WorldWideScience

Sample records for fuel contribucion al

  1. José Manuel Esteve : sus contribuciones al estudio de la profesión docente

    Directory of Open Access Journals (Sweden)

    Julio Vera

    2013-07-01

    Full Text Available Este artículo es un análisis de las contribuciones del profesor Esteve al estudio de la profesión docente. Para ello se hace un repaso a sus rasgos de personalidad, a los hitos más relevantes de su vida profesional y a algunas de sus publicaciones desde sus comienzos como profesor universitario en 1973 hasta su fallecimiento en el año 2010.This article analyzes the contributions of Professor Esteve around the teaching profession. His personality traits, the highlights of his career and some of his publications from his beginnings as a university professor in 1973 until his death in 2010 are the core of this study.

  2. Contribuciones al preprocesado, procesado y análisis en termografía infrarroja aplicados a ensayos no destructivos

    OpenAIRE

    Hidalgo-Gato García, Rafael

    2015-01-01

    RESUMEN: El empleo de las técnicas de medida mediante ensayos no destructivos por Termografía Infrarroja (TI) permite la evaluación y control de todo tipo de materiales y procesos de forma rápida, sencilla y sin contacto. Por ello, se estableció como objetivo general de este trabajo de tesis el desarrollo de contribuciones al conocimiento de las diferentes etapas del tratamiento digital de los termogramas, así como su aplicación a problemáticas reales y solución a las carencias detectadas. Es...

  3. Contribuciones de Sir Roland Fisher a la Estadística Genética

    Directory of Open Access Journals (Sweden)

    Jaime Cuadros

    2004-11-01

    Full Text Available Sir Ronald Fisher (18901962 fue profesor de genética y muchas de sus innovaciones estadísticas encontraron expresión en el desarrollo de metodología en estadística genética. Sin embargo, mientras sus contribuciones en estadística matemática son fácilmente identificadas, en genética de poblaciones compartió su supremacía con Sewall Wright (1889 1988 y J. S. S. Haldane (1892 1965. Este documento muestra algunas de las mejores contribuciones de Fisher a las bases de la estadística genética, y sus interacciones con Wright y Haldane, los cuales contribuyeron al desarrollo del tema. Con la tecnología moderna, tanto la metodología la estadística como la información genética están cambiando. No obstante, muchos de los trabajos de Fisher permanecen relevantes, y pueden aun servir como una base para investigaciones futuras en el análisis estadístico de datos de DNA. El trabajo de este autor refleja su visión del papel de Ia estadística en Ia inferencia científica expresada en 1949

  4. 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.

  5. 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

  6. Thermal behavior analysis of U-Mo/Al dispersion fuel

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-04-15

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

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

    OpenAIRE

    HO JIN RYU; YEON SOO KIM

    2014-01-01

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

  9. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  12. 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.

  13. Modeling a failure criterion for U–Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jae-Yong, E-mail: tylor@kaeri.re.kr [Korea Atomic Energy Research Institute, 111, Daedeok-Daero 989 Beon-Gil, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Tahk, Young-Wook; Kim, Hyun-Jung; Kong, Eui-Hyun; Yim, Jeong-Sik [Korea Atomic Energy Research Institute, 111, Daedeok-Daero 989 Beon-Gil, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of)

    2016-05-15

    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.

  14. 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.

  15. Mechanical analysis of UMo/Al dispersion fuel

    International Nuclear Information System (INIS)

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

    2015-01-01

    Deformation of fuel particles and mass transfer from the transverse end of fuel meat toward the meat center was observed. This caused plate thickness peaking at a location between the meat edge and the meat center. The underlying mechanism for this fuel volume transport is believed to be fission induced creep of the U–Mo/Al meat. Fuel meat swelling was measured using optical microscopy images of the cross sections of the irradiated test plates. The time-dependent meat swelling was modeled for use in numerical simulation. A distinctive discrepancy between the predicted and measured meat thickness was found at the meat ends, which was assumed to be due to creep-induced mass relocation from the meat end to the meat center region that was not considered in the meat swelling model. ABAQUS FEA simulation was performed to reproduce the observed phenomenon at the meat ends. Through the simulation, we obtained the effective creep rate constants for the interaction layers (IL) and aluminum matrix. In addition, we obtained the corresponding stress and strain analysis results that can be used to understand mechanical behavior of U–Mo/Al dispersion fuel.

  16. Mechanical analysis of UMo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon [Ulsan National Institute of Science and Technology, Department of Nuclear Engineering, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan, 689-798 (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Sohn, Dong-Seong, E-mail: dssohn@unist.ac.kr [Ulsan National Institute of Science and Technology, Department of Nuclear Engineering, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan, 689-798 (Korea, Republic of)

    2015-11-15

    Deformation of fuel particles and mass transfer from the transverse end of fuel meat toward the meat center was observed. This caused plate thickness peaking at a location between the meat edge and the meat center. The underlying mechanism for this fuel volume transport is believed to be fission induced creep of the U–Mo/Al meat. Fuel meat swelling was measured using optical microscopy images of the cross sections of the irradiated test plates. The time-dependent meat swelling was modeled for use in numerical simulation. A distinctive discrepancy between the predicted and measured meat thickness was found at the meat ends, which was assumed to be due to creep-induced mass relocation from the meat end to the meat center region that was not considered in the meat swelling model. ABAQUS FEA simulation was performed to reproduce the observed phenomenon at the meat ends. Through the simulation, we obtained the effective creep rate constants for the interaction layers (IL) and aluminum matrix. In addition, we obtained the corresponding stress and strain analysis results that can be used to understand mechanical behavior of U–Mo/Al dispersion fuel.

  17. Tensile mechanical properties of U3Si2-Al fuel plate

    International Nuclear Information System (INIS)

    Xu Yong; Hu Huawei; Zhuang Hongquan; Wang Xishu

    2003-01-01

    The fuel plate made of fuel meat, with the U 3 Si 2 -Al dispersion fuel center, and 6061 Al alloy cladding, is a new kind of fuel used in research reactors. The mechanical property data of the fuel meat is the basic data in the design of fuel group, but the mechanical property of this fuel meat has not been studied all over the world till now. In this paper, the mechanical properties of U 3 Si 2 -Al fuel meats of different sizes used in research reactors are investigated and analyzed, and at the same time the carrying capacity of tensile in different directions are also compared. In order to get more knowledge about the mechanical properties of the fuel meat, the tensile experiment has been carried out repeatedly. Considering the lower ratio of elongation and the brittleness, the microscope has been used to examine the zone of fracture after tensile test. (authors)

  18. Microstructure of the irradiated U 3Si 2/Al silicide dispersion fuel

    Science.gov (United States)

    Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J.-F.; Robinson, A. B.; Madden, J. W.; Medvedev, P. G.; Wachs, D. M.

    2011-12-01

    The silicide dispersion fuel of U 3Si 2/Al is recognized as the best performance fuel for many nuclear research and test reactors with up to 4.8 gU/cm 3 fuel loading. An irradiated U 3Si 2/Al dispersion fuel ( 235U ˜ 75%) from the high-flux side of a fuel plate (U0R040) from the Reduced Enrichment for Research and Test Reactors (RERTR)-8 test was characterized using transmission electron microscopy (TEM). The fuel was irradiated in the Advanced Test Reactor (ATR) for 105 days. The average irradiation temperature and fission density of the U 3Si 2 fuel particles for the TEM sample are estimated to be approximately 110 °C and 5.4 × 10 27 f/m 3. The characterization was performed using a 200-kV TEM. The U/Si ratio for the fuel particle and (Si + Al)/U for the fuel-matrix-interaction layer are approximately 1.1 and 4-10, respectively. The estimated average diameter, number density and volume fraction for small bubbles (<1 μm) in the fuel particle are ˜94 nm, 1.05 × 10 20 m -3 and ˜11%, respectively. The results and their implication on the performance of the U 3Si 2/Al silicide dispersion fuel are discussed.

  19. Fuel Performance Calculations for FeCrAl Cladding in BWRs

    Energy Technology Data Exchange (ETDEWEB)

    George, Nathan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Sweet, Ryan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Maldonado, G. Ivan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Wirth, Brian D. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behavior of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.

  20. 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.

  1. 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)

  2. Influence of Fuel Meat Porosity on Heat Capacities of Fuel Element Plate U3Si2-Al

    International Nuclear Information System (INIS)

    Ginting, Aslina Br.; Supardjo; Sutri Indaryati

    2007-01-01

    Analyze of heat capacities of Al powder, AIMg 2 cladding, U 3 Si 2 powder and PEB U 3 Si 2 -Al with the meat porosity of 4.9; 5.53 ; 6.25 ; 6.95 %; 7.90; 8.66% have been done. Analysis was conducted by using Differential Scanning Calorimeter (DSC) at temperature 30℃ to 450℃ with heating rate 1℃ /minute in Argon gas media. The purpose of analyze is to know the influence of increasing of fuel meat porosity on heat capacities because increasing of percentage of meat porosity will cause degradation the of heat capacities of PEB U 3 Si 2 -Al. Result of analysis showed that the heat capacities of Al powder, AIMg 2 cladding increase by temperature, while heat capacities of U 3 Si 2 powder was stable with increasing of temperature up to 450℃. Analysis of heat capacities toward PEB U 3 Si 2 -Al indicate that increasing of fuel meat porosity of caused degradation of the heat capacities of PEB U 3 Si 2 -Al. Data obtained were expected to serve the purpose of input to fabricator of research reactor fuel in for design of fuel element type silicide with high loading. (author)

  3. Fission induced swelling of U–Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Jeong, G.Y. [Ulsan National Institute of Science and Technology, 50 UNIST-gil, Eonyang-eup, Uljoo-gun, Ulsan 689-798 (Korea, Republic of); Park, J.M. [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2015-10-15

    Fission-induced swelling of U–Mo/Al dispersion fuel meat was measured using microscopy images obtained from post-irradiation examination. The data of reduced-size plate-type test samples and rod-type test samples were employed for this work. A model to predict the meat swelling of U–Mo/Al dispersion fuel was developed. This model is composed of several submodels including a model for interaction layer (IL) growth between U–Mo and Al matrix, a model for IL thickness to IL volume conversion, a correlation for the fission-induced swelling of U–Mo alloy particles, a correlation for the fission-induced swelling of IL, and models of U–Mo and Al consumption by IL growth. The model was validated using full-size plate data that were not included in the model development.

  4. Properties of U3Si2-Al dispersion fuel element and its application

    International Nuclear Information System (INIS)

    Yin Changgeng

    2001-01-01

    The properties of U 3 Si 2 fuel and U 3 Si 2 -Al dispersion fuel element are introduced, which include U-loading; the banding quality, U-homogeneity and 'dog-bone' phenomenon, the minimum thickness of cladding and the corrosion performances. The fabrication technique of fuel elements, NDT for fuel plates, assemble technique of fuel elements and the application of U 3 Si 2 -Al dispersion fuel elements in the world are introduced

  5. Solution combustion synthesis of strontium aluminate, SrAl2O4, powders: single-fuel versus fuel-mixture approach.

    Science.gov (United States)

    Ianoş, Robert; Istratie, Roxana; Păcurariu, Cornelia; Lazău, Radu

    2016-01-14

    The solution combustion synthesis of strontium aluminate, SrAl2O4, via the classic single-fuel approach and the modern fuel-mixture approach was investigated in relation to the synthesis conditions, powder properties and thermodynamic aspects. The single-fuel approach (urea or glycine) did not yield SrAl2O4 directly from the combustion reaction. The absence of SrAl2O4 was explained by the low amount of energy released during the combustion process, in spite of the highly negative values of the standard enthalpy of reaction and standard Gibbs free energy. In the case of single-fuel recipes, the maximum combustion temperatures measured by thermal imaging (482 °C - urea, 941 °C - glycine) were much lower than the calculated adiabatic temperatures (1864 °C - urea, 2147 °C - glycine). The fuel-mixture approach (urea and glycine) clearly represented a better option, since (α,β)-SrAl2O4 resulted directly from the combustion reaction. The maximum combustion temperature measured in the case of a urea and glycine fuel mixture was the highest one (1559 °C), which was relatively close to the calculated adiabatic temperature (1930 °C). The addition of a small amount of flux, such as H3BO3, enabled the formation of pure α-SrAl2O4 directly from the combustion reaction.

  6. BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

    Energy Technology Data Exchange (ETDEWEB)

    Sweet, Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-30

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and

  7. Corrosion and protection of spent Al-clad research reactor fuel during extended wet storage

    International Nuclear Information System (INIS)

    Ramanathan, Lalgudi V.

    2009-01-01

    A variety of spent research reactor fuel elements with different fuel meats, geometries and 235 U enrichments are presently stored under water in basins throughout the world. More than 90% of these fuels are clad in aluminum (Al) or its alloy and are susceptible to corrosion. This paper presents an overview of the influence of Al alloy composition, galvanic effects (Al alloy/stainless steel), crevice effects, water parameters and synergism between these parameters as well as settled solids on the corrosion of typical Al alloys used as fuel element cladding. Pitting is the main form of corrosion and is affected by water conductivity, chloride ion content, formation of galvanic couples with rack supports and settled solid particles. The extent to which these parameters influence Al corrosion varies. This paper also presents potential conversion coatings to protect the spent fuel cladding. (author)

  8. 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

  9. 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)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  11. 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)

  12. The Fabrication Problem Of U3Si2-Al Fuel With Uranium High Loading

    International Nuclear Information System (INIS)

    Supardjo

    1996-01-01

    The quality of U 3 Si 2 -Al dispersion fuel product is the main aim for each fabricator. Low loading of uranium fuel element is easily fabricated, but with the increased, uranium loading, homogeneity of uranium distribution is difficult to achieve and it always formed white spots, blister, and dogboning in the fuel plates. The problem can be eliminated by the increasing treatment of the fuel/Al powder. The precise selection of fuel/Al particles diameter is needed indeed to make easier in the homogeneous process of powder and the porosities arrangement in the fuel plates. The increasing of uranium loading at constant meat thickness will increase the meat hardness, therefore to withdraw the dogboning forming, the use of harder cladding materials is necessity

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  14. Status of U3Si2-Al fuel development in China

    International Nuclear Information System (INIS)

    Sun Rongxian; Huang Dechen; Yin Changgeng; Zhang Qindi; He Fengqi; Li Shuhua; Xiang Xingbi; Zhang Zhiyi; Wang Zhaoyue

    1995-01-01

    On the basis of the research and development work, a production line of U 3 Si 2 -Al dispersion fuel elements has been established and tens of the fuel elements have been produced. A lot of technical improvements and reliable inspection system ensure that all technical performance of the fuel elements meet International standards. (author)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  16. Contribuciones al estudiode los anfibios y reptiles de Méxicodurante el siglo XVIII y la Ilustración

    Directory of Open Access Journals (Sweden)

    Gustavo Casas Andreu

    2008-01-01

    Full Text Available Las bases de la herpetología moderna en general se establecieron en el siglo XVIII,particularmente durante la Ilustración y quienes hicieron la mayor contribución fueronfundamentalmente los naturalistas franceses. No obstante, en México se hicieron varias publicacionesde cierto relieve, en especial por los misioneros jesuitas y otros estudiosos de la Nueva España. Esinteresante mencionar que aun con la trascendencia de Linneo para la biología moderna, algunoshistoriadores de los anfibios y los reptiles señalan que por lo menos para la herpetofauna de Méxicoexistió una importante regresión, ya que era mucho mayor el conocimiento que había dejadoFrancisco Hernández en el siglo XVI. Las contribuciones de los autores de la escuela francesa comoBuffon y quienes lo sucedieron como Lacepéde y Daudin, fueron los grandes pilares de laherpetología o estudios de los anfibios y los reptiles de la manera en que la conocemos en laactualidad. Las bases establecidas en el siglo XVIII sirvieron para que con la apertura de México almundo a partir de su independencia, se entrara en una de las etapas de mayor relevancia para laherpetología del país.

  17. A review of microstructural analysis on U3Si2-Al plate-type fuel

    International Nuclear Information System (INIS)

    Ti Zhongxin; Guo Yibai

    1995-12-01

    The microstructure of U 3 Si 2 -Al plate-type fuel, that is the microstructure of fuel particles, compatibility of the fuel particles and Al matrix, fuel particles distribution, dogbone area morphology, clad and meat thickness, bone quality of clad/frame and clad/fuel core, and the effect of these factors on products quality were comprehensively investigated and analyzed by means of optical microscopy (OM), scanning electron microscopy (SEM), X-ray diffraction (XRD), energy dispersive X-ray spectrometry (EDX), image processing technique, etc.. The main results are as following: U-7.7%Si alloy contains two phases: primary U 3 Si 2 and small amount of USi (about 12%), free-uranium was not detected in fuel particles; the dogbone area is the key factor affecting fuel plate quality (1 ref., 16 figs., 4 tabs.)

  18. 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.

  19. Post-irradiation examination of prototype Al-64 wt% U3Si2 fuel rods from NRU

    International Nuclear Information System (INIS)

    Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D.

    1997-01-01

    Three prototype fuel rods containing Al-64 wt% U 3 Si 2 (3.15 gU/cm 3 ) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U 3 Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U 3 Si 2 powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37 degrees C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U 3 Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL's research reactors

  20. 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

  1. Thermal conductivity of U–Mo/Al dispersion fuel. Effects of particle shape and size, stereography, and heat generation

    International Nuclear Information System (INIS)

    Cho, Tae Won; Sohn, Dong-Seong; Kim, Yeon Soo

    2015-01-01

    This paper describes the effects of particle sphericity, interfacial thermal resistance, stereography, and heat generation on the thermal conductivity of U–Mo/Al dispersion fuel. The ABAQUS finite element method (FEM) tool was used to calculate the effective thermal conductivity of U–Mo/Al dispersion fuel by implementing fuel particles. For U–Mo/Al, the particle sphericity effect was insignificant. However, if the effect of the interfacial thermal resistance between the fuel particles and Al matrix was considered, the thermal conductivity of U–Mo/Al was increased as the particle size increases. To examine the effect of stereography, we compared the two-dimensional modeling and three-dimensional modeling. The results showed that the two-dimensional modeling predicted lower than the three-dimensional modeling. We also examined the effect of the presence of heat sources in the fuel particles and found a decrease in thermal conductivity of U–Mo/Al from that of the typical homogeneous heat generation modeling. (author)

  2. Irradiation behavior of U 6Mn-Al dispersion fuel elements

    Science.gov (United States)

    Meyer, M. K.; Wiencek, T. C.; Hayes, S. L.; Hofman, G. L.

    2000-02-01

    Irradiation testing of U 6Mn-Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U 6Mn in an unrestrained plate configuration performs similarly to U 6Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3×10 27 m-3. Fuel plate failure occurs by fission gas pressure driven `pillowing' on continued irradiation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

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

    Science.gov (United States)

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

    2014-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-15

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

  6. Study on the irradiation swelling of U3Si2-Al dispersion fuel

    International Nuclear Information System (INIS)

    Xing Zhonghu; Ying Shihao

    2001-01-01

    The dominant modeling mechanisms on irradiation swelling of U 3 Si 2 -Al dispersion fuel are introduced. The core of dispersion fuel is looked to as micro-fuel elements of continuous matrix. The formation processes of gas bubbles in the fuel phase are described through the behavior mechanisms of fission gases. The swelling in the fuel phase causes the interaction between fuel particles and metal matrix, and the metal matrix can restrain the irradiation swelling of fuel particles. The developed code can predict irradiation-swelling values according to the parameters of fuel elements and irradiation conditions, and the predicted values are in agreement with the measured results

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  9. FeCrAl/Zr dual layer fuel cladding for improved safety margin under accident scenario

    International Nuclear Information System (INIS)

    Park, D.J.; Park, J.H.; Jung, Y.I.; Kim, H.G.; Park, J.Y.; Koo, Y.H.

    2014-01-01

    For application of advanced steel as a cladding material in light water reactor (LWR), FeCrAl/Zr dual layer tube was manufactured by using a hot isostatic pressing (HIP) method. To optimize HIP condition for joining both FeCrAl and Zr alloys, HIP was carried out under various temperature conditions. Tensile test and 3-point bend test performed for measuring mechanical properties of HIPed sample. To better understand microstructural characteristics in interface region between two alloys, SEM and TEM study were conducted by using HIPed sample with different process conditions. Based on this optimization study and analyzed results, optimized HIP condition was determined and FeCrAl/Zr dual layer fuel cladding having same wall thickness with current LWR fuel cladding was manufactured. Simulated loss-of-coolant accident test was carried out using FeCrAl/Zr dual layer cladding sample and fuel integrity was measured by mechanical test. (authors)

  10. 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.

  11. The improvement of technology for high-uranium-density Al-base dispersion fuel plates

    International Nuclear Information System (INIS)

    Shouhui, Dai; Rongxian, Sun; Hejian, Mao; Baosheng, Zhao; Changgen, Yin

    1987-01-01

    An improved rolling process was developed for manufacturing Al-base dispersion fuel plates. When the fuel content in the meat increased up to 50 vol%, the non-uniformity of uranium is not more than ± 7.2%, and the minimum cladding thickness is not less than 0.32 mm. (Author)

  12. Irradiation behavior of U{sub 6}Mn-Al dispersion fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, M.K. E-mail: mitchell.meyer@anl.gov; Wiencek, T.C.; Hayes, S.L.; Hofman, G.L

    2000-04-01

    Irradiation testing of U{sub 6}Mn-Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U{sub 6}Mn in an unrestrained plate configuration performs similarly to U{sub 6}Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3x10{sup 27} m{sup -3}. Fuel plate failure occurs by fission gas pressure driven 'pillowing' on continued irradiation.

  13. 2nd Gen FeCrAl ODS Alloy Development For Accident-Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Massey, Caleb P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Edmondson, Philip D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    Extensive research at ORNL aims at developing advanced low-Cr high strength FeCrAl alloys for accident tolerant fuel cladding. One task focuses on the fabrication of new low Cr oxide dispersion strengthened (ODS) FeCrAl alloys. The first Fe-12Cr-5Al+Y2O3 (+ ZrO2 or TiO2) ODS alloys exhibited excellent tensile strength up to 800 C and good oxidation resistance in steam up to 1400 C, but very limited plastic deformation at temperature ranging from room to 800 C. To improve alloy ductility, several fabrication parameters were considered. New Fe-10-12Cr-6Al gas-atomized powders containing 0.15 to 0.5wt% Zr were procured and ball milled for 10h, 20h or 40h with Y2O3. The resulting powder was then extruded at temperature ranging from 900 to 1050 C. Decreasing the ball milling time or increasing the extrusion temperature changed the alloy grain size leading to lower strength but enhanced ductility. Small variations of the Cr, Zr, O and N content did not seem to significantly impact the alloy tensile properties, and, overall, the 2nd gen ODS FeCrAl alloys showed significantly better ductility than the 1st gen alloys. Tube fabrication needed for fuel cladding will require cold or warm working associated with softening heat treatments, work was therefore initiated to assess the effect of these fabrications steps on the alloy microstructure and properties. This report has been submitted as fulfillment of milestone M3FT 16OR020202091 titled, Report on 2nd Gen FeCrAl ODS Alloy Development for the Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program.

  14. Examen de auditoría integral a las contribuciones de mejoras por el período comprendido entre el 1 de Enero y el 31 de diciembre del 2012, en el Gobierno Municipal del Cantón Morona

    OpenAIRE

    Valencia Zabala, Liliam Margot

    2014-01-01

    En el presente trabajo de investigación trata sobre una auditoría integral al Gobierno Municipal del Cantón Morona, se han diseñado capítulos que se relacionan con el desarrollo de una auditoría integral a las contribuciones de mejoras, que ha sido desarrollado con el aporte de los funcionarios auditados y se ha evaluado con los conocimientos adquiridos en esta maestría. Se han utilizado conceptos claves, que son parte fundamental de la auditoría integral como son: auditoría de gesti...

  15. 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

  16. Post-irradiation examination of prototype Al-64 wt% U{sub 3}Si{sub 2} fuel rods from NRU

    Energy Technology Data Exchange (ETDEWEB)

    Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D. [Chalk River Labs., Ontario (Canada)

    1997-08-01

    Three prototype fuel rods containing Al-64 wt% U{sub 3}Si{sub 2} (3.15 gU/cm{sup 3}) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U{sub 3}Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U{sub 3}Si{sub 2} powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37{degrees}C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U{sub 3}Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL`s research reactors.

  17. Contribuciones bibliográficas de la Facultad de Farmacia y Bioquímica

    OpenAIRE

    Bonilla Rivera, Pablo E.

    2014-01-01

    La contribución bibliográfica de la facultad de Farmacia y Bioquímica de San Marcos es muy importante, pero tal vez poco conocida en los claustros sanmarquinos. Por ello, a partir de este número, "Ciencia e Investigación" incluirá las listas de las contribuciones hechas por los docentes de la facultad y por los exalumnos a través de sus tesis de título y de grados académicos. La contribución bibliográfica de la facultad de Farmacia y Bioquímica de San Marcos es muy importante, pero tal vez...

  18. 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.

  19. 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.

  20. Repository emplacement costs for Al-clad high enriched uranium spent fuel

    International Nuclear Information System (INIS)

    McDonell, W.R.; Parks, P.B.

    1994-01-01

    A range of strategies for treatment and packaging of Al-clad high-enriched uranium (HEU) spent fuels to prevent or delay the onset of criticality in a geologic repository was evaluated in terms of the number of canisters produced and associated repository costs incurred. The results indicated that strategies in which neutron poisons were added to consolidated forms of the U-Al alloy fuel generally produced the lowest number of canisters and associated repository costs. Chemical processing whereby the HEU was removed from the waste form was also a low cost option. The repository costs generally increased for isotopic dilution strategies, because of the substantial depleted uranium added. Chemical dissolution strategies without HEU removal were also penalized because of the inert constituents in the final waste glass form. Avoiding repository criticality by limiting the fissile mass content of each canister incurred the highest repository costs

  1. Irradiation behavior of low-enriched U/sub 6/Fe-Al dispersion fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Hofman, G.L.; Domagala, R.F.; Copeland, G.L.

    1987-10-01

    An irradiation test of miniature fuel plates containing low-enriched (20% /sup 235/U)U/sub 6/Fe dispersed and clad in Al was performed. The postirradiation examination shows U/sub 6/Fe to form extensive fission gas bubbles at burnups of only approx. = 20% of the original 20% fuel enrichment. Plate failure by fission gas-driven pillowing occurred at approx. = 40% burnup. This places U/sub 6/FE at the lowest burnup capability among low enriched dispersion fuels that have been tested for use in research and test reactors

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-06-15

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

  3. Postirradiation examination of a low enriched U3Si2-Al fuel element manufactured and irradiated at Batan, Indonesia

    International Nuclear Information System (INIS)

    Suripto, A.; Sugondo, S.; Nasution, H.

    1994-01-01

    The first low-enriched U 3 Si 2 -Al dispersion plate-type fuel element produced at the Nuclear Fuel Element Center, BATAN, Indonesia, was irradiated to a peak 235 U burnup of 62%. Postirradiation examinations performed to data shows the irradiation behavior of this element to be similar to that of U 3 Si 2 -Al plate-type fuel produced and tested at other institutions. The main effect of irradiation on the fuel plates is a thickness increase of 30--40 μm (2.5-3.0%). This thickness increase is almost entirely due to the formation of a corrosion layer (Boehmite). The contribution of fuel swelling to the thickness increase is rather small (less than 10 μm) commensurate with the burnup of the fuel and the relatively moderate as-fabricated fuel volume fraction of 27% in the fuel meat

  4. 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.

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

    International Nuclear Information System (INIS)

    Supardjo; Agoeng Kadarjono; Boybul; Aslina Br Ginting

    2016-01-01

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

  6. Irradiation tests of U3Si2-Al fuels up to very high fission densities

    International Nuclear Information System (INIS)

    Nuding, M.; Boening, K.

    2001-01-01

    The new research reactor of the Munich Technical University (TUM), the FRM-II, will have U 3 Si 2 -Al as the fuel. This fuel is considered qualified and optimally usable in the light of findings obtained in the RERTR program (Reduced Enrichment for Research and Test Reactors). The RERTR program was conducted to develop new fuel for the use of low enriched uranium (LEU) in research reactors. As the unique properties of the FRM-II in research and application are based also on achieving a very compact reactor core with highly enriched uranium (HEU), additional irradiation tests were performed on the basis of the RERTR program. They were run in close cooperation with the French Commissariat a l'Energie Atomique (CEA) in its SILOE and OSIRIS facilities, among others. After extensive evaluation, also of other studies, these tests confirm the RERTR findings about fuel swelling behavior and, consequently, the suitability of U 3 Si 2 -Al (HEU) for use in the compact core of the FRM-II. (orig.) [de

  7. Recovery of UMo alloy from UMo/Al dispersion fuel plates by dissolution

    International Nuclear Information System (INIS)

    Ren Meng; Li Jia; Liu Jinhong; Zhu Changgui

    2011-01-01

    Methods for dissolving UMo/Al dispersion fuel plates in the compounded mixed basic aqueous (NaOH and NaNO 3 ) are studied on laboratory scale. After removing the clad and the matrix of the substandard UMo/Al dispersion fuel elements, the U loss ratios are calculated and the granularity distributions of the recovered UMo alloy powder are analyzed by the metallurgical microscope. Besides, the phase structure and the composition of the recovered UMo alloy powder are analyzed by the XRD. The results indicate that as the concentration of NaOH increases, uranium loss ratio increases; but as the concentration of NaNO 3 increases, U loss ration increases firstly and then decreases subsequently; generally, the U recovery ratios are more than 99.3%. The granularity of recovered UMo powders are very small and most parts of γ-U have been oxidated to UO 2 . Therefore, further study is required to determined whether the recovered UMo alloy could be returned to the product line. (authors)

  8. 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

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

    International Nuclear Information System (INIS)

    Zweifel, Tobias

    2014-01-01

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

  10. 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)

  11. 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

  12. Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale FeCrAl Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andersson, D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Capolungo, L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wirth, B. D. [Univ. of Tennessee, Knoxville, TN (United States)

    2017-07-26

    Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced ac- cident tolerance when compared to traditional UO2 fuel zircaloy clad fuel rods. One of the potential replacement claddings are iron-chromium-alunimum (FeCrAl) alloys due to their increased oxidation resistance [1–4] and higher strength [1, 2]. While the oxidation characteristics of FeCrAl are a benefit for accident tolerance, the thermal neu- tron absorption cross section of FeCrAl is about ten times that of Zircaloy. This neutronic penalty necessitates thinner cladding. This allows for slightly larger pellets to give the same cold gap width in the rod. However, the slight increase in pellet diameter is not sufficient to compensate for the neutronic penalty and enriching the fuel beyond the current 5% limit appears to be necessary [5]. Current estimates indicate that this neutronic penalty will impose an increase in fuel cost of 15-35% [1, 2]. In addition to the neutronic disadvantage, it is anticipated that tritium release to the coolant will be larger because the permeability of hydrogen in FeCrAl is about 100 times higher than in Zircaloy [6]. Also, radiation-induced hardening and embrittlement of FeCrAl need to be fully characterized experimentally [7]. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022 [8] multiscale multiphysics modeling approaches have been used to provide insight into these the use of FeCrAl as a cladding material. The purpose of this letter report is to highlight the multiscale modeling effort for iron-chromium-alunimum (FeCrAl) cladding alloys as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program through its Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The approach taken throughout the HIP is to

  13. Contribuciones al conocimiento de las Magnoliácea de Colombia, V

    Directory of Open Access Journals (Sweden)

    Lozano Contreras Gustavo

    1978-12-01

    Full Text Available La consecución de material adicional de Magnoliaceae del sur de Colombia permite establecer dos nuevos taxa, ubicados en los géneros Dugandiodendron y Talauma, respectivamente.Al igual que Talauma dixonii Little y Dugandiodendron striatifolium (Little G. Lozano-c., conocidas de Esmeraldas, Ecuador, las nuevas especies reciben el nombre vernáculo de "Cucharillo". Como ocurre con otras magnoliáceas, la madera de estas especies es muy apreciada, por lo cual se les explota a pesar de la prohibición vigente en cuanto a tala; más aún, las especies que se describen están en peligro de extincion yes difícil obtener material completo de las mismas.  Los "cucharillos" han sido elementos importantes en la composición de la selva pluvial macrotérmica de la región tropical occidental ubicada en el Departamento de Nariño.

  14. Corrosion surveillance programme for Latin American research reactor Al-clad spent fuel in water

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Haddad, R.; Ritchie, I.

    2002-01-01

    The objectives of the IAEA sponsored Regional Technical Co-operation Project for Latin America (Argentina, Brazil, Chile, Mexico, and Peru) are to provide the basic conditions to define a regional strategy for managing spent fuel and to provide solutions, taking into consideration the economic and technological realities of the countries involved. In particular, to determine the basic conditions for managing research reactor spent fuel during operation and interim storage as well as final disposal, and to establish forms of regional cooperation in the four main areas: spent fuel characterization, safety, regulation and public communication. This paper reports the corrosion surveillance activities of the Regional Project and these are based on the IAEA sponsored co-ordinated research project (CRP) on 'Corrosion of research reactor Al-clad spent fuel in water'. The overall test consists of exposing corrosion coupon racks at different spent fuel basins followed by evaluation. (author)

  15. 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

  16. Luis Agustín García Moreno, Esther Sánchez Medina (eds.), Del Nilo al Guadalquivir. II Estudios sobre las fuentes de la conqui

    OpenAIRE

    Meouak, Mohamed

    2015-01-01

    El libro que el lector tiene entre sus manos constituye una seria apuesta por sacar definitivamente de la obscuridad los estudios dedicados al Mediterráneo centro-occidental y el norte de África entre el final de la Antigüedad y la alta Edad Media dentro del ámbito académico español. Entre mundos distintos y culturas en contacto, dicho volumen ofrece contribuciones científicas de calidad, basadas en investigaciones relativas al norte de África que permiten aumentar y matizar considerablemente...

  17. 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.

  18. 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.

  19. 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)

  20. Fuel effect on solution combustion synthesis of Co(Cr,Al)2O4 pigments

    International Nuclear Information System (INIS)

    Gilabert, J.; Palacios, M.D.; Sanz, V.; Mestre, S.

    2017-01-01

    The fuel effect on the synthesis of a ceramic pigment with a composition CoCr2−2ΨAl2ΨO4 (0≤Ψ≤1) by means of solution combustion synthesis process (SCS) has been studied. Three different fuels were selected to carry out the synthesis (urea, glycine and hexamethylentetramine (HMT)). Highly foamy pigments with very low density were obtained. Fd-3m spinel-type structure was obtained in all the experiments. Nevertheless, crystallinity and crystallite size of the spinels show significant differences with composition and fuel. The use of glycine along with the chromium-richest composition favours ion rearrangement to obtain the most ordered structure. Lattice parameter does not seem to be affected by fuel, although it evolves with Ψ according to Vegard's law. Colouring power in a transparent glaze shows important variations with composition. On the other hand, fuel effect presents a rather low influence since practically the same shades are obtained. However, it exerts certain effect on luminosity (L*). [es

  1. 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.

  2. 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

  3. Emerging methanol-tolerant AlN nanowire oxygen reduction electrocatalyst for alkaline direct methanol fuel cell.

    Science.gov (United States)

    Lei, M; Wang, J; Li, J R; Wang, Y G; Tang, H L; Wang, W J

    2014-08-11

    Replacing precious and nondurable Pt catalysts with cheap materials is a key issue for commercialization of fuel cells. In the case of oxygen reduction reaction (ORR) catalysts for direct methanol fuel cell (DMFC), the methanol tolerance is also an important concern. Here, we develop AlN nanowires with diameters of about 100-150 nm and the length up to 1 mm through crystal growth method. We find it is electrochemically stable in methanol-contained alkaline electrolyte. This novel material exhibits pronounced electrocatalytic activity with exchange current density of about 6.52 × 10(-8) A/cm(2). The single cell assembled with AlN nanowire cathodic electrode achieves a power density of 18.9 mW cm(-2). After being maintained at 100 mA cm(-2) for 48 h, the AlN nanowire-based single cell keeps 92.1% of the initial performance, which is in comparison with 54.5% for that assembled with Pt/C cathode. This discovery reveals a new type of metal nitride ORR catalyst that can be cheaply produced from crystal growth method.

  4. 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.

  5. Algunas contribuciones al debate sobre la clínica de las psicosis

    Directory of Open Access Journals (Sweden)

    Juan Manuel Rodríguez

    Full Text Available La clínica de las psicosis plantea la necesidad de establecer una modalidad de tratamiento diferente al tratamiento de las neurosis. Esta practica implica un trabajo multidisciplinario cercano y frecuente en donde se establecen algunas coordenadas fundamentales en el devenir del tratamiento. Asimismo, en el interior del grupo de trabajo se presentan algunos fenómenos similares a la dinámica propia del paciente psicotico, por ello se hace necesario considerar la locura del grupo como un factor dinámico del tratamiento. En este sentido es fundamental partir de la singularidad del paciente para desarrollar el modelo único en cada caso, en el cual la intervención es un elemento que pretende sostener aquello que falla en la dimensión delirante del paciente.

  6. Consideraciones sobre afectividad en las relaciones de enseñanza: las contribuciones de Vigotski

    Directory of Open Access Journals (Sweden)

    Raimundo Nonato de Oliveira Falabelo

    2015-10-01

    Full Text Available Este artigo tiene como objetivo presentar conclusiones de un estudio sobre la cuestión afectiva en una perspectiva inter-relacional con la cognición, como una misma unidad, que es la vida psíquica humana, lo que significa concebir que el afecto está presente y es constitutivo de toda y cualquier acción humana. Tiene como base teórica los estudios realizados por L. S. Vigotski. Echamos mano, aún, de las contribuciones de algunos de sus comentadores o interpretadores que tratan específicamente de este tema. El texto hace parte de los fundamentos teóricos de la Tesis de Doctorado sobre esta problemática, en contexto empírico, de clase, en un aula de educación de jóvenes y adultos.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-04-01

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

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

    Science.gov (United States)

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

    2018-04-01

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

  9. Cold spray deposition of Ti{sub 2}AlC coatings for improved nuclear fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Maier, Benjamin R. [University of Wisconsin, Madison, WI (United States); Garcia-Diaz, Brenda L. [Savannah River National Laboratory, Aiken, SC (United States); Hauch, Benjamin [University of Wisconsin, Madison, WI (United States); Olson, Luke C.; Sindelar, Robert L. [Savannah River National Laboratory, Aiken, SC (United States); Sridharan, Kumar, E-mail: kumar@engr.wisc.edu [University of Wisconsin, Madison, WI (United States)

    2015-11-15

    Coatings of Ti{sub 2}AlC MAX phase compound have been successfully deposited on Zircaloy-4 (Zry-4) test flats, with the goal of enhancing the accident tolerance of LWR fuel cladding. Low temperature powder spray process, also known as cold spray, has been used to deposit coatings ∼90 μm in thickness using powder particles of <20 μm. X-ray diffraction analysis showed the phase-content of the deposited coatings to be identical to the powders indicating that no phase transformation or oxidation had occurred during the coating deposition process. The coating exhibited a high hardness of about 800 H{sub K} and pin-on-disk wear tests using abrasive ruby ball counter-surface showed the wear resistance of the coating to be significantly superior to the Zry-4 substrate. Scratch tests revealed the coatings to be well-adhered to the Zry-4 substrate. Such mechanical integrity is required for claddings from the standpoint of fretting wear resistance and resisting wear handling and insertion. Air oxidation tests at 700 °C and simulated LOCA tests at 1005 °C in steam environment showed the coatings to be significantly more oxidation resistant compared to Zry-4 suggesting that such coatings can potentially provide accident tolerance to nuclear fuel cladding. - Highlights: • Deposited Ti{sub 2}AlC coatings on Zircaloy-4 substrates with a low pressure powder spray process, also known as cold spray. • Coatings have high hardness and wear resistance for both damage resistance during rod insertion and fretting wear resistance. • The oxidation resistance of Ti{sub 2}AlC coated Zircaloy-4 at 700 °C and 1005 °C was significantly superior to uncoated Zircaloy. • Cold spray of Ti{sub 2}AlC demonstrates considerable promise as a near-term solution for accident tolerant Zr-alloy fuel claddings.

  10. The UMo Powder Production Process of UMo-Al Dispersion Fuel for Research Reactor has been Studied

    International Nuclear Information System (INIS)

    Supardjo

    2007-01-01

    Development of UMo-Al dispersion fuel with low enrichment uranium ( 3 ), a relatively large range of γ phase and easily reprocessed. Using UMo alloy as nuclear fuel, uranium density can be increased until 9.0 g/cm 3 , is higher than that of U 3 Si 2 -Al fuel that has only maximum Uranium density 6.0 g/cm 3 . Because of ductility of UMo alloy, thus exact and economic powder production method is needed. Some powder production methods are mechanical crushing (milling, grinding, etc), cryogenic mechanical crushing, atomization, and Hydride-Dehydride. The mechanical crushing and cryogenic mechanical crushing methods are difficult to be performed, time consuming and have high impurity products. However, atomization and hydride-dehydride methods are performed easily, fast and have low impurity products. The product of atomization process is spherical and uniform shape, but, another processes have irregular shape. The evaluation result of some methods showed that hydride-dehydride and atomization methods are more suitable for producing UMo powder than that of another methods. (author)

  11. Deoxygenation of palm kernel oil to jet fuel-like hydrocarbons using Ni-MoS_2/γ-Al_2O_3 catalysts

    International Nuclear Information System (INIS)

    Itthibenchapong, Vorranutch; Srifa, Atthapon; Kaewmeesri, Rungnapa; Kidkhunthod, Pinit; Faungnawakij, Kajornsak

    2017-01-01

    Highlights: • The Ni-MoS_2/γ-Al_2O_3 catalysts synthesized using thiourea solution processing. • The Ni-MoS_2 showed semi-amorphous crystallinity with crystallite size of 5–10 nm. • The Ni K-edge XANES and EXAFS indicated the Ni substitution in MoS_2 structure. • A high yield of jet fuel-like hydrocarbon (>90%) from the palm kernel oil feedstock. • The HDO pathway was highly selective, while the DCO_2 and DCO pathways were minor. - Abstract: In the current study, palm kernel oil was used as a renewable feedstock for production of jet fuel-like hydrocarbons via the deoxygenation over the Ni-MoS_2/γ-Al_2O_3 catalyst. The dominant C12 fatty acid content in palm kernel oil makes it promising for jet fuel application. Synthesized by a liquid processing method with thiourea organosulfur agent, the catalyst revealed MoS_2 structure with low stacking, while Ni substitution in the MoS_2 structure and interaction with the Al_2O_3 support were determined based on the Ni K-edge XANES and EXAFS results. A high hydrodeoxygenation (HDO) activity, which as the major pathway in the deoxygenation, was observed upon application of a H_2 pressure of 30–50 bar over Ni-MoS_2/γ-Al_2O_3. The optimum product yield of approximately 92% was obtained mainly from the HDO pathway (∼60%) with 58% selectivity to C10–C12 jet fuel hydrocarbons. The flow property of the jet fuel-like hydrocarbons was more desirable than those obtained from palm olein oil-derived fuel.

  12. Metallographic analysis of irradiated U3Si2/Al fuel element plate of 2.96 gU/cm3 density

    International Nuclear Information System (INIS)

    Maman Kartaman Ajiriyanto; Aslina Br Ginting; Junaedi

    2018-01-01

    Metallographic analysis of U 3 Si 2 /Al fuel element plate has been performed in hot cell. The purpose of metallographic analysis is to study changes in PEB U 3 Si 2 /Al microstructure and AlMg 2 cladding thickness after irradiation in reactor until burn up of 56 %. The fuel element plate of irradiated U 3 Si 2 /Al was cut in top, middle and bottom positions with each size around 5 x 5 x 1.37 mm. Metallographic preparation starts from sample cutting using cutting machine with low speed and sample mounting, grinding and polishing in hot cell 104–105. Sample mounting was done by using resin for more than 10 hours followed by grinding with sand papers up to grit size of 2400 and polishing with diamond paste of size 3 to 1 micron at a rotational speed of 150 rpm for 5 minutes. Microstructure observation was performed with optical microscope in hot cell 107 at 200 times magnification. Microstructure examination reveals U 3 Si 2 particles with inverse forms and sizes, Al matrix and AlMg 2 cladding were spread along the U 3 Si 2 /Al side. Microstructure observation of irradiated U 3 Si 2 /Al has not shown good result because only topography observation of U 3 Si 2 /Al meat, Al matrix and AlMg 2 cladding can be done due to limited capability of the optical microscope in hot cell, where maximum magnification can be attained only at 200 times so that the phenomenon of interaction layer and small gas bubble can not be observed. However, U 3 Si 2 /Al microstructure of 56 % burnup, if compared to the microstructure of U 3 Si 2 /Al fuel element plate of 60 % burnup from previous researcher, shows interaction between U 3 Si 2 meat with Al matrix and the existence of layers with a thickness about 5 up to 20 microns. Meanwhile, the observed thickness of AlMg 2 cladding is greater than 0.25 mm, which indicates that irradiation does not significantly change the thickness of AlMg 2 cladding so that the overall irradiated U 3 Si 2 -Al still has good integrity and stability. (author)

  13. Study of the residual porosity in fuel plate cores based on U3O8 - Al dispersions

    International Nuclear Information System (INIS)

    Durazzo, M.

    2005-01-01

    The residual porosity in the meat of nuclear dispersion fuel plates, the fabrication voids, explains the corrosion behaviour of the meats when exposed to the water used as coolant and moderator of MTR type research reactors. The fabrication voids also explain variations in irradiation performance of many fuel dispersion for nuclear reactors. To obtain improved corrosion and irradiation performance, we must understand the fabrication factors that control the amount of void volume in fuel plate meats. The purpose of this study was to investigate the void content of aluminum-base dispersion-type U 3 O 8 -Al fuel plates depending on the characteristics of the starting fuel dispersion used to produce the fuel meat, which is fabricated by pressing. The void content depends on the U 3 O 8 concentration. For a particular U 3 O 8 content, the rolling process establishes a constant void concentration, which is called equilibrium porosity. The equilibrium quantity of voids is insensitive to the initial density of the fuel compact. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-06-01

    The purpose of this report is to collect in one publication and fit together work fragments presented in many conferences in the multi-year time span starting 2002 to the present dealing with the problem of large pore formation in U-Mo/Al dispersion fuel plates first observed in 2002. Hence, this report summarizes the excerpts from papers and reports on how we interpreted the relevant results from out-of-pile and in-pile tests and how this problem was dealt with. This report also provides a refined view to explain in detail and in a quantitative manner the underlying mechanism of the role of silicon in improving the irradiation performance of U-Mo/Al.

  15. DART model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

    International Nuclear Information System (INIS)

    Rest, J.; Hofman, G.L.

    1997-01-01

    The Dispersion Analysis Research Tool (DART) contains models for fission-gas-induced fuel swelling, interaction of fuel with the matrix aluminum, for the resultant reaction-product swelling, and for the calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U 3 SiAl-Al and U 3 Si 2 -Al for various dispersion fuel element designs with the data

  16. 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)

  17. Development of ODS FeCrAl alloys for accident-tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoelzer, David T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Unocic, Kinga A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-18

    FeCrAl alloys are prime candidates for accident-tolerant fuel cladding due to their excellent oxidation resistance up to 1400 C and good mechanical properties at intermediate temperature. Former commercial oxide dispersion strengthened (ODS) FeCrAl alloys such as PM2000 exhibit significantly better tensile strength than wrought FeCrAl alloys, which would alloy for the fabrication of a very thin (~250 m) ODS FeCrAl cladding and limit the neutronic penalty from the replacement of Zr-based alloys by Fe-based alloys. Several Fe-12-Cr-5Al ODS alloys where therefore fabricated by ball milling FeCrAl powders with Y2O3 and additional oxides such as TiO2 or ZrO2. The new Fe-12Cr-5Al ODS alloys showed excellent tensile strength up to 800 C but limited ductility. Good oxidation resistance in steam at 1200 and 1400 C was observed except for one ODS FeCrAl alloy containing Ti. Rolling trials were conducted at 300, 600 C and 800 C to simulate the fabrication of thin tube cladding and a plate thickness of ~0.6mm was reached before the formation of multiple edge cracks. Hardness measurements at different stages of the rolling process, before and after annealing for 1h at 1000 C, showed that a thinner plate thickness could likely be achieved by using a multi-step approach combining warm rolling and high temperature annealing. Finally, new Fe-10-12Cr-5.5-6Al-Z gas atomized powders have been purchased to fabricate the second generation of low-Cr ODS FeCrAl alloys. The main goals are to assess the effect of O, C, N and Zr contents on the ODS FeCrAl microstructure and mechanical properties, and to optimize the fabrication process to improve the ductility of the 2nd gen ODS FeCrAl while maintaining good mechanical strength and oxidation resistance.

