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Sample records for high-density dispersion fuels

  1. 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

  2. High density fuels using dispersion and monolithic fuel

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  3. High density fuels using dispersion and monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  4. 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.

  5. Postirradiation examination of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Strain, R.V.

    1998-01-01

    Two irradiation test vehicles, designated RERTR-2, were inserted into the Advanced Test reactor in Idaho in August 1997. These tests were designed to obtain irradiation performance information on a variety of potential new, high-density uranium alloy dispersion fuels, including U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru and U-10Mo-0.05Sn: the intermetallic compounds U 2 Mo and U-10Mo-0.-5Sn; the intermetallic compounds U 2 Mo and U 3 Si 2 were also included in the fuel test matrix. These fuels are included in the experiments as microplates (76 mm x 22 mm x 1.3mm outer dimensions) with a nominal fuel volume loading of 25% and irradiated at relatively low temperature (∼100 deg C). RERTR-1 and RERTR-2 were discharged from the reactor in November 1997 and July 1998, respectively at calculated peak fuel burnups of 45 and 71 at %-U 235 Both experiments are currently under examination at the Alpha Gamma Hot Cell Facility at Argonne National Laboratory in Chicago. This paper presents the postirradiation examination results available to date from these experiments. (author)

  6. Irradiation testing of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-01-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 'microplates'. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U10Mo-0.05Sn, U2Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of approximately 40 and 80 at.% U 235 . Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions. (author)

  7. Irradiation testing of high density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-10-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U 2 Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions

  8. 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)

  9. 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)

  10. Design of high density gamma-phase uranium alloys for LEU dispersion fuel applications

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.; Ray, Allison E.

    1998-01-01

    Uranium alloys are candidates for the fuel phase in aluminium matrix dispersion fuels requiring high uranium loading. Certain uranium alloys have been shown to have good irradiation performance at intermediate burnup. previous studies have shown that acceptable fission gas swelling behavior and fuel-aluminium interaction is possible only if the fuel alloy can be maintained in the high temperature body-centered-cubic γ-phase during fabrication and irradiation, at temperatures at which αU is the equilibrium phase. transition metals in Groups V through VIII are known to allow metastable retention of the gamma phase below the equilibrium isotherm. These metals have varying degrees of effectiveness in stabilizing the gamma phase. Certain alloys are metastable for very long times at the relatively low fuel temperatures seen in research operation. In this paper, the existing data on the gamma stability of binary and ternary uranium alloys is analysed. The mechanism and kinetics of decomposition of the gamma phase are assessed with the help of metal alloy theory. Alloys with the highest possible uranium content, good gamma-phase stability, and good neutronic performance are identified for further metallurgical studies and irradiation tests. Results from theory will be compared with experimentally generated data. (author)

  11. An investigation on fuel meats extruded with atomized U-10wt% Mo powder for uranium high-density dispersion fuel

    International Nuclear Information System (INIS)

    Kim, Chang-Kyu; Kim, Ki-Hwan; Park, Jong-Man; Lee, Don-Bae; Sohn, Dong-Seong

    1997-01-01

    The RERTR program has been making an effort to develop dispersion fuels with uranium densities of 8 to 9 g U/cm3 for research and test reactors. Using atomized U-10wt%Mo powder, fuel meats have been fabricated successfully up to 55 volume % of fuel powder. The uranium density of an extruded meat with a 55 volume % of fuel powder was obtained to be 7.7 g/cm3. A relatively high porosity of 7.3% was formed due to cracking of particles, presumably induced by the impingement among agglomerated particles. Tensile test results indicated that the strength of fuel meats with 55% volume fraction decreased some and a little of ductility was maintained. Examination on the fracture surface revealed that some U-10%Mo particles appeared to be broken by the tensile force in brittle rupture mode. The increase of broken particles in high fuel fraction is considered to be induced mainly by the impingement among agglomerated particles. Uranium loading density is assumed to be improved through the development of the better homogeneous dispersion technology. (author)

  12. Relative neutronic performance of proposed high-density dispersion fuels in water-moderated and D2O-reflected research reactors

    International Nuclear Information System (INIS)

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

    1996-01-01

    This paper provides an overview of the neutronic performance of an idealized research reactor using several high density LEU fuels that are being developed by the RERTR program. High-density LEU dispersion fuels are needed for new and existing high-performance research reactors and to extend the lifetime of fuel elements in other research reactors. This paper discusses the anticipated neutronic behavior of proposed advanced fuels containing dispersions of U 3 Si 2 , UN, U 2 Mo and several uranium alloys with Mo, or Zr and Nb. These advanced fuels are ranked based on the results of equilibrium depletion calculations for a simplified reactor model having a small H 2 O-cooled core and a D 2 O reflector. Plans have been developed to fabricate and irradiate several uranium alloy dispersion fuels in order to test their stability and compatibility with the matrix material and to establish practical loading limits

  13. Relative neutronic performance of proposed high-density dispersion fuels in water-moderated and D2O-reflected research reactors

    International Nuclear Information System (INIS)

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

    1996-01-01

    This paper provides an overview of the neutronic performance of an idealized research reactor using several high density Leu fuels that are being developed by the Rarita program. High-density Leu dispersion fuels are needed for new and existing high-performance research reactors and to extend the lifetime of fuel elements in other research reactors. This paper discusses the anticipated neutronic behavior of proposed advanced fuels containing dispersions of U 3 Si 2 , UN, U 2 Mo and several uranium alloys with Mo, or Zr and Nb. These advanced fuels are ranked based on the results of equilibrium depletion calculations for a simplified reactor model having a small H 2 O-cooled core and a D 2 O reflector. Plans have been developed to fabricate and irradiate several uranium alloy dispersion fuels in order to test their stability and compatibility with the matrix material and to establish practical loading limits. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-15

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

  15. High density fuel storage rack

    International Nuclear Information System (INIS)

    Zezza, L.J.

    1980-01-01

    High storage density for spent nuclear fuel assemblies in a pool achieved by positioning fuel storage cells of high thermal neutron absorption materials in an upright configuration in a rack. The rack holds the cells at required pitch. Each cell carries an internal fuel assembly support, and most cells are vertically movable in the rack so that they rest on the pool bottom. Pool water circulation through the cells and around the fuel assemblies is permitted by circulation openings at the top and bottom of the cells above and below the fuel assemblies

  16. High density fuel storage racks

    International Nuclear Information System (INIS)

    Groves, M.D.

    1978-01-01

    An apparatus is described for the safe and compact storage of nuclear fuel assemblies in an array of discrete open-ended neutron absorbing shields for which the theoretical minimum safe separation distance and cell pitch are known. Open-ended stainless steel end fittings are welded to each end of each shield and the end fittings are welded to each other in side-by-side relation, thereby reducing the cell pitch tolerance due to fabrication uncertainties. In addition, a multiplicity of ridges on the sides of each shield having a height equal to one half the theoretical minimum safe separation distance further reduce shield bowing tolerances. The net tolerance reduction permits a significant increase in the number of fuel assemblies that can be safely contained in a storage area of fixed size

  17. Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)], E-mail: mfarhan_73@yahoo.co.uk; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)

    2008-09-15

    The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease.

  18. Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2008-01-01

    The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease

  19. Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-08-15

    The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U-Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm{sup 3}, 7.74 gU/cm{sup 3} and 8.57 gU/cm{sup 3}. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm{sup 3} to 8.90 gU/cm{sup 3}. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.

  20. High density aseismic spent fuel storage racks

    International Nuclear Information System (INIS)

    Louvat, J.P.

    1985-05-01

    After the reasons of the development of high density aseismic spent fuel racks by FRAMATOME and LEMER, a description is presented, as also the codes, standards and regulations used to design this FRAMATOME storage rack. Tests have been carried out concerning criticality, irradiation of Cadminox, corrosion of the cell, and the seismic behaviour

  1. Reaction of unirradiated high-density fuel with aluminum

    International Nuclear Information System (INIS)

    Wiencek, T.C.; Meyer, M.K.; Prokofiev, I.G.; Keiser, D.D.

    1997-01-01

    Excellent dispersion fuel performance requires that fuel particles remain stable and do not react significantly with the surrounding aluminum matrix. A series of high-density fuels, which contain uranium densities >12 g/cm 3 , have been fabricated into plates. As part of standard processing, all of these fuels were subjected to a blister anneal of 1 h at 485 deg. C. Changes in plate thickness were measured and evaluated. From these results, suppositions about the probable irradiation properties of these fuels have been proposed. In addition, two fuels, U-10 wt% Mo and U 2 Mo, were subjected to various heat treatments and were found to be very stable in an aluminum matrix. On the basis of the experimental data, hypotheses of the irradiation behavior of these fuels are presented. (author)

  2. Reprocessing ability of high density fuels for research and test reactors

    International Nuclear Information System (INIS)

    Gay, A.; Belieres, M.

    1997-01-01

    The development of a new high density fuel is becoming a key issue for Research Reactors operators. Such a new fuel should be a Low Enrichment Uranium (LEU) fuel with a high density, to improve present in core performances. It must be compatible with the reprocessing in an industrial plant to provide a steady back-end solution. Within the framework of a work group CEA/CERCA/COGEMA on new fuel development for Research Reactors, COGEMA has performed an evaluation of the reprocessing ability of some fuel dispersants selected as good candidates. The results will allow US to classify these fuel dispersants from a reprocessing ability point of view. (author)

  3. High density UO2 powder preparation for HWR fuel

    International Nuclear Information System (INIS)

    Hwang, S. T.; Chang, I. S.; Choi, Y. D.; Cho, B. R.; Kwon, S. W.; Kim, B. H.; Moon, B. H.; Kim, S. D.; Phyu, K. M.; Lee, K. A.

    1992-01-01

    The objective of this project is to study on the preparation of method high density UO 2 powder for HWR Fuel. Accordingly, it is necessary to character ize the AUC processed UO 2 powder and to search method for the preparation of high density UO 2 powder for HWR Fuel. Therefore, it is expected that the results of this study can effect the producing of AUC processed UO 2 powder having sinterability. (Author)

  4. Behavior of high-density spent-fuel storage racks

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1986-08-01

    Included in this report is a summary of information on neutron-absorbing materials such as B 4 C in an aluminum matrix or organic binder material, stainless steel-boron and aluminum-boron alloys, and stainless steetl-clad cadmium that are used in high-density spent fuel storage racks. A list of the types of neutron-absorbing materials being used in spent fuel storage racks at domestic commercial plants is provided. Recent cases at several domestic plants where swelling of rack side plates (where the B 4 C in an aluminum matrix and B 4 C in an organic binder material were located) occurred are reviewed

  5. 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.)

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

    International Nuclear Information System (INIS)

    Taboada, Horacio; Gonzalez, Alfredo G.

    2005-01-01

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

  7. 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.)

  8. Ignition and burn in contaminated DT fuel at high densities

    International Nuclear Information System (INIS)

    Pasley, J.

    2010-01-01

    Complete text of publication follows. Radiation hydrodynamics simulations have been performed to quantify the effect of contamination upon the ignition threshold in DT at high densities. A detailed thermonuclear burn model, with multi-group multispecies ions, is incorporated alongside a multigroup diffusion approximation for thermal radiation transport. The code used is the research version of the HYADES 1D code. Acceptable levels of contamination are identified for a range of contaminant ion species. A range of different contaminant spatial distribution within the fuel are explored: i) in which the contamination is uniformly distributed throughout the fuel; ii) in which the impurity ions are confined to the hotspot, or iii) where contamination is restricted to a particular region of the hotspot (either centrally, near the surface, or at an intermediate location). Initially the fuel has a constant density with the hotspot located centrally. The overall radius of the fuel is chosen to be sufficiently large that it has no significant effect upon the success or failure of ignition. The evolution of the system is then simulated until ignition either establishes widespread thermonuclear burning, or a failure to ignite is observed. The critical ρr for ignition is found by iteration on the hotspot radius. We show that varying the spatial distribution of the contaminant within the ignition spot has little effect, so long as the total mass of contaminant is held the same. As expected, high-Z contamination is far more detrimental than that by low-Z ions. Discussion of the findings in the context of re-entrant cone-guided fast ignition is presented, in addition to a theoretical interpretation of the results.

  9. High regression rate, high density hybrid fuels, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This SBIR program will investigate high energy density novel nanofuels combined with high density binders for use with an N2O oxidizer. Terves has developed...

  10. Production of leu high density fuels at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Freim, J.B.

    1983-01-01

    A large number of fuel elements of all types are produced for both international and domestic customers by Nuclear Fuel Division of Babcock and Wilcox. A brief history of the division, included previous and present research reactor fuel element fabrication experience is discussed. The manufacturing facilities are briefly described. The fabrication of LEU fuels and economic analysis of the production are included. (A.J.)

  11. Neutronic performance of high-density LEU fuels in water-moderated and water-reflected research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Matos, J.E.

    1996-01-01

    At the Reduced Enrichment for Research and Test Reactors (RERTR) meeting in September 1994, Durand reported that the maximum uranium loading attainable with U 3 Si 2 fuel is about 6.0 g U/cm 3 . The French Commissariat a l'Energie Atomique (CEA) plan to perform irradiation tests with 5 plates at this loading. Compagnie pour L'Etude et La Realisation de Combustibles Atomiques (CERCA) has also fabricated a few uranium nitride (UN) plates with a uranium density in the fuel meat of 7.0 g/cm 3 and found that UN is compatible with the aluminum matrix at temperatures below 500 C. High density dispersion fuels proposed for development include U-Zr(4 wt%)-Nb(2 wt%), U-Mo(5 wt%), and U-Mo(9 wt%). The purpose of this note is to examine the relative neutronic behavior of these high density fuels in a typical light water-reflected and water-moderated MTR-type research reactor. The results show that a dispersion of the U-Zr-Nb alloy has the most favorable neutronic properties and offers the potential for uranium densities greater than 8.0 g/cm 3 . On the other hand, UN is the least reactive fuel because of the relatively large 14 N(n,p) cross section. For a fixed value of k eff , the required 235 U loading per fuel element is least for the U-Zr-Nb fuel and steadily increases for the U-Mo(5%), U-Mo(9%), and UN fuels. Because of volume fraction limitations, the UO 2 dispersions are only useful for uranium densities below 5.0 g/cm 3 . In this density range, however, UO 2 is more reactive than U 3 Si 2

  12. Qualification of high-density fuel manufacturing for research reactors at CNEA

    Energy Technology Data Exchange (ETDEWEB)

    Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; De La Fuente, M.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H. [CNEA, Buenos Aires (Argentina)

    2001-07-01

    CNEA, the National Atomic Energy Commission of Argentina, is at the present a qualified supplier of uranium oxide fuel for research reactors. A new objective in this field is to develop and qualify the manufacturing of LEU high-density fuel for this type of reactors. According with the international trend Silicide fuel and U-xMo fuel are included in our program as the most suitable options. The facilities to complete the qualification of high-density MTR fuels, like the manufacturing plant installations, the reactor, the pool side fuel examination station and the hot cells are fully operational and equipped to perform all the activities required within the program. The programs for both type of fuels include similar activities: development and set up of the fuel material manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of miniplates, fabrication and irradiation of full scale fuel elements, post-irradiation examination and feedback for manufacturing improvements. For silicide fuels most of these steps have already been completed. For U-xMo fuel the activities also include the development of alternative ways to obtain U-xMo powder, feasibility studies for large-scale manufacturing and the economical assessment. Set up of U-xMo fuel plate manufacturing is also well advanced and the fabrication of the first full scale prototype is foreseen during this year. (author)

  13. Seismic and structural analysis of high density/consolidated spent fuel storage racks

    International Nuclear Information System (INIS)

    Shah, S.J.; Biddle, J.R.; Bennett, S.M.; Schechter, C.B.; Harstead, G.A.; Kopecky, B.

    1995-01-01

    In many nuclear power plants, existing storage racks are being replaced with high-density racks to accommodate the increasing inventory of spent fuel. In the hypothetical design considered here, the high-density arrangement of fuel assemblies, or consolidated fuel canisters, is accomplished through the use of borated stainless steel (BSS) plates acting as neutron absorbers. The high-density fuel racks are simply supported by the pool floor with no structural connections to adjacent racks or to the pool walls or floor. Therefore, the racks are free standing and may slide and tip. Several time history, nonlinear, seismic analyses are required to account for variations in the coefficient of friction, rack loading configuration, ad the type of the seismic event. This paper presents several of the mathematical models usually used. The models include features to allow sliding and tipping of the racks and to represent the hydrodynamic coupling which can occur between fuel assemblies and rack cells, between adjacent racks, and between the racks and the reinforced concrete walls. A detailed model representing a single rack is used to evaluate the 3-D loading effects. This model is a controlling case for the stress analysis. A 2-D multi-rack model representing a row of racks between the spent fuel pool walls is used to evaluate the change in gaps between racks. The racks are analyzed for the fuel loading conditions of consolidated, full, empty, and half-loaded with fuel assemblies

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  15. Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H. [Unidad de Actividad Combustibles Nucleares Comision Nacional de Energia Atomica (CNE4), Avda. del Libertador, 8250 C1429BNO Buenos Aires (Argentina)

    2002-07-01

    One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm{sup 3}. PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

  16. Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

    International Nuclear Information System (INIS)

    Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H.

    2002-01-01

    One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm 3 . PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

  17. 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)

  18. Investigating the mechanical and barrier properties to oxygen and fuel of high density polyethylene–graphene nanoplatelet composites

    Energy Technology Data Exchange (ETDEWEB)

    Honaker, K., E-mail: honakers@egr.msu.edu; Vautard, F.; Drzal, L.T.

    2017-02-15

    Highlights: • Melt mixing used to investigate high density polyethylene and graphene nanoplatelet composite. • Addition of graphene nanoplatelets resulted in a stiffer polymer matrix. • Presence of graphene nanoplatelets causes a decrease in oxygen and fuel permeation. - Abstract: Graphene nanoplatelets (GnP) of different sizes were investigated for their ability to modify high density polyethylene (HDPE) for potential fuel system applications, focusing on compounding via melt mixing in a twin-screw extruder. Mechanical properties, crystallinity of the polymer, and permeation to oxygen and fuel were assessed as a function of GnP concentration. The surface of GnP acted as a nucleation site for the generation of HDPE crystallites, increasing the crystallinity. The flexural properties were improved, clearly influenced by platelet size and quality of dispersion. A sharp, 46% decrease of the impact resistance was observed, even at low GnP concentration (0.2 wt.%). With a 15 wt.% GnP-M-15 (platelets with a 15 μm diameter), a 73% reduction in oxygen permeation was observed and a 74% reduction in fuel vapor transmission. This correlation was similar throughout the GnP concentration range. The smaller diameter platelets had a lesser effect on the properties.

  19. High-density biosynthetic fuels: the intersection of heterogeneous catalysis and metabolic engineering.

    Science.gov (United States)

    Harvey, Benjamin G; Meylemans, Heather A; Gough, Raina V; Quintana, Roxanne L; Garrison, Michael D; Bruno, Thomas J

    2014-05-28

    Biosynthetic valencene, premnaspirodiene, and natural caryophyllene were hydrogenated and evaluated as high performance fuels. The parent sesquiterpenes were then isomerized to complex mixtures of hydrocarbons with the heterogeneous acid catalyst Nafion SAC-13. High density fuels with net heats of combustion ranging from 133-141 000 Btu gal(-1), or up to 13% higher than commercial jet fuel could be generated by this approach. The products of caryophyllene isomerization were primarily tricyclic hydrocarbons which after hydrogenation increased the fuel density by 6%. The isomerization of valencene and premnaspirodiene also generated a variety of sesquiterpenes, but in both cases the dominant product was δ-selinene. Ab initio calculations were conducted to determine the total electronic energies for the reactants and products. In all cases the results were in excellent agreement with the experimental distribution of isomers. The cetane numbers for the sesquiterpane fuels ranged from 20-32 and were highly dependent on the isomer distribution. Specific distillation cuts may have the potential to act as high density diesel fuels, while use of these hydrocarbons as additives to jet fuel will increase the range and/or time of flight of aircraft. In addition to the ability to generate high performance renewable fuels, the powerful combination of metabolic engineering and heterogeneous catalysis will allow for the preparation of a variety of sesquiterpenes with potential for pharmaceutical, flavor, and fragrance applications.

  20. Accident Analysis of High Density Storage Rack for Fresh Fuel Assemblies

    International Nuclear Information System (INIS)

    Jang, K. J.; Lee, M. J.; Jin, H. U.; Park, J. H.; Shin, S. Y.

    2009-01-01

    Recently KONES and KNF have developed the so called suspension-type High Density Storage Rack (HDSR) for fresh fuel assemblies. The USNRC OT position paper specifies that the design of the rack must ensure the functional integrity of the fuel racks under all credible fuel assembly drop events. In this context the functional integrity means the criticality safety. That is to say, the drop events must not bring any danger to the criticality safety of HDSR. This paper shows the results of the analysis carried out to demonstrate the regulatory compliance of the proposed racks under postulated accidental drop events

  1. Conversion of Mixed Plastic Wastes (High Density Polyethylene and Polypropylene) into Liquid Fuel

    International Nuclear Information System (INIS)

    Chaw Su Su Hmwe; Tint Tint Kywe; Moe Moe Kyaw

    2010-12-01

    In this study, mixed plastic wastes were converted into liquid fuels. Mixed plastic wastes used were high density polyethylene (HDPE) and polypropylene (PP). The pyrolysis of mixed plastic waste to liquid fuel was carried out with and without prepared zeolite catalyst.The catalyst was characterized by X-ray Diffraction (XRD). This catalyst was pre-treated for activation. The experiments were carried out at temperature range of 350-410C.Physical properties (density, kinematic, viscosity,refractive index)of prepared liquid fuel samples were measured. From this study, yields of liquid fuel and gas fuel were found to be 41-64% and 15-35% respectively. As for by products, char was obtained as the yield percentages from 9 to 14% and wax (yield% - 1 to 14) was formed during pyrolysis.

  2. High-Density Renewable Fuels Based on the Selective Dimerization of Pinenes

    Science.gov (United States)

    2009-01-01

    RJ-5, significant ring strain contributing to a high heat of combustion (Table 1). Bulk agricultural waste products, such as cellulose and lignin , are...compact structures and reactive olefin functionalities, these molecules have sig- nificant potential as feedstocks for high-density renewable fuels.2b,7...potential to have heating values exceeding that of JP-10. Given the favorable potential net heat of combustion for these pinene dimers, reactivity

  3. Synthesis of high density aviation fuel with cyclopentanol derived from lignocellulose

    Science.gov (United States)

    Sheng, Xueru; Li, Ning; Li, Guangyi; Wang, Wentao; Yang, Jinfan; Cong, Yu; Wang, Aiqin; Wang, Xiaodong; Zhang, Tao

    2015-03-01

    For the first time, renewable high density aviation fuels were synthesized at high overall yield (95.6%) by the Guerbet reaction of cyclopentanol which can be derived from lignocellulose, followed by the hydrodeoxygenation (HDO). The solvent-free Guerbet reaction of cyclopentanol was carried out under the co-catalysis of solid bases and Raney metals. Among the investigated catalyst systems, the combinations of magnesium-aluminium hydrotalcite (MgAl-HT) and Raney Ni (or Raney Co) exhibited the best performances. Over them, high carbon yield (96.7%) of C10 and C15 oxygenates was achieved. The Guerbet reaction products were further hydrodeoxygenated to bi(cyclopentane) and tri(cyclopentane) over a series of Ni catalysts. These alkanes have high densities (0.86 g mL-1 and 0.91 g mL-1) and can be used as high density aviation fuels or additives to bio-jet fuel. Among the investigated HDO catalysts, the 35 wt.% Ni-SiO2-DP prepared by deposition-precipitation method exhibited the highest activity.

  4. Analysis and Environmental Fate of Air Force Distillate and High Density Fuels

    Science.gov (United States)

    1981-10-01

    728.1 128 0.8 Toluenc 751.3 92 0.6 XTHDCPD 1049.6 136 66.8 NTHDCPD 1079.2 136 1.5HNN 1509.6 186 20.1 JP-1O XTHDCPD 1050.3 136 96.8 ITHDCPD 1079.6 136 1,5...deionized water and the salts listed below. Blanks of both waters were routinely extracted and analyzed for possible 4.nterferences. MNN PXTX XTHDCPD ...through 13; complete data summaries for the distillate fuels may be found in Appendix C. All com- ponents of the high density fuels except XTHDCPD of

  5. Progress in the development of very high density research and test reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    Wachs, D.M. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, Idaho 83415 (United States)

    2009-06-15

    New nuclear fuels are being developed to enable many of the most important research and test reactors worldwide to convert from high enriched uranium (HEU) fuels to low enriched uranium (LEU) fuels without significant loss in performance. The last decade of work has focused on the development of uranium-molybdenum alloy (U-Mo) based fuels and is an international effort that includes the active participation of more than ten national programs. The US RERTR program, under the NNSA's Global Threat Reduction Initiative (GTRI), is in the process of developing both dispersion and monolithic U-Mo fuel designs. While the U-Mo fuel alloy has behaved extremely well under irradiation, initial testing (circa 2003) revealed that the U-Mo fuels dispersed in aluminum had an unexpected tendency toward unstable swelling (pillowing) under high-power conditions. Technical investigations were initiated worldwide at this time by the partner programs to understand this behavior as well as to develop and test remedies. The behavior was corrected by modifying the chemistry of the U-Mo/Al interfaces in both fuel designs. In the dispersion fuel design, this was accomplished by the addition of small amounts of silicon to the aluminum matrix material. Two methods are under development for the monolithic fuel design, which include the application of a thin layer of silicon or a thin zirconium based diffusion barrier at the fuel/clad interface. This paper gives an overview of the current status of U-Mo fuel development, including basic research results, manufacturing aspects, results of the latest irradiations and post irradiation examinations, the approach to fuel performance qualification, and the scale-up and commercialization of fabrication technology. (authors)

  6. Qualification of high density aluminide fuels for the BR2 reactor

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, Andre; Gubel, Pol; Ponsard, Bernard; Pin, Thomas; Falgoux, Jean Louis

    2005-01-01

    The BR2 operation still relies on the use of 90..93% enriched HEU aluminide fuel. The availability of a limited batch of 73% enriched HEU from reprocessed BR2 uranium in Dounreay justified 10 years ago the qualification and use of this material. After some preliminary test irradiations, various batches of fuel elements were fabricated by the UKAEA-Dounreay and successfully irradiated. Due to their lower 235 U content (0.050 g 235 U/cm 2 ), these elements were always irradiated together with standard 90...93% HEU fuel elements. A mixed-core strategy was developed at this occasion for an optimal utilization, and was reported during the 4th RRFM conference (March 19-21, 2000, Colmar, France). The availability of a new batch of fresh 73% HEU material was the occasion, a few years ago, to initiate the development, fabrication and qualification of a new high density fuel element. An order was placed with CERCA to assess the optimal fabrication methods and tooling required to meet as far as possible the existing BR2 standard specifications and 235 U content (0.060 g 235 U/cm 2 ). This development phase has been already reported during the 7th RRFM conference (March 9-12, 2003, Aix-en-Provence, France). Afterwards, six lead test fuel elements were ordered for qualification by irradiation. The neutronic properties of the fuel elements were adjusted and optimized. After a short summary of the main results of the development program, this paper describes the nuclear characteristics of the high density fuel elements and comments on the nuclear follow-up of the lead test fuel elements during their irradiation for five cycles in the BR2 reactor and the return of experience for CERCA. (author)

  7. Fueling with edge recycling to high-density in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Leonard, A.W., E-mail: leonard@fusion.gat.com [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Elder, J.D. [University of Toronto Institute of Aerospace Studies, Toronto, Canada M3H 5T6 (Canada); Canik, J.M. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Groebner, R.J.; Osborne, T.H. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States)

    2013-07-15

    Pedestal fueling through edge recycling is examined with the interpretive OEDGE code for high-density discharges in DIII-D. A high current, high-density discharge is found to have a similar radial ion flux profile through the pedestal to a lower current, lower density discharge. The higher density discharge, however, has a greater density gradient indicating a pedestal particle diffusion coefficient that scales near linear with 1/I{sub p}. The time dependence of density profile is taken into account in the analysis of a discharge with low frequency ELMs. The time-dependent analysis indicates that the inferred neutral ionization source is inadequate to account for the increase in the density profile between ELMs, implying an inward density convection, or density pinch, near the top of the pedestal.

  8. New high density MTR fuel. The CEA-CERCA-COGEMA development program

    International Nuclear Information System (INIS)

    Languille, A.; Durand, J.P.; Gay, A.

    1999-01-01

    The development of a new generation of LEU, high in density and with reprocessing capacities MTR fuel, is a key issue to provide reactor operators with a smooth operation which is necessary for a long term development of Nuclear Energy. In the RRFM'98 meeting, a joint contribution of CEA, CERCA and COGEMA presented a technical classification of the potential candidates uranium alloys. In this paper this MTR working group presents the development program of a new high density fuel. This program is composed of three main steps: Basic Data analysis and collection, Plate Tests (Irradiation and Post Irradiation Examinations) and Lead Test Assemblies (Irradiation and Post Irradiation Examinations). The goal to be reached is to make this new fuel available before the end of the present US return policy. (author)

  9. Evaluation of AMPX-KENO benchmark calculations for high-density spent fuel storage racks

    International Nuclear Information System (INIS)

    Turner, S.E.; Gurley, M.K.

    1981-01-01

    The AMPX-KENO computer code package is commonly used to evaluate criticality in high-density spent fuel storage rack designs. Consequently, it is important to know the reliability that can be placed on such calculations and whether or not the results are conservative. This paper evaluates a series of AMPX-KENO calculations which have been made on selected critical experiments. The results are compared with similar analyses reported in the literature by the Oak Ridge National Laboratory and BandW. 8 refs

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  11. Technical ability of new MTR high-density fuel alloys regarding the whole fuel cycle

    International Nuclear Information System (INIS)

    Durand, J.P.; Maugard, B.; Gay, A.

    1998-01-01

    The development of new fuel alloys could provide a good opportunity to improve drastically the fuel cycle on the neutronic performances and the reprocessing point of view. Nevertheless, those parameters can only be considered if the fuel manufacture feasibility has been previously demonstrated. As a matter of fact, a MTR work group involving French partners (CEA, CERCA, COGEMA) has been set up in order to evaluate the technical ability of new fuels considering the whole fuel cycle. In this paper CERCA is presenting the preliminary results of UMo and UNbZr fuel plate manufacture, CEA is comparing to U 3 Si 2 the neutronic performances of fuels such as UMo, UN, UNbZr, while COGEMA is dealing with the reprocessing feasibility. (author)

  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. Fueling requirements of super-high-density plasmas towards innovative ignition regime

    International Nuclear Information System (INIS)

    Sakamoto, Ryuichi; Yamada, Hiroshi

    2014-01-01

    Highlights: • Self-burning scenario with internal diffusion barrier is investigated. • Peaked density profiles allow to sustain self-burning plasma at lower temperature. • Core fueling beyond internal diffusion barrier is essential to sustain peaked density. • Acceptable pellet size becomes small to prevent fusion out perturbation. • Very high velocity pellet injection beyond 10 km/s is inevitable for this scenario. - Abstract: Super-high-density plasma with an internal diffusion barrier which is observed in the Large Helical Device has been extrapolated to a fusion reactor grade plasma to explore an innovative ignition regime and to clarify essential requirements for pellet fueling. The peaked density profiles due to the internal diffusion barrier formation allow reduction in the required minimum temperature to sustain a self-burning plasma down to 10 keV. Direct core fueling beyond the internal diffusion barrier is essential to sustain the peaked density profile. Furthermore, the acceptable pellet size becomes small in terms of fusion output perturbation because the effective volume of the burning plasma becomes small with the peaked density profile. In order to sustain a self-burning plasma with an internal diffusion barrier, therefore, extremely high velocity pellet injection beyond 10 km/s is inevitable unless another solution to the core fueling is found

  14. Study of diffusion bonding in 6061 aluminum and development of future high-density fuels fabrication

    International Nuclear Information System (INIS)

    Prokofiev, I.G.; Wiencek, T.C.; McGann, D.J.

    1997-01-01

    Powder metallurgy dispersions of uranium alloys and silicides in an aluminum matrix have been developed by the RERTR program as a new generation of proliferation-resistant fuels. Testing uses fuel miniplates to simulate standard fuel with cladding and matrix in plate-type configurations. In order to seal the dispersion fuel plates, a diffusion bond must be established between the aluminum cover plates that surround the fuel meat. Four different variations of the standard method for roll-bonding 6061 aluminum were studied: mechanical cleaning, addition of a getter material, modifications to the standard chemical etching, and modifications to welding. Aluminum test pieces were subjected to a bend test after each rolling pass. Results, based on 400 samples, indicate that a reduction in thickness of at least 70% is required to produce a diffusion bond with the standard roll-bonding method, versus a 60% reduction when using a method in which the assembly was 100% welded and contained empty 9 mm holes near the frame corners. (author)

  15. Progress in developing very-high-density low-enriched-uranium fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Hayes, S.L.; Wiencek, T.C.; Strain, R.V.

    1999-01-01

    Preliminary results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-Mo alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm 3 . Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 deg. C) are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 deg. C in 8-g U/cm 3 fuel. (author)

  16. Dispersive effects in radiation transport and radiation hydrodynamics in matter at high density

    International Nuclear Information System (INIS)

    Crowley, B.J.B.

    1983-01-01

    In a recent research program (reported in AWRE 0 20/82) I have investigated the generalisation of the equations of radiation hydrodynamics when electromagnetic radiation is assumed to obey a linear-response dispersion relation of the form nω=kc where the refractive index n depends on the frequency ω and/or wave number k. From the application of the Boltzmann-Liouville transport theory to photons in the short-wavelength (geometrical optics) limit, I derive the energy and momentum equations which, when combined with a classical (Euler-Lagrange-Navier-Stokes) treatment of a fluid material medium in LTE, yield a complete dynamical theory of linear interactions (+ stimulated processes) between incoherent (thermal) radiation and dense, locally isotropic matter. The theory includes an account of pondero-motive forces and electro (magneto) striction. Moreover, it is apparently capable of being generalised to non-linear interactions in which the refractive index depends on the local specific intensity of the radiation field, and, to some extent, to the treatment of high-frequency coherent radiation. The generalisation of various approximated forms of radiation-transport theory (esp. diffusion) has been considered in detail. Some problems remain however. One such is the treatment of anomalous dispersion. Current research work is concentrating on the interesting atomic physics aspects of electromagnetic (esp. radiative) properties of a dispersive material medium

  17. Development of very-high-density low-enriched uranium fuels

    International Nuclear Information System (INIS)

    Snegrove, J.L.; Hofmann, G.L.; Trybus, C.L.; Wiencek, T.C.

    1997-01-01

    The RERTR (=Reduced Enrichment for Research and Test Reactors) program has begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun. (author)

  18. Development of very-high-density fuels by the RERTR program

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L.; Wiencek, T.C.

    1996-01-01

    The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

  19. Development of very high-density low-enriched uranium fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L.; Wiencek, T.C.

    1997-02-01

    The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and first results should be available by end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

  20. High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

    Science.gov (United States)

    Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

    2014-08-01

    In conventional nuclear fuel rods for light-water reactors, a helium-filled as-fabricated gap between the fuel and the cladding inner surface accommodates fuel swelling and cladding creep down. Because helium exhibits a very low thermal conductivity, it results in a large temperature rise in the gap. Liquid metal (LM; 1/3 weight portion each of lead, tin, and bismuth) has been proposed to be a gap filler because of its high thermal conductivity (∼100 times that of He), low melting point (∼100 °C), and lack of chemical reactivity with UO2 and water. With the presence of LM, the temperature drop across the gap is virtually eliminated and the fuel is operated at a lower temperature at the same power output, resulting in safer fuel, delayed fission gas release and prevention of massive secondary hydriding. During normal reactor operation, should an LM-bonded fuel rod failure occurs resulting in a discharge of liquid metal into the bottom of the reactor pressure vessel, it should not corrode stainless steel. An experiment was conducted to confirm that at 315 °C, LM in contact with 304 stainless steel in the PWR water chemistry environment for up to 30 days resulted in no observable corrosion. Moreover, during a hypothetical core-melt accident assuming that the liquid metal with elevated temperature between 1000 and 1600 °C is spread on a high-density concrete basement of the power plant, a small-scale experiment was performed to demonstrate that the LM-concrete interaction at 1000 °C for as long as 12 h resulted in no penetration. At 1200 °C for 5 h, the LM penetrated a distance of ∼1.3 cm, but the penetration appeared to stop. At 1400 °C the penetration rate was ∼0.7 cm/h. At 1600 °C, the penetration rate was ∼17 cm/h. No corrosion based on chemical reactions with high-density concrete occurred, and, hence, the only physical interaction between high-temperature LM and high-density concrete was from tiny cracks generated from thermal stress. Moreover

  1. Fabrication, characterization and gamma rays shielding properties of nano and micro lead oxide-dispersed-high density polyethylene composites

    Science.gov (United States)

    Mahmoud, Mohamed E.; El-Khatib, Ahmed M.; Badawi, Mohamed S.; Rashad, Amal R.; El-Sharkawy, Rehab M.; Thabet, Abouzeid A.

    2018-04-01

    Polymer composites of high-density polyethylene (HD-PE) filled with powdered lead oxide nanoparticles (PbO NPs) and bulk lead oxide (PbO Blk) were prepared with filler weight fraction [10% and 50%]. These polymer composites were investigated for radiation-shielding of gamma-rays emitted from radioactive point sources [241Am, 133Ba, 137Cs, and 60Co]. The polymer was found to decrease the heaviness of the shielding material and increase the flexibility while the metal oxide fillers acted as principle radiation attenuators in the polymer composite. The prepared composites were characterized by Fourier transform infrared spectrophotometer (FT-IR), X-ray diffraction (XRD), thermogravimetric analysis (TGA), scanning electron microscope (SEM), Brunauer-Emmett-Teller surface area (BET) and field emission transmission electron microscope (FE-TEM). The morphological analysis of the assembled composites showed that, PbO NPs and PbO Blk materials exhibited homogenous dispersion in the polymer-matrix. Thermogravimetric analysis (TGA) demonstrated that the thermal-stability of HD-PE was enhanced in the presence of both PbO Blk and PbO NPs. The results declared that, the density of polymer composites was increase with the percentage of filler contents. The highest density value was identified as 1.652 g cm-3 for 50 wt% of PbO NPs. Linear attenuation coefficients (μ) have been estimated from the use of XCOM code and measured results. Reasonable agreement was attended between theoretical and experimental results. These composites were also found to display excellent percentage of heaviness with respect to other conventional materials.

  2. Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures

    International Nuclear Information System (INIS)

    Oliveira, Fabio Branco Vaz de

    2008-01-01

    This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature, time and

  3. 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)

  4. 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)

  5. Liquid fuel obtain from polypropylene (PP-5) and high density polyethylene (HDPE-2) waste plastics mixture

    Energy Technology Data Exchange (ETDEWEB)

    Sarker, Moinuddin; Rashid, Mohammad Mamunor; Rahman, Md. Sadikur; Molla, Mohammed [Department of Research and Development, Natural State Research Inc, Stamford, (United States)

    2011-07-01

    Plastics are made by combination of small based molecules to form monomers. The monomers are then joined together by chemical polymerization mechanism to form polymers also known as plastics. These plastics contain various elements such as carbon, hydrogen, oxygen, nitrogen, chlorine and sul fur. The use of plastics is vastly expanded and it is being used in every sector of the world. However, using plastics does have a negative aspect, after use they end up in our landfill as waste causing numerous health and environmental problems. Landfill waste plastics release harmful gases due to the presence of carbon, chlorine and sul fur in them into the atmosphere causing climates to change drastically, equivalent to the effects of greenhouse gases (GHG) emission. To overcome these environmental issues, scientists have already developed many methods to converting these waste plastics into energy and fuel . We developed one new methods thermal cracking conversion to convert these waste plastics into usable liquid fuel . Thermal cracking conversion is a process to shorten the long chain hydrocarbons to produce liquid fuel in the absence of a catalyst. The thermal degradation process of the waste plastics long chain hydrocarbon to makes short chain hydrocarbon fuel. The fuel produced has been analyzed and tested according to standard methods. Key words: fuel , hydrocarbon, waste plastic, thermal degradation, conversion, GC/MS.

  6. Liquid fuel obtain from polypropylene (PP-5) and high density polyethylene (HDPE-2) waste plastics mixture

    International Nuclear Information System (INIS)

    Sarker, Moinuddin; Rashid, Mohammad Mamunor; Rahman, Md. Sadikur; Molla, Mohammed

    2011-01-01

    Plastics are made by combination of small based molecules to form monomers. The monomers are then joined together by chemical polymerization mechanism to form polymers also known as plastics. These plastics contain various elements such as carbon, hydrogen, oxygen, nitrogen, chlorine and sul fur. The use of plastics is vastly expanded and it is being used in every sector of the world. However, using plastics does have a negative aspect, after use they end up in our landfill as waste causing numerous health and environmental problems. Landfill waste plastics release harmful gases due to the presence of carbon, chlorine and sul fur in them into the atmosphere causing climates to change drastically, equivalent to the effects of greenhouse gases (GHG) emission. To overcome these environmental issues, scientists have already developed many methods to converting these waste plastics into energy and fuel . We developed one new methods thermal cracking conversion to convert these waste plastics into usable liquid fuel . Thermal cracking conversion is a process to shorten the long chain hydrocarbons to produce liquid fuel in the absence of a catalyst. The thermal degradation process of the waste plastics long chain hydrocarbon to makes short chain hydrocarbon fuel. The fuel produced has been analyzed and tested according to standard methods. Key words: fuel , hydrocarbon, waste plastic, thermal degradation, conversion, GC/MS

  7. 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

  8. Wax co-cracking synergism of high density polyethylene to alternative fuels

    Directory of Open Access Journals (Sweden)

    Magdy Motawie

    2015-09-01

    Full Text Available Attempts have been made to understand the thermal degradation of high density polyethylene (HDPE and their combined co-cracking using different ratios of HDPE and petroleum wax under nitrogen atmosphere. We have conducted the experiments using HDPE as the raw material and petroleum wax as co-feed by at 400 and 450 °C reaction temperatures. The product distribution was noted along with reaction time of 0.5–3 h for the degradation. Thermal gravimetric analysis (TGA technique was used to measure the weight change of the feedstock as a function of temperature and time. Differential scanning calorimetry (DSC was used to determine the degradation temperature. Products were characterized using gas chromatography (GC and infrared spectroscopy (FTIR, some other standard physical methods were used to determine the main properties of the liquid products. Results show that the mixed plastic-wax samples could be converted into gases, gasoline, and middle distillate depending upon the composition of feed polymer/wax ratio. It was found that the products mostly consisted of paraffin and olefin compounds, with carbon numbers of C1–C4, C5–C9 and C10–C19 in the case of gases, gasoline and middle distillate respectively.

  9. Improved method to demonstrate the structural integrity of high density fuel storage racks

    International Nuclear Information System (INIS)

    Hinderks, M.; Ungoreit, H.; Kremer, G.

