WorldWideScience

Sample records for density dispersion fuel

  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. 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)

  9. 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)

  10. 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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  12. 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)

  13. 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)

  14. 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)

  15. 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.

  16. 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

  17. 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)

  18. 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.

  19. 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

  20. 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.

  1. 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)

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

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

    Science.gov (United States)

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

    2016-06-01

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

  9. 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)

  10. 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.

  11. 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

  12. 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)

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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)

  20. 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)

  1. 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

  2. Axon density and axon orientation dispersion in children born preterm

    NARCIS (Netherlands)

    Kelly, Claire E.; Thompson, Deanne K.; Chen, Jian; Leemans, Alexander; Adamson, Christopher L.; Inder, Terrie E.; Cheong, Jeanie L Y; Doyle, Lex W.; Anderson, Peter J.

    2016-01-01

    Background Very preterm birth (VPT, <32 weeks' gestation) is associated with altered white matter fractional anisotropy (FA), the biological basis of which is uncertain but may relate to changes in axon density and/or dispersion, which can be measured using Neurite Orientation Dispersion and Density

  3. 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)

  4. Dispersal patterns of red foxes relative to population density

    Science.gov (United States)

    Allen, Stephen H.; Sargeant, Alan B.

    1993-01-01

    Factors affecting red fox (Vulpes vulpes) dispersal patterns are poorly understood but warranted investigation because of the role of dispersal in rebuilding depleted populations and transmission of diseases. We examined dispersal patterns of red foxes in North Dakota based on recoveries of 363 of 854 foxes tagged as pups and relative to fox density. Foxes were recovered up to 8.6 years after tagging; 79% were trapped or shot. Straight-line distances between tagging and recovery locations ranged from 0 to 302 km. Mean recovery distances increased with age and were greater for males than females, but longest individual recovery distances were by females. Dispersal distances were not related to population density for males (P = 0.36) or females (P = 0.96). The proportion of males recovered that dispersed was inversely related to population density (r = -0.94; n = 5; P = 0.02), but not the proportion of females (r = -0.49; n = 5; P = 0.40). Dispersal directions were not uniform for either males (P = 0.003) or females (P = 0.006); littermates tended to disperse in similar directions (P = 0.09). A 4-lane interstate highway altered dispersal directions (P = 0.001). Dispersal is a strong innate behavior of red foxes (especially males) that results in many individuals of both sexes traveling far from natal areas. Because dispersal distance was unaffected by fox density, populations can be rebuilt and diseases transmitted long distances regardless of fox abundance.

  5. 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)

  6. 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

  7. 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.

  8. 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.

  9. 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)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  11. 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

  12. Density determination of sintered ceramic nuclear fuel materials

    International Nuclear Information System (INIS)

    Landspersky, H.; Medek, J.

    1980-01-01

    The feasibility was tested of using solids for pycnometric determination of the density of uranium dioxide-based sintered ceramic fuel materials manufactured by the sol-gel method in the shape of spherical particles of 0.7 to 1.0 mm in size and of particles smaller than 200 μm. For fine particles, this is the only usable method of determining their density which is a very important parameter of the fine fraction when it is employed for the manufacture of fuel elements by vibration compacting. The method consists in compacting a mixture of pycnometric material and dispersed particles of uranium dioxide, determining the size and weight of the compact, and in calculating the density of the material measured from the weight of the oxide sample in the mixture. (author)

  13. 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

  14. 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)

  15. 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.

  16. 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.)

  17. 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)

  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. 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)

  1. 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)

  2. 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

  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. 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)

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

    International Nuclear Information System (INIS)

    Xing Zhonghu; Ying Shihao

    2000-01-01

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

  7. 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

  8. High power density carbonate fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Yuh, C.; Johnsen, R.; Doyon, J.; Allen, J. [Energy Research Corp., Danbury, CT (United States)

    1996-12-31

    Carbonate fuel cell is a highly efficient and environmentally clean source of power generation. Many organizations worldwide are actively pursuing the development of the technology. Field demonstration of multi-MW size power plant has been initiated in 1996, a step toward commercialization before the turn of the century, Energy Research Corporation (ERC) is planning to introduce a 2.85MW commercial fuel cell power plant with an efficiency of 58%, which is quite attractive for distributed power generation. However, to further expand competitive edge over alternative systems and to achieve wider market penetration, ERC is exploring advanced carbonate fuel cells having significantly higher power densities. A more compact power plant would also stimulate interest in new markets such as ships and submarines where space limitations exist. The activities focused on reducing cell polarization and internal resistance as well as on advanced thin cell components.

  9. 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.)

  10. 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...

  11. 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)

  12. 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)

  13. 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.

  14. 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

  15. 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

  16. 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.

  17. 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.

  18. 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.

  19. 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)

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. 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.

  5. High energy-density liquid rocket fuel performance

    Science.gov (United States)

    Rapp, Douglas C.

    1990-01-01

    A fuel performance database of liquid hydrocarbons and aluminum-hydrocarbon fuels was compiled using engine parametrics from the Space Transportation Engine Program as a baseline. Propellant performance parameters are introduced. General hydrocarbon fuel performance trends are discussed with respect to hydrogen-to-carbon ratio and heat of formation. Aluminum-hydrocarbon fuel performance is discussed with respect to aluminum metal loading. Hydrocarbon and aluminum-hydrocarbon fuel performance is presented with respect to fuel density, specific impulse and propellant density specific impulse.

  6. 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.

  7. 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)

  8. 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)

  9. 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.

  10. 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

  11. 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)

  12. 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)

  13. 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

  14. 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.

  15. 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.)

  16. Basic research on high-uranium density fuels for research and test reactors

    International Nuclear Information System (INIS)

    Ugajin, M.; Itoh, A.; Akabori, M.

    1992-01-01

    High-uranium density fuels, uranium silicides (U 3 Si 2 , U 3 Si) and U 6 Me-type uranium alloys (Me = Fe, Mn, Ni), were prepared and examined metallurgically as low-enriched uranium (LEU) fuels for research and test reactors. Miniature aluminum-dispersion plate-type fuel (miniplate) and aluminum-clad disk-type fuel specimens were fabricated and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Fuel-aluminum compatibility tests were conducted to elucidate the extent of reaction and to identify reaction products. The relative stability of the fuels in an aluminum matrix was established at 350degC or above. Experiments were also performed to predict the chemical form of the solid fission-products in the uranium silicide (U 3 Si 2 ) simulating a high burnup anticipated for reactor service. (author)

  17. 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

  18. 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.)

  19. Estimating the effect of urban density on fuel demand

    Energy Technology Data Exchange (ETDEWEB)

    Karathodorou, Niovi; Graham, Daniel J. [Imperial College London, London, SW7 2AZ (United Kingdom); Noland, Robert B. [Rutgers University, New Brunswick, NJ 08901 (United States)

    2010-01-15

    Much of the empirical literature on fuel demand presents estimates derived from national data which do not permit any explicit consideration of the spatial structure of the economy. Intuitively we would expect the degree of spatial concentration of activities to have a strong link with transport fuel consumption. The present paper addresses this theme by estimating a fuel demand model for urban areas to provide a direct estimate of the elasticity of demand with respect to urban density. Fuel demand per capita is decomposed into car stock per capita, fuel consumption per kilometre and annual distance driven per car per year. Urban density is found to affect fuel consumption, mostly through variations in the car stock and in the distances travelled, rather than through fuel consumption per kilometre. (author)

  20. 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

  1. 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

  2. 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)

  3. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Cirila Tacconi de

    2005-07-01

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

  5. Gamma densitometer for measuring Pu density in fuel tubes

    International Nuclear Information System (INIS)

    Winn, W.G.

    1982-01-01

    A fuel-gamma-densitometer (FGD) has been developed to examine nondestructively the uniformity of plutonium in aluminum-clad fuel tubes at the Savannah River Plant (SRP). The monitoring technique is γ-ray spectroscopy with a lead-collimated Ge(Li) detector. Plutonium density is correlated with the measured intensity of the 208 keV γ-ray from 237 U (7d) of the 241 Pu (15y) decay chain. The FGD measures the plutonium density within 0.125- or 0.25-inch-diameter areas of the 0.133- to 0.183-inch-thick tube walls. Each measurement yields a density ratio that relates the plutonium density of the measured area to the plutonium density in normal regions of the tube. The technique was used to appraise a series of fuel tubes to be irradated in an SRP reactor. High-density plutonium areas were initially identified by x-ray methods and then examined quantitatively with the FGD. The FGD reliably tested fuel tubes and yielded density ratios over a range of 0.0 to 2.5. FGD measurements examined (1) nonuniform plutonium densities or hot spots, (2) uniform high-density patches, and (3) plutonium density distribution in thin cladding regions. Measurements for tubes with known plutonium density agreed with predictions to within 2%. Attenuation measurements of the 208-keV γ-ray passage through the tube walls agreed to within 2 to 3% of calculated predictions. Collimator leakage measurements agreed with model calculations that predicted less than a 1.5% effect on plutonium density ratios. Finally, FGD measurements correlated well with x-ray transmission and fluoroscopic measurements. The data analysis for density ratios involved a small correction of about 10% for γ-shielding within the fuel tube. For hot spot examinations, limited information for this correction dictated a density ratio uncertainty of 3 to 5%

  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. High volumetric power density, non-enzymatic, glucose fuel cells.

    Science.gov (United States)

    Oncescu, Vlad; Erickson, David

    2013-01-01

    The development of new implantable medical devices has been limited in the past by slow advances in lithium battery technology. Non-enzymatic glucose fuel cells are promising replacement candidates for lithium batteries because of good long-term stability and adequate power density. The devices developed to date however use an "oxygen depletion design" whereby the electrodes are stacked on top of each other leading to low volumetric power density and complicated fabrication protocols. Here we have developed a novel single-layer fuel cell with good performance (2 μW cm⁻²) and stability that can be integrated directly as a coating layer on large implantable devices, or stacked to obtain a high volumetric power density (over 16 μW cm⁻³). This represents the first demonstration of a low volume non-enzymatic fuel cell stack with high power density, greatly increasing the range of applications for non-enzymatic glucose fuel cells.

  8. 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

  9. 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

  10. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-01

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

  12. High seed dispersal ability of Pinus canariensis in stands of contrasting density inferred from genotypic data

    Directory of Open Access Journals (Sweden)

    Unai López de Heredia

    2015-04-01

    Full Text Available Aim of the study: Models that combine parentage analysis from molecular data with spatial information of seeds and seedlings provide a framework to describe and identify the factors involved in seed dispersal and recruitment of forest species. In the present study we used a spatially explicit method (the gene shadow model in order to assess primary and effective dispersal in Pinus canariensis. Area of study: Pinus canariensis is endemic to the Canary Islands (Spain. Sampling sites were a high density forest in southern slopes of Tenerife and a low density stand in South Gran Canaria. Materials and methods: We fitted models based on parentage analysis from seeds and seedlings collected in two sites with contrasting stand density, and then compared the resulting dispersal distributions. Main results: The results showed that: 1 P. canariensis has a remarkable dispersal ability compared to other pine species; 2 there is no discordance between primary and effective dispersals, suggesting limited secondary dispersal by animals and lack of Janzen-Connell effect; and 3 low stand densities enhance the extent of seed dispersal, which was higher in the low density stand. Research highlights: The efficient dispersal mechanism of P. canariensis by wind inferred by the gene shadow model is congruent with indirect measures of gene flow, and has utility in reconstructing past demographic events and in predicting future distribution ranges for the species.

  13. 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)

  14. Development of high uranium-density fuels for use in research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ugajin, Mitsuhiro; Akabori, Mitsuo; Itoh, Akinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-02-01

    The uranium silicide U{sub 3}Si{sub 2} possesses uranium density 11.3 gU/cm{sup 3} with a congruent melting point of 1665degC, and is now successfully in use as a research reactor fuel. Another uranium silicide U{sub 3}Si and U{sub 6}Me-type uranium alloys (Me=Fe,Mn,Ni) have been chosen as new fuel materials because of the higher uranium densities 14.9 and 17.0 gU/cm{sup 3}, respectively. Experiments were carried out to fabricate miniature aluminum-dispersion plate-type and aluminum-clad disk-type fuels by using the conventional picture-frame method and a hot-pressing technique, respectively. These included the above-mentioned new fuel materials as well as U{sub 3}Si{sub 2}. Totally 14 miniplates with uranium densities from 4.0 to 6.3 gU/cm{sup 3} of fuel meat were prepared together with 28 disk-type fuel containing structurally-modified U{sub 3}Si, and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Some results of postirradiation examinations are presented. (author)

  15. Development of high uranium-density fuels for use in research reactors

    International Nuclear Information System (INIS)

    Ugajin, Mitsuhiro; Akabori, Mitsuo; Itoh, Akinori

    1996-01-01

    The uranium silicide U 3 Si 2 possesses uranium density 11.3 gU/cm 3 with a congruent melting point of 1665degC, and is now successfully in use as a research reactor fuel. Another uranium silicide U 3 Si and U 6 Me-type uranium alloys (Me=Fe,Mn,Ni) have been chosen as new fuel materials because of the higher uranium densities 14.9 and 17.0 gU/cm 3 , respectively. Experiments were carried out to fabricate miniature aluminum-dispersion plate-type and aluminum-clad disk-type fuels by using the conventional picture-frame method and a hot-pressing technique, respectively. These included the above-mentioned new fuel materials as well as U 3 Si 2 . Totally 14 miniplates with uranium densities from 4.0 to 6.3 gU/cm 3 of fuel meat were prepared together with 28 disk-type fuel containing structurally-modified U 3 Si, and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Some results of postirradiation examinations are presented. (author)

  16. Quorum sensing and density-dependent dispersal in an aquatic model system.

    Directory of Open Access Journals (Sweden)

    Simon Fellous

    Full Text Available Many organisms use cues to decide whether to disperse or not, especially those related to the composition of their environment. Dispersal hence sometimes depends on population density, which can be important for the dynamics and evolution of sub-divided populations. But very little is known about the factors that organisms use to inform their dispersal decision. We investigated the cues underlying density-dependent dispersal in inter-connected microcosms of the freshwater protozoan Paramecium caudatum. In two experiments, we manipulated (i the number of cells per microcosm and (ii the origin of their culture medium (supernatant from high- or low-density populations. We found a negative relationship between population density and rates of dispersal, suggesting the use of physical cues. There was no significant effect of culture medium origin on dispersal and thus no support for chemical cues usage. These results suggest that the perception of density - and as a result, the decision to disperse - in this organism can be based on physical factors. This type of quorum sensing may be an adaptation optimizing small scale monitoring of the environment and swarm formation in open water.

  17. 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)

  18. 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

  19. 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.

  20. 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.

  1. 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)

  2. Core fueling to produce peaked density profiles in large tokamaks

    International Nuclear Information System (INIS)

    Mikkelsen, D.R.; McGuire, K.M.; Schmidt, G.L.; Zweben, S.J.

    1994-06-01

    Peaking the density profile increases the usable bootstrap current and the average fusion power density; this could reduce the current drive power and increase the net output of power producing tokamaks. The use of neutral beams and pellet injection to produce peaked density profiles is assessed. We show that with radially ''hollow'' diffusivity profiles (and no particle pinch) moderately peaked density profiles can be produced by particle source profiles which are peaked off-axis. The fueling penetration requirements can therefore be relaxed and this greatly improves the feasibility of generating peaked density profiles in large tokamaks. In particular, neutral beam fueling does not require MeV particle energy. Even with beam voltages of ∼200 keV, however, exceptionally good particle confinement, τ p much-gt τ E is required to achieve net electrical power generation. In system with no power production requirement (e.g., neutron sources) neutral beam fueling should be capable of producing peaked density profiles in devices as large as ITER. Fueling systems with low energy cost per particle (such as cryogenic pellet injection) must be used in power producing tokamaks when τ p ∼ τ E . Simulations with pellet injection speeds of 7 km/sec show the peaking factor, n eo /left-angle n e right-angle, approaching 2

  3. 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)

  4. Shock Ignition of Thermonuclear Fuel with High Areal Density

    International Nuclear Information System (INIS)

    Betti, R.; Zhou, C. D.; Anderson, K. S.; Theobald, W.; Solodov, A. A.; Perkins, L. J.

    2007-01-01

    A novel method by C. Zhou and R. Betti [Bull. Am. Phys. Soc. 50, 140 (2005)] to assemble and ignite thermonuclear fuel is presented. Massive cryogenic shells are first imploded by direct laser light with a low implosion velocity and on a low adiabat leading to fuel assemblies with large areal densities. The assembled fuel is ignited from a central hot spot heated by the collision of a spherically convergent ignitor shock and the return shock. The resulting fuel assembly features a hot-spot pressure greater than the surrounding dense fuel pressure. Such a nonisobaric assembly requires a lower energy threshold for ignition than the conventional isobaric one. The ignitor shock can be launched by a spike in the laser power or by particle beams. The thermonuclear gain can be significantly larger than in conventional isobaric ignition for equal driver energy

  5. Shock ignition of thermonuclear fuel with high areal density.

    Science.gov (United States)

    Betti, R; Zhou, C D; Anderson, K S; Perkins, L J; Theobald, W; Solodov, A A

    2007-04-13

    A novel method by C. Zhou and R. Betti [Bull. Am. Phys. Soc. 50, 140 (2005)] to assemble and ignite thermonuclear fuel is presented. Massive cryogenic shells are first imploded by direct laser light with a low implosion velocity and on a low adiabat leading to fuel assemblies with large areal densities. The assembled fuel is ignited from a central hot spot heated by the collision of a spherically convergent ignitor shock and the return shock. The resulting fuel assembly features a hot-spot pressure greater than the surrounding dense fuel pressure. Such a nonisobaric assembly requires a lower energy threshold for ignition than the conventional isobaric one. The ignitor shock can be launched by a spike in the laser power or by particle beams. The thermonuclear gain can be significantly larger than in conventional isobaric ignition for equal driver energy.

  6. 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.

  7. 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.

  8. 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

  9. 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.

  10. Directed seed dispersal towards areas with low conspecific tree density by a scatter-hoarding rodent

    NARCIS (Netherlands)

    Hirsch, Ben T.; Kays, Roland; Pereira, Veronica E.; Jansen, Patrick A.

    2012-01-01

    Scatter-hoarding animals spread out cached seeds to reduce density-dependent theft of their food reserves. This behaviour could lead to directed dispersal into areas with lower densities of conspecific trees, where seed and seedling survival are higher, and could profoundly affect the spatial

  11. 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)

  12. 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)

  13. Inclusion of Dispersion Effects in Density Functional Theory

    DEFF Research Database (Denmark)

    Møgelhøj, Andreas

    on fitting to high-level ab initio and experimental results. The fitting scheme, based on Baysian theory, focuses on the three aspects: a) model space, b) datasets, and c) model selection. The model space consists of a flexible expansion of the exchange enhancement factor in the generalized gradient......In this thesis, applications and development will be presented within the field of van der Waals interactions in density functional theory. The thesis is based on the three projects: i) van der Waals interactions effect on the structure of liquid water at ambient conditions, ii) development......-range van der Waals interactions is essential to describe the adsorption/desorption process and commonly used generalized gradient approximation functionals are seen to be incapable of this....

  14. 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

  15. 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.

  16. 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

  17. 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.

  18. 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

  19. 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)

  20. 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

  1. 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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  3. 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

  4. 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)

  5. 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)

  6. 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)

  7. 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)

  8. 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.

  9. 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)

  10. Surface-plasmon dispersion relation for the inhomogeneous charge-density medium

    International Nuclear Information System (INIS)

    Harsh, O.K.; Agarwal, B.K.

    1989-01-01

    The surface-plasmon dispersion relation is derived for the plane-bounded electron gas when there is an inhomogeneous charge-density distribution in the plasma. The hydrodynamical model is used. Both cphi and dcphi/dx are taken to be continuous at the surface of the slab, where cphi is the scalar potential. The dispersion relation is compared with the theoretical works of Stern and Ferrell and of Harsh and Agarwal. It is also compared with the observations of Kunz. A dispersion relation for the volume-plasmon oscillations is derived which resembles the well-known relation of Bohm and Pines

  11. 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.

  12. High power density yeast catalyzed microbial fuel cells

    Science.gov (United States)

    Ganguli, Rahul

    Microbial fuel cells leverage whole cell biocatalysis to convert the energy stored in energy-rich renewable biomolecules such as sugar, directly to electrical energy at high efficiencies. Advantages of the process include ambient temperature operation, operation in natural streams such as wastewater without the need to clean electrodes, minimal balance-of-plant requirements compared to conventional fuel cells, and environmentally friendly operation. These make the technology very attractive as portable power sources and waste-to-energy converters. The principal problem facing the technology is the low power densities compared to other conventional portable power sources such as batteries and traditional fuel cells. In this work we examined the yeast catalyzed microbial fuel cell and developed methods to increase the power density from such fuel cells. A combination of cyclic voltammetry and optical absorption measurements were used to establish significant adsorption of electron mediators by the microbes. Mediator adsorption was demonstrated to be an important limitation in achieving high power densities in yeast-catalyzed microbial fuel cells. Specifically, the power densities are low for the length of time mediator adsorption continues to occur. Once the mediator adsorption stops, the power densities increase. Rotating disk chronoamperometry was used to extract reaction rate information, and a simple kinetic expression was developed for the current observed in the anodic half-cell. Since the rate expression showed that the current was directly related to microbe concentration close to the electrode, methods to increase cell mass attached to the anode was investigated. Electrically biased electrodes were demonstrated to develop biofilm-like layers of the Baker's yeast with a high concentration of cells directly connected to the electrode. The increased cell mass did increase the power density 2 times compared to a non biofilm fuel cell, but the power density

  13. 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)

  14. 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

  15. 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

  16. 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

  17. Uranium density reduction on fuel element side plates assessment

    International Nuclear Information System (INIS)

    Rios, Ilka A.; Andrade, Delvonei A.; Domingos, Douglas B.; Umbehaun, Pedro E.