  18. MARTINS: A foam/film flow model for molten material relocation in HWRs with U-Al-fueled multi-tube assemblies

    International Nuclear Information System (INIS)

    Kalimullah.

    1994-01-01

    Some special purpose heavy-water reactors (EM) are made of assemblies consisting of a number of coaxial aluminum-clad U-Al alloy fuel tubes and an outer Al sleeve surrounding the fuel tubes. The heavy water coolant flows in the annular gaps between the circular tubes. Analysis of severe accidents in such reactors requires a model for predicting the behavior of the fuel tubes as they melt and disrupt. This paper describes a detailed, mechanistic model for fuel tube heatup, melting, freezing, and molten material relocation, called MARTINS (Melting and Relocation of Tubes in Nuclear subassembly). The paper presents the modeling of the phenomena in MARTINS, and an application of the model to analysis of a reactivity insertion accident. Some models are being developed to compute gradual downward relocation of molten material at decay-heat power levels via candling along intact tubes, neglecting coolant vapor hydrodynamic forces on molten material. These models are inadequate for high power accident sequences involving significant hydrodynamic forces. These forces are included in MARTINS

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  20. Corrosion on the fuel plate nucleus based on U3 O8 - Al dispersions

    International Nuclear Information System (INIS)

    Durazzo, M.

    2005-01-01

    Samples of MTR type U 3 O 8 - Al dispersion fuel plates meats were corrosion tested in deionized water at different temperatures in the range 30 to 90 deg C. In the tests the cores were exposed to the deionized water by means of an artificially produced cladding defect. The results indicate that the meat corrosion is accompanied by hydrogen evolution. (author)

  1. Microstructure and Physical Metallurgy in Melt-Dilute Treatment Technology for Al-Based Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Adams, T.M.; Peacock, H.B.

    1998-09-01

    Major challenges associated with the direct disposal of research reactor fuel in a repository include nonproliferation and criticality control, both of which may be a concern for HEU Al-SNF. Consideration must be given to the potential desirability and/or regulatory necessity of diluting the HEU SNF to below 20 percent enrichment. The probability for a criticality event as well as the issue of proliferation are greatly decreased by reducing the enrichment. The melt-dilute technology development program is focused on the development and implementation of a treatment technology for diluting HEU Al-SNF to LEU levels and qualifying this LEU Al-SNF form for geologic repository storage. An attribute of the melt-dilute process is its applicability to various types of FRR and DRR SNF

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  3. Reactivity feedbacks 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 reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. 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. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U 3 O 8 -Al was about 2% more than the original UAl x -Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

  4. La relación entre creatividad y expertise: contribuciones teóricas y empíricas

    Directory of Open Access Journals (Sweden)

    Afonso C. T. Galvao

    2009-06-01

    Full Text Available La creatividad y el expertise son temas que están recibiendo mayor atención de los inves­tigadores en distintos campos del conocimiento. Atributos personales asociados a la alta producción creativa, factores que favorecen el desarrollo, la expresión de la creatividad y el expertise, así como los procesos implicados en el desarrollo de la alta competencia y la creatividad, son temas que han sido objeto de numerosos estudios. El presente estudio pre­senta contribuciones teóricas y estudios empíricos acerca de la creatividad y el expertise. Se describen los elementos comunes para el desarrollo de la creatividad y el expertise, así como las características de los ambientes educativos que los promueven, los modelos teóricos y estudios empíricos.

  5. Psicología comunitaria y expresiones psicosociales de la pobreza: contribuciones para la intervención en políticas públicas

    OpenAIRE

    Morais Ximenes, Verônica; Camurça Cidade, Elívia; Barbosa Nepomuceno, Bárbara

    2015-01-01

    El objetivo de este trabajo es analizar, desde la psicología comunitaria, las expresiones psicosociales de la pobreza y sus contribuciones para la intervención en políticas públicas. La psicología comunitaria critica los factores perpetuadores de aspectos materiales y simbólicos que interfieren en la constitución subjetiva de los pobres. La investigación exploratoria, de carácter cuantitativo y cualitativo, fue realizada con 417 sujetos adultos de dos comunidades de Brasil, una rural y otra u...

  6. A contribution to the analysis of the thermal behaviour of Fast Breeder fuel rods with UO{sub 2}-PuO{sub 2} fuel; Contribucion al analisis del comportamiento termico de las barras combustibles de UO{sub 2}-PuO{sub 2} de los reactores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J; Elbel, H

    1977-07-01

    The fuel of Fast Breeder Reactors which consists of Uranium and Plutonium dioxide is mainly characterized by the amount and distribution of void volume and Plutonium and the amount of oxygen. Irradiation experiments carried out with this fuel have shown that initial structure of the fuel pellet is subjected to large changes during operation. These are consequences of the radial and axial temperature gradients within the fuel rods. (Author) 54 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-06

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

  8. 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

  9. Growth of the interaction layer around fuel particles in dispersion fuel

    International Nuclear Information System (INIS)

    Olander, D.

    2009-01-01

    Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl x . The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel

  10. Use of pyrolysis gases from biogenic fuels as reductionfuels in coal dust furnaces; Einsatz von Pyrolysegasen aus biogenen Brennstoffen als Reduktionsbrennstoff in Kohlestaubfeuerungen

    Energy Technology Data Exchange (ETDEWEB)

    Ruediger, H; Greul, U; Spliethoff, H; Hein, K R.G. [Stuttgart Univ. (Germany). Inst. fuer Verfahrenstechnik und Dampfkesselwesen (IVD)

    1997-12-31

    Co-combustion of refuse-derived fuels in the form of pyrolysis gases, with coal as primary fuel, has advantages in terms of fuel ash separation and nitric oxide emissions. Biomass or sewage sludge is degassed in a pyrolysis reactor, and the gas is used as secondary fuel in a coal dust furnace. The authors investigated the influence of reaction temperature, fuel moisture and reaction atmosphere in the pyrolysis stage on the product fractions gas, tar, and residual fuel, as well as the suitability of the resulting pyrolysis gas as secondary fuel in a coal dust furnace for the purpose of reducing nitric oxide emissions. (orig) [Deutsch] Ein am IVD betriebenes Konzept der Mitverbrennung von Brennstoffen in Form von Pyrolysegasen bietet Vorteile bezueglich der Trennung der Brennstoffaschen und Stickoxidemissionen bei der Feuerung des Primaerbrennstoffes Steinkohle. Biomasse oder Klaerschlamm wird hierbei in einem Pyrolysereaktor engast und gasfoermig als Sekundaerbrennstoff in einer Kohlenstaubfeuerung eingesetzt. Untersuchungsschwerpunkte in der Pyrolysestufe des Prozesses waren die Einfluesse von Reaktionstemperatur, Brennstofffeuchte und Reaktionsatmosphaere auf die Produktfraktionen Gas, Teer und Restbrennstoff sowie die Eignung des erzeugten Pyrolysegases als Sekundaerbrennstoff in einer Kohlenstaubfeuerung zur Senkung derKohlendioxidemissione. (orig)

  11. Use of pyrolysis gases from biogenic fuels as reductionfuels in coal dust furnaces; Einsatz von Pyrolysegasen aus biogenen Brennstoffen als Reduktionsbrennstoff in Kohlestaubfeuerungen

    Energy Technology Data Exchange (ETDEWEB)

    Ruediger, H.; Greul, U.; Spliethoff, H.; Hein, K.R.G. [Stuttgart Univ. (Germany). Inst. fuer Verfahrenstechnik und Dampfkesselwesen (IVD)

    1996-12-31

    Co-combustion of refuse-derived fuels in the form of pyrolysis gases, with coal as primary fuel, has advantages in terms of fuel ash separation and nitric oxide emissions. Biomass or sewage sludge is degassed in a pyrolysis reactor, and the gas is used as secondary fuel in a coal dust furnace. The authors investigated the influence of reaction temperature, fuel moisture and reaction atmosphere in the pyrolysis stage on the product fractions gas, tar, and residual fuel, as well as the suitability of the resulting pyrolysis gas as secondary fuel in a coal dust furnace for the purpose of reducing nitric oxide emissions. (orig) [Deutsch] Ein am IVD betriebenes Konzept der Mitverbrennung von Brennstoffen in Form von Pyrolysegasen bietet Vorteile bezueglich der Trennung der Brennstoffaschen und Stickoxidemissionen bei der Feuerung des Primaerbrennstoffes Steinkohle. Biomasse oder Klaerschlamm wird hierbei in einem Pyrolysereaktor engast und gasfoermig als Sekundaerbrennstoff in einer Kohlenstaubfeuerung eingesetzt. Untersuchungsschwerpunkte in der Pyrolysestufe des Prozesses waren die Einfluesse von Reaktionstemperatur, Brennstofffeuchte und Reaktionsatmosphaere auf die Produktfraktionen Gas, Teer und Restbrennstoff sowie die Eignung des erzeugten Pyrolysegases als Sekundaerbrennstoff in einer Kohlenstaubfeuerung zur Senkung derKohlendioxidemissione. (orig)

  12. Proporcionalidad y equidad en las contribuciones. El amparo fiscal en México, 1917-1968

    Directory of Open Access Journals (Sweden)

    Carlos de Jesús Becerril Hernández

    2015-01-01

    Full Text Available El juicio de amparo ha sido comúnmente identificado como parte del proceso de la modernización jurídica liberal de la segunda mitad del siglo XIX. No obstante, el amparo fiscal, que tiene que ver directamente con el reclamo por la proporcionalidad y equidad de las contribuciones contenidas en el artículo 31, fracción IV, de la Constitución Política de los Estados Unidos Mexicanos de 1917, no fue reconocido como jurisprudencia inmediatamente aplicable; el criterio del ministro Ignacio L. Vallarta (1830-1893 sobre la incapacidad de la Corte para conocer de este recurso fue vigente hasta bien entrado el siglo XX. El amplio arco temporal (1917-1968 cubierto por este ensayo tratará de explicar, a la luz del derecho constitucional y del desempeño económico, por qué la justicia federal se negó a amparar y proteger a los causantes, bajo el argumento de que la Corte no era juez ni órgano revisor de las leyes expedidas por el Congreso.

  13. 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)

  14. Study on the Applicability of Electron Beam Welding Methods to Assembly a Fuel Compact and Al Cover Plate of Research Reactor Plate Type Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hae In; Lee, Yoon Sang; Lee, Don Dae; Jeong, Yong Jin; Kwon, Sun Chil; Kim, Soo Sung; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Among the research reactor plate type fuel fabrication processes, there is an assembly process between fuel meat compact and Al cover plates using a welding method prior to rolling process. The assembly process is such as the Al frame and Al cover plate should be welded properly as shown in Fig. 1. For welding, TIG(Tungsten Inert Gas) welding methods has been used conventionally, but in this study an electron beam welding(EB welding) technique which uses the electron beam of a high velocity for joining two materials is introduced to the assembly. The work pieces are melted as the kinetic energy of the electron beam is transformed into heat to join the two parts of the weld. The welding is often done in the conditions in a vacuum to prevent dispersion of the electron beam. The electron beam welding process has many ad-vantages such as contamination of the welds could be prevented, the penetration of the weld is deep, and also the strain of the welding area is less than other methods. In this study, to find optimal condition of the EB welding process, a welding speed, a beam current and an acceleration voltage were changed. To analyzing the welding results, the shape of the beads and defects of welding area was used. The width and depth of the beads were measured as well

  15. Study on the Applicability of Electron Beam Welding Methods to Assembly a Fuel Compact and Al Cover Plate of Research Reactor Plate Type Fuel

    International Nuclear Information System (INIS)

    Lee, Hae In; Lee, Yoon Sang; Lee, Don Dae; Jeong, Yong Jin; Kwon, Sun Chil; Kim, Soo Sung; Park, Jong Man

    2012-01-01

    Among the research reactor plate type fuel fabrication processes, there is an assembly process between fuel meat compact and Al cover plates using a welding method prior to rolling process. The assembly process is such as the Al frame and Al cover plate should be welded properly as shown in Fig. 1. For welding, TIG(Tungsten Inert Gas) welding methods has been used conventionally, but in this study an electron beam welding(EB welding) technique which uses the electron beam of a high velocity for joining two materials is introduced to the assembly. The work pieces are melted as the kinetic energy of the electron beam is transformed into heat to join the two parts of the weld. The welding is often done in the conditions in a vacuum to prevent dispersion of the electron beam. The electron beam welding process has many ad-vantages such as contamination of the welds could be prevented, the penetration of the weld is deep, and also the strain of the welding area is less than other methods. In this study, to find optimal condition of the EB welding process, a welding speed, a beam current and an acceleration voltage were changed. To analyzing the welding results, the shape of the beads and defects of welding area was used. The width and depth of the beads were measured as well

  16. 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.  

  17. Analysis of UO{sub 2}-BeO fuel under transient using fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-01

    Recent research has appointed the need to replace the classic fuel concept, used in light water reactors. Uranium dioxide has a weak point due to the low thermal conductivity, that produce high temperatures on the fuel. The ceramic composite fuel formed of uranium dioxide (UO{sub 2}), with the addition of beryllium oxide (BeO), presents high thermal conductivity compared with UO{sub 2}. The oxidation of zirconium generates hydrogen gas that can create a detonation condition. One of the preferred options are the ferritic alloys formed of iron-chromium and aluminum (FeCrAl), that should avoid the hydrogen release due to oxidation. In general, the FeCrAl alloys containing 10 - 20Cr, 3 - 5Al, and 0 - 0.12Y in weight percent. The FeCrAl alloys should exhibit a slow oxidation kinetics due to chemical composition. Resistance to oxidation in the presence of steam is improved as a function of the content of chromium and aluminum. In this way, the thermal and mechanical properties of the UO{sub 2}-BeO-10%vol, composite fuel were coupled with FeCrAl alloys and added to the fuel codes. In this work, we examine the fuel rod behavior of UO{sub 2}-10%vol-BeO/FeCrAl, including a simulated transient of reactivity. The fuels behavior shown reduced temperature with UO{sub 2}-BeO/Zr, UO{sub 2}-BeO/FeCrAl also were compared with UO{sub 2}/Zr system. The case reactivity initiated accident analyzed, reproducing the fuel rod called VA-1 using UO{sub 2}/Zr alloys and compared with UO{sub 2}-BeO/FeCrAl. (author)

  18. Status report on the irradiation testing and post-irradiation examination of low-enriched U3O8-Al and UAlsub(x)-Al fuel element by the Netherlands Energy Research Foundation (ECN)

    International Nuclear Information System (INIS)

    Pruimboom, H.; Lijbrink, E.; Otterdijk, K. von; Swanenburg de Veye, R.J.

    1984-01-01

    Within the framework of the RERTR-programme four low-enriched (20%) MTR-type fuel elements have been irradiated in the High Flux Reactor at Petten (The Netherlands) and are presently subjected to postirradiation examination. Two of the elements contain UAlsub(x)-Al and two contain U 3 O 8 -Al fuel. The test irradiation has been completed up to the target burn-up values of 50% and 75% respectively. An extensive surveillance programme carried out during the test period has confirmed the excellent in-reactor behaviour of both types. Post-irradiation examination of the 50% burn-up test elements, comprising of dimensional measurements, burn-up determination, fuel metallography and blister testing, has sofar confirmed the irradiation experiences. Good agreement between calculated and measured power and burn-up characteristics has been found. A survey of the test element characteristics, their irradiation history, the irradiation tests and the preliminary PIE results is given in the paper. (author)

  19. 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)

  20. 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

  1. 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

  2. Evaluation of Metals (Al, Fe, Zn) in Alternative Fuels by Electrochemical Impedance Spectroscopy in Two Electrode Cell

    International Nuclear Information System (INIS)

    Song, Yon Kyun; Lim, Geun Woong; Kim, Hee San

    2010-01-01

    Many kinds of alternative fuels such as biodiesel, ethanol, methanol, and natural gas have been developed in order to overcome the limited deposits in fossil fuels. In some cases, the alternative fuels have been reported to cause degrade materials. The corrosion rates of metals were measured by immersion test, a kind of time consuming test because low conductivity of these fuels was not allowed to employ electrochemical tests. With twin two-electrode cell newly designed for the study, however, electrochemical impedance spectroscopy (EIS) test was successfully applied to evaluation of the corrosion resistance (R p ) of zinc, iron, aluminum, and its alloys in an oxidized biodiesel and gasoline/ethanol solutions and the corrosion resistance from EIS was compared with the corrosion rate from immersion test. In biodiesel, R p increased in the order of zinc, iron, and aluminum, which agreed with the corrosion resistance measured from immersion test. In addition, on aluminum showing the best corrosion resistance (R p ), the effect of magnesium as an alloying element was evaluated in gasoline/ethanol solutions as well as the oxidized biodiesel. R p increased with addition of magnesium in gasoline/ethanol solutions containing chloride and the oxidized biodiesel. In the mean while, in gasoline/ethanol solutions containing formic acid, Al-Mg alloy added 1% magnesium had the highest R p and the further addition of magnesium decreased R p . It can be explained with the fact that the addition of more than 1% magnesium increases the passive current density of Al-Mg alloys

  3. 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)

  4. Estudios etnográficos sobre el desarrollo infantil en comunidades indígenas de América Latina: contribuciones, omisiones y desafíos

    Directory of Open Access Journals (Sweden)

    Carolina Remorini

    2013-09-01

    Full Text Available DOI: http://dx.doi.org/10.5007/2175-795X.2013v31n3p811 En las últimas décadas, ha habido un progreso sustancial en el estudio del Desarrollo Infantil Temprano (DIT desde perspectivas interdisciplinarias e transculturales. Varios autores reconocen la contribución de la antropología al debate sobre la primera infancia, en diálogo con la psicología, las neurociencias, la educación y la medicina. Estos estudios subrayan la importancia de los contextos ecológicos en lo que se refi ere al crecimiento físico, las trayectorias de desarrollo y los efectos sobre la salud a mediano y largo plazo. La Antropología hace hincapié en la noción de que el desarrollo de los niños depende de las interacciones que éstos entablan con su entorno inmediato y las actividades en las que participan. La importancia dada al ambiente desde una perspectiva ecológica reconoce el valor heurístico de los estudios etnográfi cos. En este trabajo caracterizamos, algunos ejes y modos de abordaje que se destacan en la producción etnográfi ca sobre infancia y desarrollo infantil en Latinoamérica, especialmente, sobre comunidades indígenas. En segundo lugar, analizamos y discutimos las posibilidades de interface entre la etnografía y las disciplinas que tradicionalmente se han ocupado del desarrollo infantil y refl exionamos sobre sus contribuciones en términos metodológicos a la comprensión de un proceso de carácter multidimensional, evitando los reduccionismos propios de cada mirada disciplinar. En relación a ello, discutimos los desafíos que la etnografía presenta en el contexto de la investigación inter y transdisciplinaria del DIT. Para concluir, reflexionamos sobre la necesidad de recuperar el DIT como objeto de estudio etnográfi co, como lo fue en los inicios de la disciplina.

  5. Pembuatan Fuel dari Liquid Hasil Pirolisis Plastik Polipropilen Melalui Proses Reforming Dengan Katalis NiO/γ-Al2O3

    Directory of Open Access Journals (Sweden)

    Mahendra Fajri Nugraha

    2013-09-01

    Full Text Available Estimasi jumlah timbulan sampah di Indonesia pada tahun 2008 mencapai 38,5 juta ton/tahun. Melihat dari sifat penyusun plastik yang tersusun dari komponen hidrokarbon minyak bumi, maka limbah plastik sangat berpotensi untuk dikonversi menjadi BBM. Tujuan penelitian ini Mempelajari proses konversi limbah plastik khususnya jenis polipropilen (PP menjadi fuel serta pengaruh berbagai macam komposisi katalis NiO/γ-Al2O3, temperatur, laju alir reaktan pada reactor reforming terhadap kualitas fuel (yield aromatis yang dihasilkan. Pada penelitian ini bahan baku yang digunakan merupakan plastik jenis Polipropilen (PP. Pada penelitian ini minyak yang telah dihasilkan pada proses pirolisis selanjutnya akan di reforming. Pada penelitian ini digunakan logam NiO dengan penyangga γ-Al2O3 (NiO/γ-Al2O3 sebagai katalis untuk proses reforming minyak hasil pirolisis plastik polipropilen. Variabel penelitian meliputi Loading Ni (% massa : 6; 10; 14, Laju alir (ml/jam : 2I7; 500; 690, Suhu reaksi (oC : 400; 450; 500. Berdasarkan hasil penelitian dan analisa diketahui bahwa % yield aromatis terbesar pada proses reforming minyak hasil pirolisis plastik polipropilen dihasilkan dengan kondisi operasi 14 % loading Ni pada katalis, temperatur reforming 500oC serta laju reaktan sebesar 217 mL/jam. Dengan kata lain dapat disimpulkan bahwa pada penelitian ini, hasil terbaik didapat pada variabel flowlaju terendah dan variabel suhu tertinggi. Kondisi operasi efektif dalam pembuatan fuel pada proses reforming diperoleh saat loading Ni pada katalis NiO/γ-Al2O3 14 %, temperatur reforming 400oC serta laju reaktan 500 mL/jam.

  6. 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

  7. Urea-nitrate combustion synthesis of MgO/MgAl2O4 nanocatalyst used in biodiesel production from sunflower oil: Influence of fuel ratio on catalytic properties and performance

    International Nuclear Information System (INIS)

    Rahmani Vahid, Behgam; Haghighi, Mohammad

    2016-01-01

    Graphical abstract: As a base catalyst for biodiesel production, MgAl 2 O 4 spinel was successfully synthesized by combustion method with MgO, as the active phase, dispersed on the catalyst surface. The nanocatalysts were characterized by XRD, FESEM, EDX, BET-BJH, TGA and FTIR analyses, so as to optimize the concentration of urea (as fuel) in the combustion synthesis. Analyzing the effect of fuel ratio on the combustion synthesized MgAl 2 O 4 , it was revealed that the synthesized base catalyst with a fuel ratio of 1.5 was of the best specifications for biodiesel production process. Future researches may investigate the catalyst reusability and mild reaction conditions, so as to achieve more economical production of biodiesel. - Highlights: • Efficient synthesis of MgAl 2 O 4 spinel by solution combustion method. • Improvement of catalytic activity and stability by optimum ratio fuel. • Enhanced dispersion of MgO over MgAl 2 O 4 spinel. • Production of biodiesel over MgO/MgAl 2 O 4 at relatively mild reaction conditions. - Abstract: MgO/MgAl 2 O 4 nanocatalyst was synthesized by a simple, cost-effective and rapid method and used in biodiesel production from sunflower oil. MgAl 2 O 4 was synthesized by combustion method at different fuel ratios and then active phase of MgO was dispersed on the samples by impregnation method. The nanocatalysts were characterized by XRD, FESEM, EDX, BET-BJH, TGA and FTIR analyses, so as to optimize the concentration of urea (as fuel) in the combustion synthesis. The physicochemical properties of the nanocatalyst confirmed the sample synthesized with fuel ratio of 1.5 has high surface area, effective morphology and texture properties. Finally, in order to evaluate catalytic activity of the samples in biodiesel production, the transesterification reaction was performed. The results indicated the catalyst prepared by combustion synthesis with a fuel ratio of 1.5 was optimum specifications for biodiesel production. Using this

  8. Contribuciones del paradigma cultural latinoamericano a la comunicación para el desarrollo. Antecedentes, textos y contextos de una relación fecunda

    OpenAIRE

    Salazar Martínez, Rafael Ángel; Universidad de Holguín “Oscar Lucero Moya”,; Portal Moreno, Rayza; Universidad de La Habana; Fonseca Valido, Rafael Ángel; Universidad de Oriente

    2016-01-01

    Durante la década de 1980 se produce una ruptura en la investigación comunicológica latinoamericana, cuyo resultado fue el nacimiento de un nuevo paradigma que se apropia y hace uso de una concepción antropológica de la cultura, desde la cual comienzan a abordarse los procesos comunicativos. Las contribuciones de este paradigma a la comunicación para el desarrollo, campo que en América Latina ya contaba con una larga tradición, han resultado particularmente significativas. A examinar y discut...

  9. 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.

  10. The development of lower enrichment fuels for Canadian research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A; Belanger, L; Grolway, C M [AECL, Atomic Energy of Canada Limited, Chalk River, ON (Canada); Foo, M T [CRNL, Combustion Engineering Superheater Ltd., Moncton, NB (Canada)

    1983-08-01

    As part of the world wide move to proliferation resistant fuels, new fuels which use reduced enrichment uranium are being developed for use in the NRX and NRU reactors. A fuel consisting of particles of a USiAl alloy dispersed in an Al matrix has been selected for development along with Al-37 wt% U alloy and Al-U{sub 3}O{sub 8} cermet as backup fuels. This report outlines the progress made in the development of the Al-USiAl and Al-37 wt% U. Results show that good quality extruded rods containing either fuel can be made with techniques similar to those used to fabricate the current NRX and NRU fuels. However, the new fuels will be more expensive to make. Although the oxidation behaviour of the Al-USiAl is not as good as that of the Al-U alloys, its corrosion behaviour in high temperature water does not seem much worse. The oxidation and aqueous corrosion of A-37 wt% U are not much different from those of the Al-U alloys currently used. (author)

  11. Balancing low cost with reliable operation in the rotordynamic design of the ALS Liquid Hydrogen Fuel Turbopump

    Science.gov (United States)

    Greenhill, L. M.

    1990-01-01

    The Air Force/NASA Advanced Launch System (ALS) Liquid Hydrogen Fuel Turbopump (FTP) has primary design goals of low cost and high reliability, with performance and weight having less importance. This approach is atypical compared with other rocket engine turbopump design efforts, such as on the Space Shuttle Main Engine (SSME), which emphasized high performance and low weight. Similar to the SSME turbopumps, the ALS FTP operates supercritically, which implies that stability and bearing loads strongly influence the design. In addition, the use of low cost/high reliability features in the ALS FTP such as hydrostatic bearings, relaxed seal clearances, and unshrouded turbine blades also have a negative influence on rotordynamics. This paper discusses the analysis conducted to achieve a balance between low cost and acceptable rotordynamic behavior, to ensure that the ALS FTP will operate reliably without subsynchronous instabilities or excessive bearing loads.

  12. 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.

  13. Contribuciones de la modelización al desarrollo de las competencias básicas

    Directory of Open Access Journals (Sweden)

    Manel Sol

    2013-06-01

    Full Text Available For many years modelling activities have not been a common practice in the classroom despite having a wide recognition of their educational interest. Burkhardt i Pollack (2006 justify it by a lack of supporting materials and Antonius et al. (2007 indicate an insufficient guidance on teaching strategies. From this starting point some ideas are proposed to overcome these difficulties. The article is organized in three sections. In the first section shown relationship between modelling and curriculum. Moreover, the catalan curriculum is highlighted as it promotes mathematical projects, in the secondary education. In fact, project research (Projecte de recerca and Research work (Treball de recerca are two compulsory subjects in Catalonia which fit in with modelling activities. The second section, some activities are shown that can serve as an example to teachers. And also the teaching of skills in modelling activities. The last section shows the contributions of modelling activities to the development of basic competencies of current curriculum thus increasing their interest.

  14. Single step synthesis of GdAlO3 powder

    International Nuclear Information System (INIS)

    Sinha, Amit; Nair, S.R.; Sinha, P.K.

    2011-01-01

    Research highlights: → First report on direct formation of GdAlO 3 powder using a novel combustion process. → Study of combustion characteristics of Gd(NO 3 ) 3 and Al(NO 3 ) 3 towards three fuels. → Preparation of highly sinterable GdAlO 3 powders through fuel-mixture approach. → Significant reduction in energy consumption for production of GdAlO 3 sintered body. - Abstract: A novel method for preparation of nano-crystalline gadolinium aluminate (GdAlO 3 ) powder, based on combustion synthesis, is reported. It was observed that aluminium nitrate and gadolinium nitrate exhibit different combustion characteristics with respect to urea, glycine and β-alanine. While urea was proven to be a suitable fuel for direct formation of crystalline α-Al 2 O 3 from its nitrate, glycine and β-alanine are suitable fuels for gadolinium nitrate for preparation of its oxide after combustion reaction. Based on the observed chemical characteristics of gadolinium and aluminium nitrates with respect to above mentioned fuels for the combustion reaction, the fuel mixture composition could be predicted that could lead to phase pure perovskite GdAlO 3 directly after the combustion reaction without any subsequent calcination step. The use of single fuel, on the other hand, leads to formation of amorphous precursor powders that call for subsequent calcination for the formation of crystalline GdAlO 3 . The powders produced directly after combustion reactions using fuel mixtures were found to be highly sinterable. The sintering of the powders at 1550 o C for 4 h resulted in GdAlO 3 with sintered density of more than 95%. T.D.

  15. Pt/Al/sub 2/O/sub 3/- carbon nanocomposite as a catalyst for fuel cells

    International Nuclear Information System (INIS)

    Naeem, R.; Ahmed, R.; Ansari, M.S.

    2013-01-01

    Catalysts comprising platinum nanoparticles (Pt NPs) on carbon support are used in fuel cells for the hydrogen and electricity production by electrochemical oxidation of methanol. However, the catalyst is not the best in terms of its performance. Considering role of the support as significant towards efficiency and durability of the catalyst, there is need for introducing novel support materials to replace carbon alone. Deposition of various metallic NPs on ceramic-carbon (hybrid) supports has been reported to improve thermal, mechanical, electrical and chemical properties of different types of catalyst. In search of better performing catalysts for proton exchange membrane fuel cells (PEMFCs), hybrid supports having different ceramic materials should be synthesized. In this regard Pt/Al/sub 2/O/sub 3/-Carbon (nanocomposites) have been synthesized and applied as promising catalysts in the PEMFCs; results obtained for the nanocomposites were compared with Pt/carbon and Pt/Al/sub 2/O/sub 3/. Vulcan carbon was purified and functionalized prior to use; presence of oxygen containing functional groups on carbon was established from the FTIR spectrum, Hybrid support (1:8 by weight ratio of ceramic and carbon) were already prepared in aqueous 2-propanol employing sonication method on to which Pt NPs (10% by weight in all the cases) were deposited by simple chemical reduction of PtCl/sub 4/ by NaBH/sub 4/ under controlled conditions. The catalysts were subjected to various characterization techniques like TGA (for thermal stability), EDX (for chemical composition), SEM (for surface morphology) and XRD (for cell-shape and -volume, material density and average crystalline size). Catalysts efficiencies for the methanol oxidation were investigated through cyclic voltammetery (CV) by comparing electrochemical surface area, peak current, exchange current density and rate constant in the acidic and basic media. Pt/Al/sub 2/O/sub 3/-carbon exhibited better catalytic efficiencies

  16. Influence of a doping by Al stainless steel on kinetics and character of interaction with the metallic nuclear fuel

    Science.gov (United States)

    Nikitin, S. N.; Shornikov, D. P.; Tarasov, B. A.; Baranov, V. G.

    2016-04-01

    Metallic nuclear fuel is a perspective kind of fuel for fast reactors. In this paper we conducted a study of the interaction between uranium-molybdenum alloy and ferritic- martensitic steels with additions of aluminum at a temperature of 700 ° C for 25 hours. The rate constants of the interaction layer growth at 700 °C is about 2.8.10-14 m2/s. It is established that doping Al stainless steel leads to decrease in interaction with uranium-molybdenum alloys. The phase composition of the interaction layer is determined.

  17. 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)

  18. Interdiffusion between U-Mo alloys and Al

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  19. Development of low-Cr ODS FeCrAl alloys for accident-tolerant fuel cladding

    Science.gov (United States)

    Dryepondt, Sebastien; Unocic, Kinga A.; Hoelzer, David T.; Massey, Caleb P.; Pint, Bruce A.

    2018-04-01

    Low-Cr oxide dispersion strengthened (ODS) FeCrAl alloys were developed as accident tolerant fuel cladding because of their excellent oxidation resistance at very high temperature, high strength and improved radiation tolerance. Fe-12Cr-5Al wt.% gas atomized powder was ball milled with Y2O3+FeO, Y2O3+ZrO2 or Y2O3+TiO2, and the resulting powders were extruded at 950 °C. The resulting fine grain structure, particularly for the Ti and Zr containing alloys, led to very high strength but limited ductility. Comparison with variants of commercial PM2000 (Fe-20Cr-5Al) highlighted the significant impact of the powder consolidation step on the alloy grain size and, therefore, on the alloy mechanical properties at T < 500 °C. These low-Cr compositions exhibited good oxidation resistance at 1400 °C in air and steam for 4 h but could not form a protective alumina scale at 1450 °C, similar to observations for fine grained PM2000 alloys. The effect of alloy grain size, Zr and Ti additions, and impurities on the alloy mechanical and oxidation behaviors are discussed.

  20. Contribuciones al conocimiento de la flora de navarra

    Directory of Open Access Journals (Sweden)

    BALDA, Angel

    2002-01-01

    Full Text Available Nuevos datos acerca de 58 táxones de plantas de interés, bien por su rareza en el terrirorio navarro o por la ampliación de área que suponen. De ellos, 12 se citan por primera vez para Navarra : Epilobium angustifolium, Fraxinus pennylvanica, Galinsoga quadriradiata, Gamochaeta falcata, Isatis tinctoria subsp. tinctoria, Ludwigia palustris, Lycopodeilla inundata, Pseudorchis albida, Ramonda myconi, Rynchospora alba, Rynchospora fusca y Spiranthes aestivalis

  1. Facile and rapid auto-combustion synthesis of nano-porous γ-Al2O3 by application of hexamethylenetetramine in fuel composition

    Science.gov (United States)

    Salem, Shiva; Salem, Amin; Parni, Mohammad Hosein; Jafarizad, Abbas

    2018-06-01

    In this article, urea, glycine and hexamethylenetetramine were blended in accordance with the mixture design algorithm to prepare γ-Al2O3 by auto-combustion technique. Aluminum nitrate was then mixed with the stoichiometric contents of prepared fuel solutions to obtain gel systems. The gels exhibited a typical self-propagating combustion behavior at low temperature, directly resulting amorphous materials. The precursors were calcined at various temperatures ranging from 700 to 900 °C. The treated powders were evaluated by determining the methylene blue (MB) adsorption efficiency. The production condition to obtain γ-Al2O3 with maximum surface area depends on fuel composition and calcination temperature. The alumina powder fabricated by this procedure was uniformly distributed and contains nano-sized secondary particles with diameter about 10-30 nm in which the average pore size is 3.2 nm induced large surface area, 240 m2g-1. The employment of hexamethylenetetramine provides a potential for synthesis of γ-Al2O3 at lower temperature, 700 °C, with maximum MB removal efficiency.

  2. High temperature steam oxidation of Al3Ti-based alloys for the oxidation-resistant surface layer on Zr fuel claddings

    International Nuclear Information System (INIS)

    Park, Jeong-Yong; Kim, Il-Hyun; Jung, Yang-Il; Kim, Hyun-Gil; Park, Dong-Jun; Choi, Byung-Kwon

    2013-01-01

    We investigated the feasibility to apply Al 3 Ti-based alloys as the surface layer for improving the oxidation resistance of Zr fuel claddings under accident conditions. Two types of Al 3 Ti-based alloys with the compositions of Al–25Ti–10Cr and Al–21Ti–23Cr in atomic percent were prepared by arc-melting followed by homogenization annealing at 1423 K for 48 h. Al–25Ti–10Cr alloy showed an L1 2 quasi-single phase microstructure with a lot of needle-shaped minor phase and pores. Al–21Ti–23Cr alloy consisted of an L1 2 matrix and Cr 2 Al as the second phase. Al 3 Ti-based alloys showed an extremely low oxidation rate in a 1473 K steam for up to 7200 s when compared to Zircaloy-4. Both alloys exhibited almost the same oxidation rate in the early stage of oxidation, but Al–25Ti–10Cr showed a little lower oxidation rate after 4000 s than Al–21Ti–23Cr. The difference in the oxidation rate between two types of Al 3 Ti-based alloys was too marginal to distinguish the oxidation behavior of each alloy. The resultant oxide exhibited almost the same characteristics in both alloys even though the microstructure was explicitly distinguished from each other. The crystal structure of the oxide formed up to 2000 s was identified as Al 2 O 3 in both alloys. The oxide morphology consisted of columnar grains whose length was almost identical to the average oxide thickness. On the basis of the results obtained, it is considered that Al 3 Ti-based alloy is one of the promising candidates for the oxidation-resistant surface layer on Zr fuel claddings

  3. Contribution to swelling analysis in ceramic nuclear fuels; Contribucion al analisis del hinchamiento de combustibles nucleares ceramicos

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Gonzalez, A

    1976-07-01

    In this work a swelling phenomenological model containing several experimental observation is presented. Observations such as dissolution and redi solution of gas atoms, bubble nucleations in defects, dislocations and grain boundary, bubbles migration and coalescence are included. The model gives a bubbles distribution according to their size and position the swelling rate is calculated through the bubble distribution changes. (Author) 17 refs.

  4. The role of various fuels on microwave-enhanced combustion synthesis of CuO/ZnO/Al2O3 nanocatalyst used in hydrogen production via methanol steam reforming

    International Nuclear Information System (INIS)

    Ajamein, Hossein; Haghighi, Mohammad; Alaei, Shervin

    2017-01-01

    Graphical abstract: CuO/ZnO/Al 2 O 3 nanocatalysts were synthesized by the fast and simple microwave enhanced combustion method. Considering that the fuel type is one of the effective parameters on quality of the prepared nanocatalysts, different fuels such as sorbitol, propylene glycol, glycerol, diethylene glycol and ethylene glycol were used. XRD, FESEM, FTIR, EDX, and BET analyses were applied to determine the physicochemical properties of fabricated nanocatalysts. The catalytic experiments were performed in a fixed bed reactor in the temperature range of 160–300 °C. The characteristic and reactivity properties of fabricated nanocatalysts proved that ethylene glycol is the suitable fuel for preparation of CuO/ZnO/Al 2 O 3 nanocatalysts via microwave enhanced combustion method. - Highlights: • Microwave combustion synthesis of CuO/ZnO/Al 2 O 3 nanocatalysts by different fuels. • Enhancement of methanol conversion at low temperatures by selecting proper fuel. • Providing a large number of combustion pores by application of ethylene glycol as fuel. • Increase of CO selectivity in steam methanol reforming by Zn(0 0 2) crystallite facet. - Abstract: A series of CuO/ZnO/Al 2 O 3 nanocatalysts were synthesized by the microwave enhanced combustion method to evaluate the influence of fuel type. Sorbitol, propylene glycol, glycerol, diethylene glycol and ethylene glycol were used as fuel. XRD results revealed that application of ethylene glycol led to highly dispersed CuO and ZnO crystals. It was more highlighted about Cu(1 1 1) crystallite facet which known as the main active site of methanol steam reforming. Moreover, using ethylene glycol resulted homogeneous morphology and narrow particles size distribution (average surface particle size is about 265 nm). Due to the significant physicochemical properties, the catalytic experiments showed that the sample prepared by ethylene glycol achieved total conversion of methanol at 260 °C. Its carbon monoxide

  5. Atomistic modeling and simulation of the role of Be and Bi in Al diffusion in U-Mo fuel

    Science.gov (United States)

    Hofman, G. L.; Bozzolo, G.; Mosca, H. O.; Yacout, A. M.

    2011-07-01

    Within the RERTR program, previous experimental and modeling studies identified Si as the alloying addition to the Al cladding responsible for inhibiting Al interdiffusion in the UMo fuel. However, difficulties with reprocessing have rendered this choice inappropriate, leading to the need to study alternative elements. In this work, we discuss the results of an atomistic modeling effort which allows for the systematic study of several possible alloying additions. Based on the behavior observed in the phase diagrams, beryllium or bismuth additions suggest themselves as possible options to replace Si. The results of temperature-dependent simulations using the Bozzolo-Ferrante-Smith (BFS) method for the energetics for varying concentrations of either element are shown, indicating that Be could have a substantial effect in stopping Al interdiffusion, while Bi does not. Details of the calculations and the dependence of the role of each alloying addition as a function of temperature and concentration (of beryllium or bismuth in Al) are shown.

  6. Atomistic modeling and simulation of the role of Be and Bi in Al diffusion in U-Mo fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hofman, G.L. [Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States); Bozzolo, G., E-mail: guille_bozzolo@yahoo.com [Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States); Mosca, H.O. [Gerencia de Investigaciones y Aplicaciones, CNEA, Av. Gral Paz 1499, B165KNA, San Martin, Buenos Aires (Argentina); Yacout, A.M. [Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States)

    2011-07-15

    Within the RERTR program, previous experimental and modeling studies identified Si as the alloying addition to the Al cladding responsible for inhibiting Al interdiffusion in the UMo fuel. However, difficulties with reprocessing have rendered this choice inappropriate, leading to the need to study alternative elements. In this work, we discuss the results of an atomistic modeling effort which allows for the systematic study of several possible alloying additions. Based on the behavior observed in the phase diagrams, beryllium or bismuth additions suggest themselves as possible options to replace Si. The results of temperature-dependent simulations using the Bozzolo-Ferrante-Smith (BFS) method for the energetics for varying concentrations of either element are shown, indicating that Be could have a substantial effect in stopping Al interdiffusion, while Bi does not. Details of the calculations and the dependence of the role of each alloying addition as a function of temperature and concentration (of beryllium or bismuth in Al) are shown.

  7. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    Science.gov (United States)

    Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

    2015-06-01

    FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

  8. 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

  9. 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

  10. Study of the aqueous corrosion mechanisms and kinetics of the AlFeNi aluminium based alloy used for the fuel cladding in the Jules Horowitz research reactor

    International Nuclear Information System (INIS)

    Wintergerst, M.

    2009-05-01

    For the Jules Horowitz new material-testing reactor (JHR), an aluminium base alloy, called AlFeNi, will be used for the cladding of the fuel plates. This alloy (Al - 1% Fe - 1% Ni - 1 % Mg), which is already used as fuel cladding, was developed for its good corrosion resistance in water at high temperatures. However, few studies dealing with the alteration process in water and the relationships with irradiation effects have been performed on this alloy. The conception of the JHR fuel requires a better knowledge of the corrosion mechanisms. Corrosion tests were performed in autoclaves at 70 C, 165 C and 250 C on AlFeNi plates representative of the fuel cladding. Several techniques were used to characterize the corrosion scale: SEM, TEM, EPMA, XRD, Raman spectroscopy. Our observations show that the corrosion scale is made of two main layers: a dense amorphous scale close to the metal and a porous crystalline scale in contact with the water. More than the morphology, the chemical compositions of both layers are different. This duplex structure results from a mixed growth mechanism: an anionic growth to develop the inner oxide and a cationic diffusion followed by a dissolution-precipitation process to form the outer one. Dynamic experiments at 70 C and corrosion kinetics measurements have demonstrated that the oxide growth process is controlled by a diffusion step associated to a dissolution/precipitation process. A corrosion mechanism of the AlFeNi alloy in aqueous media has been proposed. Then post-irradiation exams performed on irradiated fuel plates were used to investigate the effects of the irradiation on the corrosion behaviour in the reactor core. (author)

  11. BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

    Energy Technology Data Exchange (ETDEWEB)

    Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sweet, Ryan T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling. As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations

  12. 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.)

  13. Neutronics analysis on mini test fuel in the RSG-GAS core

    International Nuclear Information System (INIS)

    Tukiran S; Tagor M Sembiring

    2016-01-01

    Research on UMo fuel for research reactor has been developed. The fuel of research reactor is uranium molybdenum low enrichment with high density. For supporting the development of fuel fabrication, an neutronic analysis of mini fuel plates in the RSG-GAS core was performed. The aim of analysis is to determine the numbers of fuel cycles in the core to know the maximum fuel burn-up. The mini fuel plates of U_7Mo-Al and U_6Zr-Al with densities of 7.0 gU/cc and 5.2 gU/cc, respectively, will be irradiated in the RSG-GAS core. The size of both fuels, namely 630 x 70.75 x 1.30 mm were inserted to the 3 plates of dummy fuel. Before the fuel will be irradiated in the core, a calculation for safety analysis from neutronics and thermal-hydraulics aspects were required. However, in this paper, it will be discussed safety analysis of the U_7Mo-Al and U_6Zr-Al mini fuels from neutronic point of view. The calculation was done using WIMSD-5B and Batan-3DIFF codes. The result showed that both of the mini fuels could be irradiated in the RSG-GAS core with burn up less than 70 % within 12 cycles of operation without over limiting the safety margin. If it is compared, the power density of U_7Mo-Al mini fuel is bigger than U_6Zr-Al fuel. (author)

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  15. Contribuciones al desarrollo de la investigación en enfermería: retos y perspectivas

    Directory of Open Access Journals (Sweden)

    Myriam Durán Parra

    2012-12-01

    Full Text Available Desde la etimología del término “investigación” la palabra proviene del latín in (en y vestigare (hallar, inquirir, indagar, seguir vestigios. De ahí el uso más elemental del término en el sentido de “averiguar o describir alguna cosa”. Desde el momento en que el hombre se enfrentó a problemas y frente a ellos empezó a interrogarse sobre el porqué, como y para que, con esta indagación sobre las cosas, de una manera embrionaria, comenzó lo que hoy llamamos investigación. Con el transcurso del tiempo el ser humano ha hecho un esfuerzo importante por sobrevivir y la mejor forma de lograrlo ha sido experimentando aportando al desarrollo del conocimiento científico.Donaldson y Bottorff (1978 expresan frente a la investigación en enfermería que “es la fuente de desarrollo de conocimiento que le da la característica de disciplina, que ha de estar presente tanto en la práctica clínica como en el desarrollo de la administración y la educación, que ha de dar respuesta a las necesidades sociales que están relacionadas con procesos de salud-enfermedad, condiciones y calidad de vida de las personas, familias y comunidades, en todos los grupos etarios” (1. Por su parte Borrero-Cabal (1994, afirma que “la disciplina conlleva al sentido de rigor, de dedicación, de entrenamiento y ejercicio de los hábitos científicos de la persona para elaborar, transmitir y aprender una ciencia” (2. En enfermería esto se traduce que para poder llevar el cuidado a los pacientes de una manera técnica, científica , analítica, debe tener unas bases como la dedicación, el amor por la profesión y el gusto por la investigación reconociendo que la enfermería avanza y se fortalece a medida que crece científicamente, mientras conoce la cultura, la riqueza de la naturaleza, forma parte del país y sobre todo del mundo como una profesión que demuestra su interés en la búsqueda de su propia identidad. 

  16. Irradiation behavior of uranium oxide - Aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products and as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show that, with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 g U/cm 3 ) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼63% 235 U burnup). (author)

  17. Irradiation behavior of uranium oxide-aluminum dispersion fuel

    International Nuclear Information System (INIS)

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

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show, that with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 gm/cc) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼ 63% 235 U burnup)

  18. Multilayer (TiN, TiAlN) ceramic coatings for nuclear fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Alat, Ece, E-mail: exa179@psu.edu [Department of Materials Science and Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Motta, Arthur T. [Department of Materials Science and Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Comstock, Robert J.; Partezana, Jonna M. [Westinghouse Electric Co., Beulah Rd, Pittsburgh, PA 1332 (United States); Wolfe, Douglas E. [Department of Materials Science and Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Applied Research Laboratory, The Pennsylvania State University, 119 Materials Research Building, University Park, PA 16802 (United States)

    2016-09-15

    In an attempt to develop an accident-tolerant fuel (ATF) that can delay the deleterious consequences of loss-of-coolant-accidents (LOCA), multilayer coatings were deposited onto ZIRLO{sup ®} coupon substrates by cathodic arc physical vapor deposition (CA-PVD). Coatings were composed of alternating TiN (top) and Ti{sub 1-x}Al{sub x}N (2-layer, 4-layer, 8-layer and 16-layer) layers. The minimum TiN top coating thickness and coating architecture were optimized for good corrosion and oxidation resistance. Corrosion tests were performed in static pure water at 360 °C and 18.7 MPa for up to 90 days. The optimized coatings showed no spallation/delamination and had a maximum of 6 mg/dm{sup 2} weight gain, which is 6 times smaller than that of a control sample of uncoated ZIRLO{sup ®} which showed a weight gain of 40.2 mg/dm{sup 2}. The optimized architecture features a ∼1 μm TiN top layer to prevent boehmite phase formation during corrosion and a TiN/TiAlN 8-layer architecture which provides the best corrosion performance. - Highlights: • The first study on multilayer TiAlN and TiN ceramic coatings on ZIRLO{sup ®} coupons. • Corrosion tests were performed at 360°C and 18.7 MPa for up to 90 days. • Coatings adhered well to the substrate, and showed no spallation/delamination. • Weight gains were six times lower than those of uncoated ZIRLO{sup ®} samples. • Longer and higher temperature corrosion tests will be discussed in a further paper.

