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Sample records for gfr plate-type fuel

  1. GFR fuel and core pre-conceptual design studies

    International Nuclear Information System (INIS)

    Chauvin, N.; Ravenet, A.; Lorenzo, D.; Pelletier, M.; Escleine, J.M.; Munoz, I.; Bonnerot, J.M.; Malo, J.Y.; Garnier, J.C.; Bertrand, F.; Bosq, J.C.

    2007-01-01

    The revision of the GFR core design - plate type - has been undertaken since previous core presented at Global'05. The self-breeding searched for has been achieved with an optimized design ('12/06 E'). The higher core pressure drop was a matter of concern. First of all, the core coolability in natural circulation for pressurized conditions has been studied and preliminary plant transient calculations have been performed. The design and safety criteria are met but no more margin remains. The project is also addressing the feasibility and the design of the fuel S/A. The hexagonal shape together with the principle of closed S/A (wrapper tube) is kept. Ceramic plate type fuel element combines a high enough core power density (minimization of the Pu inventory) and plutonium and minor actinides recycling capabilities. Innovative for many aspects, the fuel element is central to the GFR feasibility. It is supported already by a significant R and D effort also applicable to a pin concept that is considered as the other fuel element of interest. This combination of fuel/core feasibility and performance analysis, safety dispositions and performances analysis will compose the 'GFR preliminary feasibility' which is a project milestone at the end of the year 2007. (authors)

  2. Evaluation of plate type fuel options for small power reactors

    International Nuclear Information System (INIS)

    Andrzejewski, Claudio de Sa

    2005-01-01

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

  3. Electromagnetic Acoustic Test of the Artificial Defects for a Plate-type Nuclear Fuel

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Kim, Dong Min; Lee, Yoon Sang; Cheong, Yong Moo

    2011-01-01

    Most research and test reactors use the nuclear fuel plates which are consisted of a fuel meat in aluminum alloy. Last year, KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of the plate-type fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done under immersion condition, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined is a non-ferromagnetic material such as aluminum with a good acousto-elastic property, which requires an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an Electromagnetic Acoustic Transducer (EMAT) technology for an automated inspection of a nuclear fuel without water

  4. Conceptual design of control rod regulating system for plate type fuels of Triga-2000 reactor

    International Nuclear Information System (INIS)

    Eko Priyono; Saminto

    2016-01-01

    Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor has been made. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor was made with refer to study result of instrument and control system which is used in BATAN'S reactor. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor consist of 4 segments that is control panel, translator, driver and display. Control panel is used for regulating, safety and display control rod, translator is used for signal processing from control panel, driver is used for driving control rod and display is used for display control rod level position. The translator was designed in 2 modes operation i.e operation by using PLC modules and IC TTL modules. These conceptual design can be used as one of reference of control rod regulating system detail design. (author)

  5. Fission rate measurements in fuel plate type assembly reactor cores

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1988-01-01

    The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

  6. Development of core technology for research reactors using plate type fuels

    International Nuclear Information System (INIS)

    Ha, Jae Joo; Lee, Doo Jeong; Park, Cheol

    2009-12-01

    Around 250 research reactors are under operation over the world. However, about 2/3 have been operated more than 30 years and demands for replacements are expected in the near future. The number of expected units is around 110, and around 55 units from 40 countries will be expected to be bid in the world market. In 2007, Netherlands started international bidding process to construct a new 80MW RR (named PALLAS) with the target of commercial operation in 2016, which will replace the existing HFR(45MW). KAERI consortium has been participated in that bid. Most of RRs use plate type fuels as a fuel assembly, Be and Graphite as a reflector. On the other hand, in Korea, the KAERI is operating the HANARO, which uses a rod type fuel assembly and heavy water as a reflector. Hence, core technologies for RRs using plate type fuels are in short. Therefore, core technologies should be secured for exporting a RR. In chapter 2, the conceptual design of PALLAS which use plate type fuels are described including core, cooling system and connected systems, layout of general components. Experimental verification tests for the plate type fuel and second shutdown system and the code verification for nuclear design are explained in Chapter 3 and 4, respectively

  7. Feasibility of Electromagnetic Acoustic Evaluation for Quality Test of a Plate-type Nuclear Fuel

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Lee, Yoon Sang; Cheong, Yong Moo

    2010-01-01

    Most research and test reactors use the nuclear fuel plates which are consisted of a fuel core in aluminum alloy. Recently KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done with water, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined within this paper is a non-ferromagnetic material such as aluminum which has a good acousto-elastic property, for an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an EMAT technology for an automated inspection of a nuclear fuel without water

  8. Heat conduction in a plate-type fuel element with time-dependent boundary conditions

    International Nuclear Information System (INIS)

    Faya, A.J.G.; Maiorino, J.R.

    1981-01-01

    A method for the solution of boundary-value problems with variable boundary conditions is applied to solve a heat conduction problem in a plate-type fuel element with time dependent film coefficient. The numerical results show the feasibility of the method in the solution of this class of problems. (Author) [pt

  9. Core conversion from rod to plate type fuel elements in research reactors

    International Nuclear Information System (INIS)

    Khattab, M.S.; Mina, A.R.

    1997-01-01

    Core thermalhydraulic analysis have been performed for rod and plate types fuel elements without altering the core bundles square grid spacer (68 mm, side) and coolant mass flow rate. The U O 2 -Mg, 10% enrichment rod type fuel elements are replaced by the MTR plate type, U-Al alloy of 20% enrichment. Coolant mass flux increased from 2000 kg/m 2 S to 5000 kg/m 2 S. Reactor power could be upgraded from 2 to 10 MW without significantly altering the steady state, thermal-hydraulic safety margins. Fuel, clad and coolant transient temperatures are determined inside the core hot channel during flow coast down using paret code. Residual heat removal system of 20% coolant capacity is necessary for upgrading reactor power to encounter the case of pumps off at 10 MW nominal operation. 6 figs., 2 tabs

  10. Prediction for the flow distribution and the pressure drop of a plate type fuel assembly

    International Nuclear Information System (INIS)

    Park, Jong Hark; Jo, Dea Sung; Chae, Hee Taek; Lee, Byung Chul

    2011-01-01

    A plate type fuel assembly widely used in many research reactors does not allow the coolant to mix with neighboring fuel channels due to the completely separated flow channels. If there is a serious inequality of coolant distribution among channels, it can reduce thermal-hydraulic safety margin, as well as it can cause a deformation of fuel plates by the pressure difference between neighboring channels, thus the flow uniformity in the fuel assembly should be confirmed. When designing a primary cooling system (PCS), the pressure drop through a reactor core is a dominant value to determine the PCS pump size. The major portion of reactor core pressure drop is caused by the fuel assemblies. However it is not easy to get a reasonable estimation of pressure drop due to the geometric complexity of the fuel assembly and the thin gaps between fuel assemblies. The flow rate through the gap is important part to determine the total flow rate of PCS, so it should be estimated as reasonable as possible. It requires complex and difficult jobs to get useful data. In this study CFD analysis to predict the flow distribution and the pressure drop were conducted on the plate type fuel assembly, which results would be used to be preliminary data to determine the PCS flow rate and to improve the design of a fuel assembly

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

    International Nuclear Information System (INIS)

    Ti Zhongxin; Guo Yibai

    1995-12-01

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

  12. Evaluation of plate type fuel elements by eddy current test method

    International Nuclear Information System (INIS)

    Frade, Rangel Teixeira

    2015-01-01

    Plate type fuel elements are used in MTR research nuclear reactors. The fuel plates are manufactured by assembling a briquette containing the fissile material inserted in a frame, with metal plates in both sides of the set, to act as a cladding. This set is rolled under controlled conditions in order to obtain the fuel plate. In Brazil, this type of fuel is manufactured by IPEN and used in the IEA-R1 reactor. After fabrication of three batches of fuel plates, 24 plates, one of them is taken, in order to verify the thickness of the cladding. For this purpose, the plate is sectioned and the thickness measurements are carried out by using optical microscopy. This procedure implies in damage of the plate, with the consequent cost. Besides, the process of sample preparation for optical microscopy analysis is time consuming, it is necessary an infrastructure for handling radioactive materials and there is a generation of radioactive residues during the process. The objective of this study was verify the applicability of eddy current test method for nondestructive measurement of cladding thickness in plate type nuclear fuels, enabling the inspection of all manufactured fuel plates. For this purpose, reference standards, representative of the cladding of the fuel plates, were manufactured using thermomechanical processing conditions similar to those used for plates manufacturing. Due to no availability of fuel plates for performing the experiments, the presence of the plate’s core was simulated using materials with different electrical conductivities, fixed to the thickness reference standards. Probes of eddy current testing were designed and manufactured. They showed high sensitivity to thickness variations, being able to separate small thickness changes. The sensitivity was higher in tests performed on the reference standards and samples without the presence of the materials simulating the core. For examination of the cladding with influence of materials simulating the

  13. Evaluation of Electron Beam Welding Performance of AA6061-T6 Plate-type Fuel Assembly

    International Nuclear Information System (INIS)

    Kim, Soo-Sung; Seo, Kyoung-Seok; Lee, Don-Bae; Park, Jong-Man; Lee, Yoon-Sang; Lee, Chong-Tak

    2014-01-01

    As one of the most commonly used heat-treatable aluminum alloys, AA6061-T6 aluminum alloy is available in a wide range of structural materials. Typically, it is used in structural members, auto-body sheet and many other applications. Generally, this alloy is easily welded by conventional GTAW (Gas Tungsten Arc Welding), LBW (Laser Beam Welding) and EBW(Electron Beam Welding). However, certain characteristics, such as solidification cracking, porosity, HAZ (Heat-affected Zone) degradation must be considered during welding. Because of high energy density and low heat input, especially LBW and EBW processes possess the advantage of minimizing the fusing zone and HAZ and producing deeper penetration than arc welding processes. In present study, to apply for the plate-type nuclear fuel fabrication and assembly, a fundamental electron beam welding experiment using AA6061-T6 aluminum alloy specimens was conducted. Furthermore, to establish the suitable welding process, and satisfy the requirements of the weld quality, EBW apparatus using an electron welding gun and vacuum chamber was developed, and preliminary investigations for optimizing the welding parameters of the specimens using AA6061-T6 aluminum plates were also performed. The EB weld quality of AA6061-T6 aluminum alloy for the plate-type fuel assembly has been also studied by the weld penetrations of side plate to end fitting and fixing bar and weld inspections using computed tomography

  14. Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

  15. Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

    International Nuclear Information System (INIS)

    Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S.

    2013-01-01

    The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

  16. Utilization of radiographic and ultrasonic testing for an evaluation of plate type fuel elements during manufacturing stages

    International Nuclear Information System (INIS)

    Brito, Mucio Jose Drummond de; Silva Junior, Silverio Ferreira da; Messias, Jose Marcos; Braga, Daniel Martins; Paula, Joao Bosco de

    2005-01-01

    Structural discontinuities can be introduced in the plate type fuel elements during the manufacturing stages due to mechanical processing conditions. The use of nondestructive testing methods to monitoring the fuel elements during the manufacturing stages presents a significant importance, contributing for manufacturing process improvement and cost reducing. This paper describes a procedure to be used detection and evaluation of structural discontinuities in plate type fuel elements during the manufacturing stages using the ultrasonic testing method and the radiographic testing method. The main results obtained are presented and discussed. (author)

  17. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    Energy Technology Data Exchange (ETDEWEB)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro, E-mail: duvan.castellanos@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: pedro.rossi@ufabc.edu.br, E-mail: pedro.carajilescov10@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharias, Modelagem e Ciências Sociais Aplicadas

    2017-07-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  18. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    International Nuclear Information System (INIS)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro

    2017-01-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  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. Fabrication of AA6061-T6 Plate Type Fuel Assembly Using Electron Beam Welding Process

    International Nuclear Information System (INIS)

    Kim, Soosung; Seo, Kyoungseok; Lee, Donbae; Park, Jongman; Lee, Yoonsang; Lee, Chongtak

    2014-01-01

    AA6061-T6 aluminum alloy is easily welded by conventional GTAW (Gas Tungsten Arc Welding), LBW (Laser Beam Welding) and EBW. However, certain characteristics, such as solidification cracking, porosity, HAZ (Heat-affected Zone) degradation must be considered during welding. Because of high energy density and low heat input, especially LBW and EBW processes possess the advantage of minimizing the fusing zone and HAZ and producing deeper penetration than arc welding processes. In present study, to apply for the nuclear fuel plate fabrication and assembly, a fundamental EBW experiment using AA6061-T6 aluminum alloy specimens was conducted. Furthermore, to establish the welding process, and satisfy the requirements of the weld quality, EBW apparatus using an electron welding gun and vacuum chamber was developed, and preliminary investigations for optimizing the welding parameters of the specimens using AA6061-T6 aluminum plates were also performed. The EB weld quality of AA6061-T6 aluminum alloy for the fuel plate assembly has been also studied by the shrinkage measurement and weld inspection using computed tomography. This study was carried out to determine the suitable welding parameters and to evaluate tensile strength of AA6061-T6 aluminum alloy. In the present experiment, satisfactory electron beam welding process of the full-sized sample was being developed. Based on this fundamental study, fabrication of the plate-type fuel assembly will be provided for the future Ki-Jang research reactor project

  1. Thermal Hydraulic Fortran Program for Steady State Calculations of Plate Type Fuel Research Reactors

    International Nuclear Information System (INIS)

    Khedr, H.

    2008-01-01

    The safety assessment of Research and Power Reactors is a continuous process over their life and that requires verified and validated codes. Power Reactor codes all over the world are well established and qualified against a real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume much more running time. On the other hand, most of the Research Reactor codes still requiring more data for validation and qualification. Therefore it is benefit for a regulatory body and the companies working in the area of Research Reactor assessment and design to have their own program that give them a quick judgment. The present paper introduces a simple one dimensional Fortran program called THDSN for steady state best estimate Thermal Hydraulic (TH) calculations of plate type fuel RRs. Beside calculating the fuel and coolant temperature distribution and pressure gradient in an average and hot channel the program calculates the safety limits and margins against the critical phenomena encountered in RR such as the burnout heat flux and the onset of flow instability. Well known TH correlations for calculating the safety parameters are used. THDSN program is verified by comparing its results for 2 and 10 MW benchmark reactors with that published in IAEA publications and good agreement is found. Also the program results are compared with those published for other programs such as PARET and TERMIC. An extension for this program is underway to cover the transient TH calculations

  2. 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

  3. 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

  4. 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

  5. Improvement of critical heat flux correlation for research reactors using plate-type fuel

    International Nuclear Information System (INIS)

    Kaminaga, Masanori; Yamamoto, Kazuyoshi; Sudo, Yukio

    1998-01-01

    In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities and are cooled by a downward flow. A core flow reversal from a steady-state forced downward flow to an upward flow due to natural convection should occur during operational transients such as Loss of the primary coolant flow'. Therefore, in the thermal hydraulic design of research reactors, critical heat flux (CHF) under a counter-current flow limitation (CCFL) or a flooding condition are important to determine safety margins of fuel against CHF during a core flow reversal. The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition. When the CHF correlation scheme was proposed, a subcooling effect for CHF correlation under CCFL condition had not been considered because of a conservative evaluation and a lack of enough CHF data to determine the subcooling effect on CHF. A too conservative evaluation is not appropriate for the design of research reactors because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effect on CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on CHF experimental results obtained under uniform and non-uniform heat flux conditions. As a result, it was made clear that CHF in this region increase linearly with an increase of the channel inlet subcooling and a new CHF correlation including the effect of channel inlet subcooling was proposed. The new correlation could be adopted under the conditions of the atmospheric pressure, the inlet subcooling less than 78K, the channel gap size between 2.25 to 5.0mm, the axial peaking factor between 1.0 to 1.6 and L/De between 71 to 174 which were the ranges investigated in this study. (author)

  6. Technical report: technical development on the silicide plate-type fuel experiment at nuclear safety research reactor

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Soyama, Kazuhiko; Ichikawa, Hiroki

    1991-08-01

    According to a reduction of fuel enrichment from 45 w/o 235 U to 20 w/o, an aluminide plate-type fuel used currently in the domestic research and material testing reactors will be replaced by a silicide plate-type one. One of the major concern arisen from this alternation is to understand the fuel behavior under simulated reactivity initiated accident (RIA) conditions, this is strongly necessary from the safety and licensing point of view. The in-core RIA experiments are, therefore, carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute (JAERI). The silicide plate-type fuel consisted of the ternary alloy of U-Al-Si as a meat with uranium density up to 4.8 g/cm 3 having thickness by 0.51 mm and the binary alloy of Al-3%Mg as a cladding by thickness of 0.38 mm. Comparison of the physical properties of this metallic plate fuel with the UO 2 -zircaloy fuel rod used conventionally in commercial light water reactors shows that the heat conductivity of the former is of the order of about 13 times greater than the latter, however the melting temperature is only one-half (1570degC). Prior to in-core RIA experiments, there were some difficulties lay in our technical path. This report summarized the technical achievements obtained through our four years work. (J.P.N.)

  7. Post-pulse detail metallographic examinations of low-enriched uranium silicide plate-type miniature fuel

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1991-10-01

    Pulse irradiation at Nuclear Safety Research Reactor (NSRR) was performed using low-enriched (19.89 w% 235 U) unirradiated silicide plate-type miniature fuel which had a density of 4.8 gU/cm 3 . Experimental aims are to understand the dimensional stability and to clarify the failure threshold of the silicide plate-type miniature fuel under power transient conditions through post-pulse detail metallographic examinations. A silicide plate-type miniature fuel was loaded into an irradiation capsule and irradiated by a single pulse. Deposited energies given in the experiments were 62, 77, 116 and 154 cal/g·fuel, which lead to corresponding peak fuel plate temperatures, 201 ± 28degC, 187 ± 10degC, 418 ± 74degC and 871 ± 74degC, respectively. Below 400degC, reliability and dimensional stability of the silicide plate fuel was sustained, and the silicide plate fuel was intact. Up to 540degC, wall-through intergranular crackings occurred in the Al-3%Mg alloy cladding. With the increase of the temperature, the melting of the aluminum cladding followed by recrystallization, the denudation of fuel core and the plate-through intergranular cracking were observed. With the increase of the temperature beyond 400degC, the bowing of fuel plate became significant. Above the temperature of 640degC molten aluminum partially reacted with the fuel core, partially flowed downward under the influence of surface tension and gravity, and partially formed agglomerations. Judging from these experimental observations, the fuel-plate above 400degC tends to reduce its dimensional stability. Despite of the apparent silicide fuel-plate failure, neither generation of pressure pulse nor that of mechanical energy occurred at all. (J.P.N.)

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

    International Nuclear Information System (INIS)

    Li Shulan; He Fengqi; Wang Qingheng; Han Jingquan

    1993-04-01

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

  9. 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

  10. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    OpenAIRE

    Itamar Iliuk; José Manoel Balthazar; Ângelo Marcelo Tusset; José Roberto Castilho Piqueira

    2016-01-01

    Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was prop...

