performance of the wavelength dispersive (WDXRF) and energy dispersive (EDXRF) X- ray fluorescence techniques was evaluated. Furthermore, two different sample preparation procedures, plain powdered and pressed powdered were evaluated. Statistical tools were used to evaluate the results and a comparison between these results and the conventional methods was done. (author)

Synthesis of Vertically-Aligned Carbon Nanotubes from Langmuir-Blodgett Films Deposited Fe Nanoparticles on Al2O3/Al/SiO2/Si Substrate.

Science.gov (United States)

Takagiwa, Shota; Kanasugi, Osamu; Nakamura, Kentaro; Kushida, Masahito

2016-04-01

In order to apply vertically-aligned carbon nanotubes (VA-CNTs) to a new Pt supporting material of polymer electrolyte fuel cell (PEFC), number density and outer diameter of CNTs must be controlled independently. So, we employed Langmuir-Blodgett (LB) technique for depositing CNT growth catalysts. A Fe nanoparticle (NP) was used as a CNT growth catalyst. In this study, we tried to thicken VA-CNT carpet height and inhibit thermal aggregation of Fe NPs by using Al2O3/Al/SiO2/Si substrate. Fe NP LB films were deposited on three typed of substrates, SiO2/Si, as-deposited Al2O3/Al/SiO2/Si and annealed Al2O3/Al/SiO2/Si at 923 K in Ar atmosphere of 16 Pa. It is known that Al2O3/Al catalyzes hydrocarbon reforming, inhibits thermal aggregation of CNT growth catalysts and reduces CNT growth catalysts. It was found that annealed Al2O3/Al/SiO2/Si exerted three effects more strongly than as-deposited Al2O3/Al/SiO2/Si. VA-CNTs were synthesized from Fe NPs-C16 LB films by thermal chemical vapor deposition (CVD) method. As a result, at the distance between two nearest CNTs 28 nm or less, VA-CNT carpet height on annealed Al2O3/Al/SiO2/Si was about twice and ten times thicker than that on SiO2/Si and that on as-deposited Al2O3/Al/SiO2/Si, respectively. Moreover, distribution of CNT outer diameter on annealed Al2O3/Al/SiO2/Si was inhibited compared to that on SiO2/Si. These results suggest that since thermal aggregation of Fe NPs is inhibited, catalyst activity increases and distribution of Fe NP size is inhibited.

Manufacture of hypoeutectic Al-Si metal powders for dispersion matriz in nuclear fuels

International Nuclear Information System (INIS)

Raffaeli, H A; Harri, S; Acosta, M; Castillo Guerra, R; Rossi, G; Fabro, J O; Rubiolo, G H

2012-01-01

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 optimized. 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. The present work deals with three techniques applied to produce metal powders of hypoeutectic Al-Si alloys: ball milling, centrifugal atomization and gas atomization. Size and microstructure of the particles are analyzed in the three techniques. The best result is found with the gas atomization system, flakes and rods morphology predominates in the produced powders, with particle sizes below 150 microns and the greater mass population (65%) is between 150 and 125 microns. The particle surface is smooth and the high solidification rate provides a good distribution of the α-Al primary and eutectic phase within each particle (author)

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)

U3Si2 Fabrication and Testing for Implementation into the BISON Fuel Performance Code

Energy Technology Data Exchange (ETDEWEB)

Knight, Travis W.

2018-04-23

A creep test stand was designed and constructed for compressive creep testing of U3Si2 pellets. This is described in Chapter 3.

• Creep testing of U3Si2 pellets was completed. In total, 13 compressive creep tests of U3Si2 pellets was successfully completed. This is reported in Chapter 3.
• Secondary creep model of U3Si2 was developed and implemented in BISON. This is described in Chapter 4.
• Properties of U3Si2 were implemented in BISON. This is described in Chapter 4.
• A resonant frequency and damping analyzer (RFDA) using impulse excitation technique (IET) was setup, tested, and used to analyze U3Si2 samples to measure Young’s and Shear Moduli which were then used to calculate the Poisson ratio for U3Si2. This is described in Chapter 5.
• Characterization of U3Si2 samples was completed. Samples were prepared and analyzed by XRD, SEM, and optical microscopy. Grain size analysis was conducted on images.

Fabrication and microstructural analysis of UN-U{sub 3}Si{sub 2} composites for accident tolerant fuel applications

Energy Technology Data Exchange (ETDEWEB)

Johnson, Kyle D., E-mail: kylej@kth.se; Raftery, Alicia M.; Lopes, Denise Adorno; Wallenius, Janne

2016-08-15

In this study, U{sub 3}Si{sub 2} was synthesized via the use of arc-melting and mixed with UN powders, which together were sintered using the SPS method. The study revealed a number of interesting conclusions regarding the stability of the system – namely the formation of a probable but as yet unidentified ternary phase coupled with the reduction of the stoichiometry in the nitride phase – as well as some insights into the mechanics of the sintering process itself. By milling the silicide powders and reducing its particle size ratio compared to UN, it was possible to form a high density UN-U{sub 3}Si{sub 2} composite, with desirable microstructural characteristics for accident tolerant fuel applications. - Highlights: • U{sub 3}Si{sub 2} fabricated from elemental uranium and silicon through arc melting. • Homogeneity of the silicides assessed through densitometry, XRD, SEM and EDS, chemical etching and optical microscopy. • UN powder fabricated using hydriding-nitriding method. • No phase transformations detected when sintering using silicide particle sizes less than UN particle size. • High density composite (98%TD) fabricated with silicide grain coating using spark plasma sintering at 1450 °C.

Fabrication and thermophysical property characterization of UN/U3Si2 composite fuel forms

Science.gov (United States)

White, J. T.; Travis, A. W.; Dunwoody, J. T.; Nelson, A. T.

2017-11-01

High uranium density composite fuels composed of UN and U3Si2 have been fabricated using a liquid phase sintering route at temperatures between 1873 K and 1973 K and spanning compositions of 10 vol% to 40 vol% U3Si2. Microstructural analysis and phase characterization revealed the formation of an U-Si-N phase of unknown structure. Microcracking was observed in the U-Si portion of the composite microstructure that likely originates from the mismatched coefficient of thermal expansion between the UN and U3Si2 leading to stresses on heating and cooling of the composite. Thermal expansion coefficient, thermal diffusivity, and thermal conductivity were characterized for each of the compositions as a function of temperature to 1673 K. Hysteresis is observed in the thermal diffusivity for the 20 vol% through 40 vol% specimens between room temperature and 1273 K, which is attributed to the microcracking in the U-Si phase. Thermal conductivity of the composites was modeled using the MOOSE framework based on the collected microstructure data. The impact of irradiation on thermal conductivity was also simulated for this class of composite materials.

Reactivity feedback coefficients of a material test research reactor fueled with high-density U{sub 3}Si{sub 2} dispersion fuels

Energy Technology Data Exchange (ETDEWEB)

Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)], E-mail: farhan73@hotmail.com; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)

2008-10-15

The reactivity feedback coefficients of a material test research reactor fueled with high-density U{sub 3}Si{sub 2} dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U{sub 3}Si{sub 2} LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 deg. C to 100 deg. C, at the beginning of life, followed the relationships (in units of {delta}k/k x 10{sup -5} K{sup -1}) -2.116 - 0.118 {rho}{sub U}, 0.713 - 37.309/{rho}{sub U} and -12.765 - 34.309/{rho}{sub U}, respectively for 4.0 {<=} {rho}{sub U} (g/cm{sup 3}) {<=} 6.0.

PEMBUATAN SUMBER RADIASI GAMMA 137Cs DENGAN AKTIVITAS 20 mCi DARI PEB U3Si2-Al PASCA IRADIASI DALAM CONTAINER STAINLESS STEEL

Directory of Open Access Journals (Sweden)

Aslina Br. Ginting

2016-03-01

Full Text Available ABSTRAK PEMBUATAN SUMBER RADIASI GAMMA ISOTOP 137Cs DENGAN AKTIVITAS 20 mCi DARI PEB U3Si2-Al PASCA IRADIASI DALAM CONTAINER STAINLESS STEEL. Kegiatan uji pasca iradiasi pelat elemen bakar (PEB U3Si2-Al banyak menghasilkan larutan dengan keaktifan yang sangat tinggi. Larutan tersebut mengandung isotop 137Cs, uranium serta transuranium yang mempunyai waktu paroh panjang dan berbahaya bagi lingkungan. Namun larutan tersebut memiliki nilai ekonomis tinggi karena dapat dimanfaatkan sebagai bahan baku untuk pembuatan sumber radiasi sinar gamma isotop 137Cs. Hal ini dapat membantu bidang industri dalam memenuhi kebutuhan sumber radioaktif dalam negeri karena selama ini kebutuhan isotop 137Cs di Indonesia masih tergantung dari industri luar negeri. Selain itu, pengadaan dan transportasi isotop 137Cs dari luar negeri serta dalam penggunaannya memerlukan persyaratan yang cukup ketat dari Badan Pengawas Tenaga Nuklir Nasional (BAPETEN, sehingga menyebabkan harga isotop 137Cs menjadi mahal sampai di Indonesia. Dengan alasan tersebut, BATAN sebagai lembaga litbang nuklir di Indonesia perlu mempelajari pembuatan sumber radiasi gamma isotop 137Cs dari larutan hasil pengujian bahan bakar nuklir U3Si2-Al pasca iradiasi. Manfaat isotop 137Cs sangat luas antara lain digunakan dalam menganalisis sampel lingkungan, industri migas, konstruksi, radiografi, perikanan, rumah sakit dan pertambangan. Pembuatan sumber radiasi gamma isotop 137Cs dimulai dari pengumpulan larutan hasil pengujian PEB U3Si2-Al. Larutan larutan hasil pengujian mengandung isotop 137Cs dan isotop lainnya dikumpulkan menjadi satu dalam botol dengan volume 65 mL. Pemisahan isotop 137Cs dari hasil fisi lainnya dilakukan dengan metode penukar kation menggunakan zeolit Lampung dengan berat 45 gr. Hasil pemisahan diperoleh 137Cs-zeolit dalam fasa padat dan isotop lainnya berada dalam fasa cair. Padatan137Cs-zeolit kering kemudian kemudian ditimbang dan diukur aktivitasnya menggunakan spektrometer

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

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%

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.

Mixing of Al into uranium silicides reactor fuels

International Nuclear Information System (INIS)

Ding, F.R.; Birtcher, R.C.; Kestel, B.J.; Baldo, P.M.

1996-11-01

SEM observations have shown that irradiation induced interaction of the aluminum cladding with uranium silicide reactor fuels strongly affects both fission gas and fuel swelling behaviors during fuel burn-up. The authors have used ion beam mixing, by 1.5 MeV Kr, to study this phenomena. RBS and the 27 Al(p, γ) 28 Si resonance nuclear reaction were used to measure radiation induced mixing of Al into U 3 Si and U 3 Si 2 after irradiation at 300 C. Initially U mixes into the Al layer and Al mixes into the U 3 Si. At a low dose, the Al layer is converted into UAl 4 type compound while near the interface the phase U(Al .93 Si .07 ) 3 grows. Under irradiation, Al diffuses out of the UAl 4 surface layer, and the lower density ternary, which is stable under irradiation, is the final product. Al mixing into U 3 Si 2 is slower than in U 3 Si, but after high dose irradiation the Al concentration extends much farther into the bulk. In both systems Al mixing and diffusion is controlled by phase formation and growth. The Al mixing rates into the two alloys are similar to that of Al into pure uranium where similar aluminide phases are formed

Effect of temperature on the expansion and microstructure Of U3 Si2-AI mini plate fuel of 3.6 g/cm3 uranium loading

International Nuclear Information System (INIS)

Ginting, A. Br.; Samosir, N.; Suparjo; Nasution, H.

2000-01-01

Expansion analysis has been conducted to 50 x 20-mm U 3 Si 2 -AI mini plate of 3.6 g/cm 3 uranium loading using dilatometer. The analysis was carried out at various temperatures of 170 o C, 350 o C and 550 o C in Argon medium with delay time 4 days. The result showed that the fuel plate was relatively stable with increasing of heating time but underwent significant expansion. Heating at 170 o C, 350 o C and 550 o C resulted in the expansion of the U 3 Si 2 -AI fuel plate of to 83-212 mum, 333-475 mum, and 433-724 mum with coefficient expansion of 24.2x10 -6 / o C - 24.3x10 -6 / o C, 25.5x10 -6 / o C - 26.2x10 -6 /'oC and 26.6 x 10 -6 / o C - 28.2 x 10 -6 / o C respectively. Microanalysis of the U 3 Si 2 -AI mini plate fuel with SEM-EDS upon heating at those temperature variation showed that microstructure change didn't occur at 170 o C, mean while interaction between AIMg2 cladding and the fuel meat appeared to take place at 350 o C and 550 o C. Data on the expansion and microstructure change of U 3 Si 2 -AI fuel plate upon heating are of great important for the manufacture/fabrication of research fuel plate to produce silicide fuel element for higher uranium loading. (author)

Radiation-induced grain subdivision and bubble formation in U3Si2 at LWR temperature

Science.gov (United States)

Yao, Tiankai; Gong, Bowen; He, Lingfeng; Harp, Jason; Tonks, Michael; Lian, Jie

2018-01-01

U3Si2, an advanced fuel form proposed for light water reactors (LWRs), has excellent thermal conductivity and a high fissile element density. However, limited understanding of the radiation performance and fission gas behavior of U3Si2 is available at LWR conditions. This study explores the irradiation behavior of U3Si2 by 300 keV Xe+ ion beam bombardment combining with in-situ transmission electron microscopy (TEM) observation. The crystal structure of U3Si2 is stable against radiation-induced amorphization at 350 °C even up to a very high dose of 64 displacements per atom (dpa). Grain subdivision of U3Si2 occurs at a relatively low dose of 0.8 dpa and continues to above 48 dpa, leading to the formation of high-density nanoparticles. Nano-sized Xe gas bubbles prevail at a dose of 24 dpa, and Xe bubble coalescence was identified with the increase of irradiation dose. The volumetric swelling resulting from Xe gas bubble formation and coalescence was estimated with respect to radiation dose, and a 2.2% volumetric swelling was observed for U3Si2 irradiated at 64 dpa. Due to extremely high susceptibility to oxidation, the nano-sized U3Si2 grains upon radiation-induced grain subdivision were oxidized to nanocrystalline UO2 in a high vacuum chamber for TEM observation, eventually leading to the formation of UO2 nanocrystallites stable up to 80 dpa.

Interaction of Al2O3xSiO2 alloyed uranium oxide with pyrocarbon coating of fuel particles under irradiation

International Nuclear Information System (INIS)

Chernikov, A.S.; Khromov, Yu.F.; Svistunov, D.E.; Chujko, E.E.

1989-01-01

Method of comparative data analysis for P O2 and P CO was used to consider interaction in fuel particle between pyrocarbon coating and fuel sample, alloyed with alumosilicate addition. Equations of interaction reactions for the case of hermetic and depressurized fuel particle are presented. Calculations of required xAl 2 O 3 XySiO 2 content, depending on oxide fuel burnup, were conducted. It was suggested to use silicon carbide for limitation of the upper level of CO pressure in fuel particle. Estimation of thermal stability of alumosilicates under conditions of uranium oxide burnup equals 1100 and 1500 deg C for Al/Si ratio in addition 1/1 and 4/1 respectively

A contribution to the kinetic study of the metatectic reaction U+U{sub 3}Si{sub 2}{yields}U{sub 3}Si; Contribucion al Estudio cinetico de la Reaccion Metatectica U+U{sub 3}Si{sub 2} U{sub 3}Si

Energy Technology Data Exchange (ETDEWEB)

Arroyo Ruiperez, J; Esteban Hernandez, J A

1962-07-01

An experimental study has been made to decide upon the advantages and drawbacks of the different methods and reagents employed in the metallography of U-Si alloys. It has been observed that all samples thermally treated to form the epsilon-phase undergo from the beginning a coalescence of the U{sub 3}Si{sub 2} particles, which makes practically useless any fine state of dispersion that might be present originally, as recommended by some authors. The coalescence of the U{sub 3}Si{sub 2} particle decreases the surface available for reaction and consequently the reaction rate. (Author) 7 refs.

Radiation Resistance of the U(Al, Si)3 Alloy: Ion-Induced Disordering

Science.gov (United States)

Yaniv, Gili; Horak, Pavel; Vacik, Jiri; Mykytenko, Natalia; Rafailov, Gennady; Dahan, Itzchak; Fuks, David; Kiv, Arik

2018-01-01

During the exploitation of nuclear reactors, various U-Al based ternary intermetallides are formed in the fuel-cladding interaction layer. Structure and physical properties of these intermetallides determine the radiation resistance of cladding and, ultimately, the reliability and lifetime of the nuclear reactor. In current research, U(Al, Si)3 composition was studied as a potential constituent of an interaction layer. Phase content of the alloy of an interest was ordered U(Al, Si)3, structure of which was reported earlier, and pure Al (constituting less than 20 vol % of the alloy). This alloy was investigated prior and after the irradiation performed by Ar ions at 30 keV. The irradiation was performed on the transmission electron microscopy (TEM, JEOL, Japan) samples, characterized before and after the irradiation process. Irradiation induced disorder accompanied by stress relief. Furthermore, it was found that there is a dose threshold for disordering of the crystalline matter in the irradiated region. Irradiation at doses equal or higher than this threshold resulted in almost solely disordered phase. Using the program “Stopping and Range of Ions in Matter” (SRIM), the parameters of penetration of Ar ions into the irradiated samples were estimated. Based on these estimations, the dose threshold for ion-induced disordering of the studied material was assessed. PMID:29393870

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

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.