  19. Partial oxidation of jet fuels over Rh/Al_2O_3. Design and reaction kinetics of sulfur-containing surrogates

    International Nuclear Information System (INIS)

    Baer, Julian Nicolaas

    2016-01-01

    The conversion of logistic fuels via catalytic partial oxidation (CPOX) on Rh/Al_2O_3 at short contact times is an efficient method for generating hydrogen-rich synthesis gas. Depending on the inlet conditions, fuel, and catalyst, high syngas yields, low by-product formation, and rates of high fuel conversion can be achieved. CPOX is relevant for mobile hydrogen generation, e.g., on board of airplanes in order to increase the fuel efficiency via fuel cell-based auxiliary power units. Jet fuels contain hundreds of different hydrocarbons and a significant amount of sulfur. The hydrocarbon composition and sulfur content of a jet fuel vary depending on distributor, origin, and refinement of the crude oil. Little is known about the influence of the various compounds on the synthesis-gas yield and the impact of sulfur on the product yield. In this work, the influence of three main chemical compounds of a jet fuel (aromatics, alkanes, and sulfur compounds) on syngas selectivity, the catalyst deactivation process, and reaction sequence is unraveled. As representative components of alkanes and aromatics, n-dodecane and 1,2,4-trimethylbenzene were chosen for ex-situ and in-situ investigations on the CPOX over Rh/Al_2O_3, respectively. Additionally, for a fixed paraffin-to-aromatics ratio, benzothiophene or dibenzothiophene were added as a sulfur component in three different concentrations. The knowledge gained about the catalytic partial oxidation of jet fuels and their surrogates is used to identify requirements for jet fuels in mobile applications based on CPOX and to optimize the overall system efficiency. The results show an influence of the surrogate composition on syngas selectivity. The tendency for syngas formation increases with higher paraffin contents. A growing tendency for by-product formation can be observed with increasing aromatics contents in the fuel. The impact of sulfur on the reaction system shows an immediate change in the product distribution. An

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  1. Status of LEU fuel development and conversion of NRU

    International Nuclear Information System (INIS)

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

    1991-01-01

    This paper reviews the status of the LEU conversion program and the progress made in the fuel development program over the last year. The results from post-irradiation examinations of prototype NRU fuel rods containing Al-U 3 Si dispersion fuel, and of mini-elements containing Al-U 3 Si 2 dispersion fuel, are presented. (orig.)

  2. 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

  3. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

    1982-01-01

    Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

  4. Part II: Oxidative Thermal Aging of Pd/Al2O3 and Pd/CexOy-ZrO2 in Automotive Three Way Catalysts: The Effects of Fuel Shutoff and Attempted Fuel Rich Regeneration

    Directory of Open Access Journals (Sweden)

    Qinghe Zheng

    2015-10-01

    Full Text Available The Pd component in the automotive three way catalyst (TWC experiences deactivation during fuel shutoff, a process employed by automobile companies for enhancing fuel economy when the vehicle is coasting downhill. The process exposes the TWC to a severe oxidative aging environment with the flow of hot (800 °C–1050 °C air. Simulated fuel shutoff aging at 1050 °C leads to Pd metal sintering, the main cause of irreversible deactivation of 3% Pd/Al2O3 and 3% Pd/CexOy-ZrO2 (CZO as model catalysts. The effect on the Rh component was presented in our companion paper Part I. Moderate support sintering and Pd-CexOy interactions were also experienced upon aging, but had a minimal effect on the catalyst activity losses. Cooling in air, following aging, was not able to reverse the metallic Pd sintering by re-dispersing to PdO. Unlike the aged Rh-TWCs (Part I, reduction via in situ steam reforming (SR of exhaust HCs was not effective in reversing the deactivation of aged Pd/Al2O3, but did show a slight recovery of the Pd activity when CZO was the carrier. The Pd+/Pd0 and Ce3+/Ce4+ couples in Pd/CZO are reported to promote the catalytic SR by improving the redox efficiency during the regeneration, while no such promoting effect was observed for Pd/Al2O3. A suggestion is made for improving the catalyst performance.

  5. Effects of MnO-Al2O3 on the grain growth and high-temperature deformation strain of UO2 fuel pellets

    International Nuclear Information System (INIS)

    Kang, Ki Won; Yang, Jae Ho; Kim, Jong Hun; Rhee, Young Woo; Kim, Dong Joo; Kim, Keon Sik; Song, Kun Woo

    2010-01-01

    The fabrication and high-temperature deformation strain of MnO-Al 2 O 3 -doped UO 2 pellets were studied. The effects of additive composition and amount on the microstructure evolution of a UO 2 pellet were investigated. The compressive creep behaviors of MnO-Al 2 O 3 -doped UO 2 pellets were examined. The results indicated that a MnO-Al 2 O 3 binary additive can effectively promote the grain growth of UO 2 pellets. In addition, the high-temperature deformation strain of the UO 2 pellet can be improved significantly with 1,000 ppm 95MnO-5Al 2 O 3 (mol%). The developed MnO-Al 2 O 3 -additive-containing UO 2 pellets can be a potential candidate for a high-burn-up fuel and a pellet-cladding interaction (PCI) remedy. (author)

  6. 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.

  7. Dissolution flowsheet for high flux isotope reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Foster, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

  8. Comunes urbanos: de la gestión colectiva al derecho a la ciudad

    Directory of Open Access Journals (Sweden)

    Mauro Castro-Coma

    2016-01-01

    Full Text Available Existe un creciente interés y una emergente literatura académica en los estudios urbanos sobre los comunes y su aplicación. Por un lado, destacan aquellas aportaciones que aplican al ámbito urbano, los estudios de Elinor Ostrom sobre las formas de autogestión de los recursos naturales de uso común. Por el otro, la literatura crítica sobre el neoliberalismo, las nuevas formas de acumulación por desposesión y las resistencias a ellas en forma de movimientos sociales también han abordado los comunes en la ciudad desde diferentes perspectivas de origen marxista. En este trabajo se ordenan y se analizan las contribuciones de ambas aproximaciones, sus puntos críticos, así como sus posibles complementariedades, para avanzar hacia una mejor conceptualización y una futura agenda de investigación sobre los comunes urbanos.

  9. CONTAINMENT EVALUATION OF BREACHED AL-SNF FOR CASK TRANSPORT

    International Nuclear Information System (INIS)

    Vinson, D. W.; Sindelar, R. L.; Iyer, N. C.

    2005-01-01

    Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site. To enter the U.S., the cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Al-SNF is subject to corrosion degradation in water storage, and many of the fuel assemblies are ''failed'' or have through-clad damage. A methodology has been developed with technical bases to show that Al-SNF with cladding breaches can be directly transported in standard casks and maintained within the allowable release rates. The approach to evaluate the limiting allowable leakage rate, L R , for a cask with breached Al-SNF for comparison to its test leakage rate could be extended to other nuclear material systems. The approach for containment analysis of Al-SNF follows calculations for commercial spent fuel as provided in NUREG/CR-6487 that adopts ANSI N14.5 as a methodology for containment analysis. The material-specific features and characteristics of damaged Al-SNF (fuel materials, fabrication techniques, microstructure, radionuclide inventory, and vapor corrosion rates) that were derived from literature sources and/or developed in laboratory testing are applied to generate the four containment source terms that yield four separate cask cavity activity densities; namely, those from fines; gaseous fission product species; volatile fission product species; and fuel assembly crud. The activity values, A 2 , are developed per the guidance of 10CFR71. The analysis is performed parametrically to evaluate maximum number of breached assemblies and exposed fuel area for a proposed shipment in a cask with a test leakage rate

  10. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

    2007-07-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

  11. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

    2007-01-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

  12. The Recovery of Uranium From The Rejected Fuel Plate Dispersion Type of U3O8-Al and U3Si2Al by NaOH

    International Nuclear Information System (INIS)

    Widodo, G; Aji, D

    1998-01-01

    The recovery of uranium from the rejected fuel plate dispersion type of U 3 O 8 -AI And U 3 Si 2 -AI with a dissolution has been performed.Each of 5 fragment of fuel plate dispersion of U 3 O 8 -AI or U 3 Si 2 Al of 1x4 cm size was put in the distilled glass content of 250 ml NaOH solution whit The concentration variation 10,15,20,25,and 30%,and than was heated at temperature of 102 o C and was stirred constantly by magnetic stirred.Uranium in the form of U 3 O 8 or U 3 Si 2 was separated by filtration and Either residu and filtrate was analyzed by potentiometry using modified Devies Gray method. From the experiment data it was found in the residu that presentation of uranium was 83.99-84.05% and 84.67-86.556% while in filtrate it was found 53.90 ppm and 69.3 ppm

  13. Polystyrene Plastic Waste Conversion into Liquid Fuel with Catalytic Cracking Process Using Al2O3 as Catalyst

    Directory of Open Access Journals (Sweden)

    Nurul Kholidah

    2018-01-01

    Full Text Available The increase in energy consumption and an increase in the plastic waste generation are two major problems that arise along with economic growth and the increase in population. Styrofoam is one type of polystyrene plastic waste that can be processed into liquid fuels by cracking process. In this study, the cracking process of polystyrene plastic waste into liquid fuel carried by the catalytic cracking process using Al2O3 as a catalyst. This study aimed to determine the effect of the catalyst weight, length of cracking time and range of temperature in the catalytic cracking process of polystyrene plastic waste into liquid fuel toward the mass and characteristics of liquid fuels produced and to determine the composition of liquid fuels produced. The catalytic cracking process of polystyrene plastic waste with catalyst was done in the fixed bed type reactor by heating the reactor with a heater, where the process took place at temperature of 150°C, 200°C, 250°C and 300°C and the length of the process was varied into 20, 40, and 60 minutes and the catalyst weight was also varied, which were 4%, 6% and 8%, while the styrofoam weight was 250 grams. From the research, the highest mass of liquid fuel derived from polystyrene catalytic cracking process was in the amount of 48.8 grams and liquid yield percentage of 19.5% at temperature of  250°C, cracking time of 60 minutes and weight of 8% catalyst, while the characteristics of liquid fuel that were approaching the characteristics of gasoline was at temperatures of 250°C, cracking time of 60 minutes and weight of 6% catalyst, in which each value of density of 0.763 g/ml, specific gravity of 0.778 and oAPI gravity of 50.2. While other liquid fuels obtained from the cracking of polystyrene were still within the tolerance range characteristic properties of gasoline. Liquid fuels produced from the catalytic cracking process was analyzed using a GC-MS, in which the analysis results indicated that liquid

  14. Status of high-density fuel plate fabrication

    International Nuclear Information System (INIS)

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

    1991-01-01

    Progress has continued on the fabrication of fuel plates with equivalent fuel zone loadings approaching 9 gU/cm 3 . Through hot isostatic pressing (HIP), successful diffusion bonds have been made with 1100 Al and 6061 Al alloys. Although additional study is necessary to optimize the procedure, these bonds demonstrated the most critical processing step for proof-of-concept hardware. Two types of prototype highly loaded fuel plates have been fabricated. The first is a fuel plate in which 0.030-in. (0.76-mm) uranium compound wires are bonded within an aluminum cladding; the second, a dispersion fuel plate with uniform cladding and fuel zone thickness. The successful fabrication of these fuel plates derives from the unique ability of the HIP process to produce diffusion bonds with minimal deformation. (orig.)

  15. Use of pyrolysis gas from coal as reburn fuel. Final report; Einsatz von kohlestaemmigem Pyrolysegas als Reduktionsbrennstoff. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Greul, U.; Magel, C.; Moersch, O.; Ruediger, H.; Storm, C.; Schnell, U.; Spliethoff, H.; Hein, K.R.G.

    1996-12-31

    The research project`s aim was to reduce nitrogen emissions from pulverized-coal furnaces by fuel staging with pyrolysis gas from coal. The test fuels were 6 German and Australian coals. The aim achieved has been the statement that the described method is an adequate means to attain to and remain below emission values of 200 mg/m{sup 3}. The method of fuel staging using coal-original gases was investigated with tests focussing the most important process parameters such as coal type, devolatilization ratio, temperature, residence time, and stoichiometry. The relevant features determined with an entrained flow reactor and with a fluidized-bed reactor were the impact of devolatilization temperatures on carbonized residue and pyrolysis products, the distribution of fuel nitrogen, and the quality of gas and tar, including the respective effects on NO{sub x} formation and reduction in staged combustion. The validation of the mathematical model was done with the experimentally obtained data. The criteria considered fundamental for achieving the NO{sub x} reduction level are temperature, air ratio, and residence time in the reduction zone of the furnace. The pyrolysis tests manifested the strong influence of the coal type and the devolatilization conditions on the composition of the gases and the attainable NO{sub x} reduction. The tars in the pyrolysis gases, with their nitrogen compounds, improve the reducing effect of available nitrogen oxides. By using pyrolysis gases from coal as reburning fuel, NO{sub x} emissions of less than 200 mg/m{sup 3} can be obtained at air ratios around 0.95. (orig./SR) [Deutsch] Das Forschungsprojekt verfolgte das Ziel mit 6 deutschen und australischen Kohlen die Stickoxidemissionen aus Kohlestaubfeuerungen durch Brennstoffstufung mit Pyrolysegas als Reduktionsbrennstoff zu verringern. Das erreichte Ziel war der Nachweis, dass mit dem beschriebenen Verfahren NO{sub x}-Emissionswerte von 200 mg/m{sup 3} erreicht und unterschritten werden

  16. NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE

    Directory of Open Access Journals (Sweden)

    Tukiran Surbakti

    2016-03-01

    Full Text Available Abstract NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE. Research of UMo fuel for research reactor has been developing  right now. The fuel of  research reactor used is uranium low enrichment with high density. For supporting the development of fuel, an assessment of mini fuel in the RSG-GAS core was performed. The mini fuel are U7Mo-Al and U6Zr-Al with densitis of 7.0gU/cc and 5.2 gU/cc, respectively. The size of both fuel are the same namely 630x70.75x1.30 mm were inserted to the 3 plates of dummy fuel. Before being irradiated in the core, a calculation for safety analysis  from neutronics and thermohydrolics aspects were required. However, in this paper will discuss safety analysis of the U7Mo-Al and U6Zr-Al mini fuels from neutronic point of view.  The calculation was done using WIMSD-5B and Batan-3DIFF code. The result showed that both of the mini fuels could be irradiated in the RSG-GAS core with burn up less than 70 % within 12 cycles of operation without over limiting the safety margin. Power density of U7Mo-Al mini fuel bigger than U6Zr-Al fuel.   Key words: mini fuel, neutronics analysis, reactor core, safety analysis   Abstrak ANALISIS NEUTRONIK ELEMEN BAKAR UJI MINI DI TERAS RSG-GAS. Penelitian tentang bahan bakar UMo untuk reaktor riset terus berkembang saat ini. Bahan bakar reaktor riset yang digunakan adalah uranium pengkayaan rendah namun densitas tinggi.  Untuk mendukung pengembangan bahan bakar dilakukan uji elemen bakar mini di teras reakror RSG-GAS dengan tujuan menentukan jumlah siklus di dalam teras sehingga tercapai fraksi bakar maksimum. Bahan bakar yang diuji adalah U7Mo-Al dengan densitas 7,0 gU/cc dan U6Zr-Al densitas 5,2 gU/cc. Ukuran kedua bahan bakar uji tersebut adalah sama 630x70,75x1,30 mm dimasukkan masing masing kedalam 3 pelat dummy bahan bakar. Sebelum diiradiasi ke dalam teras reaktor maka perlu dilakukan perhitungan keselamatan baik secara neutronik maupun termohidrolik. Dalam makalah ini

  17. Jane Austen y C. S. Lewis: de la literatura al cine. Análisis de transposición

    Directory of Open Access Journals (Sweden)

    Mercedes Sarapura Sarapura

    2016-11-01

    Full Text Available Esta es una propuesta de análisis de transposición de la literatura al cine de la novela Orgullo y Prejuicio, de Jane Austen, y del ensayo Una pena observada, de C. S. Lewis. El primer propósito en ambos casos es ahondar en los nuevos aportes de la última versión cinematográfica de la novela de Austen, así como en las características y cualidades de la versión fílmica del cuaderno de anotaciones de C. S. Lewis, a partir de las contribuciones teóricas de autores como Gerard Genette, Georffrey Wagner, Brian Mcfarlane, entre otros. El segundo objetivo se orienta al reconocimiento del intercambio entre el lenguaje cinematográfico y literario como un proceso representativo de la sociedad actual. El hecho de que las adaptaciones (históricamente vistas con gran escepticismo sean ahora continuamente realizadas y acogidas por el público, insta a una revisión y profundización de los estudios que se han hecho al respecto con el fin de conocer la situación actual de la relación existente entre cine y literatura, sus posibilidades creativas, las expectativas de las audiencias, los prejuicios y recelos de los expertos, etc.

  18. LOS APORTES DEL TRABAJO SOCIAL AL TEMA DE FAMILIA

    Directory of Open Access Journals (Sweden)

    Angela María Quintero Velásquez

    2012-08-01

    Full Text Available El artículo presenta las contribuciones del Trabajo Social en los procesos familiares del continente. Refiere modalidades de la acción tales como: economapa. formas alternativas de enfrentar los conflictos socio-familiares, grupos psicoeducativos, redes. Precisa algunos elementos de la intervención: constructivismo, enfoque de convergencia (interinstitucionalidad, intersectorialidad, interdisciplinariedad, cibernética, relaciones ecosistemicas. la familia como sistema social.

  19. Natural gas as fuel - chances and risks; Erdgas als Kraftstoff-Chancen und Risiken.

    Energy Technology Data Exchange (ETDEWEB)

    Wackertapp, H.; Wember, G. [Ruhrgas AG, Essen (Germany)

    2001-11-01

    By reducing the tax payable on natural gas as part of the ecological tax reform, the German government has created the economic framework for the break-through of the automotive fuel, natural gas. Consequently, natural gas has since become much more popular for driving motor engines. The gas industry sees its main task as the development of a natural gas filling station network. There are currently about 200 filling stations with natural gas fuelling facilities in Germany. (orig.) [German] Die Bundesregierung hat mit der in der oekologischen Steuerreform verankerten Mineraloelsteuerbeguenstigung fuer Erdgas die wirtschaftlichen Voraussetzungen fuer einen Durchbruch des neuen Kraftstoffs ''Erdgas'' geschaffen. Demzufolge hat der mobile Einsatz von Erdgas deutlich an Bedeutung gewonnen. Die Gaswirtschaft sieht den Aufbau der Betankungsinfrastruktur als ihre Hauptaufgabe an. Aktuell kann an etwa 200 Tankstellen Erdgas getankt werden. (orig.)

  20. Thermally induced dispersion mechanisms for aluminum-based plate-type fuels under rapid transient energy deposition

    International Nuclear Information System (INIS)

    Georgevich, V.; Taleyarkham, R.P.; Navarro-Valenti, S.; Kim, S.H.

    1995-01-01

    A thermally induced dispersion model was developed to analyze for dispersive potential and determine onset of fuel plate dispersion for Al-based research and test reactor fuels. Effect of rapid energy deposition in a fuel plate was simulated. Several data types for Al-based fuels tested in the Nuclear Safety Research Reactor in Japan and in the Transient Reactor Test in Idaho were reviewed. Analyses of experiments show that onset of fuel dispersion is linked to a sharp rise in predicted strain rate, which futher coincides with onset of Al vaporization. Analysis also shows that Al oxidation and exothermal chemical reaction between the fuel and Al can significantly affect the energy deposition characteristics, and therefore dispersion onset connected with Al vaporization, and affect onset of vaporization

  1. Irradiation performance of uranium-molybdenum alloy dispersion fuels

    International Nuclear Information System (INIS)

    Almeida, Cirila Tacconi de

    2005-01-01

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

  2. 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).

  3. Reproduction of the RA reactor fuel element fabrication

    International Nuclear Information System (INIS)

    Novakovic, M.

    1961-12-01

    This document includes the following nine reports: Final report on task 08/12 - testing the Ra reactor fuel element; design concept for fabrication of RA reactor fuel element; investigation of the microstructure of the Ra reactor fuel element; Final report on task 08/13 producing binary alloys with Al, Mo, Zr, Nb and B additions; fabrication of U-Al alloy; final report on tasks 08/14 and 08/16; final report on task 08/32 diffusion bond between the fuel and the cladding of the Ra reactor fuel element; Final report on task 08/33, fabrication of the RA reactor fuel element cladding; and final report on task 08/36, diffusion of solid state metals [sr

  4. Evaluation of practicability of aluminosilicate additive fuel. Influence of aluminosilicate for reprocessing and corrosion of pellet

    International Nuclear Information System (INIS)

    Matsunaga, Junji; Kashibe, Shinji; Kinoshita, Mika; Ishimoto, Shinji; Harada, Kenichi

    2014-01-01

    Al-Si-O additive fuel is a modified pellet to improve the pellet-cladding interaction (PCI) resistance. This practicability assessment concerns the effect of Al-Si-O addition on the reprocessing and steam corrosion behavior. To address these concerns, a fuel dissolution test in nitric acid and a pellet corrosion test in humidified gas were carried out using the irradiated Al-Si-O additive fuel. Regardless of the Al-Si-O concentration, the dissolution rates of all Al-Si-O additive fuels were faster than that of the standard fuel. The morphology of the insoluble residue obtained from the irradiated Al-Si-O additive fuel could be considered as acceptable for retrieval by the clarification process using a conventional precipitation model. The corrosion resistance of the irradiated Al-Si-O additive fuel to high-temperature (at 1273 K) humidified gas was comparable to or better than that of the standard fuel. The result was interpreted as being due to a large grain size effect by Al-Si-O addition. (author)

  5. Partial oxidation of jet fuels over Rh/Al{sub 2}O{sub 3}. Design and reaction kinetics of sulfur-containing surrogates

    Energy Technology Data Exchange (ETDEWEB)

    Baer, Julian Nicolaas

    2016-07-01

    The conversion of logistic fuels via catalytic partial oxidation (CPOX) on Rh/Al{sub 2}O{sub 3} at short contact times is an efficient method for generating hydrogen-rich synthesis gas. Depending on the inlet conditions, fuel, and catalyst, high syngas yields, low by-product formation, and rates of high fuel conversion can be achieved. CPOX is relevant for mobile hydrogen generation, e.g., on board of airplanes in order to increase the fuel efficiency via fuel cell-based auxiliary power units. Jet fuels contain hundreds of different hydrocarbons and a significant amount of sulfur. The hydrocarbon composition and sulfur content of a jet fuel vary depending on distributor, origin, and refinement of the crude oil. Little is known about the influence of the various compounds on the synthesis-gas yield and the impact of sulfur on the product yield. In this work, the influence of three main chemical compounds of a jet fuel (aromatics, alkanes, and sulfur compounds) on syngas selectivity, the catalyst deactivation process, and reaction sequence is unraveled. As representative components of alkanes and aromatics, n-dodecane and 1,2,4-trimethylbenzene were chosen for ex-situ and in-situ investigations on the CPOX over Rh/Al{sub 2}O{sub 3}, respectively. Additionally, for a fixed paraffin-to-aromatics ratio, benzothiophene or dibenzothiophene were added as a sulfur component in three different concentrations. The knowledge gained about the catalytic partial oxidation of jet fuels and their surrogates is used to identify requirements for jet fuels in mobile applications based on CPOX and to optimize the overall system efficiency. The results show an influence of the surrogate composition on syngas selectivity. The tendency for syngas formation increases with higher paraffin contents. A growing tendency for by-product formation can be observed with increasing aromatics contents in the fuel. The impact of sulfur on the reaction system shows an immediate change in the product

  6. 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.

  7. 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

  8. Multi-fuel driven Janus micromotors.

    Science.gov (United States)

    Gao, Wei; D'Agostino, Mattia; Garcia-Gradilla, Victor; Orozco, Jahir; Wang, Joseph

    2013-02-11

    Here the first example of a chemically powered micromotor that harvests its energy from the reactions of three different fuels is presented. The new Al/Pd Janus microspheres-prepared by depositing a Pd layer on one side of Al microparticles-are propelled efficiently by the thrust of hydrogen bubbles generated from different reactions of Al in strong acidic and alkaline environments, and by an oxygen bubble thrust produced at their partial Pd coating in hydrogen peroxide media. High speeds and long lifetimes of 200 μm s(-1) and 8 min are achieved in strong alkaline media and acidic media, respectively. The ability to autonomously adapt to the presence of a new fuel (surrounding environment), without compromising the propulsion behavior is illustrated. These data also represent the first example of a chemically powered micromotor that propels autonomously and efficiently in alkaline environments (pH > 11) without additional fuels. The ability to use multiple fuel sources to power the same micromotor offers a broader scope of operation and considerable promise for diverse applications of micromotors in different chemical environments. Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  9. Fuel cells applied to transport; Celdas de combustible aplicadas al transporte

    Energy Technology Data Exchange (ETDEWEB)

    Vargas Gonzalez, Miguel [Comision Nacional para el Ahorro de Energia, Mexico, D.F. (Mexico)

    2001-07-01

    The future demand of energy as well as the preservation of the environment have generated, in several countries, the necessity of developing alternative means for the most efficient transformation of energy that causes minimum damages to the environment. The fuel cells technology is outlined as one of the alternating means to the traditional forms of transforming the energy for residential use as well as for the automotive vehicles. At present it is in a final stage of demonstration, reason why as of year 2003 the possibility will exist on automobiles and trucks circulating normally. [Spanish] La futura demanda de energeticos asi como la preservacion del medio ambiente ha generado, en varios paises, la necesidad de desarrollar medios alternos para la transformacion mas eficiente de la energia que cause minimos danos al ambiente. La tecnologia de celdas de combustible se perfila como uno de los medios alternos a la forma tradicional de transformar la energia tanto para uso residencial como para los vehiculos automotores. En la actualidad se encuentra en una etapa final de demostracion por lo que posiblemente a partir del ano 2003 se cuente con automoviles y camiones circulando normalmente.

  10. Dissolution Flowsheet for High Flux Isotope Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Karay, N. S [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy, and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas, allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

  11. 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.

  12. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-15

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

  14. Dispersion fuel for nuclear research facilities

    International Nuclear Information System (INIS)

    Kushtym, A.V.; Belash, M.M.; Zigunov, V.V.; Slabospitska, O.O.; Zuyok, V.A.

    2017-01-01

    Designs and process flow sheets for production of nuclear fuel rod elements and assemblies TVS-XD with dispersion composition UO_2+Al are presented. The results of fuel rod thermal calculation applied to Kharkiv subcritical assembly and Kyiv research reactor VVR-M, comparative characteristics of these fuel elements, the results of metallographic analyses and corrosion tests of fuel pellets are given in this paper

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  16. Study of fuel element characteristic of SM and SMP (SM-PRIMA) fuel assemblies

    International Nuclear Information System (INIS)

    Klinov, A.V.; Kuprienko, V.A.; Lebedev, V.A.; Makhin, V.M.; Tuchnin, L.M.; Tsykanov, V.A.

    1999-01-01

    The paper discusses the techniques and results of reactor tests and post-reactor investigations of the SM reactor fuel elements and fuel elements developed in the process of designing the specialized PRIMA test reactor with the SM reactor fuel elements used as a prototype and which are referred to as the SMP fuel elements. The behavior of fuel elements under normal operating conditions and under deviation from normal operating conditions was studied to verify the calculation techniques, to check the calculation results during preparation of the SM reactor safety substantiation report and to estimate the possibility of using such fuel elements in other projects. During tests of fuel rods under deviation from normal operating conditions their advantages were shown over fuel elements, the components of which were produced using the Al-based alloys. (author)

  17. Fuel rich and fuel lean catalytic combustion of the stabilized confined turbulent gaseous diffusion flames over noble metal disc burners

    Directory of Open Access Journals (Sweden)

    Amal S. Zakhary

    2014-03-01

    Full Text Available Catalytic combustion of stabilized confined turbulent gaseous diffusion flames using Pt/Al2O3 and Pd/Al2O3 disc burners situated in the combustion domain under both fuel-rich and fuel-lean conditions was experimentally studied. Commercial LPG fuel having an average composition of: 23% propane, 76% butane, and 1% pentane was used. The thermal structure of these catalytic flames developed over Pt/Al2O3 and Pd/Al2O3 burners were examined via measuring the mean temperature distribution in the radial direction at different axial locations along the flames. Under-fuel-rich condition the flames operated over Pt catalytic disc attained high temperature values in order to express the progress of combustion and were found to achieve higher activity as compared to the flames developed over Pd catalytic disc. These two types of catalytic flames demonstrated an increase in the reaction rate with the downstream axial distance and hence, an increase in the flame temperatures was associated with partial oxidation towards CO due to the lack of oxygen. However, under fuel-lean conditions the catalytic flame over Pd catalyst recorded comparatively higher temperatures within the flame core in the near region of the main reaction zone than over Pt disc burner. These two catalytic flames over Pt and Pd disc burners showed complete oxidation to CO2 since the catalytic surface is covered by more rich oxygen under the fuel-lean condition.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  19. Properties and efficiency of a Pt/Al2O3 catalyst applied in a solid fuel thermo-accumulating furnace

    Directory of Open Access Journals (Sweden)

    SRDJAN BELOSEVIC

    2007-08-01

    Full Text Available A prototype of a solid fuel thermo-accumulating furnace has been developed. In order to achieve a higher combustion efficiency, a Pt/Al2O3 catalyst in the form of 3 ± 0.3 mm spheres was applied, which enabled further combustion of flue gases within the furnace. Experimental investigation of the influence of the catalyst on the conversion of CO has been done for different operation regimes and positions of the catalyst. Paper presents selected results regarding CO emission during wood and coal combustion. Investigations suggest a considerable effect of the catalyst and a strong influence of the catalyst position to CO emission reduction. The microstructure of the catalyst beads, characterized by selective chemisorption of CO, has shown the decrease of the number of Pt sites as a consequence of blockage by coke deposits formed during the combustion of solid fuel.

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

    Science.gov (United States)

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

    2017-10-01

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

  1. Deformation behavior of laser welds in high temperature oxidation resistant Fe-Cr-Al alloys for fuel cladding applications

    Science.gov (United States)

    Field, Kevin G.; Gussev, Maxim N.; Yamamoto, Yukinori; Snead, Lance L.

    2014-11-01

    Ferritic-structured Fe-Cr-Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability and post-weld mechanical behavior of three model alloys in a range of Fe-(13-17.5)Cr-(3-4.4)Al (wt.%) with a minor addition of yttrium using modern laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds using sub-sized, flat dog-bone tensile specimens and digital image correlation (DIC) has been carried out to determine the performance of welds as a function of alloy composition. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. For all proposed alloys, laser welding resulted in a defect free weld devoid of cracking or inclusions.

  2. Deformation behavior of laser welds in high temperature oxidation resistant Fe–Cr–Al alloys for fuel cladding applications

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G., E-mail: fieldkg@ornl.gov; Gussev, Maxim N., E-mail: gussevmn@ornl.gov; Yamamoto, Yukinori, E-mail: yamamotoy@ornl.gov; Snead, Lance L., E-mail: sneadll@ornl.gov

    2014-11-15

    Ferritic-structured Fe–Cr–Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability and post-weld mechanical behavior of three model alloys in a range of Fe–(13–17.5)Cr–(3–4.4)Al (wt.%) with a minor addition of yttrium using modern laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds using sub-sized, flat dog-bone tensile specimens and digital image correlation (DIC) has been carried out to determine the performance of welds as a function of alloy composition. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. For all proposed alloys, laser welding resulted in a defect free weld devoid of cracking or inclusions.

  3. Las políticas sociales de dependencia en España: contribuciones y consecuencias para los ancianos y sus cuidadores informales

    Directory of Open Access Journals (Sweden)

    Bibiana Escuredo

    2007-01-01

    Full Text Available Se presenta la evolución de la protección social de la dependencia en España desde la década de los noventa hasta el momento actual y se plantean las consecuencias de la falta de acuerdo político respecto a las medidas a implementar. Se analizan los elementos en los que se basan las políticas de dependencia, resaltando sus aspectos positivos y negativos, así como las consecuencias para los ancianos y sus cuidadores informales. También se pone de manifiesto cómo, ante la posibilidad de quiebra del papel de la familia en el cuidado de los dependientes, se están implementando políticas que tienen como objetivo reforzar el rol cuidador de la familia y que se enmarcan dentro de la denominada "Conciliación de la Vida Familiar y Laboral". Estas políticas, más allá de sus contribuciones positivas, implican también importantes costes personales y de oportunidad para los cuidadores informales.

  4. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    International Nuclear Information System (INIS)

    Gamble, Kyle Allan Lawrence; Hales, Jason Dean; Barani, Tommaso; Pizzocri, Davide; Pastore, Giovanni

    2016-01-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

  5. Modelling Accident Tolerant Fuel Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Hales, Jason Dean [Idaho National Laboratory; Gamble, Kyle Allan Lawrence [Idaho National Laboratory

    2016-05-01

    The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. The United States Department of Energy (DOE) through its Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is a three-year project to perform research on two accident tolerant concepts. The final outcome of the ATF HIP will be an in-depth report to the DOE Advanced Fuels Campaign (AFC) giving a recommendation on whether either of the two concepts should be included in their lead test assembly scheduled for placement into a commercial reactor in 2022. The two ATF concepts under investigation in the HIP are uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (Idaho National Laboratory, Los Alamos National Laboratory, and Argonne National Laboratory), a comprehensive multiscale approach to modeling is being used that includes atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. Model development and fuel performance analysis are critical since a full suite of experimental studies will not be complete before AFC must prioritize concepts for focused development. In this paper, we present simulations of the two proposed accident tolerance fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. Sensitivity analyses are completed using Sandia National Laboratories’ Dakota software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). We also outline the multiscale modelling approach being employed. Considerable additional work is required prior to preparing the recommendation report for the Advanced

  6. Creep Analysis of Aluminum-Based Spent Nuclear Fuel in Repository Storage

    International Nuclear Information System (INIS)

    Gong, C.; Lam, P.S.; Sindelar, R.L.

    1998-07-01

    Aluminum-clad, aluminum-based spent nuclear fuels (Al SNF) from foreign and domestic research reactors are being consolidated at the Savannah River Site (SRS). These fuels are planned to be put into dry storage followed by disposal in the federal repository. Temperature conditions in storage and disposal systems due to nuclear decay heat sources will promote creep information of the fuel elements. Excessive deformation of the Al SNF will cause gross distortion (slump) of the fuels and may cause gross cladding rupture

  7. 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

  8. Radionuclide release from research reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Curtius, H., E-mail: h.curtius@fz-juelich.de [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany); Kaiser, G.; Mueller, E.; Bosbach, D. [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany)

    2011-09-01

    Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO{sub 2} fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in {sup 235}U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO{sub 2}-fuel (LWR fuel, enrichment in {sup 235}U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Juelich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl{sub 2}-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAl{sub x}-Al and U{sub 3}Si{sub 2}-Al) was studied in 400 mL MgCl{sub 2}-rich salt brine in the presence of Fe{sup 2+} under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH){sub 3}(s) and Eu(OH){sub 3}(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu

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

    International Nuclear Information System (INIS)

    Moreira Castro, Martin Ignacio

    2015-01-01

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

  10. 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

  11. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Kawada, Toshiyuki; Hirayama, Satoshi; Yoneya, Katsutoshi.

    1980-01-01

    Purpose: To enable load-depending operation as well as moderation for the restriction of operation conditions in the present nuclear reactors, by specifying the essential ingredients and the total weight of the additives to UO 2 fuel substances. Constitution: Two or more additives selected from Al 2 O 3 , B 2 O, CaO, MgO, SiO 2 , Na 2 O and P 2 O 5 are added by the total weight of 2 - 5% to fuel substances consisting of UO 2 or a mixture of UO 2 and PuO 2 . When the mixture is sintered, the strength of the fuel elements is decreased and the fuel-cladding interactions due to the difference in the heat expansion coefficients between the ceramic fuel elements and the metal claddings are decreased to a substantially harmless degree. (Horiuchi, T.)

  12. Contribuciones al estudio de la Parasitología en Colombia

    Directory of Open Access Journals (Sweden)

    Uribe-Piedrahita César

    1947-12-01

    Full Text Available 1- Myxobolidae parasito de la vesícula biliar de una Rana de los Llanos Orientales (Resumen: Se observó un Myxobolidae, Myxobolus pfeitteri, parásito de la vesícula biliar de Rana palmipes, habitante de los Llanos orientales de Colombia. Se destaca su papel patógeno.  Creemos que es el primer Myxobolidae descrito en Colombia. / 2- Observaciones sobre un Trichomonas sp. (Resumen: Aechmea bromeliaefolia / Greigia Danielii / Guzmania costaricensis / Guzmania diffusa / Guzmania glomerata / Guzmania minor Mez var / Flammea / Pitcairnia Bakeri / Pitcairnia Barrigae / Pitcairnia costata / Pitcairnia diffusa / Pitcairnia excerta / Pitcairnia lepidopetalon / Pitcairnia squarrosa / Pitcairnia squarrosa var. Colorata / Tillandsia hospitalis / Tillandsia pulchella / Vriesia Hodgei.

  13. Thermal conductivity of fresh and irradiated U-Mo fuels

    Science.gov (United States)

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

    2018-05-01

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

  14. The Caramel fuel in OSIRIS

    International Nuclear Information System (INIS)

    Cherruau, Francois.

    1980-11-01

    This paper presents the main characteristics of the caramel fuel, a description of OSIRIS transformations that were decided in line with its conversion and the results of its operation since then. The Caramel fuel is made from sintered UO 2 pellets contained in zircaloy clads forming the plates of the fuel assembly reducing the enrichment need to as little as 3 to 10% instead of 93% enriched U/Al in the previous fuel. The first year of experience shows the capacity under a statistic scale of the caramel fuel to fulfil the most severe operation requirements for use in low and medium power research reactors

  15. Sipping test on a failed MTR fuel element

    International Nuclear Information System (INIS)

    Terremoto, Luis Antonio Albiac; Zeituni, Carlos Alberto; Silva, Antonio Teixeira e; Perrotta, Jose Augusto; Silva, Jose Eduardo Rosa da

    2002-01-01

    This work describes sipping tests performed on MTR fuel elements of the IEA-R1 research reactor, in order to determinate which one failed in the core during a routine operation of the reactor. radioactive iodine isotopes 131 I and 133 I, employed as failure indicators, were detected in samples corresponding to the fuel element IEA-156. The specific activity of each sample, as well as the average leaking rate, were measured for 137 Cs. The nuclear fuels U 3 O 8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137 Cs. (author)

  16. 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

  17. Study of the aqueous corrosion mechanisms and kinetics of the AlFeNi aluminium based alloy used for the fuel cladding in the Jules Horowitz research reactor; Etude des mecanismes et des cinetiques de corrosion aqueuse de l'alliage d'aluminium AlFeNi utilise comme gainage du combustible nucleaire de reacteurs experimentaux

    Energy Technology Data Exchange (ETDEWEB)

    Wintergerst, M.

    2009-05-15

    For the Jules Horowitz new material-testing reactor (JHR), an aluminium base alloy, called AlFeNi, will be used for the cladding of the fuel plates. This alloy (Al - 1% Fe - 1% Ni - 1 % Mg), which is already used as fuel cladding, was developed for its good corrosion resistance in water at high temperatures. However, few studies dealing with the alteration process in water and the relationships with irradiation effects have been performed on this alloy. The conception of the JHR fuel requires a better knowledge of the corrosion mechanisms. Corrosion tests were performed in autoclaves at 70 C, 165 C and 250 C on AlFeNi plates representative of the fuel cladding. Several techniques were used to characterize the corrosion scale: SEM, TEM, EPMA, XRD, Raman spectroscopy. Our observations show that the corrosion scale is made of two main layers: a dense amorphous scale close to the metal and a porous crystalline scale in contact with the water. More than the morphology, the chemical compositions of both layers are different. This duplex structure results from a mixed growth mechanism: an anionic growth to develop the inner oxide and a cationic diffusion followed by a dissolution-precipitation process to form the outer one. Dynamic experiments at 70 C and corrosion kinetics measurements have demonstrated that the oxide growth process is controlled by a diffusion step associated to a dissolution/precipitation process. A corrosion mechanism of the AlFeNi alloy in aqueous media has been proposed. Then post-irradiation exams performed on irradiated fuel plates were used to investigate the effects of the irradiation on the corrosion behaviour in the reactor core. (author)

  18. Fuels for research and test reactors, status review: July 1982

    International Nuclear Information System (INIS)

    Stahl, D.

    1982-12-01

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO 2 rod fuels. Among new fuels, those given major emphasis include H 3 Si-Al dispersion and UO 2 caramel plate fuels

  19. Sorption of selenium on Mg-Al and Mg-Al-Eu layered double hydroxides

    International Nuclear Information System (INIS)

    Curtius, H.; Paparigas, Z.; Kaiser, G.

    2008-01-01

    Salt domes represent deep geological formations which are under consideration as final repositories for irradiated research reactor fuel elements. For long-term safety aspects the mobilisation of the radionuclides due to a water ingress is intensively investigated. At the Institute of Energy Research (IEF-6), leaching experiments were performed in a hot cell facility with UAl x -Al and U 3 Si 2 -Al dispersed research reactor fuel elements in repository-relevant MgCl 2 -rich salt brines under anaerobic conditions. The fuel plates corroded completely within one year and a Mg-Al-layered double hydroxide (LDH) with chloride as interlayer anion was identified as one crystalline phase component of the corrosion products (secondary phases). This Mg-Al-LDH was synthesized, characterized, and the ability to retard europium by an incorporation process was investigated. Europium, as a representative for lanthanides, was identified to be one of the radionuclides which were found in the corrosion products. We could show that europium was incorporated in the lattice structure. LDHs have high anion exchange capacities that enhance their potential to remove anionic contaminants from aqueous systems. In this work the sorption behaviour of selenium in the chemical form as selenite (SeO 3 2- ) on Mg-Al-LDH and on Mg-Al-Eu-LDH was investigated. Especially the influence of the larger europium-III ion was of interest. It represents in the Mg-Al-Eu-LDH about 10% of the molar aluminium amount. The sorption has been experimentally studied in a wide range of pH, ionic strength, radionuclide and sorbent concentration. Both LDHs with chloride as interlayer anion were synthesized by a coprecipitation method under controlled conditions, and their main physico-chemical properties were analyzed prior to the sorption experiments. The sorption kinetics of selenite on the LDHs in water and in MgCl 2 -rich brine were rapid using a LDH concentration of 10 g/L. Equilibrium, indicated by stable p

  20. Corrosion of research reactor Al-clad spent fuel in water

    International Nuclear Information System (INIS)

    Bendereskaya, O.S.; De, P.K.; Haddad, R.; Howell, J.P.; Johnson, A.B. Jr.; Laoharojanaphand, S.; Luo, S.; Ramanathan, L.V.; Ritchie, I.; Hussain, N.; Vidowsky, I.; Yakovlev, V.

    2002-01-01

    A significant amount of aluminium-clad spent nuclear fuel from research and test reactors worldwide is currently being stored in water-filled basins while awaiting final disposition. As a result of corrosion issues, which developed from the long-term wet storage of aluminium-clad fuel, the International Atomic Energy Agency (IAEA) implemented a Co-ordinated Research Project (CRP) in 1996 on the 'Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water'. The investigations undertaken during the CRP involved ten institutes in nine different countries. The IAEA furnished corrosion surveillance racks with aluminium alloys generally used in the manufacture of the nuclear fuel cladding. The individual countries supplemented these racks with additional racks and coupons specific to materials in their storage basins. The racks were immersed in late 1996 in the storage basins with a wide range of water parameters, and the corrosion was monitored at periodic intervals. Results of these early observations were reported after 18 months at the second research co-ordination meeting (RCM) in Sao Paulo, Brazil. Pitting and crevice corrosion were the main forms of corrosion observed. Corrosion caused by deposition of iron and other particles on the coupon surfaces was also observed. Galvanic corrosion of stainless steel/aluminium coupled coupons and pitting corrosion caused by particle deposition was observed. Additional corrosion racks were provided to the CRP participants at the second RCM and were immersed in the individual basins by mid-1998. As in the first set of tests, water quality proved to be the key factor in controlling corrosion. The results from the second set of tests were presented at the third and final RCM held in Bangkok, Thailand in October 2000. An IAEA document giving details about this CRP and other guidelines for spent fuel storage is in pres. This paper presents some details about the CRP and the basis for its extension. (author)

  1. Argentine activities related to the development of low enriched fuel elements

    International Nuclear Information System (INIS)

    Giorsetti, Domingo R.; Perez, Edmundo E.

    1983-01-01

    Within the framework of the RERTR Program and supported by the technical cooperation work agreed upon between the U.S.A. and Argentina in May 1979, the CNEA Nuclear Fuel Department - Low Enriched Fuel Elements Project (ECBE Project), has carried on its own program for developing fuels with low enrichment for research and test reactors. Up to the present, its main objective has been to replace the highly enriched fuel used in its only reactor (RA-3) for research, development and radioisotopes production. The basic stages of the Argentine Program are shown in Table 1. At a meeting held in Vienna in March, 1980, the CNEA stated that its development of fuels with low enrichment would be in two fuel lines: UAl x -Al and U 3 O 8 -Al, and that its aim would be to reach uranium densities of 18-2.2 g/cm 3 for the UAI x -Al line and 2.4-3.0 g/cm 3 for the U 3 O 8 line. At the international meeting held at ANL in November, 1980, and after having received depleted uranium and uranium with 20% and 45% enrichment (purchased from the U.S.A. for manufacturing miniplates and possible standard fuels) to carry on the proposed development, CNEA anticipated -- after its first tests -- that the conditions were satisfactory for reaching uranium densities of 2.4-3.0 g/cm 3 in U 3 O 8 -Al fuel and of 2.4 g/cm 3 in UAI x -Al fuel. In February 1981, after Argentina accepted the obligation of paying for the irradiation service, authorization was obtained for irradiating miniplates in the Oak Ridge Reactor within the RERTR Program. In June 1981, the first set of miniplates was sent to Oak Ridge National Laboratory (ORNL). The maximum actual densities reached at that time were 3.12 g/cm 3 with U 3 O 8 -Al and 2.52 g/cm 3 with UAl x -Al. During a visit of the CNEA Project Technical Manager to the Argonne National Laboratory (ANL) in July 1981, and after exchanging ideas with ANL professional staff, the CNEA decided to incorporate a new line of development, that of U 3 Si-Al. Three months later

  2. Evidence of amorphous interdiffusion layer in heavy ion irradiated U–8wt%Mo/Al interfaces

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, H-Y. [Forschungsneutronenquelle Heinz Maier-Leibniz (FRM II), Technische Universität München Lichtenbergstr. 1, D-85747 Garching (Germany); Zweifel, T. [Forschungsneutronenquelle Heinz Maier-Leibniz (FRM II), Technische Universität München Lichtenbergstr. 1, D-85747 Garching (Germany); CEA, DEN, DEC, F-13108 St. Paul Lez Durance Cedex (France); Palancher, H., E-mail: herve.palancher@cea.fr [CEA, DEN, DEC, F-13108 St. Paul Lez Durance Cedex (France); Bonnin, A. [ESRF, 6 rue Jules Horowitz, 38042 Grenoble (France); Beck, L. [Tandembeschleuniger des Maier-Leibnitz-Labors (MLL), Am Coulombwall 6, D-85747 Garching (Germany); Weiser, P. [Walther Schottky Institut, Technische Universität München, Am Coulombwall 4, D-85747 Garching (Germany); Döblinger, M. [Department Chemie, Ludwig-Maximilians-Universität München (LMU), Butenandstr. 11, D-81377 München (Germany); Sabathier, C. [CEA, DEN, DEC, F-13108 St. Paul Lez Durance Cedex (France); Jungwirth, R.; Petry, W. [Forschungsneutronenquelle Heinz Maier-Leibniz (FRM II), Technische Universität München Lichtenbergstr. 1, D-85747 Garching (Germany)

    2013-09-15

    U–Mo/Al based nuclear fuels are worldwide considered as the most promising high density fuel for the conversion of high flux research and test reactors from highly enriched uranium to lower enrichment. However in-pile growth of an amorphous interdiffusion layer at the U–Mo/Al interfaces strongly limits the performances of this fuel. Several in-pile tests have been performed to optimize the composition. In this paper, a breakthrough in simulating the U–8wt%Mo/Al behavior under out-of-pile irradiation is reported. It is shown that an amorphous U–8wt%Mo/Al interdiffusion layer (IDL) is obtained by heavy ion irradiation ({sup 127}I) in a U–Mo/Al diffusion couple under controlled temperature conditions. The properties of this IDL coincide with the results obtained from in-pile tests. This methodological work clearly indicates that heavy ion irradiations could be routinely applied for optimizing composition of U–Mo/Al nuclear fuels. In other words these out-of-pile tests using ion beams could become a representative, efficient and economic step before in-pile irradiation.