    2011-01-01

    During operation of IEA-R1 research reactor, located at Instituto de Pesquisas Energeticas e Nucleares, IPEN - CNEN/SP, an abnormal oxidation on some fuel elements was noted. It was also verified, among the possible causes of the problem, that the most likely one was insufficient cooling of the elements in the core. One of the propositions to solve or minimize the problem is to reduce uranium density on fuel elements side plates. In this paper, the influence of this change on neutronic and thermal hydraulic parameters for IEA-R1 reactor is verified by simulations with the codes HAMMER and CITATION. Results are presented and discussed. (author)

  18. Uranium density reduction on fuel element side plates assessment

    Energy Technology Data Exchange (ETDEWEB)

    Rios, Ilka A. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Andrade, Delvonei A.; Domingos, Douglas B.; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    During operation of IEA-R1 research reactor, located at Instituto de Pesquisas Energeticas e Nucleares, IPEN - CNEN/SP, an abnormal oxidation on some fuel elements was noted. It was also verified, among the possible causes of the problem, that the most likely one was insufficient cooling of the elements in the core. One of the propositions to solve or minimize the problem is to reduce uranium density on fuel elements side plates. In this paper, the influence of this change on neutronic and thermal hydraulic parameters for IEA-R1 reactor is verified by simulations with the codes HAMMER and CITATION. Results are presented and discussed. (author)

  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. 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

  2. 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

  3. 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)

  4. 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

  5. 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)

  6. Dispersion relation and Landau damping of waves in high-energy density plasmas

    International Nuclear Information System (INIS)

    Zhu Jun; Ji Peiyong

    2012-01-01

    We present a theoretical investigation on the propagation of electromagnetic waves and electron plasma waves in high energy density plasmas using the covariant Wigner function approach. Based on the covariant Wigner function and Dirac equation, a relativistic quantum kinetic model is established to describe the physical processes in high-energy density plasmas. With the zero-temperature Fermi–Dirac distribution, the dispersion relation and Landau damping of waves containing the relativistic quantum corrected terms are derived. The relativistic quantum corrections to the dispersion relation and Landau damping are analyzed by comparing our results with those obtained in classical and non-relativistic quantum plasmas. We provide a detailed discussion on the Landau damping obtained in classical plasmas, non-relativistic Fermi plasmas and relativistic Fermi plasmas. The contributions of the Bohm potential, the Fermi statistics pressure and relativistic effects to the dispersion relation and Landau damping of waves are quantitatively calculated with real plasma parameters. (paper)

  7. Dispersion correction derived from first principles for density functional theory and Hartree-Fock theory.

    Science.gov (United States)

    Guidez, Emilie B; Gordon, Mark S

    2015-03-12

    The modeling of dispersion interactions in density functional theory (DFT) is commonly performed using an energy correction that involves empirically fitted parameters for all atom pairs of the system investigated. In this study, the first-principles-derived dispersion energy from the effective fragment potential (EFP) method is implemented for the density functional theory (DFT-D(EFP)) and Hartree-Fock (HF-D(EFP)) energies. Overall, DFT-D(EFP) performs similarly to the semiempirical DFT-D corrections for the test cases investigated in this work. HF-D(EFP) tends to underestimate binding energies and overestimate intermolecular equilibrium distances, relative to coupled cluster theory, most likely due to incomplete accounting for electron correlation. Overall, this first-principles dispersion correction yields results that are in good agreement with coupled-cluster calculations at a low computational cost.

  8. Assessment of Microbial Fuel Cell Configurations and Power Densities

    KAUST Repository

    Logan, Bruce E.

    2015-07-30

    Different microbial electrochemical technologies are being developed for a many diverse applications, including wastewater treatment, biofuel production, water desalination, remote power sources, and as biosensors. Current and energy densities will always be limited relative to batteries and chemical fuel cells, but these technologies have other advantages based on the self-sustaining nature of the microorganisms that can donate or accept electrons from an electrode, the range of fuels that can be used, and versatility in the chemicals that can be produced. The high cost of membranes will likely limit applications of microbial electrochemical technologies that might require a membrane. For microbial fuel cells, which do not need a membrane, questions remain on whether larger-scale systems can produce power densities similar to those obtained in laboratory-scale systems. It is shown here that configuration and fuel (pure chemicals in laboratory media versus actual wastewaters) remain the key factors in power production, rather than the scale of the application. Systems must be scaled up through careful consideration of electrode spacing and packing per unit volume of reactor.

  9. Assessment of Microbial Fuel Cell Configurations and Power Densities

    KAUST Repository

    Logan, Bruce E.; Wallack, Maxwell J; Kim, Kyoung-Yeol; He, Weihua; Feng, Yujie; Saikaly, Pascal

    2015-01-01

    Different microbial electrochemical technologies are being developed for a many diverse applications, including wastewater treatment, biofuel production, water desalination, remote power sources, and as biosensors. Current and energy densities will always be limited relative to batteries and chemical fuel cells, but these technologies have other advantages based on the self-sustaining nature of the microorganisms that can donate or accept electrons from an electrode, the range of fuels that can be used, and versatility in the chemicals that can be produced. The high cost of membranes will likely limit applications of microbial electrochemical technologies that might require a membrane. For microbial fuel cells, which do not need a membrane, questions remain on whether larger-scale systems can produce power densities similar to those obtained in laboratory-scale systems. It is shown here that configuration and fuel (pure chemicals in laboratory media versus actual wastewaters) remain the key factors in power production, rather than the scale of the application. Systems must be scaled up through careful consideration of electrode spacing and packing per unit volume of reactor.

  10. 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.

  11. A new formulation of the probability density function in random walk models for atmospheric dispersion

    DEFF Research Database (Denmark)

    Falk, Anne Katrine Vinther; Gryning, Sven-Erik

    1997-01-01

    In this model for atmospheric dispersion particles are simulated by the Langevin Equation, which is a stochastic differential equation. It uses the probability density function (PDF) of the vertical velocity fluctuations as input. The PDF is constructed as an expansion after Hermite polynomials...

  12. 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)

  13. 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)

  14. 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)

  15. Effects of UO2 fuel microstructure and density on fuel in-reactor performance

    International Nuclear Information System (INIS)

    Hansson, L.

    1988-02-01

    The volume changes of UO 2 fuel pellets, produced by neutron irradiation, can be characterized by two processes: fission spike induced densification through pore skrinkage and later fission produced induced swelling of UO 2 matrix. In-pile densification is controlled by the initial density and microstructure of the fuel, particularly by the pore size distribution. The extent of swelling depends mainly on the amount of fission products produced, but the fission gas release as well as the swelling may be reduced by increasing the grain size of UO 2 . Fabrication of fuel pellets having certain in-reactor properties requires detailed knowledge of the effects of individual fabrication parameters. The irradiation experience of fuels fabricated by using different conversion and pelletizing methods is extensive. Based on this experience, some general characteristics of stable/well-performing fuel microstructures have been summarized

  16. 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

  17. 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.

  18. 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

  19. 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)

  20. 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%.

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

    Science.gov (United States)

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

    2012-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-06-15

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

  3. Possibilities of increasing coal charge density by adding fuel oil

    Directory of Open Access Journals (Sweden)

    M. Fröhlichová

    2010-01-01

    Full Text Available The requirement of all coke-making facilities is to achieve the highest possible production of high quality coke from a chamber. It can be achieved by filling the effective capacity of the chamber with the highest possible amount of coal. One of the possibilities of meeting this requirement is to increase the charge density in the coke chamber. In case of a coke battery operating on bulk coal there are many methods to increase the charge density including the use of wetting agents in the charge. This article presents the results of the laboratory experiments aiming at the increase of the charge density using fuel oil as a wetting agent. The experiments were carried out by means of the Pitin’s device using 3 coal charges with various granularity composition and moisture content of 7, 8, 9 and 10 %.

  4. 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

  5. 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

  6. Electron density measurement of non-equilibrium atmospheric pressure plasma using dispersion interferometer

    Science.gov (United States)

    Yoshimura, Shinji; Kasahara, Hiroshi; Akiyama, Tsuyoshi

    2017-10-01

    Medical applications of non-equilibrium atmospheric plasmas have recently been attracting a great deal of attention, where many types of plasma sources have been developed to meet the purposes. For example, plasma-activated medium (PAM), which is now being studied for cancer treatment, has been produced by irradiating non-equilibrium atmospheric pressure plasma with ultrahigh electron density to a culture medium. Meanwhile, in order to measure electron density in magnetic confinement plasmas, a CO2 laser dispersion interferometer has been developed and installed on the Large Helical Device (LHD) at the National Institute for Fusion Science, Japan. The dispersion interferometer has advantages that the measurement is insensitive to mechanical vibrations and changes in neutral gas density. Taking advantage of these properties, we applied the dispersion interferometer to electron density diagnostics of atmospheric pressure plasmas produced by the NU-Global HUMAP-WSAP-50 device, which is used for producing PAM. This study was supported by the Grant of Joint Research by the National Institutes of Natural Sciences (NINS).

  7. 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)

  8. 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.)

  9. 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.

  10. The Effects of Dispersal and Predator Density on Prey Survival in an Insect-Red Clover Metacommunity.

    Science.gov (United States)

    Stasek, David J; Radl, James N; Crist, Thomas O

    2018-01-01

    Trophic interactions are often studied within habitat patches, but among-patch dispersal of individuals may influence local patch dynamics. Metacommunity concepts incorporate the effects of dispersal on local and community dynamics. There are few experimental tests of metacommunity theory using insects compared to those conducted in microbial microcosms. Using connected experimental mesocosms, we varied the density of the leafhopper Agallia constricta Van Duzee (Homoptera: Cicadellidae) and a generalist insect predator, the damsel bug (Nabis spp., Heteroptera: Nabidae), to determine the effects of conspecific and predator density and varying the time available to dispersal among mesocosms on predation rates, dispersal rates, and leafhopper survival. Conspecific and damsel bug density did not affect dispersal rates in leafhoppers, but this may be due to leafhoppers' aversion to leaving the host plants or the connecting tubes between mesocosms hindering leafhopper movement. Leafhopper dispersal was higher in high-dispersal treatments. Survival rates of A. constricta were also lowest in treatments where dispersal was not limited. This is one of the first experimental studies to vary predator density and the time available to dispersal. Our results indicate that dispersal is the key to understanding short-term processes such as prey survival in predator-prey metacommunities. Further work is needed to determine how dispersal rates influence persistence of communities in multigenerational studies. © The Author(s) 2017. Published by Oxford University Press on behalf of Entomological Society of America.

  11. 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.

  12. 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.

  13. Long-Range Corrected Hybrid Density Functionals with Damped Atom-Atom Dispersion Corrections

    Energy Technology Data Exchange (ETDEWEB)

    Chai, Jeng-Da; Head-Gordon, Martin

    2008-06-14

    We report re-optimization of a recently proposed long-range corrected (LC) hybrid density functionals [J.-D. Chai and M. Head-Gordon, J. Chem. Phys. 128, 084106 (2008)] to include empirical atom-atom dispersion corrections. The resulting functional, {omega}B97X-D yields satisfactory accuracy for thermochemistry, kinetics, and non-covalent interactions. Tests show that for non-covalent systems, {omega}B97X-D shows slight improvement over other empirical dispersion-corrected density functionals, while for covalent systems and kinetics, it performs noticeably better. Relative to our previous functionals, such as {omega}B97X, the new functional is significantly superior for non-bonded interactions, and very similar in performance for bonded interactions.

  14. Measurement of average density and relative volumes in a dispersed two-phase fluid

    Science.gov (United States)

    Sreepada, Sastry R.; Rippel, Robert R.

    1992-01-01

    An apparatus and a method are disclosed for measuring the average density and relative volumes in an essentially transparent, dispersed two-phase fluid. A laser beam with a diameter no greater than 1% of the diameter of the bubbles, droplets, or particles of the dispersed phase is directed onto a diffraction grating. A single-order component of the diffracted beam is directed through the two-phase fluid and its refraction is measured. Preferably, the refracted beam exiting the fluid is incident upon a optical filter with linearly varing optical density and the intensity of the filtered beam is measured. The invention can be combined with other laser-based measurement systems, e.g., laser doppler anemometry.

  15. 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

  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. 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)

  18. 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.

  19. 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.

  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. 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

  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. 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)

  4. Neurite orientation dispersion and density imaging in the substantia nigra in idiopathic Parkinson disease.

    Science.gov (United States)

    Kamagata, Koji; Hatano, Taku; Okuzumi, Ayami; Motoi, Yumiko; Abe, Osamu; Shimoji, Keigo; Kamiya, Kouhei; Suzuki, Michimasa; Hori, Masaaki; Kumamaru, Kanako K; Hattori, Nobutaka; Aoki, Shigeki

    2016-08-01

    We used neurite orientation dispersion and density imaging (NODDI) to quantify changes in the substantia nigra pars compacta (SNpc) and striatum in Parkinson disease (PD). Diffusion-weighted magnetic resonance images were acquired from 58 PD patients and 36 age- and sex-matched controls. The intracellular volume fraction (Vic), orientation dispersion index (OD), and isotropic volume fraction (Viso) of the basal ganglia were compared between groups. Multivariate logistic regression analysis determined which diffusion parameters were independent predictors of PD. Receiver operating characteristic (ROC) analysis compared the diagnostic accuracies of the evaluated indices. Pearson coefficient analysis correlated each diffusional parameter with disease severity. Vic in the contralateral SNpc and putamen were significantly lower in PD patients than in healthy controls (P disease severity. Multivariate logistic analysis revealed that Vic (P = 0.0000046) and mean diffusivity (P = 0.019) in the contralateral SNpc were the independent predictors of PD. In the ROC analysis, Vic in the contralateral SNpc showed the best diagnostic performance (mean cutoff, 0.62; sensitivity, 0.88; specificity, 0.83). NODDI is likely to be useful for diagnosing PD and assessing its progression. • Neurite orientation dispersion and density imaging (NODDI) is a new diffusion MRI technique • NODDI estimates neurite microstructure more specifically than diffusion tensor imaging • By using NODDI, nigrostriatal alterations in PD can be evaluated in vivo • NOODI is useful for diagnosing PD and assessing its disease progression.

  5. 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)

  6. 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

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

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Shirasu, Noriko; Ohmichi, Toshihiko

    1999-05-01

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

  8. How important is self-consistency for the dDsC density dependent dispersion correction?

    Energy Technology Data Exchange (ETDEWEB)

    Brémond, Éric; Corminboeuf, Clémence, E-mail: clemence.corminboeuf@epfl.ch [Laboratory for Computational Molecular Design, Institut des Sciences et Ingénierie Chimiques, Ecole Polytechnique Fédérale de Lausanne, CH-1015 Lausanne (Switzerland); Golubev, Nikolay [Laboratory for Computational Molecular Design, Institut des Sciences et Ingénierie Chimiques, Ecole Polytechnique Fédérale de Lausanne, CH-1015 Lausanne (Switzerland); Department of Chemistry, M. V. Lomonosov Moscow State University, Moscow 119991 (Russian Federation); Steinmann, Stephan N., E-mail: sns25@duke.edu [Laboratory for Computational Molecular Design, Institut des Sciences et Ingénierie Chimiques, Ecole Polytechnique Fédérale de Lausanne, CH-1015 Lausanne (Switzerland); Department of Chemistry, Duke University, Durham, North Carolina 27708 (United States)

    2014-05-14

    The treatment of dispersion interactions is ubiquitous but computationally demanding for seamless ab initio approaches. A highly popular and simple remedy consists in correcting for the missing interactions a posteriori by adding an attractive energy term summed over all atom pairs to standard density functional approximations. These corrections were originally based on atom pairwise parameters and, hence, had a strong touch of empiricism. To overcome such limitations, we recently proposed a robust system-dependent dispersion correction, dDsC, that is computed from the electron density and that provides a balanced description of both weak inter- and intramolecular interactions. From the theoretical point of view and for the sake of increasing reliability, we here verify if the self-consistent implementation of dDsC impacts ground-state properties such as interaction energies, electron density, dipole moments, geometries, and harmonic frequencies. In addition, we investigate the suitability of the a posteriori scheme for molecular dynamics simulations, for which the analysis of the energy conservation constitutes a challenging tests. Our study demonstrates that the post-SCF approach in an excellent approximation.

  9. 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.

  10. 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)

  11. 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)

  12. Moisture management, energy density and fuel quality in forest fuel supply chains

    Energy Technology Data Exchange (ETDEWEB)

    Tahvanainen, T. [Joensuu Science Park Ltd., Joensuu (Finland); Sikanen, L. [Joensuu Univ. (Finland); Roser, D. [Finnish Forest Research Inst., Joensuu (Finland)

    2009-07-01

    This presentation provided tools for reducing the moisture content (MC) in wood chips, as moisture is one of the main quality factors for woody biomass, along with energy density and cleanness. The amount of water in solid wood fuels has a considerable effect on transportation efficiency, combustion efficiency and emissions. Under favourable storage conditions, MC can be decreased from typical fresh cut 50-55 per cent to 20-30 per cent in relatively short periods of storing by natural or artificial drying. Minor modifications can boost natural drying in fuel wood supply chains. This natural drying effect can have significant effects on the total energy efficiency and emissions of supply chains. The effect of improved packing density on transportation phase was discussed along with the need to control chip purity and size distribution. A procedure developed at the University of Joensuu and in the Finnish Forest Research Institute was used to estimate transportation costs and emissions according to transportation fleet and MC of the transported fuel. tabs., figs.

  13. 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)

  14. 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)

  15. 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

  16. Probability density function of a puff dispersing from the wall of a turbulent channel

    Science.gov (United States)

    Nguyen, Quoc; Papavassiliou, Dimitrios

    2015-11-01

    Study of dispersion of passive contaminants in turbulence has proved to be helpful in understanding fundamental heat and mass transfer phenomena. Many simulation and experimental works have been carried out to locate and track motions of scalar markers in a flow. One method is to combine Direct Numerical Simulation (DNS) and Lagrangian Scalar Tracking (LST) to record locations of markers. While this has proved to be useful, high computational cost remains a concern. In this study, we develop a model that could reproduce results obtained by DNS and LST for turbulent flow. Puffs of markers with different Schmidt numbers were released into a flow field at a frictional Reynolds number of 150. The point of release was at the channel wall, so that both diffusion and convection contribute to the puff dispersion pattern, defining different stages of dispersion. Based on outputs from DNS and LST, we seek the most suitable and feasible probability density function (PDF) that represents distribution of markers in the flow field. The PDF would play a significant role in predicting heat and mass transfer in wall turbulence, and would prove to be helpful where DNS and LST are not always available.

  17. Biharmonic split ring resonator metamaterial: Artificially dispersive effective density in thin periodically perforated plates

    KAUST Repository

    Farhat, Mohamed

    2014-08-01

    We present in this paper a theoretical and numerical analysis of bending waves localized on the boundary of a platonic crystal whose building blocks are Split Ring Resonators (SRR). We first derive the homogenized parameters of the structured plate using a three-scale asymptotic expansion in the linearized biharmonic equation. In the limit when the wavelength of the bending wave is much larger than the typical heterogeneity size of the platonic crystal, we show that it behaves as an artificial plate with an anisotropic effective Young modulus and a dispersive effective mass density. We then analyze dispersion diagrams associated with bending waves propagating within an infinite array of SRR, for which eigen-solutions are sought in the form of Floquet-Bloch waves. We finally demonstrate that this structure displays the hallmarks of All-Angle Negative Refraction (AANR) and it leads to superlensing and ultrarefraction effects, interpreted thanks to our homogenization model as a consequence of negative and vanishing effective density, respectively. © EPLA, 2014.

  18. Validation of experimental molecular crystal structures with dispersion-corrected density functional theory calculations

    International Nuclear Information System (INIS)

    Streek, Jacco van de; Neumann, Marcus A.

    2010-01-01

    The accuracy of a dispersion-corrected density functional theory method is validated against 241 experimental organic crystal structures from Acta Cryst. Section E. This paper describes the validation of a dispersion-corrected density functional theory (d-DFT) method for the purpose of assessing the correctness of experimental organic crystal structures and enhancing the information content of purely experimental data. 241 experimental organic crystal structures from the August 2008 issue of Acta Cryst. Section E were energy-minimized in full, including unit-cell parameters. The differences between the experimental and the minimized crystal structures were subjected to statistical analysis. The r.m.s. Cartesian displacement excluding H atoms upon energy minimization with flexible unit-cell parameters is selected as a pertinent indicator of the correctness of a crystal structure. All 241 experimental crystal structures are reproduced very well: the average r.m.s. Cartesian displacement for the 241 crystal structures, including 16 disordered structures, is only 0.095 Å (0.084 Å for the 225 ordered structures). R.m.s. Cartesian displacements above 0.25 Å either indicate incorrect experimental crystal structures or reveal interesting structural features such as exceptionally large temperature effects, incorrectly modelled disorder or symmetry breaking H atoms. After validation, the method is applied to nine examples that are known to be ambiguous or subtly incorrect

  19. Biharmonic split ring resonator metamaterial: Artificially dispersive effective density in thin periodically perforated plates

    KAUST Repository

    Farhat, Mohamed; Enoch, Stefan; Guenneau, Sé bastien

    2014-01-01

    We present in this paper a theoretical and numerical analysis of bending waves localized on the boundary of a platonic crystal whose building blocks are Split Ring Resonators (SRR). We first derive the homogenized parameters of the structured plate using a three-scale asymptotic expansion in the linearized biharmonic equation. In the limit when the wavelength of the bending wave is much larger than the typical heterogeneity size of the platonic crystal, we show that it behaves as an artificial plate with an anisotropic effective Young modulus and a dispersive effective mass density. We then analyze dispersion diagrams associated with bending waves propagating within an infinite array of SRR, for which eigen-solutions are sought in the form of Floquet-Bloch waves. We finally demonstrate that this structure displays the hallmarks of All-Angle Negative Refraction (AANR) and it leads to superlensing and ultrarefraction effects, interpreted thanks to our homogenization model as a consequence of negative and vanishing effective density, respectively. © EPLA, 2014.

  20. Fast Radio Bursts’ Recipes for the Distributions of Dispersion Measures, Flux Densities, and Fluences

    Science.gov (United States)

    Niino, Yuu

    2018-05-01

    We investigate how the statistical properties of dispersion measure (DM) and apparent flux density/fluence of (nonrepeating) fast radio bursts (FRBs) are determined by unknown cosmic rate density history [ρ FRB(z)] and luminosity function (LF) of the transient events. We predict the distributions of DMs, flux densities, and fluences of FRBs taking account of the variation of the receiver efficiency within its beam, using analytical models of ρ FRB(z) and LF. Comparing the predictions with the observations, we show that the cumulative distribution of apparent fluences suggests that FRBs originate at cosmological distances and ρ FRB increases with redshift resembling the cosmic star formation history (CSFH). We also show that an LF model with a bright-end cutoff at log10 L ν (erg s‑1 Hz‑1) ∼ 34 are favored to reproduce the observed DM distribution if ρ FRB(z) ∝ CSFH, although the statistical significance of the constraints obtained with the current size of the observed sample is not high. Finally, we find that the correlation between DM and flux density of FRBs is potentially a powerful tool to distinguish whether FRBs are at cosmological distances or in the local universe more robustly with future observations.

  1. 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.

  2. 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.

  3. Halogen Bonding from Dispersion-Corrected Density-Functional Theory: The Role of Delocalization Error.

    Science.gov (United States)

    Otero-de-la-Roza, A; Johnson, Erin R; DiLabio, Gino A

    2014-12-09

    Halogen bonds are formed when a Lewis base interacts with a halogen atom in a different molecule, which acts as an electron acceptor. Due to its charge transfer component, halogen bonding is difficult to model using many common density-functional approximations because they spuriously overstabilize halogen-bonded dimers. It has been suggested that dispersion-corrected density functionals are inadequate to describe halogen bonding. In this work, we show that the exchange-hole dipole moment (XDM) dispersion correction coupled with functionals that minimize delocalization error (for instance, BH&HLYP, but also other half-and-half functionals) accurately model halogen-bonded interactions, with average errors similar to other noncovalent dimers with less charge-transfer effects. The performance of XDM is evaluated for three previously proposed benchmarks (XB18 and XB51 by Kozuch and Martin, and the set proposed by Bauzá et al.) spanning a range of binding energies up to ∼50 kcal/mol. The good performance of BH&HLYP-XDM is comparable to M06-2X, and extends to the "extreme" cases in the Bauzá set. This set contains anionic electron donors where charge transfer occurs even at infinite separation, as well as other charge transfer dimers belonging to the pnictogen and chalcogen bonding classes. We also show that functional delocalization error results in an overly delocalized electron density and exact-exchange hole. We propose intermolecular Bader delocalization indices as an indicator of both the donor-acceptor character of an intermolecular interaction and the delocalization error coming from the underlying functional.