  11. Improvement of visualization efficiency for the nondestructive inspection image of internal defects in plate type nuclear fuel

    International Nuclear Information System (INIS)

    Park, Seung Kyu; Park, Nak Kyu; Baik, Sung Hoon; Lee, Yoon Sang; Cheong, Yong Moo; Kang, Young June

    2012-01-01

    Plate type nuclear fuel has been adopted in most research reactors. The production quality of the fuel is a key part for an efficient and stable generation of thermal energy in research reactors. Thus, a nondestructive quality inspection for the internal defects of plate type nuclear fuel is a key process during the production of nuclear fuel for safety insurance. Nondestructive quality inspections based on X rays and ultrasounds have been widely used for the defect detection of plate type nuclear fuel. X ray testing is a simple and fast inspection method, and provides an image in real time as the inspection results. Thus, the testing can be carried out by a non expert field worker. However, it is hard to detect closed type defects that should be detected during the production of plate type nuclear fuel. Ultrasonic testing is a powerful tool to detect internal defects including open type and closed type defects in plate type nuclear fuel. However, the inspection process is complicated because an immersion test should be carried out in a water tank. It is also a time consuming inspection method because area testing to acquire image is based on the scanning of the point by point inspections. Among nondestructive inspection techniques, the techniques based on laser interferometry and infrared thermography have been widely used in the detection of internal defects of plate type composite materials, such as aircraft, automotive etc. While infrared thermography technique (IRT) analyses the thermal behavior of the specimen surface, laser interferometry technique (LIT) analyses the deformation field. Both techniques are useful tools for detection and evaluation of internal defects in composite materials. Especially, the laser interferometry technique can provide the depth information of internal defects. Laser interferometry technique (LIT) is a non contact inspection method faster than thermography. Also, this technique requires less energy than thermography and the

  12. Quality verification for plate-type uranium-aluminum fuel elements for use in research reactors (Revision 1) - July 1976

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    Paragraph (a) (7) of 50.34, Contents of Applications: Technical Information, of 10 CFR Part 50, Licensing of Production and Utilization Facilities, requires that each applicant for a construction permit to build a production or utilization facility include in its Preliminary Safety Analysis Report (PSAR) a description of the quality assurance program to be applied to the design, fabrication, construction, and testing of the structures, systems, and components of the facility. The Regulatory Guide presented describes a method acceptable to the NRC staff for establishing and executing a quality assurance program for verifying the quality of plate-type uranium-aluminum fuel elements used in research reactors

  13. Simplified CFD model of coolant channels typical of a plate-type fuel element: an exhaustive verification of the simulations

    Energy Technology Data Exchange (ETDEWEB)

    Mantecón, Javier González; Mattar Neto, Miguel, E-mail: javier.mantecon@ipen.br, E-mail: mmattar@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    The use of parallel plate-type fuel assemblies is common in nuclear research reactors. One of the main problems of this fuel element configuration is the hydraulic instability of the plates caused by the high flow velocities. The current work is focused on the hydrodynamic characterization of coolant channels typical of a flat-plate fuel element, using a numerical model developed with the commercial code ANSYS CFX. Numerical results are compared to accurate analytical solutions, considering two turbulence models and three different fluid meshes. For this study, the results demonstrated that the most suitable turbulence model is the k-ε model. The discretization error is estimated using the Grid Convergence Index method. Despite its simplicity, this model generates precise flow predictions. (author)

  14. Simplified CFD model of coolant channels typical of a plate-type fuel element: an exhaustive verification of the simulations

    International Nuclear Information System (INIS)

    Mantecón, Javier González; Mattar Neto, Miguel

    2017-01-01

    The use of parallel plate-type fuel assemblies is common in nuclear research reactors. One of the main problems of this fuel element configuration is the hydraulic instability of the plates caused by the high flow velocities. The current work is focused on the hydrodynamic characterization of coolant channels typical of a flat-plate fuel element, using a numerical model developed with the commercial code ANSYS CFX. Numerical results are compared to accurate analytical solutions, considering two turbulence models and three different fluid meshes. For this study, the results demonstrated that the most suitable turbulence model is the k-ε model. The discretization error is estimated using the Grid Convergence Index method. Despite its simplicity, this model generates precise flow predictions. (author)

  15. Detection of delamination defects in plate type fuel elements applying an automated C-Scan ultrasonic system

    International Nuclear Information System (INIS)

    Katchadjian, P.; Desimone, C.; Ziobrowski, C.; Garcia, A.

    2002-01-01

    For the inspection of plate type fuel elements to be used in Research Nuclear Reactors it was applied an immersion pulse-echo ultrasonic technique. For that reason an automated movement system was implemented according to the axes X, Y and Z that allows to automate the test and to show the results obtained in format of C-Scan, facilitating the immediate identification of possible defects and making repetitive the inspection. In this work problems found during the laboratory tests and factors that difficult the inspection are commented. Also the results of C-Scans over UMo fuel elements with pattern defects are shown. Finally, the main characteristics of the transducer with the one the better results were obtained are detailed. (author)

  16. 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)

  17. Implementation of a quality assurance system for the design and manufacturing of fuel assembly MTR-plate type

    International Nuclear Information System (INIS)

    Koll, J.H.

    1987-01-01

    Since more than 30 years ago, fuel assemblies (FA) of the MTR-Plate type, for research reactors, have been developed and produced using well known technologies, with different methods for the design, manufacturing, quality control and subsequent verification of FA behaviour, as well as of the design data. The FA and its reliability has been improved through the recycling of the obtained information. No nuclear accidents or major incidents have taken place that can be blamed to FA due to design, manufacturing or its use. Since the 70's, the use of Quality Assurance methodology has been increased, especially for Nuclear Power Plants, in order to ensure safety for these reactors. The use of QA for reactors for research, testing or other uses, has also been steadily increased, not only due to safety reasons, but also because of its convenience for a good operation, being presently a common requirement of the operator of the installation. Herewith is described the way the QA system that has been developed for the design, manufacturing, quality control and supply of MTR-plate type FA, at the Development Section of the Argentine Atomic Energy Commission (CNEA). (Author)

  18. 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)

  19. Standardization of specifications and inspection procedures for LEU plate-type research reactor fuels

    International Nuclear Information System (INIS)

    1988-06-01

    With the transition to high density uranium LEU fuel, fabrication costs of research reactor fuel elements have a tendency to increase because of two reasons. First, the amount of the powder of the uranium compound required increases by more than a factor of five. Second, fabrication requirements are in many cases nearer the fabrication limits. Therefore, it is important that measures be undertaken to eliminate or reduce unnecessary requirements in the specification or inspection procedures of research reactor fuel elements utilizing LEU. An additional stimulus for standardizing specifications and inspection procedures at this time is provided by the fact that most LEU conversions will occur within a short time span, and that nearly all of them will require preparation of new specifications and inspection procedures. In this sense, the LEU conversions offer an opportunity for improving the rationality and efficiency of the fuel fabrication and inspection processes. This report focuses on the standardization of specifications and inspection processes of high uranium density LEU fuels for research reactors. However, in many cases the results can also be extended directly to other research reactor fuels. 15 refs, 1 fig., 3 tabs

  20. 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

  1. 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.

  2. 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.

  3. Analysis of gamma heating at TRIGA mark reactor core Bandung using plate type fuel

    International Nuclear Information System (INIS)

    Setiyanto; Tukiran Surbakti

    2016-01-01

    In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0.87 W/g), but very low value for Lazy Susan position (lest then 0.11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. (author)

  4. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    Directory of Open Access Journals (Sweden)

    Itamar Iliuk

    2016-01-01

    Full Text Available Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.

  5. Experimental study on DNB heat flux of plate-type fuel in pressurized condition

    International Nuclear Information System (INIS)

    Komori, Yoshihiro; Oshima, Kunio; Ishitsuka, Etsuo; Sakurai, Fumio; Sudo, Yukio; Saito, Minoru; Futamura, Yoshiaki; Kaminaga, Masanori.

    1992-07-01

    Experimental study was carried out in order to determine the DNB correlation for the safety analysis of the JMTR low enrichment fuel core. Since it is essential to examine applicability and safety margin of the correlation for the safety analysis, DNB heat fluxes were measured with the test section of rectangular flow channel simulating JMTR fuel element subchannel in the pressure range of 1 ∼ 13 kg/cm 2 abs and the velocity range of 0 ∼ 4.4 m/s. Reviewing existed DNB correlations based on the experimental data, Sudo correlations scheme was selected for the JMTR safety analysis with minor modification for the high flow rate region. Comparing the correlations scheme with experimental data, allowable limit of the minimum DNBR was determined to be 1.5. (author)

  6. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL

    2014-05-01

    Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

  7. Analysis of steam explosions in plate-type, uranium-aluminum fuel test reactors

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1989-01-01

    The concern over steam explosions in nuclear reactors can be traced to prompt critical nuclear excursions in aluminum-clad/fueled test reactors, as well as to explosive events in aluminum, pulp, and paper industries. The Reactor Safety Study prompted an extensive analytical and experimental effort for over a decade. This has led to significant improvements in their understanding of the steam explosion issue for commercial light water reactors. However, little progress has been made toward applying the lessons learned from this effort to the understanding and modeling of steam explosion phenomena in aluminum-clad/fueled research and test reactors. The purposes of this paper are to (a) provide a preliminary analysis of the destructive events in test reactors, based on current understandings of steam explosions; (b) provide a proposed approach for determining the likelihood of a steam explosion event under scenarios in which molten U-Al fuel drops into a water-filled cavity; and (c) present a benchmarking study conducted to estimate peak pressure pulse magnitudes

  8. Safety criterion for burnout of the plate-type fuel in pressurized conditions

    International Nuclear Information System (INIS)

    Komori, Y.; Kaminaga, M.; Sakurai, F.; Ando, H.; Sudo, Y.; Saito, M.; Futamura, Y.

    1992-01-01

    The reduced enrichment program for JMTR is now underway and the core conversion to LEU (Low Enrichment Uranium) is scheduled to be made in 1993. Consistent with the safety guide which have been recently developed for research and test reactors in Japan, the safety analysis for the JMTR LEU conversion was conducted. In the safety analysis, DNB (Departure from Nucleate Boiling) heat flux correlation for the JMTR downflow condition was reconsidered because recent studies on burnout show that DNB heat fluxes with thin rectangular channels under low flow rate and low pressure conditions are much lower than predicted values by conventional DNB correlations. Available DNB data, however, are very limited for the JMTR operation pressure range, so that DNB experiments were conducted simulating the JMTR fuel subchannel. Based mainly on the present experimental data, the DNB correlations scheme composed of three correlations was selected for the JMTR safety analysis. Errors of the correlations scheme with experimental data were evaluated in order to determine the allowable limit of the minimum DNB ratio for preventing fuel failure. (author)

  9. ANALYSIS OF GAMMA HEATING AT TRIGA MARK REACTOR CORE BANDUNG USING PLATE TYPE FUEL

    Directory of Open Access Journals (Sweden)

    Setiyanto Setiyanto

    2016-10-01

    Full Text Available ABSTRACT In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities and central irradiation position (CIP, especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0,87 W/g, but very low value for Lazy Susan position (lest then 0,11 W/g. Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. Keywords: gamma heating, nuclear reactor, research reactor, reactor safety.   ABSTRAK Dengan dihentikannya produksi elemen bakar reaktor jenis Triga oleh produsen, maka semua reaktor TRIGA di dunia terganggu operasinya, termasuk juga reaktor TRIGA 2000 di Bandung. Untuk mendukung pengoperasian reaktor TRIGA Bandung

  10. Thermal-hydraulic Fortran program for steady-state calculations of plate-type fuel research reactors

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2008-01-01

    Full Text Available The safety assessment of research and power reactors is a continuous process covering their lifespan and requiring verified and validated codes. Power reactor codes all over the world are well established and qualified against real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume a lot of running time. On the other hand, most research reactor codes still require much more data for validation and qualification. It is, therefore, of benefit to any regulatory body to develop its own codes for the review and assessment of research reactors. The present paper introduces a simple, one-dimensional Fortran program called THDSN for steady-state thermal-hydraulic calculations of plate-type fuel research reactors. Besides calculating the fuel and coolant temperature distributions and pressure gradients in an average and hot channel, the program calculates the safety limits and margins against the critical phenomena encountered in research reactors, such as the onset of nucleate boiling, critical heat flux and flow instability. Well known thermal-hydraulic correlations for calculating the safety parameters and several formulas for the heat transfer coefficient have been used. The THDSN program was verified by comparing its results for 2 and 10 MW benchmark reactors with those published in IAEA publications and a good agreement was found. Also, the results of the program are compared with those published for other programs, such as the PARET and TERMIC.

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

    Energy Technology Data Exchange (ETDEWEB)

    Andrzejewski, Claudio de Sa

    2005-07-01

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

  12. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-15

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  15. Inert materials for the GFR fuel. Characterizations, chemical interactions and irradiation damage

    International Nuclear Information System (INIS)

    Audubert, Fabienne; Carlot, Gaoelle; Lechelle, Jacques; David, Laurent; Gomes, Severine

    2005-01-01

    In the framework of an extensive R and D Program on GFR fuel, studies on inert materials have been performed at the French Atomic Energy Commission (CEA). The inert materials would be associated with the fuel with the aim of featuring an efficient barrier to radiotoxic species with regard to the cooling circuit of the reactor. Potential matrices identified for dispersion fuels or particles fuels are SiC, TiN, ZrN, ZrC, TiC. Physical microstructural and thermal properties have been determined in order to evaluate elaboration process effects. The evolution under irradiation of thermal properties (such as conductivity, diffusivity) of the materials has been studied using heavy ions to simulate fission product irradiation. After irradiation, scanning thermal microscopy is used to investigate the thermal degradation of the materials. Thermal conductivity variations were obtained on TiC irradiated with krypton ion at an energy of 86 MeV and a fluence of 5.10 15 ions.cm -2 . They are quantified at 19 W.m -1 .K -1 . On other materials such as SiC, ZrC, TiN, no thermal conductivity contrast was shown. Reactivity between the inert matrix (SiC or TiN) and the fuel (U, Pu)N have been evaluated on powders and on ceramic samples in contact by a thermal treatment under several atmospheres. It was shown that SiC reacts with (U, Pu)N in various atmospheres making secondary phases as PuSi 2 , USi 2 , U 20 Si 16 C 3 . TiN behaviour seems to be better: the only reactivity which may take place would be a variation of the nitrogen stoichiometry in TiN and (U, Pu)N at the interface. (author)

  16. Development of neutronics and thermal hydraulics coupled code – SAC-RIT for plate type fuel and its application to reactivity initiated transient analysis

    International Nuclear Information System (INIS)

    Singh, Tej; Kumar, Jainendra; Mazumdar, Tanay; Raina, V.K.

    2013-01-01

    Highlights: • A point reactor kinetics code coupled with thermal hydraulics of plate type fuel is developed. • This code is applicable for two phase flow of coolant. • Safety analysis of IAEA benchmark reactor core is carried out. • Results agree well with the results available in literature. - Abstract: A point reactor kinetics code SAC-RIT, acronym of Safety Analysis Code for Reactivity Initiated Transient, coupled with thermal hydraulics of two phase coolant flow for plate type fuel, is developed to calculate reactivity initiated transient analysis of nuclear research and test reactors. Point kinetics equations are solved by fourth order Runge Kutta method. Reactivity feedback effect is included into the code. Solution of kinetics equations gives neutronic power and it is then fed into a thermal hydraulic code where mass, momentum and thermal energy conservation equations are solved by explicit finite difference method to find out fuel, clad and coolant temperatures during transients. In this code, all possible flow regimes including laminar flow, transient flow and turbulent flow have been covered. Various heat transfer coefficients suitable for single liquid, sub-cooled boiling, saturation boiling, film boiling and single vapor phases are incorporated in the thermal hydraulic code

  17. Gas-cooled Fast Reactor (GFR) fuel and In-Core Fuel Management

    International Nuclear Information System (INIS)

    Weaver, K.D.; Sterbentz, J.; Meyer, M.; Lowden, R.; Hoffman, E.; Wei, T.Y.C.

    2004-01-01

    The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO 2 ) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

  18. Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

    2009-08-01

    A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

  19. 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.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  1. 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)

  2. 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)

  3. 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)

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  5. Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

    Science.gov (United States)

    Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

    2014-07-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

  6. 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)

  7. Preliminary results for the Co-Rolling process fabrication of plate-type nuclear fuel based in U-10Mo monolithic meat and zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Pedrosa, Tercio A.; Brina, Jose Giovanni M.; Paula, Joao Bosco de; Lameiras, Fernando S.; Ferraz, Wilmar B.

    2013-01-01

    The fabrication process of plate-type nuclear fuel with monolithic meat is under development at CDTN. The U-10Mo alloy was chosen as the meat material due to its high density, corrosion resistance and the higher dimensional stability proportioned by the metastable gamma phase, which presents a lesser extension of the breakaway swelling phenomena occurrence during irradiation tests. The monolithic meat was cut from an U-10Mo ingot, that was induction melted at CDTN. The co-rolling process was adopted due to the higher mechanical properties and melting point of the Zircalloy-4 cladding material, which presents a lesser discrepancy in relation to the meat material properties, when compared to the aluminum 6061 alloy. Preliminary plates were obtained by means of the co-rolling process, performed at 650, 800, 950 deg C with total thickness reduction of 80%, followed by a pickling step and cold co-rolling passes. The plates were characterized through bending tests, optical microscopy and radiography. The co-rolling temperature of 800 deg C presented the best results, with a homogeneous distribution of the total thickness reduction between the cover plates and the meat, and the absence of delamination in the bending test samples. It was observed the occurrence of meat thickening in its ends, according to its longitudinal axle, parallel to the rolling direction, that is known as the d og bone , for the three co-rolling temperatures. (author)

  8. Evolution calculations of fuel for a GFR using MCNPX-C90 and Tripoli-4-D; Calculos de evolucion de combustible para un GFR usando MCNPX-C90 y TRIPOLI-4-D

    Energy Technology Data Exchange (ETDEWEB)

    Reyes R, R.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Brun, E.; Dumonteil, E.; Malvagi, F., E-mail: emeric.brun@cea.fr [Commissariat a l' Energie Atomique et aux Energies Alternative, Service d' Etude des Reacteurs et de Mathematiques Appliquees, Saclay, DEN/DM2S/SERMA/LTSD, Bat 470, 91191 Gif-sur-Yvette Cedex (France)

    2011-11-15

    Burnt calculations were realized for a fuel model based on the technology of the Gas-cooled Fast Reactor, GFR. The fuel design is based on bars. The code MCNPX-CINDER90 and the CSADA method for the burnt calculations were used. Models of homogeneous and heterogeneous fuel assembly were studied; for the burnt calculations of the fuel homogeneous model was considered the tracking of three series (Tiers) of evolution of the fission products. The Tier 1 tracks a reduced group of fission products, the Tier 2 tracks to the arrangement of fission products that are contained in the library of cross sections XSDIR of MCNPX; and the Tier 3 tracks 1325 fission products. The results were compared with those obtained with Tripoli-4-D in function of the calculation methods: 1) Explicit Euler, as method of first order; and 2) CSADA, as method of second order. According to the results was observed that the infinite multiplication factor varies in function of the fission products quantity that are tracked. The calculation time used by MCNPX-C90 with the series Tier 3 is more than double than the used by Tripoli-4-D, therefore this last code has advantage over MCNPX-C90 in the case of neutrons analysis of fast reactors. (Author)

  9. Thermal-hydraulic behavior of physical quantities at critical velocities in a nuclear research reactor core channel using plate type fuel

    Directory of Open Access Journals (Sweden)

    Sidi Ali Kamel

    2012-01-01

    Full Text Available The thermal-hydraulic study presented here relates to a channel of a nuclear reactor core. This channel is defined as being the space between two fuel plates where a coolant fluid flows. The flow velocity of this coolant should not generate vibrations in fuel plates. The aim of this study is to know the distribution of the temperature in the fuel plates, in the cladding and in the coolant fluid at the critical velocities of Miller, of Wambsganss, and of Cekirge and Ural. The velocity expressions given by these authors are function of the geometry of the fuel plate, the mechanical characteristics of the fuel plate’s material and the thermal characteristics of the coolant fluid. The thermal-hydraulic study is made under steady-state; the equation set-up of the thermal problem is made according to El Wakil and to Delhaye. Once the equation set-up is validated, the three critical velocities are calculated and then used in the calculations of the different temperature profiles. The average heat flux and the critical heat flux are evaluated for each critical velocity and their ratio reported. The recommended critical velocity to be used in nuclear channel calculations is that of Wambsganss. The mathematical model used is more precise and all the physical quantities, when using this critical velocity, stay in safe margins.