  3. Evidence of amorphous interdiffusion layer in heavy ion irradiated U–8wt%Mo/Al interfaces

    International Nuclear Information System (INIS)

    Chiang, H-Y.; Zweifel, T.; Palancher, H.; Bonnin, A.; Beck, L.; Weiser, P.; Döblinger, M.; Sabathier, C.; Jungwirth, R.; Petry, W.

    2013-01-01

    U–Mo/Al based nuclear fuels are worldwide considered as the most promising high density fuel for the conversion of high flux research and test reactors from highly enriched uranium to lower enrichment. However in-pile growth of an amorphous interdiffusion layer at the U–Mo/Al interfaces strongly limits the performances of this fuel. Several in-pile tests have been performed to optimize the composition. In this paper, a breakthrough in simulating the U–8wt%Mo/Al behavior under out-of-pile irradiation is reported. It is shown that an amorphous U–8wt%Mo/Al interdiffusion layer (IDL) is obtained by heavy ion irradiation ( 127 I) in a U–Mo/Al diffusion couple under controlled temperature conditions. The properties of this IDL coincide with the results obtained from in-pile tests. This methodological work clearly indicates that heavy ion irradiations could be routinely applied for optimizing composition of U–Mo/Al nuclear fuels. In other words these out-of-pile tests using ion beams could become a representative, efficient and economic step before in-pile irradiation

  4. The VALMONT Program: Accurate experimental techniques to support the neutronics qualification of UMo/AL

    International Nuclear Information System (INIS)

    Hedelot, J.P.; Doederlein, C.; Antony, M.; Girard, J.M.; Laval, V.; Fougeras, P.; Willermoz, G.; Leconte, P.

    2005-01-01

    The VALMONT program aimed at qualifying the HORUS3D (HOrowitz Reactor simulation Unified System) neutronics calculation route that is used for the development of the JHR core, and to verify the correct treatment of UMo/Al (20% enrichment in 235 U) fuel. The program is composed of two parts. The first part was devoted to the measurement by the oscillation technique of the reactivity effect of UAl/Mo fuel with an accuracy around 1% (1□). The second part consisted of gamma-spectroscopy experiments on a dedicated UMo/Al fuel sample in order to characterize, through axial power profiles and modified conversion ratio of 238 U measurements, the production and absorption effects inside the UAl/Mo fuel. The overall excellent agreement between high accuracy experiments and calculations allowed to qualify the HORUS3D neutronics calculation route for UMo/Al fuel. (author)

  5. Alternative Fuels in Epilepsy and Amyotrophic Lateral Sclerosis.

    Science.gov (United States)

    Tefera, Tesfaye W; Tan, Kah Ni; McDonald, Tanya S; Borges, Karin

    2017-06-01

    This review summarises the recent findings on metabolic treatments for epilepsy and Amyotrophic Lateral Sclerosis (ALS) in honour of Professor Ursula Sonnewald. The metabolic impairments in rodent models of these disorders as well as affected patients are being discussed. In both epilepsy and ALS, there are defects in glucose uptake and reduced tricarboxylic acid (TCA) cycling, at least in part due to reduced amounts of C4 TCA cycle intermediates. In addition there are impairments in glycolysis in ALS. A reduction in glucose uptake can be addressed by providing the brain with alternative fuels, such as ketones or medium-chain triglycerides. As anaplerotic fuels, such as the triglyceride of heptanoate, triheptanoin, refill the TCA cycle C4/C5 intermediate pool that is deficient, they are ideal to boost TCA cycling and thus the oxidative metabolism of all fuels.

  6. Evolution of dispersion fuel meat structure caused by interface reaction

    International Nuclear Information System (INIS)

    Xing Zhonghu; Ying Shihao

    2000-01-01

    In reactor operation, the resultant layers are formed by interdiffusion at the fuel particle-matrix interfaces of U 3 Si 2 -Al dispersion fuel. This results in the evolution of meat structure. On the basis of Monte-Carlo method, the author developed simulation method of fuel meat, and simulated the stochastic space locations of spherical fuel particles in the meat. The fuel volume fraction is 43%, and the particles are in definite size distribution. For the 13551 simulated particle samples, the evolution of meat structure is calculated with layer thickness ranging from 0 to 16 μm. The parameters of meat structure include the U 3 Si 2 fuel volume fraction, resultant layer volume fraction, Al matrix volume fraction, particle contact probability and overlap degree as functions of layer thickness

  7. Corrosion behavior of Fe-Si metallic coatings added with NiCrAlY in an environment of fuel oil ashes at 700 C

    Energy Technology Data Exchange (ETDEWEB)

    Salinas-Bravo, V.M.; Porcayo-Calderon, J.; Romero-Castanon, T. [Instituto de Investigaciones Electricas, Gerencia de Procesos Termicos., Av. Reforma 113, C.P. 62490 Col. Palmira. Temixco. Morelos (Mexico); Dominguez-Patino, G.; Gonzalez-Rodriguez, J.G. [U.A.E.M. Centro de Investigaciones en Ingenieria y Ciencias Aplicadas., Av. Universidad 1001, C.P. 62210, Col. Chamilpa. Cuernavaca, Morelos (Mexico)

    2005-07-01

    Electrochemical potentiodynamic polarization curves and immersion tests for 300 h at 700 C in a furnace have been used to evaluate the corrosion resistance of Fe-Si metallic coatings added with up to 50 wt.% of NiCrAIY. The corrosive environment was fuel oil ashes from a steam generator. The composition of fuel oil ashes includes high content of vanadium, sodium and sulfur. The results obtained show that only the addition of 20 wt.% NiCrAlY to the Fe-Si coating improves its corrosion resistance. The behavior of all tested coatings is explained by the results obtained from the analysis of every coating using electron microscopy and energy dispersive X-ray analysis. (Abstract Copyright [2005], Wiley Periodicals, Inc.)

  8. Biomethane as a fuel. Quota transfer; Biomethan als Kraftstoff. Quotenuebertragung

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-15

    Biomethane from biomass replaces fossil natural gas. Just as natural gas, it can be easily used in the transport sector. The advantages of biomethane as a fuel are numerous: high contribution to climate protection, technically acceptable admixture, price advantage over gasoline, eligibility to the biofuel quota. This brochure provides information on: (a) Role of biomethane to meet the biofuel targets of the Federal Government; (b) Eligibility of biomethane to the biofuel quota; (c) Business models for marketing of biomethane as a fuel; (d) Benefits of bio-methane for various actors.

  9. Development of technology of high density LEU dispersion fuel fabrication

    International Nuclear Information System (INIS)

    Wiencek, T.; Totev, T.

    2007-01-01

    Advanced Materials Fabrication Facilities at Argonne National Laboratory have been involved in development of LEU dispersion fuel for research and test reactors from the beginning of RERTR program. This paper presents development of technology of high density LEU dispersion fuel fabrication for full size plate type fuel elements. A brief description of Advanced Materials Fabrication Facilities where development of the technology was carried out is given. A flow diagram of the manufacturing process is presented. U-Mo powder was manufactured by the rotating electrode process. The atomization produced a U-Mo alloy powder with a relatively uniform size distribution and a nearly spherical shape. Test plates were fabricated using tungsten and depleted U-7 wt.% Mo alloy, 4043 Al and Al-2 wt% Si matrices with Al 6061 aluminum alloy for the cladding. During the development of the technology of manufacturing of full size high density LEU dispersion fuel plates special attention was paid to meet the required homogeneity, bonding, dimensions, fuel out of zone and other mechanical characteristics of the plates.

  10. Reduced enrichment fuel and its reactivity effects in the University Training Reactor Moata

    International Nuclear Information System (INIS)

    Wilson, D.J.

    1983-08-01

    Concern for nuclear proliferation is likely to preclude future supply of highly enriched uranium fuel for research reactors such as the University Training Reactor Moata. This study calculates the fuel densities necessary to maintain the reactivity per plate of the present high enrichment (90 per cent 235 U) fuel for a range of lower enrichments assuming that no geometry changes are allowed. The maximum uranium density for commercially available aluminium-type research reactor fuels is generally considered to be about 1.7 g cm -3 . With this density limitation, the minimum enrichment to maintain present reactivity per plate is about 35 per cent 235 U. For low enrichment (max. 20 per cent 235 U) fuel, the required U density is about 2.9 g cm -3 , which is beyond the expected range for UAl/sub x/-Al but within that projected for the longer term development and full qualification for U 3 O 8 -Al. Medium enrichment (nominally 45 per cent 235 U) Al/sub x/-Al would be entirely satisfactory as an immediate replacement fuel, requiring no modifications to the reactor and operating procedures, and minimal reappraisal of safety issues. Included in this study are calculations of the fuel coefficients at various enrichments, the effect of replacing standard fuel plates or complete elements with 45 per cent enriched fuel, and the reactivity to be gained by replacing 12-plate with 13-plate elements

  11. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    International Nuclear Information System (INIS)

    Souza, J.A.B.; Durazzo, M.

    2010-01-01

    IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 gU/cm 3 by using the U 3 Si 2 -Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 gU/cm 3 for the U 3 Si 2 -Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian-Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

  12. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Jose Antonio Batista de; Durazzo, Michelangelo, E-mail: jasouza@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 g U/c m3 by using the U{sub 3}Si{sub 2}-Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 g U/c m3 for the U{sub 3}Si{sub 2}-Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian- Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

  13. Tribological Properties of Ti(Al,O)/Al2O3 Composite Coating by Thermal Spraying

    Science.gov (United States)

    Salman, Asma; Gabbitas, Brian; Cao, Peng; Zhang, Deliang

    The use of thermal spray coatings provides protection to the surfaces operating in severe environments. The main goal of the current work is to investigate the possibility of using a high velocity air fuel (HVAF) thermally sprayed wear resistant Ti(Al,O)/Al2O3 coating on tool steel (H13) which is used for making dies for aluminium high pressure die casting and dummy blocks aluminium extrusion. A feedstock of Ti(Al,O)/Al2O3 composite powder was produced from a mixture of Al and TiO2 powders by high energy mechanical milling, followed by a thermal reaction process. The feedstock was then thermally sprayed using a high velocity air-fuel (HVAF) technique onto H13 steel substrates to produce a composite coating. The present study describes and compares the tribological properties such as friction and sliding wear rate of the coating both at room and high temperature (700°C). The wear resistance of the coating was investigated by a tribometer using a spherical ended alumina pin as a counter body under dry and lubricating conditions. The results showed that composite coating has lower wear rate at high temperature than at room temperature without using lubricant. The composite coating was characterized using scanning electron microscopy (SEM), optical microscopy and X-ray diffractometry (XRD). This paper reports the experimental observations and discusses the wear resistance performance of the coatings at room and high temperatures.

  14. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    Feraday, M.A.

    1993-01-01

    This paper includes some statements and remarks concerning the uranium silicide fuels for which there is significant fabrication in AECL, irradiation and defect performance experience; description of two Canadian high flux research reactors which use high enrichment uranium (HEU) and the fuels currently used in these reactors; limited fabrication work done on Al-U alloys to uranium contents as high as 40 wt%. The latter concerns work aimed at AECL fast neutron program. This experience in general terms is applied to the NRX and NRU designs of fuel

  15. 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)

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

    Science.gov (United States)

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

    2018-04-01

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

  17. Development of MTR fuel plate with U-Al dispersion core constituents

    International Nuclear Information System (INIS)

    Bressiani, Jose Carlos

    1979-01-01

    This work is a contribution to the development of fuel plates for Research Nuclear Reaction Materials Test Reactors. The plates have the core constituted by dispersions of metallic uranium in aluminum. The main topics of this work are: 1) The preparation of uranium powder with particle sizes in the 53-105μm diameter range; 2) The mixture and cold-pressing of uranium and aluminum powders for different uranium concentrations; 3) The behavior of the dispersions in the roll milling conditions; 4) Blister, radiographic, metallographic and irradiation tests for quality control of the plates. The irradiation test was performed in the IEA-R1 swimming-pool reactor using a prototype with a dispersion of aluminum and natural uranium (45 w/o ), reaching an integrated neutron flux of 8.663 X 10 18 n/cm 2 , no visual changes being noticed after the completion of the experiment. The behavior of the uranium-aluminum reaction for dispersions with 45% w/o uranium also studied. X-ray diffraction experiments showed the formation of UAl 2 UAl 3 and UAl 4 , while energy dispersive analysis of X-rays(EDAX) demonstrated that the diffusion of aluminum in uranium is the mechanism responsible for that reaction. The activation energy for the U-Al reaction was determined by dilatometric experiments yielding 20.2 kcal/mol.The aluminum-uranium reaction reaches an end when extended to 96 h at 600 deg C, namely, when all the uranium is found in the UAl 4 composition. (author)

  18. Advanced Research Reactor Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. K.; Park, H. D.; Kim, K. H. (and others)

    2006-04-15

    RERTR program for non-proliferation has propelled to develop high-density U-Mo dispersion fuels, reprocessable and available as nuclear fuel for high performance research reactors in the world. As the centrifugal atomization technology, invented in KAERI, is optimum to fabricate high-density U-Mo fuel powders, it has a great possibility to be applied in commercialization if the atomized fuel shows an acceptable in-reactor performance in irradiation test for qualification. In addition, if rod-type U-Mo dispersion fuel is developed for qualification, it is a great possibility to export the HANARO technology and the U-Mo dispersion fuel to the research reactors supplied in foreign countries in future. In this project, reprocessable rod-type U-Mo test fuel was fabricated, and irradiated in HANARO. New U-Mo fuel to suppress the interaction between U-Mo and Al matrix was designed and evaluated for in-reactor irradiation test. The fabrication process of new U-Mo fuel developed, and the irradiation test fuel was fabricated. In-reactor irradiation data for practical use of U-Mo fuel was collected and evaluated. Application plan of atomized U-Mo powder to the commercialization of U-Mo fuel was investigated.

  19. 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)

  20. Tribological properties of thermally sprayed TiAl-Al2O3 composite coating

    Science.gov (United States)

    Salman, A.; Gabbitas, B.; Li, J.; Zhang, D.

    2009-08-01

    The use of thermal spray coatings provides protection to the surfaces operating in severe environments. The main goal of the current work is to investigate the possibility of using a high velocity oxy fuel (HVOF) thermally sprayed wear resistant TiAl/Al2O3 coating on tool steel (H13) which is used for making dies for aluminium high pressure die casting. A feedstock of TiAl/Al2O3 composite powder was produced from a mixture of Al and TiO2 powders by high energy mechanical milling, followed by a thermal reaction process. The feedstock was then thermally sprayed using a high velocity oxy-fuel (HVOF) technique onto H13 steel substrates to produce a composite coating. The present study describes and compares the tribological properties such as friction and sliding wear rate of the coating both at room and high temperature (700°C). The results showed that the composite coating has lower wear rate at high temperature (700°C) than the uncoated H13 sample. At Room temperature without using lubricant there is no much significant difference between the wear rate of the coated and uncoated samples. The experimental results showed that the composite coating has great potential for high temperature application due to its lower wear rate at high temperature in comparison with the uncoated sample at the same temperature. The composite coating was characterized using scanning electron microscopy (SEM), optical microscopy and X-ray diffractometry (XRD). This paper reports the experimental observations and discusses the wear resistance performance of the coatings at room and high temperatures.

  1. Tribological properties of thermally sprayed TiAl-Al2O3 composite coating

    International Nuclear Information System (INIS)

    Salman, A; Gabbitas, B; Zhang, D; Li, J

    2009-01-01

    The use of thermal spray coatings provides protection to the surfaces operating in severe environments. The main goal of the current work is to investigate the possibility of using a high velocity oxy fuel (HVOF) thermally sprayed wear resistant TiAl/Al 2 O 3 coating on tool steel (H13) which is used for making dies for aluminium high pressure die casting. A feedstock of TiAl/Al 2 O 3 composite powder was produced from a mixture of Al and TiO 2 powders by high energy mechanical milling, followed by a thermal reaction process. The feedstock was then thermally sprayed using a high velocity oxy-fuel (HVOF) technique onto H13 steel substrates to produce a composite coating. The present study describes and compares the tribological properties such as friction and sliding wear rate of the coating both at room and high temperature (700 deg. C). The results showed that the composite coating has lower wear rate at high temperature (700deg. C) than the uncoated H13 sample. At Room temperature without using lubricant there is no much significant difference between the wear rate of the coated and uncoated samples. The experimental results showed that the composite coating has great potential for high temperature application due to its lower wear rate at high temperature in comparison with the uncoated sample at the same temperature. The composite coating was characterized using scanning electron microscopy (SEM), optical microscopy and X-ray diffractometry (XRD). This paper reports the experimental observations and discusses the wear resistance performance of the coatings at room and high temperatures.

  2. Microstructural studies on chemical interactions in U-Mo with Al

    International Nuclear Information System (INIS)

    Martins, Ilson Carlos

    2010-01-01

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

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

    OpenAIRE

    YEON SOO KIM; G.L. HOFMAN; HO JIN RYU; JONG MAN PARK; A.B. ROBINSON; D.M. WACHS

    2013-01-01

    Interaction layer growth between U-Mo alloy fuel particles and Al in a dispersion fuel is a concern due to the volume expansion and other unfavorable irradiation behavior of the interaction product. To reduce interaction layer (IL) growth, a small amount of Si is added to the Al. As a result, IL growth is affected by the Si content in the Al matrix. In order to predict IL growth during fabrication and irradiation, empirical models were developed. For IL growth prediction during fabrication an...

  4. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-04-01

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

  6. Development of the continuous casting technology for fabrication of the tubular fuels

    International Nuclear Information System (INIS)

    Kim, H. S.; Lee, Y. S.; Kim, C. K.; Lee, D. B.; Oh, S. J.

    2003-01-01

    In the irradiation test of the U-Mo dispersed nuclear fuel that is used as nuclear fuels for research reactors, it was recognized that the swelling due to reaction between U-Mo particle and Al matrix caused some failures of the fuel claddings. The development of new style nuclear fuel that could minimize the reaction between U-Mo particles and Al matrix was needed. Tube style nuclear fuel was judged to be suitable as new style nuclear fuel. We targeted to make U-Mo tube of diameter 10mm, thinner than 1mm thick, because temperature distribution of tube style nuclear fuel will be expected to have a good performance. We used continuous casting technology to make tube style nuclear fuel. In this research, we have tried to make tube using copper before we make U-Mo tube style nuclear fuel by continuous casting method. As a result of the experiment, we succeeded to make copper tube of diameter 10mm, thickness 1mm

  7. Fuel cell technology; Brennstoffzellen-Technologie

    Energy Technology Data Exchange (ETDEWEB)

    Stimming, U; Friedrich, K A; Cappadonia, M; Vogel, R

    1999-12-31

    Hydrogen from fossil or renewable sources is an important fuel for low-emission power generation in fuel cells. Methanol and maybe also ethanol can also be produced by direct electrochemical processes in low-temperature fuel cells (PEMFC, PAFC). Fuel cell systems with high operating temperatures are highly flexible with regard to fuel but tend to have material problems. On the other hand, rapid developments in materials development and the possibility of production technology transfer from the electronics industry lead one to expect a breakthrough in the near future. But in spite of this, niche market applications will prevail. Since power stations have a longer life than motor vehicles and fuel cells in mobile applications, emission reductions from fuel cell applications in road vehicles are more probable on a medium-term basis than from applications in power stations. (orig.) [Deutsch] Wasserstoff, der sowohl aus fossilen wie auch aus regenerativen Quellen erschlossen werden kann, ist ein wesentlicher Brennstoff fuer die emissionsarme Elektrizitaetsproduktion in Brennstoffzellen. Methanol und eventuell Ethanol koennen auch direkt elektrochemisch in Niedertemperaturbrennstoffzellen (PEMFC, PAFC) umgesetzt werden. Brennstoffzellensysteme mit hohen Betriebstemperaturen erlauben eine hohe Flexibilitaet bezueglich der verwendeten Brennstoffe, sind aber nach wie vor durch starke Materialprobleme belastet. Die enormen Fortschritte in der Materialentwicklung einerseits sowie ein moeglicher Transfer von Fertigungstechnologien aus der Elektronikindustrie andererseits lassen eine zukuenftige grosstechnische Nutzung von Brennstoffzellen erwarten. Die technische Einfuehrung wird dennoch nur ueber Nischenmaerkte moeglich sein. Da die mittlere Lebensdauer eines Kraftwerks deutlich hoeher ist als die eines Strassenfahrzeugs, ausserdem Brennstoffzellen auch in staerkerem Masse in Fahrzeugen eingesetzt werden koennen, sind mittelfristig Emissionen eher durch

  8. Evaluation of Test Methodologies for Dissolution and Corrosion of Al-SNF

    International Nuclear Information System (INIS)

    Wiersma, B.J.; Mickalonis, J.I.; Louthan, M.R.

    1998-09-01

    The performance of aluminum-based spent nuclear fuel (Al-SNF) in the repository will differ from that of the commercial nuclear fuels and the high level waste glasses. The program consists of evaluating three test methods

  9. Development of the fabrication technology of the simulated fuel-I, 15,000MWd/tU

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kweon Ho; Kim, D. J.; Kim, H. S.; Lee, J. W.; Yang, M. S

    2001-04-01

    It is important to get basic data to analysis physical properties, behavior in reactor and performance of the DUPIC fuel because physical properties, fission gas release, grain growth and et al. of the DUPIC fuel is different from the commercial UO2 fuel. But what directly measures physical properties et al. of DUPIC fuel being resinterred simulated spent fuel through OREOX process is very difficult in laboratory owing to its high level radiation. Then fabrication of simulated DUPIC fuel is needed to measure its properties. In this study, the sintering characterization of wet milled powder for 24 hours to fabricate 15,000MWd/tU equivalent burnup simulated fuel.

  10. Inhibition of ammonia poisoning by addition of platinum to Ru/α-Al2 O3 for preferential CO oxidation in fuel cells.

    Science.gov (United States)

    Sato, Katsutoshi; Yagi, Sho; Zaitsu, Shuhei; Kitayama, Godai; Kayada, Yuto; Teramura, Kentaro; Takita, Yusaku; Nagaoka, Katsutoshi

    2014-12-01

    In polymer electrolyte fuel cell (PEFC) systems, small amounts of ammonia (NH3 ) present in the reformate gas deactivate the supported ruthenium catalysts used for preferential oxidation (PROX) of carbon monoxide (CO). In this study, we investigated how the addition of a small amount of platinum to a Ru/α-Al2 O3 catalyst (Pt/Ru=1:9 w/w) affected the catalyst's PROX activity in both the absence and the presence of NH3 (130 ppm) under conditions mimicking the reformate conditions during steam reforming of natural gas. The activity of undoped Ru/α-Al2 O3 decreased sharply upon addition of NH3 , whereas Pt/Ru/α-Al2 O3 exhibited excellent PROX activity even in the presence of NH3 . Ruthenium K-edge X-ray absorption near-edge structure (XANES) spectra indicated that in the presence of NH3 , some of the ruthenium in the undoped catalyst was oxidized in the presence of NH3 , whereas ruthenium oxidation was not observed with Pt/Ru/α-Al2 O3 . These results suggest that ruthenium oxidation is retarded by the platinum, so that the catalyst shows high activity even in the presence of NH3 . © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  11. Performance of the MTR core with MOX fuel using the MCNP4C2 code

    International Nuclear Information System (INIS)

    Shaaban, Ismail; Albarhoum, Mohamad

    2016-01-01

    The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U 3 O 8 &PuO 2 ) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U 3 O 8 -Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U 3 O 8 -Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with 235 U and the amount of loaded 235 U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. - Highlights: • Re-cycling of the ETRR-2 reactor by MOX fuel. • Increase the number of the neutronic traps from one neutronic trap to three neutronic trap. • Calculation of the criticality safety and neutronic parameters of the ETRR-2 reactor for the U 3 O 8 -Al original fuel and the MOX fuel.

  12. CONTAINMENT ANALYSIS METHODOLOGY FOR TRANSPORT OF BREACHED CLAD ALUMINUM SPENT FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.

    2010-07-11

    Aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site and placed in interim storage in a water basin. To enter the United States, a cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Many Al-SNF assemblies have suffered corrosion degradation in storage in poor quality water, and many of the fuel assemblies are 'failed' or have through-clad damage. A methodology was developed to evaluate containment of Al-SNF even with severe cladding breaches for transport in standard casks. The containment analysis methodology for Al-SNF is in accordance with the methodology provided in ANSI N14.5 and adopted by the U. S. Nuclear Regulatory Commission in NUREG/CR-6487 to meet the requirements of 10CFR71. The technical bases for the inputs and assumptions are specific to the attributes and characteristics of Al-SNF received from basin and dry storage systems and its subsequent performance under normal and postulated accident shipping conditions. The results of the calculations for a specific case of a cask loaded with breached fuel show that the fuel can be transported in standard shipping casks and maintained within the allowable release rates under normal and accident conditions. A sensitivity analysis has been conducted to evaluate the effects of modifying assumptions and to assess options for fuel at conditions that are not bounded by the present analysis. These options would include one or more of the following: reduce the fuel loading; increase fuel cooling time; reduce the degree of conservatism in the bounding assumptions; or measure the actual leak rate of the cask system. That is, containment analysis for alternative inputs at fuel-specific conditions and

  13. 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)

  14. Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

    Energy Technology Data Exchange (ETDEWEB)

    Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

    2017-08-01

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (< 20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. This paper discusses the TEM results of the U-10Mo/Zr/Al6061 monolithic fuel plate (Plate ID: L1P09T, ~ 59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory as part of RERTR-9B irradiation campaign with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 C, respectively. A total of 5 TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (> 1 µm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ~ 30 at% and ~ 7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

  15. Atomistic modelling of the interdiffusion of Al in the UMo based fuel

    International Nuclear Information System (INIS)

    Garces, Jorge E.; Marino, Armando C.; Bozzolo, Guillermo

    2002-01-01

    The analysis of the initial stages and trends in the process of interdiffusion of Al in the UMo solid solution as a function of Mo concentration is made using the BFS method for alloys. The approach presented in this work helps to understand the exchange mechanism between adatoms and substrate atoms in the binary systems, leading to the behavior observed in the ternary Al-U-Mo system. While in Al/U Al atoms show a noticeable tendency to interdiffusion in the bulk, in Al/Mo the same atoms show a tendency to layer-by-layer growth and the formation of structures in the overlayer. In the case of Al/U-Mo, the two competing behaviors observed for Al/Mo and Al/U, translate into the role of regions rich in Mo acting as interdiffusion barriers, in excellent agreement with experimental evidence. (author)

  16. Modeling RERTR experimental fuel plates using the PLATE code

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  17. Use of plate fuel elements for the RA3 reactor

    International Nuclear Information System (INIS)

    Parodi, C.; Parkanski, D.; Higa, M.; Marajofsky, A.

    1992-01-01

    The RA3 reactor is a pool reactor, redesigned for 5 MW dissipation. Nineteen plates are used in each fuel element. The utilization of 20% enriched U, gives the possibility of the development of rod type fuel with Al/U 3 O 8 cermets. The thermohydraulic and neutronic conditions are studied in this work in order to satisfy the stipulated power. In addition, the fabrication conditions of Al/U 3 O 8 and Al/U 3 O 8 /Zr H 2 cermets with densities within the limits imposed by the thermohydraulics and neutronics conditions are studied. (author)

  18. Fuel element reactivity worth in different rings of the IPR-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gomes do Prado Souza, Rose Mary

    2008-10-29

    The thermal power of the IPR-R1 TRIGA Reactor will be upgraded from 100 kW to 250 kW. Starting core: loaded with 59 aluminum cladded fuel elements; 1.34 $ excess reactivity; and 100 kW power. It is planned to go 2.5 times the power licensed, i.e., 250 kW. This forces to enlarge the reactivity level. Nuclear reactors must have sufficient excess reactivity to compensate the negative reactivity feedback effects caused by: the fuel temperature, fuel burnup, fission poisoning production, and to allow full power operation for predetermined period of time. To provide information for the calculation of the new core arrangement, the reactivity worth of some fuel elements in the core were measured as well as the determination of the core reactivity increase in the substitution of the original fuels, cladded with aluminium, for new ones, cladded with stainless steel. The reactivity worth of fuel element was measured from the difference in critical position of the control rods, calibrated by the positive period method, before and after the fuel element was withdrawn from the core. The magnitude of reactivity increase was determined when withdrawing the original Al-clad fuel (a little burned up) and the graphite elements, and inserting a fresh Al-clad fuel element, one by one. Experimental results indicated that to obtain enough reactivity excess to increase the rector power the addition of 4 new fuel elements in the core would be sufficient: - Substitution of 4 Al-clad fuel elements in ring C for fresh stainless steel clad fuel elements; - increase the reactivity {approx_equal} 4 x 6.5 = 26 cents; - The removed 4 Al-clad F. E. (a little burned up) put in the core periphery, ring F, replacing graphite elements; - add < 4 x 39 156 cents (39 cents was measured with a fresh F.E.). Neutron source was changed from position F7 to F8. Control and Safety rods were moved from ring D to C in order to increase their reactivity worth. Regulating rod was kept at the same position, F16. Four

  19. 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)

  20. Comparison of the parameters of the IR-8 reactor with different fuel assembly designs with LEU fuel

    International Nuclear Information System (INIS)

    Vatulin, A.; Stetsky, Y.; Dobrikova, I.

    1999-01-01

    The estimation of neutron-physical, heat and hydraulic parameters of the IR-8 research reactor with low enriched uranium (LEU) fuel was performed. Two fuel assembly (FA) designs were reviewed: IRT-4M with the tubular type fuel elements and IRT-MR with the rod type fuel elements. UO 2 -Al dispersion 19.75% enrichment fuel is used in both cases. The results of the calculations were compared with main parameters of the reactor, using the current IRT-3M FA with 90% high enriched uranium (HEU) fuel. The results of these comparisons showed that during the LEU conversion of the reactor the cycle length, excess reactivity and peak power of the IRT-MR type FA are higher than for the IRT-3M type FA and IRT-4M type FA. (author)

  1. Disposal criticality analysis for aluminum-based DOE fuels

    International Nuclear Information System (INIS)

    Davis, J.W.; Gottlieb, P.

    1997-11-01

    This paper describes the disposal criticality analysis for canisters containing aluminum-based Department of Energy fuels from research reactors. Different canisters were designed for disposal of highly enriched uranium (HEU) and medium enriched uranium (MEU) fuel. In addition to the standard criticality concerns in storage and transportation, such as flooding, the disposal criticality analysis must consider the degradation of the fuel and components within the waste package. Massachusetts Institute of Technology (MIT) U-Al fuel with 93.5% enriched uranium and Oak Ridge Research Reactor (ORR) U-Si-Al fuel with 21% enriched uranium are representative of the HEU and MEU fuel inventories, respectively. Conceptual canister designs with 64 MIT assemblies (16/layer, 4 layers) or 40 ORR assemblies (10/layer, 4 layers) were developed for these fuel types. Borated stainless steel plates were incorporated into a stainless steel internal basket structure within a 439 mm OD, 15 mm thick XM-19 canister shell. The Codisposal waste package contains 5 HLW canisters (represented by 5 Defense Waste Processing Facility canisters from the Savannah River Site) with the fuel canister placed in the center. It is concluded that without the presence of a fairly insoluble neutron absorber, the long-term action of infiltrating water can lead to a small, but significant, probability of criticality for both the HEU and MEU fuels. The use of 1.5kg of Gd distributed throughout the MIT fuel and the use of carbon steels for the structural basket or 1.1 kg of Gd distributed in the ORR fuel will reduce the probability of criticality to virtually zero for both fuels

  2. Results of Microstructural Examinations of Irradiated LEU U-Mo Fuels

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  3. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

  4. Improvement of performance in low temperature solid oxide fuel cells operated on ethanol and air mixtures using Cu-ZnO-Al2O3 catalyst layer

    Science.gov (United States)

    Morales, M.; Espiell, F.; Segarra, M.

    2015-10-01

    Anode-supported single-chamber solid oxide fuel cells with and without Cu-ZnO-Al2O3 catalyst layers deposited on the anode support have been operated on ethanol and air mixtures. The cells consist of gadolinia-doped ceria electrolyte, Ni-doped ceria anode, and La0.6Sr0.4CoO3-δ-doped ceria cathode. Catalyst layers with different Cu-ZnO-Al2O3 ratios are deposited and sintered at several temperatures. Since the performance of single-chamber fuel cells strongly depends on catalytic properties of electrodes for partial oxidation of ethanol, the cells are electrochemically characterized as a function of the temperature, ethanol-air molar ratio and gas flow rate. In addition, catalytic activities of supported anode, catalytic layer-supported anode and cathode for partial oxidation of ethanol are analysed. Afterwards, the effect of composition and sintering temperature of catalyst layer on the cell performance are determined. The results indicate that the cell performance can be significantly enhanced using catalyst layers of 30:35:35 and 40:30:30 wt.% Cu-ZnO-Al2O3 sintered at 1100 °C, achieving power densities above 50 mW cm-2 under 0.45 ethanol-air ratio at temperatures as low as 450 °C. After testing for 15 h, all cells present a gradual loss of power density, without carbon deposition, which is mainly attributed to the partial re-oxidation of Ni at the anode.

  5. Protection of spent aluminum-clad research reactor fuels during extended wet storage

    International Nuclear Information System (INIS)

    Fernandes, Stela M.C.; Correa, Olandir V.; Souza, Jose A.; Ramanathan, Lalgudi V.; Antunes, Renato A.

    2013-01-01

    Aluminum-clad spent nuclear fuel from research reactors (RR) is stored in light water filled pools or basins worldwide. Many incidences of pitting corrosion of the fuel cladding has been reported and attributed to synergism in the effect of certain water parameters. Protection of spent Al-clad RR fuel with a conversion coating was proposed in 2008. Preliminary results revealed increased pitting corrosion resistance of cerium oxide coated aluminum alloys AA 1050 and AA 6061, used as RR fuel plate cladding. Further development of conversion coatings for Al alloys was carried out and this paper presents: (a) the preparation and characterization of hydrotalcite (HTC) coatings; (b) the results of laboratory tests in which the corrosion behavior of coated Al alloys in NaCl solutions was determined; (c) the results of field tests in which un-coated, boehmite coated, HTC coated and cerium modified boehmite / HTC coated AA 1050 and AA 6061 coupons were exposed to the IEA-R1 reactor spent fuel basin for extended periods. In these field tests the coupons coated with HTC from a high temperature (HT) bath and subsequently modified with Ce were the most resistant to pitting corrosion. In laboratory tests also, HT- hydrotalcite + Ce coated specimens were the most corrosion resistant in 0.01 M NaCl. The role of cerium in increasing the corrosion resistance imparted by the different conversion coatings of spent Al-clad RR fuel elements is presented. (author)

  6. A high-performance aluminum-feed microfluidic fuel cell stack

    Science.gov (United States)

    Wang, Yifei; Leung, Dennis Y. C.

    2016-12-01

    In this paper, a six-cell microfluidic fuel cell (MFC) stack is demonstrated. Low-cost aluminum is fed directly to the stack, which produces hydrogen fuel on site, through the Al-H2O reaction. This design is not only cost-efficient, but also eliminates the need for hydrogen storage. Unlike the conventional MFC stacks which generally require complex electrolyte distribution and management, the present Al-feed MFC stack requires only a single electrolyte stream, flowing successively through individual cells, which is finally utilized for hydrogen generation. In this manner, the whole system is greatly simplified while the operational robustness is also improved. With 2 M sodium hydroxide solution as electrolyte and kitchen foil Al as fuel, the present six-cell stack (in series) exhibits an open circuit voltage of nearly 6 V and a peak power density of 180.6 mWcm-2 at room temperature. In addition, an energy density of 1 Whg-1(Al) is achieved, which is quite high and comparable with its proton exchange membrane-based counterparts. Finally, pumpless operation of the present stack, together with its practical applications are successfully demonstrated, including lightening LED lights, driving an electric fan, and cell phone charging.

  7. Contribuciones al conocimiento del género Phragmotheca Cuatr. (Bombacaceae-Quararibeae

    Directory of Open Access Journals (Sweden)

    Fernández Alonso José Luis

    1996-11-01

    Full Text Available The systematic position of Phragmotheca in the tribe quararibeae is discussed and a key for the Bombacaceae with simple or unifoliolate leaves occurring in northern South America is included. The diagnostic utility of some characters in the segregation of species in Phragmotheca, is reviewed and a diagnostic key is presented. Four new species and two new infraspecific taxa are described and illustrated: P.rubriflora, P.lemniscata, P.hydra, P.siderosa subsp. megacarpa and P. amazonica varo leticiana from Colombia and P.sidereotricha from Perú.Se comenta la posición sistemática de Phragmotheca en relación con la tribu Quararibeae y se incluye una clave para la separación de los géneros de bombacáceas de hoja entera o unifoliolada del norte de Suramérica. Se revisa la utilidad de algunos caracteres en la separación de las especies y se presenta una clave diagnóstica. Se describen e ilustran 4 nuevas especies y dos taxones infraespecíficos P. rubriflora, P. lemniscata, P. hydra, P. siderosa subsp. megacarpa, P. amazonica varo leticiana de Colombia y P. sidereotricha de Perú.

  8. Hydrographic Measurements in the Western Arlboran Sea, October 1982.

    Science.gov (United States)

    1984-04-01

    about 1 km. Individual fixes based on the satellite system were accurate to about 0.5 km and on omega to about 2.0 km. 9 Table 3. Salinity Calibration...Activity Technical Note 224. 100 p. Cano, C. and F.F. de Castillejo (1972). Contribucion al Conocimiento del Mar de Alboran: III. Variaciones del

  9. Thermal-hydraulic analysis for core conversion to the use of low-enriched uranium fuels in the KUR

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Kanda, Keiji; Shibata, Toshikazu

    1985-01-01

    A feasibility study has been performed on the core conversion to the use of low-enriched uranium (LEU) fuels in the KUR. Five fuel element geometries are studied. For each fuel element, the relation between the pressure drop and the flow rate, critical heat flux, and heat fluxes for the onset of flow instability and the onset of nucleate boiling are calculated using the computer code PLTEMP3.MOD1 which has been developed for this analysis. The effect of fuel material (UAl x -Al, U 3 O 8 -Al and U 3 Si 2 -Al) on the peak fuel temperatures is also studied. A particular interest in the mixed core which may be constructed on the way to the use of LEU fuels, the change in the bypass flow rate due to the change in the gap between different fuel elements is investigated. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-15

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

  11. 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)

  12. Los Métodos de Enseñanza de Lenguas y las Teorías de Aprendizaje

    OpenAIRE

    Hernández Reinoso, Francisco Luis

    2000-01-01

    Este trabajo intenta precisar las contribuciones de la psicología a la enseñanza de lenguas a partir del análisis de sus métodos de enseñanza más reconocido, destacando en ellos como una u otra visión del aprendizaje concretada en procedimientos tiene sus antecedentes en aproximaciones teóricas de la psicología al aprendizaje de lenguas en particular o al aprendizaje en general.

  13. Behaviour of HTGR coated fuel particles at high-temperature tests

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Lyutikov, R.A.; Kurbakov, S.D.; Repnikov, V.M.; Khromonozhkin, V.V.; Soloviyov, G.I.

    1990-01-01

    At the temperature range 1200-2600 deg. C prereactor tests of TRISO fuel particles on the base of UO 2 , UC x O y and UO 2 +2Al 2 O 3 . SiO 2 kernels, and also fuel particle models with ZrC kernels were performed. Isothermal annealings carried out at temperatures of 1400-2600 deg. C, thermogradient ones at 1200-2200 deg. C (Δ T = 200-1200 deg. C/cm). It is shown that at heating to 2200 deg. C integrity of fuel particles is limited by different thermal expansion of PyC and SiC coatings, and also by thermal dissociation of SiC. At higher temperatures the failure is caused by development of high pressures within weakened fuel particles. It is found that uranium migration from alloyed fuel (UC x O y , UO 2 +2Al 2 O 3 .SiO 2 ) in the process of annealing is higher than that from UO 2 . (author)

  14. Aportaciones de las personas mayores a la sociedad: Análisis sociológico

    Directory of Open Access Journals (Sweden)

    Mª TERESA BAZO

    1996-01-01

    Full Text Available En el artículo se exponen algunos resultados de una investigación sobre actividad en la vejez y contribuciones de las personas jubiladas a la sociedad, realizada por la autora con una muestra de más de 60 personas, la mayor parte jubiladas y otras expertas en distintas áreas, residentes en el País Vasco, Navarra, Barcelona, Madrid y Zaragoza. se ha utilizado la entrevista en profundidad semiestructurada y la entrevista de grupo. los resultados expuestos son relativos, sobre todo, a las actividades altruistas de las personas jubiladas, asi como algunas contribuciones al bienestar económico de la sociedad. una conclusión es que las nuevas personas jubiladas solo necesitan oportunidades para desarrollar todo su potencial. el resto lo ponen ellas.

  15. 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)

  16. 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)

  17. Development of the Melt-Dilute Treatment Technology for Al-Based DOE Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Peacock, H.B.; Adams, T.M.; Iyer, N.C.

    1998-09-01

    Spent foreign and domestic research reactor fuel assemblies will be sent to Savannah River Site and prepared for interim and eventual geologic storage. Many of the fuel plates have been made with high enriched uranium, and during long term storage, the integrity of the fuel maybe effected if the canister is breached. To reduce the potential for criticality, proliferation, and reduce storage volume, a new treatment technology called melt-dilute is being developed at SRS. The technique will melt the spent fuel assemblies and will dilute the isotopic content to below 20%. The process is simple and versatile

  18. TiO2 and Al2O3 promoted Pt/C nanocomposites as low temperature fuel cell catalysts for electro oxidation of methanol in acidic media

    International Nuclear Information System (INIS)

    Naeem, Rabia; Ahmed, Riaz; Ansari, Muhammad Shahid

    2014-01-01

    Carbon corrosion and platinum dissolution are the two major catalyst layer degradation problems in polymer electrolyte membrane fuel cells (PEMFC). Ceramic addition can reduce the corrosion of carbon and increase the stability of catalysts. Pt/TiO 2 , Pt/TiO 2 -C, Pt/Al 2 O 3 and Pt/Al 2 O 3 -C catalysts were synthesized and characterized. Electrochemical surface area of Pt/TiO 2 -C and Pt/Al 2 O 3 -C nanocomposite catalysts was much higher than the Pt/TiO 2 and Pt/Al 2 O 3 catalysts. Peak current, specific activity and mass activity of the catalysts was also determined by cyclic voltammetry and were much higher for the carbon nanocomposites. Exchange current densities were determined from Tafel plots. Heterogeneous rates of reaction of electro oxidation of methanol were determined for all the catalysts and were substantially higher for titania catalysts as compared to alumina added catalysts. Mass activity of Pt/TiO 2 -C was much higher than mass activity of Pt/Al 2 O 3 -C. Stability studies showed that addition of ceramics have increased the catalytic activity and durability of the catalysts considerably

  19. Fuel performance analysis for the HAMP-1 mini plate test

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  20. A review of fusion device fuel cleanup systems

    International Nuclear Information System (INIS)

    Dombra, A.H.; Carney, M.

    1985-01-01

    Design options for a small fusion fuel purification system are assessed by comparing six conceptual system designs based on one of the following: a Zr/Al getter pump for in vacuo applications, a cryogenic molecular sieve adsorber at 77K, a palladium-alloy membrane diffuser, a U-bed reactor at 1170K, a two-compartment cryogenic freezer at 27-50K and 50-300K, a U-bed and non-regenerative Zr/Al gas purifier. The latter system introduces a new concept of fuel purification based on well-established techniques: recovery of purified D 2 -DT-T 2 from a helium carrier gas with the U-bed, followed by the removal of impurities from the carrier gas with the non-regenerative Zr/Al gas purifier. The main advantages of this system are simplicity, safety and relatively small quantity of tritiated waste produced by the process. The tritium in the waste is immobilized as a stable tritide of Zr/Al

  1. Vacuum Drying Tests for Storage of Aluminum Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Chen, K.F.; Large, W.S.; Sindelar, R.L.

    1998-05-01

    A total inventory of up to approximately 32,000 aluminum-based spent nuclear fuel (Al SNF) assemblies are expected to be shipped to Savannah River Site (SRS) from domestic and foreign research reactors over the next several decades. Treatment technologies are being developed as alternatives to processing for the ultimate disposition of Al SNF in the geologic repository. One technology, called Direct/Co-disposal of Al SNF, would place the SNF into a canister ready for disposal in a waste package, with or without canisters containing high-level radioactive waste glass logs, in the repository. The Al SNF would be transferred from wet storage and would need to be dried in the Al SNF canister. The moisture content inside the Al SNF canister is limited to avoid excessive Al SNF corrosion and hydrogen buildup during interim storage before disposal. A vacuum drying process was proposed to dry the Al SNF in a canister. There are two major concerns for the vacuum drying process. One is water inside the canister could become frozen during the vacuum drying process and the other one is the detection of dryness inside the canister. To vacuum dry an irradiated fuel in a heavily shielded canister, it would be very difficult to open the lid to inspect the dryness during the vacuum drying operation. A vacuum drying test program using a mock SNF assembly was conducted to demonstrate feasibility of drying the Al SNF in a canister. These tests also served as a check-out of the drying apparatus for future tests in which irradiated fuel would be loaded into a canister under water followed by drying for storage

  2. Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    El-Sahlamy, Neama M. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt)

    2017-11-15

    In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1{sub X}-A1, U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1{sub X}-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U{sub 3}Si{sub 2}-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

  3. Irradiation experience of IPEN fuel at IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

    1998-01-01

    IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

  4. Multiscale Multiphysics Developments for Accident Tolerant Fuel Concepts

    International Nuclear Information System (INIS)

    Gamble, K. A.; Hales, J. D.; Yu, J.; Zhang, Y.; Bai, X.; Andersson, D.; Patra, A.; Wen, W.; Tome, C.; Baskes, M.; Martinez, E.; Stanek, C. R.; Miao, Y.; Ye, B.; Hofman, G. L.; Yacout, A. M.; Liu, W.

    2015-01-01

    U 3 Si 2 and iron-chromium-aluminum (Fe-Cr-Al) alloys are two of many proposed accident-tolerant fuel concepts for the fuel and cladding, respectively. The behavior of these materials under normal operating and accident reactor conditions is not well known. As part of the Department of Energy's Accident Tolerant Fuel High Impact Problem program significant work has been conducted to investigate the U 3 Si 2 and FeCrAl behavior under reactor conditions. This report presents the multiscale and multiphysics effort completed in fiscal year 2015. The report is split into four major categories including Density Functional Theory Developments, Molecular Dynamics Developments, Mesoscale Developments, and Engineering Scale Developments. The work shown here is a compilation of a collaborative effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory and Anatech Corp.

  5. Informal presentations by fuel fabricators and others [contributed by A. Nishiyama, Nuclear Fuel Industries, Ltd.

    International Nuclear Information System (INIS)

    Nishiyama, A.

    1993-01-01

    This paper contains a brief summary of activities in the field of research reactor fuel fabrication in Nuclear Fuel Industries Sumitomo and Furukawa Industries. Since 1956 2 million dollars were spent for development of nuclear fuels and plant facilities including complete manufacturing and testing capabilities. Now this company is the only fuel supplier for the research reactors in Japan. The fabrication process starts with the melting, alloying, and casting of U-Al. The uranium billets are prepared by foreign fabricators. The uranium content varies from 13 to 22 wt % according to the purchaser's specifications. In making fuel plates, the picture frame method is applied. In this case, the original procedure is sufficiently effective in avoiding dogboning. The plates are finished by hot and cold roll milling and inspected dimensionally, metallurgically, and mechanically, and at the same time the blister test and X-ray radiographic tests are performed. Fuel elements are assembled by rolling flat or curved plates into side plate grooves and end-fit welding. Finished elements are tested dimensionally and hydraulically. Nominal losses during operation are less than 1% of the uranium metal. Our present capacity licensed by the Japanese Government is approximately 950 fuel elements a year. About 35 employees including engineers are engaged in development and manufacturing of fuels. Owing to the small limited demand of the research reactor fuels in Japan during the past 20 years (mostly in last 10 years), we processed only about 350 kg of highly enriched uranium and supplied approximately 1000 fuel elements to JAERI, Kyoto University, and others, and we have been suffering red-ink balance of budget every year. Some of trials in development are briefly discussed. In case of UO 2 -Al metal fuel plates, the vibratory compacting method was very popular among many researchers about 10 years ago. A lot of time and money was spent to study the economic fabrication process of

  6. Further data of silicide fuel for the LEU conversion of JMTR

    International Nuclear Information System (INIS)

    Saito, M.; Futamura, Y.; Nakata, H.; Ando, H.; Sakurai, F.; Ooka, N.; Sakakura, A.; Ugajin, M.; Shirai, E.