Attempt to produce silicide fuel elements in Indonesia

International Nuclear Information System (INIS)

Soentono, S.; Suripto, A.

1991-01-01

After the successful experiment to produce U 3 Si 2 powder and U 3 Si 2 -Al fuel plates using depleted U and Si of semiconductor quality, silicide fuel was synthesized using x -Al available at the Fuel Element Production Installation (FEPI) at Serpong, Indonesia. Two full-size U 3 Si 2 -Al fuel elements, having similar specifications to the ones of U 3 O 8 -Al for the RSG-GAS (formerly known as MPR-30), have been produced at the FEPI. All quality controls required have been imposed to the feeds, intermediate, as well as final products throughout the production processes of the two fuel elements. The current results show that these fuel elements are qualified from fabrication point of view, therefore it is expected that they will be permitted to be tested in the RSG-GAS, sometime by the end of 1989, for normal (∝50%) and above normal burn-up. (orig.)

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

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.

Optimization of neutronic characteristics of U3Si2 low enrichment fuel elements for a new design of IEA-R1 reactor core

International Nuclear Information System (INIS)

Mai, L.A.; Maiorino, J.R.; Gouvea, E.A.

1989-01-01

This work shows a study of neutronic optimization of U 3 Si 2 -Al low enrichment fuel element. This study has a goal to propose a optimized Core to be used in the research reactor IEA-R1. The external dimensions of the fuel element were maintained as constraints and the loss of reactivity along fuel life-time was defined as 'objective function', and it has been minimized by varying the fuel element dimensions. Cell calculations were made with HAMMER-TECH /3/ Code, for burnups up to 50% of U-235 initial mass. The Computer values of the objective function for several combinations of fuel element dimensions were fitted by a surface using the SAS system /9/, and it has been minimized by a Harwell subroutine /10/. (author) [pt

Production of Al2O3–SiC nano-composites by spark plasma sintering

Directory of Open Access Journals (Sweden)

Mansour Razavi

2017-07-01

Full Text Available In this paper, Al2O3–SiC composites were produced by SPS at temperatures of 1600 °C for 10 min under vacuum atmosphere. For preparing samples, Al2O3 with the second phase including of micro and nano-sized SiC powder were milled for 5 h. The milled powders were sintered in a SPS machine. After sintering process, phase studies, densification and mechanical properties of Al2O3–SiC composites were examined. Results showed that the specimens containing micro-sized SiC have an important effect on bulk density, hardness and strength. The highest relative density, hardness and strength were 99.7%, 324.6 HV and 2329 MPa, respectively, in Al2O3–20 wt% SiCmicro composite. Due to short time sintering, the growth was limited and grains still remained in nano-meter scale.

Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

International Nuclear Information System (INIS)

Peacock, H.B.

1983-08-01

Heating tests using 53 wt % U 3 O 8 -Al pellets show that an exothermic reaction occurs between 875 and 1000 0 C and takes 10 to 20 seconds to reach maximum temperature. The maximum temperature is a function of particle size of the U 3 O 8 with large particles exhibiting lower peak temperatures. The calculated energy release was 123 cal/g of U 3 O 8 -aluminum fuel. Tests using aluminum clad outer fuel tube sections gave lower peak temperatures than for pellets. No violent reactions occurred. The results are reasonably consistent with recent reported data indicating that the exothermic U 3 O 8 -Al reaction is not an important energy source. The compressive and tensile strengths of U 3 O 8 tubes above 660 0 C are low. In compression, sections with 2 psi average axial stress failed at 917 0 C, while sections with 7 psi failed at 669 0 C. Tubes with U-Al alloy cores failed at about 670 0 C with no applied load. The stresses in fuel tubes during a reactor transient may range up to several hundred psi and are less than 7 psi only in the upper part of the fuel tube

Fabrication, fabrication control and in-core follow up of 4 LEU leader fuel elements based on U3Si2 in RECH-1

International Nuclear Information System (INIS)

Chavez, J.C.; Barrera, M.; Olivares, L.; Lisboa, J.

1999-01-01

The RECH-1 MTR reactor has been converted from HEU to MEU (45% enrichment) and the decision to a LEU (20% enrichment) conversion was taken some years ago. This LEU conversion decision involved a local fuel development and fabrication based on U 3 Si 2 -Al dispersion fuel, and a fabrication qualification stage that resulted in four fuel elements fully complying with established fabrication standards for this type of fuel. This report-presents relevant points of these four leaders fuel elements fabrication, in particular a fuel plate core homogeneity control development. A summary of the intended in core follow-up studies for the leaders fuel elements is also presented here. (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.

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.

Synthesis, microstructure and magnetic properties of Fe{sub 3}Si{sub 0.7}Al{sub 0.3}@SiO{sub 2} core–shell particles and Fe{sub 3}Si/Al{sub 2}O{sub 3} soft magnetic composite core

Energy Technology Data Exchange (ETDEWEB)

Wang, Jian, E-mail: snove418562@163.com [The State Key Laboratory of Refractories and Metallurgy, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Key Laboratory for Ferrous Metallurgy and Resources Utilization of Ministry of Education, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Fan, Xi’an, E-mail: groupfxa@163.com [The State Key Laboratory of Refractories and Metallurgy, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Key Laboratory for Ferrous Metallurgy and Resources Utilization of Ministry of Education, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Wu, Zhaoyang, E-mail: wustwuzhaoyang@163.com [The State Key Laboratory of Refractories and Metallurgy, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Key Laboratory for Ferrous Metallurgy and Resources Utilization of Ministry of Education, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Li, Guangqiang [The State Key Laboratory of Refractories and Metallurgy, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Key Laboratory for Ferrous Metallurgy and Resources Utilization of Ministry of Education, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China)

2015-11-15

Fe{sub 3}Si{sub 0.7}Al{sub 0.3}@SiO{sub 2} core–shell particles and Fe{sub 3}Si/Al{sub 2}O{sub 3} soft magnetic composite core have been synthesized via a modified stöber method combined with following high temperature sintering process. Most of conductive Fe{sub 3}Si{sub 0.7}Al{sub 0.3} particles could be uniformly coated by insulating SiO{sub 2} using the modified stöber method. The Fe{sub 3}Si{sub 0.7}Al{sub 0.3}@SiO{sub 2} core–shell particles exhibited good soft magnetic properties with low coercivity and high saturation magnetization. The reaction 4Al+3SiO{sub 2}=2α-Al{sub 2}O{sub 3}+3Si took place during the sintering process. As a result the new Fe{sub 3}Si/Al{sub 2}O{sub 3} composite was formed. The Fe{sub 3}Si/Al{sub 2}O{sub 3} composite core displayed more excellent soft magnetic properties, better frequency stability at high frequencies, much higher electrical resistivity and lower core loss than the pure Fe{sub 3}Si{sub 0.7}Al{sub 0.3} core. The method of introducing insulating layers surrounding magnetic particles provides a promising route to develop new and high compact soft magnetic materials with good magnetic and electric properties. - Graphical abstract: In Fe{sub 3}Si/Al{sub 2}O{sub 3} composite, Fe{sub 3}Si phases are separated by Al{sub 2}O{sub 3} layers and the eddy currents are confined in Fe{sub 3}Si phases, thus increasing resistivity and reducing core loss. - Highlights: • Fe{sub 3}Si{sub 0.7}Al{sub 0.3}@SiO{sub 2} core–shell particles and Fe{sub 3}Si/Al{sub 2}O{sub 3} cores were prepared. • Fe{sub 3}Si{sub 0.7}Al{sub 0.3} particles could be uniformly coated by nano-sized SiO{sub 2} clusters. • Fe{sub 3}Si{sub 0.7}Al{sub 0.3}@SiO{sub 2} particles and Fe{sub 3}Si/Al{sub 2}O{sub 3} cores showed good soft magnetic properties. • Fe{sub 3}Si/Al{sub 2}O{sub 3} had lower core loss and better frequency stability than Fe{sub 3}Si{sub 0.7}Al{sub 0.3} cores.

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)

THE THERMODYNAMIC PROPERTIES OF MELTS OF DOUBLE SYSTEM MgO – Al2O3, MgO – SiO2, MgO – CaF2, Al2O3 – SiO2, Al2O3 – CaF2, SiO2 – CaF2

Directory of Open Access Journals (Sweden)

В. Судавцова

2012-04-01

Full Text Available Methodology of prognostication of thermodynamics properties of melts is presented from the coordinatesof liquidus of diagram of the state in area of equilibria a hard component is solution, on which energies ofmixing of Gibbs are expected in the double border systems of MgO – Al2O3, MgO – SiO2, MgO – CaF2,Al2O3 – SiO2, Al2O3 - CaF2, SiO2 - CaF2. For the areas of equilibrium there is quasibinary connection(MgAl2O4, Mg2SiO4, Al6Si2O13 – a grout at calculations was used equalization of Hauffe-Wagner. Theobtained data comport with literary

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.

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.

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.

Structural and microstructural characterization of U3Si2 nuclear fuel using X-ray diffraction

International Nuclear Information System (INIS)

Ichikawa, Rodrigo U.; Garcia, Rafael H.L.; Silva, Andre S.B. da; Saliba-Silva, Adonis M.; Lima, Nelson B.; Martinez, Luis G.; Turrillas, Xavier

2017-01-01

In this work, two uranium silicide powdered samples, containing 67% and 42 mol% of Si, were analyzed using X-ray diffraction (named as 67 Si and 42 Si). For structural characterization, Rietveld refinement was used to estimate cell parameters, volume fraction (weight percent) of crystalline phases and atomic positions. For the main phases, X-ray line profile analysis (XLPA) was used to estimate mean crystallite sizes and micro strains. The 67 Si sample presents higher content of USi 2( tetragonal) and the 42 Si sample presents higher content of U 3 Si 2 (tetragonal) as identified and calculated from the XRD profiles. Overall there are no appreciable structural changes and the parameters refined are in good accordance with the ones reported in the literature. Mean crystallite sizes determined by XLPA revealed small crystallites of the order of 10 1 nm and low micro strain for all samples. (author)

Irradiation behaviour of solid and hollow U{sub 3}Si fuel elements: results to 15,000 MWd/tonne U

Energy Technology Data Exchange (ETDEWEB)

Feraday, M A; Chalder, G H; Cotnam, K D

1969-06-15

U{sub 3}Si fuel elements clad in zirconium alloy sheaths have been irradiated to burnups close to 15,000 MWd/tonne U in pressurized water at 220{sup o}C, 98 bars. The results show that the external swelling can be controlled by incorporating free volume in the element. The dimensional stability of such elements is adequate to permit their use in power reactor fuel bundles. A diameter increase of 1.2% had occurred in an element initially containing 12.8% total free volume, after a burnup of 14,700 MWd/tonne U. There was no change in diameter between burnups of 5200 and 14,700 MWd/tonne U. Elements containing 3% total free volume had increased in diameter about 2.5% at 2000 MWd/tonne U compared to 0.2% at 9500 MWd/tonne U for elements containing 22% total free volume. The observed swelling in the U{sub 3}Si is discussed in terms of possible mechanisms. (author)

Reduced enrichment fuels for Canadian research reactors - Fabrication and performance

Energy Technology Data Exchange (ETDEWEB)

Wood, J C; Foo, M T; Berthiaume, L C; Herbert, L N; Schaefer, J D; Hawley, D [Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories, Chalk River, ON KOJ 1JO (Canada)

1985-07-01

Our facilities have been upgraded to manufacture fuel rods comprising dispersions of U{sub 3}Si in aluminum, to complement the dispersions of U{sub 3}Si alloyed with 1.5 and 3.0 wt% Al fabricated and tested previously. Further advances have been made in process optimization particularly in core extrusion where production rate has been doubled while maintaining high quality standards. Our mini-element irradiations of Al-61.5 wt% (U,3.5 wt% Si, 1.5 wt% Al) and Al-62.4 wt% (U,3.2 wt% Si, 30 wt% Al) have been completed successfully up to the terminal burnup of 93 atomic percent. Fuel core swelling remained marginally below 1% per 10 atomic percent burnup over the whole irradiation. Also mini-elements containing Al-72.4 wt% USiAl and Al-73.4 wt% USi*Al have been irradiated to 82 atomic percent burnup, their swelling rate marginally exceeding 1% per 10 atomic percent burnup. Three full-size 12-element NRU assemblies containing Al-62.4 wt% USi*Al have been fabricated and installed in the NRU reactor where they have performed normally without problems. The cores for four more full-size 12-element NRU assemblies containing Al-61.0 wt% U{sub 3}Si have been manufactured. (author)

Reduced enrichment fuels for Canadian research reactors - Fabrication and performance

International Nuclear Information System (INIS)

Wood, J.C.; Foo, M.T.; Berthiaume, L.C.; Herbert, L.N.; Schaefer, J.D.; Hawley, D.

1985-01-01

Our facilities have been upgraded to manufacture fuel rods comprising dispersions of U 3 Si in aluminum, to complement the dispersions of U 3 Si alloyed with 1.5 and 3.0 wt% Al fabricated and tested previously. Further advances have been made in process optimization particularly in core extrusion where production rate has been doubled while maintaining high quality standards. Our mini-element irradiations of Al-61.5 wt% (U,3.5 wt% Si, 1.5 wt% Al) and Al-62.4 wt% (U,3.2 wt% Si, 30 wt% Al) have been completed successfully up to the terminal burnup of 93 atomic percent. Fuel core swelling remained marginally below 1% per 10 atomic percent burnup over the whole irradiation. Also mini-elements containing Al-72.4 wt% USiAl and Al-73.4 wt% USi*Al have been irradiated to 82 atomic percent burnup, their swelling rate marginally exceeding 1% per 10 atomic percent burnup. Three full-size 12-element NRU assemblies containing Al-62.4 wt% USi*Al have been fabricated and installed in the NRU reactor where they have performed normally without problems. The cores for four more full-size 12-element NRU assemblies containing Al-61.0 wt% U 3 Si have been manufactured. (author)

The influence of SiO2 Addition on 2MgO-Al2O3-3.3P2O5 Glass

DEFF Research Database (Denmark)

Larsen, P.H.; Poulsen, F.W.; Berg, Rolf W.

1999-01-01

2MgO-Al2O3-3.3P2O5 glasses with increasing amounts of SiO2 are considered for sealing applications in Solid Oxide Fuel Cells (SOFC). The change in chemical durability under SOFC anode conditions and the linear thermal expansion is measured as functions of the SiO2 concentration. Raman spectroscopy...... analysis of the glasses reveals no sign of important changes in the glass structure upon SiO2 addition. Some increase in glass durability with SiO2 concentration is reported and its cause is discussed....

Recent status of development and irradiation performance for plate type fuel elements with reduced 235U enrichment at NUKEM

International Nuclear Information System (INIS)

Hrovat, M.F.; Hassel, H.W.

1984-01-01

According to the present state of development full size test fuel elements with the maximum uranium densities of 2,2 g U/cm 3 meat for UAlsub(x), 3,2 g U/cm 3 meat for U 3 O 8 and 4,8 g U/cm 3 meat for U 3 Si 2 can be fabricated at NUKEM in production scale. Special chemical procedures for the uranium recovery were developed ensuring an economic fuel fabrication process. The post irradiation examinations (PIE) of 12 UAlsub(x) (U density 2,2 g U/cm 3 meat) and U 3 O 8 (up to 3,1 g U/cm 3 meat) test plates irradiated in the ORR, Oak Ridge research reactor, were terminated. All 12 test plates show unobjectionable irradiation behavior. Extensive irradiation tests on full size fuel elements were performed. All inserted elements show perfect irradiation behavior. The PIE of the first HFR Petten U 3 O 8 fuel elements are in progress. The full size ORR U 3 Si 2 fuel elements with so far highest uranium density of 4,76 g U/cm 3 meat achieved a burnup of 50 % loss of 235 U up to May 1983. One element was withdrawn from the reactor for PIE, the second will be irradiated to a burnup of 75 % loss of 235 U. The further development is concentrated on Usub(x)Sisub(y) fuel with highest uranium density. U 3 Si miniplates with up to 6,1 g U/cm 3 meat are supplied meeting the required specification, U 3 Si miniplates with 6,7 g U/cm 3 are in fabrication. (author)

First-principles investigations on the electronic structures of U{sub 3}Si{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Wang, Tong [College of Information & Communication, Harbin Engineering University, Harbin, 150001 (China); Qiu, Nianxiang [Engineering Laboratory of Specialty Fibers and Nuclear Energy Materials, Ningbo Institute of Materials Technology and Engineering, Chinese Academy of Sciences, Ningbo, Zhejiang, 315201 (China); Wen, Xiaodong [State Key Laboratory of Coal Conversion, Institute of Coal Chemistry, Chinese Academy of Sciences, P.O. Box 165, Taiyuan, Shanxi, 030001 (China); Synfuels China, 100195 (China); Tian, Yonghui [College of Life Science, Sichuan University, Chengdu, Sichuan, 610064 (China); He, Jian [Center for Translational Medicine, Department of Biotechnology, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, Dalian, Liaoning, 116023 (China); Luo, Kan; Zha, Xianhu; Zhou, Yuhong; Huang, Qing; Lang, Jiajian [Engineering Laboratory of Specialty Fibers and Nuclear Energy Materials, Ningbo Institute of Materials Technology and Engineering, Chinese Academy of Sciences, Ningbo, Zhejiang, 315201 (China); Du, Shiyu, E-mail: dushiyu@nimte.ac.cn [Engineering Laboratory of Specialty Fibers and Nuclear Energy Materials, Ningbo Institute of Materials Technology and Engineering, Chinese Academy of Sciences, Ningbo, Zhejiang, 315201 (China)

2016-02-15

U{sub 3}Si{sub 2} has been widely utilized as a high-power uranium fuel for research reactors due to its high density of uranium. However, theoretical investigations on this material are still scarce up to now. For this reason, the computational study via density functional theory (DFT) is performed on the U{sub 3}Si{sub 2} compound in this work. The properties of U{sub 3}Si{sub 2}, such as stable crystalline structures, density of states, charge distributions, formation energy of defects, as well as the mechanical properties are explored. The calculation results show that the U{sub 3}Si{sub 2} material is metallic and brittle, which is in good agreement with the previous experimental observations. The formation energy of uranium vacancy defect is predicted to be the lowest, similar with that of UN. The theoretical investigation of this work is expected to provide new insight of uranium silicide fuels.