  4. 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)

  5. 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)

  6. 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.

  7. 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.

  8. 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.)

  9. Improvement of the environmental and operational characteristics of vehicles through decreasing the motor fuel density.

    Science.gov (United States)

    Magaril, Elena

    2016-04-01

    The environmental and operational characteristics of motor transport, one of the main consumers of motor fuel and source of toxic emissions, soot, and greenhouse gases, are determined to a large extent by the fuel quality which is characterized by many parameters. Fuel density is one of these parameters and it can serve as an indicator of fuel quality. It has been theoretically substantiated that an increased density of motor fuel has a negative impact both on the environmental and operational characteristics of motor transport. The use of fuels with a high density leads to an increase in carbonization within the engine, adversely affecting the vehicle performance and increasing environmental pollution. A program of technological measures targeted at reducing the density of the fuel used was offered. It includes a solution to the problem posed by changes in the refining capacities ratio and the temperature range of gasoline and diesel fuel boiling, by introducing fuel additives and adding butanes to the gasoline. An environmental tax has been developed which allows oil refineries to have a direct impact on the production of fuels with improved environmental performance, taking into account the need to minimize the density of the fuel within a given category of quality.

  10. Measurements relating fire radiative energy density and surface fuel consumption - RxCADRE 2011 and 2012

    Science.gov (United States)

    Andrew T. Hudak; Matthew B. Dickinson; Benjamin C. Bright; Robert L. Kremens; E. Louise Loudermilk; Joseph J. O' Brien; Benjamin S. Hornsby; Roger D. Ottmar

    2016-01-01

    Small-scale experiments have demonstrated that fire radiative energy is linearly related to fuel combusted but such a relationship has not been shown at the landscape level of prescribed fires. This paper presents field and remotely sensed measures of pre-fire fuel loads, consumption, fire radiative energy density (FRED) and fire radiative power flux density (FRFD),...

  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. Development of Advanced High Uranium Density Fuels for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

    2016-02-15

    This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

  13. 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)

  14. Dispersion corrected hartree-fock and density functional theory for organic crystal structure prediction.

    Science.gov (United States)

    Brandenburg, Jan Gerit; Grimme, Stefan

    2014-01-01

    We present and evaluate dispersion corrected Hartree-Fock (HF) and Density Functional Theory (DFT) based quantum chemical methods for organic crystal structure prediction. The necessity of correcting for missing long-range electron correlation, also known as van der Waals (vdW) interaction, is pointed out and some methodological issues such as inclusion of three-body dispersion terms are discussed. One of the most efficient and widely used methods is the semi-classical dispersion correction D3. Its applicability for the calculation of sublimation energies is investigated for the benchmark set X23 consisting of 23 small organic crystals. For PBE-D3 the mean absolute deviation (MAD) is below the estimated experimental uncertainty of 1.3 kcal/mol. For two larger π-systems, the equilibrium crystal geometry is investigated and very good agreement with experimental data is found. Since these calculations are carried out with huge plane-wave basis sets they are rather time consuming and routinely applicable only to systems with less than about 200 atoms in the unit cell. Aiming at crystal structure prediction, which involves screening of many structures, a pre-sorting with faster methods is mandatory. Small, atom-centered basis sets can speed up the computation significantly but they suffer greatly from basis set errors. We present the recently developed geometrical counterpoise correction gCP. It is a fast semi-empirical method which corrects for most of the inter- and intramolecular basis set superposition error. For HF calculations with nearly minimal basis sets, we additionally correct for short-range basis incompleteness. We combine all three terms in the HF-3c denoted scheme which performs very well for the X23 sublimation energies with an MAD of only 1.5 kcal/mol, which is close to the huge basis set DFT-D3 result.

  15. Neurite orientation dispersion and density imaging in the substantia nigra in idiopathic Parkinson disease

    International Nuclear Information System (INIS)

    Kamagata, Koji; Suzuki, Michimasa; Hori, Masaaki; Kumamaru, Kanako K.; Aoki, Shigeki; Hatano, Taku; Okuzumi, Ayami; Motoi, Yumiko; Hattori, Nobutaka; Abe, Osamu; Shimoji, Keigo; Kamiya, Kouhei

    2016-01-01

    We used neurite orientation dispersion and density imaging (NODDI) to quantify changes in the substantia nigra pars compacta (SNpc) and striatum in Parkinson disease (PD). Diffusion-weighted magnetic resonance images were acquired from 58 PD patients and 36 age- and sex-matched controls. The intracellular volume fraction (Vic), orientation dispersion index (OD), and isotropic volume fraction (Viso) of the basal ganglia were compared between groups. Multivariate logistic regression analysis determined which diffusion parameters were independent predictors of PD. Receiver operating characteristic (ROC) analysis compared the diagnostic accuracies of the evaluated indices. Pearson coefficient analysis correlated each diffusional parameter with disease severity. Vic in the contralateral SNpc and putamen were significantly lower in PD patients than in healthy controls (P < 0.00058). Vic and OD in the SNpc and putamen showed significant negative correlations (P < 0.05) with disease severity. Multivariate logistic analysis revealed that Vic (P = 0.0000046) and mean diffusivity (P = 0.019) in the contralateral SNpc were the independent predictors of PD. In the ROC analysis, Vic in the contralateral SNpc showed the best diagnostic performance (mean cutoff, 0.62; sensitivity, 0.88; specificity, 0.83). NODDI is likely to be useful for diagnosing PD and assessing its progression. (orig.)

  16. Dispersion relations of density fluctuations observed by heavy ion beam probe in the TEXT tokamak

    International Nuclear Information System (INIS)

    Ross, D.W.

    1990-09-01

    Wave numbers as functions of frequency for density fluctuations in the core of the TEXT tokamak are measured in Heavy Ion Beam Probe experiments by analyzing the relative phases of signals originating from nearby points in the plasma. The adjacent points are typically 2 cm apart, with their relative orientation (δr, δθ) depending on position (r,θ). for angular frequencies ω ≤ 10 6 /s the signals are quite coherent, leading to reasonably well-defined ''dispersion relations.'' These do not correspond to known modes of the drift wave type, i.e., ballooning or slab-like electron drift waves or ion temperature gradient modes. The effect of finite sample volume size does not significantly alter this conclusion. 25 refs., 6 figs., 3 tabs

  17. 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)

  18. 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.

  19. 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)

  20. 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.

  1. Optical sensor system for time-resolved quantification of methane densities in CH4-fueled spark ignition engines.

    Science.gov (United States)

    Golibrzuch, Kai; Digulla, Finn-Erik; Bauke, Stephan; Wackerbarth, Hainer; Thiele, Olaf; Berg, Thomas

    2017-08-01

    We present the development and the first application of an optical sensor system that allows single-cycle determination of methane (CH 4 ) concentration inside internal combustion (IC) engines. We use non-dispersive infrared absorption spectroscopy to detect the CH 4 density with a time resolution up to 33 μs at acquisition rates of 30 kHz. The measurement scheme takes advantage of the strong temperature dependence of the absorption band applying two detection channels for CH 4 that detect different spectral regions of the ν 3 anti-symmetric C-H-stretch absorption. The strategy allows the simultaneous determination of fuel concentration as well as gas temperature. We show the proof-of-concept by validation of the measurement strategy in static pressure cell experiments as well as its application to a methane-fueled IC engine using a modified spark plug probe. Our results clearly demonstrate that it is crucial to determine the CH 4 temperature in the probe volume. Due to thermal influences of the sensor probe, the temperature needed to calculate the desired quantities (fuel density, fuel concentration) significantly differs from the gas phase temperature in the rest of the combustion chamber and estimations from standard thermodynamic models, e.g., polytropic compression, will fail.

  2. Dispersion- and Exchange-Corrected Density Functional Theory for Sodium Ion Hydration.

    Science.gov (United States)

    Soniat, Marielle; Rogers, David M; Rempe, Susan B

    2015-07-14

    A challenge in density functional theory is developing functionals that simultaneously describe intermolecular electron correlation and electron delocalization. Recent exchange-correlation functionals address those two issues by adding corrections important at long ranges: an atom-centered pairwise dispersion term to account for correlation and a modified long-range component of the electron exchange term to correct for delocalization. Here we investigate how those corrections influence the accuracy of binding free energy predictions for sodium-water clusters. We find that the dual-corrected ωB97X-D functional gives cluster binding energies closest to high-level ab initio methods (CCSD(T)). Binding energy decomposition shows that the ωB97X-D functional predicts the smallest ion-water (pairwise) interaction energy and larger multibody contributions for a four-water cluster than most other functionals - a trend consistent with CCSD(T) results. Also, ωB97X-D produces the smallest amounts of charge transfer and the least polarizable waters of the density functionals studied, which mimics the lower polarizability of CCSD. When compared with experimental binding free energies, however, the exchange-corrected CAM-B3LYP functional performs best (error <1 kcal/mol), possibly because of its parametrization to experimental formation enthalpies. For clusters containing more than four waters, "split-shell" coordination must be considered to obtain accurate free energies in comparison with experiment.

  3. 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

  4. 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)

  5. 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)

  6. 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

  7. 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.

  8. 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)

  9. 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.

  10. 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)

  11. 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.

  12. 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...

  13. 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...

  14. 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

  15. 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.

  16. 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

  17. 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.

  18. Inducing Strong Density Modulation with Small Energy Dispersion in Particle Beams and the Harmonic Amplifier Free Electron Laser

    CERN Document Server

    McNeil, Brian W J; Robb, Gordon

    2005-01-01

    We present a possible method of inducing a periodic density modulation in a particle beam with little increase in the energy dispersion of the particles. The flow of particles in phase space does not obey Liouville's Theorem. The method relies upon the Kuramoto-like model of collective synchronism found in free electron generators of radiation, such as Cyclotron Resonance Masers and the Free Electron Laser. For the case of an FEL interaction, electrons initially begin to bunch and emit radiation energy with a correlated energy dispersion which is periodic with the FEL ponderomotive potential. The relative phase between potential and particles is then changed by approximately 180 degrees. The particles continue to bunch, however, there is now a correlated re-absorption of energy from the field. We show that, by repeating this relative phase change many times, a significant density modulation of the particles may be achieved with only relatively small energy dispersion. A similar method of repeated relative ele...

  19. Validation of experimental molecular crystal structures with dispersion-corrected density functional theory calculations.

    Science.gov (United States)

    van de Streek, Jacco; Neumann, Marcus A

    2010-10-01

    This paper describes the validation of a dispersion-corrected density functional theory (d-DFT) method for the purpose of assessing the correctness of experimental organic crystal structures and enhancing the information content of purely experimental data. 241 experimental organic crystal structures from the August 2008 issue of Acta Cryst. Section E were energy-minimized in full, including unit-cell parameters. The differences between the experimental and the minimized crystal structures were subjected to statistical analysis. The r.m.s. Cartesian displacement excluding H atoms upon energy minimization with flexible unit-cell parameters is selected as a pertinent indicator of the correctness of a crystal structure. All 241 experimental crystal structures are reproduced very well: the average r.m.s. Cartesian displacement for the 241 crystal structures, including 16 disordered structures, is only 0.095 Å (0.084 Å for the 225 ordered structures). R.m.s. Cartesian displacements above 0.25 A either indicate incorrect experimental crystal structures or reveal interesting structural features such as exceptionally large temperature effects, incorrectly modelled disorder or symmetry breaking H atoms. After validation, the method is applied to nine examples that are known to be ambiguous or subtly incorrect.

  20. 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

  1. 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)

  2. Validation of molecular crystal structures from powder diffraction data with dispersion-corrected density functional theory (DFT-D)

    DEFF Research Database (Denmark)

    van de Streek, Jacco; Neumann, Marcus A

    2014-01-01

    In 2010 we energy-minimized 225 high-quality single-crystal (SX) structures with dispersion-corrected density functional theory (DFT-D) to establish a quantitative benchmark. For the current paper, 215 organic crystal structures determined from X-ray powder diffraction (XRPD) data and published...

  3. High energy density additives for Hybrid Fuel Rockets to Improve Performance and Enhance Safety

    Data.gov (United States)

    National Aeronautics and Space Administration — We propose a conceptual study of prototype strained hydrocarbon molecules as high energy density additives for hybrid rocket fuels to boost the performance of these...

  4. 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.

  5. 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.

  6. 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)

  7. 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)

  8. 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)

  9. 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)

  10. 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}.

  11. Kinetic and dynamic probability-density-function descriptions of disperse turbulent two-phase flows

    Science.gov (United States)

    Minier, Jean-Pierre; Profeta, Christophe

    2015-11-01

    This article analyzes the status of two classical one-particle probability density function (PDF) descriptions of the dynamics of discrete particles dispersed in turbulent flows. The first PDF formulation considers only the process made up by particle position and velocity Zp=(xp,Up) and is represented by its PDF p (t ;yp,Vp) which is the solution of a kinetic PDF equation obtained through a flux closure based on the Furutsu-Novikov theorem. The second PDF formulation includes fluid variables into the particle state vector, for example, the fluid velocity seen by particles Zp=(xp,Up,Us) , and, consequently, handles an extended PDF p (t ;yp,Vp,Vs) which is the solution of a dynamic PDF equation. For high-Reynolds-number fluid flows, a typical formulation of the latter category relies on a Langevin model for the trajectories of the fluid seen or, conversely, on a Fokker-Planck equation for the extended PDF. In the present work, a new derivation of the kinetic PDF equation is worked out and new physical expressions of the dispersion tensors entering the kinetic PDF equation are obtained by starting from the extended PDF and integrating over the fluid seen. This demonstrates that, under the same assumption of a Gaussian colored noise and irrespective of the specific stochastic model chosen for the fluid seen, the kinetic PDF description is the marginal of a dynamic PDF one. However, a detailed analysis reveals that kinetic PDF models of particle dynamics in turbulent flows described by statistical correlations constitute incomplete stand-alone PDF descriptions and, moreover, that present kinetic-PDF equations are mathematically ill posed. This is shown to be the consequence of the non-Markovian characteristic of the stochastic process retained to describe the system and the use of an external colored noise. Furthermore, developments bring out that well-posed PDF descriptions are essentially due to a proper choice of the variables selected to describe physical systems

  12. 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.%

  13. 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

  14. 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)

  15. 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

  16. 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

  17. 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.

  18. 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

  19. Comparison of renewable fuels based on their land use using energy densities

    NARCIS (Netherlands)

    Dijkman, T. J.; Benders, R. M. J.

    2010-01-01

    In this article energy densities of selected renewable fuels are determined. Energy density is defined here as the annual energy production per hectare, taking energy inputs into account. Using 5 scenarios, consisting of 1 set focusing on technical differences and 1 set focusing on geographical

  20. High precision measurement of fuel density profiles in nuclear fusion plasmas

    NARCIS (Netherlands)

    Svensson, J.; von Hellermann, M.; Konig, R.

    2002-01-01

    This paper presents a method for deducing fuel density profiles of nuclear fusion plasmas in realtime during an experiment. A Multi Layer Perceptron (MLP) neural network is used to create a mapping between plasma radiation spectra and indirectly deduced hydrogen isotope densities. By combining

  1. SPENT NUCLEAR FUEL NUMBER DENSITIES FOR MULTI-PURPOSE CANISTER CRITICALITY CALCULATIONS

    International Nuclear Information System (INIS)

    D. A. Thomas

    1996-01-01

    The purpose of this analysis is to calculate the number densities for spent nuclear fuel (SNF) to be used in criticality evaluations of the Multi-Purpose Canister (MPC) waste packages. The objective of this analysis is to provide material number density information which will be referenced by future MPC criticality design analyses, such as for those supporting the Conceptual Design Report

  2. 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

  3. 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.

  4. Fuel density effect on near nozzle flow field in small laminar coflow diffusion flames

    KAUST Repository

    Xiong, Yuan

    2015-01-01

    Flow characteristics in small coflow diffusion flames were investigated with a particular focus on the near-nozzle region and on the buoyancy force exerted on fuels with densities lighter and heavier than air (methane, ethylene, propane, and n-butane). The flow-fields were visualized through the trajectories of seed particles. The particle image velocimetry technique was also adopted for quantitative velocity field measurements. The results showed that the buoyancy force exerted on the fuel as well as on burnt gas significantly distorted the near-nozzle flow-fields. In the fuels with densities heavier than air, recirculation zones were formed very close to the nozzle, emphasizing the importance of the relative density of the fuel to that of the air on the flow-field. Nozzle heating influenced the near-nozzle flow-field particularly among lighter fuels (methane and ethylene). Numerical simulations were also conducted, focusing specifically on the effect of specifying inlet boundary conditions for fuel. The results showed that a fuel inlet boundary with a fully developed velocity profile for cases with long tubes should be specified inside the fuel tube to permit satisfactory prediction of the flow-field. The calculated temperature fields also indicated the importance of the selection of the location of the inlet boundary, especially in testing various combustion models that include soot in small coflow diffusion flames. © 2014 The Combustion Institute.

  5. 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.

  6. 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)

  7. 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)

  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. 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)

  11. 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.

  12. Determination of the density and the viscosities of biodiesel-diesel fuel blends

    Energy Technology Data Exchange (ETDEWEB)

    Alptekin, Ertan; Canakci, Mustafa [Department of Mechanical Education, Kocaeli University, 41380 Kocaeli (Turkey); Alternative Fuels R and D Center, Kocaeli University, 41040 Kocaeli (Turkey)

    2008-12-15

    In this study, commercially available two different diesel fuels were blended with the biodiesels produced from six different vegetable oils (sunflower, canola, soybean, cottonseed, corn oils and waste palm oil). The blends (B2, B5, B10, B20, B50 and B75) were prepared on a volume basis. The key fuel properties such as density and viscosities of the blends were measured by following ASTM test methods. Generalized equations for predicting the density and viscosities for the blends were given and a mixing equation, originally proposed by Arrhenius and described by Grunberg and Nissan, was used to predict the viscosities of the blends. For all blends, it was found that there is an excellent agreement between the measured and estimated values of the density and viscosities. According to the results, the density and viscosities of the blends increased with the increase of biodiesel concentration in the fuel blend. (author)

  13. 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.

  14. Density-Dependent Formulation of Dispersion-Repulsion Interactions in Hybrid Multiscale Quantum/Molecular Mechanics (QM/MM) Models

    DEFF Research Database (Denmark)

    Curutchet, Carles; Cupellini, Lorenzo; Kongsted, Jacob

    2018-01-01

    embedding approaches, respectively, nonelectrostatic dispersion and repulsion interactions are instead commonly described through classical potentials despite their quantum mechanical origin. Here we present an extension of the Tkatchenko-Scheffler semiempirical van der Waals (vdWTS) scheme aimed......Mixed multiscale quantum/molecular mechanics (QM/MM) models are widely used to explore the structure, reactivity, and electronic properties of complex chemical systems. Whereas such models typically include electrostatics and potentially polarization in so-called electrostatic and polarizable...... at describing dispersion and repulsion interactions between quantum and classical regions within a QM/MM polarizable embedding framework. Starting from the vdWTSexpression, we define a dispersion and a repulsion term, both of them density-dependent and consistently based on a Lennard-Jones-like potential. We...

  15. On the calculation of the structure of charge-stabilized colloidal dispersions using density-dependent potentials

    International Nuclear Information System (INIS)

    Castañeda-Priego, R; Lobaskin, V; Mixteco-Sánchez, J C; Rojas-Ochoa, L F; Linse, P

    2012-01-01

    The structure of charge-stabilized colloidal dispersions has been studied through a one-component model using a Yukawa potential with density-dependent parameters examined with integral equation theory and Monte Carlo simulations. Partial thermodynamic consistency was guaranteed by considering the osmotic pressure of the dispersion from the approximate mean-field renormalized jellium and Poisson-Boltzmann cell models. The colloidal structures could be accurately described by the Ornstein-Zernike equation with the Rogers-Young closure by using the osmotic pressure from the renormalized jellium model. Although we explicitly show that the correct effective pair-potential obtained from the inverse Monte Carlo method deviates from the Yukawa shape, the osmotic pressure constraint allows us to have a good description of the colloidal structure without losing information on the system thermodynamics. Our findings are corroborated by primitive model simulations of salt-free colloidal dispersions. (paper)

  16. 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.

  17. 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.

  18. 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)

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

    Science.gov (United States)

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

    2017-08-01

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

  20. Density-Dependent Formulation of Dispersion-Repulsion Interactions in Hybrid Multiscale Quantum/Molecular Mechanics (QM/MM) Models.

    Science.gov (United States)

    Curutchet, Carles; Cupellini, Lorenzo; Kongsted, Jacob; Corni, Stefano; Frediani, Luca; Steindal, Arnfinn Hykkerud; Guido, Ciro A; Scalmani, Giovanni; Mennucci, Benedetta

    2018-03-13

    Mixed multiscale quantum/molecular mechanics (QM/MM) models are widely used to explore the structure, reactivity, and electronic properties of complex chemical systems. Whereas such models typically include electrostatics and potentially polarization in so-called electrostatic and polarizable embedding approaches, respectively, nonelectrostatic dispersion and repulsion interactions are instead commonly described through classical potentials despite their quantum mechanical origin. Here we present an extension of the Tkatchenko-Scheffler semiempirical van der Waals (vdW TS ) scheme aimed at describing dispersion and repulsion interactions between quantum and classical regions within a QM/MM polarizable embedding framework. Starting from the vdW TS expression, we define a dispersion and a repulsion term, both of them density-dependent and consistently based on a Lennard-Jones-like potential. We explore transferable atom type-based parametrization strategies for the MM parameters, based on either vdW TS calculations performed on isolated fragments or on a direct estimation of the parameters from atomic polarizabilities taken from a polarizable force field. We investigate the performance of the implementation by computing self-consistent interaction energies for the S22 benchmark set, designed to represent typical noncovalent interactions in biological systems, in both equilibrium and out-of-equilibrium geometries. Overall, our results suggest that the present implementation is a promising strategy to include dispersion and repulsion in multiscale QM/MM models incorporating their explicit dependence on the electronic density.

  1. Comment on: Power loss and electromagnetic energy density in a dispersive metamaterial medium

    OpenAIRE

    Kinsler, Paul

    2010-01-01

    By clarifying the approach of Luan (Phys. Rev. E, 2009), we can generalize the analysis of dispersive (meta)materials, and treat other material responses involving not only loss, but also gain and coherent response.