    1990-01-01

    Silicide fuel data for the safety assessment of the JMTR LEU fuel conversion are being measured. The data include fission product release, thermal properties, behaviour under accident conditions, and metallurgical characteristics. The methods used in the experiments are discussed. Results of fission products release at high temperature are described. The release of iodine from the silicide fuel is considerably lower than for U-Al alloy fuel

  7. Status report on conversion of the Georgia Tech Research Reactor to low enrichment fuel

    International Nuclear Information System (INIS)

    Karam, R.A.; Matos, J.E.; Mo, S.C.; Woodruff, W.L.

    1995-01-01

    The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the technical specifications of the GTRR so that LEU U 3 Si 2 -Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort. (author)

  8. Status report on conversion of the Georgia Tech Research Reactor to low enrichment fuel

    International Nuclear Information System (INIS)

    Karam, R.A.; Matos, J.E.; Mo, S.C.; Woodruff, W.L.

    1991-01-01

    The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the Technical Specifications of the GTRR so that LEU U 3 Si 2 -Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort

  9. Novel inorganic materials for polymer electrolyte and alkaline fuel cells

    Science.gov (United States)

    Tadanaga, Kiyoharu

    2012-06-01

    Inorganic materials with high ionic conductivity must have big advantages for the thermal and long term stability when the materials are used as the electrolyte of fuel cells. In the present paper, novel ionic conductive inorganic materials for polymer electrolyte fuel cells (PEFCs) and all solid state alkaline fuel cells (AFCs) that have been developed by our group have been reviewed. PEFCs which can operate in temperature range from 100 to 200 °C are intensively studied because of some advantages such as reduction of CO poisoning of Pt catalyst and acceleration of electrode reactions. We showed that the fuel cells using the composite membranes prepared from phosphosilicate gel powder and polyimide precursor can operate in the temperature range from 30 to 180 °C. We also found that the inorganic-organic hybrid membranes with acid-base pairs from 3-aminopropyl triethoxy silane and H2SO4 or H3PO4 show high proton conductivity under dry atmosphere, and the membranes are thermally stable at intermediate temperatures. On the other hand, because the use of noble platinum is the serious problem for the commercialization of PEFCs and because oxidation reactions are usually faster than those of acid-type fuel cells, alkaline type fuel cells, in which a nonplatinum catalyst can be used, are attractive. Recently, we have proposed an alkaline-type direct ethanol fuel cell (DEFC) using a natural clay electrolyte with non-platinum catalysts. So-called hydrotalcite clay, Mg-Al layered double hydroxide intercalated with CO32- (Mg-Al CO32- LDH), has been proved to be a hydroxide ion conductor. An alkalinetype DEFC using Mg-Al CO32- LDH as the electrolyte and aqueous solution of ethanol and potassium hydroxide as a source of fuel exhibited excellent electrochemical performance.

  10. Development of high performance hybrid rocket fuels

    Science.gov (United States)

    Zaseck, Christopher R.

    . In order to examine paraffin/additive combustion in a motor environment, I conducted experiments on well characterized aluminum based additives. In particular, I investigate the influence of aluminum, unpassivated aluminum, milled aluminum/polytetrafluoroethylene (PTFE), and aluminum hydride on the performance of paraffin fuels for hybrid rocket propulsion. I use an optically accessible combustor to examine the performance of the fuel mixtures in terms of characteristic velocity efficiency and regression rate. Each combustor test consumes a 12.7 cm long, 1.9 cm diameter fuel strand under 160 kg/m 2s of oxygen at up to 1.4 MPa. The experimental results indicate that the addition of 5 wt.% 30 mum or 80 nm aluminum to paraffin increases the regression rate by approximately 15% compared to neat paraffin grains. At higher aluminum concentrations and nano-scale particles sizes, the increased melt layer viscosity causes slower regression. Alane and Al/PTFE at 12.5 wt.% increase the regression of paraffin by 21% and 32% respectively. Finally, an aging study indicates that paraffin can protect air and moisture sensitive particles from oxidation. The opposed burner and aluminum/paraffin hybrid rocket experiments show that additives can alter bulk fuel properties, such as viscosity, that regulate entrainment. The general effect of melt layer properties on the entrainment and regression rate of paraffin is not well understood. Improved understanding of how solid additives affect the properties and regression of paraffin is essential to maximize performance. In this document I investigate the effect of melt layer properties on paraffin regression using inert additives. Tests are performed in the optical cylindrical combustor at ˜1 MPa under a gaseous oxygen mass flux of ˜160 kg/m2s. The experiments indicate that the regression rate is proportional to mu0.08rho 0.38kappa0.82. In addition, I explore how to predict fuel viscosity, thermal conductivity, and density prior to testing

  11. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, Jason Dean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barani, Tommaso [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pizzocri, Davide [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

  12. Performance of a diesel engine operating on raw coal-diesel fuel and solvent refined coal-diesel fuel slurries. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, H.P.

    1980-03-01

    Performance tests using an 11 kW single cylinder diesel engine were made to determine the effects of three different micronized coal-fuel oil slurries being considered as alternative fuels. Slurries containing 20, 32, and 40%-wt micronized raw coal in No. 2 fuel oil were used. Results are presented indicating the changes in the concentrations of SO/sub X/ and NO/sub X/ in the exhaust, exhaust opacity, power and efficiency, and in wear rates relative to operation on fuel oil No. 2. The engine was operated for 10 h at full load and 1400 rpm on al fuels except the 40%-wt slurry. This test was discontinued because of extremely poor performance.

  13. LA PERSPECTIVA HISTÓRICO-CULTURAL Y LA EDUCACIÓN ESPECIAL: CONTRIBUCIONES INICIALES Y DESARROLLOS ACTUALES

    Directory of Open Access Journals (Sweden)

    Albertina Mitjáns Martínez

    2009-01-01

    Full Text Available El objetivo principal del artículo es destacar la significación de la perspectiva histórico-cultural del desarrollo humano para el diseño de estrategias educativas favorecedoras del aprendizaje escolar de niños y niñas con deficiencias. A partir de las contribuciones de Vygotsky, en ese campo, y de los aportes de la Teoría de la Subjetividad, elaborada a partir da la perspectiva histórico-cultural, se presentan principios que, en su configuración sistémica, pueden constituir una base sólida para la reestructuración de diversos aspectos del trabajo educativo en la Educación Especial. En la primera parte del artículo son discutidas algunas de las concepciones y conceptos desarrollados por Vygotsky en su obra "Fundamentos de Defectología". La fundamentación del carácter social de la deficiencia, los conceptos de defecto primario, defecto secundario y compensación, y la consideración del carácter creativo del aprendizaje y del desarrollo de las niñas y los niños deficientes. Se presentan las ideas centrales de la Teoría de la Subjetividad elaborada por González Rey, especialmente la concepción del aprendizaje escolar como un proceso de la subjetividad irreducible a habilidades y procesos intelectuales. Por último, se fundamentan tres implicaciones educacionales derivadas de la articulación de los dos sistemas conceptuales presentados: favorecer la creación de espacios comunicativos/relacionales que contribuyan a generar en el niño nuevas producciones de sentido sobre el aprendizaje y sobre la deficiencia, estimular en ellos el desarrollo de la condición de sujetos negada por las posiciones asistencialistas y paternalistas dominantes y el enfoque cualitativo, constructivo, singular, procesual e interactivo de los procesos de diagnóstico y de evaluación educacional.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  15. Final disposition of MTR fuel

    International Nuclear Information System (INIS)

    Jonnson, Erik B.

    1996-01-01

    The final disposition of power reactor fuel has been investigated for a long time and some promising solutions to the problem have been shown. The research reactor fuels are normally not compatible with the zirkonium clad power reactor fuel and can thus not rely on the same disposal methods. The MTR fuels are typically Al-clad UAl x or U 3 Si 2 , HEU resp. LEU with essentially higher remaining enrichment than the corresponding power reactor fuel after full utilization of the uranium. The problems arising when evaluating the conditions at the final repository are the high corrosion rate of aluminum and uranium metal and the risk for secondary criticality due to the high content on fissionable material in the fully burnt MTR fuel. The newly adopted US policy to take back Foreign Research Reactor Spent Fuel of US origin for a period of ten years have given the research reactor society a reasonable time to evaluate different possibilities to solve the back end of the fuel cycle. The problem is, however, complicated and requires a solid engagement from the research reactor community. The task would be a suitable continuation of the RERTR program as it involves both the development of new fuel types and collecting data for the safe long-term disposal of the spent MTR fuel. (author)

  16. Fuel element replacement and cooling water radioactivity at the Musashi reactor

    International Nuclear Information System (INIS)

    Nozaki, T.; Honda, T.; Horiuchi, N.; Aizawa, O.; Sato, T.

    1988-01-01

    The Musashi reactor (TRIGA-II, 100kW) has been operated without any serious troubles since 1963. In 1985 the old Al-cladded fuel elements were replaced with new stainless cladded ones in order to insure a long and safe operation. By using a semi-automatic equipment the old fuel elements have been transferred into the bulk-shielding experimental pool, which was remodelled for the spent-fuel storage. In order to reduce the exposure during the transfer work, the old fuel elements were cooled in the core tank for 3 months. After the replacement, the radioactivities in the cooling water have been drastically changed. The activity of Na-24 decreased about one decade, and the activities of Cr-51, Mn-54, Mn-56, Co-58 and Co-60 increased about two decades. At this conference we will report on the following points: (1) semi-automatic equipment for the transportation of the Al-cladded spent fuel, (2) structure of spent-fuel storage pool, and (3) radioactivity change in the cooling water. (author)

  17. Interdiffusion between U-Mo alloys and Al or Al alloys at 340 deg. C. Irradiation plan

    International Nuclear Information System (INIS)

    Fortis, A.M.; Mirandou, M.; Ortiz, M.; Balart, S.; Denis, A.; Moglioni, A.; Cabot, P.

    2005-01-01

    Out of reactor interdiffusion experiments between U-Mo alloys and Al alloys made close to fuel operation temperature are needed to validate the results obtained above 500 deg. C. A study of interdiffusion between U-Mo and Al or Al alloys, out and in reactor, has been initiated. The objective is to characterize the interdiffusion layer around 250 deg. C and study the influence of neutron irradiation. Irradiation experiments will be performed in the Argentine RA3 reactor and chemical diffusion couples will be fabricated by Friction Stir Welding (FSW) technique. In this work out-of-pile diffusion experiments performed at 340 deg. C are presented. Friction Stir Welding (FSW) was used to fabricate some of the samples. One of the results is the presence of Si, in the interaction layer, coming from the Al alloy. This is promising in the sense that the absence of Al rich phases may also be expected at low temperature. (author)

  18. Synthesis and electrochemical characterization of Ti{sub x}Ta{sub y}Al{sub z}N{sub 1-δ}O{sub γ} for fuel cell catalyst supports

    Energy Technology Data Exchange (ETDEWEB)

    Wakabayashi, Ryo H.; Abruña, Héctor D., E-mail: hda1@cornell.edu; DiSalvo, Francis J., E-mail: fjd3@cornell.edu

    2017-02-15

    Quinary Ti{sub x}Ta{sub y}Al{sub z}N{sub 1-δ}O{sub γ} of various compositions have been prepared by a co-precipitation method followed by ammonolysis. The nitride samples were examined as potential catalyst supports in polymer electrolyte membrane fuel cells. The nitride products crystallized in the rock salt (NaCl) structure over a wide range of compositions. The addition of Ta and Al was highly beneficial towards improving the chemical and electrochemical stability of TiN, without a significant loss of electrical conductivity. Platinum particles were successfully deposited on the (oxy)nitride samples, and the composite samples at some compositions were found to be comparable to Pt/carbon in their stability and catalytic activity even without optimizing the Pt deposition and dispersion processes. - Graphical abstract: The effect of additions of Ta and Al into TiN structure. Shifts the lattice constant, and increases its chemical stability in acidic environment.

  19. Fuel for the Future: Biodiesel - A Case study

    Science.gov (United States)

    Lutterbach, Márcia T. S.; Galvão, Mariana M.

    High crude oil prices, concern over depletion of world reserves, and growing apprehension about the environment, encouraged the search for alternative energy sources that use renewable natural resources to reduce or replace traditional fossil fuels such as diesel and gasoline (Hill et al., 2006). Among renewable fuels, biodiesel has been attracting great interest, especially in Europe and the United States. Biodiesel is defined by the World Customs Organization (WCO) as 'a mixture of mono-alkyl esters of long-chain [C16-C18] fatty acids derived from vegetable oils or animal fats, which is a domestic renewable fuel for diesel engines and which meets the US specifications of ASTM D 6751'. Biodiesel is biodegradable and non toxic, produces 93% more energy than the fossil energy required for its production, reduces greenhouse gas emissions by 40% compared to fossil diesel (Peterson and Hustrulid, 1998; Hill et al., 2006) and stimulates agriculture.

  20. Postirradiation analysis of experimental uranium-silicide dispersion fuel plates

    International Nuclear Information System (INIS)

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

    1985-01-01

    Low-enriched uranium silicide dispersion fuel plates were irradiated to maximum burnups of 96% of 235 U. Fuel plates containing 33 v/o U 3 Si and U 3 Si 2 behaved very well up to this burnup. Plates containing 33 v/o U 3 Si-Al pillowed between 90 and 96% burnup of the fissile atoms. More highly loaded U 3 Si-Al plates, up to 50 v/o were found to pillow at lower burnups. Plates containing 40 v/o U 3 Si showed an increase swelling rate around 85% burnup. 5 refs., 10 figs

  1. Catalytic copyrolysis of particle board and polypropylene over Al-MCM-48

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hannah; Choi, Suek Ju [School of Environmental Engineering, University of Seoul, Seoul 02504 (Korea, Republic of); Kim, Ji Man [Department of Chemistry, Sungkyunkwan University, Suwon 16419 (Korea, Republic of); Jeon, Jong-Ki [Department of Chemical Engineering, Kongju National University, Cheonan 31080 (Korea, Republic of); Park, Sung Hoon; Jung, Sang-Chul [Department of Environmental Engineering, Sunchon National University, Suncheon 57922 (Korea, Republic of); Kim, Sang Chai [Department of Environmental Education, Mokpo National University, Muan 58554 (Korea, Republic of); Park, Young-Kwon, E-mail: catalica@uos.ac.kr [School of Environmental Engineering, University of Seoul, Seoul 02504 (Korea, Republic of)

    2016-10-15

    Highlights: • Al-MCM-48 was used for catalytic copyrolysis of particle board and polypropylene. • Catalytic produced mainly hydrocarbons. • The hydrocarbons produced were mainly in the diesel range. - Abstract: Particle board and polypropylene (PP) at a mixing ratio of 1:1 were copyrolyzed over two Al-MCM-48 catalysts with Si/Al ratios of 20 and 80. The catalyst characteristics were examined by measuring the Brunauer-Emmett-Teller surface area, temperature programmed desorption of ammonia, and X-ray diffraction. The main pyrolysis products of particle board were oxygenates, acids, and phenolics, whereas a large quantity of hydrocarbons within the diesel fuel range was produced from copyrolysis with polypropylene. The catalytic copyrolysis of particle board and PP over the Al-MCM-48 catalysts produced bio-oil with a much larger hydrocarbon content than that from the catalytic pyrolysis of particle board only. The hydrocarbons produced were mainly in the diesel range, highlighting the potential for the production of high-quality fuel.

  2. GEC Alsthom put their hopes on fuel cells: A large potential for PEM fuel cells; GEC Alsthom setzt auf Brennstoffzellen: Grosses Potential fuer PEMFC

    Energy Technology Data Exchange (ETDEWEB)

    Anon,

    1998-07-01

    Fuel cells are regarded as one of the high-tech products of great promise for future energy supplies. High hopes are pinned on the PEM technology: This low-temperature fuel cell is applicable to both mobile applications as well as stationary systems, i.e. from small-scale heating systems through to CHP systems of medium output. GEC Alsthom, who entered into a strategic alliance with Ballard, the market leader in fuel cell technology, are the first power plant manufacturers to commit themselves in the fuel cell market sector. (orig./CB) [Deutsch] Brennstoffzellen gelten als einer der Hoffnungstraeger der zukuenftigen Energieversorgung. Grosse Erwartungen werden dabei in die PEM-Technologie gesetzt: Diese Niedertemperatur-Brennstoffzelle eignet sich sowohl fuer mobile Anwendungen als auch fuer den stationaeren Einsatz - von kleinen Hausheizungsanlagen bis hin zu BHKW mittlerer Leistungklasse. Mit GEC Alsthom und seinem Engagement beim Marktfuehrer Ballard steigt nun erstmals ein grosser Energieanlagenhersteller verstaerkt in dieses Marktsegment ein. (orig.)

  3. An XRD technique for quantitative phase analysis of Al-U-Zr alloy

    International Nuclear Information System (INIS)

    Khan, K.B.; Kulkarni, N.K.; Jain, G.C.

    2003-01-01

    In several nuclear research reactors all over the world, Al-U alloy is used as fuel. To stabilise less brittle phase UAl 3 in Al-U alloy, a small amount of Zr (1 to 3 wt% ) is added. A rapid, non destructive and simple x-ray diffraction technique has been developed for quantitative phase analysis Al-U-Zr alloy system containing UAl 4 , UAl 3 and Al. (author)

  4. Reproduction of the RA reactor fuel element fabrication; Reprodukcija izrade gorivnog elementa za reaktor RA

    Energy Technology Data Exchange (ETDEWEB)

    Novakovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This document includes the following nine reports: Final report on task 08/12 - testing the Ra reactor fuel element; design concept for fabrication of RA reactor fuel element; investigation of the microstructure of the Ra reactor fuel element; Final report on task 08/13 producing binary alloys with Al, Mo, Zr, Nb and B additions; fabrication of U-Al alloy; final report on tasks 08/14 and 08/16; final report on task 08/32 diffusion bond between the fuel and the cladding of the Ra reactor fuel element; Final report on task 08/33, fabrication of the RA reactor fuel element cladding; and final report on task 08/36, diffusion of solid state metals. Ovaj rad sadrzi devet priloga: 1. Zavrsni izvestaj o podzadatku 08/12, ispitivanje elementa goriva reaktora RA; 2. Koncepcija izrade gorivnog elementa reaktora RA; 3. Ispitivanje mikrostrukture gorivnog elementa reaktora RA; 4. Zavrsni izvestaj o podzadatku 08/13, dobijanje binarnih legura urana sa legirajucim komponentama Al, Mo, Zr, Nb i B; 5. Dobijanje legure U-Al; 6. Zavrsni izvestaj o podzadacima 08/14 i 08/16; 7. Zavrsni izvestaj o podzadatku 08/32, difuziona veza goriva i kosuljice gorivnog elementa reaktora RA; 8. Zavrsni izvestaj o podzadatku 08/33, izrada kosuljice gorivnog elementa reaktora RA; 9. Zavrsni izvestaj o podzadatku 08/36, difuzija kod metala u cvrstom stanju.

  5. Fuel effect on solution combustion synthesis of Co(Cr,Al)2O4 pigments; Efecto del combustible en la síntesis de pigmentos Co(Cr,Al)2O4 por combustión de una disolución

    Energy Technology Data Exchange (ETDEWEB)

    Gilabert, J.; Palacios, M.D.; Sanz, V.; Mestre, S.

    2017-11-01

    The fuel effect on the synthesis of a ceramic pigment with a composition CoCr2−2ΨAl2ΨO4 (0≤Ψ≤1) by means of solution combustion synthesis process (SCS) has been studied. Three different fuels were selected to carry out the synthesis (urea, glycine and hexamethylentetramine (HMT)). Highly foamy pigments with very low density were obtained. Fd-3m spinel-type structure was obtained in all the experiments. Nevertheless, crystallinity and crystallite size of the spinels show significant differences with composition and fuel. The use of glycine along with the chromium-richest composition favours ion rearrangement to obtain the most ordered structure. Lattice parameter does not seem to be affected by fuel, although it evolves with Ψ according to Vegard's law. Colouring power in a transparent glaze shows important variations with composition. On the other hand, fuel effect presents a rather low influence since practically the same shades are obtained. However, it exerts certain effect on luminosity (L*). [Spanish] Se ha estudiado el efecto del combustible en la síntesis de pigmentos cerámicos tipo CoCr2-2ΨAl2ΨO4 (0≤Ψ≤1), obtenidos mediante síntesis por combustión de una disolución. Se seleccionaron 3 tipos de combustible diferentes: urea, glicina y hexametilentetramina. Todos los pigmentos obtenidos presentaron una textura altamente esponjosa y con muy baja densidad. Las estructuras cristalinas desarrolladas en todos los casos fueron tipo espinela Fd-3m. Sin embargo, tanto la cristalinidad como el tamaño de cristalito presentaron diferencias significativas dependiendo de la composición y del combustible utilizado. El uso de glicina, junto con las composiciones más ricas en cromo, favorece la reorganización de los iones para obtener estructuras más ordenadas y con mayor cristalinidad. El parámetro de red no parece verse afectado por el combustible, aunque sí evoluciona con Ψ de acuerdo con la Ley de Vegard. El poder colorante desarrollado

  6. Discharge behaviour of Mg-Al-Pb and Mg-Al-Pb-In alloys as anodes for Mg-air battery

    International Nuclear Information System (INIS)

    Wang, Naiguang; Wang, Richu; Peng, Chaoqun; Peng, Bing; Feng, Yan; Hu, Chengwang

    2014-01-01

    Highlights: • We investigate the effect of indium on the discharge behaviour of Mg-Al-Pb alloy. • We evaluate the performance of Mg-air batteries with Mg-Al-Pb and Mg-Al-Pb-In anodes. • We analyze the activation mechanism of Mg-Al-Pb-In alloy in the discharge process. - Abstract: The discharge behaviour of Mg-Al-Pb and Mg-Al-Pb-In alloys in 3.5 wt.% NaCl solution is investigated by electrochemical techniques, and compared with that of pure magnesium. The results show that Mg-Al-Pb-In alloy provides a more negative potential and exhibits a higher utilization efficiency in contrast with Mg-Al-Pb alloy and pure magnesium during the half-cell test at a large current density, and gives desirable discharge performance when used as anode for Mg- air battery. The peak power density of the Mg-air battery with Mg-Al-Pb-In anode is 94.5 mW cm −2 , which is comparable with those of Mg-H 2 O 2 semi-fuel batteries. Moreover, the activation mechanism of Mg-Al-Pb-In alloy during the discharge process is also analyzed

  7. Formulation and Testing of Paraffin-Based Solid Fuels Containing Energetic Additives for Hybrid Rockets

    Science.gov (United States)

    Larson, Daniel B.; Boyer, Eric; Wachs,Trevor; Kuo, Kenneth K.; Story, George

    2012-01-01

    Many approaches have been considered in an effort to improve the regression rate of solid fuels for hybrid rocket applications. One promising method is to use a fuel with a fast burning rate such as paraffin wax; however, additional performance increases to the fuel regression rate are necessary to make the fuel a viable candidate to replace current launch propulsion systems. The addition of energetic and/or nano-sized particles is one way to increase mass-burning rates of the solid fuels and increase the overall performance of the hybrid rocket motor.1,2 Several paraffin-based fuel grains with various energetic additives (e.g., lithium aluminum hydride (LiAlH4) have been cast in an attempt to improve regression rates. There are two major advantages to introducing LiAlH4 additive into the solid fuel matrix: 1) the increased characteristic velocity, 2) decreased dependency of Isp on oxidizer-to-fuel ratio. The testing and characterization of these solid-fuel grains have shown that continued work is necessary to eliminate unburned/unreacted fuel in downstream sections of the test apparatus.3 Changes to the fuel matrix include higher melting point wax and smaller energetic additive particles. The reduction in particle size through various methods can result in more homogeneous grain structure. The higher melting point wax can serve to reduce the melt-layer thickness, allowing the LiAlH4 particles to react closer to the burning surface, thus increasing the heat feedback rate and fuel regression rate. In addition to the formulation of LiAlH4 and paraffin wax solid-fuel grains, liquid additives of triethylaluminum and diisobutylaluminum hydride will be included in this study. Another promising fuel formulation consideration is to incorporate a small percentage of RDX as an additive to paraffin. A novel casting technique will be used by dissolving RDX in a solvent to crystallize the energetic additive. After dissolving the RDX in a solvent chosen for its compatibility

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

    Science.gov (United States)

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

    2015-09-01

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

  9. Annex 5 - Fabrication of U-Al alloy; Prilog 5 - Dobijanje legure U-Al

    Energy Technology Data Exchange (ETDEWEB)

    Drobnjak, Dj; Lazarevic, Dj; Mihajlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Alloy U-Al with low content of aluminium is often used for fabrication of fuel elements because it is stable under moderate neutron flux density. Additionally this type of alloys show much better characteristics than pure uranium under reactor operating conditions (temperature, mechanical load, corrosion effect of water). This report contains the analysis of the phase diagram of U-Al alloy with low content of aluminium, applied procedure for alloying and casting with detailed description of equipment. Characteristics of the obtained alloy are described and conclusions about the experiment and procedure are presented. Sistem U-Al sa niskim sadrzajem aluminijuma jedan je od cesto koriscenih za izradu gorivnih elemenata, jer je dovoljno stabilan pri umerenim gustinama fluksa. Pored toga, u uslovima karakteristicnim za rad nuklearnog reaktora (temperatura, gradijent temperature, mehanicka naprezanja, koroziono dejstvo vode) legure ovog sistema pokazuju daleko bolja svojstva od nelegiranog urana. Referat sadrzi analizu dijagrama stanja U-Al legure sa niskim sadrzajem aluminijuma, primenjeni postupak legiranja i livenja sa opisom pojedinih uredjaja i operacija. Takodje su opisana svojstva dobijene legure i dat je zakljucak o eksperimentu i tehnici rada (author)

  10. 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...

  11. Innovative nuclear fuels and applications. Part 1: limits of today's fuels and concepts for innovative fuels. Part 2: materials properties, irradiation performance and gaps in our knowledge

    International Nuclear Information System (INIS)

    Matzke, H.

    2000-01-01

    Part I of this contribution on innovative nuclear fuels gives a summary of current developments and problems of today's fuels, i.e. enriched UO 2 and UO 2 with a few % of PUO 2 (MOX fuel) or Gd 2 O 3 (as burnable neutron poison). The problems and property changes caused by high burnups (e.g. degradation of the thermal conductivity, polygonization or formation of the rim-structure) are discussed. Subsequently, the concepts for new fuels to burn excess Pu and to achieve an effective transmutation of the minor actinides Np, Am and Cm are treated. The criteria for the choice of suitable fuels and different fuel types (high Pu-content fuels, nitrides, U-free fuels, inert matrix supported fuels, cercers, cermets, etc.) are discussed. Part II of this contribution on innovative nuclear fuels deals with the properties of relevance of the different materials suggested to be used in innovative fuels which range from pure actinide fuel such as PuN and AmO 2 to spinel MgAl 2 O 4 and zircon ZrSiO 4 for inert matrix-based fuels, etc. The available knowledge on materials research aspects is summarized with emphasis on the physics of radiation damage. It is shown that significant gaps in the present knowledge exist, e.g. for the minor actinide compounds, and suggestions are made to fill these gaps in order to achieve a sufficient data base to design and operate suitable innovative fuels in a near future. (author)

  12. TMI defueling project fuel debris removal system

    International Nuclear Information System (INIS)

    Burdge, B.

    1992-01-01

    The three mile Island Unit 2 (TMI-2) pressurized water reactor loss-of-coolant accident on March 28, 1979, presented the nuclear community with many challenging remediation problems; most importantly, the removal of the fission products within the reactor containment vessel. To meet this removal problem, an air-lift system (ALS) can be used to employ compressed air to produce the motive force for transporting debris. Debris is separated from the transport stream by gravity separation. The entire method does not rely on any moving parts. Full-scale testing of the ALS at the Idaho National Engineering Laboratory (INEL) has demonstrated the capability of transporting fuel debris from beneath the LCSA into a standard fuel debris bucket at a minimum rate of 230 kg/min

  13. Premio Nobel de ciencias Económicas 1998

    Directory of Open Access Journals (Sweden)

    Sen Amartya

    1998-12-01

    Full Text Available

    La Academia Real de Ciencias de Suecia ha decidido otorgar el premio Banco de Suecia 1998 en Ciencias Económicas, en Memoria de Alfred Nobel, al Profesor Amartya Sen, del Trinity College, Reino Unido, y ciudadano indio, por sus contribuciones a la economía del bienestar.

     

     

     

  14. A study on density, porosity and grain size of unirradiated ROX fuels and simulated ROX fuels

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Shirasu, Noriko; Ohmichi, Toshihiko

    1999-05-01

    The unirradiated ROX fuels made of (1) 23wt%PuO 2 -17wt%SZR-56wt%Al 2 O 3 -4wt%MgO (ROX-SZR) and (2) 20wt%PuO 2 -29wt%ThO 2 -48wt%Al 2 O 3 -3wt%MgO (ROX-ThO 2 ) were fabricated by JAERI. Additionally, 9 simulated ROX fuels (used UO 2 instead of PuO 2 ) were fabricated by NFI. Densities, porosity and grain sizes of those were studied. Obtained results are: (1) ROX fuels: The estimated theoretical density (TD) was 5.6g/cc for the ROX-SZR and 6.2g/cc for the ROX-ThO 2 . Those were about half of that in UO 2 (10.96g/cc). As-fabricated density determined was 4.6g/cc (82%TD) for the former and 5.2g/cc (83%TD) for the latter. The %TD of those was low than that of UO 2 (95%TD). One of main reason is that the ROX fuels were sintered at temperature of 1,400degC while UO 2 was sintered at temperature of 1,700degC. The average pore diameter was about 3 μm and the porosity was 17-18%, implying that the ROX fuels were resulted in containing a lot of small pores 2 . As-fabricated density determined in the present study was about 4.5-5.5g/cc. The %TD of the simulated ROX fuels ranged about 94-98%TD. They were in the same as that of UO 2 because as high sintering temperature of 1,750degC as UO 2 . The average pore diameter was about 4-8 μm and the porosity was <6%. Grain size was revealed to be about 1-4 μm. (author)

  15. Contribuições de Vigotski para a educação de pessoas com deficiência visual Contribuciones de Vigotski para la educación de personas con deficiencia visual Vygotski's contributions for the education of visually disabled people

    Directory of Open Access Journals (Sweden)

    Adriano Henrique Nuernberg

    2008-06-01

    Full Text Available Entre as contribuições da psicologia histórico-cultural de Vigotski para a educação destacam-se aquelas que fazem referência às condições analisadas no contexto da defectologia. Esse artigo identifica algumas destas contribuições para a intervenção educacional junto a pessoas com deficiência. Para tanto, são revisados estudos fundamentados em Vigotski que apontam subsídios para a educação de pessoas com deficiência intelectual, surdez, surdocegueira, entre outras condições. Com vista a aprofundar os aportes vigotskianos para educação de pessoas com cegueira, analisa-se o conceito de sistemas psicológicos e seu valor para a reflexão sobre a intervenção educativa no contexto da deficiência visual. Conclui-se este estudo ressaltando-se que os aportes de Vigotski trazem à tona pistas concretas para a implementação de experiências educacionais que favoreçam a autonomia e a cidadania das pessoas com deficiência visual.Entre las contribuciones de la psicología histórico-cultural de Vigotski para la educación se destacan aquellas que hacen referencia a las condiciones analizadas en el contexto de la defectología. Ese artigo identifica algunas de estas contribuciones para la intervención educacional junto a personas con deficiencia. Para tanto, son revisados estudios fundamentados en Vigotski que apuntan subsidios para la educación de personas con deficiencia intelectual, sordez, sordo-ciego, entre otras condiciones. Con vista a profundizar los aportes vigotskianos para educación de personas con ceguera, se analiza el concepto de sistemas psicológicos y su valor para la reflexión sobre la intervención educativa en el contexto de la deficiencia visual. Se concluye este estudio resaltándose que los aportes de Vigotski traen al borde pistas concretas para la implementación de experiencias educacionales que favorezcan la autonomía y la ciudadanía de las personas con deficiencia visual.Among Vygotski

  16. Effectiveness of paper-structured catalyst for the operation of biodiesel-fueled solid oxide fuel cell

    Science.gov (United States)

    Quang-Tuyen, Tran; Kaida, Taku; Sakamoto, Mio; Sasaki, Kazunari; Shiratori, Yusuke

    2015-06-01

    Mg/Al-hydrotalcite (HDT)-dispersed paper-structured catalyst (PSC) was prepared by a simple paper-making process. The PSC exhibited excellent catalytic activity for the steam reforming of model biodiesel fuel (BDF), pure oleic acid methyl ester (oleic-FAME, C19H36O2) which is a mono-unsaturated component of practical BDFs. The PSC exhibited fuel conversion comparable to a pelletized catalyst material, here, conventional Ni-zirconia cermet anode for solid oxide fuel cell (SOFC) with less than one-hundredth Ni weight. Performance of electrolyte-supported cell connected with the PSC was evaluated in the feed of oleic-FAME, and stable operation was achieved. After 60 h test, coking was not observed in both SOFC anode and PSC.

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

    Directory of Open Access Journals (Sweden)

    Supardjo Supardjo

    2015-07-01

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

  18. Violencia familiar

    OpenAIRE

    Giraldo Arias, Regina; González Jaramillo, María Isabel; Beneroso, Julia; Betino, Lorena; Ceberio, Marcelo R.; De los Santos, Pía; Giovanazzi, Sandro; Hernández González, Eduardo; Inclan, Jaime; Laso, Esteban; Linares Fernández, Juan Luis; Morejón, Judith; Pereira Tercero, Roberto; Ramos, Ricardo; Ruiz López, Josefina

    2009-01-01

    Este libro reúne los pensamientos, las ideas, las propuestas y las experiencias en torno al tema de la violencia familiar de 20 autores sistémicos de habla hispana, entre ellos, algunos de los más reconocidos y destacados por sus contribuciones al paradigma en la posmodernidad. El contenido del texto, con los aportes de cada uno, ha sido cuidadosamente dispuesto por las editoras académicas, en una estructura que busca proporcionar a la complejidad del tema, un armonioso conjunto de los difere...

  19. Desempeño competitivo de Argentina y Uruguay en la leche en polvo

    OpenAIRE

    Depetris Guiguet, Edith; García Arancibia, Rodrigo; Rossini, Gustavo

    2009-01-01

    Los sectores lácteos de Argentina y Uruguay hacen importantes contribuciones tanto al producto interno bruto como a la generación de divisas provenientes de sus exportaciones. Al constituirse el MERCOSUR ambos países tenían excedentes exportables de leche en polvo entera y descremada, y aunque con similares condiciones del contexto externo impuestas por la integración y la evolución del mercado internacional, el comportamiento exportador no ha sido igual en las dos naciones. Esto lleva a cues...

  20. Ayer y hoy: El Capital financiero

    OpenAIRE

    Sergio Cabrera Morales

    2011-01-01

    Se pone de manifiesto las contribuciones realizadas en El Capital Financiero que ayudaron a entender las transformaciones del sistema económico capitalista de ese momento, ya no solo desde su perspectiva local y nacional sino de la economía mundial. Mutaciones que se centran en el desarrollo nacional y se expresan en el internacional. Pero al mismo tiempo se abordan algunas de las conclusiones en el campo de la política emprendida por R. Hilferding que al parecer van contracorriente a sus hal...

  1. Material characterization and corrosion control in wet storage of Chilean spent fuel

    International Nuclear Information System (INIS)

    Lamas, C.; Klein, J.; Escobar, I.

    2002-01-01

    Chile has two MTR type research reactors and the spent fuel will be stored in water previous to the conditioning for final disposal. One of the serious problem presented during wet storage is the phenomenon of corrosion, which depends on the water quality, the structural materials and the storage conditions. Thus, it is necessary to solve how to guarantee the integrity of the spent fuel during its wet storage. The water quality and fuel assembly materials are being characterized with the purpose to define the criteria of surveillance and control of corrosion as a function of time. The behavior of the 6061 Al and N4 Al alloys is being studied to characterize the susceptibility to pitting corrosion in solutions with chloride and cadmium as aggressive ions. The analyses were performed in a three-electrode electrochemical cell with 6061 Al and N4 Al as working electrodes. Platinum wire was the auxiliary electrode while Ag/AgCl was the reference electrode. To obtain the electrochemical characterization the polarization curves were used and the evolution of the corrosion potential of the aluminum alloys and SS 304 were measured. The electrolyte was deionized water with different concentrations of chloride and cadmium. At present, the results show that 6061 Al and N4 Al alloys are more susceptible to be attacked by pitting due to the presence of chloride than cadmium. (author)

  2. Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-26

    In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

  3. Study of the Ni-NiAl{sub 2}O{sub 4}-YSZ cermet for its possible application as an anode in solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Reyes-Rojas, A [Centro de Investigacion en Materiales Avanzados SC, Miguel de Cervantes 120, Complejo Industrial Chihuahua, Chihuahua (Mexico); Universidad Autonoma del Estado de Mexico, Facultad de Quimica, Toluca (Mexico); Esparza-Ponce, H E [Centro de Investigacion en Materiales Avanzados SC, Miguel de Cervantes 120, Complejo Industrial Chihuahua, Chihuahua (Mexico); Reyes-Gasga, J [Universidad Nacional Autonoma de Mexico, Instituto de Fisica, Mexico DF (Mexico)

    2006-05-17

    Nanocrystalline Ni-NiAl{sub 2}O{sub 4}-YSZ cermet with a possible application as anode in solid oxide fuel cells (SOFCs) has been developed. The powders were prepared by using an alternative solid-state method that includes the use of nickel acetylacetonate as an inorganic precursor to obtain a highly porous material after sintering at 1400 {sup o}C and oxide reduction (NiO -Al{sub 2}O{sub 3}-YSZ {yields} Ni-NiAl{sub 2}O{sub 4}-YSZ) at 800 {sup o}C for 8 h in a tubular reactor furnace using 10% H{sub 2}/N{sub 2}. Eight samples with 45% Ni and 55% Al{sub 2}O{sub 3}-YSZ in concentrations of Al{sub 2}O{sub 3} oxides from 10 to 80 wt% of were mixed to obtain the cermets. The obtained material was compressed using unidirectional axial pressing and calcinations from room temperature to 800 {sup o}C. Good results were registered using a heating rate of 1 {sup o}C min{sup -1} and a special ramp to avoid anode cracking. Thermal expansion, electrical conductivity, and structural characterization by thermo-mechanical analyser (TMA) techniques/methods, the four-point probe method for conductivity, scanning electron microscopy (SEM), x-ray energy dispersive spectroscopy (EDS), x-ray diffraction (XRD), and the Rietveld method were carried out. Cermets in the range 5.5 to 11% Al{sub 2}O{sub 3} present a crystal size around 200 nm. An inversion degree (I) in the NiAl{sub 2}O{sub 4} spinel structure of the cermets Ni-NiAl{sub 2}O{sub 4}-YSZ was found after the sintering and reduction processes. Good electrical conductivity and thermal expansion coefficient were obtained for the cermet with 12 wt% of spinel structure formation.

  4. Heat Generation Effects on U-Mo/Al through ABAQUS FEM Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Taewon; Jeong, Gwan Yoon; Lee, Cheol Min; Sohn Dongseong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    U-Mo/Al dispersion fuels have been considered a most promising candidate for a replacement of Highly Enriched Uranium (HEU) fuel in many research reactors. Coulson developed a FEM model which show the fuel meat realistically and compared the thermal conductivity results of two and three dimensional model. Williams also developed a FEM model which are different from the former in that it use regularly meshed unit cells. He showed a heat generation effects through FEM simulation and the effective thermal conductivity of the fuel with heat generated in the fuel particles is a little lower than that of the fuel with no heat generated. In the current work, the heat generation effects are analyzed and discussed in a wider range of volume fraction with more realistic models by using ABAQUS finite element package. The FEM model is used to determine the effective thermal conductivity of U-Mo/Al and to simulate the heat generation effects in the study. This model reflected the microscopic morphology of the fuel very well by making random distribution particles although the particle shape is considered as sphere. All simulation results show the heat generation effects although the effects are small when the volume fraction of fuels are high. When the particles are surrounded with interaction layers, the heat transfer from the particle to matrix is disturbed by interaction layers due to the low thermal conductivity of interaction layers. However this effects decreases when the sum of the volume fraction of fuels and interaction layers exceeds 40-50 vol% because a great portion of the heat must pass through fuels and interaction layers although the heat is applied on the surface. Therefore particle size and initial particle volume fractions will be the important factors for the heat generation effects when interaction layers grow during irradiations.

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  6. Waste to Watts and Water: Enabling Self-Contained Facilities Using Microbial Fuel Cells

    Science.gov (United States)

    2009-03-01

    98; “Project to Turn Beer Wastewater into Power,” ; Yokoyama et al., “Treatment of Cow-Waste Slurry,” 634; Catal et al., “Electricity Production...Fuel Cells Bulletin 2006, no. 7 (2006): 7. “Project to Turn Beer Wastewater into Power.” Fuel Cells Bulletin 2007, no. 7 (2007): 11. Rabaey, K., J...Biomass Fermentation , edited by Piet Lens, Peter Westermann, Marianne Haberbauer, and Angelo Moreno, 377–400. Integrated Environmental Technology Series

  7. Qualification process of dispersion fuels in the IEAR1 research reactor

    International Nuclear Information System (INIS)

    Domingos, D.B.; Silva, A.T.; Silva, J.E.R.

    2010-01-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a miniplate irradiation device (MID) to be placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 -Al dispersion fuels, LEU type (19,9% of 235 U) with uranium densities of, respectively, 3.0 gU/cm 3 and 4.8 gU/cm 3 . The fuel miniplates will be irradiated to nominal 235 U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and TWODB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer code LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. This paper also presents a system designed for fuel swelling evaluation. The determination of the fuel swelling will be performed by means of the fuel miniplate thickness measurements along the irradiation time. (author)

  8. The development and localization of nuclear fuel technology for KMRR

    International Nuclear Information System (INIS)

    Kim, Seong Yun; Lee, Ji Bok; Suk, Ho Chun; Kuk, Il Hyun; Hwang, Woan; Kim, Bong Goo; Park, Joo Hwan; Kim, Young Jin; Kang, Thae Khapp; Lee, Jae Choon

    1988-05-01

    This project was implemented aiming at localizing the fabrication of the KMRR fuel by october 1993. The contents of this project were divided into three parts: fuel design, fuel fabrication and process criticality analysis. In the fuel design, the radial power distribution in the fuel core was modeled and formulated taking account of the neutron flux depression in the radial direction. It was also performed to model and formulate the thermal characteristics such as the thermal conductivity and specific heat of the fuel core, U3Si-Al, the swelling and the film coefficient of heat transfer between the aluminum clad and light water coolant. The two dimensional heat transfer in the finned fuel element was equated based on the general equation governing the heat transfer in materials in order to develope a computer code, TEMP2D. TEMP2D solves finite differenced equations to calculate a two dimensional fuel temperature distribution under the steady and transient states. In the fuel fabrication, the technologies of fabricating uranium silicide fuel meat were tried by using depleted uranium as a raw material. These were extended to find the problems in technologies and to establish the ways of approach. The end product, so called fuel meat, was a metallic powder compound, U3Six(1≤x≤2), dispersed in Al matrix. The fuel meat was fabricated by the horizontal extrusion technique, and powder extrusion technique. Fabrication technologies comprise five different continuous processes: melting and casting of metallic uranium with silicon and aluminum, heat treatment, chipping and crushing, pulverizing, and extrusion. In the process criticality analysis, AMPX-KENO benchmark calculation was performed and calculational error of AMPX-KENO system was established. (Author)

  9. Three-Dimensional FIB/EBSD Characterization of Irradiated HfAl3-Al Composite

    Energy Technology Data Exchange (ETDEWEB)

    Hua, Zilong; Guillen, Donna Post; Harris, William; Ban, Heng

    2016-09-01

    A thermal neutron absorbing material, comprised of 28.4 vol% HfAl3 in an Al matrix, was developed to serve as a conductively cooled thermal neutron filter to enable fast flux materials and fuels testing in a pressurized water reactor. In order to observe the microstructural change of the HfAl3-Al composite due to neutron irradiation, an EBSD-FIB characterization approach is developed and presented in this paper. Using the focused ion beam (FIB), the sample was fabricated to 25µm × 25µm × 20 µm and mounted on the grid. A series of operations were carried out repetitively on the sample top surface to prepare it for scanning electron microscopy (SEM). First, a ~100-nm layer was removed by high voltage FIB milling. Then, several cleaning passes were performed on the newly exposed surface using low voltage FIB milling to improve the SEM image quality. Last, the surface was scanned by Electron Backscattering Diffraction (EBSD) to obtain the two-dimensional image. After 50 to 100 two-dimensional images were collected, the images were stacked to reconstruct a three-dimensional model using DREAM.3D software. Two such reconstructed three-dimensional models were obtained from samples of the original and post-irradiation HfAl3-Al composite respectively, from which the most significant microstructural change caused by neutron irradiation apparently is the size reduction of both HfAl3 and Al grains. The possible reason is the thermal expansion and related thermal strain from the thermal neutron absorption. This technique can be applied to three-dimensional microstructure characterization of irradiated materials.

  10. Application of nondestructive methods for qualification of high density fuels in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Silva, Jose E.R.; Silva, Antonio T.; Domingos, Douglas B.; Terremoto, Luis A.A.

    2011-01-01

    IPEN/CNEN-SP manufactures fuels to be used in its research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil still does not 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 (U 3 O 8 and U 3 Si 2 dispersed in Al matrix) internationally tested and qualified to be used in research reactors, and has attained 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 at IPEN/CNEN-SP to increase the uranium density of the fuels. Ten fuel miniplates (five containing U 3 O 8 -Al and five containing U 3 Si 2 -Al), with densities of 3.2 gU/cm 3 and 4.8 gU/cm 3 respectively, are being irradiated inside an irradiation device placed in a peripheral position of the IEA-R1 core. Non-destructive methods will be used to evaluate irradiation performance of the fuel miniplates after successive cycles of irradiation, by means: monitoring the reactor parameters during operation; periodic underwater visual inspection of fuel miniplates, eventual sipping test for fuel miniplates suspected of leakage and underwater measuring of the miniplate thickness for assessment of the fuel miniplate swelling. (author)

  11. Fabrication of high-uranium-loaded U/sub 3/O/sub 8/-Al developmental fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Copeland, G.L.; Martin, M.M.

    1980-12-01

    A common plate-type fuel for research and test reactors is U/sub 3/O/sub 8/ dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the /sup 235/U enrichment from above 90% to below 20% for these fuels to lessen the risk of diversion of the uranium for nonpeaceful uses. Thus, the uranium content of the fuel plates has to be increased to maintain the performance of the reactors. This paper describes work at ORNL to determine the maximal uranium loading for these fuels that can be fabricated with commercially proven materials and techniques and that can be expected to perform satisfactorily in service.

  12. Thermal-hydraulic calculations for KUHFR with reduced enrichment uranium fuel

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Shibata, Toshikazu.

    1982-01-01

    This report provides the preliminary results of the thermal-hydraulic calculations to study the safety aspects in fueling the KUHFR with reduced enrichment uranium. The calculations were based on what was outlined in the Safety Analysis Report for the KUHFR and the guidebook for research reactor core conversion, IAEA-TECDOC-233, published by the International Atomic Energy Agency. No significant differences in the thermal-hydraulic operating conditions have been found between HEU and MEU fuels. However, in LEU cases, the combination of three factors - larger power peaking with LEU fuel, smaller thermal conductivity of U 3 O 8 -Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)

  13. CFD Analysis of Hot Spot Fuel Temperature in the Control Fuel Block Assembly of a VHTR core

    International Nuclear Information System (INIS)

    Kim, Min Hwan; Tak, Nam Il; Noh, Jae Man

    2010-01-01

    The Very High Temperature Reactor (VHTR) dedicated for efficient hydrogen production requires core outlet temperatures of more than 950 .deg. C. As the outlet temperature increases, the thermal margin of the core decreases, which highlights the need for a detailed analysis to reduce its uncertainty. Tak et al. performed CFD analysis for a 1/12 fuel assembly model and compared the result with a simple unit-cell model in order to emphasize the need of a detailed CFD analysis for the prediction of hot spot fuel temperatures. Their CFD model, however, was focused on the standard fuel assembly but not on the control fuel assembly in which a considerable amount of bypass flow is expected to occur through the control rod passages. In this study, a CFD model for the control fuel block assembly is developed and applied for the hot spot analyses of PMR200 core. Not only the bypass flow but also the cross flow is considered in the analyses

  14. 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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Directory of Open Access Journals (Sweden)

    YEON SOO KIM

    2013-12-01

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

  17. Annex 5 - Fabrication of U-Al alloy

    International Nuclear Information System (INIS)

    Drobnjak, Dj.; Lazarevic, Dj.; Mihajlovic, A.