  10. Preliminary calculations of stress change of fuel pin using SiC/SiC composites for GFR with changing of thermal conductivity degradation by irradiation

    International Nuclear Information System (INIS)

    Lee, J. K.; Naganuma, M.

    2006-01-01

    Gas cooled Fast Reactor (GFR) is being researched as a candidate concept of Generation IV international Forum. As a main feature of GFR, it should be maintained high temperature and pressure of coolant gas for heat transfer efficiency. Such a demanding environment requires high-temperature-resistant structural materials distinguished from traditional steel material. Consequently, ceramics are promising candidate material of core components. Especially, Silicon Carbide fiber reinforced Silicon Carbide composites (SiC/SiC) have encouraging characteristics such as refractoriness, low activation and toughness. Application of new material to core components must be explained by the viewpoint of engineering validity. Therefore, present study surveyed that current report for mechanical strength and thermal conductivity of SiC/SiC composites. According to the reports, neutron irradiation environment degraded mechanical properties of SiC/SiC composites. To confirm applicability to core components, model of fuel pin using SiC/SiC composites was assumed with feasible mechanical properties. Furthermore, it was calculated and estimated that the stress caused by temperature variation of inner and outer side of assumed model of cladding tube. Stress was calculated by changing of input date such as thickness of cladding tube, temperature variation, thermal conductivity and linear power. In the range of this study, the most important factor was identified as degradation of thermal conductivity by irradiation. It caused a significant stress and limited a geometrical design of fuel pin. It was discussed that the differences of heat transfer between isotropic and anisotropic materials like a metal and composites. These results should be helpful not only to determine a design factor of core component but also to indicate an improvement direction of SiC/SiC composites. Through these work, reliability and safety of GFR will be increased

  11. Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

    International Nuclear Information System (INIS)

    Čerba, Štefan; Vrban, Branislav; Lüley, Jakub; Dařílek, Petr; Zajac, Radoslav; Nečas, Vladimír; Haščik, Ján

    2014-01-01

    Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice

  12. GFR demonstrator ALLEGRO design status

    International Nuclear Information System (INIS)

    Poette, C.; Malo, J.Y.; Brun-Magaud, V.; Morin, F.; Dor, I.; Mathieu, B.; Duhamel, H.; Stainsby, R.; Mikityuk, K.

    2009-01-01

    The ALLEGRO project has the ambitious goal of building and operating the first Gas Cooled Fast Reactor (GFR). It will be a low power experimental reactor with the main objective to validate on a pilot scale the specific GFR technologies (fuel element and sub-assembly, safety systems). It is a loop type, non electricity generating reactor. Its power is about 80 MW. The approach for the core includes first MOX cores loaded with some ceramic mixed carbide or nitride sub-assemblies with SiC/SiCf cladding and wrappers. When such unit test will be considered convincing enough, the diagrid and circuits are designed to accept full high temperature ceramic cores. The core neutrons can also be used to irradiate structural materials with fast neutron spectrum and in a large temperature range. The core can also include innovative irradiation fuel devices (samples or full bundles) for other reactor systems. Finally, the primary circuit can be connected to a test loop to validate the reactor coupled operation of a high temperature process or component. The paper deals with the current ALLEGRO design studies on a mid term roadmap aiming at ending the viability phase in 2012 in order to make a decision in 2013 for further detailed design and construction. Since 2005, the ALLEGRO design studies are shared in the GCFR 6th Framework Program which gathers 10 partners from 6 European countries. The paper will give an overview of recent progresses in various areas such as: - Last 3D core physics analysis of the MOX cores and their irradiation performances in terms of fast flux, dose/burnup, irradiation locations. - The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the MOX core. - Fuel handling principles and solutions. - System design and global reactor architecture which is largely influenced by the Decay Heat Removal strategy (DHR) for depressurized accidents. - An overview of the system transient analysis performed by the partners

  13. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    International Nuclear Information System (INIS)

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05)

  14. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  15. Glomerular filtration rate (GFR) and estimation of the GFR (eGFR) in ...

    African Journals Online (AJOL)

    in 2001 and has since worked at several laboratories within the Central region of the NHLS. Correspondence to: Jocelyn .... that female lean body mass was approx 15% lower).6 The ... estimation of GFR as part of management of patients with ...

  16. Estimated GFR (eGFR by prediction equation in staging of chronic kidney disease compared to gamma camera GFR

    Directory of Open Access Journals (Sweden)

    Mohammad Masum Alam

    2016-07-01

    Full Text Available Background: Glomerular filtration rate is an effective tool for diagnosis and staging of chronic kidney disease. The effect ofrenal insufficiency by different method of this tool among patients with CKD is controversial.Objective: The objec­tive of this study was to evaluate the performance of eGFR in staging of CKD compared to gamma camera based GFR.Methods: This cross sectional analytical study was conducted in the Department of Biochemistry Bangabandhu Sheikh Mujib Medical University (BSMMU with the collaboration with National Institute of Nuclear Medicine and Allied Sciences, BSMMU during the period of January 2011 to December 2012. Gama camera based GFR was estimated from DTP A reno gram and eGFR was estimated by three prediction equations. Comparison was done by Bland Altman agree­ment test to see the agreement on the measurement of GFR between three equation based eGFR method and gama camera based GFR method. Staging comparison was done by Kappa analysis to see the agreement between the stages identified by those different methods.Results: Bland-Altman agreement analysis between GFR measured by gamma camera, CG equation ,CG equation corrected by BSA and MDRD equation shows statistically significant. CKD stages determined by CG GFR, CG GFR corrected by BSA , MDRD GFR and gamma camera based GFR was compared by Kappa statistical analysis .The kappa value was 0.66, 0.77 and 0.79 respectively.Conclusions: This study findings suggest that GFR estimation by MDRD equation in CKD patients shows good agreement with gamma camera based GFR and for staging of CKD patients, eGFR by MDRD formula may be used as very effective tool in Bangladeshi population.

  17. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  18. Interim Status Report on the Design of the Gas-Cooled Fast Reactor (GFR)

    International Nuclear Information System (INIS)

    Weaver, K. D.

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above

  19. Effect of plate shapes in orifice plate type flowmeters

    International Nuclear Information System (INIS)

    Moeller, S.V.

    1984-01-01

    The study of unusual plate shapes in orifice plate type flowmeters is presented, with a view to providing data for the substitution of the plate with one centered circular orifice in those applications where its use is not possible. For this purpose, six pairs of plates with different forms, with and without chamfered edges, were made and tested in a closed water loop. Results show that, generally, the use of chamfers improves the results and, in the case of perforated and slotlike orificed plates, the narrow-ness of the fluid passage tends to make unnecessary its use. (Author) [pt

  20. Development of maintenance procedure for plate type heat exchanger taking into account preventing radioactive contamination

    International Nuclear Information System (INIS)

    Terai, Kensuke; Someki, Hiroyuki; Ueda, Yuya

    2017-01-01

    In Japanese pressurized water reactors (PWR), heat loads of spent fuel pools (SFP) is increasing due to rising spent fuels and use of mixed oxide (MOX) fuels. Therefore, SFP cooling capacities are necessary to be enhanced, and replacement of SFP coolers or installation of additional coolers is needed. On the other hand, installation spaces of SFP coolers are limited in existing buildings. Therefore, plate type heat exchangers which can be designed to be compact because of the high heat efficiency have often been adopted for SFP coolers instead of shell and tube type heat exchangers in general use. Plate type heat exchangers have to be overhauled periodically for inspection and gasket replacement. However, in plate type SFP coolers, radioactive SFP water and non-radioactive component cooling water (CCW) alternately run through between each plate. Thus there is a concern that the CCW system may be contaminated by radioactive materials from the SFP water during overhaul of the SFP cooler. In order to solve this problem, we have developed the maintenance procedure of the plate type SFP coolers to prevent CCW side contamination by coating the contaminated surfaces with strippable paint prior to disassembly. Before applying this developed maintenance procedure to actual equipment, we have performed the following verification tests. (1) Confirmation of fundamental characteristics for strippable paint. Firstly, we selected both water-based and solvent-based strippable paints. Secondly, we tested and confirmed the detachability and the drying time of the selected strippable paints respectively. Moreover we also confirmed that the selected strippable paints are appropriate materials from the viewpoint of chemical composition restriction of consumable materials used in nuclear power plant. (2) Confirmation of workability for paint filling, drying and peeling off. The strippable paints need to be peeled off after filling into plate type heat exchanger and draining

  1. Clinical study of GFR and split renal GFR in evaluating the glomerular function in patients with type 2 diabetes

    International Nuclear Information System (INIS)

    Fu Hongliang; Li Jinsong; Li Jianing; Wu Jingchuan; Yang Shurong; Gu Zhenhui; Zou Renjian; Shi Haihong

    2003-01-01

    Objective: To assess glomerular filtration function in patients with type 2 diabetes by glomerular filtration rate (GFR) and split GFR, namely left GFR (LGFR) and right GFR (RGFR). Methods: Fifty-one patients with type 2 diabetes were classified by urine albumin analysis into three groups, normalalbuminuria group (NA), microalbuminuria group(MA) and macroalbuminuria group (MAA) . Twelve patients without diabetes were included into control group. 99 Tc m -DTPA renography was performed on all these cases. GFR and split GFR were calculated by Gates formula. Results: 1) GFR, LGFR and RGFR of NA group were lower than that of the control group. 2) GFR, LGFR and RGFR were significantly correlated with the urine albumin level (r=-0.457, -0.412, -0.424, respectively, P all < 0.01). 3) In all 51 cases, there were 5 cases whose GFR were normal while split GFR were abnormal. Conclusions: 1) GFR and split GFR measurement can detect the incipient damage of glomerular function more sensitively than urine albumin analysis and show the degree of the damage correctly. 2) Split GFR measurement can improve the evaluation of the glomerular function in type 2 diabetes patients

  2. Bifurcation of cubic nonlinear parallel plate-type structure in axial flow

    International Nuclear Information System (INIS)

    Lu Li; Yang Yiren

    2005-01-01

    The Hopf bifurcation of plate-type beams with cubic nonlinear stiffness in axial flow was studied. By assuming that all the plates have the same deflections at any instant, the nonlinear model of plate-type beam in axial flow was established. The partial differential equation was turned into an ordinary differential equation by using Galerkin method. A new algebraic criterion of Hopf bifurcation was utilized to in our analysis. The results show that there's no Hopf bifurcation for simply supported plate-type beams while the cantilevered plate-type beams has. At last, the analytic expression of critical flow velocity of cantilevered plate-type beams in axial flow and the purely imaginary eigenvalues of the corresponding linear system were gotten. (authors)

  3. Fine 3D neutronic characterization of a gas-cooled fast reactor based on plate-type sub-assemblies

    International Nuclear Information System (INIS)

    Bosq, J. C.; Peneliau, Y.; Rimpault, G.; Vanier, M.

    2006-01-01

    CEA neutronic studies have allowed the definition of a first 2400 MWth reference gas-cooled fast reactor core using plate-type sub-assemblies, for which the main neutronic characteristics were calculated by the so-called ERANOS 'design calculation scheme' relying on several method approximations. The last stage has consisted in a new refine characterization, using the reference calculation scheme, in order to confirm the impact of the approximations of the design route. A first core lay-out taking into account control rods was proposed and the reactivity penalty due to the control rod introduction in this hexagonal core lay-out was quantified. A new adjusted core was defined with an increase of the plutonium content. This leads to a significant decrease of the breeding gain which needs to be recovered in future design evolutions in order to achieve the self breeding goal. Finally, the safety criteria associated to the control rods were calculated with a first estimation of the uncertainties. All these criteria are respected, even if the safety analysis of GFR concepts and the determination of these uncertainties should be further studied and improved. (authors)

  4. Neutronic evaluation of transuranics in a GFR model using MCNPX and scale 6.0

    Energy Technology Data Exchange (ETDEWEB)

    Macedo, Anderson A.P.; Castro, Victor F.; Silva, Clarysson A.M. da; Velasquez, Carlos E.; Pereira, Claubia, E-mail: macedo@nuclear.ufmg.br, E-mail: victorfariacastro@gmail.com, E-mail: clarysson@nuclear.ufmg.br, E-mail: carlosvelcab@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    In this study, a GFR core model with 100 MWt was evaluated using three different fuel compositions: a conventional (U, Pu)C and two reprocessed fuels reprocessed by UREX+ technique one spiked with depleted uranium, (U,TRU)C, and the other one reprocessed spiked with thorium, (Th,TRU)C. The reprocessed fuel came from a PWR standard fuel (33,000 MWd/T burned) with 3.1% of initial enrichment and left in the pool by 5 years. Some important nuclides were followed during burnup and k{sub inf} was evaluated for 1400 days. The results also include analysis of the B4C insertion and the temperature coefficient. The simulations were performed comparing results between MCNPX and SCALE 6.0 codes. The main goal is to validate the model and evaluate the possibility to use TRU spiked with Th in a GFR. (author)

  5. Structural assessments of plate type support system for APR1400 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Anh Tung; Namgung, Ihn, E-mail: inamgung@kings.ac.kr

    2017-04-01

    Highlights: • This paper investigates plate-type support structure for the reactor vessel of the APR 1400. • The tall column supports of APR1400 reactor challenges in seismic and severe accident events. • A plate-type support of reactor vessel was proposed and evaluated based on ASME code. • The plate-type support was assessed to show its higher rigidity than column-type. - Abstract: This paper investigates an alternative form of support structure for the reactor vessel of the APR 1400. The current reactor vessel adopts a four-column support arrangement locating on the cold legs of the vessel. Although having been successfully designed, the tall column structure challenges in seismic events. In addition, for the mitigation of severe accident consequences, the columns inhibit ex-vessel coolant flow, hence the elimination of the support columns proposes extra safety advantages. A plate-type support was proposed and evaluated for the adequacy of meeting the structural stiffness by Finite Element Analysis (FEA) approach. ASME Boiler and Pressure Vessel Code was used to verify the design. The results, which cover thermal and static structural analysis, show stresses are within allowable limits in accordance with the design code. Even the heat conduction area is increased for the plate-type of support system, the results showed that the thermal stresses are within allowable limits. A comparison of natural frequencies and mode shapes for column support and plate-type support were presented as well which showed higher fundamental frequencies for the plate-type support system resulting in greater rigidity of the support system. From the outcome of this research, the plate-type support is proven to be an alternative to current APR column type support design.

  6. Prevalence of estimated GFR reporting among US clinical laboratories.

    Science.gov (United States)

    Accetta, Nancy A; Gladstone, Elisa H; DiSogra, Charles; Wright, Elizabeth C; Briggs, Michael; Narva, Andrew S

    2008-10-01

    Routine laboratory reporting of estimated glomerular filtration rate (eGFR) may help clinicians detect kidney disease. The current national prevalence of eGFR reporting in clinical laboratories is unknown; thus, the extent of the situation of laboratories not routinely reporting eGFR with serum creatinine results is not quantified. Observational analysis. National Kidney Disease Education Program survey of clinical laboratories conducted in 2006 to 2007 by mail, web, and telephone follow-up. A national random sample, 6,350 clinical laboratories, drawn from the Federal Clinical Laboratory Improvement Amendments database and stratified by 6 major laboratory types/groupings. Laboratory reports serum creatinine results. Reporting eGFR values with serum creatinine results. Percentage of laboratories reporting eGFR along with reporting serum creatinine values, reporting protocol, eGFR formula used, and style of reporting cutoff values. Of laboratories reporting serum creatinine values, 38.4% report eGFR (physician offices, 25.8%; hospitals, 43.6%; independents, 38.9%; community clinics, 47.2%; health fair/insurance/public health, 45.5%; and others, 43.2%). Physician office laboratories have a reporting prevalence lower than other laboratory types (P laboratories reporting eGFR, 66.7% do so routinely with all adult serum creatinine determinations; 71.6% use the 4-variable Modification of Diet in Renal Disease Study equation; and 45.3% use the ">60 mL/min/1.73 m(2)" reporting convention. Independent laboratories are least likely to routinely report eGFR (50.6%; P laboratories across all strata are more likely to report eGFR (P laboratories, federal database did not have names of laboratory directors/managers (intended respondents), assumed accuracy of federal database for sample purposes. Routine eGFR reporting with serum creatinine values is not yet universal, and laboratories vary in their reporting practices.

  7. How to use… serum creatinine, cystatin C and GFR.

    Science.gov (United States)

    Pasala, Swetha; Carmody, J Bryan

    2017-02-01

    Glomerular filtration rate (GFR) is the best overall measure of kidney function. The GFR is relatively low at birth but increases through infancy and early childhood to reach adult levels of approximately 120 mL/min/1.73 m 2 by age 2. While GFR can be measured most accurately by the urinary clearance of an exogenous ideal filtration marker such as inulin, it is more clinically useful to estimate GFR using a single serum measurement of an endogenous biomarker such as creatinine or cystatin C. When in steady state, there is an inverse relationship between creatinine/cystatin C and GFR, allowing GFR to be estimated from either using simple equations. Because of the non-linear relationship between creatinine/cystatin C and GFR, relatively small initial increases in these markers represent significant decreases in GFR. While cystatin C is produced by all nucleated cells, creatinine is a waste product of muscle metabolism and is therefore influenced by diet and muscle mass/body habitus. Decreased GFR is used to diagnose and stage chronic kidney disease (CKD) using the Kidney Disease: Improving Global Outcomes system. A diagnosis of CKD requires GFR creatinine and urine output are used to diagnose acute kidney injury. It is possible to calculate a kinetic GFR when the creatinine is changing rapidly, though more complex calculations are required. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/.

  8. Feasibility study on development of plate-type heat exchanger for BWR plants

    International Nuclear Information System (INIS)

    Ohyama, Nobuhiro; Suda, Kenichi; Ogata, Hiroshi; Matsuda, Shinichi; Nagasaka, Kazuhiro; Fujii, Toshi; Nozawa, Toshiya; Ishihama, Kiyoshi; Higuchi, Tomokazu

    2004-01-01

    In order to apply plate-type heat exchanger to RCW, TCW and FPC system in BWR plants, heat test and seismic test of RCW system heat exchanger sample were carried out. The results of these tests showed new design plate-type heat exchanger satisfied the fixed pressure resistance and seismic resistance and keep the function. The evaluation method of seismic design was constructed and confirmed by the results of tests. As anti-adhesion measure of marine organism, an ozone-water circulation method, chemical-feed method and combination of circulation of hot water and air bubbling are useful in place of the chlorine feeding method. Application of the plate-type heat exchanger to BWR plant is confirmed by these investigations. The basic principles, structure, characteristics, application limit and reliability are stated. (S.Y.)

  9. Creatinine-Based and Cystatin C-Based GFR Estimating Equations and Their Non-GFR Determinants in Kidney Transplant Recipients.