  2. 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

  3. A mean-density model of ionic surfactants for the dispersion of carbon nanotubes in aqueous solutions

    Science.gov (United States)

    Joung, Young Soo

    2018-05-01

    We propose a new analytical model of ionic surfactants used for the dispersion of carbon nanotubes (CNTs) in aqueous solutions. Although ionic surfactants are commonly used to facilitate the dispersion of CNTs in aqueous solutions, understanding the dispersion process is challenging and time-consuming owing to its complexity and nonlinearity. In this work, we develop a mean-density model of ionic surfactants to simplify the calculation of interaction forces between CNTs stabilized by ionic surfactants. Using this model, we can evaluate various interaction forces between the CNTs and ionic surfactants under different conditions. The dispersion mechanism is investigated by estimating the potential of mean force (PMF) as a function of van der Waals forces, electrostatic forces, interfacial tension, and osmotic pressure. To verify the proposed model, we compare the PMFs derived using our method with those derived from molecular dynamics simulations using comparable CNTs and ionic surfactants. Notably, for stable dispersions, the osmotic pressure and interfacial energy are important for long-range and short-range interactions, respectively, in comparison with the effect of electrostatic forces. Our model effectively prescribes specific surfactants and their concentrations to achieve stable aqueous suspensions of CNTs.

  4. 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...

  5. 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.

  6. 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)

  7. 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)

  8. 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

  9. Thermophysical properties of dispersed metal materials

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Hyun; Kim, Jong Chul [KRISS, Taejeon (Korea)

    2002-04-01

    Thermal conductivities of the preliminarily fabricated U-Mo dispersion fuel meats have been measured to estimate the center temperature of the irradiation fuels. Thermal conductivities at temperatures ranging from room temperature to 500 .deg. C were calculated by measuring diffusivities, specific heat capacities and densities of dispersion fuel meats. The molybdenum content of fuel meats was varied to be 6wt%, 8wt%, and 10wt% and the volume fraction of U-Mo fuel powders were changed to be 10 vol%, 30vol%, 40 vol%, and 50 vol%. 13 refs., 39 figs., 13 tabs. (Author)

  10. Low-density moderation in the storage of PWR fuel assemblies

    International Nuclear Information System (INIS)

    Alcorn, F.M.

    1987-01-01

    The nuclear criticality safety of PWR fuel storage arrays requires that the potential of low-density moderation within the array be considered. The calculated criticality effect of low-density moderation in a typical PWR fuel assembly array is described in this paper. Calculated reactivity due to low-density moderation can vary significantly between physics codes that have been validated for well moderated systems. The availability of appropriate benchmark experiments for low-density moderation is quite limited; attempts to validate against the one set of suitable experiments at low density have been disappointing. Calculations indicate that a typical array may be unacceptable should the array be subjected to interstitial moderation equivalent to 5 % of full density water. Array parameters (such as spacing and size) will dramatically affect the calculated maximum K-eff at low-density moderation. Administrative and engineered control may be necessary to assure maintenance of safety at low-density moderation. Potential sources for low-density moderation are discussed; in general, accidentally achieving degrees of low-density moderation which might lead to a compromise of safety are not credible. (author)

  11. 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

  12. Azobenzene-functionalized carbon nanotubes as high-energy density solar thermal fuels.

    Science.gov (United States)

    Kolpak, Alexie M; Grossman, Jeffrey C

    2011-08-10

    Solar thermal fuels, which reversibly store solar energy in molecular bonds, are a tantalizing prospect for clean, renewable, and transportable energy conversion/storage. However, large-scale adoption requires enhanced energy storage capacity and thermal stability. Here we present a novel solar thermal fuel, composed of azobenzene-functionalized carbon nanotubes, with the volumetric energy density of Li-ion batteries. Our work also demonstrates that the inclusion of nanoscale templates is an effective strategy for design of highly cyclable, thermally stable, and energy-dense solar thermal fuels.

  13. 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

  14. 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)

  15. Determination of power density distribution of fuel assemblies for research reactor by directly measuring the strontium-91 activities

    International Nuclear Information System (INIS)

    Yuan, Liq-Ji

    1987-01-01

    This work described the investigations of reactor core power peaking and three dimensional power density distribution of present core configuration of Tsing Hua Open-pool reactor (THOR). An experimental program, based on non-destructive fuel gamma scanning of 91 Sr activities, provides the data of fission density distribution for individual fuel pin of four-rod TRIGA-LEU cluster or for MTR-type fuel assembly. The informations are essentially important for the safety of reactor operation and for fuel management especially for the mixed loading with three different types of fuel at present. The relative power peaking values and the power density distribution for present core are discussed. (author)

  16. Dispersion of fruit flies (Diptera: Tephritidae) at high and low densities and consequences of mismatching dispersions of wild and sterile flies

    International Nuclear Information System (INIS)

    Meats, A.

    2007-01-01

    Both wild and released (sterile) Bactrocera tryoni (Froggatt) (Diptera: Tephritidae) and wild Bactrocera papayae (Drew and Hancock) in Australia had patchy distributions and comparisons with predictions of the negative binomial model indicated that the degree of clumping was sometimes very high, particularly at low densities during eradication. An increase of mean recapture rate of sterile B. tryoni on either of 2 trap arrays was not accompanied by a reduction in its coefficient of variation and when recapture rates were high, the percentage of traps catching zero decreased only slightly with increase in recapture rate, indicating that it is not practicable to decrease the heterogeneity of dispersion of sterile flies by increasing the number released. There was often a mismatch between the dispersion patterns of the wild and sterile flies, and the implications of this for the efficiency of the sterile insect technique (SIT) were investigated with a simulation study with the observed degrees of mismatch obtained from the monitoring data and assuming the overall ratio of sterile to wild flies to be 100:1. The simulation indicated that mismatches could result in the imposed rate of increase of wild flies being up to 3.5 times higher than that intended (i.e., 0.35 instead of 0.1). The effect of a mismatch always reduces the efficiency of SIT. The reason for this asymmetry is discussed and a comparison made with host-parasitoid and other systems. A release strategy to counter this effect is suggested. (author) [es

  17. Probability density function shape sensitivity in the statistical modeling of turbulent particle dispersion

    Science.gov (United States)

    Litchford, Ron J.; Jeng, San-Mou

    1992-01-01

    The performance of a recently introduced statistical transport model for turbulent particle dispersion is studied here for rigid particles injected into a round turbulent jet. Both uniform and isosceles triangle pdfs are used. The statistical sensitivity to parcel pdf shape is demonstrated.

  18. Benchmarking of London Dispersion-Accounting Density Functional Theory Methods on Very Large Molecular Complexes.

    Science.gov (United States)

    Risthaus, Tobias; Grimme, Stefan

    2013-03-12

    A new test set (S12L) containing 12 supramolecular noncovalently bound complexes is presented and used to evaluate seven different methods to account for dispersion in DFT (DFT-D3, DFT-D2, DFT-NL, XDM, dDsC, TS-vdW, M06-L) at different basis set levels against experimental, back-corrected reference energies. This allows conclusions about the performance of each method in an explorative research setting on "real-life" problems. Most DFT methods show satisfactory performance but, due to the largeness of the complexes, almost always require an explicit correction for the nonadditive Axilrod-Teller-Muto three-body dispersion interaction to get accurate results. The necessity of using a method capable of accounting for dispersion is clearly demonstrated in that the two-body dispersion contributions are on the order of 20-150% of the total interaction energy. MP2 and some variants thereof are shown to be insufficient for this while a few tested D3-corrected semiempirical MO methods perform reasonably well. Overall, we suggest the use of this benchmark set as a "sanity check" against overfitting to too small molecular cases.

  19. Core design and fuel rod analyses of a super fast reactor with high power density

    International Nuclear Information System (INIS)

    Ju, Haitao; Cao, Liangzhi; Lu, Haoliang; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) that is presently researched in a Japanese project. One of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. A preliminary core has an average power density of 158.8W/cc. In this paper, the principle of improving the average power density is studied and the core design is improved. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. This power density is competitive with that of typical Liquid Metal Fast Breeder Reactors (LMFBR). In order to ensure the fuel rod integrity of this core design, the fuel rod behaviors on the normal operating condition are analyzed using FEMAXI-6 code. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are taken from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak(MPP), Maximum Discharge Burnup(MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900degC. (2) Maximum cladding stress in circumstance direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be

  20. Estimating Brownian motion dispersal rate, longevity and population density from spatially explicit mark-recapture data on tropical butterflies.

    Science.gov (United States)

    Tufto, Jarle; Lande, Russell; Ringsby, Thor-Harald; Engen, Steinar; Saether, Bernt-Erik; Walla, Thomas R; DeVries, Philip J

    2012-07-01

    1. We develop a Bayesian method for analysing mark-recapture data in continuous habitat using a model in which individuals movement paths are Brownian motions, life spans are exponentially distributed and capture events occur at given instants in time if individuals are within a certain attractive distance of the traps. 2. The joint posterior distribution of the dispersal rate, longevity, trap attraction distances and a number of latent variables representing the unobserved movement paths and time of death of all individuals is computed using Gibbs sampling. 3. An estimate of absolute local population density is obtained simply by dividing the Poisson counts of individuals captured at given points in time by the estimated total attraction area of all traps. Our approach for estimating population density in continuous habitat avoids the need to define an arbitrary effective trapping area that characterized previous mark-recapture methods in continuous habitat. 4. We applied our method to estimate spatial demography parameters in nine species of neotropical butterflies. Path analysis of interspecific variation in demographic parameters and mean wing length revealed a simple network of strong causation. Larger wing length increases dispersal rate, which in turn increases trap attraction distance. However, higher dispersal rate also decreases longevity, thus explaining the surprising observation of a negative correlation between wing length and longevity. © 2012 The Authors. Journal of Animal Ecology © 2012 British Ecological Society.

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

    International Nuclear Information System (INIS)

    Hey, Martin A.; Gomez Marlasca, Fernando

    2003-01-01

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

  2. 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)

  3. 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

  4. Influence of seed density and aggregation on post-dispersal weed seed predation in cereal fields

    NARCIS (Netherlands)

    Marino, P.C.; Westerman, P.R.; Pinkert, C.; Werf, van der W.

    2005-01-01

    The importance of density dependence, aggregation and background density of seeds on intensity of seed predation in cereal fields were examined in central Netherlands. Four sequential 1-week trials were conducted from 9 July to 8 August 2001 and lamb's quarters (Chenopodium album) was used as the

  5. The neutrons flux density calculations by Monte Carlo code for the double heterogeneity fuel

    International Nuclear Information System (INIS)

    Gurevich, M.I.; Brizgalov, V.I.

    1994-01-01

    This document provides the calculation technique for the fuel elements which consists of the one substance as a matrix and the other substance as the corn embedded in it. This technique can be used in the neutron flux density calculation by the universal Monte Carlo code. The estimation of accuracy is presented too. (authors). 6 refs., 1 fig

  6. The impact of residential density on vehicle usage and fuel consumption: Evidence from national samples

    DEFF Research Database (Denmark)

    Kim, Jinwon; Brownstone, David

    2013-01-01

    This paper investigates the impact of residential density on household vehicle usage and fuel consumption. We estimate a simultaneous equations system to account for the potential residential self-selection problem. While most previous studies focus on a specific region, this paper uses national...

  7. Towards High Power Density Metal Supported Solid Oxide Fuel Cell for Mobile Applications

    DEFF Research Database (Denmark)

    Nielsen, Jimmi; Persson, Åsa H.; Muhl, Thuy Thanh

    2018-01-01

    For use of metal supported solid oxide fuel cell (MS-SOFC) in mobile applications it is important to reduce the thermal mass to enable fast startup, increase stack power density in terms of weight and volume and reduce costs. In the present study, we report on the effect of reducing the Technical...

  8. Direct alcohol fuel cells: toward the power densities of hydrogen-fed proton exchange membrane fuel cells.

    Science.gov (United States)

    Chen, Yanxin; Bellini, Marco; Bevilacqua, Manuela; Fornasiero, Paolo; Lavacchi, Alessandro; Miller, Hamish A; Wang, Lianqin; Vizza, Francesco

    2015-02-01

    A 2 μm thick layer of TiO2 nanotube arrays was prepared on the surface of the Ti fibers of a nonwoven web electrode. After it was doped with Pd nanoparticles (1.5 mgPd  cm(-2) ), this anode was employed in a direct alcohol fuel cell. Peak power densities of 210, 170, and 160 mW cm(-2) at 80 °C were produced if the cell was fed with 10 wt % aqueous solutions of ethanol, ethylene glycol, and glycerol, respectively, in 2 M aqueous KOH. The Pd loading of the anode was increased to 6 mg cm(-2) by combining four single electrodes to produce a maximum peak power density with ethanol at 80 °C of 335 mW cm(-2) . Such high power densities result from a combination of the open 3 D structure of the anode electrode and the high electrochemically active surface area of the Pd catalyst, which promote very fast kinetics for alcohol electro-oxidation. The peak power and current densities obtained with ethanol at 80 °C approach the output of H2 -fed proton exchange membrane fuel cells. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  9. Dispersion forces in a four-component density functional theory framework

    OpenAIRE

    Pilemalm, Robert

    2009-01-01

    The main purpose of this thesis is to implement the Gauss--Legendre quadrature for the dispersion coefficient. This has been done and can be now be made with different number of points. The calculations with this implementation has shown that the relativistic impact on helium, neon, argon and krypton is largest for krypton, that has the highest charge of its nucleus. It was also seen that the polarizability of neon as a function of the imaginary angular frequency decreases monotonically from ...

  10. 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)

  11. Thermal Expansion and Density Data of UO2 and Simulated Fuel for Standard Reference

    International Nuclear Information System (INIS)

    Yang, Jae Hwan; Na, S. H.; Lee, J. W.; Kang, K. H.

    2010-01-01

    Standard Reference Data (SRD) is the scientific, technical data whose reliability and accuracy are evaluated by scientist group. Since SRD has a great impact on the improvement of national competitiveness by stirring up technological innovation in every sector of industries, many countries are making great efforts on establishing SRD in various areas. Data center for nuclear fuel material in Korea Atomic Energy Research Institute plays a role to providing property data of nuclear fuel material at high temperature, pressure, and radiation which are essential for the safety evaluation of nuclear power. In this study, standardization of data on thermal expansion and density of UO 2 were carried out in the temperature range from 300 K to 3100 K via uncertainty evaluation of indirectly produced data. Besides, standardization of data on thermal expansion and density of simulated fuel were also done in the temperature range from 350 K to 1750 K via uncertainty evaluation of directly produced data

  12. 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

  13. Boundedness and permanence in a class of periodic time-dependent predator-prey system with prey dispersal and predator density-independence

    International Nuclear Information System (INIS)

    Zhang Long; Teng Zhidong

    2008-01-01

    In this paper, we study two species predator-prey Lotka-Volterra type dispersal system with periodic coefficients, in which the prey species can disperse among n patches, while the density-independent predator species is confined to one of the patches and cannot disperse. Sufficient conditions on the boundedness, permanence and existence of positive periodic solution for this system are established. The theoretical results are confirmed by a special example and numerical simulations

  14. Method and apparatus for storing nuclear fuel assemblies in maximum density racks

    International Nuclear Information System (INIS)

    Wachter, W.J.; Robbins, T.R.

    1979-01-01

    A maximum density storage rack is provided for long term or semipermanent storage of spent nuclear fuel assemblies. The rack consists of storage cells arranged in a regular array, such as a checkerboard, and intended to be immersed in water. Initially, cap members are placed on alternate cells in such a manner that at least 50% of the cells are left open, some of the caps being removable. Spent fuel assemblies are then placed in the open cells until all of them are filled. The level of reactivity of each of the stored fuel assemblies is then determined by accurate calculation or by measurement, and the removable caps are removed and rearranged so that other cells are opened, permitting the storage of additional fuel assemblies in a pattern based on the actual reactivity such that criticality is prevented

  15. 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.

  16. 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)

  17. Thermal characterizations of the paraffin wax/low density polyethylene blends as a solid fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Soojong; Moon, Heejang; Kim, Jinkon, E-mail: jkkim@kau.ac.kr

    2015-08-10

    Highlights: • Regression rate of blends fuel is higher than polymer fuel. • LDPE is an effective mixing ingredient for the combustion efficiency. • Blends fuel is a uniform mixture with two degradation steps. • LDPE plays a positive role for the low sensitivity to the thermal deformation • Blends with low LDPE content can be an effective fuel for hybrid rocket application. - Abstract: Thermal characterizations of a novel solid fuel for hybrid rocket application, based on the paraffin wax blends with low density polyethylene (LDPE) concentration of 5% (SF-5) and 10% (SF-10) were conducted. Both the increased regression rate in comparison with the polymeric fuel, and the improved combustion efficiency in comparison with the pure paraffin fuel reveal that the blend fuels achieve higher combustion performance. The morphology of the shape stabilized paraffin wax/LDPE blends was characterized by the scanning electron microscopy (SEM). Although the SEM observation indicated the blends have uniform mixtures, they showed two degradation steps confirming the immiscibility of components in the crystalline phase from thermogravimetric analysis (TGA). The differential scanning calorimeter (DSC) results showed that the melting temperature of LDPE in the blends decreased with an increase of paraffin wax content. The decreasing total specific melting enthalpy of blended fuels with decreasing paraffin wax content is in fairly good agreement with the additive rule. In thermomechanical analysis (TMA), the linear coefficient of thermal expansion (LCTE) seems to decrease with an increase of LDPE loading, however, the loaded LDPE do merely affect the LCTE in case of the blends with low LDPE concentration. It was found that a blend of low concentration of LDPE with a relatively high concentration of paraffin wax can lead to a potential novel fuel for rocket application, a contrary case with respect to the field of phase change materials (PCM) where a blend of high concentration

  18. Thermal characterizations of the paraffin wax/low density polyethylene blends as a solid fuel

    International Nuclear Information System (INIS)

    Kim, Soojong; Moon, Heejang; Kim, Jinkon

    2015-01-01

    Highlights: • Regression rate of blends fuel is higher than polymer fuel. • LDPE is an effective mixing ingredient for the combustion efficiency. • Blends fuel is a uniform mixture with two degradation steps. • LDPE plays a positive role for the low sensitivity to the thermal deformation • Blends with low LDPE content can be an effective fuel for hybrid rocket application. - Abstract: Thermal characterizations of a novel solid fuel for hybrid rocket application, based on the paraffin wax blends with low density polyethylene (LDPE) concentration of 5% (SF-5) and 10% (SF-10) were conducted. Both the increased regression rate in comparison with the polymeric fuel, and the improved combustion efficiency in comparison with the pure paraffin fuel reveal that the blend fuels achieve higher combustion performance. The morphology of the shape stabilized paraffin wax/LDPE blends was characterized by the scanning electron microscopy (SEM). Although the SEM observation indicated the blends have uniform mixtures, they showed two degradation steps confirming the immiscibility of components in the crystalline phase from thermogravimetric analysis (TGA). The differential scanning calorimeter (DSC) results showed that the melting temperature of LDPE in the blends decreased with an increase of paraffin wax content. The decreasing total specific melting enthalpy of blended fuels with decreasing paraffin wax content is in fairly good agreement with the additive rule. In thermomechanical analysis (TMA), the linear coefficient of thermal expansion (LCTE) seems to decrease with an increase of LDPE loading, however, the loaded LDPE do merely affect the LCTE in case of the blends with low LDPE concentration. It was found that a blend of low concentration of LDPE with a relatively high concentration of paraffin wax can lead to a potential novel fuel for rocket application, a contrary case with respect to the field of phase change materials (PCM) where a blend of high concentration

  19. 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.

  20. 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)

  1. 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

  2. Validation of missed space-group symmetry in X-ray powder diffraction structures with dispersion-corrected density functional theory

    DEFF Research Database (Denmark)

    Hempler, Daniela; Schmidt, Martin U.; Van De Streek, Jacco

    2017-01-01

    More than 600 molecular crystal structures with correct, incorrect and uncertain space-group symmetry were energy-minimized with dispersion-corrected density functional theory (DFT-D, PBE-D3). For the purpose of determining the correct space-group symmetry the required tolerance on the atomic...... with missed symmetry were investigated by dispersion-corrected density functional theory. In 98.5% of the cases the correct space group is found....

  3. Calculation of nucleon densities in calcium, nickel, and molybdenum isotopes on the basis of the dispersive optical model

    Science.gov (United States)

    Bespalova, O. V.; Klimochkina, A. A.

    2017-09-01

    The radial distributions of proton and neutron densities in the even-even isotopes 40-70Ca and 48-78Ni and the analogous distributions of neutron densities in the even-even isotopes 92-138Mo were calculated on the basis of the mean-fieldmodel involving a dispersive optical potential. The respective root-mean-square radii and neutron-skin thicknesses were determined for the nuclei under study. In N > 40 calcium isotopes, the calculated neutron root-mean-square radius exhibits a fast growth with increasing N, and this is consistent with the prediction of the neutron-halo structure in calcium isotopes near the neutron drip line.

  4. Experimental determination of the phonon state density in the bulk and finely dispersed polycrystalline Fe1-xO

    International Nuclear Information System (INIS)

    Bulat, I.A.; Pashkovskij, Yu.L.; Semencheva, O.P.

    1990-01-01

    Density of phonon states, that is, g(ε) of polycrystalline Fe 1-x O at x=0.08 and intermediate crystallites equl to 100 and smaller, than 1μm is measured using inelastic scattering of cold neutrons at E 0 =4.43 MeV and T=293 K. Technique of separation of contributions of neutron lattice and magnetic scattering at transitions of Fe 2+ ion orbital moment which is incompletely frozen was used for measurements. Additional peculiarities of E(ε) at low energy values and density of finely dispersed powder states which is essentially suppressed, within optical frequency range are explained by presence of vacancies and defect clusters

  5. Defining the contributions of permanent electrostatics, Pauli repulsion, and dispersion in density functional theory calculations of intermolecular interaction energies

    Energy Technology Data Exchange (ETDEWEB)

    Horn, Paul R., E-mail: prhorn@berkeley.edu; Mao, Yuezhi; Head-Gordon, Martin, E-mail: mhg@cchem.berkeley.edu [Kenneth S. Pitzer Center for Theoretical Chemistry, Department of Chemistry, University of California, Berkeley, California 94720, USA and Chemical Sciences Division Lawrence Berkeley National Laboratory Berkeley, California 94720 (United States)

    2016-03-21

    In energy decomposition analysis of Kohn-Sham density functional theory calculations, the so-called frozen (or pre-polarization) interaction energy contains contributions from permanent electrostatics, dispersion, and Pauli repulsion. The standard classical approach to separate them suffers from several well-known limitations. We introduce an alternative scheme that employs valid antisymmetric electronic wavefunctions throughout and is based on the identification of individual fragment contributions to the initial supersystem wavefunction as determined by an energetic optimality criterion. The density deformations identified with individual fragments upon formation of the initial supersystem wavefunction are analyzed along with the distance dependence of the new and classical terms for test cases that include the neon dimer, ammonia borane, water-Na{sup +}, water-Cl{sup −}, and the naphthalene dimer.