    2001-01-01

    Reracking of existing fuel pools to the maximum extent is desirable from an economical point of view. This goal can be achieved by minimizing the gaps between the spent fuel storage racks. Since the rack design is aimed at enabling consolidated fuel rod storage, additional requirements arise with respect to the design and the structural analysis. The loads resulting from seismic events are decisive for the structural analysis and require a specially detailed and in-depth analysis for high seismic loads. The verification of structural integrity and functionality is performed in two phases. In the first phase the motional behavior of single racks, rows of racks and, where required, of all racks in the pool is simulated by excitation with displacement time histories under consideration of the fluid-structure interaction (FSI). The displacements from these simulations are evaluated, while the loads are utilized as input data for the structural analysis of the racks and the pool floor. The structural analyses for the racks comprise substantially stress analyses for base material and welds as well as stability analyses for the support channels and the rack outside walls. The analyses are performed in accordance with the specified codes and standards

  10. Study of diffusion bond development in 6061 aluminum and its relationship to future high density fuels fabrication.

    Energy Technology Data Exchange (ETDEWEB)

    Prokofiev, I.; Wiencek, T.; McGann, D.

    1997-10-07

    Powder metallurgy dispersions of uranium alloys and silicides in an aluminum matrix have been developed by the RERTR program as a new generation of proliferation-resistant fuels. Testing is done with miniplate-type fuel plates to simulate standard fuel with cladding and matrix in plate-type configurations. In order to seal the dispersion fuel plates, a diffusion bond must exist between the aluminum coverplates surrounding the fuel meat. Four different variations in the standard method for roll-bonding 6061 aluminum were studied. They included mechanical cleaning, addition of a getter material, modifications to the standard chemical etching, and welding methods. Aluminum test pieces were subjected to a bend test after each rolling pass. Results, based on 400 samples, indicate that at least a 70% reduction in thickness is required to produce a diffusion bond using the standard rollbonding method versus a 60% reduction using the Type II method in which the assembly was welded 100% and contained open 9mm holes at frame corners.

  11. 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)

  12. Progress in irradiation performance of experimental uranium - Molybdenum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.

    2002-01-01

    High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL alpha-gamma hot cells. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm -3 atomized and machined uranium alloy powders containing 7 wt% and 10 wt% molybdenum. In addition, two miniplates containing solid U-10 wt% Mo foils and three containing 6 g cm -3 U 3 Si 2 are part of the test. The results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from previous tests performed to lower burnup will be presented. (author)

  13. High-density and high-ρR fuel assembly for fast-ignition inertial confinement fusion

    International Nuclear Information System (INIS)

    Betti, R.; Zhou, C.

    2005-01-01

    Scaling relations to optimize implosion parameters for fast-ignition inertial confinement fusion are derived and used to design high-gain fast-ignition targets. A method to assemble thermonuclear fuel at high densities, high ρR, and with a small-size hot spot is presented. Massive cryogenic shells can be imploded with a low implosion velocity V I on a low adiabat α using the relaxation-pulse technique. While the low V I yields a small hot spot, the low α leads to large peak values of the density and areal density. It is shown that a 750 kJ laser can assemble fuel with V I ≅1.7x10 7 cm/s, α≅0.7, ρ≅400 g/cc, ρR≅3 g/cm 2 , and a hot-spot volume of less than 10% of the compressed core. If fully ignited, this fuel assembly can produce high gains of interest to inertial fusion energy applications

  14. Catalytic degradation of waste high-density polyethylene into fuel products using BaCO{sub 3} as a catalyst

    Energy Technology Data Exchange (ETDEWEB)

    Jan, M. Rasul; Shah, Jasmin; Gulab, Hussain [Institute of Chemical Sciences, University of Peshawar, N.W.F.P. (Pakistan)

    2010-11-15

    Waste high-density polyethylene (HDPE) was degraded thermally and catalytically using BaCO{sub 3} as a catalyst under different conditions of temperature, cat/pol ratio and time. The oil collected at optimum conditions (450 C, 0.1 cat/pol ratio and 2 h reaction time) was fractionated at different temperatures and fuel property of the fractions and parent oil was evaluated by their physicochemical parameters for fuel tests. The results were compared with the standard values for gasoline, kerosene and diesel oil. Boiling point distribution (BPD) curves were plotted from the gas chromatographic study of the samples and compared with that of the standard gasoline, kerosene and diesel. The oil samples were analyzed using GC/MS in order to find out their composition. The physical parameters and the composition of the parent oil and its fractions support the resemblance of the samples with the standard fuel oils. The light fractions best match with gasoline, the middle fractions match with kerosene and the heavier fractions match with diesel oil in almost all of the characteristic properties. (author)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  16. High density polyethylene (HDPE-2) and polystyrene (PS-6) waste plastic mixture turns into valuable fuel energy

    Energy Technology Data Exchange (ETDEWEB)

    Sarker, Moinuddin; Rashid, Mohammad Mamunor; Rahman, Md. Sadikur; Molla, Mohammed [Department of Research and Development, Natural State Research Inc, Stamford, (United States)

    2011-07-01

    Disposal of waste plastic is a serious concern in USA. Waste plastic generated from different cities and towns is a part of municipal solid waste. It is a matter of concern that disposal of waste plastic is causing many problems such as leaching impact on land and ground water, choking of drains, making land infertile, indiscriminate burning causes environmental hazards etc. Waste plastics being nonbiodegradable it can remain as a long period of landfill. Over 48 million tons of synthetic polymer material is produced in the United States every year. Plastic are made from limited resources such as petroleum. When waste plastic come in contact with light and starts photo degrading, it starts releasing harmful such as carbon, chlorine and sulfur causing the soil around them to decay, contributing many complications for cultivation. Waste plastics also end up in the ocean, where it becomes small particles due to the reaction caused by the sun ray and salt from the ocean. Million of ocean habitants die from consuming these small plastic particles when they mistake them for food. To solve this problem countries are resorting to dumping the waste plastics, which requires a lot of effort and money yet they are only able to recycle a fraction of waste plastics. This developed a new technology which will remove these waste plastics form landfill and ocean and convert them into useful liquid fuels. The fuels show high potential for commercialization due to the fact, its influence to the environment. Keywords: waste plastics, fuel, energy, polystyrene, high density polyethylene, thermal, environmental.

  17. Requirements for materials of dispersion fuel elements

    International Nuclear Information System (INIS)

    Samojlov, A.G.; Kashtanov, A.I.; Volkov, V.S.

    1982-01-01

    Requirements for materials of dispersion fuel elements are considered. The necessity of structural and fissile materials compatibility at maximum permissible operation temperatures and temperatures arising in a fuel element during manufacture is pointed out. The fuel element structural material must be ductile, possess high mechanical strength minimum neutron absorption cross section, sufficient heat conductivity, good corrosion resistance in a coolant and radiation resistance. The fissile material must have high fissile isotope concentration, radiation resistance, high thermal conductivity, certain porosity high melting temperature must not change the composition under irradiation

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

    Directory of Open Access Journals (Sweden)

    HO JIN RYU

    2013-12-01

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

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

    OpenAIRE

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

    2013-01-01

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

  20. Inert matrix fuel in dispersion type fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Savchenko, A.M. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)]. E-mail: sav@bochvar.ru; Vatulin, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Morozov, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Sirotin, V.L. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Dobrikova, I.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kulakov, G.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Ershov, S.A. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kostomarov, V.P. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Stelyuk, Y.I. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)

    2006-06-30

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg{sup -1} (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  1. Inert matrix fuel in dispersion type fuel elements

    Science.gov (United States)

    Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

    2006-06-01

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  2. Quality management program according to the ISO 9001: 2000 for the high density research reactor fuel development project (CADRIP)

    International Nuclear Information System (INIS)

    Mazzini, M.; Mamberto, L.; Piazza, A.F.

    2002-01-01

    The Quality is recognized at world level as the most effective and competitive strategy for the economic survival and the prosperity of organizations. In such a sense it is used to speak of the quality globalization. To be able to be competitive and to satisfy the more and more demanding expectations of the customers, the organization should use systems every time more effective and efficient. Such systems should lead to a continuous improvement of the quality and a highest satisfaction of the customers and of all the involved sectors (personnel, collaborators, stakeholders, suppliers, society). To these effects the National Commission of Atomic Energy of the Argentine Republic (CNEA) there is elaborated, ace necessary methodological instrument to guide and assessment t the process of continuous improvement of all the activities of the organization, to Quality Management Program (PGC) from which the Quality Systems of the facilities, laboratories and projects depends. The Quality System of the Development of Nuclear Fuels of High Density Project (CADRIP) was develop in accordance of the requirements of the PGC and the ISO 9001:2000, IAEA 50-C/SG-Q and AR 3.6.1. of the Nuclear Regulatory Body standards. The work describes this quality system, the systematic identification and administration of the processes used for the organization and the interaction among them. (author)

  3. Reliability analysis of dispersion nuclear fuel elements

    Science.gov (United States)

    Ding, Shurong; Jiang, Xin; Huo, Yongzhong; Li, Lin an

    2008-03-01

    Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

  4. Reliability analysis of dispersion nuclear fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Ding Shurong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)], E-mail: dsr1971@163.com; Jiang Xin [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Huo Yongzhong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)], E-mail: yzhuo@fudan.edu.cn; Li Linan [Department of Mechanics, Tianjin University, Tianjin 300072 (China)

    2008-03-15

    Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

  5. A general evaluation of the irradiation behaviour of dispersion fuels

    International Nuclear Information System (INIS)

    Hofman, G.L.

    1995-01-01

    The irradiation behaviour of aluminum-based dispersion fuels is evaluated with emphasis on metallurgical processes that control the dispersion behaviour. Phase transformations and microstructural changes resulting from fuel-matrix interactions and the effect of fissioning in fuel are discussed. (author)

  6. Uranium dispersion in the coating of weak-acid-resin-deprived HTGR fuel microspheres

    International Nuclear Information System (INIS)

    Weber, G.W.; Beatty, R.L.; Tennery, V.J.; Lackey, W.J. Jr.

    1976-02-01

    The current reference HTGR recycle fuel particle is a UO 2 /UC 2 kernel with a Triso coating comprising a low-density pyrocarbon (PyC) buffer, a high-density PyC inner LTI coating, SiC, and a high-density PyC outer LTI. The kernel is fabricated from a weak-acid ion exchange resin (WAR). Microradiographic examination of coated WAR particles has demonstrated that considerable U can be transferred from the kernel to the buffer coating during fabrication. Investigation of causes of fuel dispersion has indicated several different factors that contribute to fuel redistribution if not properly controlled. The presence of a nonequilibrium UC/sub 1-x/O/sub x/ (0 less than or equal to x less than or equal to 0.3) phase had no significant effect on initiating fuel dispersion. Gross exposure of the completed fuel kernel to ambient atmosphere or to water vapor at room temperature produced very minimal levels of dispersion. Exposure of the fuel to perchloroethylene during buffer and inner LTI deposition produced massive redistribution. Fuel redistribution observed in Triso-coated particles results from permeation of the inner LTI by HCl during SiC deposition. As the decomposition of CH 3 Cl 3 Si is used to deposit SiC, chlorine is readily available during this process. The permeability of the inner LTI coating has a marked effect on the extent of this mode of fuel dispersion. LTI permeability was determined by chlorine leaching studies to be a strong function of density, coating gas dilution, and coating temperature but relatively unaffected by application of a seal coat, variations in coating thickness, and annealing at 1800 0 C. Mechanical attrition of the kernels during processing was identified as a potential source of U-bearing fines that may be incorporated into the coating in some circumstances

  7. 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.

  8. 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

  9. 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

  10. 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.

  11. 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

  12. 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)

  13. 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

  14. 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

  15. Calculation simulation of equivalent irradiation swelling for dispersion nuclear fuel

    International Nuclear Information System (INIS)

    Cai Wei; Zhao Yunmei; Gong Xin; Ding Shurong; Huo Yongzhong

    2015-01-01

    The dispersion nuclear fuel was regarded as a kind of special particle composites. Assuming that the fuel particles are periodically distributed in the dispersion nuclear fuel meat, the finite element model to calculate its equivalent irradiation swelling was developed with the method of computational micro-mechanics. Considering irradiation swelling in the fuel particles and the irradiation hardening effect in the metal matrix, the stress update algorithms were established respectively for the fuel particles and metal matrix. The corresponding user subroutines were programmed, and the finite element simulation of equivalent irradiation swelling for the fuel meat was performed in Abaqus. The effects of the particle size and volume fraction on the equivalent irradiation swelling were investigated, and the fitting formula of equivalent irradiation swelling was obtained. The results indicate that the main factors to influence equivalent irradiation swelling of the fuel meat are the irradiation swelling and volume fraction of fuel particles. (authors)

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

    International Nuclear Information System (INIS)

    Suripto, A; Yuwono

    1998-01-01

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

  17. Detailed analysis of uranium silicide dispersion fuel swelling

    International Nuclear Information System (INIS)

    Hofmann, G.L.; Ryu, Woo-Seog

    1991-01-01

    Swelling of U 3 Si and U 3 Si 2 is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and micro structural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide dispersion fuel. (orig.)

  18. 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

  19. 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)

  20. 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)

  1. 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)

  2. 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)

  3. 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)

  4. 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)

  5. 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

  6. Fabrication of oxide dispersion strengthened ferritic clad fuel pins

    International Nuclear Information System (INIS)

    Zirker, L.R.; Bottcher, J.H.; Shikakura, S.; Tsai, C.L.

    1991-01-01

    A resistance butt welding procedure was developed and qualified for joining ferritic fuel pin cladding to end caps. The cladding are INCO MA957 and PNC ODS lots 63DSA and 1DK1, ferritic stainless steels strengthened by oxide dispersion, while the end caps are HT9 a martensitic stainless steel. With adequate parameter control the weld is formed without a residual melt phase and its strength approaches that of the cladding. This welding process required a new design for fuel pin end cap and weld joint. Summaries of the development, characterization, and fabrication processes are given for these fuel pins. 13 refs., 6 figs., 1 tab

  7. 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...

  8. 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.

  9. 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)

  10. 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

  11. 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.

  12. 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.

  13. 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)

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

    International Nuclear Information System (INIS)

    Denis, Alicia; Soba, Alejandro

    2005-01-01

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

  15. Fuel oil and dispersant toxicity to the Antarctic sea urchin (Sterechinus neumayeri).

    Science.gov (United States)

    Alexander, Frances J; King, Catherine K; Reichelt-Brushett, Amanda J; Harrison, Peter L

    2017-06-01

    The risk of a major marine fuel spill in Antarctic waters is increasing, yet there are currently no standard or suitable response methods under extreme Antarctic conditions. Fuel dispersants may present a possible solution; however, little data exist on the toxicity of dispersants or fuels to Antarctic species, thereby preventing informed management decisions. Larval development toxicity tests using 3 life history stages of the Antarctic sea urchin (Sterechinus neumayeri) were completed to assess the toxicity of physically dispersed, chemically dispersed, and dispersant-only water-accommodated fractions (WAFs) of an intermediate fuel oil (IFO 180, BP) and the chemical dispersant Slickgone NS (Dasic International). Despite much lower total petroleum hydrocarbon concentrations, physically dispersed fuels contained higher proportions of low-to-intermediate weight carbon compounds and were generally at least an order of magnitude more toxic than chemically dispersed fuels. Based on concentrations that caused 50% abnormality (EC50) values, the embryonic unhatched blastula life stage was the least affected by fuels and dispersants, whereas the larval 4-armed pluteus stage was the most sensitive. The present study is the first to investigate the possible implications of the use of fuel dispersants for fuel spill response in Antarctica. The results indicate that the use of a fuel dispersant did not increase the hydrocarbon toxicity of IFO 180 to the early life stages of Antarctic sea urchins, relative to physical dispersal. Environ Toxicol Chem 2017;36:1563-1571. © 2016 SETAC. © 2016 SETAC.

  16. 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

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. Framatome offers new high density Cadminox racks

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    Framatome have developed a new material called Cadminox for use in high density spent fuel storage racks. It is claimed that Cadminox will remain stable stable in pond storage when racks submerged in boronated water are irradiated by the spent fuel they contain. A brief description of the storage module is given, including the aseismic bearing device which minimises loads on pond walls, racks and fuel assemblies. (UK)

  2. Neutronic, thermal-hydraulics and safety calculations of a Miniplate Irradiation Device (MID) of dispersion type fuel elements

    International Nuclear Information System (INIS)

    Domingos, Douglas Borges

    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 is used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 - Al dispersion fuels, LEU type (19.75 % 235 U) with uranium densities of, respectively, 3.2 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 2DB 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 codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation should occur without adverse consequences in the IEA-R1 reactor. (author)

  3. 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.

  4. 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.

  5. 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.

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

    International Nuclear Information System (INIS)

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

    1987-10-01

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

  7. 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

  8. 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)

  9. Modeling solid-fuel dispersal during slow loss-of-flow-type transients

    International Nuclear Information System (INIS)

    DiMelfi, R.J.; Fenske, G.R.

    1981-01-01

    The dispersal, under certain accident conditions, of solid particles of fast-reactor fuel is examined in this paper. In particular, we explore the possibility that solid-fuel fragmentation and dispersal can be driven by expanding fission gas, during a slow LOF-type accident. The consequences of fragmentation are studied in terms of the size and speed of dispersed particles, and the overall quantity of fuel moved. (orig.)

  10. Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures; Desenvolvimento de um combustivel de alta densidade a base das ligas uranio-molibdenio com alta compatibilidade em altas temperaturas

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Fabio Branco Vaz de

    2008-07-01

    This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature

  11. 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

  12. Influence of the silicon content on the core corrosion properties of dispersion type fuel plates

    International Nuclear Information System (INIS)

    Calvo, C.; Saenz de Tejada, L. M.; Diaz Diaz, J.

    1969-01-01

    A new process to produce aluminium base dispersion type fuel plates has been developed at the Spanish JEN (Junta de Energia Nuclear). The dispersed fuel material is obtained by an aluminothermic process to render a stoichiometric cermet of UAI 3 and AI 2 O 3 according to the reaction. (Author)

  13. Homogeneous forming technology of composite materials and its application to dispersion nuclear fuel

    International Nuclear Information System (INIS)

    Hong, Soon Hyun; Ryu, Ho Jin; Sohn, Woong Hee; Kim, Chang Kyu

    1997-01-01

    Powder metallurgy processing technique of metal matrix composites is reviewed and its application to process homogeneous dispersion nuclear fuel is considered. The homogeneous mixing of reinforcement with matrix powders is very important step to process metal matrix composites. The reinforcement with matrix powders is very important step to process metal matrix composites. The reinforcement can be ceramic particles, whiskers or chopped fibers having high strength and high modulus. The blended powders are consolidated into billets and followed by various deformation processing, such as extrusion, forging, rolling or spinning into final usable shapes. Dispersion nuclear fuel is a class of metal matrix composite consisted of dispersed U-compound fuel particles and metallic matrix. Dispersion nuclear fuel is fabricated by powder metallurgy process such as hot pressing followed by hot extrusion, which is similar to that of SiC/Al metal matrix composite. The fabrication of homogeneous dispersion nuclear fuel is very difficult mainly due to the inhomogeneous mixing characteristics of the powders from quite different densities between uranium alloy powders and aluminum powders. In order to develop homogeneous dispersion nuclear fuel, it is important to investigate the effect of powder characteristics and mixing techniques on homogeneity of dispersion nuclear fuel. An new quantitative analysis technique of homogeneity is needed to be developed for more accurate analysis of homogeneity in dispersion nuclear fuel. (author). 28 refs., 7 figs., 1tab

  14. TREAT experimental data base regarding fuel dispersals in LMFBR loss-of-flow accidents

    International Nuclear Information System (INIS)

    Simms, R.; Fink, C.L.; Stanford, G.S.; Regis, J.P.

    1981-01-01

    The reactivity feedback from fuel relocation is a central issue in the analysis of loss-of-flow (LOF) accidents in LMFBRs. Fuel relocation has been studied in a number of LOF simulations in the TREAT reactor. In this paper the results of these tests are analyzed, using, as the principal figure of merit, the changes in equivalent fuel worth associated with the fuel motion. The equivalent fuel worth was calculated from the measured axial fuel distributions by weighting the data with a typical LMFBR fuel-worth function. At nominal power, the initial fuel relocation resulted in increases in equivalent fuel worth. Above nominal power the fuel motion was dispersive, but the dispersive driving forces could not unequivocally be identified from the experimental data

  15. A Deformation Model of TRU Metal Dispersion Fuel Rod for HYPER

    International Nuclear Information System (INIS)

    Lee, Byoung Oon; Hwang, Woan; Park, Won S.

    2002-01-01

    Deformation analysis in fuel rod design is essential to assure adequate fuel performance and integrity under irradiation conditions. An in-reactor performance computer code for a dispersion fuel rod is being developed in the conceptual design stage of blanket fuel for HYPER. In this paper, a mechanistic deformation model was developed and the model was installed into the DIMAC program. The model was based on the elasto-plasticity theory and power-law creep theory. The preliminary deformation calculation results for (TRU-Zr)-Zr dispersion fuel predicted by DIMAC were compared with those of silicide dispersion fuel predicted by DIFAIR. It appeared that the deformation levels for (TRU-Zr)-Zr dispersion fuel were relatively higher than those of silicide fuel. Some experimental tests including in-pile and out-pile experiments are needed for verifying the predictive capability of the DIMAC code. An in-reactor performance analysis computer code for blanket fuel is being developed at the conceptual design stage of blanket fuel for HYPER. In this paper, a mechanistic deformation model was developed and the model was installed into the DIMAC program. The model was based on the elasto-plasticity theory and power-law creep theory. The preliminary deformation calculation results for (TRUZr)- Zr dispersion fuel predicted by DIMAC were compared with those of silicide dispersion fuel predicted by DIFAIR. It appears that the deformation by swelling within fuel meat is very large for both fuels, and the major deformation mechanism at cladding is creep. The swelling strain is almost constant within the fuel meat, and is assumed to be zero in the cladding made of HT9. It is estimated that the deformation levels for (TRU-Zr)-Zr dispersion fuel were relatively higher than those of silicide fuel, and the dispersion fuel performance may be limited by swelling. But the predicted volume change of the (TRU-Zr)-Zr dispersion fuel models is about 6.1% at 30 at.% burnup. The value of cladding

  16. 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.

  17. High density Polyethylene plastic waste treatment with microwave heating pyrolysis method using coconut-shell activated carbon to produce alternative fuels

    Science.gov (United States)

    Juliastuti, S. R.; Hisbullah, M. I.; Abdillah, M.

    2018-03-01

    Pyrolysis is a technology that could crack polimer such as plastic waste into alternative fuels. This research uses microwave heating methode, which more efficient than conventional heating methode. The plastic waste used is 200 grams of HDPE, with feed to catalyst weight ratio are 1:1, 0.6:1, 0.4:1. Pyrolysis was run at temperatures of 250, 300, 350, & 400 °C for 15, 30 and 45 min. From the experimental result, the best variable of pyrolysis process with microwave method is at 45 minutes, at 400°C, and 1:1 feed to catalyst weight ratio. Result shows that yield of liquid and gas product is 99.22%; yield of residue is 0.78%; value of liquid product’s composition (cycloparaffin and n-paraffin) is 54.09% and concentration of methane gas is 10.2%.

  18. Some recent efforts toward high density implosions

    International Nuclear Information System (INIS)

    McClellan, G.E.

    1980-01-01

    Some recent Livermore efforts towards achieving high-density implosions are presented. The implosion dynamics necessary to compress DT fuel to 10 to 100 times liquid density are discussed. Methods of diagnosing the maximum DT density for a specific design are presented along with results to date. The dynamics of the double-shelled target with an exploding outer shell are described, and some preliminary experimental results are presented

  19. 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.

  20. 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.

  1. 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.

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

    Directory of Open Access Journals (Sweden)

    H.J. RYU

    2013-12-01

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

  3. Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Leppaenen, J.

    2007-01-01

    High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)

  4. Photoionization and High Density Gas

    Science.gov (United States)

    Kallman, T.; Bautista, M.; White, Nicholas E. (Technical Monitor)

    2002-01-01

    We present results of calculations using the XSTAR version 2 computer code. This code is loosely based on the XSTAR v.1 code which has been available for public use for some time. However it represents an improvement and update in several major respects, including atomic data, code structure, user interface, and improved physical description of ionization/excitation. In particular, it now is applicable to high density situations in which significant excited atomic level populations are likely to occur. We describe the computational techniques and assumptions, and present sample runs with particular emphasis on high density situations.

  5. High-density multicore fibers

    DEFF Research Database (Denmark)

    Takenaga, K.; Matsuo, S.; Saitoh, K.

    2016-01-01

    High-density single-mode multicore fibers were designed and fabricated. A heterogeneous 30-core fiber realized a low crosstalk of −55 dB. A quasi-single-mode homogeneous 31-core fiber attained the highest core count as a single-mode multicore fiber.......High-density single-mode multicore fibers were designed and fabricated. A heterogeneous 30-core fiber realized a low crosstalk of −55 dB. A quasi-single-mode homogeneous 31-core fiber attained the highest core count as a single-mode multicore fiber....

  6. Detailed analysis of uranium silicide dispersion fuel swelling

    International Nuclear Information System (INIS)

    Hofman, G.L.; Ryu, Woo-Seog.

    1989-01-01

    Swelling of U 3 Si and U 3 Si 2 is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and microstructural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide disperson fuel. 5 refs., 10 figs

  7. 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.

  8. 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

  9. 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.

  10. Modeling the influence of interaction layer formation on thermal conductivity of U–Mo dispersion fuel

    International Nuclear Information System (INIS)

    Burkes, Douglas E.; Casella, Andrew M.; Huber, Tanja K.

    2015-01-01

    Highlights: • Hsu equation provides best thermal conductivity estimate of U–Mo dispersion fuel. • Simple model considering interaction layer formation was coupled with Hsu equation. • Interaction layer thermal conductivity is not the most important attribute. • Effective thermal conductivity is mostly influenced by interaction layer formation. • Fuel particle distribution also influences the effective thermal conductivity. - Abstract: The Global Threat Reduction Initiative Program continues to develop existing and new test reactor fuels to achieve the maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Currently, the program is focused on assisting with the development and qualification of a fuel design that consists of a uranium–molybdenum (U–Mo) alloy dispersed in an aluminum matrix. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix, porosity that forms during fabrication of the fuel plates or rods, and upon the concentration of the dispersed phase within the matrix. This paper develops and validates a simple model to study the influence of interaction layer formation, dispersed particle size, and volume fraction of dispersed phase in the matrix on the effective conductivity of the composite. The model shows excellent agreement with results previously presented in the literature. In particular, the thermal conductivity of the interaction layer does not appear to be as important in determining the effective conductivity of the composite, while formation of the interaction layer and subsequent consumption of the matrix reveals a rather significant effect. The effective thermal conductivity of the composite can be influenced by the dispersed particle distribution by minimizing interaction

  11. DART model for thermal conductivity of U3Si2 Aluminum dispersion fuel

    International Nuclear Information System (INIS)

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

    2004-01-01

    This paper describes the primary physical models that form the basis of the DART model for calculating irradiation-induced changes in the thermal conductivity of aluminum dispersion fuel. DART calculations of fuel swelling, pore closure, and thermal conductivity are compared with measured values. (author)

  12. DART model for thermal conductivity of U3Si2 aluminum dispersion fuel

    International Nuclear Information System (INIS)

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

    1995-09-01

    This paper describes the primary physical models that form the basis of the DART model for calculating irradiation-induced changes in the thermal conductivity of aluminium dispersion fuel. DART calculations of fuel swelling, pore closure, and thermal conductivity are compared with measured values

  13. 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)

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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)

  19. Fate of dispersed marine fuel oil in sediment under pre-spill application strategy

    International Nuclear Information System (INIS)

    Jian Hua

    2004-01-01

    A comparison of the movement of dispersed oil in marine sediment under two dispersant application scenarios, applied prior to and after oil being spilled overboard, was examined. The pre-spill application scenario caused much less oil to be retained in the top sediment than post-spill scenario. The difference in oil retention in the top sediment between pre- and post-spill application scenario increased with increase in fuel oil temperature. For fuel oil above 40 o C, the difference in the effect of pre-spill application strategy under various water temperatures was negligible. When soap water was used as replacement for chemical dispersant, almost one-half as much oil was retained in the top sediment as that when using chemical dispersant. The adsorption of dispersed oil to the top sediment was almost proportionally decreased with doubling of soap dosage. (Author)

  20. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    Science.gov (United States)

    Govers, K.; Verwerft, M.

    2016-09-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

  1. 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)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  3. Structures of the particles of the condensed dispersed phase in solid fuel combustion products plasma

    International Nuclear Information System (INIS)

    Samaryan, A.A.; Chernyshev, A.V.; Nefedov, A.P.; Petrov, O.F.; Fortov, V.E.; Mikhailov, Yu.M.; Mintsev, V.B.

    2000-01-01

    The results of experimental investigations of a type of dusty plasma which has been least studied--the plasma of solid fuel combustion products--were presented. Experiments to determine the parameters of the plasma of the combustion products of synthetic solid fuels with various compositions together with simultaneous diagnostics of the degree of ordering of the structures of the particles of the dispersed condensed phase were performed. The measurements showed that the charge composition of the plasma of the solid fuels combustion products depends strongly on the easily ionized alkali-metal impurities which are always present in synthetic fuel in one or another amount. An ordered arrangement of the particles of a condensed dispersed phase in structures that form in a boundary region between the high-temperature and condensation zones was observed for samples of aluminum-coated solid fuels with a low content of alkali-metal impurities

  4. Modeling and preliminary analysis on the temperature profile of the (TRU-Zr)-Zr dispersion fuel rod for HYPER

    International Nuclear Information System (INIS)

    Lee, B. W.; Hwang, W.; Lee, B. S.; Park, W. S.

    2000-01-01

    Either TRU-Zr metal alloy or (TRU-Zr)-Zr dispersion fuel is considered as a blanket fuel for HYPER(Hybrid Power Extraction Reactor). In order to develop the code for dispersion fuel rod performance analysis under steady state condition, the fuel temperature distribution model which is the one of the most important factors in a fuel performance code has been developed in this paper,. This developed model computes the one dimensional radial temperature distribution of a cylindrical fuel rod. The temperature profile results by this model are compared with the temperature distributions of U 3 Si-A1 dispersion fuel and TRU-Zr metal alloy fuel. This model will be installed in performance analysis code for dispersion fuel

  5. 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

  6. 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)

  7. Polyisobutenylsuccinimides as detergents and dispersants in fuel: infrared spectroscopy application

    Energy Technology Data Exchange (ETDEWEB)

    Aleman-Vazquez, L.O.; Villagomez-Ibarra, J.R. [Instituto Mexicano del Petrole, San Bartolo Atepehuacan (Mexico). Gerencia de Productos Quimicos

    2001-05-01

    Polyalkenylsuccinimides were synthesized and their dispersancy properties evaluated in an internal combustion engine. The synthesis is based on the reaction of polyisobutene with maleic anhydride as the first step. The polyisobutenylsuccinic anhydride obtained reacts with primary amines in the last step to give polyalkenylsuccinimides. The results of the evaluations showed that some polyisobutenylsuccinimides reduce the deposit formation in the intake system with good efficiency. Infrared spectra of the prepared compounds allowed their identification. 16 refs., 2 tabs.

  8. 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)

  9. 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)

  10. A global-scale dispersion analysis of iodine-129 from nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Nishizawa, Masato; Suzuki, Takashi; Nagai, Haruyasu; Togawa, Orihiko

    2010-01-01

    A three-dimensional global chemical transport model, MOZART-2, is applied to investigate the global-sale dispersion of Iodine-129 from nuclear fuel reprocessing plants. The concentration and deposition of 129 I obtained by MOZART-2 are dispersed all over the Northern Hemisphere. The emission of 129 I to the atmosphere is thus important in considering the transport of 129 I to remote sites. (author)

  11. Model development of UO_2-Zr dispersion plate-type fuel behavior at early phase of severe accident and molten fuel meat relocation

    International Nuclear Information System (INIS)

    Zhang Zhuohua; Yu Junchong; Peng Shinian

    2014-01-01

    According to former study on oxygen diffusion, Nb-Zr solid reaction and UO_2-Zr solid reaction, the models of oxidation, solid reaction in fuel meat and relocation of molten fuel meat are developed based on structure and material properties of UO_2-Zr dispersion plate-type fuel, The new models can supply theoretical elements for the safety analysis of the core assembled with dispersion plate-type fuel under severe accident. (authors)

  12. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    International Nuclear Information System (INIS)

    Govers, K.; Verwerft, M.

    2016-01-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive. - Highlights: • We performed Discrete Element Methods simulation for fuel relocation and dispersal during LOCA transients. • The approach provides a mechanistic description of these phenomena. • The approach shows the ability of the technique to reproduce experimental observations. • The packing fraction in the balloon is shown to stabilize at 50–60%.

  13. Milling uranium silicide powder for dispersion nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, E.; Silva, D.G.; Souza, J.A.B.; Durazzo, M. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Riella, H.G. [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil)

    2009-07-01

    Full text: Uranium silicide (U3Si2) is presently considered the best fuel qualified so far in terms of uranium loading and performance. Stability of the U3Si2 fuel with uranium density of 4.8 g/cm3 was confirmed by burnup stability tests performed during the Reduced Enrichment for Research and Test Reactors (RERTR) program. This fuel was chosen to compose the first core of the new Brazilian Multipurpose Research Reactor (RMB), planned to be constructed in the next years. This new reactor will consume bigger quantities of U3Si2 powder, when compared with the small consumption of the IEA-R1 research reactor of IPEN-CNEN/SP, the unique MTR type research reactor operating in the country. At the present time, the milling operation of U3Si2 ingots is made manually. In order to increase the powder production capacity, the manual milling must be replaced by an automated procedure. This paper describes a new milling machine and procedure developed to produce U3Si2 powder with higher efficiency. (author)

  14. 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.

  15. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    International Nuclear Information System (INIS)

    Sinha, V.P.; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P.

    2009-01-01

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and γ-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes

  16. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, V.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)], E-mail: vedsinha@barc.gov.in; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2009-04-03

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and {gamma}-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes.

  17. 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

  18. 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)

  19. Fuel elements based on U3O8 dispersions in aluminium

    International Nuclear Information System (INIS)

    Haydt, H.M.

    1990-01-01

    A review of the Nuclear Metallurgy Division, now Nuclear and Energetic Research Institute, to the national Know-how development of fuel elements fabrication from U 3 O 8 dispersions in aluminium, during 1962 to 1977, are described. (C.G.C.)

  20. Visualization Study of Melt Dispersion Behavior for SFR with a Metallic Fuel under Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo Heo; Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Jungang Univ., Seoul (Korea, Republic of)

    2015-05-15

    The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition.

  1. 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.

  2. Improvement of the homogeneity of atomized particles dispersed in high uranium density research reactor fuels

    International Nuclear Information System (INIS)

    Kim, Chang-Kyu; Kim, Ki-Hwan; Park, Jong-Man; Lee, Yoon-Sang; Lee, Don-Bae; Sohn, Woong-Hee; Hong, Soon-Hyung

    1998-01-01

    A study on improving the homogeneous dispersion of atomized spherical particles in fuel meats has been performed in connection with the development of high uranium density fuel. In comparing various mixing methods, the better homogeneity of the mixture could be obtained as in order of Spex mill, V-shape tumbler mixer, and off-axis rotating drum mixer. The Spex mill mixer required some laborious work because of its small capacity per batch. Trough optimizing the rotating speed parameter for the V-shape tumbler mixer, almost the same homogeneity as with the Spex mill could be obtained. The homogeneity of the extruded fuel meats appeared to improve through extrusion. All extruded fuel meats with U 3 Si powder of 50-volume % had fairly smooth surfaces. The homogeneity of fuel meats by V-shaped tumbler mixer revealed to be fairly good on micrographs. (author)

  3. Powder fabrication of U-Mo alloys for nuclear dispersion fuels

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  4. Powder fabrication of U-Mo alloys for nuclear dispersion fuels

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  5. 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.

  6. 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.

  7. A model to predict failure of irradiated U–Mo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

    Highlights: • Simple model to predict failure of dispersion fuel meat designs. • Evaluated as a function of fabrication parameters and irradiation conditions. • Predictions compare well with experimental measurements of miniature fuel plates. • Interaction layer formation reduces matrix strength and increases temperature. • Si additions to the matrix appear effective only at moderate heat flux and burnup. - Abstract: Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on development and qualification of a fuel design that consists of a uranium–molybdenum (U–Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. The current paper extends a failure model originally developed for UO{sub 2}-stainless steel dispersion fuels and uses currently available thermal–mechanical property information for the materials of interest in the currently proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as onset of pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the {sup 235}U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-15

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

  9. Biodegradation of dispersed marine fuel oil in sediment under engineered pre-spill application strategy

    International Nuclear Information System (INIS)

    Hua, J.

    2006-01-01

    Biodegradation of marine fuel oil was studied by monitoring changes in residual oil and populations of microorganisms in marine sediments. Biodegradation rates for dispersant and soap water were 2.09 and 2.27 g/kg per day, respectively, under pre-application strategy, suggesting that the strategy may promote MFO dispersion and provide with sufficient source of food. The effect of temperature on the effectiveness of pre-application strategy is particularly obvious for the growth of fungi and Pseudomonas maltophilia. The effect of pre-application of soap water on the tolerance of aerobic bacteria, Escherichia coli, and P. maltophilia, was gradually diminished within 25-33 days. (author)

  10. Release behavior of fission products from irradiated dispersion fuels at high temperatures

    International Nuclear Information System (INIS)

    Iwai, Takashi; Shimizu, Michio; Nakagawa, Tetsuya

    1990-02-01

    As a framework of reduced enrichment fuel program of JMTR Project, the measurements of fission products release rates at high temperatures (600degC - 1100degC) were performed in order to take the data to use for safety evaluation of LEU fuel. Three type miniplates of dispersion silicide and aluminide fuel, 20% enrichment LEU fuel with 4.8 gU/cc (U 3 Si 2 90 %, USi 10 % and U 3 Si 2 50 %, U 3 Si 50 % dispersed in aluminium) and 45 % enrichment MEU fuel with 1.6 gU/cc, were irradiated in JMTR. The burnups attained by one cycle (22 days) irradiation were within 21.6 % - 22.5 % of initial 235 U. The specimens cut down from miniplates were measured on fission products release rates by means of new apparatus specially designed for this experiment. The specimens were heated up within 600degC - 1100degC in dry air. Then fission products such as 85 Kr, 133 Xe, 131 I, 137 Cs, 103 Ru, 129m Te were collected at each temperature and measured on release rates. In the results of measurement, the release rates of 85 Kr, 133 Xe, 131 I, 129m Te from all specimens were slightly less than that of G.W. Parker's data on U-Al alloy fuel. For 137 Cs and 103 Ru from a silicide specimen (U 3 Si 2 90 %, USi 10 % dispersed in aluminium) and 137 Cs from an aluminide specimen, the release rates were slightly higher than that of G.W. Parker's. (author)

  11. Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core

    International Nuclear Information System (INIS)

    Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.

    1985-01-01

    The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO 2 and UO 2 /metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO 2 crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process

  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. Influence of dispersing additives and blend composition on stability of marine high-viscosity fuels

    Directory of Open Access Journals (Sweden)

    Т. Н. Митусова

    2017-12-01

    Full Text Available The article offers a definition of the stability of marine high-viscosity fuel from the point of view of the colloid-chemical concept of oil dispersed systems. The necessity and importance of the inclusion in the current regulatory requirements of this quality parameter of high-viscosity marine fuel is indicated. The objects of the research are high-viscosity marine fuels, the basic components of which are heavy oil residues: fuel oil that is the atmospheric residue of oil refining and viscosity breaking residue that is the product of light thermal cracking of fuel oil. As a thinning agent or distillate component, a light gas oil was taken from the catalytic cracking unit. The stability of the obtained samples was determined through the xylene equivalent index, which characterizes the stability of marine high-viscosity fuel to lamination during storage, transportation and operation processes. To improve performance, the resulting base compositions of high-viscosity marine fuels were modified by introducing small concentrations (0.05 % by weight of stabilizing additives based on oxyethylated amines of domestic origin and alkyl naphthalenes of foreign origin.

  14. Progress in qualifying low-enriched U-Mo dispersion fuels

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  15. The dispersion of alpha and beta radioactivity to the environmental from spent fuel testing in RMI

    International Nuclear Information System (INIS)

    Yuwono, I.; Pudjadi, E.

    1996-01-01

    The destructive testing of 2 spent fuels in RMI and radioactivity air release monitoring to the environmental have been done. The monitoring equipment used alpha-beta particulate monitor, Berthold LB 150 D type. The calculations using the Gaussian plume model and distributions factor showed there were no radiological effect of alpha and beta radioactivity dispersion and contribution to the environmental. The maximum average construction of alpha and beta radioactivity are 0.002% and 0.05%. (author)

  16. 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)

  17. Development of a FBR fuel pin bundle deformation analysis code 'BAMBOO' . Development of a dispersion model and its validation

    International Nuclear Information System (INIS)

    Uwaba, Tomoyuki; Ukai, Shigeharu; Asaga, Takeo

    2002-03-01

    Bundle Duct Interaction (BDI) is one of the life limiting factors of a FBR fuel subassembly. Under the BDI condition, the fuel pin dispersion would occur mainly by the deviation of the wire position due to the irradiation. In this study the effect of the dispersion on the bundle deformation was evaluated by using the BAMBOO code and following results were obtained. (1) A new contact analysis model was introduced in BAMBOO code. This model considers the contact condition at the axial position other than the nodal point of the beam element that composes the fuel pin. This improvement made it possible in the bundle deformation analysis to cause fuel pin dispersion due to the deviations of the wire position. (2) This model was validated with the results of the out-of-pile compression test with the wire deviation. The calculated pin-to-duct and pin-to-pin clearances with the dispersion model almost agreed with the test results. Therefore it was confirmed that the BAMBOO code reasonably predicts the bundle deformation with the dispersion. (3) In the dispersion bundle the pin-to-pin clearances widely scattered. And the minimum pin-to-duct clearance increased or decreased depending on the dispersion condition compared to the no-dispersion bundle. This result suggests the possibility that the considerable dispersion would affect the thermal integrity of the bundle. (author)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  19. 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.

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

    International Nuclear Information System (INIS)

    Van den Berghe, S.

    2009-01-01

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

  1. Coupled 3D neutronic and thermohydraulic calculations for a compact fuel element with disperse UMo fuel at FRM II

    International Nuclear Information System (INIS)

    Breitkreutz, H.; Roehrmoser, A.; Petry, W.

    2010-01-01

    The newly developed X 2 program system is intended to be used for high-detail 3D calculations on compact research reactor cores. Using this system, the efforts to calculate scenarios for a new fuel element for FRM II using disperse UMo (8wt% Mo, 50% enrichment) are continued. By now, a radial symmetric core model with averaged built-in components for the D 2 O tank is used. Two different scenarios are compared: The minimum fuel density of 7.5 g U/cm 3 and 8.0 g U/cm 3 with 60 days cycle length. In addition, two 'flux loss compensating' scenarios based on 8.0 g U/cm 3 with 10% higher power/longer reactor cycles are regarded. (author)

  2. Status of fuel element technology for plate type dispersion fuels with high uranium density

    International Nuclear Information System (INIS)

    Hrovat, M.; Huschka, H.; Koch, K.H.; Nazare, S.; Ondracek, G.