Phase relations in the SiC-Al2O3-Pr2O3 system

International Nuclear Information System (INIS)

Pan, W.; Wu, L.; Jiang, Y.; Huang, Z.

2016-01-01

Phase relations in the Si-Al-Pr-O-C system, including the SiC-Al 2 O 3 -Pr 2 O 3 , the Al 2 O 3 -Pr 2 O 3 -SiO 2 and the SiC-Al 2 O 3 -Pr 2 O 3 -SiO 2 subsystems, were determined by means of XRD phase analysis of solid-state-reacted samples fabricated by using SiC, Al 2 O 3 , Pr 2 O 3 and SiO 2 powders as the starting materials. Subsolidus phase diagrams of the systems were presented. Two Pr-aluminates, namely PrAlO 3 (PrAP) and PrAl 11 O 18 (β(Pr) β-Al 2 O 3 type) were formed in the SiC-Al 2 O 3 -Pr 2 O 3 system. SiC was compatible with both of them. Pr-silicates of Pr 2 SiO 5 , Pr 2 Si 2 O 7 and Pr 9.33 Si 6 O 26 (H(Pr) apatite type) were formed owing to presence of SiO 2 impurity in the SiC powder. The presence of the SiO 2 extended the ternary system of SiC-Al 2 O 3 -Pr 2 O 3 into a quaternary system of SiC-Al 2 O 3 -SiO 2 -Pr 2 O 3 (Si-Al-Pr-O-C). SiC was compatible with Al 2 O 3 , Pr 2 O 3 and the Pr-silicates. The effect of SiO 2 on the phase relations and liquid phase sintering of SiC ceramics was discussed.

Predicted irradiation behavior of U3O8-Al dispersion fuels for production reactor applications

International Nuclear Information System (INIS)

Cronenberg, A.W.; Rest, J.

1990-01-01

Candidate fuels for the new heavy-water production reactor include uranium/aluminum alloy and U 3 O 8 -Al dispersion fuels. The U 3 O 8 -Al dispersion fuel would make possible higher uranium loadings and would facilitate uranium recycle. Research efforts on U 3 O 8 -Al fuel include in-pile irradiation studies and development of analytical tools to characterize the behavior of dispersion fuels at high-burnup. In this paper the irradiation performance of U 3 O 8 -Al is assessed using the mechanistic Dispersion Analysis Research Tool (DART) code. Predictions of fuel swelling and alteration of thermal conductivity are presented and compared with experimental data. Calculational results indicate good agreement with available data where the effects of as-fabricated porosity and U 3 O 8 -Al oxygen exchange reactions are shown to exert a controlling influence on irradiation behavior. The DART code is judged to be a useful tool for assessing U 3 O 8 -Al performance over a wide range of irradiation conditions

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.

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.

Full size U-10Mo monolithic fuel foil and fuel plate fabrication-technology development

International Nuclear Information System (INIS)

Moore, G.A.; Jue, J-F.; Rabin, B.H.; Nilles, M.J.

2010-01-01

Full-size U-10Mo foils are being developed for use in high density LEU monolithic fuel plates. The application of a zirconium barrier layer to the foil is performed using a hot co-rolling process. Aluminium clad fuel plates are fabricated using Hot Isostatic Pressing (HIP) or a Friction Bonding (FB) process. An overview is provided of ongoing technology development activities, including: the co-rolling process, foil shearing/slitting and polishing, cladding bonding processes, plate forming, plate-assembly swaging, and fuel plate characterization. Characterization techniques being employed include, Ultrasonic Testing (UT), radiography, and microscopy. (author)

High-Uranium-Loaded U3O8-Al fuel element development program. Part 1

International Nuclear Information System (INIS)

Martin, M.M.

1993-01-01

The High-Uranium-Loaded U 3 O 8 -Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U 3 O 8 -Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U 3 O 8 -Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U 3 O 8 ). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % U involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U 3 O 8 and aluminum

Application of multivariate calibration for simultaneous determination of major and minor constituents in U3Si2 by X-ray fluorescence

International Nuclear Information System (INIS)

Scapin, Marcos A.; Silva, Clayton P.; Cotrim, Marycel E.B.; Pires, Maria Aparecida F.

2013-01-01

The aim of this work is to establish and validate a methodology for a nondestructive quantitative chemical analysis method for simultaneous determination of the major constituents (U total and Si) and impurities (B, Mg, Al, Cr, Mn, Fe, Co, Ni, Cu, Zn, Mo, Cd,etc.) present in U 3 Si 2 . The method must also meet the needs of nuclear reactors for the nuclear fuel qualification type,MTR, with low cost and analysis time, while also minimizing waste generation. For this purpose, an X-ray fluorescence technique will be applied. The technique is nondestructive,aside from sample preparation procedures that do not require previous chemical treatments (dissolving, digesting), and allows for fast chemical analysis. The fundamental parameters (FP) method was applied to corrections for spectral and matrix effects. The calibration model was obtained via principal component analysis using orthogonal decomposition by the singular value decomposition method (SVD) in U 3 O 8 and U 3 Si 2 samples. The results were compared by means of statistical tests in accordance with ISO 17025 on CRMs of U 3 O 8 from New Brunswick Laboratory (NBL) and 16 U 3 Si 2 samples provided by CCN of IPEN/CNEN-SP. Multivariate calibration is a promising method for determination of major and minor constituents inU 3 Si 2 and U 3 O 8 nuclear fuel, because the precision and accuracy are statistically equivalent to volumetric analysis (U total determination), gravimetric analysis (Si determination), and ICP-OES methods (impurities determination). (author)

The status of uranium-silicon alloy fuel development for the RERTR program

International Nuclear Information System (INIS)

Domagala, R.F.; Wiencek, T.C.; Thresh, H.R.; Stahl, D.

1983-01-01

As part of the national Reduced Enrichment Research and Test Reactor (RERTR) Program, Argonne National Laboratory (ANL) is engaged in a fuel-alloy development project. The fuel alloys are dispersed in an aluminum matrix and metallurgically roll-bonded within 6061 Al alloy. To date, 'miniplates' with up to 40 vol. fuel alloy have been successfully fabricated. Thirty-one of these plates have been or are being irradiated in the Oak Ridge Reactor (ORR). Three different fuels have been used in the ANL miniplates: U 3 Si (U + 4 wt.% Si), U 3 Si 2 (U + 7.4 wt.% Si), or ''U 3 SiAl'' (U + 3.5 wt.% Si + 1.5 wt.% Al). All three are candidates for permitting higher fuel loadings and thus lower enrichments of 235 U than would be possible with either UAl x or U 3 O 8 , the current fuels for plate-type elements. The enrichment level employed at ANL is ∼19.8%. Continuing effort involves the production of miniplates with up to ∼60 vol. % fuel, the development of a technology for full-size plate fabrication, and post-irradiation examination of miniplates already removed from the ORR. (author)

Core–shell structured FeSiAl/SiO{sub 2} particles and Fe{sub 3}Si/Al{sub 2}O{sub 3} soft magnetic composite cores with tunable insulating layer thicknesses

Energy Technology Data Exchange (ETDEWEB)

Fan, Xi’an, E-mail: groupfxa@163.com [The State Key Laboratory of Refractories and Metallurgy, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Key Laboratory for Ferrous Metallurgy and Resources Utilization of Ministry of Education, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Wang, Jian, E-mail: snove418562@163.com [The State Key Laboratory of Refractories and Metallurgy, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Key Laboratory for Ferrous Metallurgy and Resources Utilization of Ministry of Education, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Wu, Zhaoyang, E-mail: wustwuzhaoyang@163.com [The State Key Laboratory of Refractories and Metallurgy, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Key Laboratory for Ferrous Metallurgy and Resources Utilization of Ministry of Education, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Li, Guangqiang, E-mail: ligq-wust@mail.wust.edu.cn [The State Key Laboratory of Refractories and Metallurgy, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China); Key Laboratory for Ferrous Metallurgy and Resources Utilization of Ministry of Education, Wuhan University of Science and Technology, Wuhan, Hubei 430081 (China)

2015-11-15

Graphical abstract: - Highlights: • FeSiAl/SiO{sub 2} core–shell particles and Fe{sub 3}Si/Al{sub 2}O{sub 3} composite cores were prepared. • SiO{sub 2} surrounding FeSiAl were replaced by Al{sub 2}O{sub 3} during sintering process. • Fe{sub 3}Si particles were separated by Al{sub 2}O{sub 3} with tunable thickness in composite cores. • Fe{sub 3}Si/Al{sub 2}O{sub 3} had lower core loss and better frequency stability than FeSiAl core. • The insulating layer between ferromagnetic particles can reduce core loss. - Abstract: FeSiAl/SiO{sub 2} core–shell particles and Fe{sub 3}Si/Al{sub 2}O{sub 3} composite cores with tunable insulating layer thicknesses have been synthesized via a modified Stöber method combined with following high temperature sintering process. Most of the conductive FeSiAl particles could be coated by insulating SiO{sub 2} using the modified Stöber method. During the sintering process, the reaction 4Al + 3SiO{sub 2} ≣ 2α-Al{sub 2}O{sub 3} + 3Si took place and the new Fe{sub 3}Si/Al{sub 2}O{sub 3} composite was formed. The Fe{sub 3}Si/Al{sub 2}O{sub 3} composite cores displayed more excellent soft magnetic properties, better frequency stability at high frequencies, much higher resistivity and lower core loss than the raw FeSiAl core. Based on this, several types of FeSiAl/SiO{sub 2} particles and Fe{sub 3}Si/Al{sub 2}O{sub 3} composite cores with tunable insulating layer thicknesses were selectively prepared by simply varying TEOS contents. The thickness of Al{sub 2}O{sub 3} insulating layer and resistivity of Fe{sub 3}Si/Al{sub 2}O{sub 3} composite cores increased with increasing the TEOS contents, while the permeability and core loss changed in the opposite direction.

CNEA/ANL collaboration program to develop an optimized version of DART validation and assessment by means of U{sub 3}Si{sub x} and U{sub 3}O{sub 8-}Al dispersed CNEA miniplate irradiation behavior.

Energy Technology Data Exchange (ETDEWEB)

Solis, D.

1998-10-16

The DART code is based upon a thermomechanical model that can predict swelling, recrystallization, fuel-meat interdiffusion and other issues related with MTR dispersed FE behavior under irradiation. As a part of a common effort to develop an optimized version of DART, a comparison between DART predictions and CNEA miniplates irradiation experimental data was made. The irradiation took place during 1981-82 for U3O8 miniplates and 1985-86 for U{sub 3}Si{sub x} at Oak Ridge Research Reactor (ORR). The microphotographs were studied by means of IMAWIN 3.0 Image Analysis Code and different fission gas bubbles distributions were obtained. Also it was possible to find and identify different morphologic zones. In both kinds of fuels, different phases were recognized, like particle peripheral zones with evidence of Al-U reaction, internal recrystallized zones and bubbles. A very good agreement between code prediction and irradiation results was found. The few discrepancies are due to local, fabrication and irradiation uncertainties, as the presence of U{sub 3}Si phase in U{sub 3}Si{sub 2} particles and effective burnup.

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.

U3Si2 behavior in H2O: Part I, flowing steam and the effect of hydrogen

Science.gov (United States)

Wood, E. Sooby; White, J. T.; Grote, C. J.; Nelson, A. T.

2018-04-01

Recent interest in U3Si2 as an advanced light water reactor fuel has driven assessment of numerous properties, but characterization of its response to H2O environments is absent from the literature. The behavior of U3Si2 in H2O containing atmospheres is investigated and presented in a two-part series of articles aimed to understand the degradation mechanism of U3Si2 in H2O. Reported here are thermogravimetric data for U3Si2 exposed to flowing steam at 250-470 °C. Additionally the response of U3Si2 to flowing Ar-6% H2 from 350 to 400 °C is presented. Microstructural degradation is observed following hours of exposure at 350 °C in steam. U3Si2 undergoes pulverization on the timescale of minutes when temperatures are increased above 400 °C. This mechanism is accelerated in flowing Ar-H2 at the same temperatures.

Microstructure studies of interdiffusion behavior of U{sub 3}Si{sub 2}/Zircaloy-4 at 800 and 1000 °C

Energy Technology Data Exchange (ETDEWEB)

He, Lingfeng, E-mail: Lingfeng.He@inl.gov; Harp, Jason M., E-mail: Jason.Harp@inl.gov; Hoggan, Rita E.; Wagner, Adrian R.

2017-04-01

Fuel swelling during normal reactor operations could lead to unfavorable chemical interactions when in contact with its cladding. As new fuel types are developed, it is crucial to understand the interaction behavior between fuel and its cladding. Diffusion experiments between U{sub 3}Si{sub 2} and Zricaloy-4 (Zry-4) were conducted at 800 and 1000 °C up to 100 h. The microstructure of pristine U{sub 3}Si{sub 2} and U{sub 3}Si{sub 2}/Zry-4 interdiffusion products were examined using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) equipped with an energy dispersive X-ray spectroscopy (EDS) system. The primary interdiffusion product observed at 800 °C is ZrSi{sub 2}, with secondary phases of U-Zr in the Zry-4, and Fe-Cr-W-Zr-Si phases at Zry-4/ZrSi{sub 2} interface and Fe-Cr-U-Si phases at ZrSi{sub 2}/U-Si interface. The primary interdiffusion products at 1000 °C were Zr{sub 2}Si, U-Zr-Fe-Ni, U, U-Zr, and a low melting point phase U{sub 6}Fe.

Production of Al2O3–SiC nano-composites by spark plasma sintering; Producción de nano-composites – SiC–Al2O3 por spark plasma sinterizado

Energy Technology Data Exchange (ETDEWEB)

Mansour Razavi; Ali Reza Farajipour; Mohammad Zakeri; Mohammad Reza Rahimipour; Ali Reza Firouzbakht

2017-11-01

In this paper, Al2O3–SiC composites were produced by SPS at temperatures of 1600°C for 10min under vacuum atmosphere. For preparing samples, Al2O3 with the second phase including of micro and nano-sized SiC powder were milled for 5h. The milled powders were sintered in a SPS machine. After sintering process, phase studies, densification and mechanical properties of Al2O3–SiC composites were examined. Results showed that the specimens containing micro-sized SiC have an important effect on bulk density, hardness and strength. The highest relative density, hardness and strength were 99.7%, 324.6 HV and 2329MPa, respectively, in Al2O3–20wt% SiCmicro composite. Due to short time sintering, the growth was limited and grains still remained in nano-meter scale. [Spanish] En este trabajo se muestran compuestos de Al2O3-SiC producidos por SPS, en vacío, a 1.600 °C durante 10 min. Para la preparación de muestras, se molieron polvos de Al2O3 durante 5 h con la segunda fase de micro-y-nano polvo de SiC. Posteriormente, estos polvos molidos se sinterizaron mediante SPS. Después del proceso de sinterización, se realizaron estudios de fase, densificación y propiedades mecánicas de los compuestos de Al2O3-SiC obtenidos. Los resultados mostraron que micro-SiC en las muestras tiene un efecto importante en su densidad aparente, dureza y resistencia. La mayor densidad relativa, dureza y resistencia fueron respectivamente del 99,7%, 324,6 HV y 2.329 MPa para Al2O3 con un 20% en peso micro-SiC. Debido al corto tiempo de sinterización, el crecimiento los granos fue limitado y se mantuvieron en escala nanométrica.