  6. Strongly coupled single-phase flow problems: Effects of density variation, hydrodynamic dispersion, and first order decay

    Energy Technology Data Exchange (ETDEWEB)

    Oldenburg, C.M.; Pruess, K. [Lawrence Berkeley Laboratory, Berkeley, CA (United States)

    1995-03-01

    We have developed TOUGH2 modules for strongly coupled flow and transport that include full hydrodynamic dispersion. T2DM models tow-dimensional flow and transport in systems with variable salinity, while T32DMR includes radionuclide transport with first-order decay of a parent-daughter chain of radionuclide components in variable salinity systems. T2DM has been applied to a variety of coupled flow problems including the pure solutal convection problem of Elder and the mixed free and forced convection salt-dome flow problem. In the Elder and salt-dome flow problems, density changes of up to 20% caused by brine concentration variations lead to strong coupling between the velocity and brine concentration fields. T2DM efficiently calculates flow and transport for these problems. We have applied T2DMR to the dispersive transport and decay of radionuclide tracers in flow fields with permeability heterogeneities and recirculating flows. Coupling in these problems occurs by velocity-dependent hydrodynamic dispersion. Our results show that the maximum daughter species concentration may occur fully within a recirculating or low-velocity region. In all of the problems, we observe very efficient handling of the strongly coupled flow and transport processes.

  7. Density-functional approach to the three-body dispersion interaction based on the exchange dipole moment

    Energy Technology Data Exchange (ETDEWEB)

    Proynov, Emil; Wang, Matthew; Kong, Jing, E-mail: jing.kong@mtsu.edu [Department of Chemistry and Center for Computational Sciences, Middle Tennessee State University, Murfreesboro, Tennessee 37132 (United States); Liu, Fenglai [Department of Chemical and Biological Engineering, University at Buffalo, Buffalo, New York 14260 (United States); Gan, Zhengting [Q-Chem Inc., 5001 Baum Boulevard, Pittsburgh, Pennsylvania 15213 (United States)

    2015-08-28

    We implement and compute the density functional nonadditive three-body dispersion interaction using a combination of Tang-Karplus formalism and the exchange-dipole moment model of Becke and Johnson. The computation of the C{sub 9} dispersion coefficients is done in a non-empirical fashion. The obtained C{sub 9} values of a series of noble atom triplets agree well with highly accurate values in the literature. We also calculate the C{sub 9} values for a series of benzene trimers and find a good agreement with high-level ab initio values reported recently in the literature. For the question of damping of the three-body dispersion at short distances, we propose two damping schemes and optimize them based on the benzene trimers data, and the fitted analytic potentials of He{sub 3} and Ar{sub 3} trimers fitted to the results of high-level wavefunction theories available from the literature. Both damping schemes respond well to the optimization of two parameters.

  8. Density and Viscosity Measurement of Diesel Fuels at Combined High Pressure and Elevated Temperature

    Directory of Open Access Journals (Sweden)

    Carl Schaschke

    2013-07-01

    Full Text Available We report the measurement of the viscosity and density of various diesel fuels, obtained from British refineries, at elevated pressures up to 500 MPa and temperatures in the range 298 K to 373 K. The measurement and prediction procedures of fluid properties under high pressure conditions is of increasing interest in many processes and systems including enhanced oil recovery, automotive engine fuel injection, braking, and hydraulic systems. Accurate data and understanding of the fluid characteristic in terms of pressure, volume and temperature is required particularly where the fluid is composed of a complex mixture or blend of aliphatic or aromatic hydrocarbons. In this study, high pressure viscosity data was obtained using a thermostatically-controlled falling sinker-type high pressure viscometer to provide reproducible and reliable viscosity data based on terminal velocity sinker fall times. This was supported with density measurements using a micro-pVT device. Both high-pressure devices were additionally capable of illustrating the freezing points of the hydrocarbon mixtures. This work has, thus, provided data that can extend the application of mixtures of commercially available fuels and to test the validity of available predictive density and viscosity models. This included a Tait-style equation for fluid compressibility prediction. For complex diesel fuel compositions, which have many unidentified components, the approach illustrates the need to apply appropriate correlations, which require accurate knowledge or prediction of thermodynamic properties.

  9. 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.

  10. Dispersal, density dependence, and population dynamics of a fungal microbe on leaf surfaces.

    Science.gov (United States)

    Woody, Scott T; Ives, Anthony R; Nordheim, Erik V; Andrews, John H

    2007-06-01

    Despite the ubiquity and importance of microbes in nature, little is known about their natural population dynamics, especially for those that occupy terrestrial habitats. Here we investigate the dynamics of the yeast-like fungus Aureobasidium pullulans (Ap) on apple leaves in an orchard. We asked three questions. (1) Is variation in fungal population density among leaves caused by variation in leaf carrying capacities and strong density-dependent population growth that maintains densities near carrying capacity? (2) Do resident populations have competitive advantages over immigrant cells? (3) Do Ap dynamics differ at different times during the growing season? To address these questions, we performed two experiments at different times in the growing season. Both experiments used a 2 x 2 factorial design: treatment 1 removed fungal cells from leaves to reveal density-dependent population growth, and treatment 2 inoculated leaves with an Ap strain engineered to express green fluorescent protein (GFP), which made it possible to track the fate of immigrant cells. The experiments showed that natural populations of Ap vary greatly in density due to sustained differences in carrying capacities among leaves. The maintenance of populations close to carrying capacities indicates strong density-dependent processes. Furthermore, resident populations are strongly competitive against immigrants, while immigrants have little impact on residents. Finally, statistical models showed high population growth rates of resident cells in one experiment but not in the other, suggesting that Ap experiences relatively "good" and "bad" periods for population growth. This picture of Ap dynamics conforms to commonly held, but rarely demonstrated, expectations of microbe dynamics in nature. It also highlights the importance of local processes, as opposed to immigration, in determining the abundance and dynamics of microbes on surfaces in terrestrial systems.

  11. Application of neurite orientation dispersion and density imaging (NODDI) to a tau pathology model of Alzheimer's disease.

    LENUS (Irish Health Repository)

    Colgan, N

    2015-10-23

    Increased hyperphosphorylated tau and the formation of intracellular neurofibrillary tangles are associated with the loss of neurons and cognitive decline in Alzheimer\\'s disease, and related neurodegenerative conditions. We applied two diffusion models, diffusion tensor imaging (DTI) and neurite orientation dispersion and density imaging (NODDI), to in vivo diffusion magnetic resonance images (dMRI) of a mouse model of human tauopathy (rTg4510) at 8.5months of age. In grey matter regions with the highest degree of tau burden, microstructural indices provided by both NODDI and DTI discriminated the rTg4510 (TG) animals from wild type (WT) controls; however only the neurite density index (NDI) (the volume fraction that comprises axons or dendrites) from the NODDI model correlated with the histological measurements of the levels of hyperphosphorylated tau protein. Reductions in diffusion directionality were observed when implementing both models in the white matter region of the corpus callosum, with lower fractional anisotropy (DTI) and higher orientation dispersion (NODDI) observed in the TG animals. In comparison to DTI, histological measures of tau pathology were more closely correlated with NODDI parameters in this region. This in vivo dMRI study demonstrates that NODDI identifies potential tissue sources contributing to DTI indices and NODDI may provide greater specificity to pathology in Alzheimer\\'s disease.

  12. 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%.

  13. 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%

  14. 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

  15. Radial power density distribution of MOX fuel rods in the IFA-651

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Ho; Koo, Yang Hyun; Joo, Hyung Kook; Cheon, Jin Sik; Oh, Je Yong; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    Two MOX fuel rods, which were fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with Korea Atomic Energy Research Institute, have been irradiated in the HBWR from June, 2000 in the framework of OECD-HRP together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is basic in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR{sub H}BWR that calculates radial power density distribution for three MOX fuel rods has been developed based on neutron physics results and DEPRESS program. The developed subroutine FACTOR{sub H}BWR gives good agreement with the physics calculation except slight under-prediction at the outer part of the pellet above the burnup of 20 MWd/kgHM. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. 24 figs., 4 tabs. (Author)

  16. Radial power density distribution of MOX fuel rods in the HBWR

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Joo, Hyung Kook; Lee, Byung Ho; Sohn, Dong Seong

    1999-07-01

    Two MOX fuel rods, which ar being fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with the Korea Atomic Energy Research Institute (KAERI), are going to be irradiated in the HBWR (Halden Boiling Water Reactor) from the beginning of 2000 in the framework of OECD Halden Reactor Programme (HRP) together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is a basic property in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR H BWR that calculates radial power density distribution for three MOX fuel rods have been developed subroutine FACTOR H BWR gives good agreement with the physics calculation except slight underprediction in the central part and a little overprediction at the outer part of the pellet. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. (author). 5 refs., 3 tabs., 24 figs

  17. Calculation of the effects of pumping, divertor configuration and fueling on density limit in a tokamak model problem

    International Nuclear Information System (INIS)

    Stacey, W. M.

    2001-01-01

    Several series of model problem calculations have been performed to investigate the predicted effect of pumping, divertor configuration and fueling on the maximum achievable density in diverted tokamaks. Density limitations due to thermal instabilities (confinement degradation and multifaceted axisymmetric radiation from the edge) and to divertor choking are considered. For gas fueling the maximum achievable density is relatively insensitive to pumping (on or off), to the divertor configuration (open or closed), or to the location of the gas injection, although the gas fueling rate required to achieve this maximum achievable density is quite sensitive to these choices. Thermal instabilities are predicted to limit the density at lower values than divertor choking. Higher-density limits are predicted for pellet injection than for gas fueling

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

    International Nuclear Information System (INIS)

    Silva, Jose Eduardo Rosa da

    2011-01-01

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

  19. 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.

  20. 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

  1. Fuel density effect on parameter of reactivity coefficient of the Innovative Research Reactor core

    International Nuclear Information System (INIS)

    Rokhmadi; Tukiran S

    2013-01-01

    The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research reactor in Indonesia right now is already 25 year old. Therefor, it is needed to design a new research reactor as a alternative called it innovative research reactor (IRR) and then as an exchanger for old research reactor. The aim of this research is to complete RRI core design data as a requirement for design license. Calculation done is to get the RRI core reactivity coefficients with 5 x 5 core configuration and 20 MW of power, has more than 40 days cycle of length. The RRI core reactivity coefficient calculation is done for new U-"9Mo-Al fuel with variation of densities. The calculation is done by using WIMSD-5B and BATAN-FUEL computer codes. The result of calculation for conceptual design showed that the equilibrium RRI core with 5 x 5 configuration, 450 g, 550 g and 700 g of fuel loadings have negative reactivity coefficients of fuel temperature, moderator temperature, void fraction and density of moderator but the values of the reactivities are very variation. This results has met the safety criteria for RRI core conceptual design. (author)

  2. Current density distribution mapping in PEM fuel cells as an instrument for operational measurements

    Energy Technology Data Exchange (ETDEWEB)

    Geske, M.; Heuer, M.; Heideck, G.; Styczynski, Z. A. [Otto-von-Guericke University Magdeburg, Chair Electric Power Networks and Renewable Energy Sources, Magdeburg (Germany)

    2010-07-01

    A newly developed measurement system for current density distribution mapping has enabled a new approach for operational measurements in proton exchange membrane fuel cells (PEMFC). Taking into account previously constructed measurement systems, a method based on a multi layer printed circuit board was chosen for the development of the new system. This type of system consists of a sensor, a special electronic device and the control and visualization PC. For the acquisition of the current density distribution values, a sensor device was designed and installed within a multilayer printed circuit board with integrated shunt resistors. Varying shunt values can be taken into consideration with a newly developed and evaluated calibration method. The sensor device was integrated in a PEM fuel cell stack to prove the functionality of the whole measurement system. A software application was implemented to visualize and save the measurement values. Its functionality was verified by operational measurements within a PEMFC system. Measurement accuracy and possible negative reactions of the sensor device during PEMFC operation are discussed in detail in this paper. The developed system enables operational measurements for different operating phases of PEM fuel cells. Additionally, this can be seen as a basis for new opportunities of optimization for fuel cell design and operation modes. (author)

  3. Current Density Distribution Mapping in PEM Fuel Cells as An Instrument for Operational Measurements

    Directory of Open Access Journals (Sweden)

    Martin Geske

    2010-04-01

    Full Text Available A newly developed measurement system for current density distribution mapping has enabled a new approach for operational measurements in proton exchange membrane fuel cells (PEMFC. Taking into account previously constructed measurement systems, a method based on a multi layer printed circuit board was chosen for the development of the new system. This type of system consists of a sensor, a special electronic device and the control and visualization PC. For the acquisition of the current density distribution values, a sensor device was designed and installed within a multilayer printed circuit board with integrated shunt resistors. Varying shunt values can be taken into consideration with a newly developed and evaluated calibration method. The sensor device was integrated in a PEM fuel cell stack to prove the functionality of the whole measurement system. A software application was implemented to visualize and save the measurement values. Its functionality was verified by operational measurements within a PEMFC system. Measurement accuracy and possible negative reactions of the sensor device during PEMFC operation are discussed in detail in this paper. The developed system enables operational measurements for different operating phases of PEM fuel cells. Additionally, this can be seen as a basis for new opportunities of optimization for fuel cell design and operation modes.

  4. 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)

  5. 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.

  6. Labeling of Pest Insects Using Radioisotopes to Study Dispersal Pattern, Migration and Estimation of Population Density

    International Nuclear Information System (INIS)

    Singgih Sutrisno

    2008-01-01

    To study insects behaviour in their habitat such as dispersal, migration and flight range, insects are needed to be labelled to trace their movement. One of the most promising labeling methodology for internal labeling is the use of radioisotopes. Radioisotopes that have been used for labeling insects are 3 H, 32 P, 14 Ca, 45 K, 35 S, 59 Fe, 60 Co, and 14 C. Insect labeling with isotopes has more advantages as compared to dyes due to isotopes used for labeling is bonded to the tissue such as 3 H, 32 P, 14 Ca, K, 131 I. Several consideration have to be taken to determine isotopes that will be used in line with the time consuming for experiments. This have to be carried out due to the phenomenon that several isotopes are toxic to insects such as 45 Ca, 59 Fe, 86 Rb, 110 Ag, 115 Cd, and 131 J. Precautions have to be fulfilled for insect radiolabeling which are save to insects, environment, easy to apply, materials are available and acceptable to the public. Radioisotope 32 P with a correct dose is very convenience to be used in such experiments due to its relatively short half live, which is only 14.3 days. If it is an stable isotope it can be kept for a long time so the sample analyzed can be conducted convenience for long periods of time. Stable elements such as Rb can be changed to be radioisotopes by bombardment of neutrons in a nuclear reactor or accelerator. Then the element that has been activated can be identified using solid scintillation counter, multichannel analyzer or can be detected using autoradiography. (author)

  7. 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)

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

    Science.gov (United States)

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

    2017-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-15

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

  10. Prediction models for density and viscosity of biodiesel and their effects on fuel supply system in CI engines

    Energy Technology Data Exchange (ETDEWEB)

    Tesfa, B.; Mishra, R.; Gu, F. [Computing and Engineering, University of Huddersfield, Queensgate, Huddersfield, HD1 3DH (United Kingdom); Powles, N. [Chemistry and Forensic Science, University of Huddersfield, Queensgate, Huddersfield, HD1 3DH (United Kingdom)

    2010-12-15

    Biodiesel is a promising non-toxic and biodegradable alternative fuel used in the transport sector. Nevertheless, the higher viscosity and density of biodiesel poses some acute problems when it is used it in unmodified engine. Taking this into consideration, this study has been focused towards two objectives. The first objective is to identify the effect of temperature on density and viscosity for a variety of biodiesels and also to develop a correlation between density and viscosity for these biodiesels. The second objective is to investigate and quantify the effects of density and viscosity of the biodiesels and their blends on various components of the engine fuel supply system such as fuel pump, fuel filters and fuel injector. To achieve first objective density and viscosity of rapeseed oil biodiesel, corn oil biodiesel and waste oil biodiesel blends (0B, 5B, 10B, 20B, 50B, 75B, and 100B) were tested at different temperatures using EN ISO 3675:1998 and EN ISO 3104:1996 standards. For both density and viscosity new correlations were developed and compared with published literature. A new correlation between biodiesel density and biodiesel viscosity was also developed. The second objective was achieved by using analytical models showing the effects of density and viscosity on the performance of fuel supply system. These effects were quantified over a wide range of engine operating conditions. It can be seen that the higher density and viscosity of biodiesel have a significant impact on the performance of fuel pumps and fuel filters as well as on air-fuel mixing behaviour of compression ignition (CI) engine. (author)

  11. 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.

  12. Structure and Stability of Molecular Crystals with Many-Body Dispersion-Inclusive Density Functional Tight Binding.

    Science.gov (United States)

    Mortazavi, Majid; Brandenburg, Jan Gerit; Maurer, Reinhard J; Tkatchenko, Alexandre

    2018-01-18

    Accurate prediction of structure and stability of molecular crystals is crucial in materials science and requires reliable modeling of long-range dispersion interactions. Semiempirical electronic structure methods are computationally more efficient than their ab initio counterparts, allowing structure sampling with significant speedups. We combine the Tkatchenko-Scheffler van der Waals method (TS) and the many-body dispersion method (MBD) with third-order density functional tight-binding (DFTB3) via a charge population-based method. We find an overall good performance for the X23 benchmark database of molecular crystals, despite an underestimation of crystal volume that can be traced to the DFTB parametrization. We achieve accurate lattice energy predictions with DFT+MBD energetics on top of vdW-inclusive DFTB3 structures, resulting in a speedup of up to 3000 times compared with a full DFT treatment. This suggests that vdW-inclusive DFTB3 can serve as a viable structural prescreening tool in crystal structure prediction.

  13. Density fluctuation in a screened Coulombic colloid dispersion: comparison of the liquid and cubic phases of lipid A-diphosphate

    International Nuclear Information System (INIS)

    Brown, Helen; Ross, D. Keith; Paradies, Henrich H.

    2004-01-01

    Light-, small-angle X-ray and neutron scattering measurements of the dynamic structure factor S(Q,t) of strong interacting dispersions of lipid A-diphosphate were recorded and analysed applying existing models of liquid state theory. Lipid A-diphosphate ordering was observed at low volume fractions (phi=2.2x10 -4 ) and at very low ionic strength (I=10 -5 M). Upon increasing the particle number density of lipid A-diphosphate a transformation of the lattices of the colloidal crystals from a BCC lattice (a=36.20 nm) to a FCC lattice (a=57.30 nm) occurred. This strongly suggests a similarity in the preformed liquid structure and the cubic colloidal phase. The fit of both S eff (Q) and the principle peak I p (Q) with the effective particle charge supports of the main conclusions drawn from the SANS experiments and the liquid state theory indicating the presence of long-range order for the dispersions of lipid A-diphosphate

  14. Validation of molecular crystal structures from powder diffraction data with dispersion-corrected density functional theory (DFT-D)

    International Nuclear Information System (INIS)

    Streek, Jacco van de; Neumann, Marcus A.

    2014-01-01

    The accuracy of 215 experimental organic crystal structures from powder diffraction data is validated against a dispersion-corrected density functional theory method. In 2010 we energy-minimized 225 high-quality single-crystal (SX) structures with dispersion-corrected density functional theory (DFT-D) to establish a quantitative benchmark. For the current paper, 215 organic crystal structures determined from X-ray powder diffraction (XRPD) data and published in an IUCr journal were energy-minimized with DFT-D and compared to the SX benchmark. The on average slightly less accurate atomic coordinates of XRPD structures do lead to systematically higher root mean square Cartesian displacement (RMSCD) values upon energy minimization than for SX structures, but the RMSCD value is still a good indicator for the detection of structures that deserve a closer look. The upper RMSCD limit for a correct structure must be increased from 0.25 Å for SX structures to 0.35 Å for XRPD structures; the grey area must be extended from 0.30 to 0.40 Å. Based on the energy minimizations, three structures are re-refined to give more precise atomic coordinates. For six structures our calculations provide the missing positions for the H atoms, for five structures they provide corrected positions for some H atoms. Seven crystal structures showed a minor error for a non-H atom. For five structures the energy minimizations suggest a higher space-group symmetry. For the 225 SX structures, the only deviations observed upon energy minimization were three minor H-atom related issues. Preferred orientation is the most important cause of problems. A preferred-orientation correction is the only correction where the experimental data are modified to fit the model. We conclude that molecular crystal structures determined from powder diffraction data that are published in IUCr journals are of high quality, with less than 4% containing an error in a non-H atom

  15. 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

  16. 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.

  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. Correlation for the estimation of the density of fatty acid esters fuels and its implications. A proposed Biodiesel Cetane Index.

    Science.gov (United States)

    Lapuerta, Magín; Rodríguez-Fernández, José; Armas, Octavio

    2010-09-01

    Biodiesel fuels (methyl or ethyl esters derived from vegetables oils and animal fats) are currently being used as a means to diminish the crude oil dependency and to limit the greenhouse gas emissions of the transportation sector. However, their physical properties are different from traditional fossil fuels, this making uncertain their effect on new, electronically controlled vehicles. Density is one of those properties, and its implications go even further. First, because governments are expected to boost the use of high-biodiesel content blends, but biodiesel fuels are denser than fossil ones. In consequence, their blending proportion is indirectly restricted in order not to exceed the maximum density limit established in fuel quality standards. Second, because an accurate knowledge of biodiesel density permits the estimation of other properties such as the Cetane Number, whose direct measurement is complex and presents low repeatability and low reproducibility. In this study we compile densities of methyl and ethyl esters published in literature, and proposed equations to convert them to 15 degrees C and to predict the biodiesel density based on its chain length and unsaturation degree. Both expressions were validated for a wide range of commercial biodiesel fuels. Using the latter, we define a term called Biodiesel Cetane Index, which predicts with high accuracy the Biodiesel Cetane Number. Finally, simple calculations prove that the introduction of high-biodiesel content blends in the fuel market would force the refineries to reduce the density of their fossil fuels. Copyright 2010 Elsevier Ireland Ltd. All rights reserved.

  19. Correction for dispersion and Coulombic interactions in molecular clusters with density functional derived methods: Application to polycyclic aromatic hydrocarbon clusters

    Science.gov (United States)

    Rapacioli, Mathias; Spiegelman, Fernand; Talbi, Dahbia; Mineva, Tzonka; Goursot, Annick; Heine, Thomas; Seifert, Gotthard

    2009-06-01

    The density functional based tight binding (DFTB) is a semiempirical method derived from the density functional theory (DFT). It inherits therefore its problems in treating van der Waals clusters. A major error comes from dispersion forces, which are poorly described by commonly used DFT functionals, but which can be accounted for by an a posteriori treatment DFT-D. This correction is used for DFTB. The self-consistent charge (SCC) DFTB is built on Mulliken charges which are known to give a poor representation of Coulombic intermolecular potential. We propose to calculate this potential using the class IV/charge model 3 definition of atomic charges. The self-consistent calculation of these charges is introduced in the SCC procedure and corresponding nuclear forces are derived. Benzene dimer is then studied as a benchmark system with this corrected DFTB (c-DFTB-D) method, but also, for comparison, with the DFT-D. Both methods give similar results and are in agreement with references calculations (CCSD(T) and symmetry adapted perturbation theory) calculations. As a first application, pyrene dimer is studied with the c-DFTB-D and DFT-D methods. For coronene clusters, only the c-DFTB-D approach is used, which finds the sandwich configurations to be more stable than the T-shaped ones.