    1961-01-01

    Alloy U-Al with low content of aluminium is often used for fabrication of fuel elements because it is stable under moderate neutron flux density. Additionally this type of alloys show much better characteristics than pure uranium under reactor operating conditions (temperature, mechanical load, corrosion effect of water). This report contains the analysis of the phase diagram of U-Al alloy with low content of aluminium, applied procedure for alloying and casting with detailed description of equipment. Characteristics of the obtained alloy are described and conclusions about the experiment and procedure are presented [sr

  18. Contribuciones al conocimiento del género Phragmotheca cuatr. 1 (Bombacaceae-Quararibeae).

    OpenAIRE

    Fernández Alonso, José Luis

    1996-01-01

    Se comenta la posicidn sistemática de Phragmotheca en relacidn con la tribu Quararibeae y se incluye una clave para la separación de los géneros de bombacáceas de hoja entera o unifoliolada del norte de Suramdrica. Se revisa la utilidad'de algunos caracteres en la separación de las especies y se presenta una clave diagnbstica. Se describen e ilustran 4 nuevas especies y dos taxones infraespeclficos P. rubriflora, P. lemniscata, P. hydra, P. siderosa subsp. megacarpa, F! amazonl...

  19. Contribuciones al conocimiento de la flora del País Vasco, II

    Directory of Open Access Journals (Sweden)

    AIZPURU, I., APERRIBAY, J.A., ASEGINOLAZA, C., GARIN, F., VIVANT, J.

    1997-01-01

    Cuscuta scandens Brot. subsp. cesatiana (BERTOL. Soó, Fallopia sachalinensis (FRIEDERICH SCHMIDT PETROP. Ronse Decraene, Galium debile Desv., Geum hispidum Fries y Myosurus minimus L.; se combina, además, Apium graveolens L. subsp. butronensis (D. GOMEZ and G. MONTSERRAT I. AIZPURU, stat. nov.

  20. Optimization of parameters in the simulation of the interdiffusion layer growth in Al-U couples

    International Nuclear Information System (INIS)

    Kniznik, Laura; Alonso, Paula R.; Gargano, Pablo H.; Rubiolo, Gerardo H.

    2009-01-01

    U-Mo alloy dispersed in aluminum is considered as a high U density fuel for research reactors. In and out of pile experiments showed a reaction layer in U-Mo/Al interphase with formation of intermetallics compounds: Al 2 U, Al 3 U and Al 4 U. Under irradiation, porosities originate an unacceptable swelling of the fuel plate. The kinetics of growth of the intermetallic compounds in the U-Mo/Al interphase is treated in the Al 3 U/Al couple as a planar moving boundary problem due to diffusion of Al and U atoms in the direction perpendicular to the interphase surface. Using data from literature, we built a thermodynamic database to be read by the Thermocalc code to calculate phase equilibria. The diffusion problem was carried out by the DICTRA simulation package which articulates data evaluated by Thermocalc with a mobility database. In a previous work we built preliminary databases, for both free energy and mobilities. In the present work, we adjust the parameters from experimental thermodynamic equilibria and concentration profiles existing in literature, and we simulate satisfactorily the growth of the Al 4 U phase. (author)

  1. The possible use of cermet fuel in the DIDO and PLUTO heavy-water research reactors

    International Nuclear Information System (INIS)

    Kennedy, T.D.A.

    1981-08-01

    As part of a study of the feasibility of using low-enrichment fuels in DIDO and PLUTO reactors the heat transfer and safety aspects involved in replacing the present U/AL-alloy (75% w/w U 235 ) fuel plates with U/AL-cermet (20% w/w U 235 ) plates, having the same outside dimensions to retain the same hydraulic characteristics, have been investigated. (U.K.)

  2. Irradiation testing of miniature fuel plates for the RERTR program

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R L; Martin, M M [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)

    1983-08-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% {sup 235}U in place of highly enriched fuel for these reactors would reduce the potential for {sup 235}U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminum-base dispersion-type fuel plates with fuel cores of 1) high uranium content U{sup 3}){sup 8}-Al being developed by ORNL, 2) high uranium content UAI{sub x}-Al being developed by EG and G Idaho, Inc., and 3) very high uranium content U{sub 3}Si-Al- being developed by ANL. The miniplates are 115-mm long by 50-mm wide with overall plate thicknesses of 1.27 or 1.52 mm. The fuel core dimensions vary according to overall plate thicknesses with a minimal clad thickness requirement of 0.20 mm. Sixty such miniplates (thirty of each thickness) can be irradiated in one test facility. The irradiation test facility, designated as HFED-1 is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The peak neutron flux measured for this experiment is 1.96 x 10{sup 18} neutrons m{sub -2} s{sub -1}. The various types of miniplates will achieve burnups of up to approximately 2.2x10{sup 27} fissions/m{sup 3} of fuel, which will require approximately eight full power months of irradiation. During reactor shutdown periods, the experiment is removed from the reactor, moved to a special poolside station, disassembled, and inspected

  3. Determination of the agricultural and process-related boundary conditions for the use of rapeseed oil and its refinery products as motor fuel. Final report; Ermittlung der landwirtschaftlichen, prozesstechnischen und verfahrenstechnischen Rahmenbedingungen fuer die Verwendung von Rapsoel und seiner Umwandlungsprodukte als Kraftstoff. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Schliephake, D; Hacker, C M

    1994-11-01

    The use of rapeseed oil as a motor fuel is viewed from various angles: Production aspects; environmental aspects; marketing aspects in agriculture; business management aspects in the Federal Republic of Germany; production and refining of rapeseed oil; use of rapeseed oil methyl ester as diesel fuel and in mineral oil processing plants; engine performance. (orig./SR) [Deutsch] In dem vorliegenden Bericht wird der Einsatz von Rapsoel als Kraftstoff aus verschiedenen Blickwinkeln betrachtet: Produktionsaspekte, Umweltaspekte, marktwirtschaftliche Aspekte in der Landwirtschaft; betriebswirtschaftliche Aspekte in der Bundesrepublik Deutschland; Herstellung und Raffination von Rapsoel; Einsatz von Rapsoelmethylester als Dieselkraftstoff und in Anlagen der Mineraloelverarbeitung; Motortechnische Untersuchungsergebnisse. (orig./SR)

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

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Cirila Tacconi de

    2005-07-01

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

  5. Proceedings of the Start-up Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 28-29 April 2014, OECD-NEA HQ

    International Nuclear Information System (INIS)

    Kurata, Masaki; Bragg-Sitton, Shannon; Pasamehmetoglu, K.; Sowder, Andrew; Koo, Yang-Hyun; Yang, Jae-Ho; Kim, Hyun-Gil; Zhou, Y.; Forgeron, T.; Guedeney, Ph.; Brachet, J.C.; Michaux, A.; Chauvin, Nathalie; Waeckel, N.; Ambard, A.; Blanpain, P.; Bischoff, J.; Zvonarev, Yu.; Verwerft, M.; Weber, M.; Lambrinou, K.; Koonen, E.; Van Dyck, S.; PETIT, Marc; Cornet, Stephanie; ); YAMAJI, Akifumi; ); Inozemtsev, V.; )

    2014-04-01

    Expert Group meeting on ATF for LWR's (M. Verwerft, SCK-CEN); 4 - Potential links with other related activities: - Recent and Current Activities of the Working Group on Fuel Safety - NEA/CSNI (M. Petit, IRSN); - Nuclear Science Committee Working Party on Scientific Issues of the Fuel Cycle, Activities on Innovative Fuels - EGIF (S. Massara for S. Cornet, OECD-NEA); - NSC/WPRS Activities on Reactor Fuel Performance (EGRFP) and IFPE database (J. Gulliford for A. Yamaji, OECD-NEA); - IAEA Coordinated Research Projects (CRP) in the area of Fuel Engineering and ATF plans (V. Inozemtsev, IAEA); 5 - Collaborative framework: Outcome of the discussion on Task Forces to be constituted within this Expert Group and Preliminary list of actions agreed during the meeting. 6 - Documents shared during the meeting by S. Bragg-Sitton (INL): - Advanced Fuels Campaign, Light Water Reactor Accident Tolerant Fuel Performance Metrics (DoE Report); - Development of advanced accident tolerant fuels for commercial LWRs (S. Bragg-Sitton, Nuclear News, March 2014, p. 83-91); 7 - Documents shared during the meeting by K. Terrani (ORNL) from a special issue of 'Journal of Nuclear Materials' on Accident Tolerant Fuels for LWRs (no. 448, 2014): - Forward for special JNM issue on accident tolerant fuels for LWRs (Preface); - Accident tolerant fuels for LWRs: A perspective (S.J. Zinkle et al.); - Silicon carbide composite for light water reactor fuel assembly applications (Ken Yueh, Kurt A. Terrani); - Stability of SiC-matrix microencapsulated fuel constituents at relevant LWR conditions (L.L. Snead et al.); - Preparation of UC 0.07-0.10 N 0.90-0.93 spheres for TRISO coated fuel particles (R.D. Hunt et al.); - Carbothermic synthesis of 820 lm uranium nitride kernels: Literature review, thermodynamics, analysis, and related experiments (T.B. Lindemer et al.); - Fission product release and survivability of UN-kernel LWR TRISO fuel (T.M. Besmann et al.); - Advanced oxidation-resistant iron

  6. Core-Shell Al-Polytetrafluoroethylene (PTFE) Configurations to Enhance Reaction Kinetics and Energy Performance for Nanoenergetic Materials.

    Science.gov (United States)

    Wang, Jun; Qiao, Zhiqiang; Yang, Yuntao; Shen, Jinpeng; Long, Zhang; Li, Zhaoqian; Cui, Xudong; Yang, Guangcheng

    2016-01-04

    The energy performance of solid energetic materials (Al, Mg, etc.) is typically restricted by a natural passivation layer and the diffusion-limited kinetics between the oxidizer and the metal. In this work, we use polytetrafluoroethylene (PTFE) as the fluorine carrier and the shielding layer to construct a new type of nano-Al based fuels. The PTFE shell not only prevents nano-Al layers from oxidation, but also assists in enhancing the reaction kinetics, greatly improving the stability and reactivity of fuels. An in situ chemical vapor deposition combined with the electrical explosion of wires (EEW) method is used to fabricate core-shell nanostructures. Studies show that by controlling the stoichiometric ratio of the precursors, the morphology of the PTFE shell and the energy performance can be easily tuned. The resultant composites exhibit superior energy output characters than that of their physically mixed Al/PTFE counterparts. This synthetic strategy might provide a general approach to prepare other high-energy fuels (Mg, Si). © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  7. Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

    Energy Technology Data Exchange (ETDEWEB)

    Ott, Larry J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristics are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate

  8. Conversion and start up of Tehran Research Reactor with LEU fuel

    International Nuclear Information System (INIS)

    Zaker, M.

    2004-01-01

    The MW Tehran Research Reactor, Highly Enriched Uranium (HEU) fuel has been converted to Low Enriched Uranium (LEU) fuel using U 3 0 8 -Al with less than 20% enriched uranium. Measured value of excess reactivity, control rod worth and other parameters indicate good agreement with computational predictions. (author)

  9. Release of fission products from irradiated aluminide fuel at high temperature

    International Nuclear Information System (INIS)

    Shibata, Toshikazu; Kanda, Keiji; Mishima, Kaichiro; Tamai, Tadaharu; Hayashi, Masatoshi; Snelgrove, James L.; Stahl, David; Matos, James E.; Travelli, Armando; Case, F. Neil; Posey, John C.

    1983-01-01

    Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel cladding material. The release of fission products from the fuel plate at temperature below 500 deg. C was found negligible. The first rapid release of fission products was observed with the occurrence of blistering at 561±1 deg. C on the plates. The next release at 585. C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 deg. C of U-Al x . The released material was mostly xenon, but small amounts of iodine and cesium were observed. (author)

  10. Release of fission products from irradiated aluminide fuel at high temperature

    International Nuclear Information System (INIS)

    Shibata, T.; Kanda, K.; Mishima, K.

    1982-01-01

    Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel-cladding material. The release of fission products from the fuel plate at temperature below 500 0 C was found negligible. The firist rapid release of fission products was observed with the occurrence of blistering at 561 +- 1 0 C on the plates. The next release at 585 0 C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 0 C of U-Al/sub x/. The released material was mostly xenon, but small amounts of iodine and cesium were observed

  11. 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

  12. Conflictos, treguas y amnistías en el mundo antiguo

    OpenAIRE

    Iriarte Goñi, Ana

    2002-01-01

    Introducción a las contribuciones presentadas al curso "Conflictos, treguas y amnistías en el mundo antiguo", desarrollado en agosto de 1999 en el marco de los XVIII Cursos de Verano de la UPV/EHU en Donostia. [ES] El dossier «Conflictos, treguas y amnistías en el mundo antiguo» supone un recorrido histórico a través del mundo griego arcaico y clásico, y del mundo romano tardorrepublicano y augusteo cuyo hilo conductor es el conflicto político y social en el mundo grecorromano. Al...

  13. IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

    Directory of Open Access Journals (Sweden)

    M.K. MEYER

    2014-04-01

    Full Text Available High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  14. Irradiation performance of U-Mo monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, M. K.; Gan, J.; Jue, J. F.; Keiser, D. D.; Perez, E.; Robinson, A.; Wachs, D. M.; Woolstenhulme, N. [Idaho National Laboratory, Idaho (Korea, Republic of); Kim, Y.S.; Hofman, G. L. [Argonne National Laboratory, Lemont (United States)

    2014-04-15

    High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  15. Improvement of the Oxidation Resistance of CoNiCrAlY Bond Coats Sprayed by High Velocity Oxygen-Fuel onto Nickel Superalloy Substrate

    Directory of Open Access Journals (Sweden)

    Alessio Fossati

    2010-11-01

    Full Text Available CoNiCrAlY powders with similar granulometry and chemical composition, but different starting reactivity toward oxygen, were sprayed onto superalloy substrates by High Velocity Oxygen-Fuel producing coatings of similar thicknesses. After spraying, samples were maintained at 1,273 K in air for different test periods of up to 5,000 hours. Morphological, microstructural, compositional and electrochemical analyses were performed on the coated samples in order to assess the high temperature oxidation resistance provided by the two different powders. The powder with higher starting reactivity towards oxygen improves the oxidation resistance of the coated samples by producing thinner and more adherent thermally grown oxide layers.

  16. Radioactive Release from Aluminum-Based Spent Nuclear Fuel in Basin Storage

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.

    1999-10-21

    The report provides an evaluation of: (1) the release rate of radionuclides through minor cladding penetrations (breaches) on aluminum-based spent nuclear fuel (AL SNF), and (2) the consequences of direct storage of breached AL SNF relative to the authorization basis for SRS basin operation.

  17. Radioactive Release from Aluminum-Based Spent Nuclear Fuel in Basin Storage

    International Nuclear Information System (INIS)

    Sindelar, R.L.

    1999-01-01

    The report provides an evaluation of: (1) the release rate of radionuclides through minor cladding penetrations (breaches) on aluminum-based spent nuclear fuel (AL SNF), and (2) the consequences of direct storage of breached AL SNF relative to the authorization basis for SRS basin operation

  18. Application of nondestructive methods for qualification of high density fuels in the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jose E.R.; Silva, Antonio T.; Domingos, Douglas B.; Terremoto, Luis A.A., E-mail: jersilva@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN/CNEN-SP manufactures fuels to be used in its research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil still does not 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 (U{sub 3}O{sub 8} and U{sub 3}Si{sub 2} dispersed in Al matrix) internationally tested and qualified to be used in research reactors, and has attained 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 at IPEN/CNEN-SP to increase the uranium density of the fuels. Ten fuel miniplates (five containing U{sub 3}O{sub 8}-Al and five containing U{sub 3}Si{sub 2}-Al), with densities of 3.2 gU/cm{sup 3} and 4.8 gU/cm{sup 3} respectively, are being irradiated inside an irradiation device placed in a peripheral position of the IEA-R1 core. Non-destructive methods will be used to evaluate irradiation performance of the fuel miniplates after successive cycles of irradiation, by means: monitoring the reactor parameters during operation; periodic underwater visual inspection of fuel miniplates, eventual sipping test for fuel miniplates suspected of leakage and underwater measuring of the miniplate thickness for assessment of the fuel miniplate swelling. (author)

  19. Corrosion analysis of AlMg2 and AlMgSi using electrochemical method

    International Nuclear Information System (INIS)

    Dian A; Maman Kartaman; Rosika K; Yanlinastuti

    2014-01-01

    Corrosion test of cladding materials and structures of research reactor fuel, AlMgSi and AlMg2 have been performed in demineralized water of pH 2 and 6.7 using an electrochemical method. Corrosion phenomenon is affected by several factor such as composition and condition of solution. The purpose of this activity is to investigate the corrosion phenomena through the determination of the parameters of corrosion and polarization curve. The materials used are AlMg2 and AlMgSi alloy in circular dish shape with an area of 1 Cm"2. Preparation of the test sample is performed through several stages polishing, cleaning and drying procedures followed ASTM G3. The electrochemical method is done by measuring the open circuit potential (OCP), polarization resistance and potentiodynamic in demineralized water of pH 2 and pH 6.7 at temperature of 25°C. The results of the OCP is the corrosion potential (Ecorr) of AlMg2 and AlMgSi each of -906.1 mV and -619.8 mV at pH 2 and -868.6 and -756.7 mV at pH 6.7 mV. The results of measurements by polarization resistance technique showed that the corrosion rate of AlMg2 and AlMgSi in safe category (<2 mpy) at pH 6.7 and at pH 2 corrosion rate increased significantly, but still in the lightweight category (<20 mpy). Potentiodynamic curves showed that the passivation at pH 6.7 is very low while the passivation at pH 2 occurs within a relatively short range potential and followed events corroded. (author)

  20. Development of the fabrication technology of the simulated DUPIC fuel

    International Nuclear Information System (INIS)

    Kang, Kweon Ho; Yang, M. S.; Bae, K. K. and others

    2000-06-01

    It is important to get basic data to analysis physical properties, behavior in reactor and performance of the DUPIC fuel because physical properties of the DUPIC fuel is different from the commercial UO 2 fuel. But what directly measures physical properties et al. of DUPIC fuel being resinterred simulated spent fuel through OREOX process is very difficult in laboratory owing to its high level radiation. Then fabrication of simulated DUPIC fuel is needed to measure its properties. In this study, processes on powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using simulated spent fuel are discribed. To fabricate simulated DUPIC fuel, the powder from 3 times OREOX and 5 times attrition milling simulated spent fuel is compacted with 1.3 ton/cm 2 . Pellets are sintered in 100% H 2 atmosphere over 10 h at 1800 deg C. Sintered densities of pellets are 10.2-10.5 g/cm 3

  1. 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)

  2. A Prediction Study of Aluminum Alloy Oxidation of the Fuel Cladding in Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tahk, Y. W.; Oh, J. Y.; Lee, B. H.; Seo, C. G.; Chae, H. T.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    U{sub 3}Si{sub 2}-Al dispersion fuel with Al cladding will be used for Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding experiences the oxidation layer growth on the surface during the reactor operation. The formation of oxides on the cladding affects fuel performance by increasing fuel temperature. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, a fresh fuel is discharged after 900 effective full power days (EFPD) with 18 cycles of 50 days loading. For the proper prediction of the aluminum oxide thickness of fuel cladding during the long residence time, a reliable model is needed. In this work, several oxide thickness prediction models are compared with the measured data from in-pile test by RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model are performed for JRTR fuel

  3. In situ metal ion contamination and the effects on proton exchange membrane fuel cell performance

    Science.gov (United States)

    Sulek, Mark; Adams, Jim; Kaberline, Steve; Ricketts, Mark; Waldecker, James R.

    Automotive fuel cell technology has made considerable progress, and hydrogen fuel cell vehicles are regarded as a possible long-term solution to reduce carbon dioxide emissions, reduce fossil fuel dependency and increase energy efficiency. Even though great strides have been made, durability is still an issue. One key challenge is controlling MEA contamination. Metal ion contamination within the membrane and the effects on fuel cell performance were investigated. Given the possible benefits of using stainless steel or aluminum for balance-of-plant components or bipolar plates, cations of Al, Fe, Ni and Cr were studied. Membranes were immersed in metal sulfide solutions of varying concentration and then assembled into fuel cell MEAs tested in situ. The ranking of the four transition metals tested in terms of the greatest reduction in fuel cell performance was: Al 3+ ≫ Fe 2+ > Ni 2+, Cr 3+. For iron-contaminated membranes, no change in cell performance was detected until the membrane conductivity loss was greater than approximately 15%.

  4. Major results on the development of high density U-Mo fuel and pin-type fuel elements executed under the Russian RERTR program and in cooperation with ANL (USA)

    International Nuclear Information System (INIS)

    Vatulin, A.; Morozov, A.; Stetsky, Y.; Suprun, V.; Dobrikova, I.; Trifonov, Y.; Mishunin, V.; Sorokin, V.

    2003-01-01

    VNIINM is active participant of 'Russian program on Reduced Enrichment for Research and Test Reactors'. Institute Works in two main directions: 1) development of new high-density fuels (HDF) and 2) development of new design of fuel elements with LEU. The development of the new type fuel element is carried out both for existing reactors, and for developing new advanced reactors. The 'TVEL' concern is coordinator of works of this program. The majority enterprises of branch (NIIAR, PIYaF, RRC KI, NZChK) take part in this work. Since 2000 these works are being conducted in cooperation with Argonne National Laboratory (USA) within the RERTR program under VNIINM with ANL contract. At the present, a large set of pre-pile investigations has been completed. All necessary fabrication procedures have been developed for utilization of U-Mo dispersion fuel in Russian-designed research reactors. For irradiation tests the pin-type mini-fuel elements with HDF dispersion fuel with LEU and the uranium density equaled to 4,0 and 6,0 g/cm 3 (up to 40 vol.%) have been manufactured. Their irradiation began in August 2003 in the MIR reactor (NIIAR, Dimitrovgrad). A large set of works for preparation of lifetime tests (WWR-M reactor in Gatchina) of two full-scale fuel assemblies with new pin-type fuel elements on basis LEU UO 2 -Al and UMo-Al fuels has been completed. The in-pile tests of fuel assemblies began in September 2003. The summary of important results of performed works and their near-term future are presented in paper. (author)

  5. Catalytic Reactor for Inerting of Aircraft Fuel Tanks

    Science.gov (United States)

    1974-06-01

    Aluminum Panels After Triphase Corrosion Test 79 35 Inerting System Flows in Various Flight Modes 82 36 High Flow Reactor Parametric Data 84 37 System...AD/A-000 939 CATALYTIC REACTOR FOR INERTING OF AIRCRAFT FUEL TANKS George H. McDonald, et al AiResearch Manufacturing Company Prepared for: Air Force...190th Street 2b. GROUP Torrance, California .. REPORT TITLE CATALYTIC REACTOR FOR INERTING OF AIRCRAFT FUEL TANKS . OESCRIP TIVE NOTEs (Thpe of refpoft

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

    International Nuclear Information System (INIS)

    Peacock, H.B.

    1984-01-01

    Research and test reactors are presently operated with aluminum-clad fuel elements containing highly enriched uranium-aluminum alloy cores. To lower the enrichment and still maintain reactivity, the uranium content of the fuel element will need to be higher than currently achievable with alloy fuels. This will necessitate conversion to other forms such as U 3 O 8 -aluminum cermets. Above the aluminum melting point, U 3 O 8 and aluminum undergo an exothermic thermite reaction and cermet fuel cores tend to keep their original shape. Both factors could affect the course and consequences of a reactor accident, and therefore prompted an investigation of the behavior of cermet fuels at elevated temperatures. Tests were carried out using pellets and extruded tube sections with 53 wt % U 3 O 8 in aluminum. This content corresponds to a theoretical uranium density of 1.9 g/cc. Results indicate that the thermite reaction occurs at about 900 0 C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U 3 O 8 -aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 660 0 C. In tension, sections failed at about the aluminum melting point. In compression with 2 psi average axial stress, failure occurred at 917 0 C, while 7 psi average axial stress produced failure at 669 0 C. (author)

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

    International Nuclear Information System (INIS)

    Peacock, H.B.

    1983-01-01

    Research and test reactors are presently operated with aluminum-clad fuel elements containing highly enriched uranium-aluminum alloy cores. To lower the enrichment and still maintain reactivity, the uranium content of the fuel element will need to be higher than currently achievable with alloy fuels. This will necessitate conversion to other forms such as U 3 O 8 -aluminum cermets. Above the aluminum melting point, U 3 O 8 and aluminum undergo an exothermic thermite reaction and cermet fuel cores tend to keep their original shape. Both factors could affect the course and consequences of a reactor accident, and prompted an investigation of the behavior of cermet fuels at elevated temperatures. Tests were carried out using pellets and extruded tube-sections with 53 wt % U 3 O 8 in aluminum. This content corresponds to a theoretical uranium density of 1.9 g/cc. Results indicate that the thermite reaction occurs at about 900 0 C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U 3 O 8 -aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 660 0 C. In tension, sections failed at about the aluminum melting point. In compression with 2-psi average axial stress, failure occurred at 917 0 C, while 7 psi average axial stress produced failure at 669 0 C

  8. 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.

  9. 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.

  10. Ausentismo y producción: el esquema de ajuste de primas por siniestralidad observada aplicado al seguro de salud en Francia

    Directory of Open Access Journals (Sweden)

    Sébastien Ménard

    2018-01-01

    Full Text Available En este trabajo se analizan los efectos de la introducción de un esquema de ajuste (experience rating en las contribuciones patronales al seguro de salud. Para ello, se desarrolla un modelo de búsqueda y emparejamiento en un mercado de trabajo segmentado. En el segmento donde los empleos son más riesgosos, las ausencias por enfermedad son más frecuentes y la producción media es más baja. El modelo es calibrado para Francia y los resultados de las simulaciones muestran que la introducción de este esquema mejora levemente la producción y el empleo. En términos cuantitativos, una variación del 1 % en la tasa de experience rating induce una reducción de 0,48 % en la duración del desempleo y de 1,07 % en los días de ausencia, así como un aumento de 0,1 % en la producción. Finalmente, se encuentra que esta política conduce también a un desempeño económico ligeramente mejor que la privatización del seguro de salud

  11. Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives

    International Nuclear Information System (INIS)

    Foerthmann, R.; Groos, E.; Gruebmeier, H.

    1975-08-01

    Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Temperature Gas-cooled Reactors led to the development of coated particles with improved fission product retention of the kernel, which represent an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates SrAl 2 Si 2 O 8 , BaAl 2 Si 2 O 8 and CsAlSi 2 O 6 in the additional aluminasilica phase of the kernel. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  13. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    fuel, Anticipated evolution of fuel in dry storage, Anticipated evolution of fuel in deep geological disposal); Boiling-water reactor fuel (Similarities, and differences with PWR fuel, Axial and radial zoning, Rod and channel box sizes, Poisoning and reactivity control, Cladding specific characteristics, Trends in fuel evolution); 3 - Liquid-metal-cooled fast reactor fuel: Fast-neutron irradiation damage in structural materials (Fast-neutron-induced damage in metals, What materials should be used?); Fuels and targets for fast-reactor transmutation (Fast reactors: reactors affording the ability to carry out effective actinide transmutation, Recycling: homogeneous, or heterogeneous?); 4 - gas-cooled reactor fuel: Particle fuel (From the initial concept to the advanced TRISO particle concept, Kernel fabrication processes, Particle coating by chemical vapor deposition, Fuel element fabrication: particle compaction, Characterization of fuel particles, and elements, From HTR fuel to VHTR and GFR fuels: the GAIA facility at CEA/Cadarache); Irradiation behavior of particle fuels (Particle fuel: a variety of failure modes for a high-strength object, The amoeba effect, Fission product behavior, and diffusion in particle fuels); Mechanical modeling of particle fuel; Very-high-temperature reactor (VHTR) fuel; Gas-cooled fast reactor (GFR) fuel (The specifications for GFR fuel, GFR fissile material, First containment baffler materials, GFR fuel element concepts); 5 - Research reactor fuels (A considerable feedback from experience, Conversion of French reactors to low-enriched (≤20% U-235)U 3 Si 2 fuel, Conversion of all reactors: R and D requirements for high-performance reactors, An 'advanced' research reactor fuel: UMo, The startup fuel for the Jules Horowitz Reactor (JHR) will still be U 3 Si 2 -Al; 6 - An instrument for future fuel research: the Jules Horowitz Reactor (JHR): Fuel irradiation experiments in JHR, JHR: a flexible instrument; 7 - Glossary-Index

  14. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    nature of spent nuclear fuel, Anticipated evolution of fuel in dry storage, Anticipated evolution of fuel in deep geological disposal); Boiling-water reactor fuel (Similarities, and differences with PWR fuel, Axial and radial zoning, Rod and channel box sizes, Poisoning and reactivity control, Cladding specific characteristics, Trends in fuel evolution); 3 - Liquid-metal-cooled fast reactor fuel: Fast-neutron irradiation damage in structural materials (Fast-neutron-induced damage in metals, What materials should be used?); Fuels and targets for fast-reactor transmutation (Fast reactors: reactors affording the ability to carry out effective actinide transmutation, Recycling: homogeneous, or heterogeneous?); 4 - gas-cooled reactor fuel: Particle fuel (From the initial concept to the advanced TRISO particle concept, Kernel fabrication processes, Particle coating by chemical vapor deposition, Fuel element fabrication: particle compaction, Characterization of fuel particles, and elements, From HTR fuel to VHTR and GFR fuels: the GAIA facility at CEA/Cadarache); Irradiation behavior of particle fuels (Particle fuel: a variety of failure modes for a high-strength object, The amoeba effect, Fission product behavior, and diffusion in particle fuels); Mechanical modeling of particle fuel; Very-high-temperature reactor (VHTR) fuel; Gas-cooled fast reactor (GFR) fuel (The specifications for GFR fuel, GFR fissile material, First containment baffler materials, GFR fuel element concepts); 5 - Research reactor fuels (A considerable feedback from experience, Conversion of French reactors to low-enriched ({<=}20% U-235)U{sub 3}Si{sub 2} fuel, Conversion of all reactors: R and D requirements for high-performance reactors, An 'advanced' research reactor fuel: UMo, The startup fuel for the Jules Horowitz Reactor (JHR) will still be U{sub 3}Si{sub 2}-Al; 6 - An instrument for future fuel research: the Jules Horowitz Reactor (JHR): Fuel irradiation experiments in JHR, JHR: a flexible

  15. Evaluation of the interfacial shear strength and residual stress of TiAlN coating on ZIRLO™ fuel cladding using a modified shear-lag model approach

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Y., E-mail: troy.liu@manchester.ac.uk [Materials Performance Centre, School of Materials, The University of Manchester, M13 9PL (United Kingdom); Bhamji, I., E-mail: imran.bhamji@manchester.ac.uk [Materials Performance Centre, School of Materials, The University of Manchester, M13 9PL (United Kingdom); Withers, P.J., E-mail: p.j.withers@manchester.ac.uk [Materials Performance Centre, School of Materials, The University of Manchester, M13 9PL (United Kingdom); Wolfe, D.E., E-mail: dew125@arl.psu.edu [The Pennsylvania State University, University Park, State College, PA 16801 (United States); Motta, A.T., E-mail: atmnuc@engr.psu.edu [The Pennsylvania State University, University Park, State College, PA 16801 (United States); Preuss, M., E-mail: michael.preuss@manchester.ac.uk [Materials Performance Centre, School of Materials, The University of Manchester, M13 9PL (United Kingdom)

    2015-11-15

    This paper investigates the residual stresses and interfacial shear strength of a TiAlN coating on Zr–Nb–Sn–Fe alloy (ZIRLO™) substrate designed to improve corrosion resistance of fuel cladding used in water-cooled nuclear reactors, both during normal and exceptional conditions, e.g. a loss of coolant event (LOCA). The distribution and maximum value of the interfacial shear strength has been estimated using a modified shear-lag model. The parameters critical to this analysis were determined experimentally. From these input parameters the interfacial shear strength between the TiAlN coating and ZIRLO™ substrate was inferred to be around 120 MPa. It is worth noting that the apparent strength of the coating is high (∼3.4 GPa). However, this is predominantly due to the large compressive residuals stress (3 GPa in compression), which must be overcome for the coating to fail in tension, which happens at a load just 150 MPa in excess of this.

  16. Determination of free acid in U(VI)-Al(III) solution by Gran plot titration

    International Nuclear Information System (INIS)

    Suh, Moo Yul; Lee, Chang Heon; Sohn, Se Chul; Kim, Jung Suk; Kim, Won Ho; Eom, Tae Yoon

    1999-01-01

    The determination method of free acid in spent U-Al nuclear fuel solutions by Gran plot titration was described. Effect of U(VI) and Al(III) on the alkalimetric titration of nitric acid was investigation in oxalate complexing media as well as in noncomplexing media. Positive biases were observed in both titration media when the end-point was estimated by the Gran plot method. It was found that the cause of the bias was U(VI) in the oxalate complexing media, but Al(III) in the noncomplexing media. The relative error was less than 1% in the titration of 0.1 M HNO 3 at a U(VI):Al(III):H + mole ratio of up to 2:12:1 as long as the pH of the oxalate titration media was sustained to be below 5.0 at the beginning of titration. The method was successfully applied to the determination of nitric acid in a solution of HANARO reactor fuel with U:Al mole ratio of 1:6

  17. Stability Study of the RERTR Fuel Microstructure

    Energy Technology Data Exchange (ETDEWEB)

    Jian Gan; Dennis Keiser; Brandon Miller; Daniel Wachs

    2014-04-01

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

  18. 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

  19. 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

  20. Experience in producing LEU fuel elements for the RSG-GAS

    International Nuclear Information System (INIS)

    Suripto, A.; Soentono, S.

    1991-01-01

    To achieve a self-reliance in the operation of the 30 MW Multipurpose Research Reactor at Serpong (the RSG-GAS), a fuel element production facility has been constructed nearby. The main task of the facility is to produce MTR type fuel and control elements containing U 3 O 8 -Al dispersion LEU fuel for the RSG-GAS. The hot commissioning activity has started in early 1988 after completion of the cold commissioning using depleted uranium in 1987, marking the beginning of the real production activity. This paper briefly describes the main features of the fuel production facility, the production experience gained so far, and its current production activity. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  2. Alternate fuels; Combustibles alternos

    Energy Technology Data Exchange (ETDEWEB)

    Romero Paredes R, Hernando; Ambriz G, Juan Jose [Universidad Autonoma Metropolitana. Iztapalapa (Mexico)

    2003-07-01

    In the definition and description of alternate fuels we must center ourselves in those technological alternatives that allow to obtain compounds that differ from the traditional ones, in their forms to be obtained. In this article it is tried to give an overview of alternate fuels to the conventional derivatives of petroleum and that allow to have a clear idea on the tendencies of modern investigation and the technological developments that can be implemented in the short term. It is not pretended to include all the tendencies and developments of the present world, but those that can hit in a relatively short term, in accordance with agreed with the average life of conventional fuels. Nevertheless, most of the conversion principles are applicable to the spectrum of carbonaceous or cellulosic materials which are in nature, are cultivated or wastes of organic origin. Thus one will approach them in a successive way, the physical, chemical and biological conversions that can take place in a production process of an alternate fuel or the same direct use of the fuel such as burning the sweepings derived from the forests. [Spanish] En la definicion y descripcion de combustibles alternos nos debemos centrar en aquellas alternativas tecnologicas que permitan obtener compuestos que difieren de los tradicionales, al menos en sus formas de ser obtenidos. En este articulo se pretende dar un panorama de los combustibles alternos a los convencionales derivados del petroleo y que permita tener una idea clara sobre las tendencias de la investigacion moderna y los desarrollos tecnologicos que puedan ser implementados en el corto plazo. No se pretende abarcar todas las tendencias y desarrollos del mundo actual, sino aquellas que pueden impactar en un plazo relativamente corto, acordes con la vida media de los combustibles convencionales. Sin embargo, la mayor parte de los principios de conversion son aplicables al espectro de materiales carbonaceos o celulosicos los cuales se

  3. Quantifying the potential impacts of fuel treatments on wildfire suppression costs volume

    Science.gov (United States)

    Matthew P. Thompson; Nicole M. Vaillant; Jessica R. Haas; Krista M. Gebert; Keith D. Stockmann

    2013-01-01

    Modeling the impacts and effects of hazardous fuel reduction treatments is a pressing issue within the wildfire management community. Prospective evaluation of fuel treatments allows for comparison of alternative treatment strategies in terms of socioeconomic and ecological impacts and facilitates analysis of tradeoffs across land management objectives (Stockmann et al...

  4. Contribuciones de la biblioteca universitaria a la internacionalización del campus: análisis sistemático de la literatura

    Directory of Open Access Journals (Sweden)

    Nancy Sánchez-Tarragó

    Full Text Available RESUMEN Los procesos de internacionalización universitaria constituyen respuesta adaptativa a un contexto mundial signado por la globalización, que implica, grosso modo, mayor interrelación política, económica y social entre países y regiones. En este contexto, es válido preguntarse cuál es el papel y las contribuciones de la biblioteca universitaria en los procesos de internacionalización del campus, teniendo en cuenta que, históricamente, su razón de ser ha sido servir de apoyo a la universidad en el desarrollo de sus funciones. Para encontrar respuestas se realizó una revisión sistemática de la literatura publicada en Ciencia de la información entre el 2005 y el 2015 y un análisis bibliométrico de esa producción científica. Se constató un crecimiento continuo de la investigación publicada, liderada por bibliotecarios de Estados Unidos, China y Canadá, que abordan diferentes aristas del trabajo bibliotecario que apoyan tanto la educación transnacional como la internacionalización ‘en casa’: servicios bibliotecarios para apoyar estudiantes y programas en el exterior, programas de alfabetización informacional, estudios sobre la influencia de las diferencias culturales y lingüísticas en la comunicación y los servicios, desarrollo de colecciones, entre otras. Se identificaron recomendaciones para la práctica bibliotecaria relacionadas con estrategias administrativas, de servicios, infraestructura, formación e investigación.

  5. Behaviour of rock-like oxide fuels under reactivity-initiated accident conditions

    International Nuclear Information System (INIS)

    Kazuyuki, Kusagaya; Takehiko, Nakamura; Makio, Yoshinaga; Hiroshi, Akie; Toshiyuki, Yamashita; Hiroshi, Uetsuka

    2002-01-01

    Pulse irradiation tests of three types of un-irradiated rock-like oxide (ROX) fuel - yttria-stabilised zirconia (YSZ) single phase, YSZ and spinel (MgAl 2 O 4 ) homogeneous mixture and particle-dispersed YSZ/spinel - were conducted in the Nuclear Safety Research Reactor to investigate the fuel behaviour under reactivity-initiated accident conditions. The ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m 3 , which was comparable to that of un-irradiated UO 2 fuel. The failure mode of the ROX fuels, however, was quite different from that of the UO 2 fuel. The ROX fuels failed with fuel pellet melting and a part of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below∼12 GJ/m 3 . The YSZ type and homogenous YSZ/spinel type ROX fuels failed by cladding burst when their temperatures peaked, while the particle-dispersed YSZ/spinel type ROX fuel seemed to have failed by cladding local melting. (author)

  6. Study of fuel control strategy based on an fuel behavior model for starting conditions; Nenryo kyodo model ni motozuita shidoji no nenryo hosei hosho ni tsuite no kosatsu

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Y; Uchida, M; Iwano, H; Oba, H [Nissan Motor Co. Ltd., Tokyo (Japan)

    1997-10-01

    We have applied a fuel behavior model to a fuel injection system which we call SOFIS (Sophisticated and Optimized Fuel Injection System) so that we get air/fuel ratio control accuracy and good driveability. However the fuel behavior under starting conditions is still not clear. To meet low emission rules and to get better driveability under starting conditions, better air/fuel ratio control is necessary. Now we have understood the ignition timing, injection timing, and injection pulse width required in such conditions. In former days, we analyzed the state of the air/fuel mixture under cold conditions and made a new fuel behavior model which considered fuel loss such as hydrocarbons and dissolution into oil and so on. Al this time, we have applied this idea to starting. We confirm this new model offers improved air/fuel ratio control. 6 refs., 9 figs., 3 tabs.

  7. Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives

    Energy Technology Data Exchange (ETDEWEB)

    Foerthmann, R.; Groos, E.; Gruebmeier, H.

    1975-08-15

    Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Temperature Gas-cooled Reactors led to the development of coated particles with improved fission product retention which represent an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates SrAl2Si2O8, BaAl2Si2O8and CsAlSi2O6 in the additional alumina-silica phase of the kernel.

  8. Selenium fuel: Surface engineering of U(Mo) particles to optimise fuel performance

    International Nuclear Information System (INIS)

    Van den Berghe, S.; Leenaers, A.; Detavernier, C.

    2010-01-01

    Recent developments on the stabilisation of U(Mo) in-pile behaviour in plate-type fuel have focussed almost exclusively on the addition of Si to the Al matrix of the fuel. This has now culminated in a qualification effort in the form of the European LEONIDAS initiative for which irradiations will start in 2010. In this framework, many discussions have been held on the Si content of the matrix needed for stabilisation of the interaction phase and the requirement for the formation of Si-rich layers around the particles during the fabrication steps. However, it is clear that the Si needs to be incorporated in the interaction phase for it to be effective, for which the currently proposed methods depend on a diffusion mechanism, which is difficult to control. This has lead to the concept of a Si coated particle as a more efficient way of incorporating the Si in the fuel by putting it immediately where it will be required : at the fuel-matrix interface. As part of the SELENIUM (Surface Engineered Low ENrIched Uranium-Molybdenum fuel) project, SCK CEN has built a sputter coater for PVD magnetron sputter coating of particles in collaboration with the University of Ghent. The coater is equipped with three 3 inch magnetron sputter heads, allowing deposition of 3 different elements or a single element at high deposition speed. The particles are slowly rotated in a drum to produce homogeneous layer thicknesses. (author)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  10. High temperature compression tests performed on doped fuels

    International Nuclear Information System (INIS)

    Duguay, C.; Mocellin, A.; Dehaudt, P.; Fantozzi, G.

    1997-01-01

    The use of additives of corundum structure M 2 O 3 (M=Cr, Al) is an effective way of promoting grain growth of uranium dioxide. The high-temperature compressive deformation of large-grained UO 2 doped with these oxides has been investigated and compared with that of pure UO 2 with a standard microstructure. Such doped fuels are expected to exhibit enhanced plasticity. Their use would therefore reduce the pellet-cladding mechanical interaction and thus improve the performances of the nuclear fuel. (orig.)

  11. Parametric study of the deformation of dispersion fuel plates

    International Nuclear Information System (INIS)

    Vieira, Edeval; Leal Neto, Ricardo Mendes; Durazzo, Michelangelo

    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, presenting results of the evolution of all parameters of technological interest, after each rolling pass, obtaining information along the fuel plate deformation during the rolling process. (author)

  12. Waste Not, Want Not: Lactate Oxidation Fuels the TCA Cycle.

    Science.gov (United States)

    Martínez-Reyes, Inmaculada; Chandel, Navdeep S

    2017-12-05

    Previous studies have demonstrated that mitochondrial respiration is essential for tumorigenesis. Hui et al. (2017) and Faubert et al. (2017) demonstrate that lactate, traditionally viewed as a waste product of anaerobic and aerobic glycolysis, is a major carbon source to fuel the mitochondrial TCA cycle in normal tissue and in tumors. Copyright © 2017 Elsevier Inc. All rights reserved.

  13. Preparation of a Ni-MgO-Al2O3 catalyst with high activity and resistance to potassium poisoning during direct internal reforming of methane in molten carbonate fuel cells

    Science.gov (United States)

    Jang, Won-Jun; Jung, You-Shick; Shim, Jae-Oh; Roh, Hyun-Seog; Yoon, Wang Lai

    2018-02-01

    Steam reforming of methane (SRM) is conducted using a series of Ni-MgO-Al2O3 catalysts for direct internal reforming (DIR) in molten carbonate fuel cells (MCFCs). Ni-MgO-Al2O3 catalysts are prepared by the homogeneous precipitation method with a variety of MgO loading amounts ranging from 3 to 15 wt%. In addition, each precursor concentrations are systemically changed (Ni: 1.2-4.8 mol L-1; Mg: 0.3-1.2 mol L-1; Al: 0.4-1.6 mol L-1) at the optimized composition (10 wt% MgO). The effects of MgO loading and precursor concentration on the catalytic performance and resistance against poisoning of the catalyst by potassium (K) are investigated. The Ni-MgO-Al2O3 catalyst with 10 wt% MgO and the original precursor concentration (Ni: 1.2 mol L-1; Mg: 0.3 mol L-1; Al: 0.4 mol L-1) exhibits the highest CH4 conversion and resistance against K poisoning even at the extremely high gas space velocity (GHSV) of 1,512,000 h-1. Excellent SRM performance of the Ni-MgO-Al2O3 catalyst is attributed to strong metal (Ni) to alumina support interaction (SMSI) when magnesium oxide (MgO) is co-precipitated with the Ni-Al2O3. The enhanced interaction of the Ni with MgO-Al2O3 support is found to protect the active Ni species against K poisoning.

  14. The ORR Whole-Core LEU Fuel Demonstration

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Snelgrove, J.L.

    1990-01-01

    The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U 3 Si 2 -Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged 235 U burnups validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of 235 U burnup support the corresponding measured quantities. In general, calculations for reaction rate distributions, control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 17 refs

  15. Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

    Science.gov (United States)

    Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

    2017-08-01

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

  16. Alcohol fuels program technical review

    Energy Technology Data Exchange (ETDEWEB)

    None

    1981-07-01

    The last issue of the Alcohol Fuels Process R/D Newsletter contained a work breakdown structure (WBS) of the SERI Alcohol Fuels Program that stressed the subcontracted portion of the program and discussed the SERI biotechnology in-house program. This issue shows the WBS for the in-house programs and contains highlights for the remaining in-house tasks, that is, methanol production research, alcohol utilization research, and membrane research. The methanol production research activity consists of two elements: development of a pressurized oxygen gasifier and synthesis of catalytic materials to more efficiently convert synthesis gas to methanol and higher alcohols. A report is included (Finegold et al. 1981) that details the experimental apparatus and recent results obtained from the gasifier. The catalysis research is principally directed toward producing novel organometallic compounds for use as a homogeneous catalyst. The utilization research is directed toward the development of novel engine systems that use pure alcohol for fuel. Reforming methanol and ethanol catalytically to produce H/sub 2/ and CO gas for use as a fuel offers performance and efficiency advantages over burning alcohol directly as fuel in an engine. An application of this approach is also detailed at the end of this section. Another area of utilization is the use of fuel cells in transportation. In-house researchers investigating alternate electrolyte systems are exploring the direct and indirect use of alcohols in fuel cells. A workshop is being organized to explore potential applications of fuel cells in the transportation sector. The membrane research group is equipping to evaluate alcohol/water separation membranes and is also establishing cost estimation and energy utilization figures for use in alcohol plant design.

  17. Metal hydride and pyrophoric fuel additives for dicyclopentadiene based hybrid propellants

    Science.gov (United States)

    Shark, Steven C.