    Science.gov (United States)

    Keddis, Mira T; Amer, Hatem; Voskoboev, Nikolay; Kremers, Walter K; Rule, Andrew D; Lieske, John C

    2016-09-07

    eGFR equations have been evaluated in kidney transplant recipients with variable performance. We assessed the performance of the Modification of Diet in Renal Disease equation and the Chronic Kidney Disease Epidemiology Collaboration equations on the basis of creatinine, cystatin C, and both (eGFR creatinine-cystatin C) compared with measured GFR by iothalamate clearance and evaluated their non-GFR determinants and associations across 15 cardiovascular risk factors. A cross-sectional cohort of 1139 kidney transplant recipients >1 year after transplant was analyzed. eGFR bias, precision, and accuracy (percentage of estimates within 30% of measured GFR) were assessed. Interaction of each cardiovascular risk factor with eGFR relative to measured GFR was determined. Median measured GFR was 55.0 ml/min per 1.73 m(2). eGFR creatinine overestimated measured GFR by 3.1% (percentage of estimates within 30% of measured GFR of 80.4%), and eGFR Modification of Diet in Renal Disease underestimated measured GFR by 2.2% (percentage of estimates within 30% of measured GFR of 80.4%). eGFR cystatin C underestimated measured GFR by -13.7% (percentage of estimates within 30% of measured GFR of 77.1%), and eGFR creatinine-cystatin C underestimated measured GFR by -8.1% (percentage of estimates within 30% of measured GFR of 86.5%). Lower measured GFR associated with older age, women, obesity, longer time after transplant, lower HDL, lower hemoglobin, lower albumin, higher triglycerides, higher proteinuria, and an elevated cardiac troponin T level but did not associate with diabetes, smoking, cardiovascular events, pretransplant dialysis, or hemoglobin A1c. These risk factor associations differed for five risk factors with eGFR creatinine, six risk factors for eGFR Modification of Diet in Renal Disease, ten risk factors for eGFR cystatin C, and four risk factors for eGFR creatinine-cystatin C. Thus, eGFR creatinine and eGFR creatinine-cystatin C are preferred over eGFR cystatin C in

  10. Equivalent linearization method for limit cycle flutter analysis of plate-type structure in axial flow

    International Nuclear Information System (INIS)

    Lu Li; Yang Yiren

    2009-01-01

    The responses and limit cycle flutter of a plate-type structure with cubic stiffness in viscous flow were studied. The continuous system was dispersed by utilizing Galerkin Method. The equivalent linearization concept was performed to predict the ranges of limit cycle flutter velocities. The coupled map of flutter amplitude-equivalent linear stiffness-critical velocity was used to analyze the stability of limit cycle flutter. The theoretical results agree well with the results of numerical integration, which indicates that the equivalent linearization concept is available to the analysis of limit cycle flutter of plate-type structure. (authors)

  11. GFR meets mTOR: value of different methods to measure and estimate GFR & (side) effects of mTOR inhibition in renal transplantation

    NARCIS (Netherlands)

    Baas, M.C.

    2011-01-01

    The subject of this thesis is twofold: where GFR and mTOR meet. Precise measurement of kidney function is difficult and cumbersome and many, simpler alternatives have been developed to determine GFR. Determination of GFR remains an approximation since the GFR itself is not a static phenomenon. This

  12. Beam Pattern Analysis of the Plate-type Waveguide Sensor for Under-Sodium Viewing

    International Nuclear Information System (INIS)

    Kim, Hoewoong; Joo, Youngsang; Park, Changgyu; Kim, Jongbum

    2013-01-01

    Sensor for under-sodium viewing (USV) in a sodium-cooled fast reactor (SFR) has been developed. In the developed WG sensor approach, the A0 mode Lamb wave is used and a thin beryllium layer is coated on the waveguide surface to improve the ultrasonic radiation ability in a sodium environment. In this work, the beam pattern radiated from the developed plate-type WG sensor is investigated analytically to understand and predict the ultrasonic beam radiation property of the WG sensor in a liquid. Analytic calculations to obtain beam patterns for two kinds of WG sensors with and without beryllium coating layers were carried out and the results were compared with those obtained by experiments. In this work, the beam pattern of the plate-type WG sensor for USV was investigated analytically. Employing the far-field approximation, the acoustic response at a given measurement position was calculated for the plate-type WG sensors with and without beryllium coating layers. The beam patterns of WG sensors were predicted by the analytic calculation and the corresponding experiments were carried out. The results showed that the far-field beam pattern radiated from the plate-type WG sensor could be well predicted by an analytic calculation. The radiation beam angles obtained by the analytical calculation were in good agreement with those obtained by experiments

  13. Use of computed tomography assessed kidney length to predict split renal GFR in living kidney donors

    Energy Technology Data Exchange (ETDEWEB)

    Gaillard, Francois; Fournier, Catherine; Leon, Carine; Legendre, Christophe [Paris Descartes University, AP-HP, Hopital Necker-Enfants Malades, Renal Transplantation Department, Paris (France); Pavlov, Patrik [Linkoeping University, Linkoeping (Sweden); Tissier, Anne-Marie; Correas, Jean-Michel [Paris Descartes University, AP-HP, Hopital Necker-Enfants Malades, Radiology Department, Paris (France); Harache, Benoit; Hignette, Chantal; Weinmann, Pierre [Paris Descartes University, AP-HP, Hopital Europeen Georges Pompidou, Nuclear Medicine Department, Paris (France); Eladari, Dominique [Paris Descartes University, and INSERM, Unit 970, AP-HP, Hopital Europeen Georges Pompidou, Physiology Department, Paris (France); Timsit, Marc-Olivier; Mejean, Arnaud [Paris Descartes University, AP-HP, Hopital Europeen Georges Pompidou, Urology Department, Paris (France); Friedlander, Gerard; Courbebaisse, Marie [Paris Descartes University, and INSERM, Unit 1151, AP-HP, Hopital Europeen Georges Pompidou, Physiology Department, Paris (France); Houillier, Pascal [Paris Descartes University, INSERM, Unit umrs1138, and CNRS Unit erl8228, AP-HP, Hopital Europeen Georges Pompidou, Physiology Department, Paris (France)

    2017-02-15

    Screening of living kidney donors may require scintigraphy to split glomerular filtration rate (GFR). To determine the usefulness of computed tomography (CT) to split GFR, we compared scintigraphy-split GFR to CT-split GFR. We evaluated CT-split GFR as a screening test to detect scintigraphy-split GFR lower than 40 mL/min/1.73 m{sup 2}/kidney. This was a monocentric retrospective study on 346 potential living donors who had GFR measurement, renal scintigraphy, and CT. We predicted GFR for each kidney by splitting GFR using the following formula: Volume-split GFR for a given kidney = measured GFR*[volume of this kidney/(volume of this kidney + volume of the opposite kidney)]. The same formula was used for length-split GFR. We compared length- and volume-split GFR to scintigraphy-split GFR at donation and with a 4-year follow-up. A better correlation was observed between length-split GFR and scintigraphy-split GFR (r = 0.92) than between volume-split GFR and scintigraphy-split GFR (r = 0.89). A length-split GFR threshold of 45 mL/min/1.73 m{sup 2}/kidney had a sensitivity of 100 % and a specificity of 75 % to detect scintigraphy-split GFR less than 40 mL/min/1.73 m{sup 2}/kidney. Both techniques with their respective thresholds detected living donors with similar eGFR evolution during follow-up. Length-split GFR can be used to detect patients requiring scintigraphy. (orig.)

  14. Use of computed tomography assessed kidney length to predict split renal GFR in living kidney donors

    International Nuclear Information System (INIS)

    Gaillard, Francois; Fournier, Catherine; Leon, Carine; Legendre, Christophe; Pavlov, Patrik; Tissier, Anne-Marie; Correas, Jean-Michel; Harache, Benoit; Hignette, Chantal; Weinmann, Pierre; Eladari, Dominique; Timsit, Marc-Olivier; Mejean, Arnaud; Friedlander, Gerard; Courbebaisse, Marie; Houillier, Pascal

    2017-01-01

    Screening of living kidney donors may require scintigraphy to split glomerular filtration rate (GFR). To determine the usefulness of computed tomography (CT) to split GFR, we compared scintigraphy-split GFR to CT-split GFR. We evaluated CT-split GFR as a screening test to detect scintigraphy-split GFR lower than 40 mL/min/1.73 m"2/kidney. This was a monocentric retrospective study on 346 potential living donors who had GFR measurement, renal scintigraphy, and CT. We predicted GFR for each kidney by splitting GFR using the following formula: Volume-split GFR for a given kidney = measured GFR*[volume of this kidney/(volume of this kidney + volume of the opposite kidney)]. The same formula was used for length-split GFR. We compared length- and volume-split GFR to scintigraphy-split GFR at donation and with a 4-year follow-up. A better correlation was observed between length-split GFR and scintigraphy-split GFR (r = 0.92) than between volume-split GFR and scintigraphy-split GFR (r = 0.89). A length-split GFR threshold of 45 mL/min/1.73 m"2/kidney had a sensitivity of 100 % and a specificity of 75 % to detect scintigraphy-split GFR less than 40 mL/min/1.73 m"2/kidney. Both techniques with their respective thresholds detected living donors with similar eGFR evolution during follow-up. Length-split GFR can be used to detect patients requiring scintigraphy. (orig.)

  15. Global Cardiovascular and Renal Outcomes of Reduced GFR.

    Science.gov (United States)

    Thomas, Bernadette; Matsushita, Kunihiro; Abate, Kalkidan Hassen; Al-Aly, Ziyad; Ärnlöv, Johan; Asayama, Kei; Atkins, Robert; Badawi, Alaa; Ballew, Shoshana H; Banerjee, Amitava; Barregård, Lars; Barrett-Connor, Elizabeth; Basu, Sanjay; Bello, Aminu K; Bensenor, Isabela; Bergstrom, Jaclyn; Bikbov, Boris; Blosser, Christopher; Brenner, Hermann; Carrero, Juan-Jesus; Chadban, Steve; Cirillo, Massimo; Cortinovis, Monica; Courville, Karen; Dandona, Lalit; Dandona, Rakhi; Estep, Kara; Fernandes, João; Fischer, Florian; Fox, Caroline; Gansevoort, Ron T; Gona, Philimon N; Gutierrez, Orlando M; Hamidi, Samer; Hanson, Sarah Wulf; Himmelfarb, Jonathan; Jassal, Simerjot K; Jee, Sun Ha; Jha, Vivekanand; Jimenez-Corona, Aida; Jonas, Jost B; Kengne, Andre Pascal; Khader, Yousef; Khang, Young-Ho; Kim, Yun Jin; Klein, Barbara; Klein, Ronald; Kokubo, Yoshihiro; Kolte, Dhaval; Lee, Kristine; Levey, Andrew S; Li, Yongmei; Lotufo, Paulo; El Razek, Hassan Magdy Abd; Mendoza, Walter; Metoki, Hirohito; Mok, Yejin; Muraki, Isao; Muntner, Paul M; Noda, Hiroyuki; Ohkubo, Takayoshi; Ortiz, Alberto; Perico, Norberto; Polkinghorne, Kevan; Al-Radaddi, Rajaa; Remuzzi, Giuseppe; Roth, Gregory; Rothenbacher, Dietrich; Satoh, Michihiro; Saum, Kai-Uwe; Sawhney, Monika; Schöttker, Ben; Shankar, Anoop; Shlipak, Michael; Silva, Diego Augusto Santos; Toyoshima, Hideaki; Ukwaja, Kingsley; Umesawa, Mitsumasa; Vollset, Stein Emil; Warnock, David G; Werdecker, Andrea; Yamagishi, Kazumasa; Yano, Yuichiro; Yonemoto, Naohiro; Zaki, Maysaa El Sayed; Naghavi, Mohsen; Forouzanfar, Mohammad H; Murray, Christopher J L; Coresh, Josef; Vos, Theo

    2017-07-01

    The burden of premature death and health loss from ESRD is well described. Less is known regarding the burden of cardiovascular disease attributable to reduced GFR. We estimated the prevalence of reduced GFR categories 3, 4, and 5 (not on RRT) for 188 countries at six time points from 1990 to 2013. Relative risks of cardiovascular outcomes by three categories of reduced GFR were calculated by pooled random effects meta-analysis. Results are presented as deaths for outcomes of cardiovascular disease and ESRD and as disability-adjusted life years for outcomes of cardiovascular disease, GFR categories 3, 4, and 5, and ESRD. In 2013, reduced GFR was associated with 4% of deaths worldwide, or 2.2 million deaths (95% uncertainty interval [95% UI], 2.0 to 2.4 million). More than half of these attributable deaths were cardiovascular deaths (1.2 million; 95% UI, 1.1 to 1.4 million), whereas 0.96 million (95% UI, 0.81 to 1.0 million) were ESRD-related deaths. Compared with metabolic risk factors, reduced GFR ranked below high systolic BP, high body mass index, and high fasting plasma glucose, and similarly with high total cholesterol as a risk factor for disability-adjusted life years in both developed and developing world regions. In conclusion, by 2013, cardiovascular deaths attributed to reduced GFR outnumbered ESRD deaths throughout the world. Studies are needed to evaluate the benefit of early detection of CKD and treatment to decrease these deaths. Copyright © 2017 by the American Society of Nephrology.

  16. Bi-layer plate-type acoustic metamaterials with Willis coupling

    Science.gov (United States)

    Ma, Fuyin; Huang, Meng; Xu, Yicai; Wu, Jiu Hui

    2018-01-01

    Dynamic effective negative parameters are principal to the representation of the physical properties of metamaterials. In this paper, a bi-layer plate-type unit was proposed with both a negative mass density and a negative bulk modulus; moreover, through analysis of these bi-layer structures, some important problems about acoustic metamaterials were studied. First, dynamic effective mass densities and the bulk modulus of the bi-layer plate-type acoustic structure were clarified through both the direct and the retrieval methods, and, in addition, the intrinsic relationship between the sound transmission (absorption) characteristics and the effective parameters was analyzed. Furthermore, the properties of dynamic effective parameters for an asymmetric bi-layer acoustic structure were further considered through an analysis of experimental data, and the modified effective parameters were then obtained through consideration of the Willis coupling in the asymmetric passive system. In addition, by taking both the clamped and the periodic boundary conditions into consideration in the bi-layer plate-type acoustic system, new perspectives were presented for study on the effective parameters and sound insulation properties in the range below the cut-off frequency. The special acoustic properties established by these effective parameters could enrich our knowledge and provide guidance for the design and installation of acoustic metamaterial structures in future sound engineering practice.

  17. STUDI KOORDINASI PERALATAN PROTEKSI OCR DAN GFR PADA PENYULANG TIBUBENENG

    Directory of Open Access Journals (Sweden)

    Indra Baskara

    2015-12-01

    Full Text Available Tibubeneng feeder equipped with protective devices over current relay (OCR and ground fault relay (GFR installed in Recloser Dama, Recloser Tandeg and relay feeder Tibubeneng in GI. Based on data from PLN Bali, there are 3 times the interference with the feeder Tibubeneng that cause coordination protection error system. The problem of protection error coordination can be addressed by the study coordinated analysis of protection systems. The analysis was performed by making a curve existing coordination OCR settings and GFR in Tibubeneng feeders and comparing it with the coordination curve setting calculation results. The calculation is performed based on the standards system of protection of sensitive, reliable, fast and remain selective. Based on the analysis of the existing curve setting, an error occurred coordination between the curve and the GFR Recloser Recloser Tandeg Dama indicated by the curves that intersect at several levels of short circuit current value and the value of grading time OCR and GFR was less than 0.4 seconds. Repair work coordination OCR and GFR in Tibubeneng feeder can be done by re-setting in accordance with the results of calculations in order to obtain protection system in accordance with the terms of the security system

  18. Pseudo-cubic thin-plate type Spline method for analyzing experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Crecy, F de

    1994-12-31

    A mathematical tool, using pseudo-cubic thin-plate type Spline, has been developed for analysis of experimental data points. The main purpose is to obtain, without any a priori given model, a mathematical predictor with related uncertainties, usable at any point in the multidimensional parameter space. The smoothing parameter is determined by a generalized cross validation method. The residual standard deviation obtained is significantly smaller than that of a least square regression. An example of use is given with critical heat flux data, showing a significant decrease of the conception criterion (minimum allowable value of the DNB ratio). (author) 4 figs., 1 tab., 7 refs.

  19. Pseudo-cubic thin-plate type Spline method for analyzing experimental data

    International Nuclear Information System (INIS)

    Crecy, F. de.

    1993-01-01

    A mathematical tool, using pseudo-cubic thin-plate type Spline, has been developed for analysis of experimental data points. The main purpose is to obtain, without any a priori given model, a mathematical predictor with related uncertainties, usable at any point in the multidimensional parameter space. The smoothing parameter is determined by a generalized cross validation method. The residual standard deviation obtained is significantly smaller than that of a least square regression. An example of use is given with critical heat flux data, showing a significant decrease of the conception criterion (minimum allowable value of the DNB ratio). (author) 4 figs., 1 tab., 7 refs

  20. Investigation of plate-type barrier ozonizers with AC and pulse power supplies

    International Nuclear Information System (INIS)

    Krasnij, V.V.; Gubarev, S.P.; Pogoghev, D.P.; Sokolova, O.T.

    2002-01-01

    In this paper the experimental results on the investigation of plate-type reactors operated on the base of barrier discharge have been presented. Different reactors with planar, strip, and trench electrodes were investigated. Such reactors operated under atmospheric pressure with ac and pulse power sources with voltage of up to 10 kV, frequency up to 12 kHz. Using atomized spectroscopy system the measurements of the main specifications of the reactors such as ozone yielding rate, the temperature in the reactor and the air flow rate were carried out

  1. GFR Prediction From Cystatin C and Creatinine in Children

    DEFF Research Database (Denmark)

    Andersen, Trine Borup; Jødal, Lars; Boegsted, Martin

    2012-01-01

    area, and SCr is serum creatinine level. The accuracy and precision of these models were compared with 7 previously published prediction models using random subsampling cross-validation. Local constants and coefficients were calculated for all models. Root mean square error, R2, and percentage...... are still not sufficiently accurate to replace exogenous markers when GFR must be determined with high accuracy....

  2. The MDRD formula does not reflect GFR in ESRD patients

    NARCIS (Netherlands)

    Grootendorst, Diana C.; Michels, Wieneke M.; Richardson, Jermaine D.; Jager, Kitty J.; Boeschoten, Elisabeth W.; Dekker, Friedo W.; Krediet, Raymond T.; Apperloo, A. J.; Bijlsma, J. A.; Boekhout, M.; Boer, W. H.; van der Boog, P. J. M.; Büller, H. R.; van Buren, M.; de Charro, F. Th; Doorenbos, C. J.; van den Dorpel, M. A.; van Es, A.; Fagel, W. J.; Feith, G. W.; de Fijter, C. W. H.; Frenken, L. A. M.; van Geelen, J. A. C. A.; Gerlag, P. G. G.; Gorgels, J. P. M. C.; Grave, W.; Huisman, R. M.; Jie, K.; Koning-Mulder, W. A. H.; Koolen, M. I.; Kremer Hovinga, T. K.; Lavrijssen, A. T. J.; Luik, A. J.; van der Meulen, J.; Parlevliet, K. J.; Raasveld, M. H. M.; van der Sande, F. M.; Schonck, M. J. M.; Schuurmans, M. M. J.; Siegert, C. E. H.; Stegeman, C. A.; Stevens, P.; Thijssen, J. G. P.; Valentijn, R. M.; Vastenburg, G. H.; Verburgh, C. A.; Vincent, H. H.; Vos, P. F.

    2011-01-01

    The Modification of Diet in Renal Disease (MDRD) equation is widely used for the estimation of glomerular filtration rate (GFR) from plasma creatinine. It has been well validated in patients with various degrees of impaired kidney function, but not in patients with end-stage renal disease (ESRD).