Modeling defect and fission gas properties in U-Si fuels

Energy Technology Data Exchange (ETDEWEB)

Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Noordhoek, Mark [Univ. of South Carolina, Columbia, SC (United States); Besmann, Theodore [Univ. of South Carolina, Columbia, SC (United States); Middleburgh, Simon C. [Westinghouse Electric Sweden, Vasteras (Sweden); Lahoda, E. J. [Westinghouse Electric Company LLC, Cranberry Woods, PA (United States); Chernatynskiy, Aleksandr [Missouri University of Science and Technology; Grimes, Robin W. [Imperial College, London (United Kingdom)

2017-04-27

Uranium silicides, in particular U₃Si₂, are being explored as an advanced nuclear fuel with increased accident tolerance as well as competitive economics compared to the baseline UO₂ fuel. They benefit from high thermal conductivity (metallic) compared to UO₂ fuel (insulator or semi-conductor) used in current Light Water Reactors (LWRs). The U-Si fuels also have higher fissile density. In order to perform meaningful engineering scale nuclear fuel performance simulations, the material properties of the fuel, including the response to irradiation environments, must be known. Unfortunately, the data available for USi fuels are rather limited, in particular for the temperature range where LWRs would operate. The ATF HIP is using multi-scale modeling and simulations to address this knowledge gap.

Modeling defect and fission gas properties in U-Si fuels

Energy Technology Data Exchange (ETDEWEB)

Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Noordhoek, Mark J. [Univ. of South Carolina, Columbia, SC (United States); Besmann, Theodore M. [Univ. of South Carolina, Columbia, SC (United States); Middleburgh, Simon C. [Westinghouse Electric Sweden, Vasteras (Sweden); Lahoda, E. J. [Westinghouse Electric Company LLC, Cranberry Woods, PA (United States); Chernatynskiy, Aleksandr [Missouri Univ. of Science and Technology, Rolla, MO (United States); Grimes, Robin W. [Imperial College, London (United Kingdom)

2017-04-14

Uranium silicides, in particular U₃Si₂, are being explored as an advanced nuclear fuel with increased accident tolerance as well as competitive economics compared to the baseline UO2 fuel. They benefit from high thermal conductivity (metallic) compared to UO₂ fuel (insulator or semi-conductor) used in current Light Water Reactors (LWRs). The U-Si fuels also have higher fissile density. In order to perform meaningful engineering scale nuclear fuel performance simulations, the material properties of the fuel, including the response to irradiation environments, must be known. Unfortunately, the data available for USi fuels are rather limited, in particular for the temperature range where LWRs would operate. The ATF HIP is using multi-scale modeling and simulations to address this knowledge gap.

Investigation for thermal stability of U3Si2 and protection methods

International Nuclear Information System (INIS)

Zhang Huiying; Sun Jichang; Sun Rongxian

1994-08-01

The thermal stability of U 3 Si 2 in Ar, N 2 and air, and the interaction between U 3 Si 2 and Al, Zr have been investigated by thermal analysis method. According to the results of thermal analysis, protection measures for various procedures have been improved. From the practice, it shows that the protection measures can ensure the safety of production and raise the product quality as well as reduce the cost effectively

Prediction of U3SI2-Al burn-up and SiC/p-AI composition effects on its thermal conductivity using metal matrix composite (MMC) model containing progressive sub-dispersion

International Nuclear Information System (INIS)

Suwardi

2000-01-01

The model takes into account the evolution of constituent volume fraction. Sub-dispersion of disperse contains fission gas bubbles that increase with bum-up. The metal matrix could contain pore and void, a different type of disperse that vary wth time. The model is previously aimed to dispersion-nuclear fuel element. The model consists of a combination of different conductance constituent of both matrix and sub-matrix. Application is carried out to predict the fuel swelling effect on thermal conductivity of U 3 SI 2 -Al dispersion, and to volume fraction effect on conductivity of SiC-particulate reinforced AI matrix. The model shows that both fuel fraction and fission gas swelling decrease the thermal conductivity. During the start-up period of swelling the conductivity increases as aluminum pore close. then decreases most linearly. SiC/p-AI conductivity decreases most linearly with particulate volume fraction, attains 57.6% of pure AI at 50 % v/v. The author conclude that the model developed is applicable for more general MMC. (author)

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.

Production of Al2O3–SiC nano-composites by spark plasma sintering

International Nuclear Information System (INIS)

Mansour Razavi; Ali Reza Farajipour; Mohammad Zakeri; Mohammad Reza Rahimipour; Ali Reza Firouzbakht

2017-01-01

In this paper, Al2O3–SiC composites were produced by SPS at temperatures of 1600°C for 10min under vacuum atmosphere. For preparing samples, Al2O3 with the second phase including of micro and nano-sized SiC powder were milled for 5h. The milled powders were sintered in a SPS machine. After sintering process, phase studies, densification and mechanical properties of Al2O3–SiC composites were examined. Results showed that the specimens containing micro-sized SiC have an important effect on bulk density, hardness and strength. The highest relative density, hardness and strength were 99.7%, 324.6 HV and 2329MPa, respectively, in Al2O3–20wt% SiCmicro composite. Due to short time sintering, the growth was limited and grains still remained in nano-meter scale. [es

System for uranium superficial density measurement in U3Si2 MTR fuel plates using radiography

International Nuclear Information System (INIS)

Hey, Martin A.; Gomez Marlasca, Fernando

2003-01-01

The paper describes a method for measuring uranium superficial density in high density uranium silicide (U 3 Si 2 ) MTR fuel plates, through the use of industrial radiography, a set of patterns built for this purpose, a transmission optical densitometer, and a quantitative model of analysis and measurement. Our choice for this particular method responds to its high accuracy, low cost and easy implementation according to the standing quality control systems. (author)

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

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.

Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

International Nuclear Information System (INIS)

Muhammad, Farhan; Majid, Asad

2009-01-01

The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U 3 Si 2 -Al followed by 0.03% for U 3 Si-Al, and 0.01% for U 3 O 8 -Al fuel. The U 3 O 8 -Al fueled reactor gave the maximum ρ excess at BOL which was 21.67% more than the original fuel followed by U 3 Si-Al which was 2.55% more, while that of U 3 Si 2 -Al was 2.50% more than the original UAl x -Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U 3 O 8 -Al followed by U 3 Si-Al and then U 3 Si 2 -Al fuel.

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

Obtenção de substratos cerâmicos no sistema Si-Al-O-N-C empregando polissiloxanos e carga de Si e Al2O3 Ceramic tapes of Si-Al-O-N-C compounds using mixtures of polyssiloxane and Si-Al2O3 fillers

Directory of Open Access Journals (Sweden)

R. M. Rocha

2005-03-01

Full Text Available A técnica de processamento de colagem por fita (tape casting tem sido amplamente utilizada na obtenção de cerâmicas para diferentes aplicações: substratos cerâmicos e estruturas planares em multicamadas para circuitos integrados e capacitores; eletrólitos sólidos para células a combustível e sensores; cerâmicas piezoelétricas para atuadores e transdutores; membranas de separação para micro-filtragem; compósitos estruturais e trocadores de calor. Neste trabalho, a técnica convencional de colagem por fita foi adaptada com a utilização do processo de pirólise controlada de misturas de polímeros e carga, empregando-se polissiloxanos e cargas de silício e alumina nas suspensões. Foram preparadas suspensões com 60% vol. de fase polimérica (polissiloxanos e metil-trietoxi-silano e 40% vol. de carga (Si e Al2O3, com diferentes concentrações dos polímeros e das cargas. As amostras na forma de substratos foram pirolisadas em atmosfera de nitrogênio a 1400 °C/2 h e 1500 °C/2 h, sendo convertidas em materiais cerâmicos no sistema Si-Al-O-N-C. O processo de pirólise foi caracterizado até a temperatura de 1000 ºC por análise termogravimétrica. As cerâmicas foram caracterizadas quanto às fases formadas, microestrutura, massa específica aparente e condutividade térmica. Os substratos cerâmicos apresentaram porosidade relativamente alta (entre 12 e 22% e baixa condutividade térmica (entre 3 e 8 W/m.K, sendo constituídos por fases cristalinas de beta-SiC, Si2ON2, O'-SiAlON, Al2O3, mulita e fase amorfa de SiOC; o Si foi observado nas amostras pirolisadas a 1400 ºC.The tape casting technique has been widely used to prepare ceramic tapes for different applications: ceramic substrates and multilayer planar structures for integrated circuits and capacitors, solid electrolytes for fuel cells and sensors, piezoelectric ceramics for actuators and transducers, membrane systems for micro-filtration, structural composites and

Structural and microstructural characterization of U{sub 3}Si{sub 2} nuclear fuel using X-ray diffraction

Energy Technology Data Exchange (ETDEWEB)

Ichikawa, Rodrigo U.; Garcia, Rafael H.L.; Silva, Andre S.B. da; Saliba-Silva, Adonis M.; Lima, Nelson B.; Martinez, Luis G. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Turrillas, Xavier, E-mail: ichikawa@usp.br, E-mail: rlgarcia@ipen.br, E-mail: andre.santos.silva@usp.br, E-mail: saliba@ipen.br, E-mail: nblima@ipen.br, E-mail: lgallego@ipen.br, E-mail: xturrillas@icmab.es [Institut de Ciència de Materials de Barcelona (ICMAB/CSIC), Cerdanyola de Vallès (Spain)

2017-07-01

In this work, two uranium silicide powdered samples, containing 67% and 42 mol% of Si, were analyzed using X-ray diffraction (named as {sup 67}Si and {sup 42}Si). For structural characterization, Rietveld refinement was used to estimate cell parameters, volume fraction (weight percent) of crystalline phases and atomic positions. For the main phases, X-ray line profile analysis (XLPA) was used to estimate mean crystallite sizes and micro strains. The {sup 67}Si sample presents higher content of USi{sub 2(}tetragonal) and the {sup 42}Si sample presents higher content of U{sub 3}Si{sub 2} (tetragonal) as identified and calculated from the XRD profiles. Overall there are no appreciable structural changes and the parameters refined are in good accordance with the ones reported in the literature. Mean crystallite sizes determined by XLPA revealed small crystallites of the order of 10{sup 1} nm and low micro strain for all samples. (author)

Program for in-pile qualification of high density silicide dispersion fuel at IPEN/CNEN-SP

International Nuclear Information System (INIS)

Silva, Jose E.R. da; Silva, Antonio T. e; Terremoto, Luis A.A.; Durazzo, Michelangelo

2009-01-01

The development of high density nuclear fuel (U 3 Si 2 -Al) with 4,8 gU/cm 3 is on going at IPEN, at this time. This fuel has been considered to be utilized at the new Brazilian Multipurpose Reactor (RMB), planned to be constructed up to 2014. As Brazil does not have hot-cell facilities available for post-irradiation analysis, an alternative qualifying program for this fuel is proposed based on the same procedures used at IPEN since 1988 for qualifying its own U 3 O 8 -Al (1,9 and 2,3 gU/cm 3 ) and U 3 Si 2 -Al (3,0 gU/cm 3 ) dispersion fuels. The fuel miniplates and full-size fuel elements irradiations should be tested at IEA-R1 core. The fuel characterization along the irradiation time should be made by means of non-destructive methods, including periodical visual inspections with an underwater video camera system, sipping tests for fuel elements suspected of leakage, and underwater dimensional measurements for swelling evaluation, performed inside the reactor pool. This work presents the program description for the qualification of the high density nuclear fuel (U 3 Si 2 -Al) with 4,8 gU/cm 3 , and describes the IPEN fuel fabrication infrastructure and some basic features of the available systems for non-destructive tests at IEA-R1 research reactor. (author)

Effect of Ni/Al2O3-SiO2 and Ni/Al2O3-SiO2 with K2O Promoter Catalysts on H2, CO and CH4 Concentration by CO2 Gasification of Rosa Multiflora Biomass

Directory of Open Access Journals (Sweden)

Tursunov Obid

2017-11-01

Full Text Available The thermal behaviour of the Rosa mutiflora biomass by thermogravimetric analysis was studied at heating rate 3 K min−1 from ambient temperature to 950 °C. TGA tests were performed in high purity carbon dioxide (99 998% with a flow rate 200 ml/min and 100 mg of sample, milled and sieved to a particle size below 250 µm. Moreover, yields of gasification products such as hydrogen (H2, carbon monoxide (CO and methane (CH4 were determined based on the thermovolumetric measurements of catalytic (Ni/Al2O3-SiO2 and Ni/Al2O3-SiO2 with K2O promoter catalysts and non-catalytic gasification of the Rosa multiflora biomass. Additionally, carbon conversion degrees are presented. Calculations were made of the kinetic parameters of carbon monoxide and hydrogen formation reaction in the catalytic and non-catalytic CO2 gasification processes. A high temperature of 950 °C along with Ni/Al2O3-SiO2and Ni/Al2O3-SiO2 with K2O promoter catalysts resulted in a higher conversion of Rosa multiflora biomass into gaseous yield production with greatly increasing of H2 and CO contents. Consequently, H2 and CO are the key factors to produce renewable energy and bio-gases (synthesis gas. The parameters obtained during the experimental examinations enable a tentative assessment of plant biomasses for the process of large-scale gasification in industrial sectors.

Interface reactions in the Al-Si-SiC and Mg-Al-Al{sub 2}O{sub 3} composite systems

Energy Technology Data Exchange (ETDEWEB)

Johnson, P.K. [Commission of the European Communities, Petten (Netherlands). Inst. for Advanced Materials; Fazal-Ur-Rehman [Imperial Coll. of Science, Technology and Medicine, London (United Kingdom). Dept. of Materials; Fox, S. [Imperial Coll. of Science, Technology and Medicine, London (United Kingdom). Dept. of Materials; Flower, H.M. [Imperial Coll. of Science, Technology and Medicine, London (United Kingdom). Dept. of Materials; West, D.R.F.

1995-12-31

Structural and compositional observations are reported on the influence of the interfaces on the mechanisms and kinetics of liquid metal-ceramic reactions in Al-SiC, Al-Si-SiC, Mg-Al{sub 2}O{sub 3} and Mg-Al-Al{sub 2}O{sub 3} composites. The aluminium based materials contained up to 20 vol% SiC in particulate form, and were produced by a spray casting process; subsequently the interface reactions were studied in samples heated to temperatures up to 1100 C. The reaction product was Al{sub 4}C{sub 3} in both Al-SiC and Al-Si-SiC composites. The influence of the crystallography and topology of the SiC particle surfaces on the nucleation of the Al{sub 4}C{sub 3} has been demonstrated; surface asperities play an important role. Growth of nuclei proceeds to form continuous reaction product layers which control the subsequent kinetics. The magnesium based composites contained 5 vol% Al{sub 2}O{sub 3} fibres (3 {mu}m in diameter), and were produced by a liquid infiltration process. SD Safimax fibres with relatively low and high porosity, and also RF Saffil fibres, with a silica binder, were investigated. Fibre porosity plays a major role in accelerating the penetration of Mg into the fibres with reaction to form MgO. Silica binder on the fibre surface transforms to MgO. The reaction rate was reduced by the presence of aluminium in the matrix. The factors controlling the reactions in the aluminium and magnesium based composites are compared. (orig.)

Assessment of semi-empirical potentials for the U-Si system

Energy Technology Data Exchange (ETDEWEB)

Baskes, Michael I. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Andersson, Anders David [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-02-03

compounds, because they are the main nuclear fuel candidates. U₃Si₅ is derived from USi₂ in the AlB₂ structure by creating 1/6 vacant sites on Si sublattice. The ordering of these vacancies will not be studied in any detail here.

Dependence of glass-forming ability on starting compositions in Y2O3–Al2O3–SiO2 system

Directory of Open Access Journals (Sweden)

Yixiang Chen

2011-09-01

Full Text Available The dependence of glass-forming ability on starting compositions in Y2O3–Al2O3–SiO2 (YAS system has been investigated by melting experiment. Transparent YAS glasses have been prepared under the condition of furnace cooling instead of quenching. It is found that, in the YAS ternary phase diagram, the compositions on the Y3Al5O12–SiO2 line and with 52-68 mol% SiO2 have a higher glass-forming ability to produce pure glass. For the compositions with too much or less SiO2 or with Y/Al = 5/3, 1/1, or 1/3, crystallization occurs with the formation of Y3Al5O12, Y2Si2O7, Al6Si2O13, or SiO2. The densities of the YAS glasses increase with decreasing SiO2 contents and increasing Y/Al ratios, and for the samples with Y/Al = 3/5 there is a good linear relationship between the density and SiO2 content.

Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin; Gamble, Kyle A.; Andersson, David; Mei, Zhi-Gang; Yacout, Abdellatif M.

2018-01-01

Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety of LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.

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

Optical properties of the Al2O3/SiO2 and Al2O3/HfO2/SiO2 antireflective coatings

Science.gov (United States)

Marszałek, Konstanty; Winkowski, Paweł; Jaglarz, Janusz

2014-01-01

Investigations of bilayer and trilayer Al2O3/SiO2 and Al2O3/HfO2/SiO2 antireflective coatings are presented in this paper. The oxide films were deposited on a heated quartz glass by e-gun evaporation in a vacuum of 5 × 10-3 [Pa] in the presence of oxygen. Depositions were performed at three different temperatures of the substrates: 100 °C, 200 °C and 300 °C. The coatings were deposited onto optical quartz glass (Corning HPFS). The thickness and deposition rate were controlled with Inficon XTC/2 thickness measuring system. Deposition rate was equal to 0.6 nm/s for Al2O3, 0.6 nm - 0.8 nm/s for HfO2 and 0.6 nm/s for SiO2. Simulations leading to optimization of the thin film thickness and the experimental results of optical measurements, which were carried out during and after the deposition process, have been presented. The optical thickness values, obtained from the measurements performed during the deposition process were as follows: 78 nm/78 nm for Al2O3/SiO2 and 78 nm/156 nm/78 nm for Al2O3/HfO2/SiO2. The results were then checked by ellipsometric technique. Reflectance of the films depended on the substrate temperature during the deposition process. Starting from 240 nm to the beginning of visible region, the average reflectance of the trilayer system was below 1 % and for the bilayer, minima of the reflectance were equal to 1.6 %, 1.15 % and 0.8 % for deposition temperatures of 100 °C, 200 °C and 300 °C, respectively.