  20. Non-density dependent pollen dispersal of Shorea maxwelliana (Dipterocarpaceae revealed by a Bayesian mating model based on paternity analysis in two synchronized flowering seasons.

    Directory of Open Access Journals (Sweden)

    Shinsuke Masuda

    Full Text Available Pollinator syndrome is one of the most important determinants regulating pollen dispersal in tropical tree species. It has been widely accepted that the reproduction of tropical forest species, especially dipterocarps that rely on insects with weak flight for their pollination, is positively density-dependent. However differences in pollinator syndrome should affect pollen dispersal patterns and, consequently, influence genetic diversity via the mating process. We examined the pollen dispersal pattern and mating system of Shorea maxwelliana, the flowers of which are larger than those of Shorea species belonging to section Mutica which are thought to be pollinated by thrips (weak flyers. A Bayesian mating model based on the paternity of seeds collected from mother trees during sporadic and mass flowering events revealed that the estimated pollen dispersal kernel and average pollen dispersal distance were similar for both flowering events. This evidence suggests that the putative pollinators - small beetles and weevils - effectively contribute to pollen dispersal and help to maintain a high outcrossing rate even during sporadic flowering events. However, the reduction in pollen donors during a sporadic event results in a reduction in effective pollen donors, which should lead to lower genetic diversity in the next generation derived from seeds produced during such an event. Although sporadic flowering has been considered less effective for outcrossing in Shorea species that depend on thrips for their pollination, effective pollen dispersal by the small beetles and weevils ensures outcrossing during periods of low flowering tree density, as occurs in a sporadic flowering event.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Effect of coating density on oxidation resistance and Cr vaporization from solid oxide fuel cell interconnects

    Science.gov (United States)

    Talic, Belma; Falk-Windisch, Hannes; Venkatachalam, Vinothini; Hendriksen, Peter Vang; Wiik, Kjell; Lein, Hilde Lea

    2017-06-01

    Manganese cobalt spinel oxides are promising materials for protective coatings for solid oxide fuel cell (SOFC) interconnects. To achieve high density such coatings are often sintered in a two-step procedure, involving heat treatment first in reducing and then in oxidizing atmospheres. Sintering the coating inside the SOFC stack during heating would reduce production costs, but may result in a lower coating density. The importance of coating density is here assessed by characterization of the oxidation kinetics and Cr evaporation of Crofer 22 APU with MnCo1.7Fe0.3O4 spinel coatings of different density. The coating density is shown to have minor influence on the long-term oxidation behavior in air at 800 °C, evaluated over 5000 h. Sintering the spinel coating in air at 900 °C, equivalent to an in-situ heat treatment, leads to an 88% reduction of the Cr evaporation rate of Crofer 22 APU in air-3% H2O at 800 °C. The air sintered spinel coating is initially highly porous, however, densifies with time in interaction with the alloy. A two-step reduction and re-oxidation heat treatment results in a denser coating, which reduces Cr evaporation by 97%.

  7. Estimation of electron temperature and density by de convolving the absorption part of the plasma dispersion function

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez D, H.; Cabral P, A.; Melendez L, L.; Lopez C, R.; Colunga S, S.; Valencia A, R.; Cruz J, S.; Gaytan G, E.; Chavez A, E

    1992-04-15

    In this work a method to estimate the temperature and density of the electron (T{sub e}, n{sub e}), based on the deconvolution of the part of absorption of the dispersion function of the plasma is suggested. The absorptive part of this function, is proportional to the convolution of a Gauss distribution with a Lorentz function. The Gaussian represents to the Maxwell function of velocities distribution of the electrons of the plasma. The Lorentzian represents to the form of it lines of an linearized electrostatic wave that spreads with reduction in the plasma. The complex variable z of the plasma dispersion function is written as: z = u + ia, where u = 2 (w-w{sub 0}) {radical} Ln 2 /{gamma}{sub G} is the dimensionless frequency variable, a = {gamma}{sub L} {radical} Ln 2 /{gamma}{sub G} is the Posener parameter, {gamma}{sub G} = k {gamma}{sup '}{sub G} where k is the wave number of the oscillatory phenomenon, {gamma}{sup '}{sub G} is the FWHM of the Gaussian and {gamma}{sub L} = 2 {alpha}, {alpha} being the damping constant; i.e the imaginary part of the frequency {omega}. In this method, it will be assumed that a wave of frequency , and of amplitude small enough to avoid non-linear effects, propagates in the plasma and decays in such a way {alpha} is the Landau damping. With this assumption, the method is only valid in the interval k < < k{sub D}, where k{sub D} is the Debye wave number. Deconvolution of the detected absorption frequency spectrum of the signal, gives the values of {gamma}{sub G} and {gamma}{sub L} from which the values of n{sub e} and T{sub e} can be deduced. (Author)

  8. Estimation of electron temperature and density by de convolving the absorption part of the plasma dispersion function

    International Nuclear Information System (INIS)

    Jimenez D, H.; Cabral P, A.; Melendez L, L.; Lopez C, R.; Colunga S, S.; Valencia A, R.; Cruz J, S.; Gaytan G, E.; Chavez A, E.

    1992-04-01

    In this work a method to estimate the temperature and density of the electron (T e , n e ), based on the deconvolution of the part of absorption of the dispersion function of the plasma is suggested. The absorptive part of this function, is proportional to the convolution of a Gauss distribution with a Lorentz function. The Gaussian represents to the Maxwell function of velocities distribution of the electrons of the plasma. The Lorentzian represents to the form of it lines of an linearized electrostatic wave that spreads with reduction in the plasma. The complex variable z of the plasma dispersion function is written as: z = u + ia, where u = 2 (w-w 0 ) √ Ln 2 /Γ G is the dimensionless frequency variable, a = Γ L √ Ln 2 /Γ G is the Posener parameter, Γ G = k Γ ' G where k is the wave number of the oscillatory phenomenon, Γ ' G is the FWHM of the Gaussian and Γ L = 2 α, α being the damping constant; i.e the imaginary part of the frequency ω. In this method, it will be assumed that a wave of frequency , and of amplitude small enough to avoid non-linear effects, propagates in the plasma and decays in such a way α is the Landau damping. With this assumption, the method is only valid in the interval k D , where k D is the Debye wave number. Deconvolution of the detected absorption frequency spectrum of the signal, gives the values of Γ G and Γ L from which the values of n e and T e can be deduced. (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. 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)

  11. 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)

  12. 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.

  13. 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

  14. Distribution of the Current Density in Electrolyte of the Pem Fuel Cell

    Directory of Open Access Journals (Sweden)

    Eugeniusz Kurgan

    2004-01-01

    Full Text Available In this paper water management in proton exchange membrane (PEM fuel cell is considered. Firt mass convervation law for water is applied. Next proton transport is described by the Nernst-Planck equation and liqid water convection velocity is eliminated by the Schlogl equation. Electro-osmotic drag coefficient is related to hydrogen index and experimentally determined swelling coefficient. Three partial differential equations for molar water concentration Cw, electric potential ϕ and water pressure Pw are formulated. Current density vector i is derived from proton flux expression. These equations together with adequate boundary conditions were solved using finite element method. The distribution of electric potential and current density in function of geometrical parametres is investigated. At the end some illustrative example is given.

  15. Assembly of high-areal-density deuterium-tritium fuel from indirectly driven cryogenic implosions.

    Science.gov (United States)

    Mackinnon, A J; Kline, J L; Dixit, S N; Glenzer, S H; Edwards, M J; Callahan, D A; Meezan, N B; Haan, S W; Kilkenny, J D; Döppner, T; Farley, D R; Moody, J D; Ralph, J E; MacGowan, B J; Landen, O L; Robey, H F; Boehly, T R; Celliers, P M; Eggert, J H; Krauter, K; Frieders, G; Ross, G F; Hicks, D G; Olson, R E; Weber, S V; Spears, B K; Salmonsen, J D; Michel, P; Divol, L; Hammel, B; Thomas, C A; Clark, D S; Jones, O S; Springer, P T; Cerjan, C J; Collins, G W; Glebov, V Y; Knauer, J P; Sangster, C; Stoeckl, C; McKenty, P; McNaney, J M; Leeper, R J; Ruiz, C L; Cooper, G W; Nelson, A G; Chandler, G G A; Hahn, K D; Moran, M J; Schneider, M B; Palmer, N E; Bionta, R M; Hartouni, E P; LePape, S; Patel, P K; Izumi, N; Tommasini, R; Bond, E J; Caggiano, J A; Hatarik, R; Grim, G P; Merrill, F E; Fittinghoff, D N; Guler, N; Drury, O; Wilson, D C; Herrmann, H W; Stoeffl, W; Casey, D T; Johnson, M G; Frenje, J A; Petrasso, R D; Zylestra, A; Rinderknecht, H; Kalantar, D H; Dzenitis, J M; Di Nicola, P; Eder, D C; Courdin, W H; Gururangan, G; Burkhart, S C; Friedrich, S; Blueuel, D L; Bernstein, L A; Eckart, M J; Munro, D H; Hatchett, S P; Macphee, A G; Edgell, D H; Bradley, D K; Bell, P M; Glenn, S M; Simanovskaia, N; Barrios, M A; Benedetti, R; Kyrala, G A; Town, R P J; Dewald, E L; Milovich, J L; Widmann, K; Moore, A S; LaCaille, G; Regan, S P; Suter, L J; Felker, B; Ashabranner, R C; Jackson, M C; Prasad, R; Richardson, M J; Kohut, T R; Datte, P S; Krauter, G W; Klingman, J J; Burr, R F; Land, T A; Hermann, M R; Latray, D A; Saunders, R L; Weaver, S; Cohen, S J; Berzins, L; Brass, S G; Palma, E S; Lowe-Webb, R R; McHalle, G N; Arnold, P A; Lagin, L J; Marshall, C D; Brunton, G K; Mathisen, D G; Wood, R D; Cox, J R; Ehrlich, R B; Knittel, K M; Bowers, M W; Zacharias, R A; Young, B K; Holder, J P; Kimbrough, J R; Ma, T; La Fortune, K N; Widmayer, C C; Shaw, M J; Erbert, G V; Jancaitis, K S; DiNicola, J M; Orth, C; Heestand, G; Kirkwood, R; Haynam, C; Wegner, P J; Whitman, P K; Hamza, A; Dzenitis, E G; Wallace, R J; Bhandarkar, S D; Parham, T G; Dylla-Spears, R; Mapoles, E R; Kozioziemski, B J; Sater, J D; Walters, C F; Haid, B J; Fair, J; Nikroo, A; Giraldez, E; Moreno, K; Vanwonterghem, B; Kauffman, R L; Batha, S; Larson, D W; Fortner, R J; Schneider, D H; Lindl, J D; Patterson, R W; Atherton, L J; Moses, E I

    2012-05-25

    The National Ignition Facility has been used to compress deuterium-tritium to an average areal density of ~1.0±0.1 g cm(-2), which is 67% of the ignition requirement. These conditions were obtained using 192 laser beams with total energy of 1-1.6 MJ and peak power up to 420 TW to create a hohlraum drive with a shaped power profile, peaking at a soft x-ray radiation temperature of 275-300 eV. This pulse delivered a series of shocks that compressed a capsule containing cryogenic deuterium-tritium to a radius of 25-35 μm. Neutron images of the implosion were used to estimate a fuel density of 500-800 g cm(-3).

  16. Gas-to-liquids synthetic fuels for use in fuel cells : reformability, energy density, and infrastructure compatibility.

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, S.; Kopasz, J. P.; Russell, B. J.; Tomlinson, H. L.

    1999-09-08

    The fuel cell has many potential applications, from power sources for electric hybrid vehicles to small power plants for commercial buildings. The choice of fuel will be critical to the pace of its commercialization. This paper reviews the various liquid fuels being considered as an alternative to direct hydrogen gas for the fuel cell application, presents calculations of the hydrogen and carbon dioxide yields from autothermal reforming of candidate liquid fuels, and reports the product gas composition measured from the autothermal reforming of a synthetic fuel in a micro-reactor. The hydrogen yield for a synthetic paraffin fuel produced by a cobalt-based Fischer-Tropsch process was found to be similar to that of retail gasoline. The advantages of the synthetic fuel are that it contains no contaminants that would poison the fuel cell catalyst, is relatively benign to the environment, and could be transported in the existing fuel distribution system.

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. 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.

  2. High-energy-density hydrogen-halogen fuel cells for advanced military applications

    International Nuclear Information System (INIS)

    Balko, E.N.; McElroy, J.F.

    1981-01-01

    It is pointed out that hydrogen-halogen fuel cell systems are particularly suited for an employment as ground power sources for military applications. The large cell potential and reversible characteristics of the H 2 Cl 2 and H 2 Br 2 couples permit high energy storage density and efficient energy conversion. When used as flow batteries, the fluid nature of the reactants in the hydrogen-halogen systems has several advantages over power sources which involve solid phases. Very deep discharge is possible without degradation of subsequent performance, and energy storage capacity is limited only by the external reactant storage volume. Very rapid chemical recharging is possible through replenishment of the reactant supply. A number of H 2 Cl 2 and H 2 Br 2 fuel cell systems have been studied. These systems use the same solid polymer electrolyte (SPE) cell technology originally developed for H2/O2 fuel cells. The results of the investigation are illustrated with the aid of a number of graphs

  3. Exceptional power density and stability at intermediate temperatures in protonic ceramic fuel cells

    Science.gov (United States)

    Choi, Sihyuk; Kucharczyk, Chris J.; Liang, Yangang; Zhang, Xiaohang; Takeuchi, Ichiro; Ji, Ho-Il; Haile, Sossina M.

    2018-03-01

    Over the past several years, important strides have been made in demonstrating protonic ceramic fuel cells (PCFCs). Such fuel cells offer the potential of environmentally sustainable and cost-effective electric power generation. However, their power outputs have lagged behind predictions based on their high electrolyte conductivities. Here we overcome PCFC performance and stability challenges by employing a high-activity cathode, PrBa0.5Sr0.5Co1.5Fe0.5O5+δ (PBSCF), in combination with a chemically stable electrolyte, BaZr0.4Ce0.4Y0.1Yb0.1O3 (BZCYYb4411). We deposit a thin dense interlayer film of the cathode material onto the electrolyte surface to mitigate contact resistance, an approach which is made possible by the proton permeability of PBSCF. The peak power densities of the resulting fuel cells exceed 500 mW cm-2 at 500 °C, while also offering exceptional, long-term stability under CO2.

  4. Effect of Side Chains on Molecular Conformation of Anthracene-Ethynylene-Phenylene-Vinylene Oligomers: A Comparative Density Functional Study With and Without Dispersion Interaction.

    Science.gov (United States)

    Dong, Chuanding; Hoppe, Harald; Beenken, Wichard J D

    2016-06-02

    Using density functional calculations with and without dispersion interaction, we studied the effects of linear octyl and branched 2-ethylhexyl side chains on the oligomer conformation of the conjugated copolymer poly(p-anthracene-ethynylene)-alt-poly(p-phenylene-vinylene). With dispersion included, the branched side chains can cause significant bending of the oligomer backbone, while without dispersion they induce mainly torsional disorder. The oligomers with mainly linear side chains keep good planarity when optimized with and without dispersion. Despite their dramatically different conformations, the calculated absorption spectra of the oligomers with various side chain combinations are very similar, indicating that the conformation of the copolymer is not the main reason for the experimentally observed different spectra of ordered and disordered phases.

  5. Improving the AGR fuel testing power density profile versus irradiation-time in the advanced test reactor

    International Nuclear Information System (INIS)

    Chang, Gray S.; Lillo, Misti A.; Maki, John T.; Petti, David A.

    2009-01-01

    The Very High Temperature gas-cooled Reactor (VHTR), which is currently being developed, achieves simplification of safety through reliance on ceramic-coated fuel particles. Each TRISO-coated fuel particle has its own containment which serves as the principal barrier against radionuclide release under normal operating and accident conditions. These fuel particles, in the form of graphite fuel compacts, are currently undergoing a series of irradiation tests in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) to support the Advanced Gas-Cooled Reactor (AGR) fuel qualification program. A representive coated fuel particle with an 235 U enrichment of 19.8 wt% was used in this analysis. The fuel burnup analysis tool used to perform the neutronics study reported herein, couples the Monte Carlo transport code MCNP, with the radioactive decay and burnup code ORIGEN2. The fuel burnup methodology known as Monte-Carlo with ORIGEN2 (MCWO) was used to evaluate the AGR experiment assembly and demonstrate compliance with ATR safety requirements. For the AGR graphite fuel compacts, the MCWO-calculated fission power density (FPD) due to neutron fission in 235 U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250degC throughout the proposed irradiation campaign of 550 effective full power days (EFPDs). Based on the MCWO-calculated FPD, a fixed gas gap size was designed to allow regulation of the fuel compact temperatures throughout the entire fuel irradiation campaign by filling the gap with a mixture of helium and neon gases. The chosen fixed gas gap can only regulate the peak fuel compact temperature in the desired range during the irradiation test if the ratio of the peak power density to the time-dependent low power density (P/T) at 550 EFPDs is less than 2.5. However, given the near constant neutron flux within the ATR driver core and the depletion of 235 U

  6. 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

  7. 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)

  8. Thermal-Hydraulic Performance of Cross-Shaped Spiral Fuel in High-Power-Density BWRs

    International Nuclear Information System (INIS)

    Conboy, Thomas; Hejzlar, Pavel

    2006-01-01

    Power up-rating of existing nuclear reactors promises to be an area of great study for years to come. One of the major approaches to efficiently increasing power density is by way of advanced fuel design, and cross-shaped spiral-fuel has shown such potential in previous studies. Our work aims to model the thermal-hydraulic consequences of filling a BWR core with these spiral-shaped pins. The helically-wound pins have a cross-section resembling a 4-petaled flower. They fill an assembly in a tight bundle, their dimensions chosen carefully such that the petals of neighboring pins contact each other at their outer-most extent in a self-supporting lattice, absent of grid spacers. Potential advantages of this design raise much optimism from a thermal-hydraulic perspective. These spiral rods possess about 40% larger surface area than traditional rods, resulting in increased cooling and a proportional reduction in average surface heat flux. The thin petal-like extensions help by lowering thermal resistance between the hot central region of the pin and the bulk coolant flow, decreasing the maximum fuel temperature by 200 deg. C according to Finite Element (COSMOS) models. However, COSMOS models also predict a potential problem area at the 'elbow' region of two adjoining petals, where heat flux peaking is twice that along the extensions. Preliminary VIPRE models, which account only for the surface area increase, predict a 22% increase in critical power. It is also anticipated that the spiral twist would provide the flowing coolant with an additional radial velocity component, and likely promote turbulence and mixing within an assembly. These factors are expected to provide further margin for increased power density, and are currently being incorporated into the VIPRE model. The reduction in pressure drop inherent in any core without grid-spacers is also expected to be significant in aiding core stability, though this has not yet been quantified. Spiral-fuel seems to be a

  9. 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

  10. Safety characteristics of mid-sized MOX fueled liquid metal reactor core of high converter type in the initiating phase of unprotected loss of flow accident. Effect of low specific fuel power density on ULOF behavior brought by employment of large diameter fuel pins

    International Nuclear Information System (INIS)

    Ishida, Masayoshi; Kawada, Kenichi; Niwa, Hajime

    2003-07-01

    Safety characteristics in core disruptive accidents (CDAs) of mid-sized MOX fueled liquid metal reactor core of high converter type have been examined by using the CDA initiating phase analysis code SAS4A. The design concept of high converter type reactor core has been studied as one of options in the category of sodium-cooled reactor in Phase II of Feasibility Study on Commercialized Fast Reactor Cycle System. An unprotected loss-of-flow accident (ULOF) has been selected as a representative CDA initiator for this study. A core concept of high converter type, which employed a large diameter fuel pin of 11.1 mm with 1.2 m core height to get a large fuel volume fraction in the core to achieve high internal conversion ratio was proposed in JFY2001. Each fuel subassembly of the core (abbreviated here as UPL120)was provided with an upper sodium plenum directly above the core to reduce the sodium void reactivity worth. Because of the large fuel pin diameter, average specific fuel power density (31 kW/kg-MOX) of UPL120 is about one half of those of conventional large MOX cores. The reactivity worth of sodium voiding is 6$ in the whole core, and -1$ in the all upper plenums. Initiating phase of ULOF accident in UPL120 under the conditions of nominal design and best estimate analysis resulted in a slightly super-prompt critical power burst. The causes of the super-prompt criticality have been identified twofold: (a) the low specific fuel power density of core reduced the effectiveness of prompt negative reactivity feedback of Doppler and axial fuel expansion effects upon increase in reactor power, and (b) the longer core height compared with conventional 1m cores brought, together with the lower specific power density, a remarkable delay in insertion of negative fuel dispersion reactivity after the onset of fuel disruption in sodium voided subassembly due to the lower linear heat rating in the top portion of the core. During the delay, burst-type fuel failures in sodium un

  11. 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.

  12. 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)

  13. Validation of molecular crystal structures from powder diffraction data with dispersion-corrected density functional theory (DFT-D).

    Science.gov (United States)

    van de Streek, Jacco; Neumann, Marcus A

    2014-12-01

    In 2010 we energy-minimized 225 high-quality single-crystal (SX) structures with dispersion-corrected density functional theory (DFT-D) to establish a quantitative benchmark. For the current paper, 215 organic crystal structures determined from X-ray powder diffraction (XRPD) data and published in an IUCr journal were energy-minimized with DFT-D and compared to the SX benchmark. The on average slightly less accurate atomic coordinates of XRPD structures do lead to systematically higher root mean square Cartesian displacement (RMSCD) values upon energy minimization than for SX structures, but the RMSCD value is still a good indicator for the detection of structures that deserve a closer look. The upper RMSCD limit for a correct structure must be increased from 0.25 Å for SX structures to 0.35 Å for XRPD structures; the grey area must be extended from 0.30 to 0.40 Å. Based on the energy minimizations, three structures are re-refined to give more precise atomic coordinates. For six structures our calculations provide the missing positions for the H atoms, for five structures they provide corrected positions for some H atoms. Seven crystal structures showed a minor error for a non-H atom. For five structures the energy minimizations suggest a higher space-group symmetry. For the 225 SX structures, the only deviations observed upon energy minimization were three minor H-atom related issues. Preferred orientation is the most important cause of problems. A preferred-orientation correction is the only correction where the experimental data are modified to fit the model. We conclude that molecular crystal structures determined from powder diffraction data that are published in IUCr journals are of high quality, with less than 4% containing an error in a non-H atom.

  14. 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.