  3. Neutronics comparative analysis of plate-type research reactor using deterministic and stochastic methods

    International Nuclear Information System (INIS)

    Liu, Shichang; Wang, Guanbo; Wu, Gaochen; Wang, Kan

    2015-01-01

    Highlights: • DRAGON and DONJON are applied and verified in calculations of research reactors. • Continuous-energy Monte Carlo calculations by RMC are chosen as the references. • “ECCO” option of DRAGON is suitable for the calculations of research reactors. • Manual modifications of cross-sections are not necessary with DRAGON and DONJON. • DRAGON and DONJON agree well with RMC if appropriate treatments are applied. - Abstract: Simulation of the behavior of the plate-type research reactors such as JRR-3M and CARR poses a challenge for traditional neutronics calculation tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity and large leakage of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON and DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic approach. The goal of this research is to examine the capability of the deterministic code system DRAGON and DONJON to reliably simulate the research reactors. The results indicate that the DRAGON and DONJON code system agrees well with the continuous-energy Monte Carlo simulation on both k eff and flux distributions if the appropriate treatments (such as the ECCO option) are applied

  4. Hydrogen Recombination Rates of Plate-type Passive Auto-catalytic Recombiner

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jongtae; Hong, Seong-Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Gun Hong [Kyungwon E-C Co., Seongnam (Korea, Republic of)

    2014-10-15

    The hydrogen mitigation system may include igniters, passive autocatalytic recombiner (PAR), and venting or dilution system. Recently PAR is commonly used as a main component of HMS in a NPP containment because of its passive nature. PARs are categorized by the shape and material of catalytic surface. Catalytic surface coated by platinum is mostly used for the hydrogen recombiners. The shapes of the catalytic surface can be grouped into plate type, honeycomb type and porous media type. Among them, the plate-type PAR is well tested by many experiments. PAR performance analysis can be approached by a multi-scale method which is composed of micro, meso and macro scales. The criterion of the scaling is the ratio of thickness of boundary layer developed on a catalytic surface to representative length of a computational domain. Mass diffusion in the boundary layer must be resolved in the micro scale analysis. In a lumped parameter (LP) analysis using a system code such as MAAP or MELCOR, the chamber of the PAR is much smaller than a computational node. The hydrogen depletion by a PAR is modeled as a source of mass and energy conservation equations. Te catalytic surface reaction of hydrogen must be modeled by a volume-averaged correlation. In this study, a micro scale analysis method is developed using libraries in OpenFOAM to evaluate a hydrogen depletion rate depending on parameters such as size and number of plates and plate arrangement. The analysis code is validated by simulating REKO-3 experiment. And hydrogen depletion analysis is conducted by changing the plate arrangement as a trial of the performance enhancement of a PAR. In this study, a numerical code for an analysis of a PAR performance in a micro scale has been developed by using OpenFOAM libraries. The physical and numerical models were validated by simulating the REKO-3 experiment. As a try to enhance the performance of the plate-type PAR, it was proposed to apply a staggered two-layer arrangement of the

  5. Hydrogen Recombination Rates of Plate-type Passive Auto-catalytic Recombiner

    International Nuclear Information System (INIS)

    Kim, Jongtae; Hong, Seong-Wan; Kim, Gun Hong

    2014-01-01

    The hydrogen mitigation system may include igniters, passive autocatalytic recombiner (PAR), and venting or dilution system. Recently PAR is commonly used as a main component of HMS in a NPP containment because of its passive nature. PARs are categorized by the shape and material of catalytic surface. Catalytic surface coated by platinum is mostly used for the hydrogen recombiners. The shapes of the catalytic surface can be grouped into plate type, honeycomb type and porous media type. Among them, the plate-type PAR is well tested by many experiments. PAR performance analysis can be approached by a multi-scale method which is composed of micro, meso and macro scales. The criterion of the scaling is the ratio of thickness of boundary layer developed on a catalytic surface to representative length of a computational domain. Mass diffusion in the boundary layer must be resolved in the micro scale analysis. In a lumped parameter (LP) analysis using a system code such as MAAP or MELCOR, the chamber of the PAR is much smaller than a computational node. The hydrogen depletion by a PAR is modeled as a source of mass and energy conservation equations. Te catalytic surface reaction of hydrogen must be modeled by a volume-averaged correlation. In this study, a micro scale analysis method is developed using libraries in OpenFOAM to evaluate a hydrogen depletion rate depending on parameters such as size and number of plates and plate arrangement. The analysis code is validated by simulating REKO-3 experiment. And hydrogen depletion analysis is conducted by changing the plate arrangement as a trial of the performance enhancement of a PAR. In this study, a numerical code for an analysis of a PAR performance in a micro scale has been developed by using OpenFOAM libraries. The physical and numerical models were validated by simulating the REKO-3 experiment. As a try to enhance the performance of the plate-type PAR, it was proposed to apply a staggered two-layer arrangement of the

  6. Experimental study on the heat transfer characteristics in corrugated and flat plate type heat exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jung Hun; Jeong, Yong Ki; Jeon, Chung Hwan; Chang, Young June [Busan National Univ., Busan (Korea, Republic of); Lim, Hyeok [DHT, Busan (Korea, Republic of)

    2003-07-01

    An experiment was performed to study heat transfer characteristics between corrugated heat exchanger and flat plate type one. While heat capacity(13.86kW) was provided constantly and the flow speed was varied from 2.8 to 17.9m/s, the temperature and the pressure drop were measured. Furthermore, heat transfer coefficient, Colburn factor and Nusselt number were calculated using them. With increase of the flow speed for both exchangers, the coefficient and the pressure drop increased, but Colburn factor decreased. The coefficient, pressure drop and Colburn factor of the corrugated type were all higher than those of the flat one, which is due to the flow interruption with recirculation and reattachment of the corrugated type. The empirical correlations of Nusselt number were suggested for the tested two heat exchangers.

  7. On the use of plate-type normal pressure cells in silos

    DEFF Research Database (Denmark)

    Ramirez, Alvaro; Nielsen, Jørgen; Ayuga, F.

    2010-01-01

    the interpretation of results. Once the cells have been delivered from the manufacturer to the researcher, they should be calibrated and validated with reference to the measurement of pressure from a granular material against a silo wall. Two related papers deal with a specific plate-type normal pressure cell...... for use in an installation of three full-scale steel silos with different hopper eccentricities (concentric, half-eccentric and full-eccentric) as part of a silo research project. It was found to be necessary to validate the performance of the cells when measuring pressures in the silos in order to arrive...... at a solid basis for the interpretation of the pressure measurements in the silo installation aforementioned. This paper presents calibration results from three investigated methods as well as results from a finite element analysis of the plate deflection of the pressure cell which were performed to evaluate...

  8. Thermal performance of plate-type loop thermosyphon at sub-atmospheric pressures

    International Nuclear Information System (INIS)

    Tsoi, Vadim; Chang, Shyy Woei; Chiang Kuei Feng; Huang, Chuan Chin

    2011-01-01

    This experimental study examines the thermal performance of a newly devised plate-type two-phase loop thermosyphon with cooling applications to electronic boards of telecommunication systems. The evaporation section is configured as the inter-connected multi channels to emulate the bridging boiling mechanism in pulsating thermosyphon. Two thermosyphon plates using water as the coolant with filling ratios (FR) of 0.22 and 0.32 are tested at sub-atmospheric pressures. The vapor-liquid flow images as well as the thermal resistances and effective spreading thermal conductivities are individually measured for each thermosyphon test plate at various heating powers. The high-speed digital images of the vapor-liquid flow structures reveal the characteristic boiling phenomena and the vapor-liquid circulation in the vertical thermosyphon plate, which assist to explore the thermal physics for this type of loop thermosyphon. The bubble agglomeration and pumping action in the inter-connected boiling channels take place at metastable non-equilibrium conditions, leading to the intermittent slug flows with a pulsation character. Such hybrid loop-pulsating thermosyphon permits the vapor-liquid circulation in the horizontal plate. Thermal resistances and spreading thermal conductivities detected from the present thermosyphon plates; the vapor chamber flat plate heat pipe and the copper plate at free and forced convective cooling conditions with both vertical and horizontal orientations are cross-examined. In most telecommunication systems and units, the electrical boards are vertical so that the thermal performance data on the vertical thermosyphon are most relevant to this particular application. - Highlights: → We examine thermal performances of plate-type loop thermosyphon. → Thermal resistances and spreading conductivities are examined. → Bubble agglomeration in inter-connected boiling channels generates intermittent slug flows with pulsations. → Boiling instability

  9. Development and application of an advanced fuel model for the safety analysis of the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Petkevich, P.

    2008-09-01

    . Within the FAST system, FRED is coupled to the TRACE code for the thermal-hydraulic modeling, so that the present work has comprised not only the development of a 2D FRED model for the plate-type GFR fuel, but also the implementation of corresponding changes in TRACE for ensuring appropriate information exchange between the two codes. The 2D thermo-mechanical model has been developed with certain assumptions. Since no experimental data exist for this fuel type, benchmarking of the new simulation tool was carried out by building up a detailed 3D model using the finite-elements code ANSYS. The 3D model has, moreover, been employed for conducting certain supplementary studies to obtain an in-depth understanding of the thermal and mechanical behavior of the fuel. It was found how the complex, multi-dimensional, heat transfer in the plate-type fuel accounts for the discrepancies between results of 3 2D and 1D simulations. Furthermore, it was shown that, under certain conditions, the temperature field can be well predicted by the 1D model with slight modifications of the solution algorithm. Other insights have been obtained from the detailed mechanical analysis. Thus, it has been shown that, during operation, cusping occurs at the pellet periphery which results in an unfavorable concentration of stresses both in pellet and cladding. Several alternative ways to optimize the fuel design and to avoid, or at least minimize, this effect have been proposed. As mentioned, the new fuel model is intended for usage in GFR transient analysis. In order to quantify the impact of the current model developments, a range of hypothetical accident events have been analyzed using the FAST code system, with and without usage of the new fuel model. It has been shown that the pure geometry effects on the temperatures are quite significant. However, for the specific honeycomb structure geometry considered, these are somewhat mitigated by the fuel and cladding expansion and the corresponding decrease

  10. Past Decline Versus Current eGFR and Subsequent Mortality Risk

    NARCIS (Netherlands)

    Naimark, David M. J.; Grams, Morgan E.; Matsushita, Kunihiro; Black, Corri; Drion, Iefke; Fox, Caroline S.; Inker, Lesley A.; Ishani, Areef; Jee, Sun Ha; Kitamura, Akihiko; Lea, Janice P.; Nally, Joseph; Peralta, Carmen Alicia; Rothenbacher, Dietrich; Ryu, Seungho; Tonelli, Marcello; Yatsuya, Hiroshi; Coresh, Josef; Gansevoort, Ron T.; Warnock, David G.; Woodward, Mark; de Jong, Paul E.

    A single determination of eGFR associates with subsequent mortality risk. Prior decline in eGFR indicates loss of kidney function, but the relationship to mortality risk is uncertain. We conducted an individual-level meta-analysis of the risk of mortality associated with antecedent eGFR slope,

  11. COOLOD-N: a computer code, for the analyses of steady-state thermal-hydraulics in plate-type research reactors

    International Nuclear Information System (INIS)

    Kaminaga, Masanori

    1990-02-01

    The COOLOD-N code provides a capability for the analysis of the steady-state thermal-hydraulics of research reactors in which plate-type fuel is employed. This code is revised version of the COOLOD code, and is applicable not only to a forced convection cooling mode, but also to a natural convection cooling mode. In the code, a function to calculate flow rate under a natural convection, and a heat transfer package which was a subroutine program to calculate heat transfer coefficient, ONB temperature and DNB heat flux, and was especially developed for the upgraded JRR-3, have been newly added to the COOLOD code. The COOLOD-N code also has a capability of calculating the heat flux at onset of flow instability as well as DNB heat flux. (author)

  12. Plate-type metamaterials for extremely broadband low-frequency sound insulation

    Science.gov (United States)

    Wang, Xiaopeng; Guo, Xinwei; Chen, Tianning; Yao, Ge

    2018-01-01

    A novel plate-type acoustic metamaterial with a high sound transmission loss (STL) in the low-frequency range ( ≤1000 Hz) is designed, theoretically proven and then experimentally verified. The thin plates with large modulus used in this paper mean that we do not need to apply tension to the plates, which is more applicable to practical engineering, the achievement of noise reduction is better and the installation of plates is more user-friendly than that of the membranes. The effects of different structural parameters of the plates on the sound-proofed performance at low-frequencies were also investigated by experiment and finite element method (FEM). The results showed that the STL can be modulated effectively and predictably using vibration theory by changing the structural parameters, such as the radius and thickness of the plate. Furthermore, using unit cells of different geometric sizes which are responsible for different frequency regions, the stacked panels with thickness ≤16 mm and weight ≤5 kg/m2 showed high STL below 2000 Hz. The acoustic metamaterial proposed in this study could provide a potential application in the low-frequency noise insulation.

  13. Comparison of creatinine and cystatin C based eGFR in the estimation of glomerular filtration rate in Indigenous Australians: The eGFR Study.

    Science.gov (United States)

    Barr, Elizabeth Lm; Maple-Brown, Louise J; Barzi, Federica; Hughes, Jaquelyne T; Jerums, George; Ekinci, Elif I; Ellis, Andrew G; Jones, Graham Rd; Lawton, Paul D; Sajiv, Cherian; Majoni, Sandawana W; Brown, Alex Dh; Hoy, Wendy E; O'Dea, Kerin; Cass, Alan; MacIsaac, Richard J

    2017-04-01

    The Chronic Kidney Disease Epidemiology Collaboration (CKD-EPI) equation that combines creatinine and cystatin C is superior to equations that include either measure alone in estimating glomerular filtration rate (GFR). However, whether cystatin C can provide any additional benefits in estimating GFR for Indigenous Australians, a population at high risk of end-stage kidney disease (ESKD) is unknown. Using a cross-sectional analysis from the eGFR Study of 654 Indigenous Australians at high risk of ESKD, eGFR was calculated using the CKD-EPI equations for serum creatinine (eGFRcr), cystatin C (eGFRcysC) and combined creatinine and cystatin C (eGFRcysC+cr). Reference GFR (mGFR) was determined using a non-isotopic iohexol plasma disappearance technique over 4h. Performance of each equation to mGFR was assessed by calculating bias, % bias, precision and accuracy for the total population, and according to age, sex, kidney disease, diabetes, obesity and c-reactive protein. Data were available for 542 participants (38% men, mean [sd] age 45 [14] years). Bias was significantly greater for eGFRcysC (15.0mL/min/1.73m 2 ; 95% CI 13.3-16.4, pcreatinine remains the preferred equation in Indigenous Australians. Copyright © 2016 The Canadian Society of Clinical Chemists. Published by Elsevier Inc. All rights reserved.

  14. Longitudinal Estimated GFR Trajectories in Patients With and Without Type 2 Diabetes and Nephropathy

    DEFF Research Database (Denmark)

    Weldegiorgis, Misghina; de Zeeuw, Dick; Li, Liang

    2017-01-01

    -renin-angiotensin-aldosterone-system antihypertensives were independently associated with a greater probability of a nonlinear eGFR trajectory. LIMITATIONS: Relatively short follow-up and no measured GFR. CONCLUSIONS: In both diabetes and nondiabetes trials, the majority of patients show a more or less linear eGFR decline. These data support...... with and without diabetes. STUDY DESIGN: Longitudinal observational study. SETTING & PARTICIPANTS: 6 clinical trials with repeated measurements of serum creatinine. PREDICTOR: Patient demographic and clinical parameters. OUTCOMES: Probability of nonlinear eGFR function trajectory calculated for each patient from...... a Bayesian model of individual eGFR trajectories. RESULTS: The median probability of a nonlinear eGFR decline in all trials was 0.26 (interquartile range, 0.13-0.48). The median probability was 0.28 in diabetes versus 0.09 in nondiabetes trials (P50% probability...

  15. Optimized, Competitive Supercritical-CO2 Cycle GFR for Gen IV Service

    International Nuclear Information System (INIS)

    M.J. Driscoll; P. Hejzlar; G. Apostolakis

    2008-01-01

    An overall plant design was developed for a gas-cooled fast reactor employing a direct supercritical Brayton power conversion system. The most important findings were that (1) the concept could be capital-cost competitive, but startup fuel cycle costs are penalized by the low core power density, specified in large part to satisfy the goal of significant post-accident passive natural convection cooling; (2) active decay heat removal is preferable as the first line of defense, with passive performance in a backup role; (3) an innovative tube-in-duct fuel assembly, vented to the primary coolant, appears to be practicable; and (4) use of the S-Co2 GFR to support hydrogen production is a synergistic application, since sufficient energy can be recuperated from the product H2 and 02 to allow the electrolysis cell to run 250 C hotter than the reactor coolant, and the water boilers can be used for reactor decay heat removal. Increasing core power density is identified as the top priority for future work on GFRs of this type

  16. Synthesis of the safety studies carried out on the GFR2400

    Energy Technology Data Exchange (ETDEWEB)

    Bertrand, F., E-mail: frederic.bertrand@cea.fr [CEA, DEN, DER, F-13108, Saint Paul-lez-Durance (France); Bassi, C. [CEA, DEN, DER, F-13108, Saint Paul-lez-Durance (France); Bentivoglio, F. [CEA, DEN, DM2S, F-38054, Grenoble (France); Audubert, F. [CEA, DEN, DEC, F-13108, Saint Paul-lez-Durance (France); Gueneau, C. [CEA, DEN, DPC, F-91191, Gif-sur-yvette (France); Rimpault, G. [CEA, DEN, DER, F-13108, Saint Paul-lez-Durance (France); Journeau, C. [CEA, DEN, DTN, F-13108, Saint Paul-lez-Durance (France)

    2012-12-15

    been preliminarily shown in several particularly challenging situations (loss of active means, unprotected transients, full depressurization). Finally, preliminary results regarding analytical studies carried out on phenomena involved in GFR2400 core degradation (physico-chemistry and neutron physics) are presented. Then, the application of the separate results aforementioned by considering results of analytical simplified thermalhydraulic calculations and of system calculations (carried out with the CATHARE2 code) have enabled a preliminary assessment of GFR2400 behaviour in case of core degradation. For some cases, such applications permitted to conclude on the problematic/begnin issue of a phenomenon (like air ingress in realistic scenarios) whereas in other cases, those applications have illustrated that more complex calculation tools coupling the various phenomena are necessary (like effects of water ingress for instance) as well as semi-integral experiments reproducing a fuel assembly degradation.

  17. GFR and Blood Lead Levels in Gas Station Workers Based on δ-Alad Gene Polymorphisms

    Directory of Open Access Journals (Sweden)

    Lantip Rujito

    2015-04-01

    showed that the proportion of ALAD genotype for ALAD 1-1, 1-2 and 2-2 were 94.7%, 5.3%, and 0% respectively. The mean of serum levels in homozygous 1-1 was 15.94 ppb and heterozygote 1-2 was 1.15 ppb. GFR of participants ranged from 71.11 mL/min to 185.20 mL/min with a mean of 117.34mL/min. There was no correlation between serum Pb and GFR (p = 0.19. Study also could not determine the correlation between GFR and ALAD gene Polymorphism. Discussion: Study then concluded that there was no correlation between blood lead levels in the GFR on each δ-ALAD genotypes. Keywords: Lead intoxication, GFR, δ-ALAD, gas station workers

  18. Nuclear fuels

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  19. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  20. Burnup-dependent core neutronics analysis of plate-type research reactor using deterministic and stochastic methods

    International Nuclear Information System (INIS)

    Liu, Shichang; Wang, Guanbo; Liang, Jingang; Wu, Gaochen; Wang, Kan

    2015-01-01

    Highlights: • DRAGON & DONJON were applied in burnup calculations of plate-type research reactors. • Continuous-energy Monte Carlo burnup calculations by RMC were chosen as references. • Comparisons of keff, isotopic densities and power distribution were performed. • Reasons leading to discrepancies between two different approaches were analyzed. • DRAGON & DONJON is capable of burnup calculations with appropriate treatments. - Abstract: The burnup-dependent core neutronics analysis of the plate-type research reactors such as JRR-3M poses a challenge for traditional neutronics calculational tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity, large leakage and the particular neutron spectrum of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the burnup-dependent core neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON & DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic one. In the first stage, the homogenizations of few-group cross sections by DRAGON and the full core diffusion calculations by DONJON have been verified by comparing with the detailed Monte Carlo simulations. In the second stage, the burnup-dependent calculations of both assembly level and the full core level were carried out, to examine the capability of the deterministic code system DRAGON & DONJON to reliably simulate the burnup-dependent behavior of research reactors. The results indicate that both RMC and DRAGON & DONJON code system are capable of burnup-dependent neutronics analysis of research reactors, provided that appropriate treatments are applied in both assembly and core levels for the deterministic codes

  1. Xenon thermal behavior in sintered titanium nitride, foreseen inert matrix for GFR

    International Nuclear Information System (INIS)

    Bes, R.