A contribution to the kinetic study of the metatectic reaction U+U3Si2→U3Si

International Nuclear Information System (INIS)

Arroyo Ruiperez, J.; Esteban Hernandez, J. A.

1962-01-01

An experimental study has been made to decide upon the advantages and drawbacks of the different methods and reagents employed in the metallography of U-Si alloys. It has been observed that all samples thermally treated to form the epsilon-phase undergo from the beginning a coalescence of the U 3 Si 2 particles, which makes practically useless any fine state of dispersion that might be present originally, as recommended by some authors. The coalescence of the U 3 Si 2 particle decreases the surface available for reaction and consequently the reaction rate. (Author) 7 refs

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.

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.

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.

The Effect of Nano sized Carbon Black on the Physical and Thermomechanical Properties of Al2O3-SiC-SiO2-C Composite

International Nuclear Information System (INIS)

Amin, M.H.; Ebrahimabadi, M.A.; Rahimipour, M.R.

2009-01-01

The effects of using nano sized carbon black in the range of 010 weight percentages on the physical and thermomechanical properties of Al 2 O 3 -Si C-SiO 2 graphite refractory composites were investigated. Nano sized carbon black addition improved the relative heat resistance and oxidation resistance of composites. The bulk density of the composites is reduced with increasing carbon black (CB) content. Increase in CB content first causes an increase in the apparent porosity, but at more than 3 wt % amount of CB, a decrease of apparent porosity was observed. The cold crushing strength (CCS) increased with increasing CB content in samples fired at 800 degree C and in samples fired at 1500 degree C when the content is increased to 3 wt %, but the CCS decreased with increasing CB content in samples fired at 1500 degree C when the CB content was less than 3 wt %. The composite without CB exhibits the highest value of CCS at firing temperature of 1500 degree C.

Study of Irradiation Effect onto Uranium silicide Fuel

International Nuclear Information System (INIS)

Suparjo

1998-01-01

The irradiation effect onto the U 3 Si-Al and U 3 Si 2 -Al dispersion type of fuel element has been studied. The fuel material performs swelling during irradiation due to boehmite (Al 2 O 3 (H 2 O)) formation in which might occurs inside the meat and on the cladding surface, the interaction between the fuel and aluminium matrix that produce U(Al,Si) 3 phase, and the formation of fission gas bubble inside the fuel. At a constant fission density, the U 3 Si-Al fuel swelling is higher than that of U 3 Si 2 -Al fuel. The swellings of both fuels increase with the increasing of fission density. The difference of swelling behavior was caused by formation of large bubble gases generated from fission product of U 3 Si fuel and distributed non-uniformly over all of fuel zone. On the other hand, the U 3 Si 2 fission produced small bubble gases, and those were uniformly distributed. The growth rate of fission gas bubble in the U 3 Si fuel has shown high diffusivity, transformation into amorph material and thus decrease its mechanical strength

Post-irradiation examination of A1-61 wt % U3Si fuel rods from the NRU reactor

International Nuclear Information System (INIS)

Sears, D.F.; Wang, N.

1997-09-01

This paper describes the post-irradiation examination of 4 intact low-enrichment uranium (LEU) fuel rods from the national research universal (NRU) reactor at the Chalk River Laboratories of AECL. The rods were irradiated during the period 1993 through 1995, under typical driver fuel operating conditions in NRU, i.e., nominal D 2 0 coolant inlet temperature 37 degrees C, inlet pressure 654 kPa and mass flow 12.4 L/s. Irradiation exposures ranged from 147 to 251 full-power days, corresponding to 40 to 84 atom % 235 U burnup. The maximum rod power was ∼2 MW, with element linear power ratings up to 68 kW/m. Post-irradiation examinations, conducted in 1997, focused on optical metallography to measure cladding oxide thickness and fuel core and cladding microstructural examinations. The cladding oxide was approximately 24 μm thick at the mid-plane of fuel rods irradiated to 251 full-power days, with small areas up to 34 μm thick on the fins. The cladding retained significant ductility after irradiation, and its microstructure appeared unchanged. Fuel core diametral increases were small (up to 4%) and within the range previously observed on A1-61 wt % U 3 Si fuel irradiated in the NRU reactor. (author)

Monte-Carlo simulation of dispersion fuel meat structure

International Nuclear Information System (INIS)

Xing Zhonghu; Ying Shihao

2003-01-01

Under the irradiation conditions in research reactors, the inter-diffusion occurs at the fuel particle and matrix interfaces of U 3 Si 2 -Al dispersion fuel. Because of the inter-diffusion reaction, the U 3 Al 7 Si 2 layer is formed around each U 3 Si 2 particle. The layer thickness grows up with irradiation duration and fission density. The formation of resultant layer causes the consumption of U 3 Si 2 fuel and aluminum matrix. This process leads to the evolution of geometrical structure of fuel meat. According to the stochastic locations of particles in dispersion, the authors developed a simulation method for the evolution of the fuel meat structure by utilizing Monte-Carlo method. Every particle is characterized by its diameter and location. The parameters of meat structure include particle size distribution, as-fabricated fuel volume fraction, resultant layer thickness, layer volume fraction, U 3 Si 2 fuel volume fraction, aluminum volume fraction, contiguity probability and inter-linkage fraction of particles. Particularly for the dispersion with as-fabricated fuel volume fraction of 43% and particle sizes in a well-defined normal distribution, more than 13000 sampling particles are simulated in the meat volume of 6 mm x 6 mm x 0.5 mm. The meat structure parameters are calculated as functions of layer thickness in the range from 0-16 μm. (authors)

Fischer-Tropsch synthesis: Support and cobalt cluster size effects on kinetics over Co/Al{sub 2}O{sub 3} and Co/SiO{sub 2} catalysts

Energy Technology Data Exchange (ETDEWEB)

Wenping Ma; Gary Jacobs; Dennis E. Sparks; Muthu K. Gnanamani; Venkat Ramana Rao Pendyala; Chia H. Yen; Jennifer L.S. Klettlinger; Thomas M. Tomsik; Burtron H. Davis [University of Kentucky, Lexington, KY (USA). Center for Applied Energy Research

2011-02-15

The influence of support type and cobalt cluster size (i.e., with average diameters falling within the range of 8-40 nm) on the kinetics of Fischer-Tropsch synthesis (FT) were investigated by kinetic tests employing a CSTR and two Co/{gamma}-Al{sub 2}O{sub 3} catalysts having different average pore sizes, and two Co/SiO{sub 2} catalysts prepared on the same support but having different loadings. A kinetic model -r{sub CO}=kP{sup a}{sub co}P{sup b}{sub H2}/(1 + mP{sub H2O}/P{sub H2}) that contains a water effect constant 'm' was used to fit the experimental data obtained with all four catalysts. Kinetic parameters suggest that both support type and average Co particle size impact FT behavior. Cobalt cluster size influenced kinetic parameters such as reaction order, rate constant, and the water effect parameter.Decreasing the average Co cluster diameter by about 30% led to an increase in the intrinsic reaction rate constant k, defined on a per g of catalyst basis, by 62-102% for the {gamma}-Al{sub 2}O{sub 3} and SiO{sub 2}-supported cobalt catalysts. Moreover, less inhibition by adsorbed CO and greater H{sub 2} dissociation on catalysts having smaller Co particles was suggested by the higher a and lower b values obtained for the measured reaction orders. Irrespective of support type, the catalysts having smaller average Co particles were more sensitive to water. Comparing the catalysts having strong interactions between cobalt and support (Co/Al{sub 2}O{sub 3}) to the ones with weak interactions (Co/SiO{sub 2}), the water effect parameters were found to be positive (indicating a negative influence on CO conversion) and negative (denoting a positive effect on CO conversion), respectively. Greater a and a/b values were observed for both Al{sub 2}O{sub 3}-supported Co catalysts, implying greater inhibition of the FT rate by strongly adsorbed CO on Co/Al{sub 2}O{sub 3} relative to Co/SiO{sub 2}. 78 refs., 4 figs., 3 tabs.

Neutronic Analysis of the RSG-GAS Compact Core without CIP Silicide 3.55 g U/cc and 4.8 g U/cc

International Nuclear Information System (INIS)

Jati S; Lily S; Tukiran S

2004-01-01

Fuel conversion from U 3 O 8 -Al to U 3 Si 2 -Al 2.96 g U/cc density in the RSG-GAS core had done successfully step by step since 36 th core until 44 th core. So that, since the 45 th core until now (48 th core) had been using full of silicide 2.96 g U/cc. Even though utilization program of silicide fuel with high density (3.55 g U/cc and 4.8 g U/cc) and optimize operation of RSG-GAS core under research. Optimalitation of core with increasing operation cycle have been analyzing about compact core. The mean of compact core is the RSG-GAS core with decrease number of IP or CIP position irradiation. In this research, the neutronic calculation to cover RSG-GAS core and RSG-GAS core without CIP that are using U 3 Si 2 -Al 2.96 g U/cc, 3.55 g U/cc and 4.8 g U/cc had done. Two core calculation done at 15 MW power using SRAC-ASMBURN code. The calculation result show that fuel conversion from 2.96 g U/cc density to 3.55 g U/cc and 4.8 g U/cc will increasing cycle length for both RSG-GAS core and RSG-GAS compact core without CIP. However, increasing of excess reactivity exceeded from nominal value of first design that 9.2%. Change of power peaking factor is not show significant value and still less than 1.4. Core fuelled with U 3 Si 2 -Al 4.8 g U/cc density have maximum discharge burn-up which exceeded from licensing value (70%). RSG-GAS compact core without CIP fuelled U 3 Si 2 -Al 2.96 g U/cc have longer cycle operation then RSG-GAS core and fulfil limitation neutronic parameter at the first design value. (author)

The conversion of NRU from HEU to LEU fuel

International Nuclear Information System (INIS)

Sears, D.F.; Atfield, M.D.; Kennedy, I.C.

1990-01-01

The program at Chalk River Nuclear Laboratories (CRNL) to develop and test low-enriched uranium fuel (LEU, 3 Si, USiAl, USi Al and U 3 Si 2 (U-3.96 wt% Si; U-3.5 wt% Si-1.5 wt% AL; U-3.2 wt%; Si-3 wt% Al; U-7.3 wt% Si, respectively). Fuel elements were fabricated with uranium loadings suitable for NRU, 3.15 gU/cm 3 , and for NRX, 4.5 gU/cm 3 , and were irradiated under normal fuel-operating conditions. Eight experimental irradiations involving 100 mini-elements and 84 full-length elements (7X12-element rods) were completed to qualify the LEU fuel and the fabrication technology. Post irradiation examinations confirmed that the performance of the LEU fuel, and that of a medium enrichment uranium (MEU, 45% U-235) alloy fuel tested as a back-up, was comparable to the HEU fuel. The uranium silicide dispersion fuel swelling was approximately linear up to burnups exceeding NRU's design terminal burnup (80 at%). NRU was partially converted to LEU fuel when the first 31 prototype fuel rods manufactured with industrial scale production equipment were installed in the reactor. The rods were loaded in NRU at a fuelling rate of about two rods per week over the period 1988 September to December. This partial LEU core (one third of a full NRU core) has allowed the reactor engineers and physicists to evaluate the bulk effects of the LEU conversion on NRU operations. As expected, the irradiation is proceeding without incident

Study on preparation of orange-emitting phosphor Y3Mg2AlSi2O12: Ce3+ for wLED

Directory of Open Access Journals (Sweden)

Yan Shirun

2017-12-01

Full Text Available Ce3+-doped garnet-structured orange-emitting phosphor Y3Mg2AlSi2O12:Ce3+ was prepared by sol-gel combustion using urea as a fuel.Effects of the reduction temperature,Ce3+ doping concentration on the structure,morphology,and photoluminescence property of the as-prepared phosphor were investigated by X-ray diffraction(XRD,scaning electron microscope(SEM,photoluminescence spectroscopy and UV-Vis reflection spectroscopy.The crystallinities,morphologies,and photoluminescence properties of the phosphors prepared by sol-gel combustion and solid-state reaction were compared.The reasons causing different performance of the phosphors were discussed.

Crack-resistant Al2O3-SiO2 glasses.

Science.gov (United States)

Rosales-Sosa, Gustavo A; Masuno, Atsunobu; Higo, Yuji; Inoue, Hiroyuki

2016-04-07

Obtaining "hard" and "crack-resistant" glasses have always been of great important in glass science and glass technology. However, in most commercial glasses both properties are not compatible. In this work, colorless and transparent xAl2O3-(100-x)SiO2 glasses (30 ≤ x ≤ 60) were fabricated by the aerodynamic levitation technique. The elastic moduli and Vickers hardness monotonically increased with an increase in the atomic packing density as the Al2O3 content increased. Although a higher atomic packing density generally enhances crack formation in conventional oxide glasses, the indentation cracking resistance increased by approximately seven times with an increase in atomic packing density in binary Al2O3-SiO2 glasses. In particular, the composition of 60Al2O3 • 40SiO2 glass, which is identical to that of mullite, has extraordinary high cracking resistance with high elastic moduli and Vickers hardness. The results indicate that there exist aluminosilicate compositions that can produce hard and damage-tolerant glasses.

Mikrostruktur dan Karakterisasi Sifat Mekanik Lapisan Cr3C2-NiAl-Al2O3 Hasil Deposisi Dengan Menggunakan High Velocity Oxygen Fuel Thermal Spray Coating

Directory of Open Access Journals (Sweden)

Edy Riyanto

2012-03-01

Full Text Available Surface coating processing of industrial component with thermal spray coatings have been applied in many industrial fields. Ceramic matrix composite coating which consists of Cr3C2-Al2O3-NiAl had been carried out to obtain layers of material that has superior mechanical properties to enhance component performance. Deposition of CMC with High Velocity Oxygen Fuel (HVOF thermal spray coating has been employed. This study aims to determine the effect of powder particle size on the microstructure, surface roughness and hardness of the layer, by varying the NiAl powder particle size. Test results show NiAl powder particle size has an influence on the mechanical properties of CMC coating. Hardness of coating increases and surface roughness values of coating decrease with smaller NiAl particle size.  

PIE Report on the KOMO-3 Irradiation Test Fuels

International Nuclear Information System (INIS)

Park, Jong Man; Ryu, H. J.; Yang, J. H.

2009-04-01

In the KOMO-3, in-reactor irradiation test had been performed for 12 kinds of dispersed U-Mo fuel rods, a multi wire fuel rod and a tube fuel rod. In this report we described the PIE results on the KOMO-3 irradiation test fuels. The interaction layer thickness between fuel particle and matrix could be reduced by using a large size U-Mo fuel particle or introducing Al-Si matrix or adding the third element in the U-Mo particle. Monolithic fuel rod of multi-wire or tube fuel was also effective in reducing the interaction layer thickness

Comparative study of gamma ray shielding and some properties of PbO–SiO2–Al2O3 and Bi2O3–SiO2–Al2O3 glass systems

International Nuclear Information System (INIS)

Singh, K.J.; Kaur, Sandeep; Kaundal, R.S.

2014-01-01

Gamma-ray shielding properties have been estimated in terms of mass attenuation coefficient, half value layer and mean free path values, whereas, structural studies have been performed in terms of density, optical band gap, glass transition temperature and longitudinal ultrasonic velocity parameters. X-ray diffraction, UV–visible, DSC and ultrasonic techniques have been used to explore the structural properties of PbO–SiO 2 –Al 2 O 3 and Bi 2 O 3 –SiO 2 –Al 2 O 3 glass systems. - Highlights: • Bi 2 O 3 –SiO 2 –Al 2 O 3 and PbO–SiO 2 –Al 2 O 3 glasses can replace conventional concretes as gamma-ray shielding materials. • Gamma-ray shielding properties improve with the addition of heavy metals. • Rigidity deteriorates with the increase in the content of heavy metals. • Bi 2 O 3 –SiO 2 –Al 2 O 3 glass system is better than PbO–SiO 2 –Al 2 O 3 glass system in terms of gamma-ray shielding as well as structural properties

Microstructure of as-fabricated UMo/Al(Si) plates prepared with ground and atomized powder

Science.gov (United States)

Jungwirth, R.; Palancher, H.; Bonnin, A.; Bertrand-Drira, C.; Borca, C.; Honkimäki, V.; Jarousse, C.; Stepnik, B.; Park, S.-H.; Iltis, X.; Schmahl, W. W.; Petry, W.