    1983-01-01

    A number of about 20 Material Test and Research Reactors in Germany and abroad is supplied with fuel elements by the company NUKEM. The power of these reactors differs widely ranging from up to about 100 MW. Consequently, the uranium density of the fuel elements in the meat varies considerably depending on the reactor type and is usually within the range from 0.4 to 1.3 g U/cm 3 if HEU is used. In order to convert these reactors to lower uranium enrichment (19.75% 235-U) extensive work is carried out at NUKEM since about two years with the goal to develop fuel elements with high U-density. This work is sponsored by the German Ministry for Research and Technology in the frame of the AF-program. This paper reports on the present state of development for fuel elements with high U-density fuels at NUKEM is reported. The development works were so far concentrated on UAl x , U 3 O 8 and UO 2 fuels which will be described in more detail. In addition fuel plates with new fuels like e.g. U-Si or U-Fe compounds are developed in collaboration with KfK. The required uranium densities for some typical reactors with low, medium, and high power are listed allowing a comparison of HEU and LEU uranium density requirements. The 235-U-content in the case of LEU is raised by 18%. Two different meat thicknesses are considered: Standard thickness of 0.5 mm; and increased thickness of 0.76 mm. From this data compilation the objective follows: in the case of conversion to LEU (19.75% 235-U-enrichment), uranium densities have to be made available up to 24 gU/cm 3 meat for low power level reactors, up to 33 gU/cm 3 meat for medium power level reactors, and between 5.75 and 7.03 g/cm 3 meat for high power level reactors according to this consideration

  3. Neutronic, thermal-hydraulics and accident analysis calculations for an irradiation device to be used in the qualification process of dispersion fuels in the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas Borges; Silva, Antonio Teixeira e; Umbehaun, Pedro Ernesto; Silva, Jose Eduardo Rosa da; Conti, Thadeu das Neves; Yamaguchi, Mitsuo [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Sao Paulo, SP (Brazil)], e-mail: douglasborgesdomingos@yahoo.com.br

    2009-07-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of an irradiation device placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U{sub 3}O{sub 8}-Al e U{sub 3}Si{sub 2}-Al dispersion fuels, LEU type (19.9% of {sup 235}U), with uranium densities of, respectively, 3.0 gU/cm{sup 3} and 4.8gU/cm{sup 3}. The fuel miniplates will be irradiated to nominal {sup 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, now in the conception phase. For the neutronic calculation, the computer code CITATION was 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 codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation of the fuel miniplates will happen without any adverse consequence in the IEA-R1 reactor. (author)

  4. Fabrication of CNT Dispersion Fluid by Wet-Jet Milling Method for Coating on Bipolar Plate of Fuel Cell

    Directory of Open Access Journals (Sweden)

    Anas Almowarai

    2015-01-01

    Full Text Available Water based carbon nanotube (CNT dispersion was produced by wet-jet milling method. Commercial CNT was originally agglomerated at the particle size of less than 1 mm. The wet-jet milling process exfoliated CNTs from the agglomerates and dispersed them into water. Sedimentation of the CNTs in the dispersion fluid was not observed for more than a month. The produced CNT dispersion was characterized by the SEM and the viscometer. CNT/PTFE composite film was formed with the CNT dispersion in this study. The electrical conductivity of the composite film increased to 10 times when the CNT dispersion, which was produced by the wet-jet milling method, was used as a constituent of the film. Moreover, the composite film was applied to bipolar plate of fuel cell and increased the output power of the fuel cell to 1.3 times.

  5. Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report

    International Nuclear Information System (INIS)

    Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier; Klennert, Lindsay A.; Nolte, Oliver; Molecke, Martin Alan; Autrusson, Bruno A.; Koch, Wolfgang; Pretzsch, Gunter Guido; Brucher, Wenzel; Steyskal, Michele D.

    2008-01-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO 2 , CeO 2 , plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively supported and

  6. Mechanical behaviors of the dispersion nuclear fuel plates induced by fuel particle swelling and thermal effect II: Effects of variations of the fuel particle diameters

    International Nuclear Information System (INIS)

    Ding Shurong; Wang Qiming; Huo Yongzhong

    2010-01-01

    In order to predict the irradiation mechanical behaviors of plate-type dispersion nuclear fuel elements, the total burnup is divided into two stages: the initial stage and the increasing stage. At the initial stage, the thermal effects induced by the high temperature differences between the operation temperatures and the room temperature are mainly considered; and at the increasing stage, the intense mechanical interactions between the fuel particles and the matrix due to the irradiation swelling of fuel particles are focused on. The large-deformation thermo-elasto-plasticity finite element analysis is performed to evaluate the effects of particle diameters on the in-pile mechanical behaviors of fuel elements. The research results indicate that: (1) the maximum Mises stresses and equivalent plastic strains at the matrix increase with the fuel particle diameters; the effects of particle diameters on the maximum first principal stresses vary with burnup, and the considered case with the largest particle diameter holds the maximum values all along; (2) at the cladding near the interface between the fuel meat and the cladding, the Mises stresses and the first principal stresses undergo major changes with increasing burnup, and different variations exist for different particle diameter cases; (3) the maximum Mises stresses at the fuel particles rise with the particle diameters.

  7. Dispersion of long-lived radionuclides from uranium mining, milling and fuel fabrication facilities

    International Nuclear Information System (INIS)

    Pettersson, H.B.L.

    1990-11-01

    The principal aim of the study was to gain further insight into the environmental dispersion of long-lived U series radionuclides from selected part of the nuclear fuel cycle and to assess the resulting exposure of members of the public. The specific objectives of this study were: 1. To determine the levels of natural radioactivity in the vicinity of two U deposits in Sweden and to establish whether U prospecting had generated significant radiological impact on man. 2. To investigate the spatial distributions of long-lived U series radionuclides caused by the dispersion of dust from the Ranger open-pit U mine in Australia. 3. To study the uptakes of long-lived U and T series radionuclides by the waterlily in order to facilitate assessment of natural exposures to the public and predictions of exposures arising from consumption of the plant due to any subsequent discharges of water from the Ranger U mine. 4. To investigate the spatial distributions of U isotopes in environmental air as a result of the release of radionuclides from the ABB-ATOM nuclear fuel factory at Vaesteraas in Sweden. In these investigations special emphasis was given to - activity ratio techniques suitable for distinguishing between natural and operation-related concentrations and for facilitating determination of the source of radionuclide uptake in the waterlily, and - the use of passive air samplers such as 'sticky vinyl' and bioindicators in investigating the aerial dispersion of radionuclides. (author)

  8. 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.

  9. Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements

    Science.gov (United States)

    Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing

    2009-08-01

    Considering the mutual actions between fuel particles and the metal matrix, the three-dimensional finite element models are developed to simulate the heat transfer behaviors of dispersion nuclear fuel plates. The research results indicate that the temperatures of the fuel plate might rise more distinctly with considering the particle swelling and the degraded surface heat transfer coefficients with increasing burnup; the local heating phenomenon within the particles appears when their thermal conductivities are too low. With rise of the surface heat transfer coefficients, the temperatures within the fuel plate decrease; the temperatures of the fuel plate are sensitive to the variations of the heat transfer coefficients whose values are lower, but their effects are weakened and slight when the heat transfer coefficients increase and reach a certain extent. Increasing the heat generation rate leads to elevating the internal temperatures. The temperatures and the maximum temperature differences within the plate increase along with the particle volume fractions. The surface thermal flux goes up along with particle volume fractions and heat generation rates, but the effects of surface heat transfer coefficients are not evident.

  10. 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.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  12. Neutronic, thermal-hydraulics and safety calculations of a Miniplate Irradiation Device (MID) of dispersion type fuel elements; Calculos neutronicos, termo-hidraulicos e de seguranca de um dispositivo para Irradiacao de miniplacas (DIM) de elementos combustiveis tipo dispersao

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas Borges

    2010-07-01

    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 is used to receive miniplates of U{sub 3}O{sub 8}-Al and U{sub 3}Si{sub 2}- Al dispersion fuels, LEU type (19.75 % {sup 235}U) with uranium densities of, respectively, 3.2 gU/cm{sup 3} and 4.8 gU/cm{sup 3}. The fuel miniplates will be irradiated to nominal {sup 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 2DB 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 codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation should occur without adverse consequences in the IEA-R1 reactor. (author)

  13. Development of ODS (oxide dispersion strengthened) ferritic-martensitic steels for fast reactor fuel cladding

    International Nuclear Information System (INIS)

    Ukai, Shigeharu

    2000-01-01

    In order to attain higher burnup and higher coolant outlet temperature in fast reactor, oxide dispersion strengthened (ODS) ferritic-martensitic steels were developed as a long life fuel cladding. The improvement in formability and ductility, which are indispensable in the cold-rolling method for manufacturing cladding tube, were achieved by controlling the microstructure using techniques such as recrystallization heat-treatment and α to γ phase transformation. The ODS ferritic-martensitic cladding tubes manufactured using these techniques have the highest internal creep rupture strength in the world as ferritic stainless steels. Strength level approaches adequate value at 700degC, which meets the requirement for commercial fast reactors. (author)

  14. Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Wang Qiming; Yan Xiaoqing [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Ding Shurong, E-mail: dsr1971@163.co [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Huo Yongzhong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)

    2010-04-01

    The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

  15. 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)

  16. 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)

  17. High density harp for SSCL linac

    International Nuclear Information System (INIS)

    Fritsche, C.T.; Krogh, M.L.; Crist, C.E.

    1993-01-01

    AlliedSignal Inc., Kansas City Division, and the Superconducting Super Collider Laboratory (SSCL) are collaboratively developing a high density harp for the SSCL linac. This harp is designed using hybrid microcircuit (HMC) technology to obtain a higher wire density than previously available. The developed harp contains one hundred twenty-eight 33-micron-diameter carbon wires on 0.38-mm centers. The harp features an onboard broken wire detection circuit. Carbon wire preparation and attachment processes were developed. High density surface mount connectors were located. The status of high density harp development will be presented along with planned future activities

  18. High density harp for SSCL linac

    International Nuclear Information System (INIS)

    Fritsche, C.T.; Krogh, M.L.

    1993-05-01

    AlliedSignal Inc., Kansas City Division, and the Superconducting Super Collider Laboratory (SSCL) are collaboratively developing a high density harp for the SSCL linac. This harp is designed using hybrid microcircuit (HMC) technology to obtain a higher wire density than previously available. The developed harp contains one hundred twenty-eight 33-micron-diameter carbon wires on 0.38-mm centers. The harp features an onboard broken wire detection circuit. Carbon wire preparation and attachment processes were developed. High density surface mount connectors were located. The status of high density harp development will be presented along with planned future activities

  19. Viscosity and attenuation of sound wave in high density deuterium

    International Nuclear Information System (INIS)

    Inoue, Kazuko; Ariyasu, Tomio

    1985-01-01

    The penetration of low frequency sound wave into the fuel deuterium is discussed as for laser fusion. The sound velocity and the attenuation constant due to viscosity are calculated for high density (n = 10 24 -- 10 27 cm -3 , T = 10 -1 -- 10 4 eV) deuterium. The shear viscosity of free electron gas and the bulk viscosity due to ion-ion interaction mainly contribute to the attenuation of sound wave. The sound wave of the frequency below 10 10 Hz can easily penetrate through the compressed fuel deuterium of diameter 1 -- 10 3 μm. (author)

  20. 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)

  1. 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.

  2. 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

  3. Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report.

    Energy Technology Data Exchange (ETDEWEB)

    Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Klennert, Lindsay A.; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

    2008-03-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO{sub 2}, CeO{sub 2}, plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively

  4. Effect of fission rate on the microstructure of coated UMo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Leenaers, A.; Parthoens, Y.; Cornelis, G.; Kuzminov, V.; Koonen, E.; Van den Berghe, S.; Ye, B.; Hofman, G. L.; Schulthess, Jason

    2017-10-01

    Compared to previous irradiation experiments containing UMo/Al dispersion fuel plates, the SELENIUM irradiation experiment performed at the SCK.CEN BR2 reactor in 2012 showed an improved plate swelling behavior. However, in the high burn-up area of the plates a significant increase in meat thickness was still measured. The origin of this increase is currently not firmly established, but it is clear from the observed microstructure that the swelling rate still is too high for practical purposes and needs to be reduced. It was stipulated that the swelling occurred at the high burnup areas which are also the high power zones at beginning of life. For that reason, an experiment was proposed to investigate the influence of fission rate (i.e. power) on some of the observed phenomena. For this purpose, a sibling plate to a high power (BOL>470 W/cm(2)) SELENIUM plate was irradiated during four BR2 cycles. The SELENIUM 1a fuel plate was submitted to a local maximum heat flux below 350 W/cm(2), throughout the full irradiation. At the end of the last cycle, the SELENIUM 1a fuel plate reached a maximum local burnup value of close to 75%U-235 compared to 70%U-235 for the SELENIUM high power plates. When comparing to the results on the SELENIUM plates, the non-destructive tests clearly show a continued linear swelling behavior of the low power irradiated fuel plate SELENIUM 1a in the high burn-up region. The influence of the fission rate is also evidenced in the microstructural examination of the fuel showing that there is no formation of interaction layer at the high burn-up region.

  5. The fabrication and performance of Canadian silicide dispersion fuel for test reactors

    International Nuclear Information System (INIS)

    Sears, D.F.; Wood, J.C.; Berthiaume, L.C.; Herbert, L.N.; Schaefer, J.D.

    1985-01-01

    Fuel fabrication effort is now concentrated on the commissioning of large-scale process equipment, defining product specifications, developing a quality assurance plan, and setting up a mini-computer material accountancy system. In the irradiation testing program, full-size NRU assemblies containing 20% enriched silicide dispersion fuel have been Irradiated successfully to burnups in the range 65-80 atomic percent. Irradiations have also been conducted on mini-elements having 1.2 mm diameter holes In their mid-sections, some drilled before irradiation and others after irradiation to 22-83 atomic percent burnup. Uranium was lost to the coolant in direct proportion to the surface area of exposed core material. Pre-irradiation in the intact condition appeared to reduce in-reactor corrosion. Fuel cores developed for the NRU reactor are dimensionally very stable, swelling by only 6-8% at the very high burnup of 93 atomic percent. Two important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles. Thermal ramping tests were conducted on irradiated silicide aspersion fuels. Small segments of fuel cores released 85 Kr starting at about 520 deg. C and peaking at about 680 deg C. After a holding period of 1 hour at 720 deg. C a secondary 85 Kr peak occurred during cooling (at about 330 deg. C) probably due to thermal contraction cracking. Whole mini-elements irradiated to 93 atomic percent burnup were also ramped thermally, with encouraging results. After about 0.25 h at 530 deg. C the aluminum cladding developed very localized small blisters, some with penetrating pin-hole cracks preventing gross pillowing or ballooning. (author)

  6. 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.

  7. The fabrication and performance of Canadian silicide dispersion fuel for test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sears, D F; Wood, J C; Berthiaume, L C; Herbert, L N; Schaefer, J D

    1985-07-01

    Fuel fabrication effort is now concentrated on the commissioning of large-scale process equipment, defining product specifications, developing a quality assurance plan, and setting up a mini-computer material accountancy system. In the irradiation testing program, full-size NRU assemblies containing 20% enriched silicide dispersion fuel have been Irradiated successfully to burnups in the range 65-80 atomic percent. Irradiations have also been conducted on mini-elements having 1.2 mm diameter holes In their mid-sections, some drilled before irradiation and others after irradiation to 22-83 atomic percent burnup. Uranium was lost to the coolant in direct proportion to the surface area of exposed core material. Pre-irradiation in the intact condition appeared to reduce in-reactor corrosion. Fuel cores developed for the NRU reactor are dimensionally very stable, swelling by only 6-8% at the very high burnup of 93 atomic percent. Two important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles. Thermal ramping tests were conducted on irradiated silicide aspersion fuels. Small segments of fuel cores released {sup 85}Kr starting at about 520 deg. C and peaking at about 680 deg C. After a holding period of 1 hour at 720 deg. C a secondary {sup 85}Kr peak occurred during cooling (at about 330 deg. C) probably due to thermal contraction cracking. Whole mini-elements irradiated to 93 atomic percent burnup were also ramped thermally, with encouraging results. After about 0.25 h at 530 deg. C the aluminum cladding developed very localized small blisters, some with penetrating pin-hole cracks preventing gross pillowing or ballooning. (author)

  8. Effect of fission rate on the microstructure of coated UMo dispersion fuel

    Science.gov (United States)

    Leenaers, A.; Parthoens, Y.; Cornelis, G.; Kuzminov, V.; Koonen, E.; Van den Berghe, S.; Ye, B.; Hofman, G. L.; Schulthess, Jason

    2017-10-01

    Compared to previous irradiation experiments containing UMo/Al dispersion fuel plates, the SELENIUM irradiation experiment performed at the SCK·CEN BR2 reactor in 2012 showed an improved plate swelling behavior. However, in the high burn-up area of the plates a significant increase in meat thickness was still measured. The origin of this increase is currently not firmly established, but it is clear from the observed microstructure that the swelling rate still is too high for practical purposes and needs to be reduced. It was stipulated that the swelling occurred at the high burnup areas which are also the high power zones at beginning of life. For that reason, an experiment was proposed to investigate the influence of fission rate (i.e. power) on some of the observed phenomena. For this purpose, a sibling plate to a high power (BOL>470 W/cm2) SELENIUM plate was irradiated during four BR2 cycles. The SELENIUM 1a fuel plate was submitted to a local maximum heat flux below 350 W/cm2, throughout the full irradiation. At the end of the last cycle, the SELENIUM 1a fuel plate reached a maximum local burnup value of close to 75%235U compared to 70%235U for the SELENIUM high power plates. When comparing to the results on the SELENIUM plates, the non-destructive tests clearly show a continued linear swelling behavior of the low power irradiated fuel plate SELENIUM 1a in the high burn-up region. The influence of the fission rate is also evidenced in the microstructural examination of the fuel showing that there is no formation of interaction layer at the high burn-up region.

  9. 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...

  10. Experimental studies of thermal and chemical interactions between molten aluminum and nuclear dispersion fuels with water

    International Nuclear Information System (INIS)

    Farahani, A.A.

    1997-01-01

    Because of the possibility of rapid physical and chemical molten fuel-water interactions during a core melt accident in noncommercial or experimental reactors, it is important to understand the interactions that might occur if these materials were to contact water. An existing vertical 1-D shock tube facility was improved and a gas sampling device to measure the gaseous hydrogen in the upper chamber of the shock tube was designed and built to study the impact of a water column driven downward by a pressurized gas onto both molten aluminum (6061 alloy) and oxide and silicide depleted nuclear dispersion fuels in aluminum matrices. The experiments were carried out with melt temperatures initially at 750 to 1,000 C and water at room temperature and driving pressures of 0.5 and 1 MPa. Very high transient pressures, in many cases even larger than the thermodynamic critical pressure of the water (∼ 20 MPa), were generated due to the interactions between the water and the crucible and its contents. The molten aluminum always reacted chemically with the water but the reaction did not increase consistently with increasing melt temperature. An aluminum ignition occurred when water at room temperature impacted 28.48 grams of molten aluminum at 980.3 C causing transient pressures greater than 69 MPa. No signs of aluminum ignition were observed in any of the experiments with the depleted nuclear dispersion fuels, U 3 O 8 -Al and U 3 Si 2 -Al. The greater was the molten aluminum-water chemical reaction, the finer was the debris recovered for a given set of initial conditions. Larger coolant velocities (larger driving pressures) resulted in more melt fragmentation but did not result in more molten aluminum-water chemical reaction. Decreasing the water temperature also resulted in more melt fragmentation and did not suppress the molten aluminum-water chemical reaction

  11. 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.

  12. Dispersants in an organic medium: synthesis and physicochemical study of dispersants for fuels and lubricants; Dispersants en milieu organique: synthese et etude physicochimique de dispersants pour carburants et lubrifiants

    Energy Technology Data Exchange (ETDEWEB)

    Dubois-Clochard, M.C.

    1998-11-19

    Carbonaceous deposits coming from the fuel and the lubricant are known to form over time at critical locations in an engine. In general, the deposits have an adverse effect on four functional areas which are the fuel metering system, the intake system, the lubrication system and the combustion chambers. These deposits can degrade vehicle performance and drive-ability, reduce fuel economy, increase fuel consumption and pollutant emissions and may lead to the destruction of the engine. In order to remedy these problems, detergent-dispersant additives are used in fuels and lubricants to avoid or decrease deposit adhesion on metallic surfaces and prevent from deposit aggregation. These products are mainly polymer surfactants and in this work, poly-iso-butenyl-succinimide of different structures have been studied. Firstly, 'comb like' polymers have been synthesized. Then they have been compared to classical di-bloc additives in terms of performance and action mechanism. These additives are adsorbed from their hydrophilic polyamine part on the acidic functions of the carbon black surface chosen as an engine deposit model and on the aluminium oxide function of an aluminium powder chosen as an engine wall model. The adsorption increases with temperature on the two solids. Their affinity with the solid surface increases with the length of the hydrophilic part. In the same way, changing the di-bloc structure for a comb like one lead to a better adsorption. At low concentration, it has been shown that the adsorption phenomenon was irreversible, due to the polymer structure of the polar part. Depending on the space required by the hydrophilic part on the solid surface, a more of less dense monolayer is formed. At higher concentrations, an important increase of the adsorbed amount appears. This phenomenon is totally reversible showing that the interactions additive / additive are weak. The dispersing efficiency of a comb like structure is better than a di-bloc one as

  13. 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.

  14. Mechanical Calculations on U-Mo Dispersion fuel plates with MAIA

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  15. High Density Digital Data Storage System

    Science.gov (United States)

    Wright, Kenneth D., II; Gray, David L.; Rowland, Wayne D.

    1991-01-01

    The High Density Digital Data Storage System was designed to provide a cost effective means for storing real-time data from the field-deployable digital acoustic measurement system. However, the high density data storage system is a standalone system that could provide a storage solution for many other real time data acquisition applications. The storage system has inputs for up to 20 channels of 16-bit digital data. The high density tape recorders presently being used in the storage system are capable of storing over 5 gigabytes of data at overall transfer rates of 500 kilobytes per second. However, through the use of data compression techniques the system storage capacity and transfer rate can be doubled. Two tape recorders have been incorporated into the storage system to produce a backup tape of data in real-time. An analog output is provided for each data channel as a means of monitoring the data as it is being recorded.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  17. Magnetization of High Density Hadronic Fluid

    DEFF Research Database (Denmark)

    Bohr, Henrik; Providencia, Constanca; da Providencia, João

    2012-01-01

    In the present paper the magnetization of a high density relativistic fluid of elementary particles is studied. At very high densities, such as may be found in the interior of a neutron star, when the external magnetic field is gradually increased, the energy of the normal phase of the fluid...... in the particle fluid. For nuclear densities above 2 to 3 rho(0), where rho(0) is the equilibrium nuclear density, the resulting magnetic field turns out to be rather huge, of the order of 10(17) Gauss....

  18. 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.%

  19. Fuel dispersal in high-speed aircraft/soil impact scenarios

    International Nuclear Information System (INIS)

    Tieszen, S.R.; Attaway, S.W.

    1996-01-01

    The objective of this study is to determine how the jet fuel contained in aircraft wing tanks disperses on impact with a soft terrain, i.e., soils, at high impact velocities. The approach used in this study is to combine experimental and numerical methods. Tests were conducted with an approximately 1/42 linear-scale mass-model of a 1/4 span section of a C-141 wing impacting a sand/clay mixture. The test results showed that within the uncertainty of the data, the percentage of incident liquid mass remaining in the crater is the same as that qualitatively described in earlier napalm bomb development studies. Namely, the percentage of fuel in the crater ranges from near zero for grazing impacts to 25%--50% for high angles of impact. To support a weapons system safety assessment (WSSA), the data from the current study have been reduced to correlations. The numerical model used in the current study is a unique coupling of a Smooth Particle Hydrodynamics (SPH) method with the transient dynamics finite element code PRONTO. Qualitatively, the splash, erosion, and soil compression phenomena are all numerically predicted. Quantitatively, the numerical method predicted a smaller crater cross section than was observed in the tests

  20. Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

    Science.gov (United States)

    Jiang, Yijie; Wang, Qiming; Cui, Yi; Huo, Yongzhong; Ding, Shurong

    2011-06-01

    Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent

  1. Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Jiang Yijie; Wang Qiming; Cui Yi; Huo Yongzhong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Ding Shurong, E-mail: dsr1971@163.com [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)

    2011-06-15

    Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent

  2. Supernovae and high density nuclear matter

    International Nuclear Information System (INIS)

    Kahana, S.

    1986-01-01

    The role of the nuclear equation of state (EOS) in producing prompt supernova explosions is examined. Results of calculations of Baron, Cooperstein, and Kahana incorporating general relativity and a new high density EOS are presented, and the relevance of these calculations to laboratory experiments with heavy ions considered. 31 refs., 6 figs., 2 tabs

  3. Supernovae and high density nuclear matter

    Energy Technology Data Exchange (ETDEWEB)

    Kahana, S.

    1986-01-01

    The role of the nuclear equation of state (EOS) in producing prompt supernova explosions is examined. Results of calculations of Baron, Cooperstein, and Kahana incorporating general relativity and a new high density EOS are presented, and the relevance of these calculations to laboratory experiments with heavy ions considered. 31 refs., 6 figs., 2 tabs.

  4. High Density GEOSAT/GM Altimeter Data

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The high density Geosat/GM altimeter data south of 30 S have finally arrived. In addition, ERS-1 has completed more than 6 cycles of its 35-day repeat track. These...

  5. Wavelength Dispersive X-ray Fluorescence Analysis of Actinides in Dissolved Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    O' Hara, David [Parallax Research, Inc., Tallahassee, FL (United States)

    2015-10-15

    There is an urgent need for an instrument that can quickly measure the concentration of Plutonium and other Actinides mixed with Uranium in liquids containing dissolved spent fuel rods. Parallax Research, Inc. proposes to develop an x-ray spectrometer capable of measuring U, Np and Pu in dissolved nuclear fuel rod material to less than 10 ppm levels to aid in material process control for these nuclear materials. Due to system noise produced by high radioactivity, previous x-ray spectrometers were not capable of low level measurements but the system Parallax proposed has no direct path for undesired radiation to get to the detector and the detector in the proposed device is well shielded from scatter and has very low dark current. In addition, the proposed spectrometer could measure these three elements simultaneously, also measuring background positions with an energy resolution of roughly 100 eV making it possible to see a small amount of Pu that would be hidden under the tail of the U peak in energy dispersive spectrometers. Another nearly identical spectrometer could be used to target Am and Cm if necessary. The proposed spectrometer needs only a tiny sample of roughly 1 micro-liter (1 mm3) and the measurement can be done with the liquid flowing in a radiation and chemical immune quartz capillary protected by a stainless steel rod making it possible to continuously monitor the liquid or to use a capillary manifold to measure other liquid streams. Unlike other methods such as mass spectroscopy where the sample must be taken to a remote facility and might take days for turn-around, the proposed measurement should take less than an hour. This spectrometer could enable near real-time measurement of U, Pu and Np in dilute dissolved spent nuclear fuel rod streams.

  6. On the possibility of reprocessing of fuel elements of dispersion type with copper matrix by pyrochemical methods

    International Nuclear Information System (INIS)

    Vasin, B.D.; Ivanov, V.A.; Shchetinskij, A.V.; Vavilov, S.K.; Savochkin, Yu.P.; Bychkov, A.V.; Kormilitsyn, M.V.

    2005-01-01

    A consideration is given to pyrochemical processes suitable for separation of uranium dioxide from structural materials when reprocessing cermet type fuel elements. The estimation of the possibility to apply liquid antimony and bismuth, potassium and copper chlorides melts is made. The specimens compacted of copper and uranium dioxide powders in a stainless steel can are used as simulators of fuel element sections. It is concluded that the dissolution of structural materials in molten salts at the stage of uranium dioxide concentration is the process of choice for reprocessing of dispersion type fuel elements [ru

  7. Irradiation behavior of uranium-molybdenum dispersion fuel: Fuel performance data from RERTR-1 and RERTR-2

    International Nuclear Information System (INIS)

    Meyer, M.K.; Clark, C.R.; Hayes, S.L.; Strain, R.V.; Hofman, G.L.; Snelgrove, J.L.; Park, J.M.; Kim, K.H.

    1999-01-01

    This paper presents quantitative data on the irradiation behavior of uranium-molybdenum fuels from the low temperature RERTR-1 and -2 experiments. Fuel swelling measurements of U-Mo fuels at ∼40% and ∼70% burnup are presented. The rate of fuel-matrix interaction layer growth is estimated. Microstructures of fuel in the pre- and postirradiation condition were compared. Based on these data, a qualitative picture of the evolution of the U-Mo fuel microstructure during irradiation has been developed. Estimates of uranium-molybdenum fuel swelling and fuel-matrix interaction under high-power research reactor operating conditions are presented. (author)

  8. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    International Nuclear Information System (INIS)

    Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

    2012-01-01

    CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

  9. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    Science.gov (United States)

    Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

    2012-08-01

    CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

  10. Operation and control of high density tokamak reactors

    International Nuclear Information System (INIS)

    Attenberger, S.E.; McAlees, D.G.

    1976-01-01

    The incentive for high density operation of a tokamak reactor was discussed. It is found that high density permits ignition in a relatively small, moderately elongated plasma with a moderate magnetic field strength. Under these conditions, neutron wall loadings approximately 4 MW/m 2 must be tolerated. The sensitivity analysis with respect to impurity effects shows that impurity control will most likely be necessary to achieve the desired plasma conditions. The charge exchange sputtered impurities are found to have an important effect so that maintaining a low neutral density in the plasma is critical. If it is assumed that neutral beams will be used to heat the plasma to ignition, high energy injection is required (approximately 250 keV) when heating is accompished at full density. A scenario is outlined where the ignition temperature is established at low density and then the fueling rate is increased to attain ignition. This approach may permit beams with energies being developed for use in TFTR to be successfully used to heat a high density device of the type described here to ignition

  11. Bioaccumulation and subacute toxicity of mechanically and chemically dispersed heavy fuel oil in sea urchin (Glyptocidaris crenulari

    Directory of Open Access Journals (Sweden)

    Bailin Yang

    2015-12-01

    Full Text Available Oil spills have a disastrous ecological impact on ecosystems but few data are available for the effects of dispersed oil on benthic marine organisms. In order to provide information for assessment, we analysed the hydrocarbon compositions of the mechanically dispersed water accommodated fraction (MDWAF and the chemically dispersed water accommodated fraction (CDWAF of No. 120 fuel oil, their bioaccumulation, and DNA damage related to oil exposure, using the sea urchin as a sentinel organism. The results show that the concentration of polycyclic aromatic hydrocarbon in the tissues of sea urchin exposed to the CDWAF is higher than that of those exposed to the MDWAF. The single cell gel electrophoresis assay results also indicated higher DNA damage from exposure to the CDWAF of oil. Thus, dispersants should be applied with caution in oil spill accidents.

  12. 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)

  13. Spin polarization in high density quark matter

    DEFF Research Database (Denmark)

    Bohr, Henrik; Panda, Prafulla K.; Providênci, Constanca

    2013-01-01

    We investigate the occurrence of a ferromagnetic phase transition in high density hadronic matter (e.g., in the interior of a neutron star). This could be induced by a four-fermion interaction analogous to the one which is responsible for chiral symmetry breaking in the Nambu-Jona-Lasinio model, ...... the so-called 2 flavor super-conducting phase to the ferromagnetic phase arises. The color-flavor-locked phase may be completely hidden by the FP....

  14. The car parking problem at high densities

    Science.gov (United States)

    Burgos, E.; Bonadeo, H.

    1989-04-01

    The radial distribution functions of random 1-D systems of sequential hard rods have been studied in the range of very high densities. It is found that as the number of samples rejected before completion increases, anomalies in the pairwise distribution functions arise. These are discussed using analytical solutions for systems of three rods and numerical simulations with twelve rods. The probabilities of different spatial orderings with respect to the sequential order are examined.

  15. 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.

  16. Computational sensitivity study of spray dispersion and mixing on the fuel properties in a gas turbine combustor

    Energy Technology Data Exchange (ETDEWEB)

    Grosshans, Holger; Szász, Robert-Zoltán [Division of Fluid Mechanics, Lund University (Sweden); Cao, Le [Key Laboratory for Aerosol-Cloud-Precipitation of China Meteorological Administration, Nanjing University of Information Science and Technology, Nanjing (China); Fuchs, Laszlo, E-mail: holger.grosshans@uclouvain.be [Department of Mechanics, KTH, Stockholm (Sweden)

    2017-04-15

    A swirl stabilized gas turbine burner has been simulated in order to assess the effects of the fuel properties on spray dispersion and fuel–air mixing. The properties under consideration include fuel surface tension, viscosity and density. The turbulence of the gas phase is modeled applying the methodology of large eddy simulation whereas the dispersed liquid phase is described by Lagrangian particle tracking. The exchange of mass, momentum and energy between the two phases is accounted for by two-way coupling. Bag and stripping breakup regimes are considered for secondary droplet breakup, using the Reitz–Diwakar and the Taylor analogy breakup models. Moreover, a model for droplet evaporation is included. The results reveal a high sensitivity of the spray structure to variations of all investigated parameters. In particular, a decrease in the surface tension or the fuel viscosity, or an increase in the fuel density, lead to less stable liquid structures. As a consequence, smaller droplets are generated and the overall spray surface area increases, leading to faster evaporation and mixing. Furthermore, with the trajectories of the small droplets being strongly influenced by aerodynamic forces (and less by their own inertia), the spray is more affected by the turbulent structures of the gaseous phase and the spray dispersion is enhanced. (paper)

  17. 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.

  18. 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.

  19. Use of a horizontal air-dispersion system to enhance biodegradation of diesel fuel contaminated soils

    International Nuclear Information System (INIS)

    Baker, J.N.; Nickerson, D.A.; Guest, P.R.; Portele, T.E.

    1993-01-01

    A horizontal air-dispersion system was designed and installed to enhance the natural biodegradation of residual diesel fuel contaminated soils at an underground storage tank (UST) facility in Seattle, Washington. This system was designed to operate in conjunction with an existing free-product recovery system which exposes more heavily contaminated soils at the capillary fringe to injected air. Results of a pilot study conducted at the facility indicate that an initial biodegradation rate of 2,200 mg of total petroleum hydrocarbons (TPH) per kg of soil per year will be achieved, making in-situ biodegradation a feasible remedial alternative for contaminated site soils. Oxygen, carbon dioxide, and hydrocarbon vapor concentrations have been monitored since full-scale startup in September 1992, using a series of vapor monitoring points (VMPs) installed in the vicinity of the aerated beds and around the perimeter of the facility. Recent monitoring data indicate that the system is capable of aerating soils at distances greater than 80 feet from the aerated beds. Oxygen utilization and carbon dioxide production measured during post-startup respiration tests indicate microbial activity has increased as a result of seven months of full-scale system operation

  20. Technology for manufacturing dispersion nuclear fuel at Instituto de Pesquisas Energeticas e Nucleares IPEN/CNEN-SP, Brazil

    International Nuclear Information System (INIS)

    Durazzo, M.; Carvalho, E.F. Urano de; Saliba-Silva, A.M.; Souza, J.A.B.; Riella, H.G.

    2008-01-01

    IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 3.5 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U 3 O 8 -Al dispersion fuel plates with 2.3 g U/cm 3 . To support the reactor power increase, higher uranium density had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicide was the chosen option. This paper describes the results of this program and the current status of silicide fuel fabrication and qualification. (author)

  1. 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)

  2. 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.

  3. High density data recording for SSCL linac

    International Nuclear Information System (INIS)

    VanDeusen, A.L.; Crist, C.

    1993-01-01

    The Superconducting Super Collider Laboratory and AlliedSignal Aerospace have collaboratively developed a high density data monitoring system for beam diagnostic activities. The 128 channel data system is based on a custom multi-channel high speed digitizer card for the VXI bus. The card is referred to as a Modular Input VXI (MIX) digitizer. Multiple MIX cards are used in the complete system to achieve the necessary high channel density requirements. Each MIX digitizer card also contains programmable signal conditioning, and enough local memory to complete an entire beam scan without assistance from the host processor

  4. 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)

  5. Inelastic neutron scattering from high-density fcc 4He

    International Nuclear Information System (INIS)

    Thomlinson, W.; Eckert, J.; Shirane, G.

    1978-01-01

    The phonon dispersion relations in high-density crystals of fcc 4 He have been measured along high-symmetry directions by the neutron-inelastic-scattering technique. A recent study of the lattice dynamics of fcc 4 He by Eckert et al. has been extended to cover the fcc phase diagram at pressures below 5 kbar. Molar volumes of 9.03, 9.43, and 9.97 cm 3 /mole have been studied in the temperature range from near the melting curve to near the fcc-hcp transition line. The phonon dispersion relations are in good agreement with a first-order self-consistent phonon theory calculation by Goldman. The observed phonon-group line shapes at large energy and momentum transfers show evidence for multiphonon scattering in agreement with calculations by Glyde. Eckert et al. reported extremely large anharmonic isochoric temperature shifts of the phonon energies. The present work studied the shifts as a function of molar volume and temperature. Mode-Grueneisen-parameter dispersion curves have been measured using the present data and earlier measurements at lower density in the fcc phase by Traylor et al. Macroscopic Grueneisen parameters have been calculated from the phonon density of states obtained from the data

  6. Plasma Diagnostics in High Density Reactors

    International Nuclear Information System (INIS)

    Daltrini, A. M.; Moshkalyov, S.; Monteiro, M. J. R.; Machida, M.; Kostryukov, A.; Besseler, E.; Biasotto, C.; Diniz, J. A.

    2006-01-01

    Langmuir electric probes and optical emission spectroscopy diagnostics were developed for applications in high density plasmas. These diagnostics were employed in two plasma sources: an electron cyclotron resonance (ECR) plasma and an RF driven inductively coupled plasma (ICP) plasma. Langmuir probes were tested using a number of probing dimensions, probe tip materials, circuits for probe bias and filters. Then, the results were compared with the optical spectroscopy measurements. With these diagnostics, analyses of various plasma processes were performed in both reactors. For example, it has been shown that species like NH radicals generated in gas phase can have critical impact on films deposited by ECR plasmas. In the ICP source, plasmas in atomic and molecular gases were shown to have different spatial distributions, likely due to nonlocal electron heating. The low-to-high density transitions in the ICP plasma were also studied. The role of metastables is shown to be significant in Ar plasmas, in contrast to plasmas with additions of molecular gases

  7. Models for Experimental High Density Housing

    Science.gov (United States)

    Bradecki, Tomasz; Swoboda, Julia; Nowak, Katarzyna; Dziechciarz, Klaudia

    2017-10-01

    The article presents the effects of research on models of high density housing. The authors present urban projects for experimental high density housing estates. The design was based on research performed on 38 examples of similar housing in Poland that have been built after 2003. Some of the case studies show extreme density and that inspired the researchers to test individual virtual solutions that would answer the question: How far can we push the limits? The experimental housing projects show strengths and weaknesses of design driven only by such indexes as FAR (floor attenuation ratio - housing density) and DPH (dwellings per hectare). Although such projects are implemented, the authors believe that there are reasons for limits since high index values may be in contradiction to the optimum character of housing environment. Virtual models on virtual plots presented by the authors were oriented toward maximising the DPH index and DAI (dwellings area index) which is very often the main driver for developers. The authors also raise the question of sustainability of such solutions. The research was carried out in the URBAN model research group (Gliwice, Poland) that consists of academic researchers and architecture students. The models reflect architectural and urban regulations that are valid in Poland. Conclusions might be helpful for urban planners, urban designers, developers, architects and architecture students.

  8. 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

  9. Research on Diamantane and other High Density Hydrocarbon Fuels.

    Science.gov (United States)

    1979-10-01

    CH2 C 2 solvent ,= : 5.21 6.87 exo- OCPD . 18.61 Isomers of THTriCPD endo ,exo ,endo-THTri CPD tI 156 238 I.. 5,939R NO METHOD RT AREA AREA % 5.21 761680...0743122-1 HEATED OCPD 5300 E.JANOSKI 773417Cut Atmos. TFJUP.. 11T. G1S. VOL- MLS. SP.GR.GO/60 F WT.PCT VOL.PCT. h:)q n n-n n n 473 900 17.0 18.7

  10. Fuel powder production from ductile uranium alloys

    International Nuclear Information System (INIS)

    Clark, C.R.; Meyer, M.K.

    1998-01-01

    Metallic uranium alloys are candidate materials for use as the fuel phase in very-high-density LEU dispersion fuels. These ductile alloys cannot be converted to powder form by the processes routinely used for oxides or intermetallics. Three methods of powder production from uranium alloys have been investigated within the US-RERTR program. These processes are grinding, cryogenic milling, and hydride-dehydride. In addition, a gas atomization process was investigated using gold as a surrogate for uranium. (author)

  11. Nanotechnology for Synthetic High Density Lipoproteins

    Science.gov (United States)

    Luthi, Andrea J.; Patel, Pinal C.; Ko, Caroline H.; Mutharasan, R. Kannan; Mirkin, Chad A.; Thaxton, C. Shad

    2014-01-01

    Atherosclerosis is the disease mechanism responsible for coronary heart disease (CHD), the leading cause of death worldwide. One strategy to combat atherosclerosis is to increase the amount of circulating high density lipoproteins (HDL), which transport cholesterol from peripheral tissues to the liver for excretion. The process, known as reverse cholesterol transport, is thought to be one of the main reasons for the significant inverse correlation observed between HDL blood levels and the development of CHD. This article highlights the most common strategies for treating atherosclerosis using HDL. We further detail potential treatment opportunities that utilize nanotechnology to increase the amount of HDL in circulation. The synthesis of biomimetic HDL nanostructures that replicate the chemical and physical properties of natural HDL provides novel materials for investigating the structure-function relationships of HDL and for potential new therapeutics to combat CHD. PMID:21087901

  12. Ground state of high-density matter

    Science.gov (United States)

    Copeland, ED; Kolb, Edward W.; Lee, Kimyeong

    1988-01-01

    It is shown that if an upper bound to the false vacuum energy of the electroweak Higgs potential is satisfied, the true ground state of high-density matter is not nuclear matter, or even strange-quark matter, but rather a non-topological soliton where the electroweak symmetry is exact and the fermions are massless. This possibility is examined in the standard SU(3) sub C tensor product SU(2) sub L tensor product U(1) sub Y model. The bound to the false vacuum energy is satisfied only for a narrow range of the Higgs boson masses in the minimal electroweak model (within about 10 eV of its minimum allowed value of 6.6 GeV) and a somewhat wider range for electroweak models with a non-minimal Higgs sector.

  13. High Density Lipoprotein and it's Dysfunction.

    Science.gov (United States)

    Eren, Esin; Yilmaz, Necat; Aydin, Ozgur

    2012-01-01

    Plasma high-density lipoprotein cholesterol(HDL-C) levels do not predict functionality and composition of high-density lipoprotein(HDL). Traditionally, keeping levels of low-density lipoprotein cholesterol(LDL-C) down and HDL-C up have been the goal of patients to prevent atherosclerosis that can lead to coronary vascular disease(CVD). People think about the HDL present in their cholesterol test, but not about its functional capability. Up to 65% of cardiovascular death cannot be prevented by putative LDL-C lowering agents. It well explains the strong interest in HDL increasing strategies. However, recent studies have questioned the good in using drugs to increase level of HDL. While raising HDL is a theoretically attractive target, the optimal approach remains uncertain. The attention has turned to the quality, rather than the quantity, of HDL-C. An alternative to elevations in HDL involves strategies to enhance HDL functionality. The situation poses an opportunity for clinical chemists to take the lead in the development and validation of such biomarkers. The best known function of HDL is the capacity to promote cellular cholesterol efflux from peripheral cells and deliver cholesterol to the liver for excretion, thereby playing a key role in reverse cholesterol transport (RCT). The functions of HDL that have recently attracted attention include anti-inflammatory and anti-oxidant activities. High antioxidant and anti-inflammatory activities of HDL are associated with protection from CVD.This review addresses the current state of knowledge regarding assays of HDL functions and their relationship to CVD. HDL as a therapeutic target is the new frontier with huge potential for positive public health implications.