    2010-11-01

    This work concerns the generation IV future nuclear reactors such as gas-cooled fast reactor (GFR) for which refractory materials as titanium nitride (TiN) are needed to surround fuel and act as a fission product diffusion barrier. This study is about Xe thermal behavior in sintered titanium nitride. Microstructure effects on Xe behavior have been studied. In this purpose, several syntheses have been performed using different sintering temperatures and initial powder compositions. Xenon species have been introduced into samples by ionic implantation. Then, samples were annealed in temperature range from 1300 C to 1600 C, these temperatures being the accidental awaited temperature. A transport of xenon towards sample surface has been observed. Transport rate seems to be slow down when increasing sintering temperature. The composition of initial powder and the crystallographic orientation of each considered grain also influence xenon thermal behavior. Xenon release has been correlated with material oxidation during annealing. Xenon bubbles were observed. Their size is proportional with xenon concentration and increases with annealing temperature. Several mechanisms which could explain Xe intragranular mobility in TiN are proposed. In addition with experiments, very low Xe solubility in TiN has been confirmed by ab initio calculations. So, bi-vacancies were found to be the most favoured Xe incorporation sites in this material. (author)

  2. [New topics regarding equations for GFR estimation based on serum creatinine and cystatin C].

    Science.gov (United States)

    Horio, Masaru

    2014-02-01

    Japanese GFR equations and CKD-EPI equations based on standardized serum creatinine and standardized cystatin C are recommended in recent Japanese CKD guides and KDIGO guidelines for CKD management, respectively. CKD-EPIcreat overestimates GFR in Japanese subjects, probably due to the difference in muscle mass between Japanese and Caucasians. Unlike CKD-EPIcreat, CKD-EPIcys performs well in Japanese subjects, indicating the advantages of using cystatin C as a GFR marker. KDIGO guidelines suggest measuring eGFRcys in adults with eGFRcreat of 45-59 ml/min/1.73 m2 who do not have markers of kidney damage if confirmation of CKD is required. Creatinine is excreted by glomerular filtration, but also secreted by the tubules. Alteration of the tubular secretion of creatinine may influence the performance of GFR equations based on serum creatinine. Multivariate analysis showed that GFR and serum albumin levels were independent parameters affecting the fractional excretion of creatinine (FE-Cr). Alteration of FE-Cr according to the serum albumin levels may be one of the reasons for the bias of GFR equations based on serum creatinine. Low GFR is a risk factor for all-cause and cardiovascular mortality in a general population. However, the relationship between eGFR and the hazard risk of events is different depending on whether cystatin C or creatinine is used to calculate eGFR. The association between eGFRcys and the hazard risk is much stronger compared with eGFRcreat. Cystatin C may be a useful alternative to creatinine for detecting a high risk of complications in a general population and subjects with CKD.

  3. Fluctuations in eGFR in relation to unenhanced and enhanced MRI and CT outpatients

    DEFF Research Database (Denmark)

    Azzouz, Manal; Rømsing, Janne; Thomsen, Henrik S

    2014-01-01

    OBJECTIVE: To study fluctuations in estimated glomerular filtration rate (eGFR) in relation to contrast medium (CM) enhanced magnetic resonance imaging (MRI) and computed tomography (CT) compared to control groups in outpatients. MATERIALS AND METHODS: eGFR was determined right before the imaging......-induced nephropathy (CIN) requirement when the definition s-creatinine ≥44μmol/l (0.5mg/dl) was used. CONCLUSIONS: eGFR in outpatients undergoing MRI or CT did vary independently of whether the patient received contrast or not. The findings probably reflect the natural variations in s-creatinine levels. This should...

  4. Assessing glomerular filtration rate (GFR) in critically ill patients with acute kidney injury - true GFR versus urinary creatinine clearance and estimating equations

    Science.gov (United States)

    2013-01-01

    Introduction Estimation of kidney function in critically ill patients with acute kidney injury (AKI), is important for appropriate dosing of drugs and adjustment of therapeutic strategies, but challenging due to fluctuations in kidney function, creatinine metabolism and fluid balance. Data on the agreement between estimating and gold standard methods to assess glomerular filtration rate (GFR) in early AKI are lacking. We evaluated the agreement of urinary creatinine clearance (CrCl) and three commonly used estimating equations, the Cockcroft Gault (CG), the Modification of Diet in Renal Disease (MDRD) and the Chronic Kidney Disease Epidemiology Collaboration (CKD-EPI) equations, in comparison to GFR measured by the infusion clearance of chromium-ethylenediaminetetraacetic acid (51Cr-EDTA), in critically ill patients with early AKI after complicated cardiac surgery. Methods Thirty patients with early AKI were studied in the intensive care unit, 2 to 12 days after complicated cardiac surgery. The infusion clearance for 51Cr-EDTA obtained as a measure of GFR (GFR51Cr-EDTA) was calculated from the formula: GFR (mL/min/1.73m2) = (51Cr-EDTA infusion rate × 1.73)/(arterial 51Cr-EDTA × body surface area) and compared with the urinary CrCl and the estimated GFR (eGFR) from the three estimating equations. Urine was collected in two 30-minute periods to measure urine flow and urine creatinine. Urinary CrCl was calculated from the formula: CrCl (mL/min/1.73m2) = (urine volume × urine creatinine × 1.73)/(serum creatinine × 30 min × body surface area). Results The within-group error was lower for GFR51Cr-EDTA than the urinary CrCl method, 7.2% versus 55.0%. The between-method bias was 2.6, 11.6, 11.1 and 7.39 ml/min for eGFRCrCl, eGFRMDRD, eGFRCKD-EPI and eGFRCG, respectively, when compared to GFR51Cr-EDTA. The error was 103%, 68.7%, 67.7% and 68.0% for eGFRCrCl, eGFRMDRD, eGFRCKD-EPI and eGFRCG, respectively, when compared to GFR51Cr-EDTA. Conclusions The study

  5. Development of the control assembly pattern and dynamic analysis of the Generation IV large gas-cooled fast reactor (GFR)

    Energy Technology Data Exchange (ETDEWEB)

    Girardin, G.

    2009-07-15

    During the past ten years, different independent factors, such as the rapidly increasing worldwide demand in energy, societal concerns about greenhouse gas emissions, and the high and volatile prices for fossil fuels, have contributed to the renewed interest in nuclear technology. In this context, the Generation IV International Forum (GIF) launched the initiative to collaborate on the research and development efforts needed for the next generation of nuclear reactors. A particular goal set for Generation IV systems is closure of the nuclear fuel cycle; they are expected to offer a better utilization of natural resources, as also a minimization of long-lived radioactive wastes. Among the systems selected by the GIF, the Gas-cooled Fast Reactor (GFR) is a highly innovative system with advanced fuel geometry and materials. The principal aim of the present research is to develop and qualify the control assembly (CA) pattern and corresponding CA implementation scheme for the 2400 MWth reference GFR design. The work has been carried out in three successive phases: (1) validation of the neutronics tools, (2) the CA pattern development and related static analysis, and (3) dynamic core behaviour studies for hypothetical CA driven transients. The deterministic code system ERANOS and its associated nuclear data libraries for fast reactors were developed and validated for sodium-cooled reactors. In order to validate ERANOS for GFR applications, a systematic reanalysis of the GFR-relevant integral data generated at PSI during the GCFR-PROTEUS experimental program of the 1970’s was undertaken. The reference PROTEUS test lattice has been analyzed with ERANOS-2.0 and its associated, adjusted nuclear data library ERALIB1. Benchmark calculations were performed with the Monte Carlo code MCNPX, allowing one to both check the deterministic results and to analyze the sensitivity to different modern data libraries. For the main reaction rate ratios, the new analysis of the GCFR

  6. Development of the control assembly pattern and dynamic analysis of the Generation IV large gas-cooled fast reactor (GFR)

    International Nuclear Information System (INIS)

    Girardin, G.

    2009-07-01

    During the past ten years, different independent factors, such as the rapidly increasing worldwide demand in energy, societal concerns about greenhouse gas emissions, and the high and volatile prices for fossil fuels, have contributed to the renewed interest in nuclear technology. In this context, the Generation IV International Forum (GIF) launched the initiative to collaborate on the research and development efforts needed for the next generation of nuclear reactors. A particular goal set for Generation IV systems is closure of the nuclear fuel cycle; they are expected to offer a better utilization of natural resources, as also a minimization of long-lived radioactive wastes. Among the systems selected by the GIF, the Gas-cooled Fast Reactor (GFR) is a highly innovative system with advanced fuel geometry and materials. The principal aim of the present research is to develop and qualify the control assembly (CA) pattern and corresponding CA implementation scheme for the 2400 MWth reference GFR design. The work has been carried out in three successive phases: (1) validation of the neutronics tools, (2) the CA pattern development and related static analysis, and (3) dynamic core behaviour studies for hypothetical CA driven transients. The deterministic code system ERANOS and its associated nuclear data libraries for fast reactors were developed and validated for sodium-cooled reactors. In order to validate ERANOS for GFR applications, a systematic reanalysis of the GFR-relevant integral data generated at PSI during the GCFR-PROTEUS experimental program of the 1970’s was undertaken. The reference PROTEUS test lattice has been analyzed with ERANOS-2.0 and its associated, adjusted nuclear data library ERALIB1. Benchmark calculations were performed with the Monte Carlo code MCNPX, allowing one to both check the deterministic results and to analyze the sensitivity to different modern data libraries. For the main reaction rate ratios, the new analysis of the GCFR

  7. 2-D FEM Simulation of Propagation and Radiation of Leaky Lamb Wave in a Plate-Type Ultrasonic Waveguide Sensor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang-Jin; Kim, Hoe-Woong; Joo, Young-Sang; Kim, Sung-Kyun; Kim, Jong-Bum [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This paper introduces the 2-D FEM simulation of the propagation and radiation of the leaky Lamb wave in and from a plate-type ultrasonic waveguide sensor conducted for the radiation beam profile analysis. The FEM simulations are performed with three different excitation frequencies and the radiation beam profiles obtained from FEM simulations are compared with those obtained from corresponding experiments. This paper deals with the 2-D FEM simulation of the propagation and radiation of the leaky Lamb wave in and from a plate-type ultrasonic waveguide sensor conducted to analyze the radiation beam profiles. The radiation beam profile results obtained from the FEM simulation show good agreement with the ones obtained from the experiment. This result will be utilized to improve the performance of the developed waveguide sensor. The quality of the visualized image is mainly affected by beam profile characteristics of the leaky wave radiated from the waveguide sensor. However, the relationships between the radiation beam profile and many parameters of the waveguide sensor are not fully revealed yet. Therefore, further parametric studies are necessary to improve the performance of the sensor and the finite element method (FEM) is one of the most effective tools for the parametric study.

  8. Development of the control assembly pattern and dynamic analysis of the generation IV large gas-cooled fast reactor (GFR)

    International Nuclear Information System (INIS)

    Girardin, G.

    2009-01-01

    Among the systems selected by the GIF, the Gas-cooled Fast Reactor (GFR) is a highly innovative system with advanced fuel geometry and materials. It is in the context of the large, 2400 MWth reference GFR design that the present doctoral research has been conducted, the principal aim having been to develop and qualify the control assembly (CA) pattern and corresponding CA implementation scheme for this system. The work has been carried out in three successive and complementary phases: (1) validation of the neutronics tools, (2) the CA pattern development and related static analysis, and (3) dynamic core behavior studies for hypothetical CA driven transients. During the first phase of the thesis, the reference PROTEUS test lattice from these experiments has been analyzed with ERANOS-2.0 and its associated, adjusted nuclear data library ERALIB1. Additionally, benchmark calculations were performed with the Monte Carlo code MCNPX, allowing one to both check the deterministic results and to analyze the sensitivity to different modern data libraries. It has been found that, for the main reaction rate ratios, the new analysis of the GCFR-PROTEUS reference lattice generally yields good agreement - within 1σ measurement uncertainty - with experimental values and with the Monte Carlo simulations. As shown by the analysis, the predictions were in somewhat better agreement in the case of the adjusted ERALIB1 library. The applicability of ERANOS-2.0/ERALIB1 as the reference neutronics tool for the GFR analysis could thus be demonstrated. Furthermore, neutronics aspects related to the novel features of the GFR, for which new experimental investigations are needed, were highlighted. In the second phase of the research, the CA pattern was developed for the GFR, based on iterative neutronics and thermal-hydraulics calculations, 2D and 3D neutronics models for the reactor core having first been set up using the reference ERANOS-2.0/ERALIB1 computational scheme. For the thermal

  9. Development and validation of GFR-estimating equations using diabetes, transplant and weight

    DEFF Research Database (Denmark)

    Stevens, L.A.; Schmid, C.H.; Zhang, Y.L.

    2009-01-01

    interactions. Equations were developed in a pooled database of 10 studies [2/3 (N = 5504) for development and 1/3 (N = 2750) for internal validation], and final model selection occurred in 16 additional studies [external validation (N = 3896)]. RESULTS: The mean mGFR was 68, 67 and 68 ml/min/ 1.73 m(2......BACKGROUND: We have reported a new equation (CKD-EPI equation) that reduces bias and improves accuracy for GFR estimation compared to the MDRD study equation while using the same four basic predictor variables: creatinine, age, sex and race. Here, we describe the development and validation...... of this equation as well as other equations that incorporate diabetes, transplant and weight as additional predictor variables. METHODS: Linear regression was used to relate log-measured GFR (mGFR) to sex, race, diabetes, transplant, weight, various transformations of creatinine and age with and without...

  10. Fluctuations in eGFR in relation to unenhanced and enhanced MRI and CT outpatients

    Energy Technology Data Exchange (ETDEWEB)

    Azzouz, Manal, E-mail: manalazzouz@gmail.com [Department of Diagnostic Radiology, Copenhagen University Hospital Herlev, Herlev Ringvej 75, DK 2730 Herlev (Denmark); Rømsing, Janne [Department of Drug Design and Pharmacology, University of Copenhagen, Universitetsparken 2, DK-2100 Copenhagen Ø (Denmark); Thomsen, Henrik S. [Department of Diagnostic Radiology, Copenhagen University Hospital Herlev, Herlev Ringvej 75, DK 2730 Herlev (Denmark)

    2014-06-15

    Objective: To study fluctuations in estimated glomerular filtration rate (eGFR) in relation to contrast medium (CM) enhanced magnetic resonance imaging (MRI) and computed tomography (CT) compared to control groups in outpatients. Materials and methods: eGFR was determined right before the imaging procedure and three days later at the department or at the patient's home. The iodine-based and gadolinium-based contrast media were the same as used for all other examinations at the department. Results: A total of 716 patients completed the study. There was a statistically significant, but not clinically relevant rise in eGFR after three days in all four groups. The average eGFR variation was 4.8 ml/min/1.73 m{sup 2}. There were large variations in eGFR between the two measurements in 45.8% of the patients as they had a change greater than ±10 ml/min/1.73 m{sup 2}. Only three patients fulfilled the contrast-induced nephropathy (CIN) requirement when the definition s-creatinine ≥44 μmol/l (0.5 mg/dl) was used. Conclusions: eGFR in outpatients undergoing MRI or CT did vary independently of whether the patient received contrast or not. The findings probably reflect the natural variations in s-creatinine levels. This should be taken into consideration when CIN is studied.

  11. Fluctuations in eGFR in relation to unenhanced and enhanced MRI and CT outpatients

    International Nuclear Information System (INIS)

    Azzouz, Manal; Rømsing, Janne; Thomsen, Henrik S.

    2014-01-01

    Objective: To study fluctuations in estimated glomerular filtration rate (eGFR) in relation to contrast medium (CM) enhanced magnetic resonance imaging (MRI) and computed tomography (CT) compared to control groups in outpatients. Materials and methods: eGFR was determined right before the imaging procedure and three days later at the department or at the patient's home. The iodine-based and gadolinium-based contrast media were the same as used for all other examinations at the department. Results: A total of 716 patients completed the study. There was a statistically significant, but not clinically relevant rise in eGFR after three days in all four groups. The average eGFR variation was 4.8 ml/min/1.73 m 2 . There were large variations in eGFR between the two measurements in 45.8% of the patients as they had a change greater than ±10 ml/min/1.73 m 2 . Only three patients fulfilled the contrast-induced nephropathy (CIN) requirement when the definition s-creatinine ≥44 μmol/l (0.5 mg/dl) was used. Conclusions: eGFR in outpatients undergoing MRI or CT did vary independently of whether the patient received contrast or not. The findings probably reflect the natural variations in s-creatinine levels. This should be taken into consideration when CIN is studied

  12. Estimating glomerular filtration rate (GFR) in children. The average between a cystatin C- and a creatinine-based equation improves estimation of GFR in both children and adults and enables diagnosing Shrunken Pore Syndrome.

    Science.gov (United States)

    Leion, Felicia; Hegbrant, Josefine; den Bakker, Emil; Jonsson, Magnus; Abrahamson, Magnus; Nyman, Ulf; Björk, Jonas; Lindström, Veronica; Larsson, Anders; Bökenkamp, Arend; Grubb, Anders

    2017-09-01

    Estimating glomerular filtration rate (GFR) in adults by using the average of values obtained by a cystatin C- (eGFR cystatin C ) and a creatinine-based (eGFR creatinine ) equation shows at least the same diagnostic performance as GFR estimates obtained by equations using only one of these analytes or by complex equations using both analytes. Comparison of eGFR cystatin C and eGFR creatinine plays a pivotal role in the diagnosis of Shrunken Pore Syndrome, where low eGFR cystatin C compared to eGFR creatinine has been associated with higher mortality in adults. The present study was undertaken to elucidate if this concept can also be applied in children. Using iohexol and inulin clearance as gold standard in 702 children, we studied the diagnostic performance of 10 creatinine-based, 5 cystatin C-based and 3 combined cystatin C-creatinine eGFR equations and compared them to the result of the average of 9 pairs of a eGFR cystatin C and a eGFR creatinine estimate. While creatinine-based GFR estimations are unsuitable in children unless calibrated in a pediatric or mixed pediatric-adult population, cystatin C-based estimations in general performed well in children. The average of a suitable creatinine-based and a cystatin C-based equation generally displayed a better diagnostic performance than estimates obtained by equations using only one of these analytes or by complex equations using both analytes. Comparing eGFR cystatin and eGFR creatinine may help identify pediatric patients with Shrunken Pore Syndrome.