2013-07-01

UMo-Al based fuel plates prepared with ground U8wt%Mo, ground U8wt%MoX (X = 1 wt%Pt, 1 wt%Ti, 1.5 wt%Nb or 3 wt%Nb) and atomized U7wt%Mo have been examined. The first finding is that that during the fuel plate production the metastable γ-UMo phases partly decomposed into two different γ-UMo phases, U2Mo and α'-U in ground powder or α″-U in atomized powder. Alloying small amounts of a third element to the UMo had no measurable effect on the stability of the γ-UMo phase. Second, the addition of some Si inside the Al matrix and the presence of oxide layers in ground and atomized samples is studied. In the case with at least 2 wt%Si inside the matrix a Silicon rich layer (SiRL) forms at the interface between the UMo and the Al during the fuel plate production. The SiRL forms more easily when an Al-Si alloy matrix - which is characterized by Si precipitates with a diameter ⩽1 μm - is used than when an Al-Si mixed powder matrix - which is characterized by Si particles with some μm diameter - is used. The presence of an oxide layer on the surface of the UMo particles hinders the formation of the SiRL. Addition of some Si into the Al matrix [7-11]. Application of a protective barrier at the UMo/Al interface by oxidizing the UMo powder [7,12]. Increase of the Mo content or use of UMo alloys with ternary element addition X (e.g. X = Nb, Ti, Pt) to stabilize the γ-UMo with respect to α-U or to control the UMo-Al interaction layer kinetics [9,12-24]. Use of ground UMo powder instead of atomized UMo powder [10,25] The points 1-3 are to limit the formation of the undesired UMo/Al layer. Especially the addition of Si into the matrix has been suggested [3,7,8,10,11,26,27]. It has been often mentioned that Silicon is efficient in reducing the Uranium-Aluminum diffusion kinetics since Si shows a higher chemical affinity to U than Al to U. Si suppresses the formation of brittle UAl4 which causes a huge swelling during the irradiation. Furthermore it enhances the

Ca2 Al2 SiO7 :Ce3+ phosphors for mechanoluminescence dosimetry.

Science.gov (United States)

Tiwari, Geetanjali; Brahme, Nameeta; Sharma, Ravi; Bisen, D P; Sao, Sanjay Kumar; Sahu, Ishwar Prasad

2016-12-01

A series of Ce 3+ ion single-doped Ca 2 Al 2 SiO 7 phosphors was synthesized by a combustion-assisted method at an initiating temperature of 600 °C. The samples were annealed at 1100 °C for 3 h and their X-ray diffraction patterns confirmed a tetragonal structure. The phase structure, particle size, surface morphology and elemental analysis were analyzed using X-ray diffraction (XRD), transmission electron microscope (TEM), scanning electron microscopy (SEM) and energy dispersive X-ray (EDX) spectroscopy techniques. Thermoluminescence (TL) intensity increased with increase in ultraviolet (UV) light exposure time up to 15 min. With further increase in the UV irradiation time the TL intensity decreases. The increase in TL intensity indicates that trap concentration increased with UV exposure time. A broad peak at 121 °C suggested the existence of a trapping level. The peak of mechanoluminescence (ML) intensity versus time curve increased linearly with increasing impact velocity of the moving piston. Mechanoluminescence intensity increased with increase in UV irradiation time up to 15 min. Under UV-irradiation excitation, the TL and ML emission spectra of Ca 2 Al 2 SiO 7 :Ce 3+ phosphor showed the characteristic emission of Ce 3+ peaking at 400 nm (UV-violet) and originating from the Ce 3+ transitions of 5d-4f ( 2 F 5/2 and 2 F 7/2 ). The photoluminescence (PL) emission spectra for Ca 2 Al 2 SiO 7 :Ce 3+ were similar to the ML/TL emission spectra. The mechanism of ML excitation and the suitability of the Ca 2 Al 2 SiO 7 :Ce 3+ phosphor for radiation dosimetry are discussed. Copyright © 2016 John Wiley & Sons, Ltd. Copyright © 2016 John Wiley & Sons, Ltd.

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

Preparation of mullite whiskers reinforced SiC/Al2O3 composites by microwave sintering

Directory of Open Access Journals (Sweden)

Wei Li

2016-12-01

Full Text Available Mullite whiskers reinforced SiC/Al2O3 composites were prepared by microwave sintering in a microwave chamber with TE666 resonant mode. Original SiC particles were coated with SiO2 using sol-gel processing and mixed with Al2O3 particles. Mullite was formed in the reaction between SiO2 and Al2O3. The isostatically pressed cylindrical pellets were sintered from 1350 °C to 1600 °C for 30 min. Physical and chemical responses were investigated by detecting changes in reflected power during the microwave sintering process. XRD was carried out to characterize the samples and showed that mullite could be formed at 1200 °C. Bridging of mullite whiskers between Al2O3 and SiC particles was observed by SEM and is due to a so-called local hot spot effect, which was the unique feature for microwave sintering. The optimized microwave sintering temperature was 1500 °C corresponding to the maximum amount of mullite whiskers within SiC/Al2O3 composites. The high electro-magnetic field enhanced the decomposition of mullite at higher temperatures above 1550 °C. The mechanical properties of mullite whiskers reinforced SiC/Al2O3 composites are much better than the SiC/Al2O3 composites without mullite whiskers.

A comparison of the metallurgical behaviour of dispersion fuels with uranium silicides and U6Fe as dispersants

International Nuclear Information System (INIS)

Nazare, S.

1984-01-01

In the past few years metallurgical studies have been carried out to develop fuel dispersions with U-densities up to 7.0 Mg U m -3 . Uranium silicides have been considered to be the prime candidates as dispersants; U 6 Fe being a potential alternative on account of its higher U-density. The objective of this paper is to compare the metallurgical behaviour of these two material combinations with regard to the following aspects: (1) preparation of the compounds U 3 Si, U 3 Si 2 and U 6 Fe; (2) powder metallurgical processing to miniature fuel element plates; (3) reaction behaviour under equilibrium conditions in the relevant portions of the ternary U-Si-Al and U-Fe-Al systems; (4) dimensional stability of the fuel plates after prolonged thermal treatment; (5) thermochemical behaviour of fuel plates at temperatures near the melting point of the cladding. Based on this data, the possible advantages of each fuel combination are discussed. (author)

Irradiation behavior of uranium-silicide dispersion fuels

International Nuclear Information System (INIS)

Hofman, G.L.; Neimark, L.A.

1984-01-01

This paper describes and analyzes the irradiation behavior of experimental fuel plates containing U 3 Si, U 3 Si-1.5 w/o Al, and U 3 Si 2 particulate fuel dispersed and clad in aluminum. The fuel is nominally 19.9%-enriched 235 U and the fuel volume fraction in the central ''meat'' section of the plates is approximately 33%. Sets of fuel plates were removed from the Oak Ridge Research reactor at burnup levels of 35, 83, and 94% 235 U depletion and examined at the Alpha-Gamma Hot-Cell Facility at Argonne National Laboratory. The results of the examination may be summarized as follows. The dimensional stability of the U 3 Si 2 and pure U 3 Si fuel was excellent throughout the entire burnup range, with uniform plate thickness increases up to a maximum of 4 mils at the highest burnup level (94% 235 U depletion). This corresponds to a meat volume increase of 11%. The swelling was partially due to solid fission products but to a larger extent to fission gas bubbles. The fission gas bubbles in U 3 Si 2 were small (submicrometer size) and very uniformly distributed, indicating great stability. To a large extent this was also the case for U 3 Si; however, larger bubbles ( 3 Si-1.5 w/o Al fuel became unstable at the higher burnup levels. Fission gas bubbles were larger than in the other two fuels and were present throughout the fuel particles. At 94% 235 U depletion, the formation of fission gas bubbles with diameters up to 20 mils caused the plates to pillow. It is proposed that aluminum in U 3 Si destabilizes fission gas bubble formation to the point of severe breakaway swelling in the prealloyed silicide fuel. (author)

U3Si2 behavior in H2O environments: Part II, pressurized water with controlled redox chemistry

Science.gov (United States)

Nelson, A. T.; Migdisov, A.; Wood, E. Sooby; Grote, C. J.

2018-03-01

Recent interest in U3Si2 as an advanced light water reactor fuel has driven assessment of numerous properties, but characterization of its response to H2O environments is sparse in available literature. The behavior of U3Si2 in H2O containing atmospheres is investigated and presented in a two-part series of articles. This work examines the behavior of U3Si2 following exposure to pressurized H2O at temperatures from 300 to 350 °C. Testing was performed using two autoclave configurations and multiple redox conditions. Use of solid state buffers to attain a controlled water chemistry is also presented as a means to test actinide-bearing systems. Buffers were used to vary the hydrogen concentration between 1 and 30 parts per million H2. Testing included UN, U3Si5, and UO2. Both UN and U3Si5 were found to rapidly pulverize in less than 50 h at 300 °C. Uranium dioxide was included as a control for the autoclave system, and was found to be minimally impacted by exposure to pressurized water at the conditions tested for extended time periods. Testing of U3Si2 at 300 °C found reasonable stability through 30 days in 1-5 ppm H2. However, pulverization was observed following 35 days. The redox condition of testing strongly affected pulverization. Characterization of the resulting microstructures suggests that the mechanism responsible for pulverization under more strongly reducing conditions differs from that previously identified. Hydride formation is hypothesized to drive this transition. Testing performed at 350 °C resulted in rapid pulverization of U3Si2 in under 50 h.

High-uranium-loaded U3O8-Al fuel element development program [contributed by N.M. Martin, ORNL

International Nuclear Information System (INIS)

Martin, M.M.

1993-01-01

The High-Uranium-Loaded U 3 O 8 -Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U 3 O 8 -Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U 3 O 8 -Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U 3 O 8 ). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U 3 O 8 and aluminum. (author)

Irradiation behaviour of a 500 mm long hollow U3Si fuel element irradiated under BLW conditions

International Nuclear Information System (INIS)

Feraday, M.A.; Chalder, G.H.; Cotnam, K.D.

1969-07-01

A 500 mm long Zircaloy-clad element of U 3 Si (4.3 wt% Si) containing a 13% central void was irradiated to an average burnup of 3600 MWd/tonne U at an average linear power output of 790 W/cm, in boiling water coolant at 55 bars pressure. A larger diameter increase (1.5%) at the mid-plane of the element than elsewhere was attributed to the reduced restraint imposed on the fuel in this area as a consequence of β annealing a section of the cold worked sheath. Diameter increases in the cold worked portions of the sheath (average 0.7%) were greater than in similar elements irradiated in pressurized water at 96 bars pressure the difference is attributed to higher linear power output of the element in this test. External swelling of the element before filling of the central void was complete is attributed to the higher silicon content of the fuel compared with previous tests. No reaction between U 3 Si and Zircaloy was observed at a fuel sheath interface temperature near 400 o C. (author)

High-uranium-loaded U3O8--Al fuel element development program

International Nuclear Information System (INIS)

Martin, M.M.

1978-01-01

The High-Uranium-Loaded U 3 O 8 --Al Fuel Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages

Interaction between uranium oxide alloyed with Al2O3·SiO2 and pyrocarbon coating during irradiation of micro fuel elements

International Nuclear Information System (INIS)

Chernikov, A.S.; Khromov, Y.F.; Svistunov, D.E.; Chuiko, E.E.

1989-01-01

The thermodynamics of the interaction between uranium oxide and carbon was previously studied in the presence of Al 2 O 3 ·SiO 2 , SiC, and UC 1.86 ; in this case, the quantity of the reacting substances does not have any effect on the attainment of the equilibrium state. Based on the obtained results, it is interesting to study the characteristic features of the interaction between the alloyed UO x cores (kernels) with the PyC-coating under the conditions involving irradiation of the micro fuel elements with thermal neutrons and the formation of solid fission products. The data concerning the characteristics of a micro fuel element (the weight of the core, its composition, etc.) are useful for carrying out a quantitative evaluation of the additives required for fixing the alkali-earth fission products by obtaining stable compounds of aluminosilicates with Ba, Sr, Rb, and Cs at different levels of depletion (burnup) of the oxide fuel. An analysis of the interaction processes in such a complex system as the irradiated alloyed uranium oxide fuel located in a micro fuel element is carried out by comparing the chemical potential of oxygen (RT ln P O 2 ) for the competing constituents of the system

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

Crack-resistant Al2O3–SiO2 glasses

Science.gov (United States)

Rosales-Sosa, Gustavo A.; Masuno, Atsunobu; Higo, Yuji; Inoue, Hiroyuki

2016-01-01

Obtaining “hard” and “crack-resistant” glasses have always been of great important in glass science and glass technology. However, in most commercial glasses both properties are not compatible. In this work, colorless and transparent xAl2O3–(100–x)SiO2 glasses (30 ≤ x ≤ 60) were fabricated by the aerodynamic levitation technique. The elastic moduli and Vickers hardness monotonically increased with an increase in the atomic packing density as the Al2O3 content increased. Although a higher atomic packing density generally enhances crack formation in conventional oxide glasses, the indentation cracking resistance increased by approximately seven times with an increase in atomic packing density in binary Al2O3–SiO2 glasses. In particular, the composition of 60Al2O3•40SiO2 glass, which is identical to that of mullite, has extraordinary high cracking resistance with high elastic moduli and Vickers hardness. The results indicate that there exist aluminosilicate compositions that can produce hard and damage-tolerant glasses. PMID:27053006

Effect of Y2O3 on microstructure and mechanical properties of hypereutectic Al-20% Si alloy

Institute of Scientific and Technical Information of China (English)

YANG Ya-feng; XU Chang-lin; WANG Hui-yuan; LIU Chang; JIANG Qi-chuan

2006-01-01

The effect of Y2O3 on the microstructure and mechanical properties of the hypereutectic Al-20%Si(mass fraction) alloy was investigated. The results show that, with the addition of Y2O3 into the Al-P-Ti-TiC modifier, the average size of primary silicon in th.e Al-20%Si alloy modified by Al-P-Ti-TiC-Y2O3 modifier (approximately 15μm or less) is significantly reduced, and the morphology of eutectic silicon changes from coarse acicular and plate like to refined fibrous. The Brinell hardness (HB189) and tensile strength (301 MPa) of Al-20%Si alloy modified by the Al-P-Ti-TiC-Y2O3 increase by 11.6% and 10.7%, respectively, for the alloys afrer heat treatment.

Synthesis of geopolymer from spent FCC: Effect of SiO2/Al2O<3 and Na2O/SiO2 molar ratios

Directory of Open Access Journals (Sweden)

Trochez, J. J.

2015-03-01

Full Text Available This paper assesses the feasibility of using a spent fluid catalytic cracking catalyst (SFCC as precursor for the production of geopolymers. The mechanical and structural characterization of alkali-activated SFCC binders formulated with different overall (activator + solid precursor SiO2/Al2O3 and Na2O/SiO2 molar ratios are reported. Formation of an aluminosilicate ‘geopolymer’ gel is observed under all conditions of activation used, along with formation of zeolites. Increased SiO2/Al2O3 induces the formation of geopolymers with reduced mechanical strength, for all the Na2O/SiO2 ratios assessed, which is associated with excess silicate species supplied by the activator. This is least significant at increased alkalinity conditions (higher Na2O/SiO2 ratios, as larger extents of reaction of the spent catalyst are achieved. SiO2/Al2O3 and Na2O/SiO2 ratios of 2.4 and 0.25, respectively, promote the highest compressive strength (67 MPa. This study elucidates the great potential of using SFCC as precursor to produce sustainable ceramic-like materials via alkali-activation.Este artículo estudia la factibilidad de usar un catalizador gastado del proceso de craqueo (SFCC para la producción de geopolímeros. Se evalúan las características mecánicas y estructurales de los geopolímeros producidos con diferentes relaciones molares (activador + precursor solido de SiO2/Al2O3 y Na2O/SiO2. La formación de un gel geopolimérico de tipo aluminosilicato se observa a las diferentes condiciones evaluadas, así como la formación de zeolitas. Un incremento en la relación SiO2/Al2O3 genera geopolímeros de baja resistencia mecánica, a las diferentes relaciones molares Na2O/SiO2 evaluadas, como consecuencia del exceso de especies silicato provenientes del activador. Este efecto es menos significativo al incrementar las condiciones de alcalinidad (mayores relaciones Na2O/SiO2, ya que un mayor grado de reacción del catalizador gastado es alcanzado. Las

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

Non destructive examination of UN / U-Si fuel pellets using neutrons (preliminary assessment)

Energy Technology Data Exchange (ETDEWEB)

Bourke, Mark Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Voit, Stewart Lancaster [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Losko, Adrian S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tremsin, Anton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-03-31

Tomographic imaging and diffraction measurements were performed on nine pellets; four UN/ U Si composite formulations (two enrichment levels), three pure U₃Si₅ reference formulations (two enrichment levels) and two reject pellets with visible flaws (to qualify the technique). The U-235 enrichments ranged from 0.2 to 8.8 wt.%. The nitride/silicide composites are candidate compositions for use as Accident Tolerant Fuel (ATF). The monophase U₃Si₅ material was included as a reference. Pellets from the same fabrication batches will be inserted in the Advanced Test Reactor at Idaho during 2016. The goal of the Advanced Non-destructive Fuel Examination work package is the development and application of non-destructive neutron imaging and scattering techniques to ceramic and metallic nuclear fuels. Data reported in this report were collected in the LANSCE run cycle that started in September 2015 and ended in March 2016. Data analysis is ongoing; thus, this report provides a preliminary review of the measurements and provides an overview of the characterized samples.