  14. 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.

  15. Modélisation de la combustion de fuels lourds prenant en compte la dispersion des asphaltènes Modeling Heavy Fuel-Oil Combustion (While Considering Or Including Asphaltene Dispersion

    Directory of Open Access Journals (Sweden)

    Audibert F.

    2006-11-01

    Full Text Available Divers modèles, ayant pour but de prédire le taux d'imbrûlés solides lors de la combustion du fuel lourd, ont été mis au point dans le passé. Les paramètres entrant en ligne de compte sont le plus souvent les teneurs en résidus lourds hydrocarbonés (asphaltènes précipités au pentane ou à l'heptane et carbone Conradson et en métaux : c'est le cas des modèles Exxon et Shell développés respectivement en 1979 et 1981. D'autres modèles tiennent compte, en plus de la composition du fuel, de son mode d'atomisation, de son mode de diffusion dans le foyer et de la cinétique de combustion : on peut citer les travaux du Laboratoire Energie du MIT publiés en 1986. Néanmoins, ces facteurs ne sont pas les seuls à intervenir : l'expérience a montré que l'état de dispersion des asphaltènes peut jouer également un grand rôle, notamment dans le cas d'installations de combustion à injection mécanique, pour lesquelles la dispersion des gouttelettes n'est pas aussi fine que pour des installations munies d'une injection assistée par la vapeur. Cette influence de la dispersion des asphaltènes sur la combustion a été mise en évidence dans le passé par l'utilisation d'additifs dispersants et également par la combustion de fuels lourds constitués par dilution d'asphaltes précipités au pentane avec un gas-oil de cracking catalytique de raffinerie (LCO. Ce sont ces fuels que l'on a considérés dans la présente étude. L'effet de ce facteur dispersion n'a pas été quantifié jusqu'alors, la difficulté étant de définir une grandeur mesurable représentant la répartition des agglomérats d'asphaltènes. Dans cette étude, on a essayé en un premier temps de faire une approche fractale de la répartition des asphaltènes à partir de clichés (préparés par la société Total, cette méthode ayant déjà été utilisée avec succès pour décrire des structures d'aspects comparables. Malheureusement, on s'est heurté à des

  16. 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.

  17. 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.

  18. High density thermite mixture for shaped charge ordnance disposal

    Directory of Open Access Journals (Sweden)

    Tamer Elshenawy

    2017-10-01

    Full Text Available The effect of thermite mixture based on aluminum and ferric oxides for ammunition neutralization has been studied and tested. Thermochemical calculations have been carried out for different percentage of Al using Chemical Equilibrium Code to expect the highest performance thermite mixture used for shaped charge ordnance disposal. Densities and enthalpy of different formulations have been calculated and demonstrated. The optimized thermite formulation has been prepared experimentally using cold iso-static pressing technique, which exhibited relatively high density and high burning rate thermite mixture. The produced green product compacted powder mixture was tested against small caliber shaped charge bomblet for neutralization. Theoretical and experimental results showed that the prepared thermite mixture containing 33% of aluminum as a fuel with ferric oxide can be successfully used for shaped charge ordnance disposal.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

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

    Science.gov (United States)

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

    2015-03-01

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

  1. Characterization of dispersed type fuel miniplates based in alloy UMo by evaluation of changes volumetrics and thermal conductivity

    International Nuclear Information System (INIS)

    Salinas Valero, Pablo Ignacio

    2016-01-01

    The development of new technologies in the nuclear area is extremely important to achieve greater efficiency and security in the production of electrical energy in the case of power reactors and for the production of radioisotopes and neutrons in research reactors. Throughout history, uranium-based nuclear fuels evolved in parallel with the requirements of nuclear reactors, this emphasis was increased when the RERTR program was created, which restricts the use of fuels with a maximum enrichment of 20% of the isotope U 235 (fissile isotope), which makes it necessary to increase the mass of uranium to compensate the amount of fissile material to maintain a neutron flux necessary for the reactors to operate with the same power. The search for new nuclear fuels has reached the UMo alloy with which densities of 18 gU/cm 3 are achieved in type fuels and 8 gU/cm 3 in dispersed type fuels, properties under irradiation due to their cubic crystalline structure. This type of fuel, when used dispersed in an aluminum matrix, becomes thermodynamically unstable by increasing the fission temperature of the U 235 isotope, due to this, compounds of lower density are formed, which causes an increase in volume (swelling). ). This swelling is studied throughout the present work, to relate the changes of UMo-Al / 4% volume of thermally induced miniecography in thermal treatments, with the purpose of evaluating changes in the thermal conductivity of the material. In this study it was detected that the swelling in miniplates is related in some way to the reduction of thermal conductivity, it was also recorded that the volume of change is irregular increasing and decreasing its volume according to the hours of induced swelling. The purpose of this work is to contribute to the development of dispersed fuels based on the UMo alloy in order to control the variables and reduce the probability of faults and possible accidents, such as fission products, or an increase in temperature in the core

  2. 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.

  3. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-12-15

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  4. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    International Nuclear Information System (INIS)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-01-01

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60% 235 U; the mini-rods were irradiated to an average burnup of ∼ 85% 235 U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%

  5. High density implosion experiments at Nova

    International Nuclear Information System (INIS)

    Cable, M.D.; Hatchett, S.P.; Nelson, M.B.; Lerche, R.A.; Murphy, T.J.; Ress, D.B.

    1994-01-01

    Deuterium filled glass microballoons are used as indirectly driven targets for implosion experiments at the Nova Laser Fusion Facility. High levels of laser precision were required to achieve fuel densities and convergences to an ignition scale hot spot. (AIP) copyright 1994 American Institute of Physics

  6. High-density lipoprotein cholesterol: How High

    Directory of Open Access Journals (Sweden)

    G Rajagopal

    2012-01-01

    Full Text Available The high-density lipoprotein cholesterol (HDL-C is considered anti-atherogenic good cholesterol. It is involved in reverse transport of lipids. Epidemiological studies have found inverse relationship of HDL-C and coronary heart disease (CHD risk. When grouped according to HDL-C, subjects having HDL-C more than 60 mg/dL had lesser risk of CHD than those having HDL-C of 40-60 mg/dL, who in turn had lesser risk than those who had HDL-C less than 40 mg/dL. No upper limit for beneficial effect of HDL-C on CHD risk has been identified. The goals of treating patients with low HDL-C have not been firmly established. Though many drugs are known to improve HDL-C concentration, statins are proven to improve CHD risk and mortality. Cholesteryl ester transfer protein (CETP is involved in metabolism of HDL-C and its inhibitors are actively being screened for clinical utility. However, final answer is still awaited on CETP-inhibitors.

  7. High-density hybrid interconnect methodologies

    International Nuclear Information System (INIS)

    John, J.; Zimmermann, L.; Moor, P.De; Hoof, C.Van

    2003-01-01

    Full text: The presentation gives an overview of the state-of-the-art of hybrid integration and in particular the IMEC technological approaches that will be able to address future hybrid detector needs. The dense hybrid flip-chip integration of an array of detectors and its dedicated readout electronics can be achieved with a variety of solderbump techniques such as pure Indium or Indium alloys, Ph-In, Ni/PbSn, but also conducting polymers... Particularly for cooled applications or ultra-high density applications, Indium solderbump technology (electroplated or evaporated) is the method of choice. The state-of-the-art of solderbump technologies that are to a high degree independent of the underlying detector material will be presented and examples of interconnect densities between 5x1E4cm-2 and 1x1E6 cm-2 will be demonstrated. For several classes of detectors, flip-chip integration is not allowed since the detectors have to be illuminated from the top. This applies to image sensors for EUV applications such as GaN/AlGaN based detectors and to MEMS-based sensors. In such cases, the only viable interconnection method has to be through the (thinned) detector wafer followed by a solderbump-based integration. The approaches for dense and ultra-dense through-the-wafer interconnect 'vias' will be presented and wafer thinning approaches will be shown

  8. HIGH DENSITY QCD WITH HEAVY-IONS

    CERN Multimedia

    The Addendum 1 to Volume 2 of the CMS Physics TDR has been published The Heavy-Ion analysis group completed the writing of a TDR summarizing the CMS plans in using heavy ion collisions to study high density QCD. The document was submitted to the LHCC in March and presented in the Open Session of the LHCC on May 9th. The study of heavy-ion physics at the LHC is promising to be very exciting. LHC will open a new energy frontier in ultra-relativistic heavy-ion physics. The collision energy of heavy nuclei at sNN = 5.5 TeV will be thirty times larger than what is presently available at RHIC. We will certainly probe quark and gluon matter at unprecedented values of energy density. The prime goal of this research programme is to study the fundamental theory of the strong interaction - Quantum Chromodynamics (QCD) - in extreme conditions of temperature, density and parton momentum fraction (low-x). Such studies, with impressive experimental and theoretical advances in recent years thanks to the wealth of high-qua...

  9. High-density oxidized porous silicon

    International Nuclear Information System (INIS)

    Gharbi, Ahmed; Souifi, Abdelkader; Remaki, Boudjemaa; Halimaoui, Aomar; Bensahel, Daniel

    2012-01-01

    We have studied oxidized porous silicon (OPS) properties using Fourier transform infraRed (FTIR) spectroscopy and capacitance–voltage C–V measurements. We report the first experimental determination of the optimum porosity allowing the elaboration of high-density OPS insulators. This is an important contribution to the research of thick integrated electrical insulators on porous silicon based on an optimized process ensuring dielectric quality (complete oxidation) and mechanical and chemical reliability (no residual pores or silicon crystallites). Through the measurement of the refractive indexes of the porous silicon (PS) layer before and after oxidation, one can determine the structural composition of the OPS material in silicon, air and silica. We have experimentally demonstrated that a porosity approaching 56% of the as-prepared PS layer is required to ensure a complete oxidation of PS without residual silicon crystallites and with minimum porosity. The effective dielectric constant values of OPS materials determined from capacitance–voltage C–V measurements are discussed and compared to FTIR results predictions. (paper)

  10. Effect of palladium dispersion on the capture of toxic components from fuel gas by palladium-alumina sorbents

    Energy Technology Data Exchange (ETDEWEB)

    John P. Baltrus; Evan J. Granite; Erik C. Rupp; Dennis C. Stanko; Bret Howard; Henry W. Pennline [US DOE National Energy Technology Laboratory, Pittsburgh, PA (United States)

    2011-05-15

    The dispersion and location of Pd in alumina-supported sorbents prepared by different methods was found to influence the performance of the sorbents in the removal of mercury, arsine, and hydrogen selenide from a simulated fuel gas. When Pd is well dispersed in the pores of the support, contact interaction with the support is maximized, Pd is less susceptible to poisoning by sulfur, and the sorbent has better long-term activity for adsorption of arsine and hydrogen selenide, but poorer adsorption capacity for Hg. As the contact interaction between Pd and the support is lessened the Pd becomes more susceptible to poisoning by sulfur, resulting in higher capacity for Hg, but poorer long-term performance for adsorption of arsenic and selenium. 18 refs., 3 figs., 4 tabs.

  11. 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)

  12. 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

  13. 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)

  14. Simulation of fuel dispersion in the MYRRHA-FASTEF primary coolant with CFD and SIMMER-IV

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, Sophia, E-mail: sophia.buckingham@vki.ac.be [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Planquart, Philippe [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Eboli, Marica [University of Pisa, Largo Lucio Lazzarino 2, 56122 Pisa (Italy); Moreau, Vincent [CRS4, Science and Technology Park Polaris – Piscina Manna, 09010 Pula (Italy); Van Tichelen, Katrien [SCK-CEN, Boeretang 200, 2400 Mol (Belgium)

    2015-12-15

    Highlights: • A comparison between CFD and system codes applied to long-term dispersion of fuel particles inside the MYRRHA reactor is proposed. • Important accumulations at the free-surface level are to be expected. • The risk of core blockage should not be neglected. • Numerical approach and modeling assumptions have a strong influence on the simulation results and accuracy. - Abstract: The objective of this work is to assess the behavior of fuel redistribution in heavy liquid metal nuclear systems under fuel pin failure conditions. Two different modeling approaches are considered using Computational Fluid Dynamics (CFD) codes and a system code, applied to the MYRRHA facility primary coolant loop version 1.4. Two different CFD models are constructed: the first is a single-phase steady model prepared in ANSYS Fluent, while the second is a two-phase model based on the volume of fluid (VOF) method in STARCCM+ to capture the upper free-surface dynamics. Both use a Lagrangian tracking approach with oneway coupling to follow the particles throughout the reactor. The system code SIMMER-IV is used for the third model, without neutronic coupling. Although limited regarding the fluid dynamic aspects compared to the CFD codes, comparisons of particle distributions highlight strong similarities despite quantitative discrepancies in the size of fuel accumulations. These disparities should be taken into account while performing the safety analysis of nuclear systems and developing strategies for accident mitigation.

  15. High density linear systems for fusion power

    International Nuclear Information System (INIS)

    Ellis, W.R.; Krakowski, R.A.

    1975-01-01

    The physics and technological limitations and uncertainties associated with the linear theta pinch are discussed in terms of a generalized energy balance, which has as its basis the ratio (Q/sub E/) of total electrical energy generated to net electrical energy consumed. Included in this total is the virtual energy of bred fissile fuel, if a hybrid blanket is used, as well as the actual of real energy deposited in the blanket by the fusion neutron. The advantages and disadvantages of the pulsed operation demanded by the linear theta pinch are also discussed

  16. Comparison of the socioeconomic impacts of international fuel service centers versus dispersed nuclear facilities

    International Nuclear Information System (INIS)

    Braid, R.B. Jr.

    1979-01-01

    The paper investigates a variety of community impacts including: public services, fiscal issues, economic matters, land and water use, political and social cohesion, and legal considerations. Comparisons of socioeconomic impacts of colocated versus dispersed sites are made on the basis of the size of the impacted communities, the size and type of nuclear facility, and the facility's construction time frame. The paper concludes that, under similar circumstances, most of the socioeconomic impacts of colocated nuclear facilities would be somewhat less than the sum of the impacts associated with equivalent dispersed sites. While empirical data is non-existent, the paper contends, however, that because the socioeconomic impacts of colocated facilities are so great and readily identifiable to a public unskilled in making comparisons with the dispersed alternative, the facilities will likely generate so much public opposition that IFSCs will probably prove infeasible

  17. 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)

  18. Properties of recycled high density polyethylene and coffee dregs composites

    Directory of Open Access Journals (Sweden)

    Sibele Piedade Cestari

    2013-01-01

    Full Text Available Composites of recycled high density polyethylene (HDPE-R and coffee dregs (COFD were elaborated. The blends were made at the proportions of 100-0, 90-10, 80-20, 70-30, 60-40, 50-50 and 40-60% polymer-filler ratio. The materials were evaluated through scanning electron microscopy (SEM, differential scanning calorimetry (DSC, thermogravimetry/derivative thermogravimetry (TGA, and compressive resistance test. The compounding was done using a two-stage co-kneader system extruder, and then cylindrical specimens were injection molded. All composites had a fine dispersion of the COFD into the polymeric matrix. The composites degraded in two steps. The first one was in a temperature lower than the neat HDPE, but higher than the average processing temperature of the polymer. The melting temperature and the degree of crystallinity of the composites resulted similar to the neat HDPE ones. The compressive moduli of the composites resulted similar to the neat polymer one. The results show that these composites have interesting properties as a building material.

  19. 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.

  20. 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.

  1. Method and device for fabricating dispersion fuel comprising fission product collection spaces

    Science.gov (United States)

    Shaber, Eric L; Fielding, Randall S

    2015-05-05

    A method of fabricating a nuclear fuel comprising a fissile material, one or more hollow microballoons, a phenolic resin, and metal matrix. The fissile material, phenolic resin and the one or more hollow microballoons are combined. The combined fissile material, phenolic resin and the hollow microballoons are heated sufficiently to form at least some fissile material carbides creating a nuclear fuel particle. The resulting nuclear fuel particle comprises one or more fission product collection spaces. In a preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by forming the fissile material into microspheres. The fissile material microspheres are then overcoated with the phenolic resin and microballoon. In another preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by overcoating the microballoon with the fissile material, and phenolic resin.

  2. 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)

  3. High-temperature deformation and processing maps of Zr-4 metal matrix with dispersed coated surrogate nuclear fuel particles

    Science.gov (United States)

    Chen, Jing; Liu, Huiqun; Zhang, Ruiqian; Li, Gang; Yi, Danqing; Lin, Gaoyong; Guo, Zhen; Liu, Shaoqiang

    2018-06-01

    High-temperature compression deformation of a Zr-4 metal matrix with dispersed coated surrogate nuclear fuel particles was investigated at 750 °C-950 °C with a strain rate of 0.01-1.0 s-1 and height reduction of 20%. Scanning electron microscopy was utilized to investigate the influence of the deformation conditions on the microstructure of the composite and damage to the coated surrogate fuel particles. The results indicated that the flow stress of the composite increased with increasing strain rate and decreasing temperature. The true stress-strain curves showed obvious serrated oscillation characteristics. There were stable deformation ranges at the initial deformation stage with low true strain at strain rate 0.01 s-1 for all measured temperatures. Additionally, the coating on the surface of the surrogate nuclear fuel particles was damaged when the Zr-4 matrix was deformed at conditions of high strain rate and low temperature. The deformation stability was obtained from the processing maps and microstructural characterization. The high-temperature deformation activation energy was 354.22, 407.68, and 433.81 kJ/mol at true strains of 0.02, 0.08, and 0.15, respectively. The optimum deformation parameters for the composite were 900-950 °C and 0.01 s-1. These results are expected to provide guidance for subsequent determination of possible hot working processes for this composite.

  4. 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)

  5. 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, J.A.B.; Durazzo, M., E-mail: jasouza@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2010-07-01

    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 gU/cm{sup 3} 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 gU/cm{sup 3} 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)

  6. Effects of temperature and wave conditions on chemical dispersion efficacy of heavy fuel oil in an experimental flow-through wave tank.

    Science.gov (United States)

    Li, Zhengkai; Lee, Kenneth; King, Thomas; Boufadel, Michel C; Venosa, Albert D

    2010-09-01

    The effectiveness of chemical dispersants (Corexit 9500 and SPC 1000) on heavy fuel oil (IFO180 as test oil) has been evaluated under different wave conditions in a flow-through wave tank. The dispersant effectiveness was determined by measuring oil concentrations and droplet size distributions. An analysis of covariance (ANCOVA) model indicated that wave type and temperature significantly (p or = 400 microm). Copyright 2010 Elsevier Ltd. All rights reserved.

  7. Highly water-dispersible, mixed ionic-electronic conducting, polymer acid-doped polyanilines as ionomers for direct methanol fuel cells.

    Science.gov (United States)

    Murthy, Arun; Manthiram, Arumugam

    2011-06-28

    Highly water-dispersible polymer acid-doped polyanilines have been synthesized and evaluated as an alternative for expensive Nafion ionomers in the anode of direct methanol fuel cells (DMFC). These polymers as ionomers lead to higher performance in single cell DMFC compared to Nafion ionomers due to mixed ionic-electronic conduction, water dispersibility, and co-catalytic activity. This journal is © The Royal Society of Chemistry 2011

  8. Computational simulation of the microstructure of irradiation damaged regions for the plate type fuel of UO2 microspheres dispersed in stainless steel matrix

    International Nuclear Information System (INIS)

    Reis, S.C. dos; Lage, A.F.; Braga, D.; Ferraz, W.B.

    2006-01-01

    Plate type fuel elements have high efficiency of thermal transference what benefits the heat flux with high rates of power output. In reactor cores, fuel elements, in general, are subject to a high neutrons flux, high working temperatures, severe corrosion conditions, direct interference of fission products that result from nuclear reactions and radiation interaction-matter. For plate type fuels composed of ceramic particles dispersed in metallic matrix, one can observe the damage regions that arise due to the interaction fission products in the metallic matrix. Aiming at evaluating the extension of the damage regions in function of the particles and its diameters, in this paper, computational geometric simulations structure of plate type fuel cores, composed of UO 2 microspheres dispersed in stainless steel in several fractions of volume and diameters were carried out. The results of the simulations were exported to AutoCAD R where it was possible its visualization and analysis. (author)

  9. 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

  10. 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

  11. In situ ceramic layer growth on coated fuel particles dispersed in a zirconium metal matrix

    Science.gov (United States)

    Terrani, K. A.; Silva, C. M.; Kiggans, J. O.; Cai, Z.; Shin, D.; Snead, L. L.

    2013-06-01

    The extent and nature of the chemical interaction between the outermost coating layer of coated fuel particles embedded in zirconium metal during fabrication of metal matrix microencapsulated fuels were examined. Various particles with outermost coating layers of pyrocarbon, SiC, and ZrC have been investigated in this study. ZrC-Zr interaction was the least substantial, while the PyC-Zr reaction can be exploited to produce a ZrC layer at the interface in an in situ manner. The thickness of the ZrC layer in the latter case can be controlled by adjusting the time and temperature during processing. The kinetics of ZrC layer growth is significantly faster from what is predicted using literature carbon diffusivity data in ZrC. SiC-Zr interaction is more complex and results in formation of various chemical phases in a layered aggregate morphology at the interface.

  12. 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.

  13. Combustion Enhancement of Liquid Fuels via Nanoparticle Additions: Screening, Dispersion, and Characterization

    Science.gov (United States)

    2015-06-04

    burning rates of nitromethane (a monopropellant) solutions with both n-Al and fumed silica 8 and functionalized graphene 9 in a liquid propellant...These include but are not limited to aluminum, boron, boron carbide (B4C), carbon ( graphene ), titanium, and tungsten nano-sized particles. When... paint at one corner. To dilute the particle number density to an appropriate level for imaging, the particle/fuel mixes were agitated using a

  14. The obtainment of highly concentrated uranium pellets for plate type (MTR) fuel by dispersion of uranium aluminides in aluminium

    International Nuclear Information System (INIS)

    Morando, R.A.; Raffaeli, H.A.; Balzaretti, D.E.

    1980-01-01

    The use of the intermetallic UAl 3 for manufacturing plate type MTR fuel with 20% U 235 enriched uranium and a density of about 20 kg/m 3 is analyzed. The technique used is the dispersion of UAl 3 particles in aluminium powder. The obtainment of the UAl 3 intermetallic was performed by fusion in an induction furnace in an atmosphere of argon at a pressure of 0.7 BAR (400 mm) using an alumina melting pot. To make the aluminide powder and attain the wished granulometry a cutting and a rotating crusher were used. Aluminide powders of different granulometries and different pressures of compactation were analyzed. In each case the densities were measured. The compacts were colaminated with the 'Picture Frame' technique at temperatures of 490 and 0 deg C with excellent results from the manufacturing view point. (M.E.L.) [es

  15. Highly dispersed Pt-Ni nanoparticles on nitrogen-doped carbon nanotubes for application in direct methanol fuel cells.

    Science.gov (United States)

    Jiang, Shujuan; Ma, Yanwen; Tao, Haisheng; Jian, Guoqiang; Wang, Xizhang; Fan, Yining; Zhu, Jianmin; Hu, Zheng

    2010-06-01

    Binary Pt-Ni alloyed nanoparticles supported on nitrogen-doped carbon nanotubes (NCNTs) have been facilely constructed without pre-modification by making use of the active sites in NCNTs due to the N-participation. So-obtained binary Pt-Ni alloyed nanoparticles have been highly dispersed on the outer surface of the support with the size of about 3-4 nm. The electrochemical properties of the catalysts for methanol oxidation have been systematically evaluated. Binary Pt-Ni alloyed composites with molar ratio (Pt:Ni) of 3:2 and 3:1 present enhanced electrocatalytic activities and improved tolerance to CO poisoning as well as the similar stability, in comparison with the commercial Pt/C catalyst and the monometallic Pt/NCNTs catalysts. These results imply that so-constructed nanocomposite catalysts have the potential for applications in direct methanol fuel cells.

  16. Steady- and transient-state analysis of fully ceramic microencapsulated fuel with randomly dispersed tristructural isotropic particles via two-temperature homogenized model-I: Theory and method

    International Nuclear Information System (INIS)

    Lee, Yoon Hee; Cho, Bum Hee; Cho, Nam Zin

    2016-01-01

    As a type of accident-tolerant fuel, fully ceramic microencapsulated (FCM) fuel was proposed after the Fukushima accident in Japan. The FCM fuel consists of tristructural isotropic particles randomly dispersed in a silicon carbide (SiC) matrix. For a fuel element with such high heterogeneity, we have proposed a two-temperature homogenized model using the particle transport Monte Carlo method for the heat conduction problem. This model distinguishes between fuel-kernel and SiC matrix temperatures. Moreover, the obtained temperature profiles are more realistic than those of other models. In Part I of the paper, homogenized parameters for the FCM fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure are obtained by (1) matching steady-state analytic solutions of the model with the results of particle transport Monte Carlo method for heat conduction problems, and (2) preserving total enthalpies in fuel kernels and SiC matrix. The homogenized parameters have two desirable properties: (1) they are insensitive to boundary conditions such as coolant bulk temperatures and thickness of cladding, and (2) they are independent of operating power density. By performing the Monte Carlo calculations with the temperature-dependent thermal properties of the constituent materials of the FCM fuel, temperature-dependent homogenized parameters are obtained

  17. Improved GAMMA 10 tandem mirror confinement in high density plasma

    International Nuclear Information System (INIS)

    Yatsu, K.; Cho, T.; Higaki, H.; Hirata, M.; Hojo, H.; Ichimura, M.; Ishii, K.; Ishimoto, Y.; Itakura, A.; Katanuma, I.; Kohagura, J.; Minami, R.; Nakashima, Y.; Numakura, T.; Saito, T.; Saosaki, S.; Takemura, Y.; Tatematsu, Y.; Yoshida, M.; Yoshikawa, M.

    2003-01-01

    GAMMA 10 experiments have advanced in high density experiments after the last IAEA fusion energy conference in 2000 where we reported the production of the high density plasma through use of ion cyclotron range of frequency heating at a high harmonic frequency and neutral beam injection in the anchor cells. However, the diamagnetic signal of the plasma decreased when electron cyclotron resonance heating was applied for the potential formation. Recently a high density plasma has been obtained without degradation of the diamagnetic signal and with much improved reproducibility than before. The high density plasma was attained through adjustment of the spacing of the conducting plates installed in the anchor transition regions. The potential confinement of the plasma has been extensively studied. Dependences of the ion confinement time, ion-energy confinement time and plasma confining potential on plasma density were obtained for the first time in the high density region up to a density of 4x10 18 m -3 . (author)

  18. Report on achievements in fiscal 1998. Development of technologies to put photovoltaic power generation systems into practical use - Demonstrative research on photovoltaic power generation system (Study on grid interconnection technique for dispersed photovoltaic systems under high-density connection); 1998 nendo taiyoko hatsuden system jitsuyoka gijutsu kaihatsu seika hokokusho. Taiyoko hatsuden system no jissho kenkyu (komitsudo renkei gijutsu no kenkyu)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-01

    Interconnecting photovoltaic systems with power transmission systems under high density affects power quality, protection, maintenance and stability of the transmission lines. As measures to deal with this issue, investigations are being made on (1) elucidation of effects imposed on transmission lines, (2) establishment of countermeasure technologies, and (3) technological options leading to higher value addition. In Item (1), with an objective to identify the current status, evaluations were given on prevention of independent operation of commercially available inverters, and on their stabilizing performance against system fluctuation. The evaluations were performed by conducting a test for multiple unit operation in parallel and a single unit performance test. The test result indicated that, while the prevention performance can be satisfied, maloperation has occurred frequently due to the system fluctuation, and that voltage rise due to the inverter was suppressed effectively by using the simultaneous control of active and reactive powers. In Item (2), a demonstration test was launched on an inverter incorporating a new prevention device. The effective means to suppress voltage rise in the high-voltage power transmission lines is the discrete voltage suppression by controlling reactive power. In addition, a proposal was made on a new voltage and phase detection method that can be used at short circuit of the high-voltage transmission lines. In Item (3), having a photovoltaic system contain a small size batteries was found effective in suppressing the power generation output variation, and in smoothing the loads. (NEDO)

  19. Postirradiation Examination Of U3O8-AL Plate Type Dispersion Fuel Element

    International Nuclear Information System (INIS)

    Nasution-Hasbullah; Sugondo; Amin, D.L.; Siti-Amini

    1996-01-01

    Postirradiation examination of plate type spent fuel element RIE-01 has been carried out in order to observer its physical changes and performance under irradiation in the reactor. The irradiation has been time more than two years with a declared burnup of 51.04 %. The examination included visual and dimensional measurement, measurement of burn-up distribution, wipe test and metallographic analysis. The results showed that all fuel plates retained their integrity. The colour changes were occurred on most of the plates significant suggesting that it was generated from the oxide layer formation. From gamma-scanning examination it could be deducted that the highest burn-up distribution of the plate was at position of 30 cm from the bottom. A more homogeneous distribution was found in the middle plate of the bundle. The increased plate thickness, as revealed by dimensional measurements as in agreement with the burn-up distribution pattern. Despite the changes observed in could be concluded that all changes occurred were still within the allowable limits and therefore it can recommended that an increase of the burn-up level above 51,04 % is still quite possible

  20. Thermomechanical DART code improvements for LEU VHD dispersion and monolithic fuel element analysis

    International Nuclear Information System (INIS)

    Taboada, H.; Saliba, R.; Moscarda, M.V.; Rest, J.

    2005-01-01

    A collaboration agreement between ANL/US DOE and CNEA Argentina in the area of Low Enriched Uranium Advanced Fuels has been in place since October 16, 1997 under the Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy. An annex concerning DART code optimization has been operative since February 8, 1999. Previously, as a part of this annex a visual FASTDART version and also a DART THERMAL version were presented during RERTR 2000, 2002 and RERTR 2003 Meetings. During this past year the following activities were completed: Optimization of DART TM code Al diffusion parameters by testing predictions against reliable data from RERTR experiments. Improvements on the 3-D thermo-mechanical version of the code for modeling the irradiation behavior of LEU U-Mo monolithic fuel. Concerning the first point, by means of an optimization of parameters of the Al diffusion through the interaction product theoretical expression, a reasonable agreement between DART temperature calculations with reliable RERTR PIE data was reached. The 3-D thermomechanical code complex is based upon a finite element thermal-elastic code named TERMELAS, and irradiation behavior provided by the DART code. An adequate and progressive process of coupling calculations of both codes at each time step is currently developed. Compatible thermal calculation between both codes was reached. This is the first stage to benchmark and validate against RERTR PIE data the coupling process. (author)

  1. Development of a Fast Breeder Reactor Fuel Bundle Deformation Analysis Code - BAMBOO: Development of a Pin Dispersion Model and Verification by the Out-of-Pile Compression Test

    International Nuclear Information System (INIS)

    Uwaba, Tomoyuki; Ito, Masahiro; Ukai, Shigeharu

    2004-01-01

    To analyze the wire-wrapped fast breeder reactor fuel pin bundle deformation under bundle/duct interaction conditions, the Japan Nuclear Cycle Development Institute has developed the BAMBOO computer code. This code uses the three-dimensional beam element to calculate fuel pin bowing and cladding oval distortion as the primary deformation mechanisms in a fuel pin bundle. The pin dispersion, which is disarrangement of pins in a bundle and would occur during irradiation, was modeled in this code to evaluate its effect on bundle deformation. By applying the contact analysis method commonly used in the finite element method, this model considers the contact conditions at various axial positions as well as the nodal points and can analyze the irregular arrangement of fuel pins with the deviation of the wire configuration.The dispersion model was introduced in the BAMBOO code and verified by using the results of the out-of-pile compression test of the bundle, where the dispersion was caused by the deviation of the wire position. And the effect of the dispersion on the bundle deformation was evaluated based on the analysis results of the code

  2. Renewable carbohydrates are a potential high-density hydrogen carrier

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Y.-H. Percival [Biological Systems Engineering Department, 210-A Seitz Hall, Virginia Polytechnic Institute and State University, Blacksburg, VA 24061 (United States); Institute for Critical Technology and Applied Sciences (ICTAS), Virginia Polytechnic Institute and State University, Blacksburg, VA 24061 (United States); DOE BioEnergy Science Center (BESC), Oak Ridge, TN 37831 (United States)

    2010-10-15

    The possibility of using renewable biomass carbohydrates as a potential high-density hydrogen carrier is discussed here. Gravimetric density of polysaccharides is 14.8 H{sub 2} mass% where water can be recycled from PEM fuel cells or 8.33% H{sub 2} mass% without water recycling; volumetric densities of polysaccharides are >100 kg of H{sup 2}/m{sup 3}. Renewable carbohydrates (e.g., cellulosic materials and starch) are less expensive based on GJ than are other hydrogen carriers, such as hydrocarbons, biodiesel, methanol, ethanol, and ammonia. Biotransformation of carbohydrates to hydrogen by cell-free synthetic (enzymatic) pathway biotransformation (SyPaB) has numerous advantages, such as high product yield (12 H{sub 2}/glucose unit), 100% selectivity, high energy conversion efficiency (122%, based on combustion energy), high-purity hydrogen generated, mild reaction conditions, low-cost of bioreactor, few safety concerns, and nearly no toxicity hazards. Although SyPaB may suffer from current low reaction rates, numerous approaches for accelerating hydrogen production rates are proposed and discussed. Potential applications of carbohydrate-based hydrogen/electricity generation would include hydrogen bioreactors, home-size electricity generators, sugar batteries for portable electronics, sugar-powered passenger vehicles, and so on. Developments in thermostable enzymes as standardized building blocks for cell-free SyPaB projects, use of stable and low-cost biomimetic NAD cofactors, and accelerating reaction rates are among the top research and development priorities. International collaborations are urgently needed to solve the above obstacles within a short time. (author)

  3. Influence of the silicon content on the core corrosion properties of dispersion type fuel plates; Influencia del Contenido en silicio sobre la corrosion acuosa de los nucleos de placas combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Calvo, C; Saenz de Tejada, L M; Diaz Diaz, J

    1969-07-01

    A new process to produce aluminium base dispersion type fuel plates has been developed at the Spanish JEN (Junta de Energia Nuclear). The dispersed fuel material is obtained by an aluminothermic process to render a stoichiometric cermet of UAI{sub 3} and AI{sub 2}O{sub 3} according to the reaction. (Author)

  4. Phenomenology of high density disruptions in the TFTR tokamak

    International Nuclear Information System (INIS)

    Fredrickson, E.D.; McGuire, K.M.; Bell, M.G.

    1993-01-01

    Studies of high density disruptions on TFTR, including a comparison of minor and major disruptions at high density, provide important new information regarding the nature of the disruption mechanism. Further, for the first time, an (m,n)=(1,1) 'cold bubble' precursor to high density disruptions has been experimentally observed in the electron temperature profile. The precursor to major disruptions resembles the 'vacuum bubble' model of disruptions first proposed by B.B. Kadomtsev and O.P. Pogutse (Sov. Phys. - JETP 38 (1974) 283). (author). Letter-to-the-editor. 25 refs, 3 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  6. Impingement heat flux by dispersed molten metal fuel on a horizontal stainless steel structure

    International Nuclear Information System (INIS)

    Gabor, J.D.; Purviance, R.T.; Aeschlimann, R.W.; Spencer, B.W.

    1989-01-01

    Although the Integral Fast Reactor (IFR) possesses inherent safety features, an assessment of the consequences of melting of the metal fuel is necessary for risk analysis. As part of this effort an experimental study was conducted to determine the depths of sodium at 600 C required for pour streams of various molten uranium alloys (U, U-5 wt % Zr, U-10 wt % Zr, and U-10 wt % Fe) to break up and solidify. The quenched particulate material, which was in the shape of filaments and sheets, formed coolable beds because of the high void-age (∼0.9) and large particle size (∼10 mm). In a test with a 0.15-m sodium depth, the fragments from a pure uranium pour stream did not completely solidify but formed an agglomerated mass which did not fuse to the base plate. However, the agglomerated fragments of U-10 wt % Fe eutectic fused to the stainless steel base plate. An analysis of the temperature response of a 25-mm thick base plate was made by volume averaging the properties of the sodium and metal phases and assuming two semi-infinite solids coming into contact. Good agreement was obtained with the data during the initial 5 to 10 s of the contact period. 16 refs., 5 figs., 2 tabs

  7. The Influence of Decreased Levels of High Density Lipoprotein ...

    African Journals Online (AJOL)

    very low density lipoprotein cholesterol, and triglyceride were assayed. ... Abiodun and Gwarzo: Association of high density lipoprotein cholesterol with haemolysis in sickle cell disease ... analyses were carried out to determine the correlation.

  8. The high density effects in the Drell-Yan process

    International Nuclear Information System (INIS)

    Betemps, M.A.; Gay Ducati, M.B.; Ayala Filho, A.L.

    2003-01-01

    The high density effects in the Drell-Yan process (q q-bar → γ * →l + l - ) are investigated for pA collisions at RHIC and LHC energies. In particular, we use a set of nuclear parton distributions that describes the present nuclear eA and pA data in the DGLAP approach including the high density effects introduced in the perturbative Glauber-Mueller approach. (author)

  9. Preparation of highly dispersed PEM fuel cell catalysts using electroless deposition methods

    Energy Technology Data Exchange (ETDEWEB)

    Beard, K.D.; Schaal, M.T.; Van Zee, J.W.; Monnier, J.R. [Department of Chemical Engineering, University of South Carolina, Swearingen Engineering Center, 301 South Main Street, Columbia, SC 29208 (United States)

    2007-03-30

    A methodology for the electroless deposition (ED) of PtCl{sub 6}{sup 2-} using dimethylamine borane (DMAB) on a Rh-seeded carbon support has been developed for electrochemical and fuel cell applications. This procedure required seeding the carbon with a Rh-precursor catalyst via wet impregnation prior to the exposure of an aqueous ED bath containing PtCl{sub 6}{sup 2-}, DMAB, and sodium citrate (complexing/stabilizing agent). Kinetic parameters that affect the extent and rate of PtCl{sub 6}{sup 2-} deposition include concentrations of PtCl{sub 6}{sup 2-}, DMAB, and sodium citrate as well as pH and concentrations of Rh seed sites. A linear relationship between rate and extent of PtCl{sub 6}{sup 2-} deposition and DMAB and Rh concentrations was found while the citrate concentration had little effect on rate and a modest effect on extent. Lastly, extent of PtCl{sub 6}{sup 2-} deposition showed a maximum with respect to pH. Characterization of the Rh-seeded, carbon support by transmission electron microscopy (TEM) shows that the Rh particle diameters remain constant at 33-43 Aa as the Rh weight loading increases from 0.4% to 2.2% to 4.4%. Further, after deposition of similar loadings of Pt, TEM analysis shows Pt particle diameters decrease with increasing Rh loading, since equal amounts of Pt were deposited on greater numbers of Rh seed particles. This pattern suggests a shell-core geometry, where Pt is deposited more or less uniformly around a Rh core. (author)

  10. Highly dispersed TaOx nanoparticles prepared by electrodeposition as oxygen reduction electrocatalysts for polymer electrolyte fuel cells

    KAUST Repository

    Seo, Jeongsuk

    2013-06-06

    Based on the chemical stability of group IV and V elements in acidic solutions, TaOx nanoparticles prepared by electrodeposition in an ethanol-based Ta plating bath at room temperature were investigated as novel nonplatinum electrocatalysts for the oxygen reduction reaction (ORR) in polymer electrolyte fuel cells (PEFCs). Electrodeposition conditions of Ta complexes and subsequent various heat treatments for the deposited TaOx were examined for the best performance of the ORR. TaOx particles on carbon black (CB), electrodeposited at a constant potential of -0.5 V Ag/AgCl for 10 s and then heat-treated by pure H2 flow at 523 K for 1 h, showed excellent catalytic activity with an onset potential of 0.93 VRHE (for 2 μA cm-2) for the ORR. Surface characterizations of the catalysts were performed by scanning transmission electron microscopy (STEM), transmission electron microscopy (TEM), and energy dispersive X-ray spectroscopy (EDS). The loading amounts of the electrodeposited material on the CB were determined by inductively coupled plasma atomic emission spectroscopy (ICP-AES). All the physical results suggested that high dispersion of TaOx particles on the CB surface with 2-3 nm size was critical and key for high activity. The chemical identity and modified surface structure for the deposited TaOx catalysts before and after H 2 heat treatment were analyzed by X-ray photoelectron spectroscopy (XPS). The formation of more exposed active sites on the electrode surface and enhanced electroconductivity of the tantalum oxide promoted from the H 2 treatment greatly improved the ORR performance of the electrodeposited TaOx nanoparticles on CB. Finally, the highly retained ORR activity after an accelerated durability test in an acidic solution confirmed and proved the chemical stability of the oxide nanoparticles. The high utilization of the electrodeposited TaOx nanoparticles uniformly dispersed on CB for the ORR was comparable to that of commercial Pt/CB catalysts

  11. Fabrication of cermet fuel for fast reactor

    International Nuclear Information System (INIS)

    Mishra, Sudhir; Kumar, Arun; Kutty, T.R.G.; Kamath, H.S.

    2011-01-01

    Mixed oxide (MOX) (U,Pu)O 2 , and metallic (U,Pu ,Zr) fuels are considered promising fuels for the fast reactor. The fuel cycle of MOX is well established. The advantages of the oxide fuel are its easy fabricability, good performance in the reactor and a well established reprocessing technology. However the problems lie in low thermal conductivity , low density of the fuel leading to low breeding ratio and consequently longer doubling time. The metallic fuel has the advantages of high thermal conductivity, higher metal density and higher coefficient of linear expansion. The higher coefficient of linear expansion is good from the safety consideration (negative reactivity factor). Because of higher metal density it offers highest breeding ratio and shortest doubling time. Metallic fuel disadvantages comprise large swelling at high burnup, fuel cladding interaction and lower margin between operating and melting temperature. The optimal solution may lie in cermet fuel (U, PuO 2 ), where PuO 2 is dispersed in U metal matrix and combines the favorable features of both the fuel types. The advantages of this fuel include high thermal conductivity, larger margin between melting and operating temperature, ability to retain fission product etc. The matrix being of high density metal the advantage of high breeding ratio is also maintained. In this report some results of fabrication of cermet pellet comprising of UO 2 /PuO 2 dispersed in U metal powder through classical powder metallurgy route and characterization are presented. (author)

  12. High density high performance plasma with internal diffusion barrier in Large Helical Device

    International Nuclear Information System (INIS)

    Sakamoto, R.; Kobayashi, M.; Miyazawa, J.

    2008-10-01

    A attractive high density plasma operational regime, namely an internal diffusion barrier (IDB), has been discovered in the intrinsic helical divertor configuration on the Large Helical Device (LHD). The IDB which enables core plasma to access a high density/high pressure regime has been developed. It is revealed that the IDB is reproducibly formed by pellet fueling in the magnetic configurations shifted outward in major radius. Attainable central plasma density exceeds 1x10 21 m -3 . Central pressure reaches 1.5 times atmospheric pressure and the central β value becomes fairly high even at high magnetic field, i.e. β(0)=5.5% at B t =2.57 T. (author)

  13. In vitro experiments for the development of a high density (HD) barium sulfate contrast medium

    International Nuclear Information System (INIS)

    Klein, J.