  13. Comparing the difference of measured GFR of ectopic pelvic kidney between anterior and posterior imaging processing in renal dynamic imaging

    International Nuclear Information System (INIS)

    Li Baojun; Zhao Deshan

    2014-01-01

    Objective: To compare and analyze the difference of measured glomerular filtration rate (GFR) of ectopic pelvic kidney between anterior and posterior imaging processing in renal dynamic imaging. Methods: There were 10 patients collected retrospectively, with ectopic kidneys in pelvic cavity confirmed by ultrasound, CT, renal dynamic imaging and other imaging modalities. All images of ectopic kidneys in renal dynamic imaging were processed by anterior and posterior methods respectively. The ectopic kidney was only processed in anterior imaging, ectopic kidney and contralateral normal kidney were processed in posterior imaging. Total GFR equalled the sum of GFR of normal kidney in posterior imaging and GFR of ectopic kidney in anterior imaging, was compared with total GFR of two kidneys in posterior imaging and GFR in two-sample method. All correlation analysis were completed between GFRs from three methods and all patients were followed up. Statistically paired t-test and bivariate correlation analysis test were used. Results: The mean GFR of ectopic kidney in anterior imaging equal to (27.48±12.24) ml/(min · 1.73 m 2 ). It was more than GFR [(10.71 ±4.74) ml/ (min · 1.73 m 2 )] in posterior imaging above 46% (t=5.481, P<0.01). There was no significant difference (t=-2.238, P>0.05), but better correlation (r=0.704, P<0.05) between total GFR in anterior imaging and GFR in two-sample method. There was significant difference (t=4.629, P<0.01)and worse correlation (r=0.576, P>0.05) between total GFR in posterior imaging and GFR in two-sample method. Conclusion: Comparing with GFR in posterior imaging, GFR in anterior imaging can more truly reflect function condition of ectopic pelvic kidney in renal dynamic imaging. (authors)

  14. A new surface fractal dimension for displacement mode shape-based damage identification of plate-type structures

    Science.gov (United States)

    Shi, Binkai; Qiao, Pizhong

    2018-03-01

    Vibration-based nondestructive testing is an area of growing interest and worthy of exploring new and innovative approaches. The displacement mode shape is often chosen to identify damage due to its local detailed characteristic and less sensitivity to surrounding noise. Requirement for baseline mode shape in most vibration-based damage identification limits application of such a strategy. In this study, a new surface fractal dimension called edge perimeter dimension (EPD) is formulated, from which an EPD-based window dimension locus (EPD-WDL) algorithm for irregularity or damage identification of plate-type structures is established. An analytical notch-type damage model of simply-supported plates is proposed to evaluate notch effect on plate vibration performance; while a sub-domain of notch cases with less effect is selected to investigate robustness of the proposed damage identification algorithm. Then, fundamental aspects of EPD-WDL algorithm in term of notch localization, notch quantification, and noise immunity are assessed. A mathematical solution called isomorphism is implemented to remove false peaks caused by inflexions of mode shapes when applying the EPD-WDL algorithm to higher mode shapes. The effectiveness and practicability of the EPD-WDL algorithm are demonstrated by an experimental procedure on damage identification of an artificially-induced notched aluminum cantilever plate using a measurement system of piezoelectric lead-zirconate (PZT) actuator and scanning laser Doppler vibrometer (SLDV). As demonstrated in both the analytical and experimental evaluations, the new surface fractal dimension technique developed is capable of effectively identifying damage in plate-type structures.

  15. A GFR benchmark comparison of transient analysis codes based on the ETDR concept

    International Nuclear Information System (INIS)

    Bubelis, E.; Coddington, P.; Castelliti, D.; Dor, I.; Fouillet, C.; Geus, E. de; Marshall, T.D.; Van Rooijen, W.; Schikorr, M.; Stainsby, R.

    2007-01-01

    A GFR (Gas-cooled Fast Reactor) transient benchmark study was performed to investigate the ability of different code systems to calculate the transition in the core heat removal from the main circuit forced flow to natural circulation cooling using the Decay Heat Removal (DHR) system. This benchmark is based on a main blower failure in the Experimental Technology Demonstration Reactor (ETDR) with reactor scram. The codes taking part into the benchmark are: RELAP5, TRAC/AAA, CATHARE, SIM-ADS, MANTA and SPECTRA. For comparison purposes the benchmark was divided into several stages: the initial steady-state solution, the main blower flow run-down, the opening of the DHR loop and the transition to natural circulation and finally the 'quasi' steady heat removal from the core by the DHR system. The results submitted by the participants showed that all the codes gave consistent results for all four stages of the benchmark. In the steady-state the calculations revealed some differences in the clad and fuel temperatures, the core and main loop pressure drops and in the total Helium mass inventory. Also some disagreements were observed in the Helium and water flow rates in the DHR loop during the final natural circulation stage. Good agreement was observed for the total main blower flow rate and Helium temperature rise in the core, as well as for the Helium inlet temperature into the core. In order to understand the reason for the differences in the initial 'blind' calculations a second round of calculations was performed using a more precise set of boundary conditions

  16. Recording blood pressure and eGFR in primary care after the Belgrade screening study.

    Science.gov (United States)

    Lezaic, Visnja; Marinkovic, Jelena; Milutinovic, Zoran; Jovanovic-Vasiljevic, Nada; Vujicic, Vesna; Pejovic, Branka; Kalabic, Snezana; Djukanovic, Ljubica

    2018-11-01

    In 2009, Belgrade nephrologists and general practitioners from thirteen health centers carried out screening for chronic kidney disease (CKD). Three years later, medical records of patients from four health centers participating in the screening study were retrospectively analyzed in order to check whether general practitioners had continued to control patients at risk for CKD in accordance with the recommendations provided. The study included 460 patients who visited their doctor at least once in the three-year period. Data on blood pressure, ACEI use, estimated glomerular filtration rate (eGFR) and comorbidities were taken from patients' medical records. Blood pressure was not recorded in any of the three years in 42.8% and eGFR in 36.7% of the patients, but blood pressure was registered every year in 7.8% and eGFR in 4.3% of them. Over the three years, the relative number of patients with recorded blood pressure decreased from 41.7% to 17.8%, and with recorded eGFR from 41.7% to 21.5%. Multivariate linear regression found that Health Center, systolic and diastolic blood pressure and presence of hypertension were negatively associated with number of years with recorded blood pressure. Health Center, systolic blood pressure and sum of years with recorded eGFR below 60 ml/min/1.73m 2 were associated with number of years with recorded eGFR. Under-recording of blood pressure and eGFR in primary care health centers suggests lack of adherence to current guidelines and insufficient care of CKD patients. This implies the necessity for continuous education of physicians.

  17. A method to eliminate the effect of protein binding on GFR

    International Nuclear Information System (INIS)

    Prabakaran, K.; Fernandes, V.; Nour, R.

    1998-01-01

    Full text: Plasma clearance of 99m Tc-DTPA is a standard method for GFR assessment. Protein binding (PB) of 99m Tc-DTPA is thought to affect the accuracy of the GFR measurement. Therefore, eliminating PB portion would improve the accuracy of this measurement. Methods to eliminate PB are usually complex, but a simple Amicon micropartition system has been proposed to eliminate the PB effect. This study is aimed to assess if this simple method effectively excluded the PB fraction and improved the accuracy of the GFR measurement. Therefore, eliminating PB portion would improve the accuracy of this measurement. Methods to eliminate PB are usually complex, but a simple Amicon micropartition system separates free 99m Tc-DTPA from the PB fraction after a 10 min centrifugation of patient plasma. The ultrafiltrate obtained is totally free of PB 99m Tc-DTPA. 20 consecutive patients had GFRs performed using the two sample method with blood drawn at 60 and 150 min. All samples had an Ultrafiltrate 99m Tc-DTPA and normal plasma 99m Tc-DTPA. The GFRs were calculated from both samples. The findings showed a mean PB of 6.5 + 3.9% and 13 + 5.0% at 60 min and 150 min respectively for our DTPA. The mean GFR from normal plasma 85.7 + 26.4 mL/min and from ultrafiltrate = 97.2 + 28.8 mL/min. Statistical analysis using Student''s ''t'' test shows that there is a significant difference between the values of GFR (P < 0.05). Our result confirms that the PB of DTPA is substantial and affects the GFR estimation significantly. The method of ultrafiltration eliminates the PB, and very simple to use. This method could be used on a routine basis and/or to assess the suitability of the DTPA used in various departments for GFR measurement

  18. Expression of GDNF and GFR alpha 1 in mouse taste bud cells.

    Science.gov (United States)

    Takeda, Masako; Suzuki, Yuko; Obara, Nobuko; Uchida, Nobuhiko; Kawakoshi, Kentaro

    2004-11-01

    GDNF (glial cell line-derived neurotrophic factor) affects the survival and maintenance of central and peripheral neurons. Using an immunocytochemical method, we examined whether the taste bud cells in the circumvallate papillae of normal mice expressed GDNF and its GFR alpha 1 receptor. Using double immunostaining for either of them and NCAM, PGP 9.5, or alpha-gustducin, we additionally sought to determine what type of taste bud cells expressed GDNF or GFR alpha 1, because NCAM is reported to be expressed in type-III cells, PGP 9.5, in type-III and some type-II cells, and alpha-gustducin, in some type-II cells. Normal taste bud cells expressed both GDNF and GFR alpha 1. The percentage of GDNF-immunoreactive cells among all taste bud cells was 31.63%, and that of GFR alpha 1-immunoreactive cells, 83.21%. Confocal laser scanning microscopic observations after double immunostaining showed that almost none of the GDNF-immunoreactive cells in the taste buds were reactive with anti-NCAM or anti-PGP 9.5 antibody, but could be stained with anti-alpha-gustducin antibody. On the other hand, almost all anti-PGP 9.5- or anti-alpha-gustducin-immunoreactive cells were positive for GFR alpha 1. Thus, GDNF-immunoreactive cells did not include type-III cells, but type-II cells, which are alpha-gustducin-immunoreactive; on the other hand, GFR alpha 1-immunoreactive cells included type-II and -III cells, and perhaps type-I cells. We conclude that GDNF in the type-II cells may exert trophic actions on type-I, -II, and -III taste bud cells by binding to their GFR alpha 1 receptors.

  19. A comparison of the performances of an artificial neural network and a regression model for GFR estimation.

    Science.gov (United States)

    Liu, Xun; Li, Ning-shan; Lv, Lin-sheng; Huang, Jian-hua; Tang, Hua; Chen, Jin-xia; Ma, Hui-juan; Wu, Xiao-ming; Lou, Tan-qi

    2013-12-01

    Accurate estimation of glomerular filtration rate (GFR) is important in clinical practice. Current models derived from regression are limited by the imprecision of GFR estimates. We hypothesized that an artificial neural network (ANN) might improve the precision of GFR estimates. A study of diagnostic test accuracy. 1,230 patients with chronic kidney disease were enrolled, including the development cohort (n=581), internal validation cohort (n=278), and external validation cohort (n=371). Estimated GFR (eGFR) using a new ANN model and a new regression model using age, sex, and standardized serum creatinine level derived in the development and internal validation cohort, and the CKD-EPI (Chronic Kidney Disease Epidemiology Collaboration) 2009 creatinine equation. Measured GFR (mGFR). GFR was measured using a diethylenetriaminepentaacetic acid renal dynamic imaging method. Serum creatinine was measured with an enzymatic method traceable to isotope-dilution mass spectrometry. In the external validation cohort, mean mGFR was 49±27 (SD) mL/min/1.73 m2 and biases (median difference between mGFR and eGFR) for the CKD-EPI, new regression, and new ANN models were 0.4, 1.5, and -0.5 mL/min/1.73 m2, respectively (P30% from mGFR) were 50.9%, 77.4%, and 78.7%, respectively (Psource of systematic bias in comparisons of new models to CKD-EPI, and both the derivation and validation cohorts consisted of a group of patients who were referred to the same institution. An ANN model using 3 variables did not perform better than a new regression model. Whether ANN can improve GFR estimation using more variables requires further investigation. Copyright © 2013 National Kidney Foundation, Inc. Published by Elsevier Inc. All rights reserved.

  20. Rethinking CKD Evaluation: Should We Be Quantifying Basal or Stimulated GFR to Maximize Precision and Sensitivity?

    Science.gov (United States)

    Molitoris, Bruce A.

    2017-01-01

    Chronic kidney disease (CKD) remains an increasing clinical problem. Although clinical risk factors and biomarkers for development and progression of CKD have been identified, there is no commercial surveillance technology to definitively diagnose and quantify the severity and progressive loss of glomerular filtration rate (GFR) in CKD. This has limited the study of potential therapies to late stages of CKD when FDA-registerable events are more likely. Since patient outcomes, including the rate of CKD progression, correlate with disease severity, and effective therapy may require early intervention, being able to diagnose and stratify patients by their level of decreased kidney function early on is key for translational progress. In addition, renal reserve, defined as the increase in GFR following stimulation, may improve the quantification of GFR based solely on basal levels. Various groups are developing and characterizing optical measurement techniques utilizing new minimally invasive or non-invasive approaches for quantifying basal and stimulated kidney function. This development has the potential to allow widespread individualization of therapy at an earlier disease stage. Therefore, the purposes of this review are to suggest why quantifying stimulated GFR, by activating renal reserve, may be advantageous in patients and review fluorescent technologies to deliver patient-specific GFR. PMID:28223001

  1. Cross-Sectional Evaluation of Kidney Function in Hospitalized Patients: Estimated GFR Versus Renal Scintigraphy

    Directory of Open Access Journals (Sweden)

    Domenico Santoro

    2014-12-01

    Full Text Available Background/Aims: Accurate staging of chronic kidney disease (CKD is very important. We tried to identify difference in GFR evaluation between CKD-EPI and Gates method with renal scintigraphy and which variables are associated with these differences. Methods: We retrospectively reviewed the records of 341 patients who underwent dynamic renal scintigraphy in the last 5 years. Patients were categorized according to KDIGO staging I to V, using the eGFR calculated with the CKD-EPI equation. Secondarily, we stratified patients according to treatment with renin-angiotensin system (RAS inhibitors. Results: Gates method tends to underestimate GFR especially in CKD stage I (mean -22.2 ml/min and II (mean -12.5 ml/min. The division in quartiles of ages showed an underestimation of GFR only in the first quartile of age (Conclusion: The assessment of GFR by the Gates method must be carefully considered in the early stages of CKD, especially in younger patients. Moreover, the difference is more pronounced in patients treated with RAS inhibitors. Longitudinal studies will prove which method better predicts cardiovascular or renal events.

  2. 'Scan GFR' as an alternative to 51Cr-EDTA clearance

    International Nuclear Information System (INIS)

    Osman, E.A.; Clarke, M.B.; Barrett, J.J.; Parsons, V.; Dulwich Hospital, London

    1989-01-01

    In this report 39 patients had their GFR assessed by 51 Cr-EDTA in a single shot-multiple sampling technique and also by measuring the quantitative renal uptake of 99m Tc-DTPA during a standard renal imaging test. The results showed a correlation coefficient of 0.96 between the two techniques. By using linear regression analysis the GFR was derived from the % renal uptake of DTPA in a further series of 24 patients in whom the 51 Cr-EDTA method was also employed. The two methods were compared and showed a correlation coefficient of 0.94 with a mean GFR difference of 7 ± 4.5 ml/min over a 51 Cr-EDTA range of 38-150 ml/min. (orig.) [de

  3. Effect of food and activity on the reproducibility of isotopic GFR estimation

    International Nuclear Information System (INIS)

    Wilkinson, J.; Fleming, J.S.; Waller, D.G.

    1990-01-01

    The reproducibility of the plasma clearance of 99 Tc m DTPA was studied in 26 patients under standardized conditions with the subject fasting and at rest. The coefficient of variation of duplicate measurements in patients with glomerular filtration rates (GRFs) ranging from 11-103 ml min -1 was 8%. Mean GFR following a breakfast containing 670 kcal and 31 g protein was increased significantly from 40.7±28.1 ml min -1 to 43.6±30.8 ml min -1 . When fasted but permitted free exercise there was no consistent trend in GFR but the coefficient of variation of duplicate estimates increased significantly to 12.1%. It is recommended that routine GFR measurement should be carried out fasting or following a light diet with restricted activity. (author)

  4. Cystatin C levels in healthy kidney donors and its correlation with GFR by creatinine clearance

    International Nuclear Information System (INIS)

    Ayub, S.; Khan, S.; Zafar, M.N.

    2014-01-01

    Objective: To determine Serum Cystatin C (S.CysC) levels in healthy potential kidney donors and its correlation with Serum Creatinine (S.Cr), Glomerular filtration rate (GFR) by 24 hour urinary Creatinine clearance (CCL) and GFR by formulae of Cockcroft Gault (CCG) and Modification of diet in Renal Disease (MDRD). Methods: A Cross sectional study was conducted at Sindh Institute of Urology and Transplantation (SIUT), Karachi, between June and December 2012. One hundred and three potential healthy kidney donors were enrolled in the study to measure their S.CysC and correlate it with S.Cr, CCL and GFR by CCG and MDRD. Statistical analysis was done by SPSS 17. Results: The mean age of the healthy kidney donors was 32.19+8.27 years with a M:F ratio of 1.86:1. The mean Serum Creatinine (S.Cr) was 0.86+0.18 mg/dl and mean S.CysC was 0.88+0.12 mg/dl. S.CysC showed significant correlation with S.Cr (r = 0.78, p<0.001), CCL (r = 0.67, p<0.001), GFR CCG (r = 0.54, p<0.001) and GFR MDRD (r = 0.67, p<0.001). Correlation of S.CysC was better than S.Cr for CCL, S.Cr (0.60) vs S.CysC (0.67) and GFR CCG, S.Cr (0.41) vs S.CysC (0.54). Correlation was comparable for MDRD, S.Cr (0.67) vs S.Cys (0.67). Conclusion: S.CysC is better marker of kidney function in potential healthy kidney donors. It is a reliable, convenient and economical marker that can be used especially in routine clinical practice. (author)

  5. Effect of Genetic African Ancestry on eGFR and Kidney Disease

    Science.gov (United States)

    Nadkarni, Girish N.; Belbin, Gillian; Lotay, Vaneet; Wyatt, Christina; Gottesman, Omri; Bottinger, Erwin P.; Kenny, Eimear E.; Peter, Inga

    2015-01-01

    Self-reported ancestry, genetically determined ancestry, and APOL1 polymorphisms are associated with variation in kidney function and related disease risk, but the relative importance of these factors remains unclear. We estimated the global proportion of African ancestry for 9048 individuals at Mount Sinai Medical Center in Manhattan (3189 African Americans, 1721 European Americans, and 4138 Hispanic/Latino Americans by self-report) using genome-wide genotype data. CKD-EPI eGFR and genotypes of three APOL1 coding variants were available. In admixed African Americans and Hispanic/Latino Americans, serum creatinine values increased as African ancestry increased (per 10% increase in African ancestry, creatinine values increased 1% in African Americans and 0.9% in Hispanic/Latino Americans; P≤1x10−7). eGFR was likewise significantly associated with African genetic ancestry in both populations. In contrast, APOL1 risk haplotypes were significantly associated with CKD, eGFRblack on the basis of ≥50% African ancestry resulted in higher eGFR for 14.7% of Hispanic/Latino Americans and lower eGFR for 4.1% of African Americans, affecting CKD staging in 4.3% and 1% of participants, respectively. Reclassified individuals had electrolyte values consistent with their newly assigned CKD stage. In summary, proportion of African ancestry was significantly associated with normal-range creatinine and eGFR, whereas APOL1 risk haplotypes drove the associations with CKD. Recalculation of eGFR on the basis of genetic ancestry affected CKD staging and warrants additional investigation. PMID:25349204

  6. Estimated GFR Decline as a Surrogate End Point for Kidney Failure

    DEFF Research Database (Denmark)

    Lambers Heerspink, Hiddo J; Weldegiorgis, Misghina; Inker, Lesley A

    2014-01-01

    A doubling of serum creatinine value, corresponding to a 57% decline in estimated glomerular filtration rate (eGFR), is used frequently as a component of a composite kidney end point in clinical trials in type 2 diabetes. The aim of this study was to determine whether alternative end points defin...