  15. Durability of Low Platinum Fuel Cells Operating at High Power Density

    Energy Technology Data Exchange (ETDEWEB)

    Polevaya, Olga [Nuvera Fuel Cells Inc.; Blanchet, Scott [Nuvera Fuel Cells Inc.; Ahluwalia, Rajesh [Argonne National Lab; Borup, Rod [Los-Alamos National Lab; Mukundan, Rangachary [Los-Alamos National Lab

    2014-03-19

    Understanding and improving the durability of cost-competitive fuel cell stacks is imperative to successful deployment of the technology. Stacks will need to operate well beyond today’s state-of-the-art rated power density with very low platinum loading in order to achieve the cost targets set forth by DOE ($15/kW) and ultimately be competitive with incumbent technologies. An accelerated cost-reduction path presented by Nuvera focused on substantially increasing power density to address non-PGM material costs as well as platinum. The study developed a practical understanding of the degradation mechanisms impacting durability of fuel cells with low platinum loading (≤0.2mg/cm2) operating at high power density (≥1.0W/cm2) and worked out approaches for improving the durability of low-loaded, high-power stack designs. Of specific interest is the impact of combining low platinum loading with high power density operation, as this offers the best chance of achieving long-term cost targets. A design-of-experiments approach was utilized to reveal and quantify the sensitivity of durability-critical material properties to high current density at two levels of platinum loading (the more conventional 0.45 mgPt.cm–1 and the much lower 0.2 mgPt.cm–2) across several cell architectures. We studied the relevance of selected component accelerated stress tests (AST) to fuel cell operation in power producing mode. New stress tests (NST) were designed to investigate the sensitivity to the addition of electrical current on the ASTs, along with combined humidity and load cycles and, eventually, relate to the combined city/highway drive cycle. Changes in the cathode electrochemical surface area (ECSA) and average oxygen partial pressure on the catalyst layer with aging under AST and NST protocols were compared based on the number of completed cycles. Studies showed elevated sensitivity of Pt growth to the potential limits and the initial particle size distribution. The ECSA loss

  16. 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

  17. 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.

  18. The influence of changes in the VVER-1000 fuel assembly shape during operation on the power density distribution

    Energy Technology Data Exchange (ETDEWEB)

    Shishkov, L. K., E-mail: Shishkov-LK@nrcki.ru; Gorodkov, S. S.; Mikailov, E. F.; Sukhino-Homenko, E. A.; Sumarokova, A. S., E-mail: Sumarokova-AS@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    2016-12-15

    A new approach to calculation of the coefficients of sensitivity of the fuel pin power to deviations in gap sizes between fuel assemblies of the VVER-1000 reactor during its operation is proposed. It is shown that the calculations by the MCU code should be performed for a full-size model of the core to take the interference of the gap influence into account. In order to reduce the conservatism of calculations, the coolant density and coolant temperature feedbacks should be taken into account, as well as the fuel burnup.

  19. A mathematical model of the maximum power density attainable in an alkaline hydrogen/oxygen fuel cell

    Science.gov (United States)

    Kimble, Michael C.; White, Ralph E.

    1991-01-01

    A mathematical model of a hydrogen/oxygen alkaline fuel cell is presented that can be used to predict the polarization behavior under various power loads. The major limitations to achieving high power densities are indicated and methods to increase the maximum attainable power density are suggested. The alkaline fuel cell model describes the phenomena occurring in the solid, liquid, and gaseous phases of the anode, separator, and cathode regions based on porous electrode theory applied to three phases. Fundamental equations of chemical engineering that describe conservation of mass and charge, species transport, and kinetic phenomena are used to develop the model by treating all phases as a homogeneous continuum.

  20. 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)

  1. 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

  2. Evaluation of the effect of reactant gases mass flow rates on power density in a polymer electrolyte membrane fuel cell

    Science.gov (United States)

    Kahveci, E. E.; Taymaz, I.

    2018-03-01

    In this study it was experimentally investigated the effect of mass flow rates of reactant gases which is one of the most important operational parameters of polymer electrolyte membrane (PEM) fuel cell on power density. The channel type is serpentine and single PEM fuel cell has an active area of 25 cm2. Design-Expert 8.0 (trial version) was used with four variables to investigate the effect of variables on the response using. Cell temperature, hydrogen mass flow rate, oxygen mass flow rate and humidification temperature were selected as independent variables. In addition, the power density was used as response to determine the combined effects of these variables. It was kept constant cell and humidification temperatures while changing mass flow rates of reactant gases. From the results an increase occurred in power density with increasing the hydrogen flow rates. But oxygen flow rate does not have a significant effect on power density within determined mass flow rates.

  3. Application of London-type dispersion corrections to the solid-state density functional theory simulation of the terahertz spectra of crystalline pharmaceuticals.

    Science.gov (United States)

    King, Matthew D; Buchanan, William D; Korter, Timothy M

    2011-03-14

    The effects of applying an empirical dispersion correction to solid-state density functional theory methods were evaluated in the simulation of the crystal structure and low-frequency (10 to 90 cm(-1)) terahertz spectrum of the non-steroidal anti-inflammatory drug, naproxen. The naproxen molecular crystal is bound largely by weak London force interactions, as well as by more prominent interactions such as hydrogen bonding, and thus serves as a good model for the assessment of the pair-wise dispersion correction term in systems influenced by intermolecular interactions of various strengths. Modifications to the dispersion parameters were tested in both fully optimized unit cell dimensions and those determined by X-ray crystallography, with subsequent simulations of the THz spectrum being performed. Use of the unmodified PBE density functional leads to an unrealistic expansion of the unit cell volume and the poor representation of the THz spectrum. Inclusion of a modified dispersion correction enabled a high-quality simulation of the THz spectrum and crystal structure of naproxen to be achieved without the need for artificially constraining the unit cell dimensions.

  4. Cure characteristics, crosslink density and degree of filler dispersion of kaolin-filled natural rubber compounds in the presence of alkanolamide

    Science.gov (United States)

    Surya, I.; Hayeemasae, N.; Ginting, M.

    2018-03-01

    The effects of alkanolamide (ALK) addition on cure characteristics, crosslink density and degree of filler dispersion of kaolin-filled natural rubber (NR) compounds were investigated. The kaolin filler was incorporated into NR compounds with a fixed loading, 30.0 phr. The ALK was prepared from Refined Bleached Deodorized Palm Stearin (RBDPS), a waste product of cooking oil production, and diethanolamine. The ALK is an oily material and added into the filled NR compounds as a rubber additive at different loadings, 0.0, 3.0, 5.0 and 7.0. The kaolin-filled NR compounds with and without ALK were vulcanized using a semi-efficient vulcanization system. It was found that ALK decreased the scorch and cure times and improved filler dispersion of the kaolin-filled NR compounds. The higher the ALK loading, the shorter were the scorch and cure times. It was also found that ALK increased the crosslink density of kaolin-filled NR compound up to 5.0 phr of loading. Due to its oily properties, The ALK acted as an internal plasticizer which decreased the minimum torque and improved the degree of kaolin dispersion in NR phases. The higher the ALK loading; the lower the minimum torque and better the filler dispersion.

  5. Relationships between population density, fine-scale genetic structure, mating system and pollen dispersal in a timber tree from African rainforests.

    Science.gov (United States)

    Duminil, J; Daïnou, K; Kaviriri, D K; Gillet, P; Loo, J; Doucet, J-L; Hardy, O J

    2016-03-01

    Owing to the reduction of population density and/or the environmental changes it induces, selective logging could affect the demography, reproductive biology and evolutionary potential of forest trees. This is particularly relevant in tropical forests where natural population densities can be low and isolated trees may be subject to outcross pollen limitation and/or produce low-quality selfed seeds that exhibit inbreeding depression. Comparing reproductive biology processes and genetic diversity of populations at different densities can provide indirect evidence of the potential impacts of logging. Here, we analysed patterns of genetic diversity, mating system and gene flow in three Central African populations of the self-compatible legume timber species Erythrophleum suaveolens with contrasting densities (0.11, 0.68 and 1.72 adults per ha). The comparison of inbreeding levels among cohorts suggests that selfing is detrimental as inbred individuals are eliminated between seedling and adult stages. Levels of genetic diversity, selfing rates (∼16%) and patterns of spatial genetic structure (Sp ∼0.006) were similar in all three populations. However, the extent of gene dispersal differed markedly among populations: the average distance of pollen dispersal increased with decreasing density (from 200 m in the high-density population to 1000 m in the low-density one). Overall, our results suggest that the reproductive biology and genetic diversity of the species are not affected by current logging practices. However, further investigations need to be conducted in low-density populations to evaluate (1) whether pollen limitation may reduce seed production and (2) the regeneration potential of the species.

  6. Relationships between population density, fine-scale genetic structure, mating system and pollen dispersal in a timber tree from African rainforests

    Science.gov (United States)

    Duminil, J; Daïnou, K; Kaviriri, D K; Gillet, P; Loo, J; Doucet, J-L; Hardy, O J

    2016-01-01

    Owing to the reduction of population density and/or the environmental changes it induces, selective logging could affect the demography, reproductive biology and evolutionary potential of forest trees. This is particularly relevant in tropical forests where natural population densities can be low and isolated trees may be subject to outcross pollen limitation and/or produce low-quality selfed seeds that exhibit inbreeding depression. Comparing reproductive biology processes and genetic diversity of populations at different densities can provide indirect evidence of the potential impacts of logging. Here, we analysed patterns of genetic diversity, mating system and gene flow in three Central African populations of the self-compatible legume timber species Erythrophleum suaveolens with contrasting densities (0.11, 0.68 and 1.72 adults per ha). The comparison of inbreeding levels among cohorts suggests that selfing is detrimental as inbred individuals are eliminated between seedling and adult stages. Levels of genetic diversity, selfing rates (∼16%) and patterns of spatial genetic structure (Sp ∼0.006) were similar in all three populations. However, the extent of gene dispersal differed markedly among populations: the average distance of pollen dispersal increased with decreasing density (from 200 m in the high-density population to 1000 m in the low-density one). Overall, our results suggest that the reproductive biology and genetic diversity of the species are not affected by current logging practices. However, further investigations need to be conducted in low-density populations to evaluate (1) whether pollen limitation may reduce seed production and (2) the regeneration potential of the species. PMID:26696137

  7. Recent status and future aspect of plate type fuel element technology with high uranium density at NUKEM

    International Nuclear Information System (INIS)

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

    1983-01-01

    According to the present state of development full size test fuel elements with UAl x , U 3 O 8 , and U 3 Si 2 fuel were fabricated at Nukem in production scale. The maximum uranium densities amount to 1.8 g/cc for UAI x , 2.9 g/cc for U 3 O 8 , and 4.76 g/cc for U 3 Si 2 . The irradiation performance of these fuel elements is good: Up to the end of September 1982 the following burnups were achieved: 73% with UA1 x , 60% with U 3 O 8 , 39% with U 3 Si 2 ; no defects could be detected. For an economical fuel element production with reduced 235-U enrichment chemical uranium recycling methods were developed allowing immediate scrap recovery at minimum waste generation. In addition test plates with UAl x and U 3 O 8 fuel were successfully irradiated in the ORR up to a burnup of 75 %. The relatively high uranium meat densities of these test plates amount to 2.2 g/cc for UAI x , and 3.14 g/cc for U 3 O 8 fuel. Apart from plates with standard geometry also plates with increased meat thickness were inserted. (author)

  8. Post irradiation examination of HANARO nucler mini-element fuel (metallographic and density test)

    International Nuclear Information System (INIS)

    Yoo, Byung Ok; Hong, K. P.; Park, D. G.; Choo, Y. S.; Baik, S. J.; Kim, K. H.; Kim, H. C.; Jung, Y. H.

    2001-05-01

    The post irradiation examination of a HANARO mini-element nuclear fuel, KH96C-004, was done in June 6, 2000. The purpose of this project is to evaluate the in-core performance and reliability of mini-element nuclear fuel for HANARO developed by the project T he Nuclear Fuel Material Development of Research Reactor . And, in order to examine the performance of mini-element nuclear fuel in normal output condition, the post irradiation examination of a nuclear fuel bundle composed by 6 mini nuclear fuel rods and 12 dummy fuel rods was performed. Based on these examination results, the safety and reliability of HANARO fuel and the basic data on the design of HANARO nuclear fuel can be ensured and obtained,

  9. Neutronic studies of the long life core concept: Part 1, Design and performance of 1000 MWe uranium oxide fueled low power density LMR cores

    International Nuclear Information System (INIS)

    Orechwa, Y.

    1987-04-01

    The parametric behavior of some key neutronic performance parameters for low power density LMR cores fueled with uranium oxide is investigated. The results are compared to reference homogeneous and heterogeneous cores with normal fuel management and Pu fueling. It can be concluded that with respect to minimizing the initial fissile mass and thereby economizing on the inventory costs and carrying charges, the superior neutron economy of the LMR fuel cycle is best exploited through normal fuel management with Pu recycling. In the once-through mode the LMR fuel cycle has disadvantages due to a higher fissile inventory and is not competitive with the LWR fuel cycle

  10. Density functional theory calculations of H/D isotope effects on polymer electrolyte membrane fuel cell operations

    Energy Technology Data Exchange (ETDEWEB)

    Yanase, Satoshi; Oi, Takao [Sophia Univ., Tokyo (Japan). Faculty of Science and Technology

    2015-10-01

    To elucidate hydrogen isotope effects observed between fuel and exhaust hydrogen gases during polymer electrolyte membrane fuel cell operations, H-to-D reduced partition function ratios (RPFRs) for the hydrogen species in the Pt catalyst phase of the anode and the electrolyte membrane phase of the fuel cell were evaluated by density functional theory calculations on model species of the two phases. The evaluation yielded 3.2365 as the value of the equilibrium constant of the hydrogen isotope exchange reaction between the two phases at 39 C, which was close to the experimentally estimated value of 3.46-3.99 at the same temperature. It was indicated that H{sup +} ions on the Pt catalyst surface of the anode and H species in the electrolyte membrane phase were isotopically in equilibrium with one another during fuel cell operations.

  11. The relation between pre-eruptive bubble size distribution, ash particle morphology, and their internal density: Implications to volcanic ash transport and dispersion models

    Science.gov (United States)

    Proussevitch, Alexander

    2014-05-01

    Parameterization of volcanic ash transport and dispersion (VATD) models strongly depends on particle morphology and their internal properties. Shape of ash particles affects terminal fall velocities (TFV) and, mostly, dispersion. Internal density combined with particle size has a very strong impact on TFV and ultimately on the rate of ash cloud thinning and particle sedimentation on the ground. Unlike other parameters, internal particle density cannot be measured directly because of the micron scale sizes of fine ash particles, but we demonstrate that it varies greatly depending on the particle size. Small simple type ash particles (fragments of bubble walls, 5-20 micron size) do not contain whole large magmatic bubbles inside and their internal density is almost the same as that of volcanic glass matrix. On the other side, the larger compound type ash particles (>40 microns for silicic fine ashes) always contain some bubbles or the whole spectra of bubble size distribution (BSD), i.e. bubbles of all sizes, bringing their internal density down as compared to simple ash. So, density of the larger ash particles is a function of the void fraction inside them (magmatic bubbles) which, in turn, is controlled by BSD. Volcanic ash is a product of the fragmentation of magmatic foam formed by pre-eruptive bubble population and characterized by BSD. The latter can now be measured from bubble imprints on ash particle surfaces using stereo-scanning electron microscopy (SSEM) and BubbleMaker software developed at UNH, or using traditional high-resolution X-Ray tomography. In this work we present the mathematical and statistical formulation for this problem connecting internal ash density with particle size and BSD, and demonstrate how the TFV of the ash population is affected by variation of particle density.

  12. Molecularly Engineered Azobenzene Derivatives for High Energy Density Solid-State Solar Thermal Fuels.

    Science.gov (United States)

    Cho, Eugene N; Zhitomirsky, David; Han, Grace G D; Liu, Yun; Grossman, Jeffrey C

    2017-03-15

    Solar thermal fuels (STFs) harvest and store solar energy in a closed cycle system through conformational change of molecules and can release the energy in the form of heat on demand. With the aim of developing tunable and optimized STFs for solid-state applications, we designed three azobenzene derivatives functionalized with bulky aromatic groups (phenyl, biphenyl, and tert-butyl phenyl groups). In contrast to pristine azobenzene, which crystallizes and makes nonuniform films, the bulky azobenzene derivatives formed uniform amorphous films that can be charged and discharged with light and heat for many cycles. Thermal stability of the films, a critical metric for thermally triggerable STFs, was greatly increased by the bulky functionalization (up to 180 °C), and we were able to achieve record high energy density of 135 J/g for solid-state STFs, over a 30% improvement compared to previous solid-state reports. Furthermore, the chargeability in the solid state was improved, up to 80% charged from 40% charged in previous solid-state reports. Our results point toward molecular engineering as an effective method to increase energy storage in STFs, improve chargeability, and improve the thermal stability of the thin film.

  13. Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Savander, V. I.; Shumskiy, B. E., E-mail: borisshumskij@yandex.ru [National Research Nuclear University MEPhI (Russian Federation); Pinegin, A. A. [National Research Center Kurchatov Institute (Russian Federation)

    2016-12-15

    The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  15. 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

  16. In depth fusion flame spreading with a deuterium—tritium plane fuel density profile for plasma block ignition

    International Nuclear Information System (INIS)

    Malekynia, B.; Razavipour, S. S.

    2012-01-01

    Solid-state fuel ignition was given by Chu and Bobin according to the hydrodynamic theory at x = 0 qualitatively. A high threshold energy flux density, i.e., E* = 4.3 × 10 12 J/m 2 , has been reached. Recently, fast ignition by employing clean petawatt—picosecond laser pulses was performed. The anomalous phenomena were observed to be based on suppression of prepulses. The accelerated plasma block was used to ignite deuterium—tritium fuel at solid-state density. The detailed analysis of the thermonuclear wave propagation was investigated. Also the fusion conditions at x ≠ 0 layers were clarified by exactly solving hydrodynamic equations for plasma block ignition. In this paper, the applied physical mechanisms are determined for nonlinear force laser driven plasma blocks, thermonuclear reaction, heat transfer, electron—ion equilibration, stopping power of alpha particles, bremsstrahlung, expansion, density dependence, and fluid dynamics. New ignition conditions may be obtained by using temperature equations, including the density profile that is obtained by the continuity equation and expansion velocity. The density is only a function of x and independent of time. The ignition energy flux density, E* t , for the x ≠ 0 layers is 1.95 × 10 12 J/m 2 . Thus threshold ignition energy in comparison with that at x = 0 layers would be reduced to less than 50 percent. (physics of gases, plasmas, and electric discharges)

  17. Carbon deposition thresholds on nickel-based solid oxide fuel cell anodes II. Steam:carbon ratio and current density

    Science.gov (United States)

    Kuhn, J.; Kesler, O.

    2015-03-01

    For the second part of a two part publication, coking thresholds with respect to molar steam:carbon ratio (SC) and current density in nickel-based solid oxide fuel cells were determined. Anode-supported button cell samples were exposed to 2-component and 5-component gas mixtures with 1 ≤ SC ≤ 2 and zero fuel utilization for 10 h, followed by measurement of the resulting carbon mass. The effect of current density was explored by measuring carbon mass under conditions known to be prone to coking while increasing the current density until the cell was carbon-free. The SC coking thresholds were measured to be ∼1.04 and ∼1.18 at 600 and 700 °C, respectively. Current density experiments validated the thresholds measured with respect to fuel utilization and steam:carbon ratio. Coking thresholds at 600 °C could be predicted with thermodynamic equilibrium calculations when the Gibbs free energy of carbon was appropriately modified. Here, the Gibbs free energy of carbon on nickel-based anode support cermets was measured to be -6.91 ± 0.08 kJ mol-1. The results of this two part publication show that thermodynamic equilibrium calculations with appropriate modification to the Gibbs free energy of solid-phase carbon can be used to predict coking thresholds on nickel-based anodes at 600-700 °C.

  18. Structure and stability of acrolein and allyl alcohol networks on Ag(111) from density functional theory based calculations with dispersion corrections

    Science.gov (United States)

    Ferullo, Ricardo M.; Branda, Maria Marta; Illas, Francesc

    2013-11-01

    The interaction of acrolein and allyl alcohol with the Ag(111) surface has been studied by means of periodic density functional theory based calculations including explicitly dispersion terms. Different coverage values have been explored going from isolated adsorbed molecules to isolated dimers, interacting dimers or ordered overlayers. The inclusion of the dispersion terms largely affects the calculated values of the adsorption energy and also the distance between adsorbed molecule and the metallic surface but much less the adsorbate-adsorbate interactions. Owing to the large dipole moment of acrolein, the present calculations predict that at high coverage this molecule forms a stable extensive two-dimensional network on the surface, caused by the alignment of the adsorbate dipoles. For the case of allyl alcohol, dimers and complex networks exhibit similar stability.

  19. 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.

  20. 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.

  1. Inverse link between density and dispersal distance in butterflies: field evidence from six co-occurring species

    Czech Academy of Sciences Publication Activity Database

    Konvička, Martin; Zimmermann, Kamil; Klímová, M.; Hůla, V.; Fric, Zdeněk

    2012-01-01

    Roč. 54, č. 1 (2012), s. 91-101 ISSN 1438-3896 R&D Projects: GA MŠk LC06073; GA ČR GAP505/10/2167 Institutional research plan: CEZ:AV0Z50070508 Keywords : animal movement * dispersal * landscape occupancy Subject RIV: EH - Ecology, Behaviour Impact factor: 1.923, year: 2012 http://www.springerlink.com/content/b1156640p50pw285/fulltext.pdf

  2. 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

  3. Spontaneous oscillations of cell voltage, power density, and anode exit CO concentration in a PEM fuel cell.

    Science.gov (United States)

    Lu, Hui; Rihko-Struckmann, Liisa; Sundmacher, Kai

    2011-10-28

    The spontaneous oscillations of the cell voltage and output power density of a PEMFC (with PtRu/C anode) using CO-containing H(2) streams as anodic fuels have been observed during galvanostatic operating. It is ascribed to the dynamic coupling of the CO adsorption (poisoning) and the electrochemical CO oxidation (reactivating) processes in the anode chamber of the single PEMFC. Accompanying the cell voltage and power density oscillations, the discrete CO concentration oscillations at the anode outlet of the PEMFC were also detected, which directly confirms the electrochemical CO oxidation taking place in the anode chamber during galvanostatic operating. This journal is © the Owner Societies 2011

  4. 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

  5. Interfaces in ceramic nuclear fuels

    International Nuclear Information System (INIS)

    Reeve, K.D.

    Internal interfaces in all-ceramic dispersion fuels (such as these for HTGRs) are discussed for two classes: BeO-based dispersions, and coated particles for graphite-based fuels. The following points are made: (1) The strength of a two-phase dispersion is controlled by the weaker dispersed phase bonded to the matrix. (2) Differential expansion between two phases can be controlled by an intermediate buffer zone of low density. (3) A thin ceramic coating should be in compression. (4) Chemical reaction between coating and substrate and mass transfer in service should be minimized. The problems of the nuclear fuel designer are to develop coatings for fission product retention, and to produce radiation-resistant interfaces. 44 references, 18 figures

  6. Effects of heat and water transport on the performance of polymer electrolyte membrane fuel cell under high current density operation

    International Nuclear Information System (INIS)

    Tabuchi, Yuichiro; Shiomi, Takeshi; Aoki, Osamu; Kubo, Norio; Shinohara, Kazuhiko

    2010-01-01

    Key challenges to the acceptance of polymer electrolyte membrane fuel cells (PEMFCs) for automobiles are the cost reduction and improvement in its power density for compactness. In order to get the solution, the further improvement in a fuel cell performance is required. In particular, under higher current density operation, water and heat transport in PEMFCs has considerable effects on the cell performance. In this study, the impact of heat and water transport on the cell performance under high current density was investigated by experimental evaluation of liquid water distribution and numerical validation. Liquid water distribution in MEA between rib and channel area is evaluated by neutron radiography. In order to neglect the effect of liquid water in gas channels and reactant species concentration distribution in the flow direction, the differential cell was used in this study. Experimental results suggested that liquid water under the channel was dramatically changed with rib/channel width. From the numerical study, it is found that the change of liquid water distribution was significantly affected by temperature distribution in MEA between rib and channel area. In addition, not only heat transport but also water transport through the membrane also significantly affected the cell performance under high current density operation.