Synthesis of Nanoscale CaO-Al2O3-SiO2-H2O and Na2O-Al2O3-SiO2-H2O Using the Hydrothermal Method and Their Characterization

Directory of Open Access Journals (Sweden)

Jingbin Yang

2017-06-01

Full Text Available C-A-S-H (CaO-Al2O3-SiO2-H2O and N-A-S-H (Na2O-Al2O3-SiO2-H2O have a wide range of chemical compositions and structures and are difficult to separate from alkali-activated materials. Therefore, it is difficult to analyze their microscopic properties directly. This paper reports research on the synthesis of C-A-S-H and N-A-S-H particles with an average particle size smaller than 300 nm by applying the hydrothermal method. The composition and microstructure of the products with different CaO(Na2O/SiO2 ratios and curing conditions were characterized using XRD, the RIR method, FTIR, SEM, TEM, and laser particle size analysis. The results showed that the C-A-S-H system products with a low CaO/SiO2 ratio were mainly amorphous C-A-S-H gels. With an increase in the CaO/SiO2 ratio, an excess of Ca(OH2 was observed at room temperature, while in a high-temperature reaction system, katoite, C4AcH11, and other crystallized products were observed. The katoite content was related to the curing temperature and the content of Ca(OH2 and it tended to form at a high-temperature and high-calcium environment, and an increase in the temperature renders the C-A-S-H gels more compact. The main products of the N-A-S-H system at room temperature were amorphous N-A-S-H gels and a small amount of sodalite. An increase in the curing temperature promoted the formation of the crystalline products faujasite and zeolite-P. The crystallization products consisted of only zeolite-P in the high-temperature N-A-S-H system and its content were stable above 70%. An increase in the Na2O/SiO2 ratio resulted in more non-bridging oxygen and the TO4 was more isolated in the N-A-S-H structure. The composition and microstructure of the C-A-S-H and N-A-S-H system products synthesized by the hydrothermal method were closely related to the ratio of the raw materials and the curing conditions. The results of this study increase our understanding of the hydration products of alkali

Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

International Nuclear Information System (INIS)

Souza, Jose Antonio Batista de

2011-01-01

IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm 3 for U 3 Si 2 -Al dispersion-based and 2.3 gU/cm 3 for U 3 O 8 -Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm 3 in U 3 Si 2 -Al dispersion and 3.2 gU/cm 3 U 3 O 8 -Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U 3 Si 2 -Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U 3 O 8 -Al dispersion fuel plates with 3.2 gU/cm 3 showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U 3 Si 2 production at 4.8 gU/cm 3 , with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)

In vitro cytotoxicity of SiO2 or ZnO nanoparticles with different sizes and surface charges on U373MG human glioblastoma cells

Directory of Open Access Journals (Sweden)

Kim JE

2014-12-01

Full Text Available Jung-Eun Kim,1,* Hyejin Kim,1,* Seong Soo A An,2 Eun Ho Maeng,3 Meyoung-Kon Kim,4 Yoon-Jae Song1 1Department of Life Science, 2Department of Bionano Technology, Gachon University, Seongnam-Si, South Korea; 3Korea Testing and Research Institute, Seoul, South Korea; 4Department of Biochemistry and Molecular Biology, Korea University Medical School and College, Seoul, South Korea *These authors contributed equally to this work Abstract: Silicon dioxide (SiO2 and zinc oxide (ZnO nanoparticles are widely used in various applications, raising issues regarding the possible adverse effects of these metal oxide nanoparticles on human cells. In this study, we determined the cytotoxic effects of differently charged SiO2 and ZnO nanoparticles, with mean sizes of either 100 or 20 nm, on the U373MG human glioblastoma cell line. The overall cytotoxicity of ZnO nanoparticles against U373MG cells was significantly higher than that of SiO2 nanoparticles. Neither the size nor the surface charge of the ZnO nanoparticles affected their cytotoxicity against U373MG cells. The 20 nm SiO2 nanoparticles were more toxic than the 100 nm nanoparticles against U373MG cells, but the surface charge had little or no effect on their cytotoxicity. Both SiO2 and ZnO nanoparticles activated caspase-3 and induced DNA fragmentation in U373MG cells, suggesting the induction of apoptosis. Thus, SiO2 and ZnO nanoparticles appear to exert cytotoxic effects against U373MG cells, possibly via apoptosis. Keyword: apoptosis

Application of bias correction methods to improve U3Si2 sample preparation for quantitative analysis by WDXRF

International Nuclear Information System (INIS)

Scapin, Marcos A.; Guilhen, Sabine N.; Azevedo, Luciana C. de; Cotrim, Marycel E.B.; Pires, Maria Ap. F.

2017-01-01

The determination of silicon (Si), total uranium (U) and impurities in uranium-silicide (U 3 Si 2 ) samples by wavelength dispersion X-ray fluorescence technique (WDXRF) has been already validated and is currently implemented at IPEN's X-Ray Fluorescence Laboratory (IPEN-CNEN/SP) in São Paulo, Brazil. Sample preparation requires the use of approximately 3 g of H 3 BO 3 as sample holder and 1.8 g of U 3 Si 2 . However, because boron is a neutron absorber, this procedure precludes U 3 Si 2 sample's recovery, which, in time, considering routinely analysis, may account for significant unusable uranium waste. An estimated average of 15 samples per month are expected to be analyzed by WDXRF, resulting in approx. 320 g of U 3 Si 2 that would not return to the nuclear fuel cycle. This not only impacts in production losses, but generates another problem: radioactive waste management. The purpose of this paper is to present the mathematical models that may be applied for the correction of systematic errors when H 3 BO 3 sample holder is substituted by cellulose-acetate {[C 6 H 7 O 2 (OH) 3-m (OOCCH 3 )m], m = 0∼3}, thus enabling U 3 Si 2 sample’s recovery. The results demonstrate that the adopted mathematical model is statistically satisfactory, allowing the optimization of the procedure. (author)

Prediction of fracture strength in Al2O3/SiCp ceramic matrix nanocomposites

Directory of Open Access Journals (Sweden)

Z. Zhang and D.L. Chen

2007-01-01

Full Text Available Based primarily on a recent publication [S.M. Choi, H. Awaji, Sci. Tech. Adv. Mater. 6 (2005 2–10.], where the dislocations around the nano-sized particles in the intra-granular type of ceramic matrix nanocomposites (CMNCs were modeled, dislocation activities in Al2O3/SiCp CMNCs were discussed in relation to the processing conditions. The dislocations around the nano-sized particles, caused by the thermal mismatch between the ceramic matrix and nano-sized particles, were assumed to hold out the effect of Orowan-like strengthening, although the conventional Owowan loops induced by the movement of dislocations were unlikely in the ceramic matrix at room temperature. A model involving the yield strength of metal matrix nanocomposites (MMNCs, where the Owowan strengthening effect was taken into consideration, was thus modified and extended to predict the fracture strength of the intra-granular type of CMNCs without and with annealing. On the basis of the characteristics of dislocations in the CMNCs, the load-bearing effect and Orowan-like strengthening were considered before annealing, while the load-bearing effect and enhanced dislocation density strengthening were taken into account after annealing. The model prediction was found to be in agreement with the experimental data of Al2O3/SiCp nanocomposites reported in the literature.

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)

Estudo da viabilidade de obtenção de cerâmicas de SiC por infiltração espontânea de misturas eutéticas de Y2O3-AlN, Y2O3-Al2O3, R2O3-AlN Study of the viability to produce SiC ceramics by Y2O3-AlN, Y2O3-Al2O3, R2O3-AlN spontaneous infiltration

Directory of Open Access Journals (Sweden)

G. C. R. Garcia

2008-06-01

Full Text Available As cerâmicas de carbeto de silício, SiC, apresentam excelentes propriedades quando obtidas por infiltração de determinados líquidos. Na infiltração o tempo de contato entre o líquido e o SiC a temperaturas elevadas é muito curto, diminuindo a probabilidade de formação dos produtos gasosos que interferem negativamente na resistência da peça final, como ocorre na sinterização via fase líquida. O objetivo deste trabalho é mostrar uma correlação entre molhabilidade e capacidade de infiltração de alguns aditivos em compactos de SiC. Foram preparados compactos de SiC por prensagem isostática a frio e posterior pré-sinterização via fase sólida. Nesses compactos foram infiltradas misturas de Y2O3-AlN, Y2O3-Al2O3, R2O3-AlN, nas composições eutéticas, 10 ºC acima da temperatura de fusão das respectivas misturas por 4, 8 e 12 min. Após infiltração, as amostras foram analisadas quanto à densidade aparente e real, fases cristalinas, microestrutura e grau de infiltração, sendo que as amostras infiltradas com Y2O3-AlN apresentaram melhores resultados.Silicon carbide ceramics, SiC, obtained by liquid infiltration have shown excellent properties. In infiltration process the contact time of the liquid with SiC at elevated temperature is short, decreasing the probability to form gaseous products that contribute negatively in the final product properties. This phenomenon occurs during SiC liquid phase sintering. The purpose of the present study was to investigate the correlation between wettability and infiltration tendency of some additives in SiC compacts. SiC compacts were prepared by cold isostatic pressing followed by solid phase pre-sintering. Into the compacts were introduced Y2O3-AlN, Y2O3-Al2O3, R2O3-AlN liquids with eutectic compositions at a temperature 10 ºC higher than the melting point of each mixture for 4, 8 and 12 min. Before infiltration, the samples were analyzed by determining densities, crystalline phases

Exploration of R₂XM₂ (R=Sc, Y, Ti, Zr, Hf, rare earth; X=main group element; M=transition metal, Si, Ge): Structural Motifs, the novel Compound Gd₂AlGe₂ and Analysis of the U₃Si₂ and Zr₃Al₂ Structure Types

Energy Technology Data Exchange (ETDEWEB)

McWhorter, Sean William [Iowa State Univ., Ames, IA (United States)

2006-01-01

electrons can be accounted for. A series of electronic structure calculations were performed on models with the U₃Si₂ and Zr₃Al₂ structures, using Zr and A1 as the building blocks. The starting point for these models was the U₃Si₂ structure, and models were created to simulate the transition from the idealized U₃Si₂ structure to the distorted Zr₃Al₂ structure. Analysis of the band structures of the models has shown that the transition from the U₃Si₂ structure to the Zr₃Al₂ structure lifts degeneracies along the Λ → Z direction, indicating a Peierls-type mechanism for the displacement occurring in the positions of the Zr atoms.

Irradiation behaviour of a 500 mm long hollow U{sub 3}Si fuel element irradiated under BLW conditions

Energy Technology Data Exchange (ETDEWEB)

Feraday, M A; Chalder, G H; Cotnam, K D

1969-07-15

A 500 mm long Zircaloy-clad element of U{sub 3}Si (4.3 wt% Si) containing a 13% central void was irradiated to an average burnup of 3600 MWd/tonne U at an average linear power output of 790 W/cm, in boiling water coolant at 55 bars pressure. A larger diameter increase (1.5%) at the mid-plane of the element than elsewhere was attributed to the reduced restraint imposed on the fuel in this area as a consequence of {beta} annealing a section of the cold worked sheath. Diameter increases in the cold worked portions of the sheath (average 0.7%) were greater than in similar elements irradiated in pressurized water at 96 bars pressure the difference is attributed to higher linear power output of the element in this test. External swelling of the element before filling of the central void was complete is attributed to the higher silicon content of the fuel compared with previous tests. No reaction between U{sub 3}Si and Zircaloy was observed at a fuel sheath interface temperature near 400{sup o}C. (author)

TEM Nanostructural Study of Al-6Si-3Cu-Mg Melt-Spun Ribbons

Directory of Open Access Journals (Sweden)

Ismeli Alfonso López

2008-01-01

Full Text Available Three quaternary Al-6Si-3Cu-xMg (x = 0.59, 3.80, and 6.78 wt.% alloys were produced by melt-spun and characterized using X-ray diffractometry (XRD, transmission electron microscopy (TEM, and microhardness techniques. Obtained second phases were Al2Cu( for the alloy with 0.59% Mg and Al5Cu2Mg8Si6 (Q for the alloys with 3.80 and 6.78% Mg. These phases are present as 30–50 nm or as 5–10 nm nanoparticles. Alloying elements content in solid solution increased, mainly for Si and Mg. The high alloying elements content in solid solution and the small -Al cell size for melt-spun alloys leads to microhardness values about 2 times higher than those of ingot counterparts. The microhardness increase for melt-spun alloys with 3.80 and 6.78% Mg depends on Mg content in solid solution.

Microstructural analysis of as-processed U-10 wt.%Mo monolithic fuel plate in AA6061 matrix with Zr diffusion barrier

Energy Technology Data Exchange (ETDEWEB)

Perez, E.; Yao, B. [Advanced Materials Processing and Analysis Center, Department of Mechanical, Materials and Aerospace Engineering, University of Central Florida, Orlando, FL 32816 (United States); Keiser, D.D. [Nuclear Fuels and Materials Division, Idaho National Laboratory, Scoville, ID 83415 (United States); Sohn, Y.H., E-mail: ysohn@mail.ucf.ed [Advanced Materials Processing and Analysis Center, Department of Mechanical, Materials and Aerospace Engineering, University of Central Florida, Orlando, FL 32816 (United States)

2010-07-01

For higher U-loading in low-enriched U-10 wt.%Mo fuels, monolithic fuel plate clad in AA6061 is being developed as a part of Reduced Enrichment for Research and Test Reactor (RERTR) program. This paper reports the first characterization results from a monolithic U-10 wt.%Mo fuel plate with a Zr diffusion barrier that was fabricated as part of a plate fabrication campaign for irradiation testing in the Advanced Test Reactor (ATR). Both scanning and transmission electron microscopy (SEM and TEM) were employed for analysis. At the interface between the Zr barrier and U-10 wt.%Mo, going from Zr to U(Mo), UZr{sub 2}, {gamma}-UZr, Zr solid-solution and Mo{sub 2}Zr phases were observed. The interface between AA6061 cladding and Zr barrier plate consisted of four layers, going from Al to Zr, (Al, Si){sub 2}Zr, (Al, Si)Zr{sub 3} (Al, Si){sub 3}Zr, and AlSi{sub 4}Zr{sub 5}. Irradiation behavior of these intermetallic phases is discussed based on their constituents. Characterization of as-fabricated phase constituents and microstructure would help understand the irradiation behavior of these fuel plates, interpret post-irradiation examination, and optimize the processing parameters of monolithic fuel system.

Microstructural analysis of as-processed U-10 wt.%Mo monolithic fuel plate in AA6061 matrix with Zr diffusion barrier

Science.gov (United States)

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

2010-07-01

For higher U-loading in low-enriched U-10 wt.%Mo fuels, monolithic fuel plate clad in AA6061 is being developed as a part of Reduced Enrichment for Research and Test Reactor (RERTR) program. This paper reports the first characterization results from a monolithic U-10 wt.%Mo fuel plate with a Zr diffusion barrier that was fabricated as part of a plate fabrication campaign for irradiation testing in the Advanced Test Reactor (ATR). Both scanning and transmission electron microscopy (SEM and TEM) were employed for analysis. At the interface between the Zr barrier and U-10 wt.%Mo, going from Zr to U(Mo), UZr 2, γ-UZr, Zr solid-solution and Mo 2Zr phases were observed. The interface between AA6061 cladding and Zr barrier plate consisted of four layers, going from Al to Zr, (Al, Si) 2Zr, (Al, Si)Zr 3 (Al, Si) 3Zr, and AlSi 4Zr 5. Irradiation behavior of these intermetallic phases is discussed based on their constituents. Characterization of as-fabricated phase constituents and microstructure would help understand the irradiation behavior of these fuel plates, interpret post-irradiation examination, and optimize the processing parameters of monolithic fuel system.

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.

Full core operation in JRR-3 with LEU fuels

International Nuclear Information System (INIS)

Murayama, Y.; Issiki, M.

1995-01-01

The new JRR-3 a 20MWT swimming pool type research reactor, is made up of plate type LEU fuel elements with U-Al x fuel at 2.2 gU/cm 3 . Reconstruction work for the new JR-3 was a good success, and common operation started in November 1990, and 7 cycles (26 days operation/cycle) have passed. We have no experience in using such a high uranium density fuel element with aluminide fuel. So we plan to examine the condition of the irradiated fuel elements with three methods, that is, measurement of the value of FFD in operation, observation of external view of the fuels in refueling work and postirradiation examination after maximum burn-up will be established. In the results of the first two methods, the fuel elements of JRR-3 is burned up normally and have no evidence of failure. (author)

Radiation Resistance of the U(Al, Si)(3) Alloy: Ion-Induced Disordering

Czech Academy of Sciences Publication Activity Database

Meshi, L.; Yaniv, G.; Horák, Pavel; Vacík, Jiří; Mykytenko, N.; Rafailov, G.; Dahan, I.; Fuks, D.; Kiv, A.