    1986-01-01

    In vitro experiments with the high-density (HD) barium meal Falibaryt HD are described. Several charges of BaSO 4 were tested together with certain additives influencing dispersion, stability of the suspension, flowability, surface tension etc. Particle size spectra were measured by the manufacturer, VEB Fahlberg-List. With a simple PVC test plate containing several grooves simulating small details (areae gastricae) the diagnostic capabilities of the HD contrast medium were evaluated in an in vitro test. The developed barium meal Falibaryt HD is in its physical and chemical parameters comparable with Prontobario-HD, one of the best HD barium meals. (author)

  14. High-density-plasma diagnostics in magnetic-confinement fusion

    International Nuclear Information System (INIS)

    Jahoda, F.C.

    1982-01-01

    The lectures will begin by defining high density in the context of magnetic confinement fusion research and listing some alternative reactor concepts, ranging from n/sub e/ approx. 2 x 10 14 cm -3 to several orders of magnitude greater, that offer potential advantages over the main-line, n/sub e/ approx. 1 x 10 14 cm -3 , Tokamak reactor designs. The high density scalings of several major diagnostic techniques, some favorable and some disadvantageous, will be discussed. Special emphasis will be given to interferometric methods, both electronic and photographic, for which integral n/sub e/dl measurements and associated techniques are accessible with low wavelength lasers. Reactor relevant experience from higher density, smaller dimension devices exists. High density implies high β, which implies economies of scale. The specialized features of high β diagnostics will be discussed

  15. BCS Theory of Hadronic Matter at High Densities

    DEFF Research Database (Denmark)

    Bohr, Henrik; Panda, Prafulla K.; Providencia, Constanca

    2012-01-01

    The equilibrium between the so-called 2SC and CFL phases of strange quark matter at high densities is investigated in the framework of a simple schematic model of the NJL type. Equal densities are assumed for quarks u, d and s. The 2SC phase is here described by a color-flavor symmetric state, in...

  16. Antioxidant activity of high-density lipoprotein (HDL) using different ...

    African Journals Online (AJOL)

    HDL is a potent antioxidant in terms of inhibition of lipid peroxidation, ROS production and LDL oxidation. These may to some extent add to the antiatherogenic beyond reverse-cholesterol transport properties of HDL. Keywords: high-density lipoprotein; reverse cholesterol transport; apolipoprotein A1; antioxidant; in vitro.

  17. Morphodynamics of supercritical high-density turbidity currents

    NARCIS (Netherlands)

    Cartigny, M.

    2012-01-01

    Seafloor and outcrop observations combined with numerical and physical experiments show that turbidity currents are likely 1) to be in a supercritical flow state and 2) to carry high sediment concentrations (being of high-density). The thesis starts with an experimental study of bedforms

  18. Fluorescent Fe K Emission from High Density Accretion Disks

    Science.gov (United States)

    Bautista, Manuel; Mendoza, Claudio; Garcia, Javier; Kallman, Timothy R.; Palmeri, Patrick; Deprince, Jerome; Quinet, Pascal

    2018-06-01

    Iron K-shell lines emitted by gas closely orbiting black holes are observed to be grossly broadened and skewed by Doppler effects and gravitational redshift. Accordingly, models for line profiles are widely used to measure the spin (i.e., the angular momentum) of astrophysical black holes. The accuracy of these spin estimates is called into question because fitting the data requires very high iron abundances, several times the solar value. Meanwhile, no plausible physical explanation has been proffered for why these black hole systems should be so iron rich. The most likely explanation for the super-solar iron abundances is a deficiency in the models, and the leading candidate cause is that current models are inapplicable at densities above 1018 cm-3. We study the effects of high densities on the atomic parameters and on the spectral models for iron ions. At high densities, Debye plasma can affect the effective atomic potential of the ions, leading to observable changes in energy levels and atomic rates with respect to the low density case. High densities also have the effec of lowering energy the atomic continuum and reducing the recombination rate coefficients. On the spectral modeling side, high densities drive level populations toward a Boltzman distribution and very large numbers of excited atomic levels, typically accounted for in theoretical spectral models, may contribute to the K-shell spectrum.

  19. Sputtered thin films for high density tape recording

    NARCIS (Netherlands)

    Nguyen, L.T.

    This thesis describes the investigation of sputtered thin film media for high density tape recording. As discussed in Chapter 1, to meet the tremendous demand of data storage, the density of recording tape has to be increased continuously. For further increasing the bit density the key factors are:

  20. Ultra-stretchable Interconnects for high-density stretchable electronics

    NARCIS (Netherlands)

    Shafqat, S.; Hoefnagels, J.P.M.; Savov, A.; Joshi, S.; Dekker, R.; Geers, M.G.D.

    2017-01-01

    The exciting field of stretchable electronics (SE) promises numerous novel applications, particularly in-body and medical diagnostics devices. However, future advanced SE miniature devices will require high-density, extremely stretchable interconnects with micron-scale footprints, which calls for

  1. Positron camera with high-density avalanche chambers

    International Nuclear Information System (INIS)

    Manfrass, D.; Enghardt, W.; Fromm, W.D.; Wohlfarth, D.; Hennig, K.

    1988-01-01

    The results of an extensive investigation of the properties of high-density avalanche chambers (HIDAC) are presented. This study has been performed in order to optimize the layout of HIDAC detectors, since they are intended to be applied as position sensitive detectors for annihilation radiation in a positron emission tomograph being under construction. (author)

  2. Role of Lipids in Spheroidal High Density Lipoproteins

    NARCIS (Netherlands)

    Vuorela, Timo; Catte, Andrea; Niemela, Perttu S.; Hall, Anette; Hyvonen, Marja T.; Marrink, Siewert-Jan; Karttunen, Mikko; Vattulainen, Ilpo

    2010-01-01

    We study the structure and dynamics of spherical high density lipoprotein (HDL) particles through coarse-grained multi-microsecond molecular dynamics simulations. We simulate both a lipid droplet without the apolipoprotein A-I (apoA-I) and the full HDL particle including two apoA-I molecules

  3. Role of lipids in spheroidal high density lipoproteins

    NARCIS (Netherlands)

    Vuorela, T.A.; Catte, A.; Niemelä, P.S.; Hall, A.; Hyvönen, M.T.; Marrink, S.J.; Karttunen, M.E.J.; Vattulainen, I.

    2010-01-01

    We study the structure and dynamics of spherical high density lipoprotein (HDL) particles through coarse-grained multi-microsecond molecular dynamics simulations. We simulate both a lipid droplet without the apolipoprotein A-I (apoA-I) and the full HDL particle including two apoA-I molecules

  4. Interfacial stick–slip transition in hydroxyapatite filled high density ...

    Indian Academy of Sciences (India)

    Unknown

    flow curves of composites and that of unfilled system remain identical. Filler addition lowers the .... Injection moulding grade high density polyethylene,. HD6070EA, was ... rheometer (Rosand Precision Ltd., UK) using version. 6⋅10 software. .... Bagley E B, Cabbot I M and West D C 1958 J. Appl. Phys. 29. 109. Blyler L L and ...

  5. Spontaneous magnetization in high-density quark matter

    DEFF Research Database (Denmark)

    Tsue, Yasuhiko; da Providência, João; Providência, Constanca

    2015-01-01

    It is shown that spontaneous magnetization occurs due to the anomalous magnetic moments of quarks in high-density quark matter under the tensor-type four-point interaction. The spin polarized condensate for each flavor of quark appears at high baryon density, which leads to the spontaneous magnet...

  6. The Influence of Decreased Levels of High Density Lipoprotein ...

    African Journals Online (AJOL)

    Background: Changes in lipoproteins levels in sickle cell disease (SCD) patients are well.known, but the physiological ramifications of the low levels observed have not been entirely resolved. Aim: The aim of this study is to evaluate the impact of decreased levels of high density lipoprotein cholesterol (HDL.c) on ...

  7. 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.

  8. Development and implementation of computational geometric model for simulation of plate type fuel fabrication process with microspheres dispersed in metallic matrix

    International Nuclear Information System (INIS)

    Lage, Aldo M.F.; Reis, Sergio C.; Braga, Daniel M.; Santos, Armindo; Ferraz, Wilmar B.

    2005-01-01

    In this report it is presented the development of a geometric model to simulate the plate type fuel fabrication process with fuels microspheres dispersed in metallic matrix, as well as its software implementation. The developed geometric model encloses the steps of pellets pressing and sintering, as well as the plate rolling passes. The model permits the simulation of structures, where the values of the various variables of the fabrication processes can be studied and modified. The following variables were analyzed: microspheres diameters, density of the powder/microspheres mixing, microspheres density, fuel volume fraction, sintering densification, and rolling passes number. In the model implementation, which was codified in DELPHI programming language, systems of structured analysis techniques were utilized. The structures simulated were visualized utilizing the AutoCAD applicative, what permitted to obtain planes sections in diverse directions. The objective of this model is to enable the analysis of the simulated structures and supply information that can help in the improvement of the dispersion microspheres fuel plates fabrication process, now in development at CDTN (Centro de Desenvolvimento da Tecnologia Nuclear) in cooperation with the CTMSP (Centro Tecnologico da Marinha em Sao Paulo). (author)

  9. 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.

  10. Qualification status of LEU [low enriched uranium] fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.

    1987-01-01

    Sufficient data has been obtained from tests of high-density, low-enriched fuels for research and test reactors to declare them qualified for use. These fuels include UZrH x (TRIGA fuel) and UO 2 (SPERT fuel) for rod-type reactors and UAl x , U 3 O 8 , U 3 Si 2 , and U 3 Si dispersed in aluminium for plate-type reactors. Except for U 3 Si, the allowable fission density for LEU applications is limited only by the available 235 U. Several reactors are now using these fuels, and additional conversions are in progress. The basic performance characteristics and limits, if any, of the qualified low-enriched (and medium-enriched) fuels are discussed. Continuing and planned work to qualify additional fuels is also discussed. (Author)

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

    International Nuclear Information System (INIS)

    Medvedev, Pavel G.

    2009-01-01

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

  12. High density plasmas formation in Inertial Confinement Fusion and Astrophysics

    International Nuclear Information System (INIS)

    Martinez-Val, J. M.; Minguez, E.; Velarde, P.; Perlado, J. M.; Velarde, G.; Bravo, E.; Eliezer, S.; Florido, R.; Garcia Rubiano, J.; Garcia-Senz, D.; Gil de la Fe, J. M.; Leon, P. T.; Martel, P.; Ogando, F.; Piera, M.; Relano, A.; Rodriguez, R.; Garcia, C.; Gonzalez, E.; Lachaise, M.; Oliva, E.

    2005-01-01

    In inertially confined fusion (ICF), high densities are required to obtain high gains. In Fast Ignition, a high density, low temperature plasma can be obtained during the compression. If the final temperature reached is low enough, the electrons of the plasma can be degenerate. In degenerate plasmas. Bremsstrahlung emission is strongly suppressed an ignition temperature becomes lower than in classical plasmas, which offers a new design window for ICF. The main difficulty of degenerate plasmas in the compression energy needed for high densities. Besides that, the low specific heat of degenerate electrons (as compared to classical values) is also a problem because of the rapid heating of the plasma. Fluid dynamic evolution of supernovae remnants is a very interesting problem in order to predict the thermodynamical conditions achieved in their collision regions. Those conditions have a strong influence in the emission of light and therefore the detection of such events. A laboratory scale system has been designed reproducing the fluid dynamic field in high energy experiments. The evolution of the laboratory system has been calculated with ARWEN code, 2D Radiation CFD that works with Adaptive Mesh Refinement. Results are compared with simulations on the original system obtained with a 3D SPH astrophysical code. New phenomena at the collision plane and scaling of the laboratory magnitudes will be described. Atomic physics for high density plasmas has been studied with participation in experiments to obtain laser produced high density plasmas under NLTE conditions, carried out at LULI. A code, ATOM3R, has been developed which solves rate equations for optically thin plasmas as well as for homogeneous optically thick plasmas making use of escape factors. New improvements in ATOM3R are been done to calculate level populations and opacities for non homogeneous thick plasmas in NLTE, with emphasis in He and H lines for high density plasma diagnosis. Analytical expression

  13. Fluidisation and dispersion behaviour of small high density pellicular expanded bed adsorbents

    DEFF Research Database (Denmark)

    Theodossiou, Irini; Elsner, H.D.; Thomas, Owen R. T.

    2002-01-01

    correlation for characterisation of expanded bed systems is questioned. Residence time distribution studies using acetone tracers, demonstrated that in comparison to existing commercial supports, the small pellicular prototype materials generally possessed far superior hydrodynamic properties, which augurs......, fluidisation behaviour was poorly predicted from the Richardson-Zaki correlation, with experimentally determined values of the expansion index being considerably higher than the theoretical values. The reasons for these discrepancies are discussed in detail and the validity of applying this widely used...

  14. High density data storage principle, technology, and materials

    CERN Document Server

    Zhu, Daoben

    2009-01-01

    The explosive increase in information and the miniaturization of electronic devices demand new recording technologies and materials that combine high density, fast response, long retention time and rewriting capability. As predicted, the current silicon-based computer circuits are reaching their physical limits. Further miniaturization of the electronic components and increase in data storage density are vital for the next generation of IT equipment such as ultra high-speed mobile computing, communication devices and sophisticated sensors. This original book presents a comprehensive introduction to the significant research achievements on high-density data storage from the aspects of recording mechanisms, materials and fabrication technologies, which are promising for overcoming the physical limits of current data storage systems. The book serves as an useful guide for the development of optimized materials, technologies and device structures for future information storage, and will lead readers to the fascin...

  15. Noise reduction in muon tomography for detecting high density objects

    International Nuclear Information System (INIS)

    Benettoni, M; Checchia, P; Cossutta, L; Furlan, M; Gonella, F; Pegoraro, M; Garola, A Rigoni; Ronchese, P; Vanini, S; Viesti, G; Bettella, G; Bonomi, G; Donzella, A; Subieta, M; Zenoni, A; Calvagno, G; Cortelazzo, G; Zanuttigh, P; Calvini, P; Squarcia, S

    2013-01-01

    The muon tomography technique, based on multiple Coulomb scattering of cosmic ray muons, has been proposed as a tool to detect the presence of high density objects inside closed volumes. In this paper a new and innovative method is presented to handle the density fluctuations (noise) of reconstructed images, a well known problem of this technique. The effectiveness of our method is evaluated using experimental data obtained with a muon tomography prototype located at the Legnaro National Laboratories (LNL) of the Istituto Nazionale di Fisica Nucleare (INFN). The results reported in this paper, obtained with real cosmic ray data, show that with appropriate image filtering and muon momentum classification, the muon tomography technique can detect high density materials, such as lead, albeit surrounded by light or medium density material, in short times. A comparison with algorithms published in literature is also presented

  16. Operation and control of high density tokamak reactors

    International Nuclear Information System (INIS)

    Attenberger, S.E.; McAlees, D.G.

    1976-01-01

    The incentive for high density operation of a tokamak reactor is discussed. The plasma size required to attain ignition is determined. Ignition is found to be possible in a relatively small system provided other design criteria are met. These criteria are described and the technology developments and operating procedures required by them are outlined. The parameters for such a system and its dynamic behavior during the operating cycle are also discussed

  17. Volume generation of negative ions in high density hydrogen discharges

    International Nuclear Information System (INIS)

    Hiskes, J.R.; Karo, A.M.

    1983-01-01

    A parametric survey is made of a high-density tandem two-chamber hydrogen negative ion system. The optimum extracted negative ion current densities are sensitive to the atom concentration in the discharge and to the system scale length. For scale lengths ranging from 10 cm to 0.1 cm optimum current densities range from of order 1 to 100 mA cm -2 , respectively

  18. Apparatus and method for generating high density pulses of electrons

    International Nuclear Information System (INIS)

    Lee, C.; Oettinger, P.E.

    1981-01-01

    An apparatus and method are described for the production of high density pulses of electrons using a laser energized emitter. Caesium atoms from a low pressure vapour atmosphere are absorbed on and migrate from a metallic target rapidly heated by a laser to a high temperature. Due to this heating time being short compared with the residence time of the caesium atoms adsorbed on the target surface, copious electrons are emitted which form a high current density pulse. (U.K.)

  19. Relativistic many-body theory of high density matter

    International Nuclear Information System (INIS)

    Chin, S.A.

    1977-01-01

    A fully relativistic quantum many-body theory is applied to the study of high-density matter. The latter is identified with the zero-temperature ground state of a system of interacting baryons. In accordance with the observed short-range repulsive and long-range attractive character of the nucleon--nucleon force, baryons are described as interacting with each other via a massive scalar and a massive vector meson exchange. In the Hartree approximation, the theory yields the same result as the mean-field theory, but with additional vacuum fluctuation corrections. The resultant equation of state for neutron matter is used to determine properties of neutron stars. The relativistic exchange energy, its corresponding single-particle excitation spectrum, and its effect on the neutron matter equation of state, are calculated. The correlation energy from summing the set of ring diagrams is derived directly from the energy-momentum tensor, with renormalization carried out by adding counterterms to the original Lagrangian and subtracting purely vacuum expectation values. Terms of order g 4 lng 2 are explicitly given. Effects of scalar-vector mixing are discussed. Collective modes corresponding to macroscopic density fluctuation are investigated. Two basic modes are found, a plasma-like mode and zero sound, with the latter dominant at high density. The stability and damping of these modes are studied. Last, the effect of vacuum polarization in high-density matter is examined

  20. Evaporation of carbon using electrons of a high density plasma

    International Nuclear Information System (INIS)

    Muhl, S.; Camps, E.; Escobar A, L.; Garcia E, J.L.; Olea, O.

    1999-01-01

    The high density plasmas are used frequently in the preparation of thin films or surface modification, for example to nitridation. In these processes, are used mainly the ions and the neutrals which compose the plasma. However, the electrons present in the plasma are not used, except in the case of chemical reactions induced by collisions, although the electron bombardment usually get hot the work piece. Through the adequate polarization of a conductor material, it is possible to extract electrons from a high density plasma at low pressure, that could be gotten the evaporation of this material. As result of the interaction between the plasma and the electron flux with the vapor produced, this last will be ionized. In this work, it is reported the use of this novelty arrangement to prepare carbon thin films using a high density argon plasma and a high purity graphite bar as material to evaporate. It has been used substrates outside plasma and immersed in the plasma. Also it has been reported the plasma characteristics (temperature and electron density, energy and ions flux), parameters of the deposit process (deposit rate and ion/neutral rate) as well as the properties of the films obtained (IR absorption spectra and UV/Vis, elemental analysis, hardness and refractive index. (Author)

  1. High-density cervical ureaplasma urealyticum colonization in pregnant women

    Directory of Open Access Journals (Sweden)

    Ranđelović Gordana

    2006-01-01

    Full Text Available Background/aim: Ureaplasma urealyticum, a common commensal of the female lower genital tract, has been observed as an important opportunistic pathogen during pregnancy. The aims of this study were to determine the degree of cervical colonization with U. urealyticum in pregnant women with risk pregnancy and in pregnant women with normal term delivery and to evaluate the correlation between high-density cervical U. urealyticum colonization and premature rupture of membranes (PROM as well. Methods. This research was conducted on the samples comprising 130 hospitalized pregnant women with threatening preterm delivery and premature rupture of membranes. The control group consisted of 39 pregnant women with term delivery without PROM. In addition to standard bacteriological examination and performing direct immunofluorescence test to detect Chlamydia trachomatis, cervical swabs were also examined for the presence of U. urealyticum and Mycoplasma hominis by commercially available Mycofast Evolution 2 test (International Microbio, France. Results. The number of findings with isolated high-density U. urealyticum in the target group was 69 (53.08%, while in the control group was 14 (35.90%. Premature rupture of membranes (PROM occurred in 43 (33.08% examinees: 29 were pPROM, and 14 were PROM. The finding of U.urealyticum ≥104 was determined in 25 (58.14% pregnant women with rupture, 17 were pPROM, and 8 were PROM. There was statistically significant difference in the finding of high-density U. urealyticum between the pregnant women with PROM and the control group (χ² = 4.06, p < 0.05. U. urealyticum was predominant bacterial species found in 62.79% of isolates in the PROM cases, while in 32.56% it was isolated alone. Among the 49 pregnant women with preterm delivery, pPROM occurred in 29 (59.18% examinees, and in 70.83% of pregnant women with findings of high-density U. urealyticum pPROM was observed. Conclusion. Cervical colonization with U

  2. Characterization of the high density plasma etching process of CCTO thin films for the fabrication of very high density capacitors

    International Nuclear Information System (INIS)

    Altamore, C; Tringali, C; Sparta', N; Marco, S Di; Grasso, A; Ravesi, S

    2010-01-01

    In this work the feasibility of CCTO (Calcium Copper Titanate) patterning by etching process is demonstrated and fully characterized in a hard to etch materials etcher. CCTO sintered in powder shows a giant relative dielectric constant (10 5 ) measured at 1 MHz at room temperature. This feature is furthermore coupled with stability from 10 1 Hz to 10 6 Hz in a wide temperature range (100K - 600K). In principle, this property can allow to fabricate very high capacitance density condenser. Due to its perovskite multi-component structure, CCTO can be considered a hard to etch material. For high density capacitor fabrication, CCTO anisotropic etching is requested by using high density plasma. The behavior of etched CCTO was studied in a HRe- (High Density Reflected electron) plasma etcher using Cl 2 /Ar chemistry. The relationship between the etch rate and the Cl 2 /Ar ratio was also studied. The effects of RF MHz, KHz Power and pressure variation, the impact of HBr addiction to the Cl 2 /Ar chemistry on the CCTO etch rate and on its selectivity to Pt and photo resist was investigated.

  3. Characterization of the high density plasma etching process of CCTO thin films for the fabrication of very high density capacitors

    Energy Technology Data Exchange (ETDEWEB)

    Altamore, C; Tringali, C; Sparta' , N; Marco, S Di; Grasso, A; Ravesi, S [STMicroelectronics, Industial and Multi-segment Sector R and D, Catania (Italy)

    2010-02-15

    In this work the feasibility of CCTO (Calcium Copper Titanate) patterning by etching process is demonstrated and fully characterized in a hard to etch materials etcher. CCTO sintered in powder shows a giant relative dielectric constant (10{sup 5}) measured at 1 MHz at room temperature. This feature is furthermore coupled with stability from 10{sup 1} Hz to 10{sup 6} Hz in a wide temperature range (100K - 600K). In principle, this property can allow to fabricate very high capacitance density condenser. Due to its perovskite multi-component structure, CCTO can be considered a hard to etch material. For high density capacitor fabrication, CCTO anisotropic etching is requested by using high density plasma. The behavior of etched CCTO was studied in a HRe- (High Density Reflected electron) plasma etcher using Cl{sub 2}/Ar chemistry. The relationship between the etch rate and the Cl{sub 2}/Ar ratio was also studied. The effects of RF MHz, KHz Power and pressure variation, the impact of HBr addiction to the Cl{sub 2}/Ar chemistry on the CCTO etch rate and on its selectivity to Pt and photo resist was investigated.

  4. Investigation of Y2O3 distribution in the welded joints of the fast reactor fuel claddings made of oxide dispersion strengthened steel

    International Nuclear Information System (INIS)

    Tabakin, E.M.; Kuz'min, S.V.; Ivanovich, Yu.V.; Ukai, Sh.; Kaito, T.; Seki, M.

    2007-01-01

    The study results of Y 2 O 3 distribution in welded joints of claddings from oxide dispersion strengthened steel produced by the technique of powder metallurgy are given in this paper. Change of content and distribution uniformity of yttrium oxide in welds in comparison with metal shell is the result of using flash welding of thin-walled fuel claddings. It is shown that concentration and yttrium oxide distribution uniformity in the cross section of weld, made by pulse laser welding is more high as compared with argon-arc welding [ru

  5. The effect of fuel burnup and dispersed water intrusion on the criticality of spent high-level nuclear fuel in a geologic repository

    International Nuclear Information System (INIS)

    Culbreth, W.G.; Zielinski, P.R.

    1994-01-01

    Studies of the spent fuel waste package have been conducted through the use of a Monte-Carlo neutron simulation program to determine the ability of the fuel to sustain a chain reaction. These studies have included fuel burnup and the effect of water mists on criticality. Results were compared with previous studies. In many criticality studies of spent fuel waste packages, fresh fuel with an enrichment as high as 4.5% is used as the conservative (worst) case. The actual spent fuel has a certain amount of burnup that decreases the concentration of fissile uranium and increases the amount of radionuclides present. The LWR Radiological Data Base from OCRWM has been used to determine the relative radionuclide ratios and KENO 5.1 was used to calculate values of the effective multiplication factor, k eff . Spent fuel is not capable of sustaining a chain reaction unless a suitable moderator, such as water, is present. A completely flooded container has been treated as the worst case for criticality. Results of a previous report that demonstrated that k eff actually peaked at a water-to-mixture ratio of 13% were analyzed for validity. In the present study, these results did not occur in the SCP waste package container

  6. Biomimetic High Density Lipoprotein Nanoparticles For Nucleic Acid Delivery

    Science.gov (United States)

    McMahon, Kaylin M.; Mutharasan, R. Kannan; Tripathy, Sushant; Veliceasa, Dorina; Bobeica, Mariana; Shumaker, Dale K.; Luthi, Andrea J.; Helfand, Brian T.; Ardehali, Hossein; Mirkin, Chad A.; Volpert, Olga; Thaxton, C. Shad

    2014-01-01

    We report a gold nanoparticle-templated high density lipoprotein (HDL AuNP) platform for gene therapy which combines lipid-based nucleic acid transfection strategies with HDL biomimicry. For proof-of-concept, HDL AuNPs are shown to adsorb antisense cholesterylated DNA. The conjugates are internalized by human cells, can be tracked within cells using transmission electron microscopy (TEM), and regulate target gene expression. Overall, the ability to directly image the AuNP core within cells, the chemical tailorability of the HDL AuNP platform, and the potential for cell-specific targeting afforded by HDL biomimicry make this platform appealing for nucleic acid delivery. PMID:21319839

  7. High Density Thermal Energy Storage with Supercritical Fluids

    Science.gov (United States)

    Ganapathi, Gani B.; Wirz, Richard

    2012-01-01

    A novel approach to storing thermal energy with supercritical fluids is being investigated, which if successful, promises to transform the way thermal energy is captured and utilized. The use of supercritical fluids allows cost-affordable high-density storage with a combination of latent heat and sensible heat in the two-phase as well as the supercritical state. This technology will enhance penetration of several thermal power generation applications and high temperature water for commercial use if the overall cost of the technology can be demonstrated to be lower than the current state-of-the-art molten salt using sodium nitrate and potassium nitrate eutectic mixtures.

  8. Optically Addressed Nanostructures for High Density Data Storage

    Science.gov (United States)

    2005-10-14

    beam to sub-wavelength resolutions. X. Refereed Journal Publications I. M. D. Stenner , D. J. Gauthier, and M. A. Neifeld, "The speed of information in a...profiles for high-density optical data storage," Optics Communications, Vol.253, pp.56-69, 2005. 5. M. D. Stenner , D. J. Gauthier, and M. A. Neifeld, "Fast...causal information transmission in a medium with a slow group velocity," Physical Review Letters, Vol.94, February 2005. 6. M. D. Stenner , M. A

  9. Low-frequency oscillations at high density in JFT-2

    International Nuclear Information System (INIS)

    Maeno, Masaki; Katagiri, Masaki; Suzuki, Norio; Fujisawa, Noboru

    1977-12-01

    Low-frequency oscillations in a plasma were measured with magnetic probes and Si surface-barrier detectors, and behaviour of the high density plasmas was studied. The plasma current profile in the phase of decreasing density after the interruption of gas input is more peaked than during gas input. The introduction of hydrogen during a discharge results in a reduction of the impurities flux. The increase of density by fast gas input is limited with a negative voltage spike. Immediately before a negative voltage spike, oscillations of m=1,2 grow, leading to the spike. (auth.)

  10. Possible new form of matter at high density

    International Nuclear Information System (INIS)

    Lee, T.D.

    1974-01-01

    As a preliminary to discussion of the possibility of new forms of matter at high density, questions relating to the vacuum and vacuum excitation are considered. A quasi-classical approach to the development of abnormal nuclear states is undertaken using a Fermi gas of nucleons of uniform density. Discontinuous transitions are considered in the sigma model (tree approximation) followed by brief consideration of higher order loop diagrams. Production and detection of abnormal nuclear states are discussed in the context of high energy heavy ion collisions. Remarks are made on motivation for such research. 8 figures

  11. Characterization of composite high density polyethylene and layered zirconium phosphate

    International Nuclear Information System (INIS)

    Lino, Adan S.; Silva, Daniela F.; Mendes, Luis C.

    2011-01-01

    Zirconium phosphate (ZrP) (2 w%), synthesized by direct precipitation method, was used in the preparation of composite with high density polyethylene (HDPE), through extrusion processing in the molten state. Wide angle x-ray diffraction (WAXD), stress-strain mechanical analysis and scanning electron microscopy (SEM) techniques were used for ZrP, neat polymer and composite mechanical and morphologic characterization. Although there was a slight increase in the Young modulus, WAXD and SEM analysis showed that the intercalation of the HDPE matrix in the filler galleries did not occur, probably due to the insufficient lamellae spacing to intercalate the polymer chains. Then, a microcomposite was achieved. (author)

  12. Structure of high-density amorphous ice under pressure

    International Nuclear Information System (INIS)

    Klotz, S.; Hamel, G.; Loveday, J.S.; Nelmes, R.J.; Guthrie, M.; Soper, A.K.

    2002-01-01

    We report in situ neutron diffraction studies of high-density amorphous ice (HDA) at 100 K at pressures up to 2.2 GPa. We find that the compression is achieved by a strong contraction (∼20%) of the second neighbor coordination shell, so that at 2.2 GPa it closely approaches the first coordination shell, which itself remains intact in both structure and size. The hydrogen bond orientations suggest an absence of hydrogen bonding between first and second shells and that HDA has increasingly interpenetrating hydrogen bond networks under pressure

  13. Development status of metallic, dispersion and non-oxide advanced and alternative fuels for power and research reactors

    International Nuclear Information System (INIS)

    2003-09-01

    The current thermal power reactors use less than 1% of the energy contained in uranium. Long term perspectives aiming at a better economical extraction of the potential supplied by uranium motivated the development of new reactor types and, of course, new fuel concepts. Most of them dated from the sixties including liquid metal cooled fast (FR) and high temperature gas cooled (HTGR) reactors. Unfortunately, these impulses slowed down during the last twenty years; nuclear energy had to face political and consensus problems, in particular in the United States of America and in Europe, resulting from the consequences of the TMI and Chernobyl accidents. Good economical results obtained by the thermal power reactors also contributed to this process. During the last twenty years mainly France, India, Japan and the Russian Federation have maintained a relatively high level of technological development with appropriate financial items, in particular, in fuel research for the above mentioned reactor types. China and South Africa are now progressing in development of FR/HTGR and HTGR technologies, respectively. The purpose of this report is not only to summarise knowledge accumulated in the fuel research since the beginning of the sixties. This subject has been well covered in literature up to the end of the eighties. This report rather concentrates on the 'advanced fuels 'for the current different types of reactors including metallic, carbide and nitride fuels for fast reactors, so-called 'cold' fuels and fuels to burn excessive ex-weapons plutonium in thermal power reactors, alternative fuels for small size and research reactors. Emphasis has been put on the aspects of fabrication and irradiation behaviour of these fuels; available basic data concerning essential properties that help to understand the phenomena have been mentioned as well. This report brings complementary information to the earlier published monographs and concerns developments carried out after the early

  14. 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)

  15. Multiplexed, high density electrophysiology with nanofabricated neural probes.

    Directory of Open Access Journals (Sweden)

    Jiangang Du

    Full Text Available Extracellular electrode arrays can reveal the neuronal network correlates of behavior with single-cell, single-spike, and sub-millisecond resolution. However, implantable electrodes are inherently invasive, and efforts to scale up the number and density of recording sites must compromise on device size in order to connect the electrodes. Here, we report on silicon-based neural probes employing nanofabricated, high-density electrical leads. Furthermore, we address the challenge of reading out multichannel data with an application-specific integrated circuit (ASIC performing signal amplification, band-pass filtering, and multiplexing functions. We demonstrate high spatial resolution extracellular measurements with a fully integrated, low noise 64-channel system weighing just 330 mg. The on-chip multiplexers make possible recordings with substantially fewer external wires than the number of input channels. By combining nanofabricated probes with ASICs we have implemented a system for performing large-scale, high-density electrophysiology in small, freely behaving animals that is both minimally invasive and highly scalable.

  16. Mixing of high density solution in vertical upward flow

    International Nuclear Information System (INIS)

    Kumamaru, Hiroshige; Hosogi, Nobuyoshi; Komada, Toshiaki; Fujiwara, Yoshiki

    1999-01-01

    Experimental and analytical studies have been performed in order to provide fundamental data and a numerical calculation model on the mixing of boric acid solution, injected from the standby liquid control system (SLCS), under a low natural circulation flow during an ATWS in a BWR. First, fundamental experiments on the mixing of high-density solution in vertically-upward water flow have been performed by using a small apparatus. Mixing patterns observed in the experiments have been classified to two groups, i.e. complete mixing (entrainment) and incomplete mixing (entrainment). In the complete mixing, the injected high-density solution is mixed (entrained) completely into the vertically-upward water flow. From the experiments, the minimum water flow rates in which the complete mixing (entrainment) is achieved have been obtained for various solution densities and solution injection rates. Secondly, two-dimensional numerical calculations have been performed. A continuity equation for total fluid, momentum equations in two directions and a continuity equation for solute are solved by using the finite difference method for discretization method and by following the MAC method for solution procedure. The calculations have predicted nearly the minimum water flow rate in which the complete mixing is achieved, while the calculations have been performed only for one combination of the solution density and solution injection rate until now. (author)

  17. Ultra-Stretchable Interconnects for High-Density Stretchable Electronics

    Directory of Open Access Journals (Sweden)

    Salman Shafqat

    2017-09-01

    Full Text Available The exciting field of stretchable electronics (SE promises numerous novel applications, particularly in-body and medical diagnostics devices. However, future advanced SE miniature devices will require high-density, extremely stretchable interconnects with micron-scale footprints, which calls for proven standardized (complementary metal-oxide semiconductor (CMOS-type process recipes using bulk integrated circuit (IC microfabrication tools and fine-pitch photolithography patterning. Here, we address this combined challenge of microfabrication with extreme stretchability for high-density SE devices by introducing CMOS-enabled, free-standing, miniaturized interconnect structures that fully exploit their 3D kinematic freedom through an interplay of buckling, torsion, and bending to maximize stretchability. Integration with standard CMOS-type batch processing is assured by utilizing the Flex-to-Rigid (F2R post-processing technology to make the back-end-of-line interconnect structures free-standing, thus enabling the routine microfabrication of highly-stretchable interconnects. The performance and reproducibility of these free-standing structures is promising: an elastic stretch beyond 2000% and ultimate (plastic stretch beyond 3000%, with <0.3% resistance change, and >10 million cycles at 1000% stretch with <1% resistance change. This generic technology provides a new route to exciting highly-stretchable miniature devices.

  18. Manufacture of sintered bricks of high density from beryllium oxide

    International Nuclear Information System (INIS)

    Pointud, R.; Rispal, Ch.; Le Garec, M.

    1959-01-01

    Beryllium oxide bricks of nuclear purity 100 x 100 x 50 and 100 x 100 x 100 mm of very high density (between 2.85 and 3.00) are manufactured by sintering under pressure in graphite moulds at temperatures between 1,750 and 1,850 deg. C, and under a pressure of 150 kg/cm 2 . The physico-chemical state of the saw material is of considerable importance with regard to the success of the sintering operation. In addition, a study of the sintering of a BeO mixture with 3 to 5 per cent of boron introduced in the form of boric acid, boron carbide or elementary boron shows that high densities can only be obtained by sintering under pressure. For technical reasons of manufacture, only the mixture based on boron carbide is used. The sintering is carried out in graphite moulds at 1500 deg. C under 150 kg/cm 2 pressure, and bricks can be obtained with density between 2,85 and 2,90. Laboratory studies and the industrial manufacture of various sinters are described in detail. (author) [fr

  19. Some evidence for the high density phase of excitons in CdS

    International Nuclear Information System (INIS)

    Rueckmann, I.; May, V.; Voigt, J.

    1980-01-01

    Reflection spectra without and with additional pumping are measured in CdS at 1.8 and 77 K, respectively, starting from very low pump intensities. At low pump intensities a special behaviour of the reflection minimum is found indicating the importance of an exciton dead layer. The experimental spectra up to the highest pump intensities can be fitted very well taking into account spatial dispersion, exciton dead layer, and density dependent excitonic parameters. Polarizability, damping, and layer thickness change continuously in the whole range of excitation intensities. On the other side, the transverse dielectric function of a high density gas of interacting 1s excitons is calculated. A comparison between theoretically and experimentally obtained excitonic parameters at different densities shows good agreement. Hence, the importance of many-exciton interaction for the reflection spectra up to vanishing at highest pump intensities is concluded. (author)

  20. 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

  1. Thermal and mechanical properties of injection molded recycled high density polyethylene blends with virgin isotactic polypropylene

    International Nuclear Information System (INIS)

    Madi, N.K.

    2013-01-01

    Highlights: ► Recycled high density polyethylene and isotactic polypropylene blends have been prepared by melt compounding. ► Thermal study showed that iPP is not well dispersed into the rHDPE matrix. ► Tensile testing shows that there is strong correlation between the thermal properties and the tensile behavior of rHDPE/ipp blends. - Abstract: Polymer blending has become an important field in polymer research and especially in the area of recycling. In this research the target was to reduce the polymer waste problem. Therefore, recycled high density polyethylene (rHDPE) and virgin isotactic polypropylene (vPP) blends containing upto 30 wt% of vPP have been prepared by melt compounding method using injection molding at 220 °C. The thermal properties, thermal degradation and the mechanical properties of the polymer blends were studied using differential scanning calorimetry (DSC), Thermogravimetric analysis (TGA), and tensile testing method. DSC study shows that in all the blends there are two melting peaks, one around the melting temperature of rHDPE and another one around the melting point of vPP, indicating that vPP is not well dispersed into the rHDPE matrix. The changes in the heat of fusion for the rHDPE/iPP polymer blends versus vPP content suggests that incorporating vPP affects the crystallinity of the system. TGA analysis of the polymer blends shows that parts of rHDPE with 95/5 upto 80/20 of vPP are mostly stable composition which brings about valuable stabilization to the rHDPE. Tensile testing shows that there is strong correlation between the thermal properties and the tensile behavior of rHDPE/vpp blends

  2. Development and Characterization of Gas Diffusion Layer Using Carbon Slurry Dispersed by Ammonium Lauryl Sulfate for Proton Exchange Member Fuel Cells

    Science.gov (United States)

    Villacorta, Rashida

    Gas diffusion layers (GDLs) are a critical and essential part of proton exchange membrane fuel cells (PEMFCs). They carry out various important functions such as transportation of reactants to and from the reaction sites. The material properties and structural characteristics of the substrate and the microporous layer strongly influence fuel cell performance. The microporous layer of the GDLs was fabricated with the carbon slurry dispersed in water containing ammonium lauryl sulfate (ALS) using the wire rod coating method. GDLs were fabricated with different materials to compose the microporous layer and evaluated the effects on PEMFC power output performance. The consistency of the carbon slurry was achieved by adding 25 wt. % of PTFE, a binding agent with a 75:25 ratio of carbon (Pureblack and vapor grown carbon fiber). The GDLs were investigated in PEMFC under various relative humidity (RH) conditions using H2/O2 and H2/Air. GDLs were also fabricated with the carbon slurry dispersed in water containing sodium dodecyl sulfate (SDS) and multiwalled carbon nanotubes (MWCNTs) with isopropyl alcohol (IPA) based for fuel cell performance comparison. MWCNTs and SDS exhibits the highest performance at 60% and 70% RH with a peak power density of 1100 mW.cm-2 and 850 mW.cm-2 using air and oxygen as an oxidant. This means that the gas diffusion characteristics of these two samples were optimum at 60 and 70 % RH with high limiting current density range. It was also found that the composition of the carbon slurry, specifically ALS concentration has the highest peak power density of 1300 and 500mW.cm-2 for both H2/O 2 and H2/Air at 100% RH. However, SDS and MWCNTs demonstrates the lowest power density using air and oxygen as an oxidants at 100% RH.

  3. Improving the circular economy via hydrothermal processing of high-density waste plastics.

    Science.gov (United States)

    Helmer Pedersen, Thomas; Conti, Federica

    2017-10-01

    Rising environmental concerns on climate changes are causing an increasing attention on circular economies. The plastic economy, in particular, is in focus due to the accelerating consumption of plastics, mainly derived from virgin feedstock, combined with the lack of plastic recycling strategies. This work presents a novel outlook on the potential of using supercritical hydrothermal processing of waste plastic fractions for tertiary recycling. The study investigates hydrothermal processing of nine different, high-density types of plastics into original resin monomers and other value-added chemical compounds. The outlook presents conversion yields, carbon balances, and chemical details on the products obtained. It is found that all the investigated resins are prone to hydrothermal treatment, and that high yields of monomers and high value compounds (up to nearly 100%), suitable for chemicals and fuels applications, can be obtained. For instance, for polycarbonate, styrene-butadiene, poly(lactic acid), poly(ethylene terephthalate), and poly(butylene terephthalate), original monomeric compounds can be reclaimed for manufacturing new resins. The promising results presented demonstrate that hydrothermal processing of high-density plastics is a prospective technology for increasing the circularity of the plastic economy. Copyright © 2017 Elsevier Ltd. All rights reserved.

  4. A distributed current stimulator ASIC for high density neural stimulation.

    Science.gov (United States)

    Jeong Hoan Park; Chaebin Kim; Seung-Hee Ahn; Tae Mok Gwon; Joonsoo Jeong; Sang Beom Jun; Sung June Kim

    2016-08-01

    This paper presents a novel distributed neural stimulator scheme. Instead of a single stimulator ASIC in the package, multiple ASICs are embedded at each electrode site for stimulation with a high density electrode array. This distributed architecture enables the simplification of wiring between electrodes and stimulator ASIC that otherwise could become too complex as the number of electrode increases. The individual ASIC chip is designed to have a shared data bus that independently controls multiple stimulating channels. Therefore, the number of metal lines is determined by the distributed ASICs, not by the channel number. The function of current steering is also implemented within each ASIC in order to increase the effective number of channels via pseudo channel stimulation. Therefore, the chip area can be used more efficiently. The designed chip was fabricated with area of 0.3 mm2 using 0.18 μm BCDMOS process, and the bench-top test was also conducted to validate chip performance.

  5. Automatic Calibration of High Density Electric Muscle Stimulation

    DEFF Research Database (Denmark)

    Knibbe, Jarrod; Strohmeier, Paul; Boring, Sebastian

    2017-01-01

    . (2) EMS requires time consuming, expert calibration -- confining these interaction techniques to the lab. EMS arrays have been shown to increase stimulation resolution, but as calibration complexity increases exponentially as more electrodes are used, we require heuristics or automated procedures......Electric muscle stimulation (EMS) can enable mobile force feedback, support pedestrian navigation, or confer object affordances. To date, however, EMS is limited by two interlinked problems. (1) EMS is low resolution -- achieving only coarse movements and constraining opportunities for exploration...... for successful calibration. We explore the feasibility of using electromyography (EMG) to auto-calibrate high density EMS arrays. We determine regions of muscle activity during human-performed gestures, to inform stimulation patterns for EMS-performed gestures. We report on a study which shows that auto...