    The purpose of this study is to investigate the use of reactive energetic fuel additives that have the potential to increase the combustion performance of hybrid rocket propellants in terms of solid fuel regression rate and combustion efficiency. Additives that can augment the combustion flame zone in a hybrid rocket motor by means of increased energy feedback to the fuel grain surface are of great interest. Metal hydrides have large volumetric hydrogen densities, which gives these materials high performance potential as fuel additives in terms of specifc impulse. The excess hydrogen and corresponding base metal may also cause an increase in the hybrid rocket solid fuel regression rate. Pyrophoric additives also have potential to increase the solid fuel regression rate by reacting more readily near the burning fuel surface providing rapid energy feedback. An experimental performance evaluation of metal hydride fuel additives for hybrid rocket motor propulsion systems is examined in this study. Hypergolic ignition droplet tests and an accelerated aging study revealed the protection capabilities of Dicyclopentadiene (DCPD) as a fuel binder, and the ability for unaided ignition. Static hybrid rocket motor experiments were conducted using DCPD as the fuel. Sodium borohydride (NabH4) and aluminum hydride (AlH3) were examined as fuel additives. Ninety percent rocket grade hydrogen peroxide (RGHP) was used as the oxidizer. In this study, the sensitivity of solid fuel regression rate and characteristic velocity (C*) efficiency to total fuel grain port mass flux and particle loading is examined. These results were compared to HTPB combustion performance as a baseline. Chamber pressure histories revealed steady motor operation in most tests, with reduced ignition delays when using NabH4 as a fuel additive. The addition of NabH4 and AlH3 produced up to a 47% and 85% increase in regression rate over neat DCPD, respectively. For all test conditions examined C* efficiency ranges

  18. Caracterización curricular de la Unidad de Formación Humana y Bioética de la Universidad de San Buenaventura Medellín

    OpenAIRE

    Sánchez Sánchez, Luis Fernando; Moreno López, Verónica

    2011-01-01

    Ubicación en Biblioteca USB Medellín (San Benito): CD-2033t El presente informe recoge los desarrollos y resultados investigativos que entorno al currículo de la formación socio ¿ humanista de la Universidad de San Buenaventura (USB) Medellín, se generaron como propuesta para optar al título de magister en educación.El desarrollo del proyecto generó amplias contribuciones a la formación de los estudiantes de la Maestría en Educación, permitiéndoles conocer las bondades y la complejidad de ...

  19. Extended storage of spent fuel

    International Nuclear Information System (INIS)

    1992-10-01

    This document is the final report on the IAEA Co-ordinated Research Programme on the Behaviour of Spent Fuel and Storage Facility Components during Long Term Storage (BEFAST-II, 1986-1991). It contains the results on wet and dry spent fuel storage technologies obtained from 16 organizations representing 13 countries who participated in the co-ordinated research programme. Considerable quantities of spent fuel continue to arise and accumulate. Many countries are investigating the option of extended spent fuel storage prior to reprocessing or fuel disposal. Wet storage continues to predominate as an established technology with the construction of additional away-from-reactor storage pools. However, dry storage is increasingly used with most participants considering dry storage concepts for the longer term. Depending on the cladding type options of dry storage in air or inert gas are proposed. Dry storage is becoming widely used as a supplement to wet storage for zirconium alloy clad oxide fuels. Storage periods as long as under wet conditions appear to be feasible. Dry storage will also continue to be used for Al clad and Magnox type fuel. Enhancement of wet storage capacity will remain an important activity. Rod consolidation to increase wet storage capacity will continue in the UK and is being evaluated for LWR fuel in the USA, and may start in some other countries. High density storage racks have been successfully introduced in many existing pools and are planned for future facilities. For extremely long wet storage (≥50 years), there is a need to continue work on fuel integrity investigations and LWR fuel performance modelling. it might be that pool component performance in some cases could be more limiting than the FA storage performance. It is desirable to make concerted efforts in the field of corrosion monitoring and prediction of fuel cladding and poll component behaviour in order to maintain good experience of wet storage. Refs, figs and tabs

  20. Chemical looping combustion. Fuel conversion with inherent CO2 capture

    Energy Technology Data Exchange (ETDEWEB)

    Brandvoll, Oeyvind

    2005-07-01

    Chemical looping combustion (CLC) is a new concept for fuel energy conversion with CO2 capture. In CLC, fuel combustion is split into separate reduction and oxidation processes, in which a solid carrier is reduced and oxidized, respectively. The carrier is continuously recirculated between the two vessels, and hence direct contact between air and fuel is avoided. As a result, a stoichiometric amount of oxygen is transferred to the fuel by a regenerable solid intermediate, and CLC is thus a variant of oxy-fuel combustion. In principle, pure CO2 can be obtained from the reduction exhaust by condensation of the produced water vapour. The thermodynamic potential and feasibility of CLC has been studied by means of process simulations and experimental studies of oxygen carriers. Process simulations have focused on parameter sensitivity studies of CLC implemented in 3 power cycles; CLC-Combined Cycle, CLC-Humid Air Turbine and CLC-Integrated Steam Generation. Simulations indicate that overall fuel conversion ratio, oxidation temperature and operating pressure are among the most important process parameters in CLC. A promising thermodynamic potential of CLC has been found, with efficiencies comparable to, - or better than existing technologies for CO2 capture. The proposed oxygen carrier nickel oxide on nickel spinel (NiONiAl) has been studied in reduction with hydrogen, methane and methane/steam as well as oxidation with dry air. It has been found that at atmospheric pressure and temperatures above 600 deg C, solid reduction with dry methane occurs with overall fuel conversion of 92%. Steam methane reforming is observed along with methane cracking as side reactions, yielding an overall selectivity of 90% with regard to solid reduction. If steam is added to the reactant fuel, coking can be avoided. A methodology for long-term investigation of solid chemical activity in a batch reactor is proposed. The method is based on time variables for oxidation. The results for NiONiAl

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  2. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements; Procedimentos de fabricacao de elementos combustiveis a base de dispersoes com alta concentracao de uranio

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Jose Antonio Batista de

    2011-07-01

    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{sup 3} for U{sub 3}Si{sub 2}-Al dispersion-based and 2.3 gU/cm{sup 3} for U{sub 3}O{sub 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{sup 3} in U{sub 3}Si{sub 2}-Al dispersion and 3.2 gU/cm{sup 3} U{sub 3}O{sub 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{sub 3}Si{sub 2}-Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U{sub 3}O{sub 8}-Al dispersion fuel plates with 3.2 gU/cm{sup 3} showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U{sub 3}Si{sub 2} production at 4.8 gU/cm{sup 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)

  3. Predicting Argentine Jet Fuel Prices

    Science.gov (United States)

    2007-03-01

    economics, engineering, and the social and biological sciences (Kutner, 2005:2), but successful application of this method requires not only a deep...situations that affected the country. This crisis had politic, economic and social implications. Five presidents governed in a two month period. The...Scheimberg Sebastian. “Comentario al trabajo de German Coloma: Analisys of the behavior of the Argentine fuel market”. Political Economy Asociation

  4. 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)

  5. Insect Optic Glomeruli: Exploration of a Universal Circuit for Sensorimotor Processing

    Science.gov (United States)

    2012-01-20

    based  on  an  algorithm   published  by  Cohen...Cajal,  SR,  Sánchez  D  (1915)  Contribucion  al   conocimiento  de  los  centros  nerviosos   de  los  insectos...Drosophila   revealed  by  green  fluorescent  protein-­‐ based  Ca2+  imaging.  J  Neurosci  24:   6507–6514.    

  6. Evolution of fuel plate parameters during deformation in rolling

    Energy Technology Data Exchange (ETDEWEB)

    Durazzo, M., E-mail: mdurazzo@ipen.br [Nuclear and Energy Research Institute – IPEN/CNEN-SP, São Paulo (Brazil); Vieira, E.; Urano de Carvalho, E.F. [Nuclear and Energy Research Institute – IPEN/CNEN-SP, São Paulo (Brazil); Riella, H.G. [Nuclear and Energy Research Institute – IPEN/CNEN-SP, São Paulo (Brazil); Chemical Engineering Department, Santa Catarina Federal University, Florianópolis (Brazil)

    2017-07-15

    The Nuclear and Energy Research Institute – IPEN/CNEN-SP routinely produces the nuclear fuel necessary for operating its research reactor, IEA-R1. This fuel consists of fuel plates containing U{sub 3}Si{sub 2}-Al composites as the meat, which are fabricated by rolling. The rolling process currently deployed was developed based on information obtained from literature, which was used as a premise for defining the current manufacturing procedures, according to a methodology with an essentially empirical character. Despite the current rolling process being perfectly stable and highly reproducible, it is not well characterized and is therefore not fully known. The objective of this work is to characterize the rolling process for producing dispersion fuel plates. Results regarding the evolution of the main parameters of technological interest, after each rolling pass, are presented. Some defects that originated along the fuel plate deformation during the rolling process were characterized and discussed. The fabrication procedures for manufacturing the fuel plates are also presented. - Highlights: •Evolution of defects when manufacturing dispersion fuel plates. •Aspects of dispersion fuel plates fabrication. •What happen during the manufacturing of dispersion fuel plates? •Clarifying the deformation of fuel plates by rolling.

  7. Experimental assessment of fuel-cladding interactions

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-29

    A range of fuel concepts designed to better tolerate accident scenarios and reactor transients are currently undergoing fundamental development at national laboratories as well as university and industrial partners. Pellet-clad mechanical and chemical interaction can be expected to affect fuel failure rates experienced during steady state operation, as well as dramatically impact the response of the fuel form under loss of coolant and other accident scenarios. The importance of this aspect of fuel design prompted research initiated by AFC in FY14 to begin exploratory efforts to characterize this phenomenon for candidate fuelcladding systems of immediate interest. Continued efforts in FY15 and FY17 aimed to better understand and simulate initial pellet-clad interaction with little-to-no pressure on the pellet-clad interface. Reported here are the results from 1000 h heat treatments at 400, 500, and 600°C of diffusion couples pairing UN with a FeCrAl alloy, SiC, and Zr-based cladding candidate sealed in evacuated quartz ampoules. No gross reactions were observed, though trace elemental contaminants were identified.

  8. Cost and energy consumption estimates for the aluminum-air battery anode fuel cycle

    Science.gov (United States)

    1990-01-01

    At the request of DOE's Office of Energy Storage and Distribution (OESD), Pacific Northwest Laboratory (PNL) conducted a study to generate estimates of the energy use and costs associated with the aluminum anode fuel cycle of the aluminum-air (Al-air) battery. The results of this analysis indicate that the cost and energy consumption characteristics of the mechanically rechargeable Al-air battery system are not as attractive as some other electrically rechargeable electric vehicle battery systems being developed by OESD. However, there are distinct advantages to mechanically rechargeable batteries, which may make the Al-air battery (or other mechanically rechargeable batteries) attractive for other uses, such as stand-alone applications. Fuel cells, such as the proton exchange membrane (PEM), and advanced secondary batteries may be better suited to electric vehicle applications.

  9. Study of the uniform corrosion of an aluminium alloy used for the fuel cladding of the Jules Horowitz experimental reactor

    International Nuclear Information System (INIS)

    Wintergerst, M.

    2008-01-01

    For the Jules Horowitz new material testing reactor, an aluminium base alloy, AlFeNi, will be used for the cladding of the fuel plates. Taking into account the thermal properties of the alloy and of its oxide, the corrosion of the fuel cans presents many problems. The aim of this thesis is to provide a growing kinetic of the oxide layer at the surface of the AlFeNi fuel can in order to predict the life time of fuel element. Thus the mechanism of degradation of the cladding will be describe in order to integrate the different parameters of the operating reactor. (A.L.B.)

  10. Numerical Simulations of Hollow-Cone Injection and Gasoline Compression Ignition Combustion With Naphtha Fuels

    KAUST Repository

    Badra, Jihad A.

    2016-01-29

    Gasoline compression ignition (GCI), also known as partially premixed compression ignition (PPCI) and gasoline direct injection compression ignition (GDICI), engines have been considered an attractive alternative to traditional spark ignition (SI) engines. Lean-burn combustion with the direct injection of fuel eliminates throttle losses for higher thermodynamic efficiencies, and the precise control of the mixture compositions allows better emission performance such as NOx and particulate matter (PM). Recently, low octane gasoline fuel has been identified as a viable option for the GCI engine applications due to its longer ignition delay characteristics compared to diesel and lighter evaporation compared to gasoline fuel (Chang et al., 2012, "Enabling High Efficiency Direct Injection Engine With Naphtha Fuel Through Partially Premixed Charge Compression Ignition Combustion," SAE Technical Paper No. 2012-01-0677). The feasibility of such a concept has been demonstrated by experimental investigations at Saudi Aramco (Chang et al., 2012, "Enabling High Efficiency Direct Injection Engine With Naphtha Fuel Through Partially Premixed Charge Compression Ignition Combustion," SAE Technical Paper No. 2012-01-0677; Chang et al., 2013, "Fuel Economy Potential of Partially Premixed Compression Ignition (PPCI) Combustion With Naphtha Fuel," SAE Technical Paper No. 2013-01-2701). The present study aims to develop predictive capabilities for low octane gasoline fuel compression ignition (CI) engines with accurate characterization of the spray dynamics and combustion processes. Full three-dimensional simulations were conducted using converge as a basic modeling framework, using Reynolds-averaged Navier-Stokes (RANS) turbulent mixing models. An outwardly opening hollow-cone spray injector was characterized and validated against existing and new experimental data. An emphasis was made on the spray penetration characteristics. Various spray breakup and collision models have been

  11. Development of high-strength aluminum alloys for basket in transport and storage cask for high burn-up spent fuel

    International Nuclear Information System (INIS)

    Maeguchi, T.; Sakaguchi, Y.; Kamiwaki, Y.; Ishii, M.; Yamamoto, T.

    2004-01-01

    Mitsubishi Heavy Industries, Ltd. (MHI) has developed high-strength borated aluminum alloys (high-strength B-Al alloys), suitable for application to baskets in transport and storage casks for high burn-up spent fuels. Aluminum is a suitable base material for the baskets due to its low density and high thermal conductivity. The aluminum basket would reduce weight of the cask, and effectively release heat generated by spent fuels. MHI had already developed borated aluminum alloys (high-toughness B-Al alloy), and registered them as ASME Code Case ''N-673''. However, there has been a strong demand for basket materials with higher strength in the case of MSF (Mitsubishi Spent Fuel) casks for high-burn up spent fuels, since the basket is required to stand up to higher stress at higher temperature. The high-strength basket material enables the design of a compact cask under a limitation of total size and weight. MHI has developed novel high-strength B-Al alloys which meet these requirements, based on a new manufacturing process. The outline of mechanical and metallurgical characteristics of the high-strength B-Al alloys is described in this paper

  12. The use of blunted Maya crude as alternating fuel to fuel oil; Uso de crudo Maya despuntado como combustiononble alterno al combustoleo

    Energy Technology Data Exchange (ETDEWEB)

    Salinas Bravo, Victor M.; Diego Marin, Antonio; Porcayo C; Jesus [Instituto de Investigaciones Electricas, Temixco, Morelos (Mexico)

    2001-07-01

    The Mexican fuel oil is characterized by its high viscosity and elevated contents of sulfur and vanadium. This particularity determines a problematic characteristic in the steam generators that use this fuel. Our country produces three types of crude petroleum: Olmeca, Istmo and Maya. The difference between these crudes remains in its density and the sulfur content. The Olmeca crude and the Istmo are classified as super-light and light, respectively; whereas the Maya crude is classified as a heavy crude, with a density of 1.0 to 0.92 g/cm and a minimum sulfur content of 3.3 %. For this reason the processing of the Maya crude is seen as an alternative. A table with the physicochemical analysis and properties of a fuel oil and the blunted Maya crude is given, and another one with the quantitative chemical analysis of some metals in fuel oil and in the blunted Maya crude. The stability tests of the fuel are studied giving a comparison of the effect of the aging time on the viscosity of a fuel oil and of the blunted Maya crude at 60 Celsius degrees. As well as a graph of thermograms obtained of a blunted Maya crude sample and of a typical fuel oil. Finally the corrosion tests are described that allow to know in real time the degradation process by corrosion of the materials. [Spanish] El combustoleo mexicano se caracteriza por su alta viscosidad y elevados contenidos de azufre y vanadio. Esta particularidad determina una problematica caracteristica en los generadores de vapor que utilizan este combustible. Nuestro pais produce tres tipos de petroleo crudo: Olmeca, Istmo y Maya. La diferencia entre estos crudos estriba en su densidad y el contenido de azufre. El crudo Olmeca y el Istmo estan clasificados como superligero y ligero, respectivamente; mientras que el crudo Maya esta clasificado como un crudo pesado, con una densidad de 1.0 a 0.92 g/cm y un contenido minimo de azufre de 3.3 %.Con el proposito se busca la alternativa de procesar el crudo Maya. Se da una tabla

  13. 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

  14. High temperature compression tests performed on doped fuels

    Energy Technology Data Exchange (ETDEWEB)

    Duguay, C.; Mocellin, A.; Dehaudt, P. [Commissariat a l`Energie Atomique, CEA Grenoble (France); Fantozzi, G. [INSA Lyon - GEMPPM, Villeurbanne (France)

    1997-12-31

    The use of additives of corundum structure M{sub 2}O{sub 3} (M=Cr, Al) is an effective way of promoting grain growth of uranium dioxide. The high-temperature compressive deformation of large-grained UO{sub 2} doped with these oxides has been investigated and compared with that of pure UO{sub 2} with a standard microstructure. Such doped fuels are expected to exhibit enhanced plasticity. Their use would therefore reduce the pellet-cladding mechanical interaction and thus improve the performances of the nuclear fuel. (orig.) 5 refs.

  15. The fuel index method as a new planning instrument in the erection and redevelopment of residential estates; Das Brennstoffkennzahl-Verfahren als neues Planungsinstrument fuer Neubau- und Sanierungsvorhaben

    Energy Technology Data Exchange (ETDEWEB)

    Elsenberger, U. [Industrie und Communal Consulting GmbH, Aachen (Germany); Boese, M.; Diefenbach, N.; Petersdorff, C.; Schultze, K. [Rheinisch-Westfaelische Technische Hochschule Aachen (Germany). 3. Physikalisches Inst. A

    1998-12-31

    The fuel index method aims at the largest possible exploitation of conservation potentials in residential thermal energy supply. The introduction of a thermal quality standard for thermal energy systems lays the foundation in process engineering terms for a competition between environment-friendly thermal energy supply and thermal insulation, with the aim to save energy and cost. Using higher-grade thermal energy at higher cost than current energy costs is approved of if its use involves an ambitious conservation aim. In this way, particularly rational energy techniques are promoted. (orig.) [Deutsch] Das Brennstoffkennzahl-Verfahren zielt auf eine effektive Erschliessung weitgehender Einsparziele bei der Waermeversorgung von Wohngebieten. Mit der Einfuehrung der Waermequalitaet als Guetemass eines Waermeversorgungssystems werden die verfahrenstechnischen Grundlagen fuer eine auf kostenguenstige Energieeinsparung ausgerichtete Konkurrenz zwischen umweltfreundlicher Waermeversorgung und Waermeschutz geschaffen. Waerme hoeherer Qualitaet darf bei Vorgabe eines ehrgeizigen Einsparziels teurer sein als heute ueblich. Damit wird der Einsatz besonders rationeller Energietechniken gefoerdert. (orig.)

  16. Synthesis and characterization of nanocrystalline Cu-Al coatings

    International Nuclear Information System (INIS)

    Lau, M.L.; He, J.; Schweinfest, R.; Ruehle, M.; Levi, C.G.; Lavernia, E.J.

    2003-01-01

    Commercially pure Cu and Al powders were blended in a 90:10 ratio by weight and then mechanically milled in methanol or in liquid nitrogen. The milled powders, as well as as-blended (non-milled) powder, were deposited as coatings using high velocity oxygen fuel thermal spraying. Scanning and transmission electron microscopy techniques were used to investigate the microstructure of the powders and coatings. The results showed that milling of the powders in methanol induced the conversion of most of the Al into amorphous Al 2 O 3 , precluding the desired mechanical alloying. This experimental observation was consistent with available thermodynamic data. In contrast, cryomilling exhibited no significant oxidation and induced mechanical alloying of the powders, albeit incomplete. The non-milled powder generated a coating with a bimodal grain structure consisting of fine Cu grains and coarse Al grains. Amorphous oxide regions and coarse Al grains were observed intermixed with the finer Cu matrix in the coatings sprayed using the powders milled in methanol. Coatings based on cryomilled powders consisted primarily of equiaxed Cu grains and twinned martensite regions, with occasional inclusion of elongated amorphous Al 2 O 3 regions

  17. Irradiation behavior of modified high-performance nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jungwirth, Rainer

    2011-11-03

    To overcome the problem of UMo/Al fuel swelling, four different possibilities have been identified: (i) the modification of the Al matrix by adding diffusion limiting elements (ii) the insertion of a diffusion barrier at the interface UMo-Al (iii) further alloying the UMo with a third element to stabilize the γ-UMo phase (iv) a combination of means (i)-(iii). In consequence, 20 different UMoX/AlY (X=Si, Ti, Mg, Bi, with and without oxidation layer; Y=Nb, Ti, Pt) samples have been examined before and after irradiation with Iodine at 80MeV. First it has been shown, that a protective oxidation layer on the UMo grains does not prevent the formation of a interdiffusion layer. In contrast, additions to the Al matrix can be reduced to the self-acting formation of a protective layer at the UMo/Al interface. Additions to the UMo to stabilize the γ-UMo upon heating are of minor importance since irradiation reverses the phase decomposition of UMo.

  18. Development of very-high-density low-enriched-uranium fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Trybus, C.L.; Wiencek, T.C.

    1997-01-01

    Following a hiatus of several years and following its successful development and qualification of 4.8 g U cm -3 U 3 Si 2 -Al dispersion fuel for application with low-enriched uranium in research and test reactors, the US Reduced Enrichment for Research and Test Reactors program has embarked on the development of even-higher-density fuels. Our goal is to achieve uranium densities of 8-9 g cm -3 in aluminum-based dispersion fuels. Achieving this goal will require the use of high-density, γ-stabilized uranium alloy powders in conjunction with the most-advanced fuel fabrication techniques. Key issues being addressed are the reaction of the fuel alloys with aluminum and the irradiation behavior of the fuel alloys and any reaction products. Test irradiations of candidate fuels in very-small (micro) plates are scheduled to begin in the Advanced Test Reactor during June, 1997. Initial results are expected to be available in early 1998. We are performing out-of-reactor studies on the phase structure of the candidate alloys on diffusion of the matrix material into the aluminum. In addition, we are modifying our current dispersion fuel irradiation behavior model to accommodate the new fuels. Several international partners are participating in various phases of this work. (orig.)

  19. R and D for back-end options for irradiated research reactor fuel in Germany

    International Nuclear Information System (INIS)

    Bruecher, H.; Curtius, H.; Fachinger, J.

    2001-01-01

    Out of 11.5 t of irradiated fuel arising from German research reactors until the end of this decade, 3.9 t are intended to be returned to the USA, and 2.3 t are expected to be recycled for reuse of uranium. The remaining 5.3 t, as well as the fuel irradiated after the year 2010, will have to follow the domestic back-end option of extended dry interim storage in Castor-type casks, followed by disposal in a deep geological repository. R and D is going on in the Research Centre Juelich to investigate the long-term behaviour of U-Al based fuel in a salt repository. First results from leaching experiments show I) a fast dissolution of the fuel with mobilization of its radionuclide inventory, and 2) the following formation of amorphous Al-Mg-hydroxide phases. Long-lived actinides from the fuel were shown to be fixed in these phases and hence immobilized. Future R and D will be to investigate the nature and stability of these phases for long-term safety assessments. Investigations will have to be extended to cover alternative disposal sites (granite clay) as well as different (e.g. silicon based) fuels. (author)

  20. Comparison of quartz tuning forks and AlN-based extensional microresonators for viscosity measurements in oil/fuel mixtures

    Science.gov (United States)

    Toledo, J.; Manzaneque, T.; Hernando-García, J.; Vazquez, J.; Ababneh, A.; Seidel, H.; Lapuerta, M.; Sánchez-Rojas, J. L.

    2013-05-01

    In-situ monitoring of the physical properties of liquids is of great interest in the automotive industry. For example, lubricants are subject to dilution with diesel fuel as a consequence of late-injection processes, which are necessary for regenerating diesel particulate filters. This dilution can be determined by tracking the viscosity and the density of the lubricant. Here we report the test of two in-plane movement based resonators to explore their capability to monitor oil dilution with diesel and biodiesel. One of the resonators is the commercially available millimeter-sized quartz tuning fork, working at 32.7 kHz. The second resonator is a state-of-the-art micron-sized AlN-based rectangular plate, actuated in the first extensional mode in the MHz range. Electrical impedance measurements were carried out to characterize the performance of the structures in various liquid media in a wide range of viscosities. These measurements were completed with the development of low-cost electronic circuits to track the resonance frequency and the quality factor automatically, these two parameters allow to obtain the viscosity of various fluids under investigation, as in the case of dilution of lubricant SAE 15W40 and biodiesel.

  1. Fabrication of fully ceramic microencapsulated fuel by hot pressing

    International Nuclear Information System (INIS)

    Lee, H. G.; Kim, D. J; Park, J. Y.; Kim, W. J.; Lee, S. J

    2014-01-01

    Fully ceramic microencapsulated(FCM) nuclear fuel is one of the recently suggested concept to enhance stability nuclear fuel itself. The requirements to increase the accident tolerance of nuclear fuel are mainly two parts: First, the performance has to be maintained compared to the existing UO 2 nuclear fuel and zircaloy cladding system under the normal operation condition. Second, under the severe accident condition, the high temperature structural integrity has to be kept and the generation rate of hydrogen has to be decrease largely. FCM nuclear fuel consists of tristructural isotropic(TRISO) fuel particle and SiC matrix. The relative thermal conductivity of the SiC matrix as compared to UO 2 is quite good, yielding as-irradiated fuel centerline temperature compared to high temperature for the existing fuel leading to reduced stored energy in the core and reduced operational release of fission products from the fuel. Generally SiC ceramics are fabricated via liquid phase sintering due to strong covalent bonding property and low self-diffusivity coefficient. Hot pressing is very effective method to conduct sintering of SiC powder including different second phase. In this study, SiC-matrix composite including TRISO particles were sintered by hot pressing with Al 2 O 3 -Y 2 O 3 additive system. Various sintering condition were investigated to obtain high relative density above 95%. The internal distribution of TRISO particles within SiC-matrix composite was observed by x-ray radiograph. From the analysis of the cross-section of SiC-matrix composite, the fracture of TRISO particles was investigated. In order to uniform distribution of TRISO particle embedded in the SiC matrix, SiC powder overcoating is considered. SiC matrix composite including TRISO was fabricated by hot pressing. FCM pallets with full density were obtained with Al 2 O 3 -Y 2 O 3 additive system. From the microstructure image, the effect of the sintering additive contents and sintering mechanism

  2. Closed Fuel Cycle Waste Treatment Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, J. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Collins, E. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Crum, J. V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, S. M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Garn, T. G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gombert, D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maio, V. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Matyas, J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Nenoff, T. M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Riley, B. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sevigny, G. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Thallapally, P. K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, J. H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-02-01

    This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form

  3. A Study on Rack Thickness Effect for Spent Fuel Pool Storage

    International Nuclear Information System (INIS)

    Kim, Mi Jin; Lee, Hee-Jae; Sohn, Dong-Seong

    2015-01-01

    For the effective storage of used fuel, the development of high performance neutron absorbing materials is needed. One of the major concern for the used fuel storage is the assurance to keep subcriticality of the storage rack and the high performance neutron absorbing material is the vital part to assure this requirement. According to NRC guide line, the k-effective of the spent fuel storage racks must not exceed 0.95. To ensure its safety, subcriticality analysis is required. Subcriticality analysis of the used storage in spent fuel pool have been performed by different authors. Criticality calculations for light water reactor spent fuel storage rack were carried out by Jae et al. They used AMPX-KENO IV code and considered the effect of rack pitch and rack thickness for consolidated fuel. The criticality analysis has performed at Gd 0.2 and 1 wt% according to thickness change. As thickness increases, the volume of the spent fuel pool rack increases. Therefore, absorbing material also increases according to thickness

  4. A Study on Rack Thickness Effect for Spent Fuel Pool Storage

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Mi Jin; Lee, Hee-Jae; Sohn, Dong-Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    For the effective storage of used fuel, the development of high performance neutron absorbing materials is needed. One of the major concern for the used fuel storage is the assurance to keep subcriticality of the storage rack and the high performance neutron absorbing material is the vital part to assure this requirement. According to NRC guide line, the k-effective of the spent fuel storage racks must not exceed 0.95. To ensure its safety, subcriticality analysis is required. Subcriticality analysis of the used storage in spent fuel pool have been performed by different authors. Criticality calculations for light water reactor spent fuel storage rack were carried out by Jae et al. They used AMPX-KENO IV code and considered the effect of rack pitch and rack thickness for consolidated fuel. The criticality analysis has performed at Gd 0.2 and 1 wt% according to thickness change. As thickness increases, the volume of the spent fuel pool rack increases. Therefore, absorbing material also increases according to thickness.

  5. Crítica ecológica al concepto de Desarrollo y nuevas alternativas desde América Latina

    Directory of Open Access Journals (Sweden)

    Ana Luíza Matos de Oliveira

    2014-12-01

    Full Text Available Este artículo derivado de investigación analiza el concepto de Desarrollo desde sus orígenes, presentando la crítica ecológica a este concepto, en especial por parte de la teoría del decrecimiento, muestra algunas contribuciones de los pueblos originarios a las constituciones de países de tendencia de izquierda en América Latina en los años recientes y problematiza la actual inserción de América Latina en el comercio internacional. Se encontró que contrarrestar a la fuerza económica, política y discursiva del agronegocio y del “progreso”, apoyado en el “desarrollo” no es algo simple. En América Latina hay tensiones entre el concepto de desarrollo convencional e intentos por construir un nuevo camino, un camino latinoamericano. Hay resistencias entre la (reprimarización y la inserción internacional de América Latina, también existen tensiones en el campo, conuna disputa entre el agronegocio como modelo de desarrollo, modernidad y progreso y otras formas de relacionarse con el medio ambiente y con el otro. Se concluye que desde América Latina es posible crear un nuevo camino, no necesariamente en aras de imitar el mundo desarrollado, sino considerando los limites ecológicos y sociales del capitalismo y las contribuciones de los pueblos originarios.

  6. 'Diesel regenerativ' as fuel for passenger cars

    Energy Technology Data Exchange (ETDEWEB)

    Zimon, Anja; Krahl, Juergen [Coburg Univ. of Applied Sciences and Arts (Germany); Schroeder, Olaf; Fey, Barbara; Munack, Axel [Thuenen Institute, Braunschweig (Germany); Bockey, Dieter [Union for the Promotion of Oil and Protein Plants, Berlin

    2013-06-01

    Among the multitude of possible biofuels, hydrotreated vegetable oil (HVO) presents one possible means of partially replacing diesel fuel. In the project presented here, HVO was used with an admixture of biodiesel in the amount of 2% and 7% in two different vehicle fleets. HVO and biodiesel were made from domestic rapeseed oil. Both fuels trade under the name Diesel regenerative. The test vehicles were cars of the emission standards Euro 3 to Euro 6 that had been previously fueled with fossil diesel fuel, each for different periods. All vehicles were tested for regulated emissions at the beginning and the end of the project. In summary, emission reductions for hydrocarbons, carbon monoxide and particulate matter were identified for Diesel regenerative in comparison to fossil diesel fuel. However, nitrogen oxides were slightly increased for Diesel regenerative. Until now, this increase was only known in the literature to be associated with paraffinic fuel exhaust gases such as GTL. Moreover, decreases in nitrogen oxide when using pure HVO versus DF were published for utility vehicles (Warnecke et al., 2012). (orig.)

  7. Correlative Microscopy of Alpha Prime Precipitation in Neutron-Irradiated Fe-Cr-Al Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, Samuel A. [Univ. of Wisconsin, Madison, WI (United States)

    2016-12-01

    Fe-Cr-Al alloys are currently being considered for accident tolerant light water reactor fuel cladding applications due to their superior high temperature oxidation and corrosion resistance compared to Zr-based alloys. This work represents the current state-of-the-art on both techniques for analysis of α' precipitate microstructures and the processes and mechanisms governing its formation in neutron-irradiated Fe-Cr-Al alloys.

  8. Furfural and other furane-derivates as diesel fuel additives. Furfural und andere Furanderivate als Zusatz zum Dieselkraftstoff

    Energy Technology Data Exchange (ETDEWEB)

    Nemling, O.

    1982-01-01

    Alternative energies such as fuel additives or fuel-substitutes are on their way upwards as they may help to reduce the dependency on crude and crude-based fuels. A number of solutions has already been indicated for carburettor engines whereas a lot more of research and development is needed to find a safe alternative to diesel fuels. Possibilities of using furfural, furfuralcohol and furan-2-carbonic acid-methylester as additives to diesel or as a substitute were investigated. A survey on the characteristic properties of these substances is given, then several tests are set up in order to describe the properties of diesel additive mixtures at low temperatures. Based on these tests two mixtures are proposed both with a furfural/butanole mixture. Furfural is suitable as diesel additive, a solubilizer is needed though.

  9. Nafion®/H-ZSM-5 composite membranes with superior performance for direct methanol fuel cells

    NARCIS (Netherlands)

    Yildirim, M.H.; Curos, Anna Roca; Motuzas, Julius; Motuzas, J.; Julbe, Anne; Stamatialis, Dimitrios; Wessling, Matthias

    2009-01-01

    Solution cast composite direct methanol fuel cell membranes (DEZ) based on DE2020 Nafion® dispersion and in-house prepared H-ZSM-5 zeolites with different Si/Al ratios were prepared and thoroughly characterized for direct methanol fuel cell (DMFC) applications. All composite membranes have indeed

  10. 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

  11. Corrosion of non-irradiated UAl{sub x}-Al fuel in the presence of clay pore solution. A quantitative XRD secondary phase analysis applying the DDM method

    Energy Technology Data Exchange (ETDEWEB)

    Neumann, Andreas [Halle-Wittenberg Univ. (Germany). Dept. of Mineralogy and Geochemistry; RWTH Aachen Univ. (Germany). Inst. of Crystallography; Klinkenberg, Martina; Curtius, Hildegard [Forschungszentrum Juelich GmbH (Germany). Inst. of Energy and Climate Research, IEK-6 Nuclear Waste Management

    2017-04-01

    Corrosion experiments with non-irradiated metallic UAl{sub x}-Al research reactor fuel elements were carried out in autoclaves to identify and quantify the corrosion products. Such compounds, considering the long-term safety assessment of final repositories, can interact with the released inventory and this constitutes a sink for radionuclide migration in formation waters. Therefore, the metallic fuel sample was subjected to clay pore solution to investigate its process of disintegration by analyzing the resulting products and the remnants, i.e. the secondary phases. Due to the fast corrosion rate a full sample disintegration was observed within the experimental period of 1 year at 90 C. The obtained solids were subdivided into different grain size fractions and prepared for analysis. The elemental analysis of the suspension showed that, uranium and aluminum are concentrated in the solids, whereas iron was mainly dissolved. Non-ambient X-ray diffraction (XRD) combined with the derivative difference minimization (DDM) method was applied for the qualitative and quantitative phase analysis (QPA) of the secondary phases. Gypsum and hemihydrate (bassanite), residues of non-corroded nuclear fuel, hematite, and goethite were identified. The quantitative phase analysis showed that goethite is the major crystalline phase. The amorphous content exceeded 80 wt% and hosted the uranium. All other compounds were present to a minor content. The obtained results by XRD were well supported by complementary scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS) analysis.

  12. Production of Solar Fuels by Photoelectrochemical Conversion of Carbon Dioxide

    OpenAIRE

    Irtem, Ibrahim Erdem

    2017-01-01

    Growing global emission of carbon dioxide gas (CO2) reflects the world’s energy dependence on fossil fuels. The conversion of CO2 emission into value-added products, like fuels completes a circular CO2 economy which requires a renewable energy conversion and storage system. Amongst a few, photo/electrochemistry has been particularly appealing thanks to its energy efficiency and enormous potential for industrial applications. Formic acid (HCOOH) production from CO2 reduction appears as an al...

  13. High density dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, G.L.

    1996-01-01

    A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm 3 of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm -3 with U 3 Si 2 as fuel. High-density uranium compounds offer no real density advantage over U 3 Si 2 and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U 3 Si has approximately a 30% higher uranium density but the density of the U 6 X compounds would yield the factor 1.5 needed to achieve 9 g cm -3 uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure α-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic γ phase at low temperatures where normally α phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing

  14. Mercadotecnia internacional: instrumento para el comercio global

    Directory of Open Access Journals (Sweden)

    Elizabeth Guadalupe Ramírez Moreno

    2013-07-01

    Full Text Available La investigación tiene por objeto analizar la mercadotecnia internacional y sus contribuciones al comercio internacional. En muchas empresas toman la decisión de expandir su producción hacia otro u otros países. Así mismo sin embargo hay factores que se deben de tomar en cuenta antes de competir en un mercado hacia el extranjero. Todos los temas referentes del marketing internacional se han expandido en su ámbito internacional.

  15. El burro cuántico de Sancho Panza (Los Nobel en Física 2003

    Directory of Open Access Journals (Sweden)

    Salvador Galindo Uribarri

    2004-01-01

    Full Text Available EL premio Nobel en Física 2003 fue compartido por Leggett, Abrikosov y Ginzburg, por sus contribuciones al entendimiento de fenómenos a bajas temperaturas. Este ensayo explica, a un nivel elemental, sus teorías. Además este trabajo presenta una serie de eventos –relacionados con las teorías mencionadas– que forman una historia de tres animales cuánticos: un gato, un calamar y un burro.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  17. Implementing Monitored Natural Attenuation and Expediting Closure at Fuel-Release Sites

    Science.gov (United States)

    2004-08-01

    gasoline, kerosene, diesel, and jet fuel (e.g., Jamison et al., 1975; Atlas , 1981, 1984, and 1988; Young, 1984; Bartha , 1986; Wilson et al., 1986 and...Supporting data. Gas Research Institute, Chicago, Illinois. Atlas , R. M. 1981. Microbial degradation of petroleum hydrocarbons - an Environmental... Microbial Ecology 12:155-172 Battelle. 1984. Chemical Attenuation Rates Coefficients, and Constants in Leachate Migration. Vol I: Critical Review

  18. Helium in inert matrix dispersion fuels

    International Nuclear Information System (INIS)

    Veen, A. van; Konings, R.J.M.; Fedorov, A.V.

    2003-01-01

    The behaviour of helium, an important decay product in the transmutation chains of actinides, in dispersion-type inert matrix fuels is discussed. A phenomenological description of its accumulation and release in CERCER and CERMET fuel is given. A summary of recent He-implantation studies with inert matrix metal oxides (ZrO 2 , MgAl 2 O 4 , MgO and Al 2 O 3 ) is presented. A general picture is that for high helium concentrations helium and vacancy defects form helium clusters which convert into over-pressurized bubbles. At elevated temperature helium is released from the bubbles. On some occasions thermal stable nano-cavities or nano-pores remain. On the basis of these results the consequences for helium induced swelling and helium storage in oxide matrices kept at 800-1000 deg. C will be discussed. In addition, results of He-implantation studies for metal matrices (W, Mo, Nb and V alloys) will be presented. Introduction of helium in metals at elevated temperatures leads to clustering of helium to bubbles. When operational temperatures are higher than 0.5 melting temperature, swelling and helium embrittlement might occur

  19. Recycling process of Mn-Al doped large grain UO2 pellets

    International Nuclear Information System (INIS)

    Nam, Ik Hui; Yang, Jae Ho; Rhee, Young Woo; Kim, Dong Joo; Kim, Jong Hun; Kim, Keon Sik; Song, Kun Woo

    2010-01-01

    To reduce the fuel cycle costs and the total mass of spent light water reactor (LWR) fuels, it is necessary to extend the fuel discharged burn-up. Research on fuel pellets focuses on increasing the pellet density and grain size to increase the uranium contents and the high burnup safety margins for LWRs. KAERI are developing the large grain UO 2 pellet for the same purpose. Small amount of additives doping technology are used to increase the grain size and the high temperature deformation of UO 2 pellets. Various promising additive candidates had been developed during the last 3 years and the MnO-Al 2 O 3 doped UO 2 fuel pellet is one of the most promising candidates. In a commercial UO 2 fuel pellet manufacturing process, defective UO 2 pellets or scraps are produced and those should be reused. A common recycling method for defective UO 2 pellets or scraps is that they are oxidized in air at about 450 .deg. C to make U 3 O 8 powder and then added to UO 2 powder. In the oxidation of a UO 2 pellet, the oxygen propagates along the grain boundary. The U 3 O 8 formation on the grain boundary causes a spallation of the grains. So, size and shape of U 3 O 8 powder deeply depend on the initial grain size of UO 2 pellets. In the case of Mn-Al doped large grain pellets, the average grain size is about 45μm and about 5 times larger than a typical un-doped UO 2 pellet which has grain size of about 8∼10μm. That big difference in grain size is expected to cause a big difference in recycled U 3 O 8 powder morphology. Addition of U 3 O 8 to UO 2 leads to a drop in the pellet density, impeding a grain growth and the formation of graph- like pore segregates. Such degradation of the UO 2 pellet properties by adding the recycled U 3 O 8 powder depend on the U 3 O 8 powder properties. So, it is necessary to understand the property and its effect on the pellet of the recycled U 3 O 8 . This paper shows a preliminary result about the recycled U 3 O 8 powder which was obtained by

  20. Improving performance with accident tolerant-fuels

    International Nuclear Information System (INIS)

    Gomes, Daniel S.; Muniz, Rafael O.R.; Giovedi, Claudia; Universidade de São Paulo

    2017-01-01

    After the Fukushima reactor accident, efforts to improve risk management in nuclear operations have included the intensification of research on accident-tolerant fuels (ATFs). In this investigation, the physical properties of recently developed ATFs were compared with those of the current standard fuel, UO 2 - Zr. The goals for innovative fuel design include a rigorous characterization of the thermal, mechanical, and chemical considerations. The intentions are to lengthen the burnup cycle, raise the power density, and improve safety. Fuels must have a high uranium density - above that supported by UO 2 - and possess a coating that exhibits better oxidation resistance than Zircaloys. ATFs such as U 3 Si 2 , UN, and UC contain a higher uranium density and thermal conductivity than UO 2 , providing significant benefits. The ideal combination of fuel and cladding must increase performance in a loss-of-coolant accident. However, U 3 Si 2 , UN, and UC have a disadvantage; their respective swelling rates are higher than that of UO 2 . These ATFs also have thermal conductivities approximately four times higher than that of UO 2 . A study was conducted investigating the hydrogen generated by the oxidation of zirconium alloys in contact with steam using cladding options such as Fe-Cr-Al and silicon carbide. It was confirmed that ferritic alloys offer a better response under severe conditions, because of their mechanical properties as creep rate. The findings of this study indicate that advanced fuels should replace UO 2 - Zr as the fuel system of choice. (author)

  1. Improving performance with accident tolerant-fuels

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

    After the Fukushima reactor accident, efforts to improve risk management in nuclear operations have included the intensification of research on accident-tolerant fuels (ATFs). In this investigation, the physical properties of recently developed ATFs were compared with those of the current standard fuel, UO{sub 2} - Zr. The goals for innovative fuel design include a rigorous characterization of the thermal, mechanical, and chemical considerations. The intentions are to lengthen the burnup cycle, raise the power density, and improve safety. Fuels must have a high uranium density - above that supported by UO{sub 2} - and possess a coating that exhibits better oxidation resistance than Zircaloys. ATFs such as U{sub 3}Si{sub 2}, UN, and UC contain a higher uranium density and thermal conductivity than UO{sub 2}, providing significant benefits. The ideal combination of fuel and cladding must increase performance in a loss-of-coolant accident. However, U{sub 3}Si{sub 2}, UN, and UC have a disadvantage; their respective swelling rates are higher than that of UO{sub 2}. These ATFs also have thermal conductivities approximately four times higher than that of UO{sub 2}. A study was conducted investigating the hydrogen generated by the oxidation of zirconium alloys in contact with steam using cladding options such as Fe-Cr-Al and silicon carbide. It was confirmed that ferritic alloys offer a better response under severe conditions, because of their mechanical properties as creep rate. The findings of this study indicate that advanced fuels should replace UO{sub 2} - Zr as the fuel system of choice. (author)

  2. Multiple Surrogate Modeling for Wire-Wrapped Fuel Assembly Optimization

    International Nuclear Information System (INIS)

    Raza, Wasim; Kim, Kwang-Yong

    2007-01-01

    In this work, shape optimization of seven pin wire wrapped fuel assembly has been carried out in conjunction with RANS analysis in order to evaluate the performances of surrogate models. Previously, Ahmad and Kim performed the flow and heat transfer analysis based on the three-dimensional RANS analysis. But numerical optimization has not been applied to the design of wire-wrapped fuel assembly, yet. Surrogate models are being widely used in multidisciplinary optimization. Queipo et al. reviewed various surrogates based models used in aerospace applications. Goel et al. developed weighted average surrogate model based on response surface approximation (RSA), radial basis neural network (RBNN) and Krigging (KRG) models. In addition to the three basic models, RSA, RBNN and KRG, the multiple surrogate model, PBA also has been employed. Two geometric design variables and a multi-objective function with a weighting factor have been considered for this problem

  3. 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

  4. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    Feraday, M.A.

    1993-01-01

    For a period of about 10 years AECL had a significant program looking into the possibility of developing U 3 Si as a high density replacement for the UO 2 pellet fuel in use in CANDU power reactors. The element design consisted of a Zircaloy-clad U 3 Si rod containing suitable voidage to accommodate swelling. We found that the binary U 3 Si could not meet the defect criterion for our power reactors, i.e., one month in 300 degree C water with a defect in the sheath and no significant damage to the element. Since U 3 Si could not do the job, a new corrosion resistant ternary U-Si-Al alloy was developed and patented. Fuel elements containing this alloy came close to meeting the defect criterion and showed slightly better irradiation stability than U 3 Si. Shortly after this, the program was terminated for other reasons. We have made much of this experience available to the Low Enrichment Fuel Development Program and will be glad to supply further data to assist this program

  5. Laboratory Investigations of Ni-Al Coatings Exposed to Conditions Simulating Biomass Firing

    DEFF Research Database (Denmark)

    Wu, Duoli; Okoro, Sunday Chukwudi; Dahl, Kristian Vinter

    2016-01-01

    Fireside corrosion is a key problem when using biomass fuels in power plants. A possible solution is to apply corrosion resistant coatings. The present paper studies the corrosion and interdiffusion behaviour of a Ni-Al diffusion coating on austenitic stainless steel (TP347H). Ni-Al coatings were...... prepared by electrolytic deposition of nickel followed by pack aluminizing performed at 650˚C. A uniform and dense Ni-Al coating with an outer layer of Ni2Al3 and an inner Ni layer was formed. Samples were exposed to 560°C for 168h in an atmosphere simulating biomass combustion. This resulted in localized...... corrosion attack. Interdiffusion was studied by isothermal heat treatment in static air at 650˚C or 700˚C for up to 3000h. The Ni2Al3 gradually transformed into NiAl and Ni3Al during the interdiffusion process. Porosity developed at the interface between the Ni-Al coating and the Ni layer and expanded...

  6. Main results and status of the development of LEU fuel for Russian research reactors

    International Nuclear Information System (INIS)

    Vatulin, A.; Morozov, A.; Suprun, V.; Dobrikova, I.

    2005-01-01

    VNIINM develops low enrichment uranium (LEU) fuel on base U-Mo alloys and a novel design of pin-type fuel elements. The development is carried out both for existing reactors, and for new advanced designs of reactors. The work is carried on the following main directions: - irradiate LEU U-Mo dispersion fuel (the uranium density up to 6,0 g/cm 3 ) in two Russian research reactors: MIR (RIAR, Dimitrovgrad) as pin type fuel mini-elements and in WWR-M (PINP, Gatchina) within full-scaled fuel assembly (FA) with pin type fuel elements; - finalize development of design and fabrication process of IRT type FA with pin type fuel elements; - develop methods of reducing of U-Mo fuel --Al matrix interaction under irradiation; - develop fabricating methods of fuel elements on base of monolithic U-Mo fuel. The paper generally reviews the results of calculation, design and technology investigations accomplished by now. (author)

  7. First commercial fuel cells in emergency power supplies; Erste kommerzielle Brennstoffzellen im Notstrombereich. Praxistest als USV fuer eine Mobilfunkbasisstation

    Energy Technology Data Exchange (ETDEWEB)

    Trachte, U.