  7. Directional Freezing of Nanocellulose Dispersions Aligns the Rod-Like Particles and Produces Low-Density and Robust Particle Networks.

    Science.gov (United States)

    Munier, Pierre; Gordeyeva, Korneliya; Bergström, Lennart; Fall, Andreas B

    2016-05-09

    We show that unidirectional freezing of nanocellulose dispersions produces cellular foams with high alignment of the rod-like nanoparticles in the freezing direction. Quantification of the alignment in the long direction of the tubular pores with X-ray diffraction shows high orientation of cellulose nanofibrils (CNF) and cellulose nanocrystals (CNC) at particle concentrations above 0.2 wt % (CNC) and 0.08 wt % (CNF). Aggregation of CNF by pH decrease or addition of salt significantly reduces the particle orientation; in contrast, exceeding the concentration where particles gel by mobility constraints had a relatively small effect on the orientation. The dense nanocellulose network formed by directional freezing was sufficiently strong to resist melting. The formed hydrogels were birefringent and displayed anisotropic laser diffraction patterns, suggesting preserved nanocellulose alignment and cellular structure. Nondirectional freezing of the hydrogels followed by sublimation generates foams with a pore structure and nanocellulose alignment resembling the structure of the initial directional freezing.

  8. 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)

  9. 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.

  10. 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

  11. 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

  12. Effects of density functionals and dispersion interactions on geometries, bond energies and harmonic frequencies of Etbnd UX3 (E = N, P, CH; X = H, F, Cl)

    Science.gov (United States)

    Pandey, Krishna Kumar; Patidar, Pankaj; Patidar, Sunil Kumar; Vishwakarma, Ravi

    2014-12-01

    Quantum-chemical calculations have been performed to evaluate the geometries, bonding nature and harmonic frequencies of the compounds [Etbnd UX3] at DFT, DFT-D3, DFT-D3(BJ) and DFT-dDSc levels using different density functionals BP86, BLYP, PBE, revPBE, PW91, TPSS and M06-L. The stretching frequency of Utbnd N bond in [Ntbnd UF3] calculated with DFT/BLYP closely resembles with the experimental value. The performance of different density functionals for accurate Utbnd N vibrational frequencies follows the order BLYP > revPBE > BP86 > PW91 > TPSS > PBE > M06-L. The BLYP functional gives accurate value of the Utbnd E bond distances. The uranium atom in the studied compounds [Etbnd UX3] is positively charged. Upon going from [Etbnd UF3] to [Etbnd UCl3], the partial Hirshfeld charge on uranium atom decreases because of the lower electronegativity of chlorine compared to flourine. The Gopinathan-Jug bond order for Utbnd E bonds ranges from 2.90 to 3.29. The Utbnd E bond dissociation energies vary with different density functionals as M06-L < TPSS < BLYP < revPBE < BP86 < PBE ≈ PW91. The orbital interactions ΔEorb, in all studied compounds [Etbnd UX3] are larger than the electrostatic interaction ΔEelstat, which means the Utbnd N bonds in these compound have greater degree of covalent character (in the range 63.8-77.2%). The Usbnd E σ-bonding interaction is the dominant bonding interaction in the nitride and methylidyne complexes while it is weaker in [Ptbnd UX3]. The dispersion energy contributions to the total bond dissociation energies are rather small. Compared to the Grimme's D3(BJ) corrections, the Corminboeuf's dispersion corrections are larger with metaGGA functionals (TPSS, M06-L) while smaller with GGA functionals.

  13. 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

  14. 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

  15. Preparation method of Ni@Pt/C nanocatalyst affects the performance of direct borohydride-hydrogen peroxide fuel cell: Improved power density and increased catalytic oxidation of borohydride.

    Science.gov (United States)

    Hosseini, Mir Ghasem; Mahmoodi, Raana

    2017-08-15

    The Ni@Pt/C electrocatalysts were synthesized using two different methods: with sodium dodecyl sulfate (SDS) and without SDS. The metal loading in synthesized nanocatalysts was 20wt% and the molar ratio of Ni: Pt was 1:1. The structural characterizations of Ni@Pt/C electrocatalysts were investigated by field emission scanning electron microscopy (FE-SEM), energy-dispersive X-ray spectroscopy (EDX), X-ray diffraction (XRD), transmission electron microscopy (TEM) and high-resolution transmission electron microscopy (HR-TEM). The electrocatalytic activity of Ni@Pt/C electrocatalysts toward BH 4 - oxidation in alkaline medium was studied by means of cyclic voltammetry (CV), chronopotentiometry (CP), chronoamperometry (CA) and electrochemical impedance spectroscopy (EIS). The results showed that Ni@Pt/C electrocatalyst synthesized without SDS has superior catalytic activity toward borohydride oxidation (22016.92Ag Pt -1 ) in comparison with a catalyst prepared in the presence of SDS (17766.15Ag Pt -1 ) in NaBH 4 0.1M at 25°C. The Membrane Electrode Assembly (MEA) used in fuel cell set-up was fabricated with catalyst-coated membrane (CCM) technique. The effect of Ni@Pt/C catalysts prepared with two methods as anode catalyst on the performance of direct borohydride-hydrogen peroxide fuel cell was studied. The maximum power density was obtained using Ni@Pt/C catalyst synthesized without SDS at 60°C, 1M NaBH 4 and 2M H 2 O 2 (133.38mWcm -2 ). Copyright © 2017 Elsevier Inc. All rights reserved.

  16. 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)

  17. Current density and polarization curves for radial flow field patterns applied to PEMFCs (Proton Exchange Membrane Fuel Cells)

    International Nuclear Information System (INIS)

    Cano-Andrade, S.; Hernandez-Guerrero, A.; Spakovsky, M.R. von; Damian-Ascencio, C.E.; Rubio-Arana, J.C.

    2010-01-01

    A numerical solution of the current density and velocity fields of a 3-D PEM radial configuration fuel cell is presented. The energy, momentum and electrochemical equations are solved using a computational fluid dynamics (CFD) code based on a finite volume scheme. There are three cases of principal interest for this radial model: four channels, eight channels and twelve channels placed in a symmetrical path over the flow field plate. The figures for the current-voltage curves for the three models proposed are presented, and the main factors that affect the behavior of each of the curves are discussed. Velocity contours are presented for the three different models, showing how the fuel cell behavior is affected by the velocity variations in the radial configuration. All these results are presented for the case of high relative humidity. The favorable results obtained for this unconventional geometry seems to indicate that this geometry could replace the conventional commercial geometries currently in use.

  18. Experimental study on density, thermal conductivity, specific heat, and viscosity of water-ethylene glycol mixture dispersed with carbon nanotubes

    Directory of Open Access Journals (Sweden)

    Ganeshkumar Jayabalan

    2017-01-01

    Full Text Available This article presents the effect of adding multi wall carbon nanotubes (MWCNT in water – ethylene glycol mixture on density and various thermophysical properties such as thermal conductivity, specific heat and viscosity. Density of nanofluids was measured using standard volumetric flask method and the data showed a good agreement with the mixing theory. The maximum thermal conductivity enhancement of 11 % was noticed for the nanofluids with 0.9 wt. %. Due to lower specific heat of the MWCNT, the specific heat of the nanofluids decreased in proportion with the MWCNT concentration. The rheological analysis showed that the transition region from shear thinning to Newtonian extended to the higher shear stress range compared to that of base fluids. Viscosity ratio of the nanofluids augmented anomalously with respect to increase in temperature and about 2.25 fold increase was observed in the temperature range of 30 – 40 ˚C. The modified model of Maron and Pierce predicted the viscosity of the nanofluids with the inclusion of effect of aspect ratio of MWCNT and nanoparticle aggregates.

  19. FIBWR2 evaluation of fuel thermal limits during density wave oscillaions in BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Nik, N.; Rajan, S.R.; Karasulu, M. [New York Power Authority, White Plains, NY (United States)

    1995-09-01

    Analyses were performed to evaluate hydraulic and thermal margin responses of three different BWR fuel designs subjected to the same periodic power/flow oscillations, such as those that might be exhibited during an instability event. The power/flow versus time information from the oscillations was used as a forcing function to calculate the hydraulic response and the MCPR performance of the limiting fuel bundles during the regional oscillations using the analytical code FIBWR2. The results of the calculations were used to determine the thermal margin variation as a function of oscillation magnitude.

  20. Energy hyperspace for stacking interaction in AU/AU dinucleotide step: Dispersion-corrected density functional theory study.

    Science.gov (United States)

    Mukherjee, Sanchita; Kailasam, Senthilkumar; Bansal, Manju; Bhattacharyya, Dhananjay

    2014-01-01

    Double helical structures of DNA and RNA are mostly determined by base pair stacking interactions, which give them the base sequence-directed features, such as small roll values for the purine-pyrimidine steps. Earlier attempts to characterize stacking interactions were mostly restricted to calculations on fiber diffraction geometries or optimized structure using ab initio calculations lacking variation in geometry to comment on rather unusual large roll values observed in AU/AU base pair step in crystal structures of RNA double helices. We have generated stacking energy hyperspace by modeling geometries with variations along the important degrees of freedom, roll, and slide, which were chosen via statistical analysis as maximally sequence dependent. Corresponding energy contours were constructed by several quantum chemical methods including dispersion corrections. This analysis established the most suitable methods for stacked base pair systems despite the limitation imparted by number of atom in a base pair step to employ very high level of theory. All the methods predict negative roll value and near-zero slide to be most favorable for the purine-pyrimidine steps, in agreement with Calladine's steric clash based rule. Successive base pairs in RNA are always linked by sugar-phosphate backbone with C3'-endo sugars and this demands C1'-C1' distance of about 5.4 Å along the chains. Consideration of an energy penalty term for deviation of C1'-C1' distance from the mean value, to the recent DFT-D functionals, specifically ωB97X-D appears to predict reliable energy contour for AU/AU step. Such distance-based penalty improves energy contours for the other purine-pyrimidine sequences also. © 2013 Wiley Periodicals, Inc. Biopolymers 101: 107-120, 2014. Copyright © 2013 Wiley Periodicals, Inc.

  1. 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

  2. On-line micro-volume introduction system developed for lower density than water extraction solvent and dispersive liquid–liquid microextraction coupled with flame atomic absorption spectrometry

    International Nuclear Information System (INIS)

    Anthemidis, Aristidis N.; Mitani, Constantina; Balkatzopoulou, Paschalia; Tzanavaras, Paraskevas D.

    2012-01-01

    Highlights: ► A dispersive liquid–liquid micro extraction method for lead and copper determination. ► A micro-volume transportation system for extractant solvent lighter than water. ► Analysis of natural water samples. - Abstract: A simple and fast preconcentration/separation dispersive liquid–liquid micro extraction (DLLME) method for metal determination based on the use of extraction solvent with lower density than water has been developed. For this purpose a novel micro-volume introduction system was developed enabling the on-line injection of the organic solvent into flame atomic absorption spectrometry (FAAS). The effectiveness and efficiency of the proposed system were demonstrated for lead and copper preconcentration in environmental water samples using di-isobutyl ketone (DBIK) as extraction solvent. Under the optimum conditions the enhancement factor for lead and copper was 187 and 310 respectively. For a sample volume of 10 mL, the detection limit (3 s) and the relative standard deviation were 1.2 μg L −1 and 3.3% for lead and 0.12 μg L −1 and 2.9% for copper respectively. The developed method was evaluated by analyzing certified reference material and it was applied successfully to the analysis of environmental water samples.

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

    International Nuclear Information System (INIS)

    Nuding, M.; Boening, K.

    2001-01-01

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

  4. Apparatus for dynamic monitoring of the size, shape, and density of fuel pellets

    International Nuclear Information System (INIS)

    Domoratskii, E.P.

    1994-01-01

    The objective of this study was to examine the structure, principle of operation, and technical characteristics of the KOMBI automated system for non-destructive monitoring of gas-cooled reactor fuel pellets. A detailed description of the apparatus was provided, and the technical characteristics were also presented

  5. Transport dynamics of a high-power-density matrix-type hydrogen-oxygen fuel cell

    Science.gov (United States)

    Prokopius, P. R.; Hagedorn, N. H.

    1974-01-01

    Experimental transport dynamics tests were made on a space power fuel cell of current design. Various operating transients were introduced and transport-related response data were recorded with fluidic humidity sensing instruments. Also, sampled data techniques were developed for measuring the cathode-side electrolyte concentration during transient operation.

  6. Validation of missed space-group symmetry in X-ray powder diffraction structures with dispersion-corrected density functional theory.

    Science.gov (United States)

    Hempler, Daniela; Schmidt, Martin U; van de Streek, Jacco

    2017-08-01

    More than 600 molecular crystal structures with correct, incorrect and uncertain space-group symmetry were energy-minimized with dispersion-corrected density functional theory (DFT-D, PBE-D3). For the purpose of determining the correct space-group symmetry the required tolerance on the atomic coordinates of all non-H atoms is established to be 0.2 Å. For 98.5% of 200 molecular crystal structures published with missed symmetry, the correct space group is identified; there are no false positives. Very small, very symmetrical molecules can end up in artificially high space groups upon energy minimization, although this is easily detected through visual inspection. If the space group of a crystal structure determined from powder diffraction data is ambiguous, energy minimization with DFT-D provides a fast and reliable method to select the correct space group.

  7. Dopant activation mechanism of Bi wire-δ-doping into Si crystal, investigated with wavelength dispersive fluorescence x-ray absorption fine structure and density functional theory.

    Science.gov (United States)

    Murata, Koichi; Kirkham, Christopher; Shimomura, Masaru; Nitta, Kiyofumi; Uruga, Tomoya; Terada, Yasuko; Nittoh, Koh-Ichi; Bowler, David R; Miki, Kazushi

    2017-04-20

    We successfully characterized the local structures of Bi atoms in a wire-δ-doped layer (1/8 ML) in a Si crystal, using wavelength dispersive fluorescence x-ray absorption fine structure at the beamline BL37XU, in SPring-8, with the help of density functional theory calculations. It was found that the burial of Bi nanolines on the Si(0 0 1) surface, via growth of Si capping layer at 400 °C by molecular beam epitaxy, reduced the Bi-Si bond length from [Formula: see text] to [Formula: see text] Å. We infer that following epitaxial growth the Bi-Bi dimers of the nanoline are broken, and the Bi atoms are located at substitutional sites within the Si crystal, leading to the shorter Bi-Si bond lengths.

  8. Mathematical modeling of current density distribution in composite cathode of solid oxide fuel cells. Paper no. IGEC-1-099

    International Nuclear Information System (INIS)

    Kenney, B.; Karan, K.

    2005-01-01

    Cathodes processes in a solid oxide fuel cell (SOFC) are thought to dominate the overall electrochemical losses. One strategy for minimizing the cathode electrochemical losses in a state-of-the-art SOFC that utilize lanthanum-strontium-manganate (LSM) electrocatalyst and yttria-stabilized-zirconia (YSZ) electrolyte is to utilize composite cathodes comprising a mixture of LSM and YSZ. Composite cathodes improve performance by extending the active reaction zone from electrolyte-electrode interface to throughout the electrode. In this study, a two-dimensional composite cathode model was developed to assess cathode performance in terms of current density distributions. The model results indicate that geometric and microstructural parameters strongly influence current density distribution. In addition electrode composition affects magnitude and distribution of current. An optimum composition for equal-sized LSM/YSZ is 40 vol% LSM and 60 vol% YSZ at 900 o C. (author)

  9. Efficient, High Power Density Hydrocarbon-Fueled Solid Oxide Stack System, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Precision Combustion, Inc. (PCI) proposes to develop and demonstrate an innovative high power density design for direct internal reforming of regolith off-gases...

  10. Efficient, high power density hydrocarbon-fueled solid oxide stack system, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Precision Combustion, Inc. (PCI) proposes to develop and demonstrate an innovative high power density design for direct internal reforming of regolith off-gases...

  11. Direct phase selection of initial phases from single-wavelength anomalous dispersion (SAD) for the improvement of electron density and ab initio structure determination.

    Science.gov (United States)

    Chen, Chung-De; Huang, Yen-Chieh; Chiang, Hsin-Lin; Hsieh, Yin-Cheng; Guan, Hong-Hsiang; Chuankhayan, Phimonphan; Chen, Chun-Jung

    2014-09-01

    Optimization of the initial phasing has been a decisive factor in the success of the subsequent electron-density modification, model building and structure determination of biological macromolecules using the single-wavelength anomalous dispersion (SAD) method. Two possible phase solutions (φ1 and φ2) generated from two symmetric phase triangles in the Harker construction for the SAD method cause the well known phase ambiguity. A novel direct phase-selection method utilizing the θ(DS) list as a criterion to select optimized phases φ(am) from φ1 or φ2 of a subset of reflections with a high percentage of correct phases to replace the corresponding initial SAD phases φ(SAD) has been developed. Based on this work, reflections with an angle θ(DS) in the range 35-145° are selected for an optimized improvement, where θ(DS) is the angle between the initial phase φ(SAD) and a preliminary density-modification (DM) phase φ(DM)(NHL). The results show that utilizing the additional direct phase-selection step prior to simple solvent flattening without phase combination using existing DM programs, such as RESOLVE or DM from CCP4, significantly improves the final phases in terms of increased correlation coefficients of electron-density maps and diminished mean phase errors. With the improved phases and density maps from the direct phase-selection method, the completeness of residues of protein molecules built with main chains and side chains is enhanced for efficient structure determination.

  12. The spectroscopic (FT-IR, FT-Raman, dispersive Raman and NMR) study of ethyl-6-chloronicotinate molecule by combined density functional theory.

    Science.gov (United States)

    Karabacak, Mehmet; Calisir, Zuhre; Kurt, Mustafa; Kose, Etem; Atac, Ahmet

    2016-01-15

    In this study, ethyl-6-chloronicotinate (E-6-ClN) molecule is recorded in the region 4000-400 cm(-1) and 3500-100 cm(-1) (FT-IR, FT-Raman and dispersive Raman, respectively) in the solid phase. ((1))H and ((13))C nuclear magnetic resonance (NMR) spectra are recorded in DMSO solution. The structural and spectroscopic data of the molecule are obtained for two possible isomers (S1 and S2) from DFT (B3LYP) with 6-311++G(d,p) basis set calculations. The geometry of the molecule is fully optimized, vibrational spectra are calculated and fundamental vibrations are assigned on the basis of the potential energy distribution (PED) of the vibrational modes. ((1))H and ((13))C NMR chemical shifts are calculated by using the gauge-invariant atomic orbital (GIAO) method. The electronic properties, such as excitation energies, oscillator strengths, wavelengths, HOMO and LUMO energies, are performed by time-dependent density functional theory (TD-DFT). Total and partial density of state and overlap population density of state diagrams analysis are presented for E-6-ClN molecule. Furthermore, frontier molecular orbitals (FMO), molecular electrostatic potential, and thermodynamic features are performed. In addition to these, reduced density gradient of the molecule is performed and discussed. As a conclusion, the calculated results are compared with the experimental spectra of the title compound. The results of the calculations are applied to simulate the vibrational spectra of the molecule, which show excellent agreement with the observed ones. The theoretical and tentative results will give us a detailed description of the structural and physicochemical properties of the molecule. Natural bond orbital analysis is done to have more information stability of the molecule arising from charge delocalization, and to reveal the information regarding charge transfer within the molecules. Copyright © 2015 Elsevier B.V. All rights reserved.

  13. Direct phase selection of initial phases from single-wavelength anomalous dispersion (SAD) for the improvement of electron density and ab initio structure determination

    International Nuclear Information System (INIS)

    Chen, Chung-De; Huang, Yen-Chieh; Chiang, Hsin-Lin; Hsieh, Yin-Cheng; Guan, Hong-Hsiang; Chuankhayan, Phimonphan; Chen, Chun-Jung

    2014-01-01

    A novel direct phase-selection method to select optimized phases from the ambiguous phases of a subset of reflections to replace the corresponding initial SAD phases has been developed. With the improved phases, the completeness of built residues of protein molecules is enhanced for efficient structure determination. Optimization of the initial phasing has been a decisive factor in the success of the subsequent electron-density modification, model building and structure determination of biological macromolecules using the single-wavelength anomalous dispersion (SAD) method. Two possible phase solutions (ϕ 1 and ϕ 2 ) generated from two symmetric phase triangles in the Harker construction for the SAD method cause the well known phase ambiguity. A novel direct phase-selection method utilizing the θ DS list as a criterion to select optimized phases ϕ am from ϕ 1 or ϕ 2 of a subset of reflections with a high percentage of correct phases to replace the corresponding initial SAD phases ϕ SAD has been developed. Based on this work, reflections with an angle θ DS in the range 35–145° are selected for an optimized improvement, where θ DS is the angle between the initial phase ϕ SAD and a preliminary density-modification (DM) phase ϕ DM NHL . The results show that utilizing the additional direct phase-selection step prior to simple solvent flattening without phase combination using existing DM programs, such as RESOLVE or DM from CCP4, significantly improves the final phases in terms of increased correlation coefficients of electron-density maps and diminished mean phase errors. With the improved phases and density maps from the direct phase-selection method, the completeness of residues of protein molecules built with main chains and side chains is enhanced for efficient structure determination

  14. 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)

  15. 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

  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. 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)

  18. 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

  19. Towards High Power Density Metal Supported Solid Oxide Fuel Cell for Mobile Applications

    DEFF Research Database (Denmark)

    Nielsen, Jimmi; Persson, Åsa Helen; Muhl, Thuy

    2017-01-01

    For use of metal supported SOFC in mobile applications it is important to reduce the thermal mass to enable fast start up, increase stack power density in terms of weight and volume and reduce costs. In the present study, we report on the effect of reducing the support layer thickness of 313 μm...

  20. Melting Point and Viscosity Behavior of High Energy Density Missile Fuels

    Science.gov (United States)

    1982-09-01

    CLASSIFICATION OF THIS PAGE (f,n Date Eneed . etrahydrodi(cyclopentadiene) ( XTHDCPD or JP-lO). HNN and HXX each have two crystalline forms. The solid-solid...suggesting solid solution formation on crystallization. The experimental m.p. curves for the binary/isomer I - XTHDCPD system could be used to predict m.p...liquidus temperature, of any/fuel blend of HNN, HXX, isomer I and XTHDCPD of kno composition. It )as found that the maximum m.p. specification of -54 C

  1. 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 co