  6. A high-density lipoprotein-mediated drug delivery system.

    Science.gov (United States)

    Mo, Zhong-Cheng; Ren, Kun; Liu, Xing; Tang, Zhen-Li; Yi, Guang-Hui

    2016-11-15

    High-density lipoprotein (HDL) is a comparatively dense and small lipoprotein that can carry lipids as a multifunctional aggregate in plasma. Several studies have shown that increasing the levels or improving the functionality of HDL is a promising target for treating a wide variety of diseases. Among lipoproteins, HDL particles possess unique physicochemical properties, including naturally synthesized physiological components, amphipathic apolipoproteins, lipid-loading and hydrophobic agent-incorporating characteristics, specific protein-protein interactions, heterogeneity, nanoparticles, and smaller size. Recently, the feasibility and superiority of using HDL particles as drug delivery vehicles have been of great interest. In this review, we summarize the structure, constituents, biogenesis, remodeling, and reconstitution of HDL drug delivery systems, focusing on their delivery capability, characteristics, applications, manufacturing, and drug-loading and drug-targeting characteristics. Finally, the future prospects are presented regarding the clinical application and challenges of using HDL as a pharmacodelivery carrier. Copyright © 2016 Elsevier B.V. All rights reserved.

  7. Single-Readout High-Density Memristor Crossbar

    KAUST Repository

    Zidan, M. A.

    2016-01-07

    High-density memristor-crossbar architecture is a very promising technology for future computing systems. The simplicity of the gateless-crossbar structure is both its principal advantage and the source of undesired sneak-paths of current. This parasitic current could consume an enormous amount of energy and ruin the readout process. We introduce new adaptive-threshold readout techniques that utilize the locality and hierarchy properties of the computer-memory system to address the sneak-paths problem. The proposed methods require a single memory access per pixel for an array readout. Besides, the memristive crossbar consumes an order of magnitude less power than state-of-the-art readout techniques.

  8. High-Density Lipoproteins and the Immune System

    Directory of Open Access Journals (Sweden)

    Hidesuke Kaji

    2013-01-01

    Full Text Available High-density lipoprotein (HDL plays a major role in vasodilation and in the reduction of low-density lipoprotein (LDL oxidation, inflammation, apoptosis, thrombosis, and infection; however, HDL is now less functional in these roles under certain conditions. This paper focuses on HDL, its anti-inflammation behavior, and the mechanisms by which HDL interacts with components of the innate and adaptive immune systems. Genome-wide association studies (GWAS and proteomic studies have elucidated important molecules involved in the interaction between HDL and the immune system. An understanding of these mechanisms is expected to be useful for the prevention and treatment of chronic inflammation due to metabolic syndrome, atherosclerosis, or various autoimmune diseases.

  9. High-current discharge channel contraction in high density gas

    International Nuclear Information System (INIS)

    Rutberg, Ph. G.; Bogomaz, A. A.; Pinchuk, M. E.; Budin, A. V.; Leks, A. G.; Pozubenkov, A. A.

    2011-01-01

    Research results for discharges at current amplitudes of 0.5-1.6 MA and current rise rate of ∼10 10 A/s are presented. The discharge is performed in the hydrogen environment at the initial pressure of 5-35 MPa. Initiation is implemented by a wire explosion. The time length of the first half-period of the discharge current is 70-150 μs. Under such conditions, discharge channel contraction is observed; the contraction is followed by soft x-ray radiation. The phenomena are discussed, which are determined by high density of the gas surrounding the discharge channel. These phenomena are increase of the current critical value, where the channel contraction begins and growth of temperature in the axis region of the channel, where the initial density of the gas increases.

  10. Diquark Bose Condensates in High Density Matter and Instantons

    International Nuclear Information System (INIS)

    Rapp, R.; Shuryak, E.; Schaefer, T.; Velkovsky, M.

    1998-01-01

    Instantons lead to strong correlations between up and down quarks with spin zero and antisymmetric color wave functions. In cold and dense matter, n b >n c ≅1 fm -3 and T c ∼50 thinspthinspMeV, these pairs Bose condense, replacing the usual left-angle bar qq right-angle condensate and restoring chiral symmetry. At high density, the ground state is a color superconductor in which diquarks play the role of Cooper pairs. An interesting toy model is provided by QCD with two colors: it has a particle-antiparticle symmetry which relates left-angle bar qq right-angle and left-angle qq right-angle condensates. copyright 1998 The American Physical Society

  11. High-Density Near-Field Optical Disc Recording

    Science.gov (United States)

    Shinoda, Masataka; Saito, Kimihiro; Ishimoto, Tsutomu; Kondo, Takao; Nakaoki, Ariyoshi; Ide, Naoki; Furuki, Motohiro; Takeda, Minoru; Akiyama, Yuji; Shimouma, Takashi; Yamamoto, Masanobu

    2005-05-01

    We developed a high-density near-field optical recording disc system using a solid immersion lens. The near-field optical pick-up consists of a solid immersion lens with a numerical aperture of 1.84. The laser wavelength for recording is 405 nm. In order to realize the near-field optical recording disc, we used a phase-change recording media and a molded polycarbonate substrate. A clear eye pattern of 112 GB capacity with 160 nm track pitch and 50 nm bit length was observed. The equivalent areal density is 80.6 Gbit/in2. The bottom bit error rate of 3 tracks-write was 4.5× 10-5. The readout power margin and the recording power margin were ± 30.4% and ± 11.2%, respectively.

  12. EVALUATION OF SOME PLUM CULTIVARS IN A HIGH DENSITY SYSTEM

    Directory of Open Access Journals (Sweden)

    Madalina Butac

    2014-12-01

    Full Text Available Three plum cultivars bred in Romania (‘Carpatin’, ‘Centenar’ and ‘Tita’ were tested together with several standards (‘Cacanska Rodna’ and ‘Stanley’ in a high density experimental orchard established at Pitesti - Maracineni in the spring of 2009, with spacing 4 x 2.25 m. Trees were trained as spindles, grafted on ‘Saint Julian’ rootstock. In the orchard the following characteristics were evaluated: tree vigour based upon measuring of trunk-diameter, yields in kg/tree, time of fruit ripening and basic parameters of fruit quality. All Romanian varieties were characterized by earliness and large fruit, but production was relatively small. Instead, foreign varieties were characterized by high productivity in the 4th year after planting.

  13. The stability of the High-Density Z-Pinch

    International Nuclear Information System (INIS)

    Glasser, A.H.; Nebel, R.A.

    1989-01-01

    Fiber-initiated High Density Z-Pinches at Los Alamos, NRL, and Karlsruhe have shown anomalously good stability. Kink modes are never seen, and sausage modes are at least delayed until late in the discharge. The success of these devices in reaching fusion conditions may depend on maintaining and understanding this anomalous stability. We have developed two numerical methods to study the stability in the regime where fluid theory is valid. While our methods are applicable to all modes, we will describe them only for the m = 0 sausage mode. The appearance of sausage modes late in the discharge and the total absence of kink modes suggest that an understanding of sausage modes is more urgent, and it is also simpler. 14 refs., 8 figs

  14. The stability of the high-density z-pinch

    International Nuclear Information System (INIS)

    Glasser, A.H.; Nebel, R.A.

    1989-01-01

    Fiber-initiated High Density Z-Pinches at Los Alamos, NRL, and Karlsruhe have shown anomalously good stability. Kink modes are never seen, and sausage modes are at least delayed until late in the discharge. The success of these devices in reaching fusion conditions may depend on maintaining and understanding this anomalous stability. We have developed two numerical methods to study the stability in the regime where fluid theory is valid. While our methods are applicable to all modes, we will describe them only for the m=0 sausage mode. The appearance of sausage modes late in the discharge and the total absence of kink modes suggest that an understanding of sausage modes is more urgent, and it is also simpler

  15. Corrosion of high-density sintered tungsten alloys. Part 2

    International Nuclear Information System (INIS)

    Batten, J.J.; Moore, B.T.

    1988-12-01

    The behaviour of four high-density sintered tungsten alloys has been evluated and compared with that of pure tungsten. Rates of corrosion during the cyclic humidity and the salt mist tests were ascertained from weight loss measurements. Insight into the corrosion mechanism was gained from the nature of the corrosion products and an examination of the corroded surfaces. In the tests, the alloy 95% W, 2.5% Ni, 1.5% Fe was the most corrosion resistant. The data showed that copper as an alloying element accelerates corrosion of tungsten alloys. Both attack on the tungsten particles and the binder phase were observed together with tungsten grain loss. 6 refs., 3 tabs.,

  16. High density high-TC ceramic superconductors by hot pressing

    International Nuclear Information System (INIS)

    Mak, S.; Chaklader, A.C.D.

    1989-01-01

    High density and high T C superconductor specimens, YBa 2 Cu 3 O x , have been produced by hot-pressing. The factors studied are the effect of hot pressing on the density, the oxygen stoichiometry, the crystal structure, and the critical temperature. Hot pressing followed by heat treatment increased the density of the specimen to 93%. The hot pressing itself did not significantly affect the oxygen content in the specimen, and although the crystal structure appeared to be orthorhombic, the specimens were not superconducting above liquid nitrogen temperature. The superconductivity was restored after head treatment in oxygen. The highest critical temperature (T C ) of the hot pressed pellets was 82K, which was slightly lower than the T C that could be obtained with the cold pressed/sintered pellets. (6 refs., 5 figs., tab.)

  17. Evaluation of the radiation resistance of high-density polyethylene

    International Nuclear Information System (INIS)

    Dougherty, D.R.; Adams, J.W.; Barletta, R.R.

    1984-03-01

    Mechanical tests following gamma irradiation and creep tests during irradiation have been conducted on high-density polyethylene (HDPE) to provide data to help assess the adequacy of this material for use in high integrity containers (HICs). Two types of HDPE, a highly cross-linked rotationally molded material and a non-cross-linked blow molded material, were used in these tests. Gamma-ray irradiations were performed at several dose rates in environments of air, Barnwell and Hanford backfill soils, and ion-exchange resins. The results of tensile and bend tests on these materials following irradiation are presented along with results on creep during irradiation. 8 references, 9 figures, 2 tables

  18. The Pulsed High Density Experiment (PHDX) Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Slough, John P. [Univ. of Washington, Seattle, WA (United States); Andreason, Samuel [Univ. of Washington, Seattle, WA (United States)

    2017-04-27

    The purpose of this paper is to present the conclusions that can be drawn from the Field Reversed Configuration (FRC) formation experiments conducted on the Pulsed High Density experiment (PHD) at the University of Washington. The experiment is ongoing. The experimental goal for this first stage of PHD was to generate a stable, high flux (>10 mWb), high energy (>10 KJ) target FRC. Such results would be adequate as a starting point for several later experiments. This work focuses on experimental implementation and the results of the first four month run. Difficulties were encountered due to the initial on-axis plasma ionization source. Flux trapping with this ionization source acting alone was insufficient to accomplish experimental objectives. Additional ionization methods were utilized to overcome this difficulty. A more ideal plasma source layout is suggested and will be explored during a forthcoming work.

  19. High-Density Stacked Ru Nanocrystals for Nonvolatile Memory Application

    International Nuclear Information System (INIS)

    Ping, Mao; Zhi-Gang, Zhang; Li-Yang, Pan; Jun, Xu; Pei-Yi, Chen

    2009-01-01

    Stacked ruthenium (Ru) nanocrystals (NCs) are formed by rapid thermal annealing for the whole gate stacks and embedded in memory structure, which is compatible with conventional CMOS technology. Ru NCs with high density (3 × 10 12 cm −2 ), small size (2–4 nm) and good uniformity both in aerial distribution and morphology are formed. Attributed to the higher surface trap density, a memory window of 5.2 V is obtained with stacked Ru NCs in comparison to that of 3.5 V with single-layer samples. The stacked Ru NCs device also exhibits much better retention performance because of Coulomb blockade and vertical uniformity between stacked Ru NCs

  20. Single-Readout High-Density Memristor Crossbar

    KAUST Repository

    Zidan, M. A.; Omran, Hesham; Naous, Rawan; Salem, Ahmed Sultan; Fahmy, H. A. H.; Lu, W. D.; Salama, Khaled N.

    2016-01-01

    High-density memristor-crossbar architecture is a very promising technology for future computing systems. The simplicity of the gateless-crossbar structure is both its principal advantage and the source of undesired sneak-paths of current. This parasitic current could consume an enormous amount of energy and ruin the readout process. We introduce new adaptive-threshold readout techniques that utilize the locality and hierarchy properties of the computer-memory system to address the sneak-paths problem. The proposed methods require a single memory access per pixel for an array readout. Besides, the memristive crossbar consumes an order of magnitude less power than state-of-the-art readout techniques.

  1. High density submicron magnetoresistive random access memory (invited)

    Science.gov (United States)

    Tehrani, S.; Chen, E.; Durlam, M.; DeHerrera, M.; Slaughter, J. M.; Shi, J.; Kerszykowski, G.

    1999-04-01

    Various giant magnetoresistance material structures were patterned and studied for their potential as memory elements. The preferred memory element, based on pseudo-spin valve structures, was designed with two magnetic stacks (NiFeCo/CoFe) of different thickness with Cu as an interlayer. The difference in thickness results in dissimilar switching fields due to the shape anisotropy at deep submicron dimensions. It was found that a lower switching current can be achieved when the bits have a word line that wraps around the bit 1.5 times. Submicron memory elements integrated with complementary metal-oxide-semiconductor (CMOS) transistors maintained their characteristics and no degradation to the CMOS devices was observed. Selectivity between memory elements in high-density arrays was demonstrated.

  2. The alterations in high density polyethylene properties with gamma irradiation

    Science.gov (United States)

    Zaki, M. F.; Elshaer, Y. H.; Taha, Doaa. H.

    2017-10-01

    In the present investigation, high density polyethylene (HDPE) polymer has been used to study the alterations in its properties under gamma-irradiation. Physico-chemical properties have been investigated with different spectroscopy techniques, Fourier Transform Infrared spectroscopy (FTIR), X-ray diffraction (XRD), biocompatibility properties, as well as, mechanical properties change. The FT-IR analysis shows the formation of new band at 1716 cm-1 that is attributed to the oxidation of irradiated polymer chains, which is due to the formation of carbonyl groups (C˭O). XRD patterns show that a decrease in the crystallite size and increase in the Full Width at Half Maximum (FWHM). This means that the crystallinity of irradiated samples is decreased with increase in gamma dose. The contact angle measurements show an increase in the surface free energy as the gamma irradiation increases. The measurements of mechanical properties of irradiated HDPE samples were discussed.

  3. Relaxation Time of High-Density Amorphous Ice

    Science.gov (United States)

    Handle, Philip H.; Seidl, Markus; Loerting, Thomas

    2012-06-01

    Amorphous water plays a fundamental role in astrophysics, cryoelectron microscopy, hydration of matter, and our understanding of anomalous liquid water properties. Yet, the characteristics of the relaxation processes taking place in high-density amorphous ice (HDA) are unknown. We here reveal that the relaxation processes in HDA at 110-135 K at 0.1-0.2 GPa are of collective and global nature, resembling the alpha relaxation in glassy material. Measured relaxation times suggest liquid-like relaxation characteristics in the vicinity of the crystallization temperature at 145 K. By carefully relaxing pressurized HDA for several hours at 135 K, we produce a state that is closer to the ideal glass state than all HDA states discussed so far in literature.

  4. High-density amorphous ice: A path-integral simulation

    Science.gov (United States)

    Herrero, Carlos P.; Ramírez, Rafael

    2012-09-01

    Structural and thermodynamic properties of high-density amorphous (HDA) ice have been studied by path-integral molecular dynamics simulations in the isothermal-isobaric ensemble. Interatomic interactions were modeled by using the effective q-TIP4P/F potential for flexible water. Quantum nuclear motion is found to affect several observable properties of the amorphous solid. At low temperature (T = 50 K) the molar volume of HDA ice is found to increase by 6%, and the intramolecular O-H distance rises by 1.4% due to quantum motion. Peaks in the radial distribution function of HDA ice are broadened with respect to their classical expectancy. The bulk modulus, B, is found to rise linearly with the pressure, with a slope ∂B/∂P = 7.1. Our results are compared with those derived earlier from classical and path-integral simulations of HDA ice. We discuss similarities and discrepancies with those earlier simulations.

  5. Characterization of High Density Concrete by Ultrasonic Goniometer

    International Nuclear Information System (INIS)

    Suhairy Sani; Mohamad Pauzi Ismail; Noor Azreen Masenwat; Nasharuddin Isa; Mohamad Haniza Mahmud

    2014-01-01

    This paper described the results of ultrasonic goniometer measurements on concrete containing hematite. Local hematite stones were used as aggregates to produce high density concrete for application in X-and gamma shielding. Concrete cube samples (150 mm x 150 mm x 150 mm) containing hematite as coarse aggregates were prepared by changing mix ratio, water to cement ratio (w/ c) and types of fine aggregate. All samples were cured in water for 7 days. After 28 days of casting, the concrete cubes were then cut into small size of about 10 mm x 20 mm x 30 mm so that it can be fitted into goniometer specimen holder. From this measurement, longitudinal, shear and surface Rayleigh waves in the concrete can be determined. The measurement results are explained and discussed. (author)

  6. Microelectromechanical high-density energy storage/rapid release system

    Science.gov (United States)

    Rodgers, M. Steven; Allen, James J.; Meeks, Kent D.; Jensen, Brian D.; Miller, Samuel L.

    1999-08-01

    One highly desirable characteristic of electrostatically driven microelectromechanical systems (MEMS) is that they consume very little power. The corresponding drawback is that the force they produce may be inadequate for many applications. It has previously been demonstrated that gear reduction units or microtransmissions can substantially increase the torque generated by microengines. Operating speed, however, is also reduced by the transmission gear ratio. Some applications require both high speed and high force. If this output is only required for a limited period of time, then energy could be stored in a mechanical system and rapidly released upon demand. We have designed, fabricated, and demonstrated a high-density energy storage/rapid release system that accomplishes this task. Built using a 5-level surface micromachining technology, the assembly closely resembles a medieval crossbow. Energy releases on the order of tens of nanojoules have already been demonstrated, and significantly higher energy systems are under development.

  7. Preparations of high density (Th,U)O2 pellets

    International Nuclear Information System (INIS)

    Akabori, Mitsuo; Ikawa, Katsuichi

    1986-07-01

    Preparations of high density and homogeneous (Th,U)O 2 pellets by a powder metallurgy method were examined. (Th,U)O 2 powders were prepared by calcining coprecipitates of ammonium uranate and thorium hydroxide derived from nitrates and mixed sols, and by calcining mixed oxalates precipitated from nitrates. (Th,U)O 2 pellets were characterized with respect to sinterability, lattice parameter, microstructure, homogeneity and stoichiometry. Sintering atmospheres had a significant effect upon all the properties of the derived pellets. The sinterability of (Th,U)O 2 was most favourable in oxidizing and reducing atmospheres for ThO 2 -rich and UO 2 -rich compositions, respectively, and can be enhanced by presence of water vapour in sintering atmospheres. In addition, highly homogeneous (Th,U)O 2 pellets with 99 % in theoretical density were derived from the sol powders. (author)

  8. Neutron shielding properties of a borated high-density glass

    Directory of Open Access Journals (Sweden)

    Saeed Aly Abdallah

    2017-01-01

    Full Text Available The neutron shielding properties of a borated high density glass system was characterized experimentally. The total removal macroscopic cross-section of fast neutrons, slow neutrons as well as the linear attenuation coefficient of total gamma rays, primary in addition to secondary, were measured experimentally under good geometric condition to characterize the attenuation properties of (75-x B2O3-1Li2O-5MgO-5ZnO-14Na2O-xBaO glassy system. Slabs of different thicknesses from the investigated glass system were exposed to a collimated beam of neutrons emitted from 252Cf and 241Am-Be neutron sources in order to measure the attenuation properties of fast and slow neutrons as well as total gamma rays. Results confirmed that barium borate glass was suitable for practical use in the field of radiation shielding.

  9. Biominetic High Density Lipoproteins for the Delivery of Therapeutic Oligonucleotides

    Science.gov (United States)

    Tripathy, Sushant

    Advances in nanotechnology have brought about novel inorganic and hybrid nanoparticles with unique physico-chemical properties that make them suitable for a broad range of applications---from nano-circuitry to drug delivery. A significant part of those advancements have led to ground-breaking discoveries that have changed the approaches to formulation of therapeutics against diseases, such as cancer. Now-a-days the focus does not lie solely on finding a candidate small-molecule therapeutic with minimal adverse effects, but researchers are looking up to nanoparticles to improve biodistribution and biocompatibility profile of clinically proven therapeutics. The plethora of conjugation chemistries offered by currently extant inorganic nanoparticles have, in recent years, led to great leaps in the field of biomimicry---a modality that promises high biocompatibility. Further, in the pursuit of highly specific therapeutic molecules, researchers have turned to silencing oligonucleotides and some have already brought together the strengths of nanoparticles and silencing oligonucleotides in search of an efficacious therapy for cancer with minimal adverse effects. This dissertation work focuses on such a biomimetic platform---a gold nanoparticle based high density lipoprotein biomimetic (HDL NP), for the delivery of therapeutic oligonucleotides. The first chapter of this body of work introduces the molecular target of the silencing oligonucleotides---VEGFR2, and its role in the progression of solid tumor cancers. The background information also covers important aspects of natural high density lipoproteins (HDL), especially their innate capacity to bind and deliver exogenous and endogenous silencing oligonucleotides to tissues that express their high affinity receptor SRB1. We subsequently describe the synthesis of the biomimetic HDL NP and its oligonucleotide conjugates, and establish their biocompatibility. Further on, experimental data demonstrate the efficacy of silencing

  10. Rethinking reverse cholesterol transport and dysfunctional high-density lipoproteins.

    Science.gov (United States)

    Gillard, Baiba K; Rosales, Corina; Xu, Bingqing; Gotto, Antonio M; Pownall, Henry J

    2018-04-12

    Human plasma high-density lipoprotein cholesterol concentrations are a negative risk factor for atherosclerosis-linked cardiovascular disease. Pharmacological attempts to reduce atherosclerotic cardiovascular disease by increasing plasma high-density lipoprotein cholesterol have been disappointing so that recent research has shifted from HDL quantity to HDL quality, that is, functional vs dysfunctional HDL. HDL has varying degrees of dysfunction reflected in impaired reverse cholesterol transport (RCT). In the context of atheroprotection, RCT occurs by 2 mechanisms: one is the well-known trans-hepatic pathway comprising macrophage free cholesterol (FC) efflux, which produces early forms of FC-rich nascent HDL (nHDL). Lecithin:cholesterol acyltransferase converts HDL-FC to HDL-cholesteryl ester while converting nHDL from a disc to a mature spherical HDL, which transfers its cholesteryl ester to the hepatic HDL receptor, scavenger receptor B1 for uptake, conversion to bile salts, or transfer to the intestine for excretion. Although widely cited, current evidence suggests that this is a minor pathway and that most HDL-FC and nHDL-FC rapidly transfer directly to the liver independent of lecithin:cholesterol acyltransferase activity. A small fraction of plasma HDL-FC enters the trans-intestinal efflux pathway comprising direct FC transfer to the intestine. SR-B1 -/- mice, which have impaired trans-hepatic FC transport, are characterized by high plasma levels of a dysfunctional FC-rich HDL that increases plasma FC bioavailability in a way that produces whole-body hypercholesterolemia and multiple pathologies. The design of future therapeutic strategies to improve RCT will have to be formulated in the context of these dual RCT mechanisms and the role of FC bioavailability. Copyright © 2018 National Lipid Association. Published by Elsevier Inc. All rights reserved.

  11. Properties of matter at ultra-high densities

    International Nuclear Information System (INIS)

    Banerjee, B.; Chitre, S.M.

    1975-01-01

    The recent discovery of pulsars and their subsequent identification with neutron stars has given a great impetus to the study of the behaviour of matter at ultra high densities. The object of these studies is to calculate the equation of state as a function of density. In this paper, the properties of electrically neutral, cold (T=0) matter at unusually high densities has been reviewed. The physics of the equation of state of such matter divides quite naturally in four density ranges. (i) At the very lowest densities the state of minimum energy is a lattice of 56 Fe atoms. This state persists upto 10 7 g/cm 3 . (ii) In the next density region the nuclei at the lattice sites become neutron rich because the high electron Fermi energy makes inverse beta decay possible. (iii) At a density 4.3 x 10 11 the nuclei become so neutron rich that the neutrons start 'dripping' out of the nuclei and form a gas. This density range is characterised by large, neutron-rich nuclei immersed in a neutron gas. (iv) At a density 2.4 x 10 14 g/cm 3 , the nuclei disappear and a fluid of uniform neutron matter with a small percentage of protons and electrons results. The above four density ranges have been discussed in detail as the equation of state is now well established upto the nuclear density 3 x 10 14 g/cm 3 . The problems of extending the equation of state beyond this density are also touched upon. (author)

  12. Pulsed high-density plasmas for advanced dry etching processes

    International Nuclear Information System (INIS)

    Banna, Samer; Agarwal, Ankur; Cunge, Gilles; Darnon, Maxime; Pargon, Erwine; Joubert, Olivier

    2012-01-01

    Plasma etching processes at the 22 nm technology node and below will have to satisfy multiple stringent scaling requirements of microelectronics fabrication. To satisfy these requirements simultaneously, significant improvements in controlling key plasma parameters are essential. Pulsed plasmas exhibit considerable potential to meet the majority of the scaling challenges, while leveraging the broad expertise developed over the years in conventional continuous wave plasma processing. Comprehending the underlying physics and etching mechanisms in pulsed plasma operation is, however, a complex undertaking; hence the full potential of this strategy has not yet been realized. In this review paper, we first address the general potential of pulsed plasmas for plasma etching processes followed by the dynamics of pulsed plasmas in conventional high-density plasma reactors. The authors reviewed more than 30 years of academic research on pulsed plasmas for microelectronics processing, primarily for silicon and conductor etch applications, highlighting the potential benefits to date and challenges in extending the technology for mass-production. Schemes such as source pulsing, bias pulsing, synchronous pulsing, and others in conventional high-density plasma reactors used in the semiconductor industry have demonstrated greater flexibility in controlling critical plasma parameters such as ion and radical densities, ion energies, and electron temperature. Specifically, plasma pulsing allows for independent control of ion flux and neutral radicals flux to the wafer, which is key to eliminating several feature profile distortions at the nanometer scale. However, such flexibility might also introduce some difficulty in developing new etching processes based on pulsed plasmas. Therefore, the main characteristics of continuous wave plasmas and different pulsing schemes are compared to provide guidelines for implementing different schemes in advanced plasma etching processes based on

  13. The glass transition in high-density amorphous ice.

    Science.gov (United States)

    Loerting, Thomas; Fuentes-Landete, Violeta; Handle, Philip H; Seidl, Markus; Amann-Winkel, Katrin; Gainaru, Catalin; Böhmer, Roland

    2015-01-01

    There has been a long controversy regarding the glass transition in low-density amorphous ice (LDA). The central question is whether or not it transforms to an ultraviscous liquid state above 136 K at ambient pressure prior to crystallization. Currently, the most widespread interpretation of the experimental findings is in terms of a transformation to a superstrong liquid above 136 K. In the last decade some work has also been devoted to the study of the glass transition in high-density amorphous ice (HDA) which is in the focus of the present review. At ambient pressure HDA is metastable against both ice I and LDA, whereas at > 0.2 GPa HDA is no longer metastable against LDA, but merely against high-pressure forms of crystalline ice. The first experimental observation interpreted as the glass transition of HDA was made using in situ methods by Mishima, who reported a glass transition temperature T g of 160 K at 0.40 GPa. Soon thereafter Andersson and Inaba reported a much lower glass transition temperature of 122 K at 1.0 GPa. Based on the pressure dependence of HDA's T g measured in Innsbruck, we suggest that they were in fact probing the distinct glass transition of very high-density amorphous ice (VHDA). Very recently the glass transition in HDA was also observed at ambient pressure at 116 K. That is, LDA and HDA show two distinct glass transitions, clearly separated by about 20 K at ambient pressure. In summary, this suggests that three glass transition lines can be defined in the p-T plane for LDA, HDA, and VHDA.

  14. Propofol Anesthesia and Sleep: A High-Density EEG Study

    Science.gov (United States)

    Murphy, Michael; Bruno, Marie-Aurelie; Riedner, Brady A.; Boveroux, Pierre; Noirhomme, Quentin; Landsness, Eric C.; Brichant, Jean-Francois; Phillips, Christophe; Massimini, Marcello; Laureys, Steven; Tononi, Giulio; Boly, Melanie

    2011-01-01

    Study Objectives: The electrophysiological correlates of anesthetic sedation remain poorly understood. We used high-density electroencephalography (hd-EEG) and source modeling to investigate the cortical processes underlying propofol anesthesia and compare them to sleep. Design: 256-channel EEG recordings in humans during propofol anesthesia. Setting: Hospital operating room. Patients or Participants: 8 healthy subjects (4 males) Interventions: N/A Measurements and Results: Initially, propofol induced increases in EEG power from 12–25 Hz. Loss of consciousness (LOC) was accompanied by the appearance of EEG slow waves that resembled the slow waves of NREM sleep. We compared slow waves in propofol to slow waves recorded during natural sleep and found that both populations of waves share similar cortical origins and preferentially propagate along the mesial components of the default network. However, propofol slow waves were spatially blurred compared to sleep slow waves and failed to effectively entrain spindle activity. Propofol also caused an increase in gamma (25–40 Hz) power that persisted throughout LOC. Source modeling analysis showed that this increase in gamma power originated from the anterior and posterior cingulate cortices. During LOC, we found increased gamma functional connectivity between these regions compared to the wakefulness. Conclusions: Propofol anesthesia is a sleep-like state and slow waves are associated with diminished consciousness even in the presence of high gamma activity. Citation: Murphy M; Bruno MA; Riedner BA; Boveroux P; Noirhomme Q; Landsness EC; Brichant JF; Phillips C; Massimini M; Laureys S; Tononi G; Boly M. Propofol anesthesia and sleep: a high-density EEG study. SLEEP 2011;34(3):283-291. PMID:21358845

  15. Uranium hexafluoride reconversion used for dispersion fuel elements fabrication for IEAR-1/SP reactor; Reconversao de hexafluoreto de uranio para a fabricacao de combustiveis na forma de dispersoes para o reator IEA-R1/SP

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, E.F. Urano de; Lainetti, P.E.; Gomes, R.P. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    1996-07-01

    In this paper are described the main chemical process employed in the Chemical Processes Division of the Fuel Technology Department - IPEN for conversion of enriched UF{sub 6} in ammonium diuranate - DUA and uranium tetrafluoride - UF{sub 4}. These activities have assured the continuity of fuel elements production at IPEN since 1984. The uranium recovery from scraps of the fuel elements production and the purification processes are also described. Those compounds are important intermediate products in the fabrication routine and in development dispersed fuel elements with higher uranium loading for IEA{sub R}1 research reactor power increase program. (author)

  16. Gas dispersion concentration of trace inorganic contaminants from fuel gas and analysis using head-column field-amplified sample stacking capillary electrophoresis.

    Science.gov (United States)

    Yang, Jianmin; Li, Hai-Fang; Li, Meilan; Lin, Jin-Ming

    2012-08-21

    The presence of inorganic elements in fuel gas generally accelerates the corrosion and depletion of materials used in the fuel gas industry, and even leads to serious accidents. For identification of existing trace inorganic contaminants in fuel gas in a portable way, a highly efficient gas-liquid sampling collection system based on gas dispersion concentration is introduced in this work. Using the constructed dual path gas-liquid collection setup, inorganic cations and anions were simultaneously collected from real liquefied petroleum gas (LPG) and analyzed by capillary electrophoresis (CE) with indirect UV absorbance detection. The head-column field-amplified sample stacking technique was applied to improve the detection limits to 2-25 ng mL(-1). The developed collection and analytical methods have successfully determined existing inorganic contaminants in a real LPG sample in the range of 4.59-138.69 μg m(-3). The recoveries of cations and anions with spiked LPG samples were between 83.98 and 105.63%, and the relative standard deviations (RSDs) were less than 7.19%.

  17. 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.

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

    International Nuclear Information System (INIS)

    Ugajin, Mitsuhiro; Itoh, Akinori; Akabori, Mitsuo

    1993-06-01

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

  19. Highly dispersed TaOx nanoparticles prepared by electrodeposition as oxygen reduction electrocatalysts for polymer electrolyte fuel cells

    KAUST Repository

    Seo, Jeongsuk; Zhao, Lan; Cha, Dong Kyu; Takanabe, Kazuhiro; Katayama, Masao; Kubota, Jun; Domen, Kazunari

    2013-01-01

    for the oxygen reduction reaction (ORR) in polymer electrolyte fuel cells (PEFCs). Electrodeposition conditions of Ta complexes and subsequent various heat treatments for the deposited TaOx were examined for the best performance of the ORR. TaOx particles

  20. 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.

  1. Effects of gamma irradiation on polypropylene, polypropylene + high density polyethylene and polypropylene + high density polyethylene + wood flour

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, J.; Albano, C.; Davidson, E.; Poleo, R. [Universidad Central de Venezuela, Caracas (Venezuela). Escuela de Quimica; Gonzalez, J.; Ichazo, M. [Universidad Simon Bolivar, Dept. de Mecanica, Caracas (Venezuela); Chipara, M. [Research Institute for Electrotechnics, Bucharest (Romania)

    2001-04-01

    The effect of the gamma-irradiation on the mechanical properties of the composites, Polypropylene (PP), PP+high density Polyethylene (HDPE), PP+ HDPE+wood flour, where HDPE is virgin and recycled, was studied. This paper discusses the behavior of the composites after exposure to various doses of gamma irradiation (1-7 MRads) in the presence of oxygen. The dependence of mechanical properties on the integral dose for a constant dose rate of 0.48 MRads/h confirms the influence of the irradiation. Strong effects on the elongation at break and break strength is noticed. The mathematical analysis suggests for the PP+r-HDPE a bimolecular process of the elongation at break. On the order hand, for the PP+HDPE a complex process is represented for a three exponential equation. (orig.)

  2. Effects of gamma irradiation on polypropylene, polypropylene + high density polyethylene and polypropylene + high density polyethylene + wood flour

    International Nuclear Information System (INIS)

    Reyes, J.; Albano, C.; Davidson, E.; Poleo, R.; Chipara, M.

    2001-01-01

    The effect of the gamma-irradiation on the mechanical properties of the composites, Polypropylene (PP), PP+high density Polyethylene (HDPE), PP+ HDPE+wood flour, where HDPE is virgin and recycled, was studied. This paper discusses the behavior of the composites after exposure to various doses of gamma irradiation (1-7 MRads) in the presence of oxygen. The dependence of mechanical properties on the integral dose for a constant dose rate of 0.48 MRads/h confirms the influence of the irradiation. Strong effects on the elongation at break and break strength is noticed. The mathematical analysis suggests for the PP+r-HDPE a bimolecular process of the elongation at break. On the order hand, for the PP+HDPE a complex process is represented for a three exponential equation. (orig.)

  3. Effect of Modified and Nonmodified Carbon Nanotubes on the Rheological Behavior of High Density Polyethylene Nanocomposite

    Directory of Open Access Journals (Sweden)

    Adewunmi A. Ahmad

    2013-01-01

    Full Text Available This paper reports the results of studies on the rheological behavior of nanocomposites of high density polyethylene (HDPE with pristine multiwall carbon nanotubes (CNT as well as phenol and 1-octadecanol (C18 functionalized CNT at 1, 2, 3, 4, 5, and 7 wt% loading. The viscosity reduction at 1 wt% CNT follows the order, pristine CNT < phenol functionalized CNT < C18 functionalized CNT. As the filler loading increases from 1 to 2, 3, and 4 wt%, neat HDPE and filled HDPE systems show similar moduli and viscosity, particularly in the low frequency region. As the filler loading increases further to 5 and 7 wt%, the viscosity and moduli become greater than the neat HDPE. The storage modulus, tan, and the Cole-Cole plots show that CNT network formation occurs at higher CNT loading. The critical CNT loading or the rheological percolation threshold, where network formation occurs is found to be strongly dependant on the functionalization of CNT. For pristine CNT, the rheological percolation threshold is around 4 wt%, but for functionalized CNT it is around 7 wt%. The surface morphologies of CNT and functionalized CNT at 1 wt% loading showed good dispersion while at 7 wt% loading, dispersion was also achieved, but there are few regions with agglomeration of CNT.

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

    Directory of Open Access Journals (Sweden)

    ALEKSEY. L. IZHUTOV

    2013-12-01

    The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; the mini-rods were irradiated to an average burnup of ∼ 85%235U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  5. Wake High-Density Electroencephalographic Spatiospectral Signatures of Insomnia

    Science.gov (United States)

    Colombo, Michele A.; Ramautar, Jennifer R.; Wei, Yishul; Gomez-Herrero, Germán; Stoffers, Diederick; Wassing, Rick; Benjamins, Jeroen S.; Tagliazucchi, Enzo; van der Werf, Ysbrand D.; Cajochen, Christian; Van Someren, Eus J.W.

    2016-01-01

    Study Objectives: Although daytime complaints are a defining characteristic of insomnia, most EEG studies evaluated sleep only. We used high-density electroencephalography to investigate wake resting state oscillations characteristic of insomnia disorder (ID) at a fine-grained spatiospectral resolution. Methods: A case-control assessment during eyes open (EO) and eyes closed (EC) was performed in a laboratory for human physiology. Participants (n = 94, 74 female, 21–70 y) were recruited through www.sleepregistry.nl: 51 with ID, according to DSM-5 and 43 matched controls. Exclusion criteria were any somatic, neurological or psychiatric condition. Group differences in the spectral power topographies across multiple frequencies (1.5 to 40 Hz) were evaluated using permutation-based inference with Threshold-Free Cluster-Enhancement, to correct for multiple comparisons. Results: As compared to controls, participants with ID showed less power in a narrow upper alpha band (11–12.7 Hz, peak: 11.7 Hz) over bilateral frontal and left temporal regions during EO, and more power in a broad beta frequency range (16.3–40 Hz, peak: 19 Hz) globally during EC. Source estimates suggested global rather than cortically localized group differences. Conclusions: The widespread high power in a broad beta band reported previously during sleep in insomnia is present as well during eyes closed wakefulness, suggestive of a round-the-clock hyperarousal. Low power in the upper alpha band during eyes open is consistent with low cortical inhibition and attentional filtering. The fine-grained HD-EEG findings suggest that, while more feasible than PSG, wake EEG of short duration with a few well-chosen electrodes and frequency bands, can provide valuable features of insomnia. Citation: Colombo MA, Ramautar JR, Wei Y, Gomez-Herrero G, Stoffers D, Wassing R, Benjamins JS, Tagliazucchi E, van der Werf YD, Cajochen C, Van Someren EJW. Wake high-density electroencephalographic spatiospectral

  6. Coextrusion applied to the construction of fuel elements in solid or powder form; Coextrusion appliquee a la realisation d'elements combustibles massifs ou disperses

    Energy Technology Data Exchange (ETDEWEB)

    Montagne, R; Meny, L [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    In this issue is described, in the first part, a realisation process of fuel elements for nuclear reactors. A contact as good as possible is achieved between the fuel and the can by both elements simultaneous extrusion. In this way a real weld is work out between the two metals. This weld can be improved by a thermic treatment that bring a diffusion. In this article are described the test carried out on these co extruded elements. In the second part, the fabrication of dispersed fuel elements studied: a 30 per cent weight U uranium-aluminium alloy is used, valuable with 20 per cent enriched uranium. The dimensions of the fuel element have been fixed at: external diameter: 30 mm, internal diameter: 24 mm, length of the core: 300 mm, thickness of the can: 0,4 mm. The method of fabrication is pressing of the mixed uranium and aluminium powders in an aluminium can, and extrusion at 500 deg. C.; one end is directly canned by extrusion and the other by welding of an aluminium plug. The results of the first test are described. (author) [French] Dans ce memoire est decrit, en premiere partie, un procede d'obtention d'elements combustibles pour reacteurs atomiques. Un contact aussi bon que possible est realise entre le combustible et la gaine grace au filage simultane des deux elements. Une veritable soudure est ainsi realisee entre les deux metaux. Celle-ci peut ensuite etre amelioree par un traitement thermique provoquant une diffusion. Les essais effectues sur ces elements coextrudes sont decrits dans cet article. Dans une deuxieme partie, la fabrication d'elements combustibles disperses est etudiee, avec un alliage uranium-aluminium a 30 pour cent en poids d'uranium, valable pour un enrichissement de l'uranium de 20 pour cent. Les dimensions des elements combustibles ont ete fixees a: diametre exterieur: 30 mm, diametre interieur: 24 mm, longueur du noyau: 300 mm, epaisseur de la gaine: 0,4 mm. La methode de fabrication est le pressage dans un pot en aluminium du

  7. Atmospheric Dispersion Modeling of 137Cs generated from Nuclear Spent Fuel under Hypothetic Accidental Condition in the BNPP Area

    Science.gov (United States)

    Lee, Jongkuk; Lee, Kwan-Hee; Yook, Daesik; Kim, Sung Il; Lee, Byung Soo

    2016-04-01

    This study presents the results of atmosphere dispersion modeling using CALPUFF code that are based on computational simulation to evaluate the environmental characteristics of the Barakah nuclear power plant (BNPP) in west area of UAE. According to meteorological data analysis (2012~2013), the winds from the north(7.68%) and west(9.05%) including NNW(41.63%), NW(28.55%), and WNW(6.31%) winds accounted for more than 90% of the wind directions. East(0.2%) and south(0.6%) direction wind, including ESE(0.31%), SE(0.38%), and SSE(0.38%) were rarely distributed during the simulation period. Seasonal effects were not showed. However, a discrepancy in the tendency between daytime and night-time was observed. Approximately 87% of the wind speed was distributed below 5.4m/s (17%, 47% and 23% between the speeds of 0.5-1.8m/s 1.8-3.3m/s and 3.3-5.4m/s, respectively) during the annual period. Seasonal wind speed distribution results presented very similar pattern of annual distribution. Wind speed distribution of day and night, on the other hand, had a discrepancy with annual modeling results than seasonal distribution in some sections. The results for high wind speed (more than 10.8m/s) showed that this wind blew from the west. This high wind speed is known locally as the 'Shamal', which occurs rarely, lasting one or two days with the strongest winds experienced in association with gust fronts and thunderstorms. Six variations of cesium-137 (137Cs) dispersion test were simulated under hypothetic severe accidental condition. The 137Cs dispersion was strongly influenced by the direction and speed of the main wind. From the test cases, east-south area of the BNPP site was mainly influenced by 137Cs dispersion. A virtual receptor was set and calculated for observation of the 137Cs movement and accumulation. Surface roughness tests were performed for the analysis of topographic conditions. According to the surface condition, there are various surface roughness length. Four types

  8. BASIC program to compute uranium density and void volume fraction in laboratory-scale uranium silicide aluminum dispersion plate-type fuel

    International Nuclear Information System (INIS)

    Ugajin, Mitsuhiro

    1991-05-01

    BASIC program simple and easy to operate has been developed to compute uranium density and void volume fraction for laboratory-scale uranium silicide aluminum dispersion plate-type fuel, so called miniplate. An example of the result of calculation is given in order to demonstrate how the calculated void fraction correlates with the microstructural distribution of the void in a miniplate prepared in our laboratory. The program is also able to constitute data base on important parameters for miniplates from experimentally-determined values of density, weight of each constituent and dimensions of miniplates. Utility programs pertinent to the development of the BASIC program are also given which run in the popular MS-DOS environment. All the source lists are attached and brief description for each program is made. (author)

  9. Using NEG-pumping near a high density internal target

    Energy Technology Data Exchange (ETDEWEB)

    Gruber, Alexander; Marton, Johann; Widmann, Eberhard; Zmeskal, Johann [Stefan Meyer Institut fuer Subatomare Physik, OeAW (Germany); Orth, Herbert [GSI, Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany)

    2009-07-01

    The universal detector PANDA will be constructed at the future high-energy antiproton storage ring HESR at FAIR (Facility for Antiproton and Ion Research, GSI/Darmstadt). It will use antiproton beams (1.5 to 15 GeV/c) for hadron physics in the charmonium region. The Stefan Meyer Institut (SMI) contributes to major parts of the PANDA detector like the hydrogen cluster-jet target and the vacuum system of the antiproton - target interaction zone. To ensure low background, the residual gas load in the interaction zone and in the antiproton beam-pipe has to be minimised. Most of the gas load will come from the high density internal hydrogen target. As the detector will cover almost the full solid angle, the installation of pumps near the interaction zone is impossible. Therefore the use of NEG (non-evaporative-getter) coated beam pipes has been considered as an alternative. Two setups with NEG coated tubes have been installed at SMI as prototypes of the PANDA interaction zone. General parameters of the NEG-film, its outgassing behaviour, the pumping speed and the pumping capacity for hydrogen have been tested. The results of the studies on the PANDA-interaction region are presented.