2018-01-01

Roč. 11, č. 2 (2018), č. článku 228. ISSN 1996-1944 R&D Projects: GA ČR(CZ) GBP108/12/G108 Institutional support: RVO:61389005 Keywords : U-Al-Si * ion-irradiation * transmission electron microscopy * structural defects * disordering Subject RIV: JI - Composite Materials OBOR OECD: Composites (including laminates, reinforced plastics, cermets, combined natural and synthetic fibre fabrics Impact factor: 2.654, year: 2016

Study on Microstructure and Mechanical Properties of Hypereutectic Al-18Si Alloy Modified with Al-3B.

Science.gov (United States)

Gong, Chunjie; Tu, Hao; Wu, Changjun; Wang, Jianhua; Su, Xuping

2018-03-20

An hypereutectic Al-18Si alloy was modified via an Al-3B master alloy. The effect of the added Al-3B and the modification temperature on the microstructure, tensile fracture morphologies, and mechanical properties of the alloy were investigated using an optical microscope, Image-Pro Plus 6.0, a scanning electron microscope, and a universal testing machine. The results show that the size of the primary Si and its fraction decreased at first, and then increased as an additional amount of Al-3B was added. When the added Al-3B reached 0.2 wt %, the fraction of the primary Si in the Al-18Si alloy decreased with an increase in temperature. Compared with the unmodified Al-18Si alloy, the tensile strength and elongation of the alloy modified at 850 °C with 0.2 wt % Al-3B increased by 25% and 81%, respectively. The tensile fracture of the modified Al-18Si alloy exhibited partial ductile fracture characteristics, but there were more areas with ductile characteristics compared with that of the unmodified Al-18Si alloy.

Liquidus Temperature of SrO-Al2O3-SiO2 Glass-Forming Compositions

DEFF Research Database (Denmark)

Abel, Brett M.; Morgan, James M.; Mauro, John C.

2013-01-01

. In the composition range of interest for industrial glasses, Tliq tends to decrease with increasing strontium-to-alumina ratio. We find that cristobalite, mullite, and slawsonite are the dominant devitrification phases for the compositions with high SiO2, SiO2+Al2O3, and SrO contents, respectively. By comparison...... with the phase diagrams for CaO-Al2O3-SiO2 and MgO-Al2O3-SiO2 systems, we have found that for the highest [RO]/[Al2O3] ratios, Tliq exhibits a minimum value for R = Ca. Based on the phase diagram established here, the composition of glass materials, for example, for liquid crystal display substrates, belonging...... to the SrO-Al2O3-SiO2 family may be designed with a more exact control of the glass-forming ability by avoiding the regions of high liquidus temperature....

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)

UV and IR laser induced ablation of Al2O3/SiN:H and a-Si:H/SiN:H

Directory of Open Access Journals (Sweden)

Schutz-Kuchly T.

2014-01-01

Full Text Available Experimental work on laser induced ablation of thin Al2O3(20 nm/SiN:H (70 nm and a-Si:H (20 nm/SiN:H (70 nm stacks acting, respectively, as p-type and n-type silicon surface passivation layers is reported. Results obtained using two different laser sources are compared. The stacks are efficiently removed using a femtosecond infra-red laser (1030 nm wavelength, 300 fs pulse duration but the underlying silicon surface is highly damaged in a ripple-like pattern. This collateral effect is almost completely avoided using a nanosecond ultra-violet laser (248 nm wavelength, 50 ns pulse duration, however a-Si:H flakes and Al2O3 lace remain after ablation process.

Wearproof composition coatings on the basis of SiC-AL2O3 for restoration and reiforcement of the components of aircraft ground support equipment

Directory of Open Access Journals (Sweden)

О. П. Уманський

2013-07-01

Full Text Available On the ground of research of a contact interaction of the melts of the system Ni–Al with the ceramics of SiC–Al2O3 content, the possibility of wearproof coating deposition of the system SiC–Al2O3–Ni–Al by gas-flame techniques has been proved. Technological features of their acquisition also have been studied. The structure of coatings from composition material that contains the SiC–Al2O3 wearproof component and Ni–Al metallic binder, deposited by the method of high velocity air fuel deposition (HVAF on medium-carbon steel steels has been researched. Tribotechnical descriptions of the deposited coatings under the conditions of friction without lubricating materials in the air environment in wide range of speed-load modes of the “pin–on–disk” layout have been studied. The features and regularities of their wear mechanisms retaining the constant speed and constant load have been determined

Comparison of irradiation behavior of different uranium silicide dispersion fuel element designs

International Nuclear Information System (INIS)

Hofman, G.L.; Rest, J.; Snelgrove, J.L.

1995-01-01

Calculations of fuel swelling of U 3 SiAl-Al and U 3 Si 2 were performed for various dispersion fuel element designs. Breakaway swelling criteria in the form of critical fuel volume fractions were derived with data obtained from U 3 SiAl-Al plate irradiations. The results of the analysis show that rod-type elements remain well below the pillowing threshold. However, tubular fuel elements, which behave essentially like plates, will likely develop pillows or blisters at around 90% 235 U burnup. The U 3 Si 2 -Al compounds demonstrate stable swelling behavior throughout the entire burnup range for all fuel element designs

Microstructure and Mechanical Properties of Multiphase Strengthened Al/Si/Al_2O_3/SiO_2/MWCNTs Nano composites Sintered by In Situ Vacuum Hot Pressing

International Nuclear Information System (INIS)

Li, J.; Jiang, X.; Zhu, D.; Zhu, M.; Shao, Z.; Johnson, S.; Luo, Z.

2015-01-01

Eutectic Al/Si binary alloy is technically one of the most important Al casting alloys due to its high corrosion resistance, evident shrinkage reduction, low thermal expansion coefficient, high fluidity, and good weldability. In this work, multi phased Al/Si matrix nano composites reinforced with Al_2O_3 and multi walled carbon nano tubes (MWCNTs) have been sintered by an in situ vacuum hot-pressing method. The alumina Al_2O_3 nanoparticles were introduced by an in situ reaction of Al with SiO_2. Microstructure and mechanical properties of the sintered Al/Si/Al_2O_3/SiO_2/MWCNTs nano composites with different alumina contents were investigated. The mechanical properties were determined by micro-Vickers hardness and compressive and shear strength tests. The results demonstrated that in situ alumina and MWCNTs had impacts on microstructure and mechanical properties of the nano composites. Based on the mechanical properties and microstructure of the nano composites, strengthening and fracture mechanisms by multiple reinforcements were analyzed

Crystallization kinetics of BaO-Al2O3-SiO2 glasses

Science.gov (United States)

Bansal, Narottam P.; Hyatt, Mark J.

1989-01-01

Barium aluminosilicate glasses are being investigated as matrix materials in high-temperature ceramic composites for structural applications. Kinetics of crystallization of two refractory glass compositions in the barium aluminosilicate system were studied by differential thermal analysis (DTA), X-ray diffraction (XRD), and scanning electron microscopy (SEM). From variable heating rate DTA, the crystallization activation energies for glass compositions (wt percent) 10BaO-38Al2O3-51SiO2-1MoO3 (glass A) and 39BaO-25Al2O3-35SiO2-1MoO3 (glass B) were determined to be 553 and 558 kJ/mol, respectively. On thermal treatment, the crystalline phases in glasses A and B were identified as mullite (3Al2O3-2SiO2) and hexacelsian (BaO-Al2O3-2SiO2), respectively. Hexacelsian is a high-temperature polymorph which is metastable below 1590 C. It undergoes structural transformation into the orthorhombic form at approximately 300 C accompanied by a large volume change which is undesirable for structural applications. A process needs to be developed where stable monoclinic celsian, rather than hexacelsian, precipitates out as the crystal phase in glass B.

Development of U6Fe-Al dispersions for the use of LEU in research and test reactors

International Nuclear Information System (INIS)

Nazare, S.

1983-01-01

For some time now, efforts are being made to develop fuel dispersions that would permit the use of low (approx. 20% 235-U) enriched uranium (LEU) instead of the currently used highly (approx. 93% 235-U) enriched uranium (HEU) in research and test reactors. Since penalties in the performance of the reactor have to be avoided, the 235-U content in the dispersion has at least to be retained at current levels. On account of their high U-densities, the major development effort has been focussed on the uranium silicides (U 3 Si, U 3 Si(Al), and U 3 Si 2 -based dispersions). With silicides as dispersants, it is possible to fabricate fuel element plates with U-densities in the dispersion of about 6.0 gU/cm 3 . In comparison to the silicides, the U 6 Fe-phase offers several advantages namely: higher U-density (approx. 17.0 gU/cm 3 ); relative ease of formation compared to U 3 Si; possible advantages with regard to reprocessing of the spent fuel due to the absence of silicon. The studies outlined here were performed with a view to investigating the preparation, reaction behavior and dimensional stability after heat treatment of U 6 Fe-Al dispersions

Development of U6Fe-Al dispersions for the use of LEU in research and test reactors

International Nuclear Information System (INIS)

Nazare, S.

1983-01-01

For some time now, efforts are being made to develop fuel dispersions that would permit the use of low (∼ 20% 235-U) enriched uranium (LEU) instead of the currently used highly (∼ 93% 235-U) enriched uranium (HEU) in research and test reactors. Since penalties in the performance of the reactor have to be avoided, the 235-U content in the dispersion has at least to be retained at current levels. On account of their high U-densities, the major development effort has been focussed on the uranium silicides [U 3 Si, U 3 Si(Al), and U 3 Si 2 - based dispersions. With silicides as dispersants, it is possible to fabricate fuel element plates with U-densities in the dispersion of about 6.0 g U/cm 3 . In comparison to the silicides, the U 6 Fe-phase offers several advantages namely: - higher U-density (∼ 17.0 g U/cm 3 ); - relative ease of formation compared to U 3 Si; - possible advantages with regard to reprocessing of the spent fuel due to the absence of silicon. The studies outlined here were therefore performed with a view to investigating the preparation, reaction behaviour and dimensional stability after heat treatment of U 6 Fe-Al dispersions

Photoluminescence of the Mg2Al4Si5O18-Al2O3-MgAl2O4-SiO2 ceramic system containing Fe3+ and Cr3+ as impurity ions

Science.gov (United States)

Sosman, L. P.; López, A.; Pedro, S. S.; Papa, A. R. R.

2018-02-01

This work presents the results of photoluminescence, excitation and radiative decay time for a ceramic system containing Mg2Al4Si5O18-Al2O3-MgAl2O4-SiO2 with Fe3+ and Cr3+ as impurity ions. Emission data were obtained using several excitation wavelengths and the excitation data were acquired for the most intense emission bands. The optical results were analyzed according to the Tanabe-Sugano (TS) theory from which the crystalline field parameter Dq and Racah parameters B and C were obtained for the Fe3+ and Cr3+ sites. The results indicate that the Fe3+ and Cr3+ ions occupy tetrahedral and octahedral sites, respectively. The emission from Fe3+ and Cr3+ ions causes an intense and broad band ranging between 350 nm and 850 nm, showing that this material is a potential tunable radiation source at room temperature.

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}.

Synergetic Effect of Ni2P/SiO2 and γ-Al2O3 Physical Mixture in Hydrodeoxygenation of Methyl Palmitate

Directory of Open Access Journals (Sweden)

Ivan V. Shamanaev

2017-11-01

Full Text Available The Ni2P/SiO2 catalyst, which was prepared by in situ temperature-programmed reduction and in the mixture with the inert (SiC, SiO2 or acidic (γ-Al2O3 material was studied in methyl palmitate hydrodeoxygenation (HDO. Methyl palmitate HDO was carried out at temperatures of 270–330 °C, H2/feed volume ratio of 600 Nm3/m3, and H2 pressure of 3.0 MPa. Ni2P/SiO2 catalyst, diluted with γ-Al2O3 showed a higher activity than Ni2P/SiO2 catalyst diluted with SiC or SiO2. The conversion of methyl palmitate increased significantly in the presence of γ-Al2O3 most probably due to the acceleration of the acid-catalyzed reaction of ester hydrolysis. The synergism of Ni2P/SiO2 and γ-Al2O3 in methyl palmitate HDO can be explained by the cooperation of the metal sites of Ni2P/SiO2 and the acid sites of γ-Al2O3 in consecutive metal-catalyzed and acid-catalyzed reactions of HDO. The obtained results let us conclude that the balancing of metal and acid sites plays an important role in the development of the efficient catalyst for the HDO of fatty acid esters over supported phosphide catalysts.

Interaction Layer Characteristics in U-xMo Dispersion/Monolithic Fuels

International Nuclear Information System (INIS)

Porter, D.L.

2010-01-01

Published data concerning the interaction layer (IL) formed between U-xMo fuel alloy and aluminum (Al)-based matrix or cladding materials was reviewed, including the effects of silicon (Si) content in the matrix/cladding, molybdenum (Mo) content in the fuel, pre irradiation thermal treatments, irradiation, and test temperature. The review revealed that tests conducted in the laboratory produce results different from those conducted in an irradiation environment. However, the laboratory testing relates well to thermal treatments performed prior to irradiation and helps in understanding the effects that these pre irradiation treatments have on in reactor performance. A pre-formed, Si-enriched IL seems to be important in delaying the onset of rapid growth of fission gas bubbles at low irradiation temperatures. Several other conclusions can be drawn: (1) An IL with phases akin to UAl3 is desired for optimum fuel performance, but at low temperatures, and especially in an irradiation atmosphere, the desired (Al+Si)/(U+Mo) ratio of three is difficult to produce. When the fuel operating temperature is low, it is important to create a pre-irradiation IL, enriched in Si. This pre-formed IL is relatively stable, performs well in terms of swelling resistance, and prevents rapid IL growth during irradiation. (2) At higher operating temperatures (>150-170 C), IL formation in reactor may not be so dependent on pre-irradiation IL formation, especially at high burnup; a pre-fabricated IL seems to be less stable at high burnup and high operating temperature. Moreover, the (Al+SI)/(U+Mo) ratio of three occurs more often at higher temperature. For these two reasons, it is important at high operating temperature to also have a matrix with significant Si content to create an IL in reactor with the right characteristics. (3) Out-of-reactor testing seems to indicate that Si in the matrix material is required in some concentration (2%, 5%, ?) to provide for a thin, Si-enriched IL formed

Interaction Layer Characteristics in U-xMo Dispersion/Monolithic Fuels

Energy Technology Data Exchange (ETDEWEB)

D. L. Porter

2010-11-01

Published data concerning the interaction layer (IL) formed between U-xMo fuel alloy and aluminum (Al)-based matrix or cladding materials was reviewed, including the effects of silicon (Si) content in the matrix/cladding, molybdenum (Mo) content in the fuel, pre irradiation thermal treatments, irradiation, and test temperature. The review revealed that tests conducted in the laboratory produce results different from those conducted in an irradiation environment. However, the laboratory testing relates well to thermal treatments performed prior to irradiation and helps in understanding the effects that these pre irradiation treatments have on in reactor performance. A pre-formed, Si-enriched IL seems to be important in delaying the onset of rapid growth of fission gas bubbles at low irradaiiation temperatures. Several other conclusions can be drawn: 1. An IL with phases akin to UAl3 is desired for optimum fuel performance, but at low temperatures, and especially in an irradiation atmosphere, the desired (Al+Si)/(U+Mo) ratio of three is difficult to produce. When the fuel operating temperature is low, it is important to create a pre-irradiation IL, enriched in Si. This pre-formed IL is relatively stable, performs well in terms of swelling resistance, and prevents rapid IL growth during irradiation. 2. At higher operating temperatures (>150–170°C), IL formation in reactor may not be so dependent on pre-irradiation IL formation, especially at high burnup; a pre-fabricated IL seems to be less stable at high burnup and high operating temperature. Moreover, the (Al+SI)/(U+Mo) ratio of three occurs more often at higher temperature. For these two reasons, it is important at high operating temperature to also have a matrix with significant Si content to create an IL in reactor with the right characteristics. 3. Out-of-reactor testing seems to indicate that Si in the matrix material is required in some concentration (2%, 5%, ?) to provide for a thin, Si-enriched IL formed

Post-irradiation examination of U3SIX-AL fuel element manufactured and irradiated in Argentina

International Nuclear Information System (INIS)

Ruggirello, Gabriel; Calabroni, Hector; Sanchez, Miguel; Hofman, Gerard

2002-01-01

As a part of CNEA's qualification program as a supplier of low enriched Al-U 3 Si 2 dispersion fuel elements for research reactors, a post irradiation examination (PIE) of the first prototype of this kind, called P-04, manufactured and irradiated in Argentina, was carried out. The main purpose of this work was to set up various standard PIE techniques in the hot cell, looking forward to the next steps of the qualification program, as well as to acquire experience on the behaviour of this nuclear material and on the control of the manufacturing process. After an appropriate cooling period, on May 2000 the P-04 was transported to the hot cell in Ezeiza Atomic Centre. Non destructive and destructive tests were performed following the PIE procedures developed in Argonne National Laboratory (ANL), this mainly included dimensional measurement, microstructural observations and chemical burn-up analyses. The methodology and results of which are outlined in this report. The results obtained show a behaviour consistent with that of other fuel elements of the same kind, tested previously. On the other hand the results of this PIE, specially those concerning burn-up analysis and stability and corrosion behaviour of the fuel plates, will be of use for the IAEA Regional Program on the characterization of MTR spent fuel. (author)

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