    2008-07-01

    This article takes a look at the testing of a fuel-cell-based uninterruptible power supply for a cell-phone base station. The working principles of various types of fuel cell are examined and their areas of application are discussed. The use of polymer-electrolyte-membrane (PEM) fuel cells as a back-up power supply for a cell-phone base station was examined by the University of Applied Sciences in Lucerne, Switzerland. Several designs and models of fuel cells were examined. The installation of the cell-phone station is described and experience gained from the use of fuel cells in practice is presented and discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  9. Comparison of platinum/MWCNTs Nanocatalysts Synthesis Processes for Proton Exchange Membrane Fuel Cells

    Science.gov (United States)

    Liu, Xuan

    Due to the growing concerns on the depletion of petroleum based energy resources and climate change; fuel cell technologies have received much attention in recent years. Proton exchange membrane fuel cell (PEMFCs) features high energy conversion efficiency and nearly zero greenhouse gas emissions, because of its combination of the hydrogen oxidation reaction (HOR) at anode side and oxygen reduction reaction (ORR) at cathode side. Synthesis of Pt nanoparticles supported on multi walled carbon nanotubes (MWCNTs) possess a highly durable electrochemical surface area (ESA) and show good power output on proton exchange membrane (PEM) fuel cell performance. Platinum on multi-walled carbon nanotubes (MWCNTs) support were synthesized by two different processes to transfer PtCl62- from aqueous to organic phase. While the first method of Pt/MWCNTs synthesis involved dodecane thiol (DDT) and octadecane thiol (ODT) as anchoring agent, the second method used ammonium lauryl sulfate (ALS) as the dispersion/anchoring agent. The particle size and distribution of platinum were examined by high-resolution transmission electron microscope (HRTEM). The TEM images showed homogenous distribution and uniform particle size of platinum deposited on the surface of MWCNTs. The single cell fuel cell performance of the Pt/MWCNTs synthesized thiols and ALS based electrode containing 0.2 (anode) and 0.4 mg (cathode) Pt.cm-2 were evaluated using Nafion-212 electrolyte with H2 and O2 gases at 80 °C and ambient pressure. The catalyst synthesis with ALS is relatively simple compared to that with thiols and also showed higher performance (power density reaches about 1070 mW.cm -2). The Electrodes with Pt/MWCNTs nanocatalysts synthesized using ALS were characterized by cyclic voltammetry (CV) for durability evaluation using humidified H2 and N2 gases at room temperature (21 °C) along with commercial Pt/C for comparison. The ESA measured by cyclic voltammetry between 0.15 and 1.2 V showed significant

  10. Cermet fuel behaviour and properties in ADS reactors

    International Nuclear Information System (INIS)

    Haas, D.; Fernandez, A.; Staicu, D.; Somers, J.; Maschek, W.; Chen, X.

    2007-01-01

    irradiation conditions in these two new fuel irradiation experiments. References 1. Haas and al. ICENES Conference 2005 Brussels 2. Jaecki and al. GLOBAL 2005 3. Scaffidi-Argentina and al. ICAPP 2006 4. W. Maschek, X. Chen, C. Matzerath Boccaccini, A. Rineiski, J. Wallenius , Sobolev, P. Smith, R. Thetford, J.P. Ottaviani, S.Pillon, D. Haas, First Results of Safety Analyses for ADTs with CERCER and CERMET Fuels within the EUROTRANS-AFTRA Program, Ninth Information Exchange Meeting, Actinide and Fission Product Partitioning and Transmutation, France, 25-29 September 2006

  11. Adsorptive on-board desulfurization over multiple cycles for fuel-cell-based auxiliary power units operated by different types of fuels

    Science.gov (United States)

    Neubauer, Raphael; Weinlaender, Christof; Kienzl, Norbert; Bitschnau, Brigitte; Schroettner, Hartmuth; Hochenauer, Christoph

    2018-05-01

    On-board desulfurization is essential to operate fuel-cell-based auxiliary power units (APU) with commercial fuels. In this work, both (i) on-board desulfurization and (ii) on-board regeneration performance of Ag-Al2O3 adsorbent is investigated in a comprehensive manner. The herein investigated regeneration strategy uses hot APU off-gas as the regeneration medium and requires no additional reagents, tanks, nor heat exchangers and thus has remarkable advantages in comparison to state-of-the-art regeneration strategies. The results for (i) show high desulfurization performance of Ag-Al2O3 under all relevant operating conditions and specify the influence of individual operation parameters and the combination of them, which have not yet been quantified. The system integrated regeneration strategy (ii) shows excellent regeneration performance recovering 100% of the initial adsorption capacity for all investigated types of fuels and sulfur heterocycles. Even the adsorption capacity of the most challenging dibenzothiophene in terms of regeneration is restored to 100% over 14 cycles of operation. Subsequent material analyses proved the thermal and chemical stability of all relevant adsorption sites under APU off-gas conditions. To the best of our knowledge, this is the first time 100% regeneration after adsorption of dibenzothiophene is reported over 14 cycles of operation for thermal regeneration in oxidizing atmospheres.

  12. 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)

  13. Investigation of electrorefining of metallic alloy fuel onto solid Al cathodes

    International Nuclear Information System (INIS)

    Cassayre, L.; Malmbeck, R.; Masset, P.; Rebizant, J.; Serp, J.; Soucek, P.; Glatz, J.-P.

    2007-01-01

    This work concerned the electrorefining of UZr and UPuZr alloys on a solid aluminium cathode, in the LiCl-KCl eutectic melt containing U 3+ , Pu 3+ , Np 3+ , Zr 2+ or Zr 4+ , Am 3+ , Nd 3+ , Y 3+ , Ce 3+ and Gd 3+ chlorides. During constant current electrolyses, the use of a cathodic cut-off potential (-1.25 V versus Ag/AgCl) allowed to selectively deposit actinides (mainly U), while lanthanides remained in the salt. The aim was to determine the maximal load achievable on a single aluminium electrode. The total exchange charge was 4300 C, which represents the deposition of 3.72 g of actinides in 4.17 g Al, yielding a composition of 44.6 wt% An in Al. It was shown that the melting of the cathode contributed to increase the total amount of actinides deposited on the aluminium

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

    Cho, Tae-Won; Sohn, Dong-Seong

    2014-01-01

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

  16. Production of jet fuel from alternative source

    Energy Technology Data Exchange (ETDEWEB)

    Eller, Zoltan; Papp, Anita; Hancsok, Jenoe [Pannonia Univ., Veszprem (Hungary). MOL Dept. of Hydrocarbon and Coal Processing

    2013-06-01

    Recent demands for low aromatic content jet fuels have shown significant increase in the last 20 years. This was generated by the growing of aviation. Furthermore, the quality requirements have become more aggravated for jet fuels. Nowadays reduced aromatic hydrocarbon fractions are necessary for the production of jet fuels with good burning properties, which contribute to less harmful material emission. In the recent past the properties of gasolines and diesel gas oils were continuously severed, and the properties of jet fuels will be more severe, too. Furthermore, it can become obligatory to blend alternative components into jet fuels. With the aromatic content reduction there is a possibility to produce high energy content jet fuels with the desirable properties. One of the possibilities is the blending of biocomponents from catalytic hydrogenation of triglycerides. Our aim was to study the possibilities of producing low sulphur and aromatic content jet fuels in a catalytic way. On a CoMo/Al{sub 2}O{sub 3} catalyst we studied the possibilities of quality improving of a kerosene fraction and coconut oil mixture depending on the change of the process parameters (temperature, pressure, liquid hourly space velocity, volume ratio). Based on the quality parameters of the liquid products we found that we made from the feedstock in the adequate technological conditions products which have a high smoke point (> 35 mm) and which have reduced aromatic content and high paraffin content (90%), so these are excellent jet fuels, and their stack gases damage the environment less. (orig.)

  17. Composite fuel behaviour under and after irradiation

    International Nuclear Information System (INIS)

    Dehaudt, P.; Mocellin, A.; Eminet, G.; Caillot, L.; Delette, G.; Bauer, M.; Viallard, I.

    1997-01-01

    Two kinds of composite fuels have been irradiated in the SILOE reactor. They are made of UO 2 particles dispersed in a molybdenum metallic (CERMET) or a MgAl 2 O 4 ceramic (CERCER) matrix. The irradiation conditions have allowed to reach a 50000 MWd/t U burn-up in these composite fuels after a hundred equivalent full power days long irradiation. The irradiation is controlled by a continuous measure of the pellet centre line temperature. It allows to have information about the TANOX rods thermal behaviour and the fuels thermal conductivities in comparing the centre line temperature versus linear power curves among themselves. Our results show that the CERMET centre line temperature is much lower than the CERCER and UO 2 ones: 520 deg. C against 980 deg. C at a 300W/cm linear power. After pin puncturing tests the rods are dismantled to recover each fuel pellet. In the CERCER case, the cladding peeling off has revealed that the fuel came into contact with the cladding and that some of the pellets were linked together. Optical microscopy observations show a changing of the MgAl 2 O 4 matrix state around the UO 2 particles at the pellets periphery. This transformation may have caused a swelling and would be at the origin of the pellet-cladding and the pellet-pellet interactions. No specific damage is seen after irradiation. The CERMET pellets are not cracked and remain as they were before irradiation. The CERCER crack network is slightly different from that observed in UO 2 . Kr retention was evaluated by annealing tests under vacuum at 1580 deg. C or 1700 deg. C for 30 minutes. The CERMET fission gas release is lower than the CERCER one. Inter- and intragranular fission gas bubbles are observed in the UO 2 particles after heat treatments. The CERCER pellet periphery has also cracked and the matrix has transformed again around UO 2 particles to present a granular and porous aspect. (author). 4 refs, 6 figs, 2 tabs

  18. Experimental studies of thermal and chemical interactions between oxide and silicide nuclear fuels with water

    Energy Technology Data Exchange (ETDEWEB)

    farahani, A.A.; Corradini, M.L. [Univ. of Wisconsi, Madison, WI (United States)

    1995-09-01

    Given some transient power/cooling mismatch is a nuclear reactor and its inability to establish the necessary core cooling, energetic fuel-coolant interactions (FCI`s commonly called `vapor explosions`) could occur as a result of the core melting and coolant contact. Although a large number of studies have been done on energetic FCI`s, very few experiments have been performed with the actual fuel materials postulated to be produced in severe accidents. Because of the scarcity of well-characterized FCI data for uranium allows in noncommercial reactors (cermet and silicide fuels), we have conducted a series of experiments to provide a data base for the foregoing materials. An existing 1-D shock-tube facility was modified to handle depleted radioactive materials (U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al). Our objectives have been to determine the effects of the initial fuel composition and temperature and the driving pressure (triggering) on the explosion work output, dynamic pressures, transient temperatures, and the hydrogen production. Experimental results indicate limited energetics, mainly thermal interactions, for these fuel materials as compared to aluminum where more chemical reactions occur between the molten aluminum and water.

  19. Effect of additives in sintering UO2-7wt%Gd2O3 fuel pellets

    International Nuclear Information System (INIS)

    Santos, L.R.; Riella, H.G.

    2009-01-01

    Gadolinium has been used as burnable poison for reactivity control in modern PWRs. The incorporation of Gd 2 O 3 powder directly into the UO 2 powder enables longer fuel cycles and optimized fuel utilization. Nevertheless, processing by this method leads to difficulties while obtaining sintered pellets with the minimum required density. The process for manufacturing UO 2 - Gd 2 O 3 generates scraps that should be reused. The main scraps are green and sintered pellets, which must be calcined to U 3 O 8 to return to the fabrication process. Also, the incorporation of Gd 2 O 3 in UO 2 requires the use of an additive to improve the sintering process, in order to achieve the physical properties specified for the mixed fuel, mainly density and microstructure. This paper describes the effect of the addition of fabrication scraps on the properties of the UO 2 -Gd 2 O 3 fuel. Aluminum hydroxide Al(OH) 3 was also incorporated to the fuel as a sintering aid. The results shown that the use of 2000 ppm of Al(OH) 3 as additive allow to fabricate good pellets with up to 10 wt% of recycled scraps. (author)

  20. Formulation, Casting, and Evaluation of Paraffin-Based Solid Fuels Containing Energetic and Novel Additives for Hybrid Rockets

    Science.gov (United States)

    Larson, Daniel B.; Desain, John D.; Boyer, Eric; Wachs, Trevor; Kuo, Kenneth K.; Borduin, Russell; Koo, Joseph H.; Brady, Brian B.; Curtiss, Thomas J.; Story, George

    2012-01-01

    This investigation studied the inclusion of various additives to paraffin wax for use in a hybrid rocket motor. Some of the paraffin-based fuels were doped with various percentages of LiAlH4 (up to 10%). Addition of LiAlH4 at 10% was found to increase regression rates between 7 - 10% over baseline paraffin through tests in a gaseous oxygen hybrid rocket motor. Mass burn rates for paraffin grains with 10% LiAlH4 were also higher than those of the baseline paraffin. RDX was also cast into a paraffin sample via a novel casting process which involved dissolving RDX into dimethylformamide (DMF) solvent and then drawing a vacuum on the mixture of paraffin and RDX/DMF in order to evaporate out the DMF. It was found that although all DMF was removed, the process was not conducive to generating small RDX particles. The slow boiling generated an inhomogeneous mixture of paraffin and RDX. It is likely that superheating the DMF to cause rapid boiling would likely reduce RDX particle sizes. In addition to paraffin/LiAlH4 grains, multi-walled carbon nanotubes (MWNT) were cast in paraffin for testing in a hybrid rocket motor, and assorted samples containing a range of MWNT percentages in paraffin were imaged using SEM. The fuel samples showed good distribution of MWNT in the paraffin matrix, but the MWNT were often agglomerated, indicating that a change to the sonication and mixing processes were required to achieve better uniformity and debundled MWNT. Fuel grains with MWNT fuel grains had slightly lower regression rate, likely due to the increased thermal conductivity to the fuel subsurface, reducing the burning surface temperature.

  1. 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)

  2. Lifecycle Cost Assessment of Fuel Cell Technologies for Soldier Power System Applications. Paper and Presentation for the 43rd Power Sources Conference held 8-9 July 2008, Philadelphia, PA

    Science.gov (United States)

    2008-07-09

    PEMFC in Federal Markets,” 2007 Fuel Cell Seminar, San Antonio, TX, 17 October 2007. 7. Fok, K., “Metal Hydride Fuel Cells: Increases in Power...Lauderdale, FL, March 17-20, 2008. 10. Zhao J., et al, “Reclaim/recycle of Pt/C catalysts for PEMFC ,” Energy Conversion and Management, vol. 48...hydrogen PEMFC or SOFC systems – Baratto et al, Journal of Power Sources – Citigroup, Dist. Telecom Backup – Battelle, Fuel Cell Seminar 2007 • Fuel

  3. French LEU fuel for research reactor with emphasis on the Osiris experience of core conversion and reactor operation with the new fuel

    International Nuclear Information System (INIS)

    Cerles, J.-M.

    1981-09-01

    One of the various activities carried out in France concerned with the design, fabrication and development of nuclear fuels was the development by the CEA of a plate type fuel (Caramel fuel). A Caramel fuel element is in the form of a plate consisting of two tight covering zircaloy sheets in which the UO 2 platelets are confined themselves within the network of a zircaloy grid. The plane geometry provides an effective means of overcoming the drawback of poor uranium oxide conductivity, and makes it possible to combine high specific power with low fuel temperature. The chief advantages of this fuel are the following: it is a very low enriched fuel. It can be used in research reactors demanding high volumetric powers and neutron fluxes, with a required enrichment significantly lower than 20% 235 U. The difference between the densities of UO 2 matrix and U-Al, 10.3 and 1.6 g/cm respectively, leads to a higher uranium charge, making it possible to reduce the enrichment to between 3 and 10%. Owing to fuel dispersion, any loss of tightness only puts a small amount of fissile material in contact with the coolant, thus limiting any contamination of the primary circuit. Another safety factor is the operating temperature, which is considerably lower than the temperature at which fission gases are liberated

  4. CONTRIBUCIÓN DEL SISTEMA DE PENSIONES PRIVADO DE CAPITALIZACIÓN INDIVIDUAL AL DESARROLLO DEL MERCADO DE CAPITALES EN BOLIVIA, 1997– 2009

    Directory of Open Access Journals (Sweden)

    Pamela Córdova Olivera

    2011-01-01

    Full Text Available El establecimiento del Sistema de Pensiones de Capitalización Individual en Bolivia aprobado en el año 1996, constituye el reemplazo de un sistema tradicional de reparto colectivo, de contribuciones comunes y administradas por el Estado, por un sistema de capitalización individual, de contribuciones definidas y administradas por el sector privado. A más de una década de su establecimiento, es posible realizar una estimación más completa y precisa de sus resultados e impactos. Este estudio presenta una evaluación cualitativa y cuantitativa del impacto de los Fondos de Pensiones de Capitalización Individual sobre el desarrollo del Mercado de Capitales en Bolivia en lo que respecta a su regulación, composición, tamaño y profundidad financiera considerando alternativamente la participación del sector privado y público. Como base de demostración empírica, se plantea un modelo econométrico basado en el análisis de series de tiempo cointegradas que hacen hincapié en los factores financieros concentrados en una variable de profundidad financiera, cuyo resultado confirma el importante aporte del Sistema de Pensiones Privado a la mejora en la calidad de regulación, tamaño, trasparencia y crecimiento del mercado financiero en Bolivia.

  5. RERTR program activities related to the development and application of new LEU fuels

    International Nuclear Information System (INIS)

    Travelli, A.

    1983-01-01

    The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U 3 Si 2 -Al and U 3 Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm 3 each year, from the current 1.7 g U/cm 3 to the 7.0 g U/cm 3 which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years

  6. The advanced neutron source three-element-core fuel grading

    International Nuclear Information System (INIS)

    Gehin, J.C.

    1995-01-01

    The proposed Advanced Neutron Source (ANS) pre-conceptual design consists of a two-element 330 MW f nuclear reactor fueled with highly-enriched uranium and is cooled, moderated, and reflected with heavy water. Recently, the ANS design has been changed to a three-element configuration in order to permit a reduction of the enrichment, if required, while maintaining or improving the thermal-hydraulic margins. The core consists of three annular fuel elements composed of involute-shaped fuel plates. Each fuel plate has a thickness of 1.27 mm and consists of a fuel meat region Of U 3 Si 2 -Al (50% enriched in one case that was proposed) and an aluminum filler region between aluminum cladding. The individual plates are separated by a 1.27 mm coolant channel. The three element core has a fuel loading of 31 kg of 235 U which is sufficient for a 17-day fuel cycle. The goal in obtaining a new fuel grading is to maximize important temperature margins. The limits imposed axe: (1) Limit the temperature drop over the cladding oxide layer to less than 119 degrees C to avoid oxide spallation. (2) Limit the fuel centerline temperature to less than 400 degrees C to avoid fuel damage. (3) Limit the cladding wall temperature to less than the coolant. incipient-boiling temperature to avoid coolant boiling. Other thermal hydraulic conditions, such as critical heat flux, are also considered

  7. Investigation of electrorefining of metallic alloy fuel onto solid Al cathodes

    Energy Technology Data Exchange (ETDEWEB)

    Cassayre, L. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany); Malmbeck, R. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany)]. E-mail: rikard.malmbeck@cec.eu.int; Masset, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany); Rebizant, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany); Serp, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany); Soucek, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany); Glatz, J.-P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany)

    2007-01-15

    This work concerned the electrorefining of UZr and UPuZr alloys on a solid aluminium cathode, in the LiCl-KCl eutectic melt containing U{sup 3+}, Pu{sup 3+}, Np{sup 3+}, Zr{sup 2+} or Zr{sup 4+}, Am{sup 3+}, Nd{sup 3+}, Y{sup 3+}, Ce{sup 3+} and Gd{sup 3+} chlorides. During constant current electrolyses, the use of a cathodic cut-off potential (-1.25 V versus Ag/AgCl) allowed to selectively deposit actinides (mainly U), while lanthanides remained in the salt. The aim was to determine the maximal load achievable on a single aluminium electrode. The total exchange charge was 4300 C, which represents the deposition of 3.72 g of actinides in 4.17 g Al, yielding a composition of 44.6 wt% An in Al. It was shown that the melting of the cathode contributed to increase the total amount of actinides deposited on the aluminium.

  8. Nuclear fuel pellet production method and nuclear fuel pellet

    International Nuclear Information System (INIS)

    Yuda, Ryoichi; Ito, Ken-ichi; Masuda, Hiroshi.

    1993-01-01

    In a method of manufacturing nuclear fuel pellets by compression-molding UO 2 powders followed by sintering, a sintering agent having a composition of about 40 to 80 wt% of SiO 2 and the balance of Al 2 O 3 , a sintering agent at a ratio of 10 to 500 ppm based on the total amount of UO 2 and UO 2 powders are mixed, compression molded and then sintered at a sintering temperature of about 1500 of 1800degC. The UO 2 particles have an average grain size of about 20 to 60μm, most of the crystal grain boundary thereof is coated with a glassy or crystalline alumina silicate phase, and the porosity is about 1 to 4 vol%. With such a constitution, the sintering agent forms a single liquid phase eutectic mixture during sintering, to promote a surface reaction between nuclear fuel powders by a liquid phase sintering mechanism, increase their density and promote the crystal growth. Accordingly, it is possible to lower the softening temperature, improve the creep velocity of the pellets and improve the resistance against pellet-clad interaction. (T.M.)

  9. 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.

  10. Anticorrosion ion implantation of fragments of zirconium fuel can specimens

    International Nuclear Information System (INIS)

    Kalin, B.A.; Osipov, V.V.; Volkov, N.V.; Khernov, V.Yu.

    2001-01-01

    Aimed at the study of specific features of oxide film formation in the initial stage of Eh110 and Eh635 alloy fuel can oxidation the modification of tubular specimen surfaces is performed using an ion mixing technique, and the structure of oxide films produced in a steam-water environment is investigated. Using the method of vacuum vapor deposition the outer surface of specimens is coated with alloying element films irradiated by a polyenergetic Ar + ion beam with a 10 keV mean energy up to radiation doses of (7-10) x 10 17 ion/cm 2 . Monatomic (Al, Fe, Cu, Cr, Mo, Sn) or diatomic (Al-Fe, Al-Mo, Al-Sn, Fe-Cu, Fe-Mo, Fe-Sn, Cr-Mo, Cr-Sn) implantation into a zirconium cladding occurs under irradiation effect. The positive influence of combined intrusion of Al and other elements is revealed. The presence of Al atoms enhances the oxide film structure. The least ZeO 2 film thickness is observed when alloying with molybdenum, Al-Fe, Al-Mo and Al-Sn [ru

  11. Investigations of uraniumsilicide-based dispersion fuels for the use of low enrichment uranium (LEU) in research and test reactors

    International Nuclear Information System (INIS)

    Nazare, S.

    1982-07-01

    The work presents at the outset, a review of the preparation and properties of uranium silicides (U 3 Si and U 3 Si 2 ) in so far as these are relevant for their use as dispersants in research reactor fuels. The experimental work deals with the preparation and powder metallurgical processing of Al-clad miniature fuel element plates with U 3 Si- und U 3 Si-Al up to U-densities of 6.0 g U/cm 3 . The compatibility of these silicides with the Al-matrix under equilibrium conditions (873 K) and the influence of the reaction on the dimensional stability of the miniplates is described and discussed. (orig.) [de

  12. Fuel Exhaling Fuel Cell.

    Science.gov (United States)

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  13. Arte, ciencia y ambiente construido : Precisiones y propuestas

    OpenAIRE

    Aliata, Fernando; Rosenfeld, Elías; Sessa, Emilio

    1999-01-01

    El presente documento fue elaborado por los autores, profesores de la Facultad de Arquitectura y Urbanismo de la Universidad Nacional de La Plata, como contribución a la discusión que se esta desarrollando en diversos ámbitos académicos de la facultad y del país sobre el estado del campo disciplinario y sus facetas productivas. Consideramos al presente como documento de trabajo esperando recibir opiniones y contribuciones sobre el mismo. Facultad de Arquitectura y Urbanismo

  14. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Guenther, R.J.; Johnson, A.B. Jr.; Lund, A.L.; Gilbert, E.R.

    1994-11-01

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl x , UAl x -Al and U 3 O 8 -Al cermets, U-5% fissium, UMo, UZrH x , UErZrH, UO 2 -stainless steel cermet, and U 3 O 8 -stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified

  15. Status Report on Irradiation Capsules Designed to Evaluate FeCrAl-UO2 Interactions

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-06-24

    This status report provides the background and current status of a series of irradiation capsules that were designed and are being built to test the interactions between candidate FeCrAl cladding for enhanced accident tolerant applications and prototypical enriched commercial UO2 fuel in a neutron radiation environment. These capsules will test the degree, if any, of fuel cladding chemical interactions (FCCI) between FeCrAl and UO2. The capsules are to be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory to burn-ups of 10, 30, and 50 GWd/MT with a nominal target temperature at the interfaces between the pellets and clad of 350°C.

  16. Contribuciones del Censo de Población y Viviendas 2012 al desarrollo socioeconómico del municipio Mayarí, Holguín, Cuba

    Directory of Open Access Journals (Sweden)

    Rafael Ángel Salazar Martínez

    2014-06-01

    Full Text Available El presente trabajo establece las coordenadas para la utilización eficiente de los resultados que arroje el Censo del Población y Viviendas del 2012, en función del desarrollo socioeconómico del municipio Mayarí, territorio que implementa desde ese mismo año un Programa Integral de Desarrollo Municipal (PIDM, cuya significación estratégica radica en el hecho de constituir la experiencia piloto a nivel nacional, para su generalización hacia otros municipios del país, respondiendo a los Lineamientos de la Política Económica y Social del Partido y la Revolución, en lo referente al desarrollo de proyectos locales. En este contexto, resulta de vital importancia el manejo y acceso a fuentes de información confiables y pertinentes, como las que ofrece el Censo de Población y Viviendas, que se traduzca luego en una mayor consistencia a la hora de tomar decisiones y gestionar todos los procesos que intervienen en la ejecución de este Programa.

  17. Use of ion beams to simulate reaction of reactor fuels with their cladding

    International Nuclear Information System (INIS)

    Birtcher, R.C.; Baldo, P.

    2006-01-01

    Processes occurring within reactor cores are not amenable to direct experimental observation. Among major concerns are damage, fission gas accumulation and reaction between the fuel and its cladding all of which lead to swelling. These questions can be investigated through simulation with ion beams. As an example, we discuss the irradiation driven interaction of uranium-molybdenum alloys, intended for use as low-enrichment reactor fuels, with aluminum, which is used as fuel cladding. Uranium-molybdenum coated with a 100 nm thin film of aluminum was irradiated with 3 MeV Kr ions to simulate fission fragment damage. Mixing and diffusion of aluminum was followed as a function of irradiation with RBS and nuclear reaction analysis using the 27 Al(p,γ) 28 Si reaction which occurs at a proton energy of 991.9 keV. During irradiation at 150 deg. C, aluminum diffused into the uranium alloy at a irradiation driven diffusion rate of 30 nm 2 /dpa. At a dose of 90 dpa, uranium diffusion into the aluminum layer resulted in formation of an aluminide phase at the initial interface. The thickness of this phase grew until it consumed the aluminum layer. The rapid diffusion of Al into these reactor fuels may offer explanation of the observation that porosity is not observed in the fuel particles but on their periphery

  18. Effect of Air Pollutants on Rain Water Characteristics in Hammam Al-Aleel District/ Al-Erej Village South of Mosul City

    Directory of Open Access Journals (Sweden)

    Ayad F. Qasim

    2012-06-01

    Full Text Available This study focused on Hammam Al-Aleel district/ Al-Erej village which is located south of Mosul city. This area has been suffering from spreading the air-pollutants resulted from the chimneys of Hammam Al-Aleel cement factories. Different areas were chosen for collecting samples of the rain water from December 2008 to March 2009. The study showed that rain-water is generally basic. The tests show an increase in ( EC, Mg+2, Ca+2, Cl-1, SO4-2 and NO-1,  due to the air pollutants emitted from local industries. The concentration of heavy elements represented by  Pb and Cd were high and this increase is due to the source of these elements from burning of the crude oil which is used in running the rotating kiln to produce the clinker in cement factory, In addition to the fuel used for  running vehicle and the friction between tires and  road.

  19. Viability of inert matrix fuel in reducing plutonium amounts in reactors

    International Nuclear Information System (INIS)

    2006-08-01

    Reactors worldwide have produced more than 2000 tonnes of plutonium, contained in spent fuel or as separated forms through reprocessing. Disposition of fissile materials has become a primary concern of nuclear non-proliferation efforts. There is a significant interest in IAEA Member States to develop proliferation resistant nuclear fuel cycles for incineration of plutonium such as inert matrix fuels (IMFs). The present report summarises R and D work on inert matrix fuel for plutonium and (to a lesser extent) minor actinide stock-pile reduction, and discusses the possible strategies to include inert matrix fuel approaches to the nuclear fuel cycle. The publication reviews the status of potential IMF candidates and describes several identified candidate materials for both fast and thermal reactors: MgO, ZrO2, SiC, Zr alloy, SiAl, ZrN; some of these have undergone test irradiations and post-irradiation examination. Also discussed are modelling of IMF fuel performance and safety analysis. System studies have identified strategies for both implementation of IMF fuel as homogeneous or heterogeneous phases, as assemblies or core loadings and in existing reactors in the shorter term, as well as in new reactors in the longer term

  20. Elastic Modulus Measurement of ORNL ATF FeCrAl Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, Zachary T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    Elastic modulus and Poisson’s ratio for a number of wrought FeCrAl alloys, intended for accident tolerant fuel cladding application, are determined via resonant ultrasonic spectroscopy. The results are reported as a function of temperature from room temperature to 850°C. The wrought alloys were in the fully annealed and unirradiated state. The elastic modulus for the wrought FeCrAl alloys is at least twice that of Zr-based alloys over the temperature range of this study. The Poisson’s ratio of the alloys was 0.28 on average and increased very slightly with increasing temperature.

  1. 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.

  2. 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.

  3. K-Basin spent nuclear fuel characterization data report 2

    International Nuclear Information System (INIS)

    Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, S.C.; Pyecha, T.D.; Thornton, T.A.

    1996-03-01

    An Integrated Process Strategy has been developed to package, condition, transport, and store in an interim storage facility the spent nuclear fuel (SNF) currently residing in the K-Basins at Hanford. Information required to support the development of the condition process and to support the safety analyses must be obtained from characterization testing activities conducted on fuel samples from the Basins. Some of the information obtained in the testing was reported in PNL-10778, K-Basin Spent Nuclear Fuel Characterization Data Report (Abrefah et al. 1995). That report focused on the physical, dimensional, metallographic examinations of the first K-West (KW) Basin SNF element to be examined in the Postirradiation Testing Laboratory (PTL) hot cells; it also described some of the initial SNF conditioning tests. This second of the series of data reports covers the subsequent series of SNF tests on the first fuel element. These tests included optical microscopy analyses, conditioning (drying and oxidation) tests, ignition tests, and hydrogen content tests

  4. K-Basin spent nuclear fuel characterization data report 2

    Energy Technology Data Exchange (ETDEWEB)

    Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, S.C.; Pyecha, T.D.; Thornton, T.A.

    1996-03-01

    An Integrated Process Strategy has been developed to package, condition, transport, and store in an interim storage facility the spent nuclear fuel (SNF) currently residing in the K-Basins at Hanford. Information required to support the development of the condition process and to support the safety analyses must be obtained from characterization testing activities conducted on fuel samples from the Basins. Some of the information obtained in the testing was reported in PNL-10778, K-Basin Spent Nuclear Fuel Characterization Data Report (Abrefah et al. 1995). That report focused on the physical, dimensional, metallographic examinations of the first K-West (KW) Basin SNF element to be examined in the Postirradiation Testing Laboratory (PTL) hot cells; it also described some of the initial SNF conditioning tests. This second of the series of data reports covers the subsequent series of SNF tests on the first fuel element. These tests included optical microscopy analyses, conditioning (drying and oxidation) tests, ignition tests, and hydrogen content tests.

  5. Application of gamma spectrometry technique in combination with weighing for material balance taking in the production of highly enriched U-A1 fuel

    International Nuclear Information System (INIS)

    Serin, P.A.

    1975-07-01

    The purpose of this project is to obtain the data on material balance for a batch of highly enriched U-Al alloys (used in the NRX and NRU reactors) during production of fuel, using gamma spectrometry (mainly the 186 KeV photopeak) and weighing, and to determine operational data of the Agency's single channel stabilized spectrometer (SAM-1) for measurement of the product typical for the production of highly enriched U-Al fuel (U-Al billets, fuel elements, scrap). The data collected indicates that gamma spectrometry using the single channel stabilized spectrometer is a valid non-destructive method of determining quantitatively U-235 content of U-Al alloy in the form of cast billets or extruded fuel elements providing that adequate standards are available. An accuracy of better than + 1% relative can be obtained using a simple jig to provide reproducible counting geometry. Count rates should be kept well below the saturation level of the detector and counter, preferably by a lead collimator in front of the detector. This non-destructive method is not easily applicable to scrap because of the inability to maintain constant geometry and to prepare standards closely similar in size and shape to the samples

  6. Contribuciones al diseño e implementación de un sistema de alta resolución para tomografía por emisión

    OpenAIRE

    Guerra Gutiérrez, Pedro

    2011-01-01

    La tomografía por emisión combina biología, física y computación para dar lugar a imágenes del organismo que en última instancia se relacionan con el comportamiento celular, habiendo demostrado ser una técnica de gran interés en oncología, neurología y radiología, tanto en investigación como en rutina clínica. La técnica se fundamenta en la detección de la radiación gamma generada por un radiofármaco que previamente se ha inyectado al paciente, para lo cual el equipo de tomografía dispone un ...

  7. Characterizing and packaging BN-350 spent fuel for long-term dry storage

    International Nuclear Information System (INIS)

    Lambert, J. D. B.; Bolshinsky, I.; Haues, S.L.; Allen, K.J.; Howden, E.A.; Hill, R.N.; Planchon, H.P.; Staples, P.; Karaulov, V.N.; Blynskij, A.P.; Yakovlev, I.K.; Maev, V.; Dumchev, I. A.

    2000-01-01

    The Republic of Kazakhstan is being assisted by the U.S. Department of Energy in preparing spent fuel from the BN-350 fast reactor for long term dry storage. Argonne National Laboratory was assigned responsibility for the physical and nuclear characterization of the spent fuel, for the design and safety analysis of 6-pac and 4-pac canisters used to contain spent fuel assemblies for storage, and for the design, testing and installation of a closure station at the reactor in which the canisters of fuel are dried, filled with inert gas and welded shut. This paper briefly describes the specialized components and equipment used, the process followed, and experience gained in packaging the spent fuel. Olsen et al and Schaefer separately discuss overall safety and criticality considerations of the packaging process in parallel papers to this conference

  8. U-Mo fuels handbook. Version 1.0

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  9. Development of nuclear fuel materials for research reactor

    International Nuclear Information System (INIS)

    Kim, Chang Kyu; Park, H. D.; Kim, K. H.; Lee, J. T.; Ryu, W. S.; Hwang, W.; Kim, H. N.; Kim, H. I.; Kwon, H. I.; Park, C.; Lee, B. C.; Park, J. M.; Lee, C. S.; Chae, H. T.; Im, N. J.; Cho, M. S.; Im, I. C.; Nam, C.; Lee, D. B.; Goh, Y. M.; Kim, J. D.; Ahn, H. S.; Woo, Y. M.; Chang, S. J.; Cho, H. D.

    1997-09-01

    This project has aimed at the development of U 3 Si dispersion fuel for the localization of HANARO fuel and the application of atomization process to advanced RERTR fuel development. The design criteria were established through the modified computer codes. Design documents were prepared and issued. The acceptable co-extrusion cladding was achieved. The electron beam welding technology has been developed and the sealing of the end plug and cladding was accomplished without defects. The atomization fuel meats have about 200% higher elongation and about 20% higher than comminution fuel meats. The thermal compatibility test showed that atomization fuel have about 30% higher stability that the comminution fuel. The pressure drops of 18 rods fuel assembly and 36 rods fuel assembly were measured to have 213 kPa and 205 kPa respectively. Apparent wear was not found in endurance test. The irradiation fuel was designed and fabricated by using low enriched uranium metal following the developed Q/A system. The safety analysis of irradiation fuel assembly was performed through linear power calculation by using MCNP4A code and centerline temperature calculation by using DIFAIR code. The quality assurance system has been established. The quality inspection technologies were developed. By acquiring the license, low enriched uranium of 100 kg as well as depleted uranium can be used. U 3 Si 2 -Al fuel swelled less than comminution fuel irrespective of temperature and fuel fraction in a compatibility test. The atomized U-10wt.%Mo powder were found to have gamma phase of isotropic structure. Gamma structure remained with a little swelling without any structure change at 400 deg C for 100 hours. Irradiation miniplate and test rig were designed preliminary manufactured. Thermal hydraulic and linear power calculations were performed by using PLTEMP and MCNP4A computer codes respectively. The hydraulic test showed that the pressure drop met the HANARO requirement. The vibration

  10. Development of a treatment technology for diluting highly enriched AL-based DOE spent nuclear fuel: principles and practices

    International Nuclear Information System (INIS)

    Adams, T.M.; Duncan, A.J.; Peacock, H.B.; Fisher, D.L.

    2001-01-01

    The Savannah River Site is the U.S. Department of Energy's preferred site for return and treatment of all aluminum-base, spent, research and test reactor fuel assemblies. There are over 20,000 spent fuel assemblies now stored in different countries around the world, and by 2035 many will be returned to SRS for treatment and interim storage. Interim storage canisters at SRS will be sent to a Mined Geologic Repository for long-term geologic storage. Laboratory and plant layout of a full-scale process for irradiated fuel has been completed. Tests of the off gas absorption system have been initiated using both surrogate and irradiated RERTR mini fuel plates. An pilot-scale L-Area Experimental Facility (LEF) is planned to validate induction furnace operations, remote handling, and the off gas system for trapping volatile elements under plant operating conditions. (authors)

  11. Hydrogen production from bio-fuels using precious metal catalysts

    Science.gov (United States)

    Pasel, Joachim; Wohlrab, Sebastian; Rotov, Mikhail; Löhken, Katrin; Peters, Ralf; Stolten, Detlef

    2017-11-01

    Fuel cell systems with integrated autothermal reforming unit require active and robust catalysts for H2 production. Thus, an experimental screening of catalysts for autothermal reforming of commercial biodiesel fuel was performed. Catalysts consisted of a monolithic cordierite substrate, an oxide support (γ-Al2O3) and Pt, Ru, Ni, PtRh and PtRu as active phase. Experiments were run by widely varying the O2/C and H2O/C molar ratios at different gas hourly space velocities. Fresh and aged catalysts were characterized by temperature programmed methods and thermogravimetry to find correlations with catalytic activity and stability.

  12. Progress on LEU very high density fuel and target development in Argentina

    International Nuclear Information System (INIS)

    Balart, S.; Cabot, P.; Calzetta, O.; Duran, A.; Garces, J.; Hermida, J.D.; Manzini, A.; Pasqualini, E.; Taboada, H.

    2006-01-01

    Since last RRFM meeting, CNEA has continued on new LEU fuel and target development activities. Main goals are the plan to convert our RA-6 reactor from HEU to a new LEU core, to get a comprehensive understanding of U-Mo/Al compounds phase formation in dispersed and monolithic fuels, to develop possible solutions to VHD dispersed and monolithic fuels technical problems, to optimize techniques to recover U from silicide scrap samples as cold test for radiowaste separation for final conditioning of silicide spent fuels. and to improve the diffusion of LEU target and radiochemical technology for radioisotope production. Future plans include: - Completion of the RA-6 reactor conversion to LEU; - Improvement on fuel development and production facilities to implement new technologies, including NDT techniques to assess bonding quality; - Irradiation of miniplates and full scale fuel assembly at RA-3 and plans to perform irradiation on higher power and temperature regime reactors; - Optimization of LEU target and radiochemical techniques for radioisotope production. (author)

  13. Quantitative determination of uranium distribution homogeneity in MTR fuel type plates

    International Nuclear Information System (INIS)

    Ferrufino, Felipe Bonito Jaldin

    2011-01-01

    IPEN/CNEN-SP produces the fuel to supply its nuclear research reactor IEA-R1. The fuel is assembled with fuel plates containing an U 3 Si 2 -Al composite meat. A good homogeneity in the uranium distribution inside the fuel plate meat is important from the standpoint of irradiation performance. Considering the lower power of reactor IEA-R1, the uranium distribution in the fuel plate has been evaluated only by visual inspection of radiographs. However, with the possibility of IPEN to manufacture the fuel for the new Brazilian Multipurpose Reactor (RMB), with higher power, it urges to develop a methodology to determine quantitatively the uranium distribution into the fuel. This paper presents a methodology based on X-ray attenuation, in order to quantify the uranium concentration distribution in the meat of the fuel plate by using optical densities in radiographs and comparison with standards. The results demonstrated the inapplicability of the method, considering the current specification for the fuel plates due to the high intrinsic error to the method. However, the study of the errors involved in the methodology, seeking to increase their accuracy and precision, can enable the application of the method to qualify the final product. (author)

  14. Verification to the RSG-GAS fuel discharge burn-up using SRAC2006 module of COREBN/HIST

    International Nuclear Information System (INIS)

    J-Susilo; T-M-Sembiring; G-R-Sunaryo; M-Imron

    2018-01-01

    For 30 years operation, some of the modifications to the RSG GAS core has been done, that are changes included the type of fuel from U 3 O 8 -Al to U 3 Si 2 -Al with the same density 2.96 gU/cc, the loading pattern of standard fuel elements/fuel control elements from 6/1 & 6/2 to 5/1 pattern, and in core fuel management calculation tool has been change from IAFUEL to BATAN-FUEL. To obtain an extension of the operating license for the next 10 years, the RSG-GAS Periodic Safety Assessment Document is need to prepared. According to the Regulatory Body Chairman Regulation No. 2 2015, RSG-GAS safety assessment should be done independently. As part of this assessment the fuel discharge burn-up must be estimated. In this research, to ensure that the misposition of fuel element in the core has not occurred, the investigation to the document operating report related the fuel placement has been done. Therefore, by using 78 th to 93 rd operation data, verify of the fuel discharge burn-up of the RSG-GAS has been performed by using SRAC2006 module of COREBN/HIST. In addition, the results of these calculations are also made comparative with the operating report data that is calculated by using BATAN-FUEL. Maximum fuel discharge burn-up (57.73 % of U-235) was verified still under permissible value determined by the regulatory body (<60 % of U-235). Maximum differences value between two computer codes was about 2.12 % of U-235 (3.80 %) that is fuel at the B-7 position. Fuel discharge burn-up of RSG-GAS showed almost the same value for each the operation cycle, range of 1.52 % of U-235. So it can be concluded that the RSG-GAS core operation over the last ten years was in good fuel management performance, in accordance with the design. BATAN-FUEL has been conformed well enough with COREBN/HIST. (author)

  15. 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

  16. Annex 2 - Design concept of RA reactor fuel element fabrication; Prilog 2 - Koncepcija izrade gorivnog elementa reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovic, M; Mihajlovic, A; Ristic, M; Simic, G [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This document shows the construction and description of the individual parts of the fuel element. Planned procedures for fabrication of Al cladding and the fuel element structure are presented. Within the analysis of the results, claddings obtained by applying the proposed procedure are presented as well as the future plans. Dat je prikaz konstrukcije sa opisom i ulogom pojedinih sastavnih delova i sklopa gorivnog elementa. Izneta su predvidjanja za izradu, kako Al kosuljice, tako i sklopa gorivnog elementa. U analizi dosadasnjih rezultata, prikazane su kosuljice dobijene po navedenim postupcima, kao i smernice daljeg rada (author)

  17. Reduced Gravity Studies of Soret Transport Effects in Liquid Fuel Combustion

    Science.gov (United States)

    Shaw, Benjamin D.

    2004-01-01

    Soret transport, which is mass transport driven by thermal gradients, can be important in practical flames as well as laboratory flames by influencing transport of low molecular weight species (e.g., monatomic and diatomic hydrogen). In addition, gas-phase Soret transport of high molecular weight fuel species that are present in practical liquid fuels (e.g., octane or methanol) can be significant in practical flames (Rosner et al., 2000; Dakhlia et al., 2002) and in high pressure droplet evaporation (Curtis and Farrell, 1992), and it has also been shown that Soret transport effects can be important in determining oxygen diffusion rates in certain classes of microgravity droplet combustion experiments (Aharon and Shaw, 1998). It is thus useful to obtain information on flames under conditions where Soret effects can be clearly observed. This research is concerned with investigating effects of Soret transport on combustion of liquid fuels, in particular liquid fuel droplets. Reduced-gravity is employed to provide an ideal (spherically-symmetrical) experimental model with which to investigate effects of Soret transport on combustion. The research will involve performing reduced-gravity experiments on combustion of liquid fuel droplets in environments where Soret effects significantly influence transport of fuel and oxygen to flame zones. Experiments will also be performed where Soret effects are not expected to be important. Droplets initially in the 0.5 to 1 mm size range will be burned. Data will be obtained on influences of Soret transport on combustion characteristics (e.g., droplet burning rates, droplet lifetimes, gas-phase extinction, and transient flame behaviors) under simplified geometrical conditions that are most amenable to theoretical modeling (i.e., spherical symmetry). The experiments will be compared with existing theoretical models as well as new models that will be developed. Normal gravity experiments will also be performed.

  18. Numerical studies of spray combustion processes of palm oil biodiesel and diesel fuels using reduced chemical kinetic mechanisms

    KAUST Repository

    Kuti, Olawole; Sarathy, Mani; Nishida, Keiya; Roberts, William L.

    2014-01-01

    Spray combustion processes of palm oil biodiesel (PO) and conventional diesel fuels were simulated using the CONVERGE CFD code. Thermochemical and reaction kinetic data (115 species and 460 reactions) by Luo et al. (2012) and Lu et al. (2009) (68

  19. Electrostatic interactions for directed assembly of high performance nanostructured energetic materials of Al/Fe2O3/multi-walled carbon nanotube (MWCNT)

    International Nuclear Information System (INIS)

    Zhang, Tianfu; Ma, Zhuang; Li, Guoping; Wang, Zhen; Zhao, Benbo; Luo, Yunjun

    2016-01-01

    Electrostatic self-assembly in organic solvent without intensively oxidative or corrosive environments, was adopted to prepare Al/Fe 2 O 3 /MWCNT nanostructured energetic materials as an energy generating material. The negatively charged MWCNT was used as a glue-like agent to direct the self-assembly of the well dispersed positively charged Al (fuel) and Fe 2 O 3 (oxide) nanoparticles. This spontaneous assembly method without any surfactant chemistry or other chemical and biological moieties decreased the aggregation of the same nanoparticles largely, moreover, the poor interfacial contact between the Al (fuel) and Fe 2 O 3 (oxide) nanoparticles was improved significantly, which was the key characteristic of high performance nanostructured energetic materials. In addition, the assembly process was confirmed as Diffusion-Limited Aggregation. The assembled Al/Fe 2 O 3 /MWCNT nanostructured energetic materials showed excellent performance with heat release of 2400 J/g, peak pressure of 0.42 MPa and pressurization rate of 105.71 MPa/s, superior to that in the control group Al/Fe 2 O 3 nanostructured energetic materials prepared by sonication with heat release of 1326 J/g, peak pressure of 0.19 MPa and pressurization rate of 33.33 MPa/s. Therefore, the approach, which is facile, opens a promising route to the high performance nanostructured energetic materials. - Graphical abstract: The negatively charged MWCNT was used as a glue-like agent to direct the self-assembly of the well dispersed positively charged Al (fuel) and Fe 2 O 3 (oxide) nanoparticles. - Highlights: • A facile spontaneous electrostatic assembly strategy without surfactant was adopted. • The fuels and oxidizers assembled into densely packed nanostructured composites. • The assembled nanostructured energetic materials have excellent performance. • This high performance energetic material can be scaled up for practical application. • This strategy can be applied into other nanostructured

  20. Viability of thin wall tube forming of ATF FeCrAl

    Energy Technology Data Exchange (ETDEWEB)

    Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aydogan, Eda [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Anderoglu, Osman [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lavender, Curt [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-16

    Fabrication of thin walled tubing of FeCrAl alloys is critical to its success as a candidate enhanced accident-tolerant fuel cladding material. Alloys that are being investigated are Generation I and Generation II FeCrAl alloys produced at ORNL and an ODS FeCrAl alloy, MA-956 produced by Special Metals. Gen I and Gen II FeCrAl alloys were provided by ORNL and MA-956 was provided by LANL (initially produced by Special Metals). Three tube development efforts were undertaken. ORNL led the FeCrAl Gen I and Gen II alloy development and tube processing studies through drawing tubes at Rhenium Corporation. LANL received alloys from ORNL and led tube processing studies through drawing tubes at Century Tubing. PNNL led the development of tube processing studies on MA-956 through pilger processing working with Sandvik Corporation. A summary of the recent progress on tube development is provided in the following report and a separate ORNL report: ORNL/TM-2015/478, “Development and Quality Assessments of Commercial Heat Production of ATF FeCrAl Tubes”.