  10. Statistical mechanics of high-density bond percolation

    Science.gov (United States)

    Timonin, P. N.

    2018-05-01

    High-density (HD) percolation describes the percolation of specific κ -clusters, which are the compact sets of sites each connected to κ nearest filled sites at least. It takes place in the classical patterns of independently distributed sites or bonds in which the ordinary percolation transition also exists. Hence, the study of series of κ -type HD percolations amounts to the description of classical clusters' structure for which κ -clusters constitute κ -cores nested one into another. Such data are needed for description of a number of physical, biological, and information properties of complex systems on random lattices, graphs, and networks. They range from magnetic properties of semiconductor alloys to anomalies in supercooled water and clustering in biological and social networks. Here we present the statistical mechanics approach to study HD bond percolation on an arbitrary graph. It is shown that the generating function for κ -clusters' size distribution can be obtained from the partition function of the specific q -state Potts-Ising model in the q →1 limit. Using this approach we find exact κ -clusters' size distributions for the Bethe lattice and Erdos-Renyi graph. The application of the method to Euclidean lattices is also discussed.

  11. Low fasting low high-density lipoprotein and postprandial lipemia

    Directory of Open Access Journals (Sweden)

    Sorodila Konstandina

    2004-07-01

    Full Text Available Abstract Background Low levels of high density lipoprotein (HDL cholesterol and disturbed postprandial lipemia are associated with coronary heart disease. In the present study, we evaluated the variation of triglyceride (TG postprandially in respect to serum HDL cholesterol levels. Results Fifty two Greek men were divided into 2 main groups: a the low HDL group (HDL p = 0.002. The low HDL group had significantly higher TG at 4, 6 and 8 h postprandially compared to the controls (p = 0.006, p = 0.002, and p p = 0.017 compared to the matched-control group. ROC analysis showed that fasting TG ≥ 121 mg/dl have 100% sensitivity and 81% specificity for an abnormal TG response (auc = 0.962, p Conclusions The delayed TG clearance postprandially seems to result in low HDL cholesterol even in subjects with low fasting TG. The fasting TG > 121 mg/dl are predictable for abnormal response to fatty meal.

  12. High density turbulent plasma processes from a shock tube

    International Nuclear Information System (INIS)

    Oyedeji, O.; Johnson, J.A. III

    1991-01-01

    We have finished the first stages of our experimental and theoretical investigations on models for energy and momentum transport and for photon-particle collision processes in a turbulent quasi-stationary high density plasma. The system is explored by beginning to determine the turbulence phenomenology associated with an ionizing shock wave. The theoretical underpinnings are explored for phonon particle collisions by determining the collisional redistribution function, using Lioville Space Green's Function, which will characterize the inelastic scattering of the radiation from one frequency to another. We have observed that a weak magnetic field tends to increase the apparent random-like behaviors in a collisional turbulent plasma. On the theoretical side, we have been able to achieve a form for the collisional redistribution function. It remains to apply these concepts to a stationary turbulent plasma in the reflected ionizing shock wave and to exercise the implications of evaluations of the collisional redistribution function for such a system when it is probed by a strong radiation source. These results are discussed in detail in the publications, which have resulted from the this effort, cited at the end of the report

  13. High-field, high-density tokamak power reactor

    International Nuclear Information System (INIS)

    Cohn, D.R.; Cook, D.L.; Hay, R.D.; Kaplan, D.; Kreischer, K.; Lidskii, L.M.; Stephany, W.; Williams, J.E.C.; Jassby, D.L.; Okabayashi, M.

    1977-11-01

    A conceptual design of a compact (R 0 = 6.0 m) high power density (average P/sub f/ = 7.7 MW/m 3 ) tokamak demonstration power reactor has been developed. High magnetic field (B/sub t/ = 7.4 T) and moderate elongation (b/a = 1.6) permit operation at the high density (n(0) approximately 5 x 10 14 cm -3 ) needed for ignition in a relatively small plasma, with a spatially-averaged toroidal beta of only 4%. A unique design for the Nb 3 Sn toroidal-field magnet system reduces the stress in the high-field trunk region, and allows modularization for simpler disassembly. The modest value of toroidal beta permits a simple, modularized plasma-shaping coil system, located inside the TF coil trunk. Heating of the dense central plasma is attained by the use of ripple-assisted injection of 120-keV D 0 beams. The ripple-coil system also affords dynamic control of the plasma temperature during the burn period. A FLIBE-lithium blanket is designed especially for high-power-density operation in a high-field environment, and gives an overall tritium breeding ratio of 1.05 in the slowly pumped lithium

  14. [Residual risk: The roles of triglycerides and high density lipoproteins].

    Science.gov (United States)

    Grammer, Tanja; Kleber, Marcus; Silbernagel, Günther; Scharnagl, Hubert; März, Winfried

    2016-06-01

    In clinical trials, the reduction of LDL-cholesterol (LDL-C) with statins reduces the incidence rate of cardiovascular events by approximately one third. This means, that a sizeable "residual risk" remains. Besides high lipoprotein (a), disorders in the metabolism of triglyceride-rich lipoproteins and high density liproteins have been implicated as effectors of the residual risk. Both lipoprotein parameters correlate inversely with each other. Therefore, the etiological contributions of triglycerides and / or of HDL for developing cardiovascular disease can hardly be estimated from either observational studies or from intervention studies. The largely disappointing results of intervention studies with inhibitors of the cholesteryl ester transfer protein and in particular the available set of genetically-epidemiological studies suggest that in the last decade, the importance of HDL cholesterol has been overvalued, while the importance of triglycerides has been underestimated. High triglycerides not always atherogenic, but only if they are associated with the accumulation relatively cholesterol-enriched, incompletely catabolized remnants of chylomicrons and very low density lipoproteins (familial type III hyperlipidemia, metabolic syndrome, diabetes mellitus). The normalization of the concentration of triglycerides and remnants by inhibiting the expression of apolipoprotein C3 is hence a new, promising therapeutic target. © Georg Thieme Verlag KG Stuttgart · New York.

  15. Silicon micromachining using a high-density plasma source

    International Nuclear Information System (INIS)

    McAuley, S.A.; Ashraf, H.; Atabo, L.; Chambers, A.; Hall, S.; Hopkins, J.; Nicholls, G.

    2001-01-01

    Dry etching of Si is critical in satisfying the demands of the micromachining industry. The micro-electro-mechanical systems (MEMS) community requires etches capable of high aspect ratios, vertical profiles, good feature size control and etch uniformity along with high throughput to satisfy production requirements. Surface technology systems' (STS's) high-density inductively coupled plasma (ICP) etch tool enables a wide range of applications to be realized whilst optimizing the above parameters. Components manufactured from Si using an STS ICP include accelerometers and gyroscopes for military, automotive and domestic applications. STS's advanced silicon etch (ASE TM ) has also allowed the first generation of MEMS-based optical switches and attenuators to reach the marketplace. In addition, a specialized application for fabricating the next generation photolithography exposure masks has been optimized for 200 mm diameter wafers, to depths of ∼750 μm. Where the profile is not critical, etch rates of greater than 8 μm min -1 have been realized to replace previous methods such as wet etching. This is also the case for printer applications. Specialized applications that require etching down to pyrex or oxide often result in the loss of feature size control at the interface; this is an industry wide problem. STS have developed a technique to address this. The rapid progression of the industry has led to development of the STS ICP etch tool, as well as the process. (author)

  16. High-Density Infrared Surface Treatments of Refractories

    Energy Technology Data Exchange (ETDEWEB)

    Tiegs, T.N.

    2005-03-31

    Refractory materials play a crucial role in all energy-intensive industries and are truly a crosscutting technology for the Industries of the Future (IOF). One of the major mechanisms for the degradation of refractories and a general decrease in their performance has been the penetration and corrosion by molten metals or glass. Methods and materials that would reduce the penetration, wetting, and corrosive chemistry would significantly improve refractory performance and also maintain the quality of the processed liquid, be it metal or glass. This report presents the results of an R&D project aimed at investigating the use of high-density infrared (HDI) heating to surface treat refractories to improve their performance. The project was a joint effort between Oak Ridge National Laboratory (ORNL) and the University of Missouri-Rolla (UMR). HDI is capable of heating the near-surface region of materials to very high temperatures where sintering, diffusion, and melting can occur. The intended benefits of HDI processing of refractories were to (1) reduce surface porosity (by essentially sealing the surface to prevent liquid penetration), (2) allow surface chemistry changes to be performed by bonding an adherent coating onto the underlying refractory (in order to inhibit wetting and/or improve corrosion resistance), and (3) produce noncontact refractories with high-emissivity surface coatings.

  17. High-density polyethylene dosimetry by transvinylene FTIR analysis

    DEFF Research Database (Denmark)

    McLaughlin, W.L.; Silverman, J.; Al-Sheikhly, M.

    1999-01-01

    and electrons. The useful dose range of 0.053 cm thick high-density polyethylene film (rho = 0.961 g cm(-3); melt index = 0.8 dg min(-1)), for irradiations by (60)Co gamma radiation and 2.0 and 0.4 MeV electron beams in deaerated atmosphere (Na gas), is about 50-10(3) kGy for FTIR transvinylene......The formation of transvinylene unsaturation, -CH=CH-, due to free-radical or cationic-initiated dehydrogenation by irradiation, is a basic reaction in polyethylene and is useful for dosimetry at high absorbed doses. The radiation-enhanced infrared absorption having a maximum at nu = 965 cm......(-l) (lambda = 10.36 mu m) is stable in air and can be measured by Fourier-transform infrared (FTIR) spectrophotometry. The quantitative analysis is a useful means of product end-point dosimetry for radiation processing with gamma rays and electrons, where polyethylene is a component of the processed product...

  18. High Density Lipoprotein: A Therapeutic Target in Type 2 Diabetes

    Directory of Open Access Journals (Sweden)

    Philip J. Barter

    2013-09-01

    Full Text Available High density lipoproteins (HDLs have a number of properties that have the potential to inhibit the development of atherosclerosis and thus reduce the risk of having a cardiovascular event. These protective effects of HDLs may be reduced in patients with type 2 diabetes, a condition in which the concentration of HDL cholesterol is frequently low. In addition to their potential cardioprotective properties, HDLs also increase the uptake of glucose by skeletal muscle and stimulate the synthesis and secretion of insulin from pancreatic β cells and may thus have a beneficial effect on glycemic control. This raises the possibility that a low HDL concentration in type 2 diabetes may contribute to a worsening of diabetic control. Thus, there is a double case for targeting HDLs in patients with type 2 diabetes: to reduce cardiovascular risk and also to improve glycemic control. Approaches to raising HDL levels include lifestyle factors such as weight reduction, increased physical activity and stopping smoking. There is an ongoing search for HDL-raising drugs as agents to use in patients with type 2 diabetes in whom the HDL level remains low despite lifestyle interventions.

  19. High-density matter: current status and future challenges

    Directory of Open Access Journals (Sweden)

    Stone J. R.

    2015-01-01

    Full Text Available There are many fascinating processes in the Universe which we observe in more and more in detail thanks to increasingly sophisticated technology. One of the most interesting phenomena is the life cycle of stars, their birth, evolution and death. If the stars are massive enough, they end their lives in the core-collapse supernova explosion, the one of the most violent events in the Universe. As the result, the densest objects in the Universe, neutron stars and/or black holes are created. Naturally, the physical basis of these events should be understood in line with observation. The current status of our knowledge of processes in the life of stars is far from adequate for their true understanding. We show that although many models have been constructed their detailed ability to describe observations is limited or non-existent. Furthermore the general failure of all models means that we cannot tell which are heading in the right direction. A possible way forward in modeling of high-density matter is outlined, exemplified by the quark-meson-coupling model (QMC. This model has a natural explanation for the saturation of nuclear forces and depends on very few adjustable parameters, strongly constrained by the underlying physics. Latest QMC results for compact objects and finite nuclei are presented.

  20. High-Density Quantum Sensing with Dissipative First Order Transitions.

    Science.gov (United States)

    Raghunandan, Meghana; Wrachtrup, Jörg; Weimer, Hendrik

    2018-04-13

    The sensing of external fields using quantum systems is a prime example of an emergent quantum technology. Generically, the sensitivity of a quantum sensor consisting of N independent particles is proportional to sqrt[N]. However, interactions invariably occurring at high densities lead to a breakdown of the assumption of independence between the particles, posing a severe challenge for quantum sensors operating at the nanoscale. Here, we show that interactions in quantum sensors can be transformed from a nuisance into an advantage when strong interactions trigger a dissipative phase transition in an open quantum system. We demonstrate this behavior by analyzing dissipative quantum sensors based upon nitrogen-vacancy defect centers in diamond. Using both a variational method and a numerical simulation of the master equation describing the open quantum many-body system, we establish the existence of a dissipative first order transition that can be used for quantum sensing. We investigate the properties of this phase transition for two- and three-dimensional setups, demonstrating that the transition can be observed using current experimental technology. Finally, we show that quantum sensors based on dissipative phase transitions are particularly robust against imperfections such as disorder or decoherence, with the sensitivity of the sensor not being limited by the T_{2} coherence time of the device. Our results can readily be applied to other applications in quantum sensing and quantum metrology where interactions are currently a limiting factor.

  1. High density operation on the HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Xiang Gao

    2000-01-01

    The structure of the operation region has been studied in the HT-7 superconducting tokamak, and progress on the extension of the HT-7 ohmic discharge operation region is reported. A density corresponding to 1.2 times the Greenwald limit was achieved by RF boronization. The density limit appears to be connected to the impurity content and the edge parameters, so the best results are obtained with very clean plasmas and peaked electron density profiles. The peaking factors of electron density profiles for different current and line averaged densities were observed. The density behaviour and the fuelling efficiency for gas puffing (20-30%), pellet injection (70-80%) and molecular beam injection (40-50%) were studied. The core crash sawteeth and MHD behaviour, which were induced by an injected pellet, were observed and the events correlated with the change of current profile and reversed magnetic shear. The MARFE phenomena on HT-7 are summarized. The best correlation has been found between the total input ohmic power and the product of the edge line averaged density and Z eff . HT-7 could be easily operated in the high density region MARFE-free using RF boronization. (author)

  2. The physics of the high density Z-pinch

    International Nuclear Information System (INIS)

    Glasser, A.H.; Hammel, J.E.; Lewis, H.R.

    1988-01-01

    The fiber-initiated High-Density Z-Pinch (HDZP) is a novel concept in which fusion plasma could be produced by applying 2 MV along a thin filament of frozen deuterium, 20-30 μm in diameter, 5-10 cm long. The megamp-range currents that result would ohmically heat the fiber to fusion temperatures in 100 ns while maintaining nearly constant radius. The plasma pressure would be held stably by the self-magnetic field for many radial sound transit times during the current-rise phase while, in the case of D-T, a significant fraction of the fiber undergoes thermonuclear fusion. This paper presents results of Los Alamos HDZP studies. Existing and new experiments are described. A succession of theoretical studies, including 1D self-similar and numerical studies of the hot plasma phase, 1D and 2D numerical studies of the cold startup phase, and 3D numerical studies of stability in the hot regime, are then presented. 9 refs., 4 figs

  3. High-density housing that works for all

    Energy Technology Data Exchange (ETDEWEB)

    Hasan, Arif

    2010-03-15

    In an urbanising world, the way people fit into cities is vastly important - socially, economically, environmentally, even psychologically. So density, or the number of people living in a given area, is central to urban design and planning. Both governments and markets tend to get density wrong, leading to overcrowding, urban sprawl or often both. A case in point are the high-rise buildings springing up throughtout urban Asia - perceived as key features of that widely touted concept, the 'world-class city'. While some may offer a viable solution to land pressures and density requirements, many built to house evicted or resettled 'slum' dwellers are a social and economic nightmare - inconveniently sited, overcrowded and costly. New evidence from Karachi, Pakistan, reveals a real alternative. Poor people can create liveable high-density settlements as long as community control, the right technical assistance and flexible designs are in place. A city is surely 'world-class' only when it is cosmopolitan – built to serve all, including the poorest.

  4. High-Density Superconducting Cables for Advanced ACTPol

    Science.gov (United States)

    Pappas, C. G.; Austermann, J.; Beall, J. A.; Duff, S. M.; Gallardo, P. A.; Grace, E.; Henderson, S. W.; Ho, S. P.; Koopman, B. J.; Li, D.; McMahon, J.; Nati, F.; Niemack, M. D.; Niraula, P.; Salatino, M.; Schillaci, A.; Schmitt, B. L.; Simon, S. M.; Staggs, S. T.; Stevens, J. R.; Vavagiakis, E. M.; Ward, J. T.; Wollack, E. J.

    2016-07-01

    Advanced ACTPol (AdvACT) is an upcoming Atacama Cosmology Telescope (ACT) receiver upgrade, scheduled to deploy in 2016, that will allow measurement of the cosmic microwave background polarization and temperature to the highest precision yet with ACT. The AdvACT increase in sensitivity is partly provided by an increase in the number of transition-edge sensors (TESes) per array by up to a factor of two over the current ACTPol receiver detector arrays. The high-density AdvACT TES arrays require 70 \\upmu m pitch superconducting flexible cables (flex) to connect the detector wafer to the first-stage readout electronics. Here, we present the flex fabrication process and test results. For the flex wiring layer, we use a 400-nm-thick sputtered aluminum film. In the center of the cable, the wiring is supported by a polyimide substrate, which smoothly transitions to a bare (uncoated with polyimide) silicon substrate at the ends of the cable for a robust wedge wire-bonding interface. Tests on the first batch of flex made for the first AdvACT array show that the flex will meet the requirements for AdvACT, with a superconducting critical current above 1 mA at 500 mK, resilience to mechanical and cryogenic stress, and a room temperature yield of 97 %.

  5. High-Density Quantum Sensing with Dissipative First Order Transitions

    Science.gov (United States)

    Raghunandan, Meghana; Wrachtrup, Jörg; Weimer, Hendrik

    2018-04-01

    The sensing of external fields using quantum systems is a prime example of an emergent quantum technology. Generically, the sensitivity of a quantum sensor consisting of N independent particles is proportional to √{N }. However, interactions invariably occurring at high densities lead to a breakdown of the assumption of independence between the particles, posing a severe challenge for quantum sensors operating at the nanoscale. Here, we show that interactions in quantum sensors can be transformed from a nuisance into an advantage when strong interactions trigger a dissipative phase transition in an open quantum system. We demonstrate this behavior by analyzing dissipative quantum sensors based upon nitrogen-vacancy defect centers in diamond. Using both a variational method and a numerical simulation of the master equation describing the open quantum many-body system, we establish the existence of a dissipative first order transition that can be used for quantum sensing. We investigate the properties of this phase transition for two- and three-dimensional setups, demonstrating that the transition can be observed using current experimental technology. Finally, we show that quantum sensors based on dissipative phase transitions are particularly robust against imperfections such as disorder or decoherence, with the sensitivity of the sensor not being limited by the T2 coherence time of the device. Our results can readily be applied to other applications in quantum sensing and quantum metrology where interactions are currently a limiting factor.

  6. LTCC magnetic components for high density power converter

    Science.gov (United States)

    Lebourgeois, Richard; Labouré, Eric; Lembeye, Yves; Ferrieux, Jean-Paul

    2018-04-01

    This paper deals with multilayer magnetic components for power electronics application and specifically for high frequency switching. New formulations based on nickel-zinc-copper spinel ferrites were developed for high power and high frequency applications. These ferrites can be sintered at low temperature (around 900°C) which makes them compatible with the LTCC (Low Temperature Co-fired Ceramics) technology. Metallic parts of silver or gold can be fully integrated inside the ferrite while guaranteeing the integrity of both the ferrite and the metal. To make inductors or transformers with the required properties, it is mandatory to have nonmagnetic parts between the turns of the winding. Then it is essential to find a dielectric material, which can be co-sintered both with the ferrite and the metal. We will present the solution we found to this problem and we will describe the results we obtained for a multilayer co-sintered transformer. We will see that these new components have good performance compared with the state of the art and are very promising for developing high density switching mode power supplies.

  7. Acrolein impairs the cholesterol transport functions of high density lipoproteins.

    Science.gov (United States)

    Chadwick, Alexandra C; Holme, Rebecca L; Chen, Yiliang; Thomas, Michael J; Sorci-Thomas, Mary G; Silverstein, Roy L; Pritchard, Kirkwood A; Sahoo, Daisy

    2015-01-01

    High density lipoproteins (HDL) are considered athero-protective, primarily due to their role in reverse cholesterol transport, where they transport cholesterol from peripheral tissues to the liver for excretion. The current study was designed to determine the impact of HDL modification by acrolein, a highly reactive aldehyde found in high abundance in cigarette smoke, on the cholesterol transport functions of HDL. HDL was chemically-modified with acrolein and immunoblot and mass spectrometry analyses confirmed apolipoprotein crosslinking, as well as acrolein adducts on apolipoproteins A-I and A-II. The ability of acrolein-modified HDL (acro-HDL) to serve as an acceptor of free cholesterol (FC) from COS-7 cells transiently expressing SR-BI was significantly decreased. Further, in contrast to native HDL, acro-HDL promotes higher neutral lipid accumulation in murine macrophages as judged by Oil Red O staining. The ability of acro-HDL to mediate efficient selective uptake of HDL-cholesteryl esters (CE) into SR-BI-expressing cells was reduced compared to native HDL. Together, the findings from our studies suggest that acrolein modification of HDL produces a dysfunctional particle that may ultimately promote atherogenesis by impairing functions that are critical in the reverse cholesterol transport pathway.

  8. High-Density Superconducting Cables for Advanced ACTPol

    Science.gov (United States)

    Pappas, C. G.; Austermann, J.; Beall, J. A.; Duff, S. M.; Gallardo, P. A.; Grace, E.; Henderson, S. W.; Ho, S. P.; Koopman, B. J.; Li, D.; hide

    2016-01-01

    Advanced ACTPol (AdvACT) is an upcoming Atacama Cosmology Telescope (ACT) receiver upgrade, scheduled to deploy in 2016, that will allow measure- ment of the cosmic microwave background polarization and temperature to the highest precision yet with ACT. The AdvACT increase in sensitivity is partly provided by an increase in the number of transition-edge sensors (TESes) per array by up to a factor of two over the current ACTPol receiver detector arrays. The high-density AdvACT TES arrays require 70 µ m pitch superconducting flexible cables (flex) to connect the detec- tor wafer to the first-stage readout electronics. Here, we present the flex fabrication process and test results. For the flex wiring layer, we use a 400-nm-thick sputtered alu- minum film. In the center of the cable, the wiring is supported by a polyimide substrate, which smoothly transitions to a bare (uncoated with polyimide) silicon substrate at the ends of the cable for a robust wedge wire-bonding interface. Tests on the first batch of flex made for the first AdvACT array show that the flex will meet the requirements for AdvACT, with a superconducting critical current above 1 mA at 500 mK, resilience to mechanical and cryogenic stress, and a room temperature yield of 97%.

  9. The infant mirror neuron system studied with high density EEG.

    Science.gov (United States)

    Nyström, Pär

    2008-01-01

    The mirror neuron system has been suggested to play a role in many social capabilities such as action understanding, imitation, language and empathy. These are all capabilities that develop during infancy and childhood, but the human mirror neuron system has been poorly studied using neurophysiological measures. This study measured the brain activity of 6-month-old infants and adults using a high-density EEG net with the aim of identifying mirror neuron activity. The subjects viewed both goal-directed movements and non-goal-directed movements. An independent component analysis was used to extract the sources of cognitive processes. The desynchronization of the mu rhythm in adults has been shown to be a marker for activation of the mirror neuron system and was used as a criterion to categorize independent components between subjects. The results showed significant mu desynchronization in the adult group and significantly higher ERP activation in both adults and 6-month-olds for the goal-directed action observation condition. This study demonstrate that infants as young as 6 months display mirror neuron activity and is the first to present a direct ERP measure of the mirror neuron system in infants.

  10. High-Density Stretchable Electrode Grids for Chronic Neural Recording.

    Science.gov (United States)

    Tybrandt, Klas; Khodagholy, Dion; Dielacher, Bernd; Stauffer, Flurin; Renz, Aline F; Buzsáki, György; Vörös, János

    2018-04-01

    Electrical interfacing with neural tissue is key to advancing diagnosis and therapies for neurological disorders, as well as providing detailed information about neural signals. A challenge for creating long-term stable interfaces between electronics and neural tissue is the huge mechanical mismatch between the systems. So far, materials and fabrication processes have restricted the development of soft electrode grids able to combine high performance, long-term stability, and high electrode density, aspects all essential for neural interfacing. Here, this challenge is addressed by developing a soft, high-density, stretchable electrode grid based on an inert, high-performance composite material comprising gold-coated titanium dioxide nanowires embedded in a silicone matrix. The developed grid can resolve high spatiotemporal neural signals from the surface of the cortex in freely moving rats with stable neural recording quality and preserved electrode signal coherence during 3 months of implantation. Due to its flexible and stretchable nature, it is possible to minimize the size of the craniotomy required for placement, further reducing the level of invasiveness. The material and device technology presented herein have potential for a wide range of emerging biomedical applications. © 2018 The Authors. Published by WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  11. Using NEG-pumping near a high density internal target

    International Nuclear Information System (INIS)

    Gruber, A.; Marton, J.; Widmann, E.; Zmeskal, J.; Orth, H.

    2008-01-01

    Full text: The Stefan Meyer Institut (SMI) is part of the international PANDA collaboration. The universal detector will be constructed at the future high-energy antiproton storage ring HESR at FAIR (Facility for Antiproton and Ion Research, GSI/Darmstadt). PANDA will use antiproton beams (1.5 to 15 GeV/c) for hadron physics in the charmonium region. SMI contributes to major parts of the PANDA detector like the hydrogen cluster-jet target and the vacuum system of the antiproton - target interaction zone. To ensure low background, the residual gas load in the interaction zone and in the antiproton beam-pipe has to be minimized. Most of the gas load, of course will come from the high density internal hydrogen target. Since the PANDA detector will cover almost the full solid angle, the installation of pumps near the interaction zone is impossible. Therefore, the use of NEG (non-evaporative-getter) coated beam pipes has been considered as an alternative. Two setups with NEG coated tubes have been installed at SMI as prototypes of the PANDA interaction zone. The outgassing behavior, the pumping speed and the pumping capacity for hydrogen have been tested. The status of the studies of the interaction region will be presented. (author)

  12. High U-density nuclear fuel development with application of centrifugal atomization technology

    International Nuclear Information System (INIS)

    Kim, Chang Kyu; Kim, Ki Hwan; Lee, Don Bae

    1997-01-01

    In order to simplify the preparation process and improve the properties of uranium silicide fuels prepared by mechanical comminution, a fuel fabrication process applying rotating-disk centrifugal atomization technology was invented in KAERI in 1989. The major characteristic of atomized U 3 Si and U 3 Si 2 powders have been examined. The out-pile properties, including the thermal compatibility between atomized particle and aluminum matrix in uranium silicide dispersion fuels, have generally showed a superiority to the comminuted fuels. Moreover, the RERTR (reduced enrichment for research and test reactors) program, which recently begins to develop very-high-density uranium alloy fuels, including U-Mo fuels, requires the centrifugal atomization process to overcome the contaminations of impurities and the difficulties of the comminution process. In addition, a cooperation with ANL in the U.S. has been performed to develop high-density fuels with an application of atomization technology since December 1996. If the microplate and miniplate irradiation tests of atomized fuels, which have been performed with ANL, demonstrated the stability and improvement of in-reactor behaviors, nuclear fuel fabrication technology by centrifugal atomization could be most-promising to the production method of very-high-uranium-loading fuels. (author). 22 refs., 2 tabs., 12 figs

  13. Assessing photocatalytic power of g-C3N4 for solar fuel production: A first-principles study involving quasi-particle theory and dispersive forces.

    Science.gov (United States)

    Osorio-Guillén, J M; Espinosa-García, W F; Moyses Araujo, C

    2015-09-07

    First-principles quasi-particle theory has been employed to assess catalytic power of graphitic carbon nitride, g-C3N4, for solar fuel production. A comparative study between g-h-triazine and g-h-heptazine has been carried out taking also into account van der Waals dispersive forces. The band edge potentials have been calculated using a recently developed approach where quasi-particle effects are taken into account through the GW approximation. First, it was found that the description of ground state properties such as cohesive and surface formation energies requires the proper treatment of dispersive interaction. Furthermore, through the analysis of calculated band-edge potentials, it is shown that g-h-triazine has high reductive power reaching the potential to reduce CO2 to formic acid, coplanar g-h-heptazine displays the highest thermodynamics force toward H2O/O2 oxidation reaction, and corrugated g-h-heptazine exhibits a good capacity for both reactions. This rigorous theoretical study shows a route to further improve the catalytic performance of g-C3N4.

  14. Steady- and Transient-State Analyses of Fully Ceramic Microencapsulated Fuel with Randomly Dispersed Tristructural Isotropic Particles via Two-Temperature Homogenized Model—I: Theory and Method

    Directory of Open Access Journals (Sweden)

    Yoonhee Lee

    2016-06-01

    Full Text Available As a type of accident-tolerant fuel, fully ceramic microencapsulated (FCM fuel was proposed after the Fukushima accident in Japan. The FCM fuel consists of tristructural isotropic particles randomly dispersed in a silicon carbide (SiC matrix. For a fuel element with such high heterogeneity, we have proposed a two-temperature homogenized model using the particle transport Monte Carlo method for the heat conduction problem. This model distinguishes between fuel-kernel and SiC matrix temperatures. Moreover, the obtained temperature profiles are more realistic than those of other models. In Part I of the paper, homogenized parameters for the FCM fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure are obtained by (1 matching steady-state analytic solutions of the model with the results of particle transport Monte Carlo method for heat conduction problems, and (2 preserving total enthalpies in fuel kernels and SiC matrix. The homogenized parameters have two desirable properties: (1 they are insensitive to boundary conditions such as coolant bulk temperatures and thickness of cladding, and (2 they are independent of operating power density. By performing the Monte Carlo calculations with the temperature-dependent thermal properties of the constituent materials of the FCM fuel, temperature-dependent homogenized parameters are obtained.

  15. Analysis of the star formation histories of galaxies in different environments: from low to high density

    Science.gov (United States)

    Ortega-Minakata, René A.

    2015-11-01

    formation histories of galaxies in our catalogue were found to be strongly dependent on their morphology, and consequently on their emission-line activity. Star-forming galaxies are late types that have the youngest populations; Sy2-dominated galaxies show a mixture of young and old populations, while LINER-likes have older populations; and passive, early-type galaxies have the oldest populations and have no current star formation. This is consistent with the idea that the processes that fix or change the morphology of galaxies are more internal and modulated by their mass, and are tightly related to how much gas is available to stimulate or stop star formation or AGN activity. In contrast, the star formation histories of galaxies were found to be only weakly dependent on their environmental density, with isolated galaxies showing somewhat younger populations than galaxies in high-density environments. This is consistent with the weak morphology-density relation found for the general population of galaxies, and supports the idea that morphology and formation history are tightly related and, while the processes that change the morphology are more common in the cluster environment, the environmental density itself does not directly affect much the formation history of galaxies. The stellar mass of galaxies seems to modulate their activity and morphology: massive galaxies formed more rapidly in the past, efficiently converting their gas into stars, leaving little or no gas to form stars or fuel AGN activity later on, thus making them low-intensity active galaxies or passive galaxies. The formation of these massive galaxies would then only depend on the local density of protogalaxies, so a high merger rate in environments similar to compact groups of galaxies in the past would result in an early-type galaxy, effectively explaining the relation between mass, activity, morphology, stellar mean age and velocity dispersion of galaxies. The catalogue presented here is useful and

  16. Experimental study of high density foods for the Space Operations Center

    Science.gov (United States)

    Ahmed, S. M.

    1981-01-01

    The experimental study of high density foods for the Space Operations Center is described. A sensory evaluation of the high density foods was conducted first to test the acceptability of the products. A shelf-life study of the high density foods was also conducted for three different time lengths at three different temperatures. The nutritional analysis of the high density foods is at present incomplete.

  17. Nondestructive, energy-dispersive, x-ray fluorescence analysis of actinide stream concentrations from reprocessed nuclear fuels

    International Nuclear Information System (INIS)

    Camp, D.C.; Ruhter, W.D.

    1979-01-01

    In one plan for reprocessing LWR spent fuel, after separation from fission products and transplutonics, part of the U and all of the Pu in a nitrate solution will form a coprocessed stream which is then evaporated and sent to a hold tank for accounting. The remaining U fraction will be purified and sent to a separate storage tank. These two streams can be monitored using x-ray fluorescence analysis. This report discusses equipment, spectra, cell calibration, and dynamic concentration measurements. 7 figures

  18. Phase analyses of silicide or nitride coated U–Mo and U–Mo–Ti particle dispersion fuel after out-of-pile annealing

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woo Jeong [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong, Daejeon 305-353 (Korea, Republic of); Palancher, Hervé [CEA, DEN, DEC, F-13108 Saint Paul Lez Durance Cedex (France); Ryu, Ho Jin, E-mail: hojinryu@kaist.ac.kr [Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong, Daejeon 305-701 (Korea, Republic of); Park, Jong Man; Nam, Ji Min [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong, Daejeon 305-353 (Korea, Republic of); Bonnin, Anne [CEA, DEN, DEC, F-13108 Saint Paul Lez Durance Cedex (France); ESRF, 6, rue J. Horowitz, F-38000 Grenoble Cedex (France); Honkimäki, Veijo [ESRF, 6, rue J. Horowitz, F-38000 Grenoble Cedex (France); Charollais, François [CEA, DEN, DEC, F-13108 Saint Paul Lez Durance Cedex (France); Lemoine, Patrick [CEA, DEN, DISN, 91191 Gif sur Yvette (France)

    2014-03-15

    Highlights: • Silicide or nitride layers were coated on atomized U–Mo or U–Mo–Ti powder. • The constituent phases after annealing were identified through high-energy XRD. • U{sub 3}Si{sub 5} and U{sub 4}Mo(Mo{sub x}Si{sub 1−x})Si{sub 2} were identified in the silicide coating layers. • UN was identified for U–Mo particles and UN and U{sub 4}N{sub 7} formed on U–Mo–Ti particles. -- Abstract: The coating of silicide or nitride layers on U–7 wt%Mo or U–7 wt%Mo–1 wt%Ti particles has been proposed for the minimization of the interaction phase growth in U–Mo/Al dispersion fuel during irradiation. Out-of-pile annealing tests show reduced inter-diffusion by forming silicide or nitride protective layers on U–Mo and U–Mo–Ti particles. To characterize the constituent phases of the coated layers on U–Mo and U–Mo–Ti particles and the interaction phases of coated U–Mo and U–Mo–Ti particle dispersed Al matrix fuel, synchrotron X-ray diffraction experiments have been performed. It was identified that silicide coating layers consisted mainly of U{sub 3}Si{sub 5} and U{sub 4}Mo(Mo{sub x}Si{sub 1−x})Si{sub 2}, and nitride coating layers were composed of mainly UN and U{sub 4}N{sub 7}. The interaction phases obtained after annealing of coated U–Mo and U–Mo–Ti particle dispersion samples were identical to those found in U–Mo/Al–Si and U–Mo/Al systems. Nitride-coated particles showed less interaction formation than silicide-coated particles after annealing at 580 °C for 1 h owing to the higher susceptibility to breakage of the silicide coating layers during hot extrusion.

  19. Ultra-high Density SNParray in Neuroblastoma Molecular Diagnostics

    Directory of Open Access Journals (Sweden)

    Inge M. Ambros

    2014-08-01

    Full Text Available Neuroblastoma serves as a paradigm for applying tumor genomic data for determining patient prognosis and thus for treatment allocation. MYCN status, i.e. amplified vs. non-amplified, was one of the very first biomarkers in oncology to discriminate aggressive from less aggressive or even favorable clinical courses of neuroblastoma. However, MYCN amplification is by far not the only genetic change associated with unfavorable clinical courses: so called segmental chromosomal aberrations, i.e. gains or losses of chromosomal fragments, can also indicate tumor aggressiveness. The clinical use of these genomic aberrations has, however, been hampered for many years by methodical and interpretational problems. Only after reaching worldwide consensus on markers, methodology, and data interpretation, information on SCAs has recently been implemented in clinical studies. Now, a number of collaborative studies within COG, GPOH and SIOPEN use genomic information to stratify therapy for patients with localized and metastatic disease. Recently, new types of DNA based aberrations influencing the clinical behavior of neuroblastomas have been described. Deletions or mutations of genes like ATRX and a phenomenon referred to as chromothripsis are all assumed to correlate with an unfavorable clinical behavior. However, these genomic aberrations need to be scrutinized in larger studies applying the most appropriate techniques. Single nucleotide polymorphism (SNP arrays have proven successful in deciphering genomic aberrations of cancer cells; these techniques, however, are usually not applied in the daily routine. Here, we present an ultra-high density (UHD SNParray technique which is, because of its high specificity and sensitivity and the combined copy number and allele information, highly appropriate for the genomic diagnosis of neuroblastoma and other malignancies.

  20. High density microelectronics package using low temperature cofirable ceramics

    International Nuclear Information System (INIS)

    Fu, S.-L.; Hsi, C.-S.; Chen, L.-S.; Lin, W. K.

    1997-01-01

    Low Temperature Cofired Ceramics (LTCC) is a relative new thick film process and has many engineering and manufacturing advantages over both the sequential thick film process and high temperature cofired ceramic modules. Because of low firing temperature, low sheet resistance metal conductors, commercial thick film resistors, and thick film capacitors can be buried in or printed on the substrates. A 3-D multilayer ceramic substrate can be prepared via laminating and co-firing process. The packing density of the LTCC substrates can be increased by this 3-D packing technology. At Kaohsiung Polytechnic Institute (KPI), a LTCC substrate system has been developed for high density packaging applications, which had buried surface capacitors and resistors. The developed cordierite-glass ceramic substrate, which has similar thermal expansion as silicon chip, is a promising material for microelectronic packaging. When the substrates were sintered at temperatures between 850-900 degree centigrade, a relative density higher than 96 % can be obtained. The substrate had a dielectric constant between 5.5 and 6.5. Ruthenium-based resistor pastes were used for resistors purposes. The resistors fabricated in/on the LTCC substrates were strongly depended on the microstructures developed in the resistor films. Surface resistors were laser trimmed in order to obtain specific values for the resistors. Material with composition Pb(Fe 2/3 W 1/3 ) x (Fe l/2 Nb l/2 ) y Ti 2 O 3 was used as dielectric material of the capacitor in the substrate. The material can be sintered at temperatures between 850-930 degree centigrade, and has dielectric constant as high as 26000. After cofiring, good adhesion between dielectric and substrate layers was obtained. Combing the buried resistors and capacitors together with the lamination of LTCC layer, a 3-dimensional multilayered ceramic package was fabricated. (author)

  1. Degradation of high density lipoprotein in cultured rat luteal cells

    International Nuclear Information System (INIS)

    Rajan, V.P.; Menon, K.M.J.

    1986-01-01

    In rat ovary luteal cells, degradation of high density lipoprotein (HDL) to tricholoracetic acid (TCA)-soluble products accounts for only a fraction of the HDL-derived cholesterol used for steroidogenesis. In this study the authors have investigated the fate of 125 I]HDL bound to cultured luteal cells using pulse-chase technique. Luteal cell cultures were pulse labeled with [ 125 I]HDL 3 and reincubated in the absence of HDL. By 24 h about 50% of the initallay bound radioactivity was released into the medium, of which 60-65% could be precipitated with 10% TCA. Gel filtration of the chase incubation medium on 10% agarose showed that the amount of TCA-soluble radioactivity was nearly completely accounted for by a sharp peak in the low molecular weight region which was identified as 96% monoiodotyrosine by paper chromatography. The TCA-precipitable radioactivity was nearly completely accounted for by a sharp peak in the low molecular weight region which was identified as 96% monoiodotyrosine by paper chromatography. The TCA-precipitable radioactivity eluted over a wide range of molecular weights (15,000-80,000), and there was very little intact HDL present. Electrophoresis of the chase medium showed that component of the TCA-precipitable portion had mobility similar to apo AI. Lysosomal inhibitors of receptor-mediated endocytosis had no effect on the composition or quantity of radioactivity released during chase incubation. The results show that HDL 3 binding to luteal cells is followed by complete degradation of the lipoprotein, although the TCA-soluble part does not reflect the extent of degradation

  2. Surface interactions involved in flashover with high density electronegative gases.

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, Keith Conquest; Warne, Larry Kevin; Jorgenson, Roy Eberhardt; Wallace, Zachariah Red; Lehr, Jane Marie

    2010-01-01

    This report examines the interactions involved with flashover along a surface in high density electronegative gases. The focus is on fast ionization processes rather than the later time ionic drift or thermalization of the discharge. A kinetic simulation of the gas and surface is used to examine electron multiplication and includes gas collision, excitation and ionization, and attachment processes, gas photoionization and surface photoemission processes, as well as surface attachment. These rates are then used in a 1.5D fluid ionization wave (streamer) model to study streamer propagation with and without the surface in air and in SF6. The 1.5D model therefore includes rates for all these processes. To get a better estimate for the behavior of the radius we have studied radial expansion of the streamer in air and in SF6. The focus of the modeling is on voltage and field level changes (with and without a surface) rather than secondary effects, such as, velocities or changes in discharge path. An experiment has been set up to carry out measurements of threshold voltages, streamer velocities, and other discharge characteristics. This setup includes both electrical and photographic diagnostics (streak and framing cameras). We have observed little change in critical field levels (where avalanche multiplication sets in) in the gas alone versus with the surface. Comparisons between model calculations and experimental measurements are in agreement with this. We have examined streamer sustaining fields (field which maintains ionization wave propagation) in the gas and on the surface. Agreement of the gas levels with available literature is good and agreement between experiment and calculation is good also. Model calculations do not indicate much difference between the gas alone versus the surface levels. Experiments have identified differences in velocity between streamers on the surface and in the gas alone (the surface values being larger).

  3. High density microelectronics package using low temperature cofirable ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Fu, S -L; Hsi, C -S; Chen, L -S; Lin, W K [Kaoshiung Polytechnic Institute Ta-Hsu, Kaoshiung (China)

    1998-12-31

    Low Temperature Cofired Ceramics (LTCC) is a relative new thick film process and has many engineering and manufacturing advantages over both the sequential thick film process and high temperature cofired ceramic modules. Because of low firing temperature, low sheet resistance metal conductors, commercial thick film resistors, and thick film capacitors can be buried in or printed on the substrates. A 3-D multilayer ceramic substrate can be prepared via laminating and co-firing process. The packing density of the LTCC substrates can be increased by this 3-D packing technology. At Kaohsiung Polytechnic Institute (KPI), a LTCC substrate system has been developed for high density packaging applications, which had buried surface capacitors and resistors. The developed cordierite-glass ceramic substrate, which has similar thermal expansion as silicon chip, is a promising material for microelectronic packaging. When the substrates were sintered at temperatures between 850-900 degree centigrade, a relative density higher than 96 % can be obtained. The substrate had a dielectric constant between 5.5 and 6.5. Ruthenium-based resistor pastes were used for resistors purposes. The resistors fabricated in/on the LTCC substrates were strongly depended on the microstructures developed in the resistor films. Surface resistors were laser trimmed in order to obtain specific values for the resistors. Material with composition Pb(Fe{sub 2/3}W{sub 1/3}){sub x}(Fe{sub l/2}Nb{sub l/2}){sub y}Ti{sub 2}O{sub 3} was used as dielectric material of the capacitor in the substrate. The material can be sintered at temperatures between 850-930 degree centigrade, and has dielectric constant as high as 26000. After cofiring, good adhesion between dielectric and substrate layers was obtained. Combing the buried resistors and capacitors together with the lamination of LTCC layer, a 3-dimensional multilayered ceramic package was fabricated. (author)

  4. Pharmacologic management of isolated low high-density lipoprotein syndrome.

    Science.gov (United States)

    Bermúdez, Valmore; Cano, Raquel; Cano, Clímaco; Bermúdez, Fernando; Arraiz, Nailet; Acosta, Luis; Finol, Freddy; Pabón, María Rebeca; Amell, Anilsa; Reyna, Nadia; Hidalgo, Joaquin; Kendall, Paúl; Manuel, Velasco; Hernández, Rafael

    2008-01-01

    High-density lipoprotein (HDL) cholesterol is a heterogeneous group of lipoproteins exhibiting a variety of properties like prostacyclin production stimulation, decrease in platelet aggregation, endothelial cell apoptosis inhibition, and low-density lipoprotein oxidation blockade. Epidemiologic studies have shown an inverse relation between HDL cholesterol levels and cardiovascular risk. Low HDL cholesterol is associated with increased risk for myocardial infarction, stroke, sudden death, peripheral artery disease, and postangioplasty restenosis. In contrast, high HDL levels are associated with longevity and protection against atherosclerotic disease development. Given the evolving epidemic of obesity, diabetes mellitus, and metabolic syndrome, the prevalence of low HDL will continue to rise. In the United States, low HDL is present in 35% of men, 15% of women, and approximately 63% of patients with coronary artery disease. Data extracted from the Framingham study highlight that 1-mg increase in HDL levels decreases by 2% to 3% the risk of cardiovascular disease. There is no doubt regarding clinical importance about isolated low HDL, but relatively few clinicians consider a direct therapeutic intervention of this dyslipidemia. In this sense, lifestyle measures should be the first-line strategy to manage low HDL levels. On the other hand, pharmacologic options include niacin, fibrates, and statins. Fibrates appear to reduce risk preferentially in patients with low HDL with metabolic syndrome, whereas statins reduce risk across all levels of HDL. Torcetrapib, a cholesteryl esters transfer protein inhibitor, represented a hope to raise this lipoprotein; however, all clinical trials on this drug had ceased after ILLUMINATE, RADIANCE and ERASE trials had recorded an increase in mortality, rates of myocardial infarction, angina, and heart failure. In the near future, drugs as beta-glucans, Apo-A1 mimetic peptides, and ACAT inhibitors, are the new promises to treat this

  5. Neutron shielding properties of a new high-density concrete

    International Nuclear Information System (INIS)

    Lorente, A.; Gallego, E.; Vega Carrillo, H.R.; Mendez, R.

    2008-01-01

    The neutron shielding properties of a new high-density concrete (commercially available under the name Hormirad TM , developed in Spain by the company CT-RAD) have been characterized both experimentally and by Monte Carlo calculations. The shielding properties of this concrete against photons were previously studied and the material is being used to build bunkers, mazes and doors in medical accelerator facilities with good overall results. In this work, the objective was to characterize the material behaviour against neutrons, as well as to test alternative mixings including boron compounds in an effort to improve neutron shielding efficiency. With that purpose, Hormirad TM slabs of different thicknesses were exposed to an 241 Am-Be neutron source under controlled conditions in the neutron measurements laboratory of the Nuclear Engineering Department at UPM. The original mix, which includes a high fraction of magnetite, was then modified by adding different proportions of anhydrous borax (Na 2 B 4 O 7 ). In order to have a reference against common concrete used to shield medical accelerator facilities, the same experiment was repeated with ordinary (HA-25) concrete slabs. In parallel to the experiments, Monte Carlo calculations of the experiments were performed with MCNP5. The experimental results agree reasonably well with the Monte Carlo calculations. Therefore, the first and equilibrium tenth-value layers have been determined for the different types of concrete tested. The results show an advantageous behaviour of the Hormirad TM concrete, in terms of neutron attenuation against real thickness of the shielding. Borated concretes seem less practical since they did not show better neutron attenuation with respect to real thickness and their structural properties are worse. The neutron attenuation properties of Hormirad TM for typical neutron spectra in clinical LINAC accelerators rooms have been also characterized by Monte Carlo calculation. (author)

  6. Deposition and characterisation of copper for high density interconnects

    International Nuclear Information System (INIS)

    McCusker, N.

    1999-09-01

    Copper has been deposited by sputtering and investigated for application as high density interconnects, with a view to maximising its performance and reliability. A sputter deposition process using gettering has been developed, which produces consistently pure, low resistivity films. A relationship between film thickness and resistivity has been explained by studying the grain growth process in copper films using atomic force microscopy. The Maydas-Shatzkes model has been used to separate the contributions of grain boundary and surface scattering to thin film resistivity, in copper and gold. Stress and texture in copper film have been studied. Annealing has been used to promote grain growth and texture development. Electromigration has been studied in copper and aluminium interconnects using a multi-line accelerated test set-up. A difference in failure distributions and void morphologies has been explained by an entirely different damage mechanism. The importance of surface/interface migration in electromigration damage of copper lines has been established and explained using a grain boundary-grooving model. A tantalum overlayer was found to extend the lifetime of copper lines. A composite sputtering target has been used to deposit copper/zirconium alloy films. The composition of the alloys was studied by Rutherford backscattering, Auger and secondary neutral mass spectrometry. The alloy films had an improved electromigration lifetime. A surface controlled mechanism is proposed to explain the advantage. A metal oxide semiconductor (MOS) capacitor technique is used to investigate barrier reliability. Tungsten is shown to be an effective diffusion barrier for copper, up to 700 deg. C. (author)

  7. Modelling high density phenomena in hydrogen fibre Z-pinches

    International Nuclear Information System (INIS)

    Chittenden, J.P.

    1990-09-01

    The application of hydrogen fibre Z-pinches to the study of the radiative collapse phenomenon is studied computationally. Two areas of difficulty, the formation of a fully ionized pinch from a cryogenic fibre and the processes leading to collapse termination, are addressed in detail. A zero-D model based on the energy equation highlights the importance of particle end losses and changes in the Coulomb logarithm upon collapse initiation and termination. A 1-D Lagrangian resistive MHD code shows the importance of the changing radial profile shapes, particularly in delaying collapse termination. A 1-D, three fluid MHD code is developed to model the ionization of the fibre by thermal conduction from a high temperature surface corona to the cold core. Rate equations for collisional ionization, 3-body recombination and equilibration are solved in tandem with fluid equations for the electrons, ions and neutrals. Continuum lowering is found to assist ionization at the corona-core interface. The high density plasma phenomena responsible for radiative collapse termination are identified as the self-trapping of radiation and free electron degeneracy. A radiation transport model and computational analogues for the effects of degeneracy upon the equation of state, transport coefficients and opacity are implemented in the 1-D, single fluid model. As opacity increases the emergent spectrum is observed to become increasingly Planckian and a fall off in radiative cooling at small radii and low frequencies occurs giving rise to collapse termination. Electron degeneracy terminates radiative collapse by supplementing the radial pressure gradient until the electromagnetic pinch force is balanced. Collapse termination is found to be a hybrid process of opacity and degeneracy effects across a wide range of line densities with opacity dominant at large line densities but with electron degeneracy becoming increasingly important at lower line densities. (author)

  8. Biomimetic High-Density Lipoproteins from a Gold Nanoparticle Template

    Science.gov (United States)

    Luthi, Andrea Jane

    For hundreds of years the field of chemistry has looked to nature for inspiration and insight to develop novel solutions for the treatment of human diseases. The ability of chemists to identify, mimic, and modifiy small molecules found in nature has led to the discovery and development of many important therapeutics. Chemistry on the nanoscale has made it possible to mimic natural, macromolecular structures that may also be useful for understanding and treating diseases. One example of such a structure is high-density lipoprotein (HDL). The goal of this work is to use a gold nanoparticle (Au NP) as a template to synthesize functional mimics of HDL and characterize their structure and function. Chapter 1 details the structure and function of natural HDL and how chemistry on the nanoscale provides new strategies for mimicking HDL. This Chapter also describes the first examples of using nanoparticles to mimic HDL. Chapter 2 reports the synthesis and characterization of biomimetic HDL using different sizes of Au NPs and different surface chemistries and how these variables can be used to tailor the properties of biomimetic HDL. From these studies the optimal strategy for synthesizing biomimetic HDL was determined. In Chapter 3, the optimization of the synthesis of biomimetic HDL is discussed as well as a full characterization of its structure. In addition, the work in this chapter shows that biomimetic HDL can be synthesized on a large scale without alterations to its structure or function. Chapter 4 focuses on understanding the pathways by which biomimetic HDL accepts cholesterol from macrophage cells. The results of these studies demonstrate that biomimetic HDL is able to accept cholesterol by both active and passive pathways of cholesterol efflux. In Chapter 5 the preliminary results of in vivo studies to characterize the pharmacokinetics and pharmacodynamics of biomimetic HDL are presented. These studies suggest that biomimetic HDL traffics through tissues prone to

  9. Human endothelial progenitor cells internalize high-density lipoprotein.

    Directory of Open Access Journals (Sweden)

    Kaemisa Srisen

    Full Text Available Endothelial progenitor cells (EPCs originate either directly from hematopoietic stem cells or from a subpopulation of monocytes. Controversial views about intracellular lipid traffic prompted us to analyze the uptake of human high density lipoprotein (HDL, and HDL-cholesterol in human monocytic EPCs. Fluorescence and electron microscopy were used to investigate distribution and intracellular trafficking of HDL and its associated cholesterol using fluorescent surrogates (bodipy-cholesterol and bodipy-cholesteryl oleate, cytochemical labels and fluorochromes including horseradish peroxidase and Alexa Fluor® 568. Uptake and intracellular transport of HDL were demonstrated after internalization periods from 0.5 to 4 hours. In case of HDL-Alexa Fluor® 568, bodipy-cholesterol and bodipy-cholesteryl oleate, a photooxidation method was carried out. HDL-specific reaction products were present in invaginations of the plasma membrane at each time of treatment within endocytic vesicles, in multivesicular bodies and at longer periods of uptake, also in lysosomes. Some HDL-positive endosomes were arranged in form of "strings of pearl"- like structures. HDL-positive multivesicular bodies exhibited intensive staining of limiting and vesicular membranes. Multivesicular bodies of HDL-Alexa Fluor® 568-treated EPCs showed multilamellar intra-vacuolar membranes. At all periods of treatment, labeled endocytic vesicles and organelles were apparent close to the cell surface and in perinuclear areas around the Golgi apparatus. No HDL-related particles could be demonstrated close to its cisterns. Electron tomographic reconstructions showed an accumulation of HDL-containing endosomes close to the trans-Golgi-network. HDL-derived bodipy-cholesterol was localized in endosomal vesicles, multivesicular bodies, lysosomes and in many of the stacked Golgi cisternae and the trans-Golgi-network Internalized HDL-derived bodipy-cholesteryl oleate was channeled into the lysosomal

  10. High density internal transport barriers for burning plasma operation

    Energy Technology Data Exchange (ETDEWEB)

    Ridolfini, V Pericoli [Associazione EURATOM-ENEA sulla Fusione, CR Frascati, Rome (Italy); Barbato, E [Associazione EURATOM-ENEA sulla Fusione, CR Frascati, Rome (Italy); Buratti, P [Associazione EURATOM-ENEA sulla Fusione, CR Frascati, Rome (Italy)] (and others)

    2005-12-15

    A tokamak plasma with internal transport barriers (ITBs) is the best candidate for a steady ITER operation, since the high energy confinement allows working at plasma currents (I{sub p}) lower than the reference scenario. To build and sustain an ITB at the ITER high density ({>=}10{sup 20} m{sup -3}) and largely dominant electron (e{sup -}) heating is not trivial in most existing tokamaks. FTU can instead meet both requests, thanks to its radiofrequency heating systems, lower hybrid (LH, up to 1.9 MW) and electron cyclotron (EC up to 1.2 MW). By the combined use of them, ITBs are obtained up to peak densities n{sub e0} > 1.3 x 10{sup 20} m{sup -3}, with central e{sup -} temperatures T{sub e0} {approx} 5.5 keV, and are sustained for as long as the heating pulse is applied (>35 confinement times, {tau}{sub E}). At n{sub e0} {approx} 0.8 x 10{sup 20} m{sup -3} T{sub e0} can be larger than 11 keV. Almost full current drive (CD) and an overall good steadiness is attained within about one {tau}{sub E}, 20 times faster than the ohmic current relaxation time. The ITB extends over a central region with an almost flat or slightly reversed q profile and q{sub min} {approx} 1.3 that is fully sustained by off-axis lower hybrid current drive. Consequent to this is the beneficial good alignment of the bootstrap current, generated by the ITB large pressure gradients, with the LH driven current. Reflectometry shows a clear change in the turbulence close to the ITB radius, consistent with the reduced e{sup -} transport. Ions (i{sup +}) are significantly heated via collisions, but thermal equilibrium with electrons cannot be attained since the e{sup -}-i{sup +} equipartition time is always 4-5 times longer than {tau}{sub E}. No degradation of the overall ion transport, rather a reduction of the i{sup +} heat diffusivity, is observed inside the ITB. The global confinement has been improved up to 1.6 times over the scaling predictions. The ITB radius can be controlled by adjusting the

  11. Electron Cyclotron Resonance Heating of a High-Density Plasma

    DEFF Research Database (Denmark)

    Hansen, F. Ramskov

    1986-01-01

    Various schemes for electron cyclotron resonance heating of tokamak plasmas with the ratio of electron plasma frequency to electron cyclotron frequency, "»pe/^ce* larger than 1 on axis, are investigated. In particular, a mode conversion scheme is investigated using ordinary waves at the fundamental...... of the electron cyclotron frequency. These are injected obliquely from the outside of the tokamak near an optimal angle to the magnetic field lines. This method involves two mode conversions. The ordinary waves are converted into extraordinary waves near the plasma cut-off layer. The extraordinary waves...... are subsequently converted into electrostatic electron Bernstein waves at the upper hybrid resonance layer, and the Bernstein waves are completely absorbed close to the plasma centre. Results are presented from ray-tracinq calculations in full three-dimensional geometry using the dispersion function for a hot non...

  12. MEASUREMENT AND CORRELATION OF THE MASS TRANSFER COEFFICIENT FOR A LIQUID-LIQUID SYSTEM WITH HIGH DENSITY DIFFERENCE

    Directory of Open Access Journals (Sweden)

    Zhixian Huang

    Full Text Available Abstract To investigate the mass transfer behavior of a liquid-liquid system with high density difference (∆ρ≈500 kg/m3, single drop experiments were performed by using the ternary chloroform-ethanol-water system. The mass transfer direction was from the dispersed phase to the continuous phase, while the aqueous phase was dispersed in chloroform to generate drops. The influences of drop diameter, initial solute concentration and temperature on the mass transfer were investigated. The effects of the drop diameter and initial solute concentration on interfacial instability of droplets hanging in the continuous phase were also observed. For the purpose of correlation, a mass transfer enhancement factor F was introduced and then correlated as a function of dimensionless variables. The modified correlation from the mass transfer coefficient model was found to fit well with the experimental values.

  13. High density plasma heating in the Tokamak à configuration variable

    International Nuclear Information System (INIS)

    Curchod, L.

    2011-04-01

    The Tokamak à Configuration Variable (TCV) is a medium size magnetic confinement thermonuclear fusion experiment designed for the study of the plasma performances as a function of its shape. It is equipped with a high power and highly flexible electron cyclotron heating (ECH) and current drive (ECCD) system. Up to 3 MW of 2 nd harmonic EC power in ordinary (O 2 ) or extraordinary (X 2 ) polarization can be injected from TCV low-field side via six independently steerable launchers. In addition, up to 1.5 MW of 3 rd harmonic EC power (X 3 ) can be launched along the EC resonance from the top of TCV vacuum vessel. At high density, standard ECH and ECCD are prevented by the appearance of a cutoff layer screening the access to the EC resonance at the plasma center. As a consequence, less than 50% of TCV density operational domain is accessible to X 2 and X 3 ECH. The electron Bernstein waves (EBW) have been proposed to overcome this limitation. EBW is an electrostatic mode propagating beyond the plasma cutoff without upper density limit. Since it cannot propagate in vacuum, it has to be excited by mode conversion of EC waves in the plasma. Efficient electron Bernstein waves heating (EBH) and current drive (EBCD) were previously performed in several fusion devices, in particular in the W7-AS stellarator and in the MAST spherical tokamak. In TCV, the conditions for an efficient O-X-B mode conversion (i.e. a steep density gradient at the O 2 plasma cutoff) are met at the edge of high confinement (H-mode) plasmas characterized by the appearance of a pedestal in the electron temperature and density profiles. TCV experiments have demonstrated the first EBW coupling to overdense plasmas in a medium aspect-ratio tokamak via O-X-B mode conversion. This thesis work focuses on several aspects of ECH and EBH in low and high density plasmas. Firstly, the experimental optimum angles for the O-X-B mode conversion is successfully compared to the full-wave mode conversion calculation

  14. Treat upgrade fuel fabrication

    International Nuclear Information System (INIS)

    Davidson, K.V.; Schell, D.H.

    1979-01-01

    An extrusion and thermal treatment process was developed to produce graphite fuel rods containing a dispersion of enriched UO 2 . These rods will be used in an upgraded version of the Transient Reactor Test Facility (TREAT). The improved fuel provides a higher graphite matrix density, better fuel dispersion and higher thermal capabilities than the existing fuel

  15. Coupled neutronics and thermal hydraulics of high density cores for FRM II

    Energy Technology Data Exchange (ETDEWEB)

    Breitkreutz, Harald

    2011-03-04

    According to the 'Verwaltungsvereinbarung zwischen Bund und Land vom 30.5.2003' and its updating on 13.11.2010, the Forschungs-Neutronenquelle Heinz Maier-Leibnitz, Frm II, has to convert its fuel element to an uranium enrichment which is significantly lower than the current 93%, in case this is economically reasonable and doesn't impact the reactor performance immoderate. In the framework of this conversion, new calculations regarding neutronics and thermal hydraulics for the anticipated core configurations have to be made. The computational power available nowadays allows for detailed 3D calculations, on the neutronic as well as on the thermal hydraulic side. In this context, a new program system, 'X{sup 2}', was developed. It couples the Monte Carlo code McnpX, the computational fluid dynamics code Cfx and the burn-up code sequence MonteBurns. The codes were modified and extended to meet the requirements of the coupled calculation concept. To verify the new program system, highly detailed calculations for the current fuel element were made and compared to simulations and measurements that were performed in the past. The results strengthen the works performed so far and show that the original, conservative approach overestimates all critical thermal hydraulic values. Using the CFD software, effects like the impact of the combs that fix the fuel plates and the pressure drop at the edges of the fuel plates were studied in great detail for the first time. Afterwards, a number of possible new fuel elements with lower enrichment, based on disperse and monolithic UMo (uranium with 8 wt.-% Mo) were analysed. A number of straight-forward conversion scenarios was discussed, showing that a further compaction of the fuel element, an extended cycle length or an increased reactor power is needed to compensate the flux loss, which is caused by the lower enrichment. This flux loss is in excess of 7%. The discussed new fuel elements include a 50

  16. Coupled neutronics and thermal hydraulics of high density cores for FRM II

    International Nuclear Information System (INIS)

    Breitkreutz, Harald

    2011-01-01

    According to the 'Verwaltungsvereinbarung zwischen Bund und Land vom 30.5.2003' and its updating on 13.11.2010, the Forschungs-Neutronenquelle Heinz Maier-Leibnitz, Frm II, has to convert its fuel element to an uranium enrichment which is significantly lower than the current 93%, in case this is economically reasonable and doesn't impact the reactor performance immoderate. In the framework of this conversion, new calculations regarding neutronics and thermal hydraulics for the anticipated core configurations have to be made. The computational power available nowadays allows for detailed 3D calculations, on the neutronic as well as on the thermal hydraulic side. In this context, a new program system, 'X 2 ', was developed. It couples the Monte Carlo code McnpX, the computational fluid dynamics code Cfx and the burn-up code sequence MonteBurns. The codes were modified and extended to meet the requirements of the coupled calculation concept. To verify the new program system, highly detailed calculations for the current fuel element were made and compared to simulations and measurements that were performed in the past. The results strengthen the works performed so far and show that the original, conservative approach overestimates all critical thermal hydraulic values. Using the CFD software, effects like the impact of the combs that fix the fuel plates and the pressure drop at the edges of the fuel plates were studied in great detail for the first time. Afterwards, a number of possible new fuel elements with lower enrichment, based on disperse and monolithic UMo (uranium with 8 wt.-% Mo) were analysed. A number of straight-forward conversion scenarios was discussed, showing that a further compaction of the fuel element, an extended cycle length or an increased reactor power is needed to compensate the flux loss, which is caused by the lower enrichment. This flux loss is in excess of 7%. The discussed new fuel elements include a 50% enriched disperse UMo core with

  17. Nondestructive, energy-dispersive x-ray fluorescence analysis of product-stream concentrations from reprocessed LWR fuels

    International Nuclear Information System (INIS)

    Camp, D.C.; Ruhter, W.D.; Benjamin, S.

    1979-01-01

    Energy-dispersive x-ray fluorescence analysis can be used for quantitative on-line monitoring of the product concentrations in single- or dual-element process streams in a reprocessing plant. The 122-keV gamma ray from 57 Co is used to excite the K x-rays of uranium and/or plutonium in nitric acid solution streams. A collimated HPGe detector is used to measure the excited x-ray intensities. Net solution radioactivity may be measured by eclipsing the exciting radiation, or by measuring it simultaneously with a second detector. The technique is nondestructive and noninvasive, and is easily adapted directly to pipes containing the solution of interest. The dynamic range of the technique extends from below 1 to 500 g/l. Measurement times depend on concentration, but better than 1% counting statistics can be obtained in 100 s for 400 g/l concentrations, and in 1000 s for as little as 10 g/l. Calibration accuracies of 0.3% or better over the entire dynamic range can be achieved easily using carefully prepared standards. Computer-based analysis equipment allows concentration changes in flowing streams to be dynamically monitored. Changes in acid normality of the stream will affect the concentration determined, hence it must also be determined by measuring the intensity of a transmitted 57 Co beam. The computer/disk-based pulse-height analysis system allows all necessary calculations to be done on-line. Experimental requirements for an in-plant installation or a test and evaluation are discussed

  18. Study on severe accident fuel dispersion behavior in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.; Xiang, J.Y.

    1995-01-01

    Core flow blockage events are a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel could result in core heatup and melting under full coolant flow condition. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Heat transfer between melt particle and coolant, which affects particle breakup, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to relative motion of particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. Results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also core debris tends to move together upon melting and entrainment

  19. Study on severe accident fuel dispersion behavior in the advanced neutron source reactor at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S. [Oak Ridge National Lab., TN (United States)] [and others

    1995-09-01

    Core flow blockage events have been determined to represent a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel in a few adjacent blocked coolant channels out of several hundred channels, could also result in core heatup and melting under full coolant flow condition in other coolant channels. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Hat transfer between melt particle and coolant, which affects the particle breakup characteristics, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to the relative motion of the particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. The results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also they are entrained and move together in a cloud.

  20. Dispersion and thermal interactions of molten metal fuel settling on a horizontal steel plate through a sodium pool

    International Nuclear Information System (INIS)

    Gabor, J.D.; Purviance, R.T.; Aeschlimann, R.W.; Spencer, B.W.

    1989-01-01

    Although the Integral Fast Reactor (IFR) possesses inherent safety features, an assessment of the consequences of melting of the metal fuel is necessary for risk analysis. As part of this effort an experimental study was conducted to determine the depths of sodium at 600 C required for pour streams of various molten uranium alloys (U, U-5 wt % Zr, U-10 wt % Zr, and U-10 wt % Fe) to break up and solidify. The quenched particulate material, which was in the shape of filaments and sheets, formed coolable beds because of the high voidage (∼0.9) and large particle size (∼10 mm). In a test with a 0.15-m sodium depth, the fragments from a pure uranium pour stream did not completely solidify but formed an agglomerated mass which did not fuse to the base plate. However, the agglomerated fragments of U-10 wt % Fe eutectic fused to the stainless steel base plate. An analysis of the temperature response of a 25-mm thick base plate was made by volume averaging the properties of the sodium and metal particle phases and assuming two semi-infinite solids coming into contact. Good agreement was obtained with the data during the initial 5 to 10 s of the contact period. 16 refs., 5 figs., 1 tab

  1. Effect of expanded graphite on the phase change materials of high density polyethylene/wax blends

    Energy Technology Data Exchange (ETDEWEB)

    AlMaadeed, M.A., E-mail: m.alali@qu.edu.qa [Center for Advanced Materials, Qatar University, 2713 Doha (Qatar); Labidi, Sami [Center for Advanced Materials, Qatar University, 2713 Doha (Qatar); Krupa, Igor [QAPCO Polymer Chair, Center for Advanced Materials, Qatar University, P.O. Box 2713, Doha (Qatar); Karkri, Mustapha [Université Paris-Est CERTES, 61 avenue du Général de Gaulle, 94010 Créteil (France)

    2015-01-20

    Highlights: • Expanded graphite (EG) and low melting point (42.3 °C) wax were added to HDPE to form phase change material. • EG was well dispersed in the composites and did not affect the melting or crystallization of the HDPE matrix. • EG increased the thermal stability of the composites by reducing chain mobility and inhibiting degradation. • The addition of a relatively small quantity of EG enhances the heat conduction in the composite. • HDPE/40% RT42 that contained up to 15% EG demonstrated excellent mechanical and thermal properties and can be used as PCM. - Abstract: Phase change materials fabricated from high density polyethylene (HDPE) blended with 40 or 50 wt% commercial wax (melting point of 43.08 °C) and up to 15 wt% expanded graphite (EG) were studied. Techniques including scanning electron microscope (SEM), differential scanning calorimetry (DSC), thermogravimetric analysis (TGA), and an experimental device to measure diffusivity and conductivity (DICO) were used to determine the microstructural, mechanical and thermal properties of the composites. The composites possessed good mechanical properties. Additionally, no leaching was observed during material processing or characterization. Although the Young’s modulus increased with the addition of EG, no significant changes in tensile strength were detected. The maximum Young’s modulus achieved was 650 MPa for the HDPE/40% wax composite with 15 wt% EG. The EG was well dispersed within the composites and did not affect the melting or crystallization of the HDPE matrix. The incorporation of EG increased the thermal stability of the composites by reducing chain mobility and inhibiting degradation. The intensification of thermal conductivity occurred with increasing fractions of EG, which was attributed to the high thermal conductivity of graphite. The maximum quantity of heat stored by latent heat was found for the HDPE/40% wax composite with EG. The addition of a relatively small quantity

  2. Effect of expanded graphite on the phase change materials of high density polyethylene/wax blends

    International Nuclear Information System (INIS)

    AlMaadeed, M.A.; Labidi, Sami; Krupa, Igor; Karkri, Mustapha

    2015-01-01

    Highlights: • Expanded graphite (EG) and low melting point (42.3 °C) wax were added to HDPE to form phase change material. • EG was well dispersed in the composites and did not affect the melting or crystallization of the HDPE matrix. • EG increased the thermal stability of the composites by reducing chain mobility and inhibiting degradation. • The addition of a relatively small quantity of EG enhances the heat conduction in the composite. • HDPE/40% RT42 that contained up to 15% EG demonstrated excellent mechanical and thermal properties and can be used as PCM. - Abstract: Phase change materials fabricated from high density polyethylene (HDPE) blended with 40 or 50 wt% commercial wax (melting point of 43.08 °C) and up to 15 wt% expanded graphite (EG) were studied. Techniques including scanning electron microscope (SEM), differential scanning calorimetry (DSC), thermogravimetric analysis (TGA), and an experimental device to measure diffusivity and conductivity (DICO) were used to determine the microstructural, mechanical and thermal properties of the composites. The composites possessed good mechanical properties. Additionally, no leaching was observed during material processing or characterization. Although the Young’s modulus increased with the addition of EG, no significant changes in tensile strength were detected. The maximum Young’s modulus achieved was 650 MPa for the HDPE/40% wax composite with 15 wt% EG. The EG was well dispersed within the composites and did not affect the melting or crystallization of the HDPE matrix. The incorporation of EG increased the thermal stability of the composites by reducing chain mobility and inhibiting degradation. The intensification of thermal conductivity occurred with increasing fractions of EG, which was attributed to the high thermal conductivity of graphite. The maximum quantity of heat stored by latent heat was found for the HDPE/40% wax composite with EG. The addition of a relatively small quantity

  3. Curaua fiber reinforced high-density polyethylene composites: effect of impact modifier and fiber loading

    Directory of Open Access Journals (Sweden)

    Jaqueline Albano de Morais

    Full Text Available Abstract Short fibers are used in thermoplastic composites to increase their tensile and flexural resistance; however, it often decreases impact resistance. Composites with short vegetal fibers are not an exception to this behavior. The purpose of this work is to produce a vegetal fiber reinforced composite with improved tensile and impact resistance in relation to the polymer matrix. We used poly(ethylene-co-vinyl acetate, EVA, to recover the impact resistance of high density polyethylene, HDPE, reinforced with Curauá fibers, CF. Blends and composites were processed in a corotating twin screw extruder. The pure polymers, blends and composites were characterized by differential scanning calorimetry, thermogravimetry, infrared spectroscopy, scanning electron microscopy, tensile mechanical properties and Izod impact resistance. EVA used as impact modifier in the HDPE matrix exhibited a co-continuous phase and in the composites the fibers were homogeneously dispersed. The best combination of mechanical properties, tensile, flexural and impact, were obtained for the formulations of composites with 20 wt. % of CF and 20 to 40 wt. % of EVA. The composite prepared with 20 wt. % EVA and containing 30 wt. % of CF showed impact resistance comparable to pure HDPE and improved tensile and flexural mechanical properties.

  4. Diffusion of multiwall carbon nanotubes (MWCNTs) through a high density polyethylene (HDPE) geomembrane.

    Science.gov (United States)

    Saheli, P T; Rowe, R K; Petersen, E J; O'Carroll, D M

    2017-05-01

    The new applications for carbon nanotubes (CNTs) in various fields and consequently their greater production volume have increased their potential release to the environment. Landfills are one of the major locations where carbon nanotubes are expected to be disposed and it is important to ensure that they can limit the release of CNTs. Diffusion of multiwall carbon nanotubes (MWCNTs) dispersed in an aqueous media through a high-density polyethylene (HDPE) geomembrane (as a part of the landfill barrier system) was examined. Based on the laboratory tests, the permeation coefficient was estimated to be less than 5.1×10 -15 m 2 /s. The potential performance of a HDPE geomembrane and geosynthetic clay liner (GCL) as parts of a composite liner in containing MWCNTs was modelled for six different scenarios. The results suggest that the low value of permeation coefficient of an HDPE geomembrane makes it an effective diffusive barrier for MWCNTs and by keeping the geomembrane defects to minimum during the construction (e.g., number of holes and length of wrinkles) a composite liner commonly used in municipal solid waste landfills will effectively contain MWCNTs.

  5. Modeling benzene permeation through drinking water high density polyethylene (HDPE) pipes.

    Science.gov (United States)

    Mao, Feng; Ong, Say Kee; Gaunt, James A

    2015-09-01

    Organic compounds such as benzene, toluene, ethyl benzene and o-, m-, and p-xylene from contaminated soil and groundwater may permeate through thermoplastic pipes which are used for the conveyance of drinking water in water distribution systems. In this study, permeation parameters of benzene in 25 mm (1 inch) standard inside dimension ratio (SIDR) 9 high density polyethylene (HDPE) pipes were estimated by fitting the measured data to a permeation model based on a combination of equilibrium partitioning and Fick's diffusion. For bulk concentrations between 6.0 and 67.5 mg/L in soil pore water, the concentration-dependent diffusion coefficients of benzene were found to range from 2.0×10(-9) to 2.8×10(-9) cm2/s while the solubility coefficient was determined to be 23.7. The simulated permeation curves of benzene for SIDR 9 and SIDR 7 series of HDPE pipes indicated that small diameter pipes were more vulnerable to permeation of benzene than large diameter pipes, and the breakthrough of benzene into the HDPE pipe was retarded and the corresponding permeation flux decreased with an increase of the pipe thickness. HDPE pipes exposed to an instantaneous plume exhibited distinguishable permeation characteristics from those exposed to a continuous source with a constant input. The properties of aquifer such as dispersion coefficients (DL) also influenced the permeation behavior of benzene through HDPE pipes.

  6. Activity and Stability of Dispersed Multi Metallic Pt-based Catalysts for CO Tolerance in Proton Exchange Membrane Fuel Cell Anodes.

    Science.gov (United States)

    Hassan, Ayaz; Ticianelli, Edson A

    2018-01-01

    Studies aiming at improving the activity and stability of dispersed W and Mo containing Pt catalysts for the CO tolerance in proton exchange membrane fuel cell (PEMFC) anodes are revised for the following catalyst systems: (1) a carbon supported PtMo electrocatalyst submitted to heat treatments; (2) Pt and PtMo nanoparticles deposited on carbon-supported molybdenum carbides (Mo2C/C); (3) ternary and quaternary materials formed by PtMoFe/C, PtMoRu/C and PtMoRuFe/C and; (4) Pt nanoparticles supported on tungsten carbide/carbon catalysts and its parallel evaluation with carbon supported PtW catalyst. The heat-treated (600 oC) Pt-Mo/C catalyst showed higher hydrogen oxidation activity in the absence and in the presence of CO and better stability, compared to all other Mo-containing catalysts. PtMoRuFe, PtMoFe, PtMoRu supported on carbon and Pt supported on Mo2C/C exhibited similar CO tolerances but better stability, as compared to as-prepared PtMo supported on carbon. Among the tungsten-based catalysts, tungsten carbide supported Pt catalyst showed reasonable performance and reliable stability in comparison to simple carbon supported PtW catalyst, though an uneven level of catalytic activity towards H2 oxidation in presence of CO is observed for the former as compared to Mo containing catalyst. However, a small dissolution of Mo, Ru, Fe and W from the anodes and their migration toward cathodes during the cell operation is observed. These results indicate that the fuel cell performance and stability has been improved but not yet totally resolved.

  7. Activity and Stability of Dispersed Multi Metallic Pt-based Catalysts for CO Tolerance in Proton Exchange Membrane Fuel Cell Anodes

    Directory of Open Access Journals (Sweden)

    AYAZ HASSAN

    2018-04-01

    Full Text Available ABSTRACT Studies aiming at improving the activity and stability of dispersed W and Mo containing Pt catalysts for the CO tolerance in proton exchange membrane fuel cell (PEMFC anodes are revised for the following catalyst systems: (1 a carbon supported PtMo electrocatalyst submitted to heat treatments; (2 Pt and PtMo nanoparticles deposited on carbon-supported molybdenum carbides (Mo2C/C; (3 ternary and quaternary materials formed by PtMoFe/C, PtMoRu/C and PtMoRuFe/C and; (4 Pt nanoparticles supported on tungsten carbide/carbon catalysts and its parallel evaluation with carbon supported PtW catalyst. The heat-treated (600 oC Pt-Mo/C catalyst showed higher hydrogen oxidation activity in the absence and in the presence of CO and better stability, compared to all other Mo-containing catalysts. PtMoRuFe, PtMoFe, PtMoRu supported on carbon and Pt supported on Mo2C/C exhibited similar CO tolerances but better stability, as compared to as-prepared PtMo supported on carbon. Among the tungsten-based catalysts, tungsten carbide supported Pt catalyst showed reasonable performance and reliable stability in comparison to simple carbon supported PtW catalyst, though an uneven level of catalytic activity towards H2 oxidation in presence of CO is observed for the former as compared to Mo containing catalyst. However, a small dissolution of Mo, Ru, Fe and W from the anodes and their migration toward cathodes during the cell operation is observed. These results indicate that the fuel cell performance and stability has been improved but not yet totally resolved.

  8. Applications of Neutron Scattering in the Chemical Industry: Proton Dynamics of Highly Dispersed Materials, Characterization of Fuel Cell Catalysts, and Catalysts from Large-Scale Chemical Processes

    Science.gov (United States)

    Albers, Peter W.; Parker, Stewart F.

    The attractiveness of neutron scattering techniques for the detailed characterization of materials of high degrees of dispersity and structural complexity as encountered in the chemical industry is discussed. Neutron scattering picks up where other analytical methods leave off because of the physico-chemical properties of finely divided products and materials whose absorption behavior toward electromagnetic radiation and electrical conductivity causes serious problems. This is demonstrated by presenting typical applications from large-scale production technology and industrial catalysis. These include the determination of the proton-related surface chemistry of advanced materials that are used as reinforcing fillers in the manufacture of tires, where interrelations between surface chemistry, rheological properties, improved safety, and significant reduction of fuel consumption are the focus of recent developments. Neutron scattering allows surface science studies of the dissociative adsorption of hydrogen on nanodispersed, supported precious metal particles of fuel cell catalysts under in situ loading at realistic gas pressures of about 1 bar. Insight into the occupation of catalytically relevant surface sites provides valuable information about the catalyst in the working state and supplies essential scientific input for tailoring better catalysts by technologists. The impact of deactivation phenomena on industrial catalysts by coke deposition, chemical transformation of carbonaceous deposits, and other processes in catalytic hydrogenation processes that result in significant shortening of the time of useful operation in large-scale plants can often be traced back in detail to surface or bulk properties of catalysts or materials of catalytic relevance. A better understanding of avoidable or unavoidable aspects of catalyst deactivation phenomena under certain in-process conditions and the development of effective means for reducing deactivation leads to more energy

  9. Influence sur les imbrûlés solides de composés métalliques particuliers et du taux de dispersion des asphaltènes dans les fuels lourds Influence of Unburned Solids Made of Unusual Metal Compounds and of the Asphaltene Dispersion Rate in Heavy Fuel Oils

    Directory of Open Access Journals (Sweden)

    Audibert F.

    2006-11-01

    Full Text Available La diversité croissante des origines de bruts a conduit, en combustion, a prendre en considération dans les modèles la teneur en métaux en plus du carbone Conradson ou des asphaltènes au C7 dans les fuels lourds. De tels modèles ont été développés par Exxon (1979 et Shell (1981 notamment. Des travaux récents faits à l'Institut Français du Pétrole (IFP sur une chaudière de 2 MW ont montré l'influence de composés métalliques particuliers présents dans le fuel oil sous forme de sulfures imprégnant des particules de carbone poreux. Ces microparticules peuvent être générées lorsque l'on applique des conditions opératoires sévères à la viscoréduction de fuels résiduaires en présence d'hydrogène et d'un catalyseur approprié. Ces microparticules se sont révélées très actives en combustion et ont montré que la concentration du métal n'était pas le seul facteur à prendre en considération mais que la manière dont il était combiné pouvait être prépondérante. Pour étendre le domaine d'application des modèles, d'autres paramètres comme les conditions de fonctionnement de la chaudière et l'atomisation du fuel ont été pris en compte en plus des paramètres d'influence propres au fuel (travaux du Laboratoire Énergie du MIT publications 1986. En ce qui concerne la prédiction d'émissions particulaires une méthode complémentaire aux tests de résidu de Conradson et d'insolubilité à l'heptane normal a été appliquée à I'IFP dans le cadre d'un programme de valorisation des huiles lourdes en association avec les compagnies pétrolières Elf et Total France. Cette méthode consiste à évaluer à la microscopie électronique par transmission le taux de dispersion des asphaltènes, selon une méthode développée par Total France (M. Peyrot. Il existe une relation entre ce taux de dispersion et l'émission particulaire. Ce phénomène a été nettement observé dans le cas de fuels lourds constitués par

  10. Experimental operation of the RA reactor with 4 fuel channels containing 80% enriched dispersion fuel - Operational Report; Radni izvestaj - Eksperimentalna kampanja reaktora RA sa 4 kanala sa 80% obogacenim disperzionim gorivom

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Milosevic, M; Cupac, S; Kozomara, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-12-15

    Start of utilization of the new 80% enriched dispersion nuclear fuel is underway in the RA reactor core. Both economic and technical analyses were in favor of introducing the new fuel elements gradually into the RA reactor core. Thus overall theoretical and experimental analyses as well as other preparations are directed to transition regime based on gradual introducing of new fuel into the core, i.e. reactor core with two types of fuel. The objective of these analyses and preparation is establishment of conditions for safe reactor operation during transition period. The analyses and preparations are almost completed. The experimental data about fuel burnup during a time period of operation at nominal power i.e. daily decrease of excess reactivity is missing. This data is needed for planning the refueling (quantity of fresh fuel and frequency of refueling) during the transient period. This data can be obtained only by normal operation of the reactor during a period of time significantly longer than the period of attaining equilibrium poisoning, as time between two D{sub 2}O condensate overflows into the RA reactor core. Thus a ten day experimental campaign was planned to be done in December 1976. This report presents the most important results of safety analyses and preparation which show that, during this experimental period, the reactor operation is absolutely safe taking into account the most important parameters influencing reactor safety, as reactivity, thermal and temperature limits for fuel and the reactor, etc. Data to be obtained during this experimental campaign are significant because they would enable definition of future supply of fresh fuel during the transition period. [Serbo-Croat] Predstoji pocetak koriscenja novog 80% obogacenog uranskog disperzionog goriva u reaktoru RA. Ekonomske i tehnicke analize dale su prednost postupku postepenog uvodjenja novog goriva u reaktor RA. Prema tome, obimne teorijske i eksperimentalne analize i druge pripreme

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  12. Production of High Density Aviation Fuels via Novel Zeolite Catalyst Routes

    Science.gov (United States)

    1989-10-23

    108. 325 144. Scherzer, J., "The Preparation and Characterization of Aluminum- deficient Zeolites," in Catalytic Materials: Relationship between...10),TIC(100),TFC(100),TI(100),TF(100) 4 DIMENSION Ti(iO),T2(1O),T3(1O),TK1(1O), TK2 (1O),TK3(O) 5 DIMENSION DPISOF(100),TIT2(80,5),DPISOI(IOO),NVSB(100...0CP3(N))/1O.O 32 RSC(N) - (CPI(N)/TK1(N)-CP2(N)/ TK2 (N))* TK2 (N)/CP2(N) 33 RCELL(N)-(CP2(N)/ TK2 (N)-CP3(N)/TK3(N))*TC/CP3(N) 34 ARSC-ARSC+RSC(N) 35