  7. Cystatin C Falsely Underestimated GFR in a Critically Ill Patient with a New Diagnosis of AIDS

    Directory of Open Access Journals (Sweden)

    Caitlin S. Brown

    2016-01-01

    Full Text Available Cystatin C has been suggested to be a more accurate glomerular filtration rate (GFR surrogate than creatinine in patients with acquired immunodeficiency syndrome (AIDS because it is unaffected by skeletal muscle mass and dietary influences. However, little is known about the utility of this marker for monitoring medications in the critically ill. We describe the case of a 64-year-old female with opportunistic infections associated with a new diagnosis of AIDS. During her course, she experienced neurologic, cardiac, and respiratory failure; yet her renal function remained preserved as indicated by an eGFR ≥ 120 mL/min and a urine output > 1 mL/kg/hr without diuresis. The patient was treated with nephrotoxic agents; therefore cystatin C was assessed to determine if cachexia was resulting in a falsely low serum creatinine. Cystatin C measured 1.50 mg/L which corresponded to an eGFR of 36 mL/min. Given the >60 mL/min discrepancy, serial 8-hour urine samples were collected and a GFR > 120 mL/min was confirmed. It is unclear why cystatin C was falsely elevated, but we hypothesize that it relates to the proinflammatory state with AIDS, opportunistic infections, and corticosteroids. More research is needed before routine use of cystatin C in this setting can be recommended.

  8. Comparison of the predictive performance of eGFR formulae for mortality and graft failure in renal transplant recipients.

    LENUS (Irish Health Repository)

    He, Xiang

    2009-02-15

    To date, efforts have focused on assessing estimated glomerular filtration rate (eGFR) formulae against measured GFR. However, a more appropriate clinical gold standard is one conveying a defined clinical disadvantage. In renal transplantation, these measures are mortality and graft failure.

  9. Distribution of estimated glomerular filtration rate (eGFR) values in patients receiving contrast-enhanced magnetic resonance imaging

    International Nuclear Information System (INIS)

    Shimoji, Keigo; Aoki, Shigeki; Nakanishi, Atsushi

    2012-01-01

    The aim of this study was to elucidate the distribution of estimated glomerular filtration rate (eGFR) values in patients who underwent gadolinium-based contrast agent (GBCA)-enhanced magnetic resonance imaging (MRI) at different types of hospitals. We retrospectively studied 2,550 patients who underwent MRI at five institutions. We recorded the date and value of each patient's eGFR test. The distribution of eGFR values was compared with that in the general Japanese population. A total of 84.3% of patients had their eGFRs evaluated before GBCA-enhanced MRI. Of these, 84.7% were evaluated within 3 months before the GBCA-enhanced MRI, and 1.3% were evaluated on the day of the GBCA-enhanced MRI. A total of 87.2% of patients tested had an eGFR of ≥60 ml/min/1.73 m 2 ; 12.8% had an eGFR of 2 , and no patients had an eGFR of 2 . The rate of renal function evaluation differed among hospitals. The prevalence of low eGFR values was greater in Juntendo Tokyo Koto Geriatric Medical Center than in the other hospitals, and the prevalence of low eGFR values was greater in patients who underwent GBCA-enhanced MRI than in the general Japanese population. (author)

  10. Comparison of the usefulness of selected formulas for GFR estimation in patients with diagnosed chronic kidney disease

    Directory of Open Access Journals (Sweden)

    Paweł Wróbel

    2018-03-01

    Conclusions: CKD-EPI and abbreviated MDRD formulas have a similar usefulness in GFR value estimation in patients with diagnosed chronic kidney disease. Lower eGFR values achieved using abbreviated MDRD formula and CKD-EPI equation in comparison with Bjornsson’s formula may result in an increased number of patients diagnosed with CKD.

  11. Measurement of glomerular filtration rate (GFR) with 99mTc-DTPA and semiconductors minidetectors: further validation of the method

    International Nuclear Information System (INIS)

    Mahlstedt, J.; Muehlbauer, J.; Schrott, K.H.; Wolf, F.

    1982-01-01

    A technique measuring external radiation with semiconductor minidetectors (SCM) by use of a portable solid state memory after bolus injection of 1mCi 99mTc-DTPA is described. The correlation of the calculated clearances to plasma creatinine levels and to endogenous creatinine clearance as well is satisfactory. This new technique is compared with a standard steady state technique (GFR-SS) using 51 Cr-EDTA. GFR-SCM is strongly correlated to GFR-SS as standard method, therefore the data provides substantial evidence that GFR-SCM is a reliable method of GFR measurement. The advantages are: fast availability of the test result especially when calculated by use of a minicomputer; ease of performance in routine work; convenience for the patient without need of compliance and low costs for the detector system

  12. Prevention of criticality accidents. Fuel elements storage

    International Nuclear Information System (INIS)

    Canavese, S.I.; Capadona, N.M.

    1990-01-01

    Before the need to store fuel elements of the plate type MTR (Materials Testing Reactors), produced with enriched uranium at 20% in U235 for research reactors, it requires the design of a deposit for this purpose, which will give intrinsic security at a great extent and no complaints regarding its construction, is required. (Author) [es

  13. Clinical significance of determination of GFR, urinary albumin and UAlb/UCr in patients with type 2 diabetes

    International Nuclear Information System (INIS)

    Zhang Li; Li Sumei; Zhang Rong; Zhai Fei

    2010-01-01

    Objective: To study the diagnostic value of determination of glomerular filtration rate (GFR), Urinary albumin (UAlb) and UAlb/UCr for early diabetic nephropathy in patients with type 2 Diabetes. Methods: UAlb, UAlb/UCr were measured with radioimmunoassay (RIA) and GFR measured with Tcm DTPA dynamic renal imaging in 102 DM2 patients with diabetic nephropathy (DN) and 23 DM2 patients without nephropathy. Results: In diabetics with nephropathy with a disease course of less than five years (n=38), the GFR, UAlb/UCr were significantly higher than those in diabetics without nephropathy. With prolongation of the disease, GFR gradually declined and UAlb/UCr rose further and in patients with nephropathy with disease courses over 10 years, the GFR was significantly lower and UAlb/UCr significantly higher than those in diabetics without nephropathy (all P<0.05). The GFR was negatively correlated with course of disease, BMI and UAlb/UCr (r=-0.691, -0.631, -0.698, respectively, P<0.01). Conclusion: GFR can reflect the degree of renal damage of diabetic patients,and may be helpful for evaluation of progression of DN when combined with assay of albuminuria. (authors)

  14. Association of Reduced eGFR and Albuminuria with Serious Fall Injuries among Older Adults.

    Science.gov (United States)

    Bowling, C Barrett; Bromfield, Samantha G; Colantonio, Lisandro D; Gutiérrez, Orlando M; Shimbo, Daichi; Reynolds, Kristi; Wright, Nicole C; Curtis, Jeffrey R; Judd, Suzanne E; Franch, Harold; Warnock, David G; McClellan, William; Muntner, Paul

    2016-07-07

    Falls are common and associated with adverse outcomes in patients on dialysis. Limited data are available in earlier stages of CKD. We analyzed data from 8744 Reasons for Geographic and Racial Differences in Stroke Study participants ≥65 years old with Medicare fee for service coverage. Serious fall injuries were defined as a fall-related fracture, brain injury, or joint dislocation using Medicare claims. Hazard ratios (HRs) for serious fall injuries were calculated by eGFR and albumin-to-creatinine ratio (ACR). Among 2590 participants with CKD (eGFRfall injury compared with age-matched controls without a fall injury was calculated. Overall, 1103 (12.6%) participants had a serious fall injury over 9.9 years of follow-up. The incidence rates per 1000 person-years of serious fall injuries were 21.7 (95% confidence interval [95% CI], 20.3 to 23.2), 26.6 (95% CI, 22.6 to 31.3), and 38.3 (95% CI, 31.2 to 47.0) at eGFR levels ≥60, 45-59, and fall injuries were 0.91 (95% CI, 0.76 to 1.09) and 1.09 (95% CI, 0.86 to 1.37) for eGFR=45-59 and fall and age-matched controls were 21.0% and 5.5%, respectively. Elevated ACR but not lower eGFR was associated with serious fall injuries. Evaluation for fall risk factors and fall prevention strategies should be considered for older adults with elevated ACR. Copyright © 2016 by the American Society of Nephrology.

  15. Association of eGFR-Related Loci Identified by GWAS with Incident CKD and ESRD.

    Directory of Open Access Journals (Sweden)

    Carsten A Böger

    2011-09-01

    Full Text Available Family studies suggest a genetic component to the etiology of chronic kidney disease (CKD and end stage renal disease (ESRD. Previously, we identified 16 loci for eGFR in genome-wide association studies, but the associations of these single nucleotide polymorphisms (SNPs for incident CKD or ESRD are unknown. We thus investigated the association of these loci with incident CKD in 26,308 individuals of European ancestry free of CKD at baseline drawn from eight population-based cohorts followed for a median of 7.2 years (including 2,122 incident CKD cases defined as eGFR <60ml/min/1.73m(2 at follow-up and with ESRD in four case-control studies in subjects of European ancestry (3,775 cases, 4,577 controls. SNPs at 11 of the 16 loci (UMOD, PRKAG2, ANXA9, DAB2, SHROOM3, DACH1, STC1, SLC34A1, ALMS1/NAT8, UBE2Q2, and GCKR were associated with incident CKD; p-values ranged from p = 4.1e-9 in UMOD to p = 0.03 in GCKR. After adjusting for baseline eGFR, six of these loci remained significantly associated with incident CKD (UMOD, PRKAG2, ANXA9, DAB2, DACH1, and STC1. SNPs in UMOD (OR = 0.92, p = 0.04 and GCKR (OR = 0.93, p = 0.03 were nominally associated with ESRD. In summary, the majority of eGFR-related loci are either associated or show a strong trend towards association with incident CKD, but have modest associations with ESRD in individuals of European descent. Additional work is required to characterize the association of genetic determinants of CKD and ESRD at different stages of disease progression.

  16. The Correlation Between the GFR and the Renal Dimensions in Glomerulopathy Patients: Comparison of 2D and 3D Ultrasound

    International Nuclear Information System (INIS)

    Kim, Gyoung Min; Lee, Hak Jong; Hwang, Sung Il; Chin, Ho Jun

    2011-01-01

    We wanted to determine the correlation between the renal length as measured on two dimensional (2D) ultrasonography (US) and the renal parenchymal volume as measured with a new three-dimensional (3D) volume probe ultrasound system. We also wanted to determine the correlation between the renal length or renal parenchymal volume and the glomerular filtration rate (GFR) in patients with glomerulopathy. From July 2007 to December 2007, 26 patients who were pathologically confirmed to have glomerulopathy by biopsy were enrolled. Renal length was measured with 2D US and the renal parenchymal volume was measured with 3D US just prior to biopsy. The GFR was obtained from the electronic medical records. Pearson's correlation coefficients were used to analyze the correlation between the renal length and the renal parenchymal volume, the correlation between the renal length and the GFR and the correlation between the renal parenchymal volume and the GFR. The renal length and the renal parenchymal volume showed strong positive correlation (r = 0.850, p = 0.0001). The correlation coefficient between the renal length and the GFR was 0.623 (p = 0.0007) and the correlation coefficient between the renal volume and the GFR was 0.590 (p = 0.0015). Both the renal length and renal parenchymal volume showed apparently positive correlations with the GFR in glomerulopathy patients. The renal length showed strong positive correlations with the renal parenchymal volume. Both the renal length and the renal parenchymal volume showed apparently positive correlations with the GFR in glomerulopathy patients. In glomerulopathy patients, the renal dimensions measured by ultrasound can reflect the status of the GFR, and the measurement of the 2D renal length could be sufficient for follow up. Further studies are needed to evaluate the role of 3D US for assessing patients with renal disease

  17. Determination of single-kidney glomerular filtration rate (GFR) with CT urography versus renal dynamic imaging Gates method

    Energy Technology Data Exchange (ETDEWEB)

    You, Shan [Hebei North University, Department of Graduate, Zhangjiakou City, Hebei Province (China); Ma, XianWu; Zhang, ChangZhu; Li, Qiang [Qiqihar Chinese Medicine Hospital, Department of Radiology, Qigihar City, Heilongjiang Province (China); Shi, WenWei; Zhang, Jing; Yuan, XiaoDong [The 309th Hospital of Chinese People' s Liberation Army, Department of Radiology, Beijing (China)

    2018-03-15

    To present a single-kidney CT-GFR measurement and compare it with the renal dynamic imaging Gates-GFR. Thirty-six patients with hydronephrosis referred for CT urography and 99mTc-DTPA renal dynamic imaging were prospectively included. Informed consent was obtained from all patients. The CT urography protocol included non-contrast, nephrographic, and excretory phase imaging. The total CT-GFR was calculated by dividing the CT number increments of the total urinary system between the nephrographic and excretory phase by the products of iodine concentration in the aorta and the elapsed time, then multiplied by (1- Haematocrit). The total CT-GFR was then split into single-kidney CT-GFR by a left and right kidney proportionality factor. The results were compared with single-kidney Gates-GFR by using paired t-test, correlation analysis, and Bland-Altman plots. Paired difference between single-kidney CT-GFR (45.02 ± 13.91) and single-kidney Gates-GFR (51.21 ± 14.76) was 6.19 ± 5.63 ml/min, p<0.001, demonstrating 12.1% systematic underestimation with ±11.03 ml/min (±21.5%) measurement deviation. A good correlation was revealed between both measurements (r=0.87, p<0.001). The proposed single-kidney CT-GFR correlates and agrees well with the reference standard despite a systematic underestimation, therefore it could be a one-stop-shop for evaluating urinary tract morphology and split renal function. (orig.)

  18. GFR, serum creatinine and 24-hour urine protein in evaluating renal function of patients with diabetes mellitus

    International Nuclear Information System (INIS)

    Chi Xiaohua; Li Guiping; Liu Feng; Wang Bing; Du Li; Deng Zhifang; Li Wei

    2013-01-01

    Background: Diabetes nephropathy is a common complication of diabetes mellitus patients. Early detection of renal impairment can improve the quality of life of patients. Purpose: The value of total GFR, serum creatinine, 24-hour urine protein excretion in diabetes mellitus patients with renal impairment were evaluated. Methods: A retrospective analysis of 147 patients with diabetes undergoing routine renal dynamic imaging was undertaken. The cases were divided into three groups according to the illness duration: group I of not more than five years, group 2 of five to ten years, Gr.3: more than ten years. The 22 renal transplant donors were selected as the normal control group, The total GFR, serum creatinine and 24-hour urinary protein excretion of all patients were measured before the treatments, and the data were statistically analyzed. Results: There was no significant differences in renal function between the two kidneys of in the diabetes mellitus patients (P=0.536). Serum creatinine and total GFR had significant correlation (R 2 =0.762), but no significant relationship between the 24-hour urine protein and the total GFR or serum creatinine. In the early and middle times of renal function impairment, the total GFR and serum creatinine have significant difference in different time periods (P<0.05). During the mid-late times of renal function impairment, total GFR and serum creatinine have no statistically significant differences (P value is 0.781, 0.297). 24-hour urine protein quality had no statistical differences in each stage. However: the total GFR is more sensitive than the serum creatinine in evaluation of early impairing of renal function. Conclusions: There is significant correlation between serum creatinine and total GFR. Both of them can reflect the degree of diabetic renal injury, but the total GFR is more sensitive than serum creatinine in early degree. 24-hour urine protein quantitative can not evaluate the degree of impaired renal function alone

  19. Determination of single-kidney glomerular filtration rate (GFR) with CT urography versus renal dynamic imaging Gates method

    International Nuclear Information System (INIS)

    You, Shan; Ma, XianWu; Zhang, ChangZhu; Li, Qiang; Shi, WenWei; Zhang, Jing; Yuan, XiaoDong

    2018-01-01

    To present a single-kidney CT-GFR measurement and compare it with the renal dynamic imaging Gates-GFR. Thirty-six patients with hydronephrosis referred for CT urography and 99mTc-DTPA renal dynamic imaging were prospectively included. Informed consent was obtained from all patients. The CT urography protocol included non-contrast, nephrographic, and excretory phase imaging. The total CT-GFR was calculated by dividing the CT number increments of the total urinary system between the nephrographic and excretory phase by the products of iodine concentration in the aorta and the elapsed time, then multiplied by (1- Haematocrit). The total CT-GFR was then split into single-kidney CT-GFR by a left and right kidney proportionality factor. The results were compared with single-kidney Gates-GFR by using paired t-test, correlation analysis, and Bland-Altman plots. Paired difference between single-kidney CT-GFR (45.02 ± 13.91) and single-kidney Gates-GFR (51.21 ± 14.76) was 6.19 ± 5.63 ml/min, p<0.001, demonstrating 12.1% systematic underestimation with ±11.03 ml/min (±21.5%) measurement deviation. A good correlation was revealed between both measurements (r=0.87, p<0.001). The proposed single-kidney CT-GFR correlates and agrees well with the reference standard despite a systematic underestimation, therefore it could be a one-stop-shop for evaluating urinary tract morphology and split renal function. (orig.)

  20. Development of multi-group xs libraries for the gfr 2400 reactor

    International Nuclear Information System (INIS)

    Cerba, Š.; Vrban, B.; Lüley, J.; Necas, V.

    2016-01-01

    GFR 2400 is considered as a conceptual design of the large scale GEN IV Gas-Cooled Fast Reactor. In general, the GEN IV technologies are seen as reliable but also very challenging reactor concepts. Since GFR 2400 lacks any experimental data, the questions on its safety are even more complex and the assessment of its performance could be made only based on computational experience. The paper deals with the development process of multi-group XS libraries based on a hybrid deterministic-Stochastic methodology, using the NJOY99, TRANSX, DIF3D, PARTISN and MCNP5 codes. A new optimized 25 group SBJ E 71 2 5G cross section library was developed based on ENDF/B-VII.1 evaluated data, ZZ-KAFAX-E70 background cross sections and GFR 2400 neutron spectrum. The created library was validated through integral experiments evaluated on the HEX-Z deterministic models in DIF3D. The results were also compared with MCNP5 calculations. (authors)

  1. Prediction of hospital mortality by changes in the estimated glomerular filtration rate (eGFR).

    LENUS (Irish Health Repository)

    Berzan, E

    2015-03-01

    Deterioration of physiological or laboratory variables may provide important prognostic information. We have studied whether a change in estimated glomerular filtration rate (eGFR) value calculated using the (Modification of Diet in Renal Disease (MDRD) formula) over the hospital admission, would have predictive value. An analysis was performed on all emergency medical hospital episodes (N = 61964) admitted between 1 January 2002 and 31 December 2011. A stepwise logistic regression model examined the relationship between mortality and change in renal function from admission to discharge. The fully adjusted Odds Ratios (OR) for 5 classes of GFR deterioration showed a stepwise increased risk of 30-day death with OR\\'s of 1.42 (95% CI: 1.20, 1.68), 1.59 (1.27, 1.99), 2.71 (2.24, 3.27), 5.56 (4.54, 6.81) and 11.9 (9.0, 15.6) respectively. The change in eGFR during a clinical episode, following an emergency medical admission, powerfully predicts the outcome.

  2. Analysis of a convection loop for GFR post-LOCA decay heat removal

    International Nuclear Information System (INIS)

    Williams, W.C.; Hejzlar, P.; Saha, P.

    2004-01-01

    A computer code (LOCA-COLA) has been developed at MIT for steady state analysis of convective heat transfer loops. In this work, it is used to investigate an external convection loop for decay heat removal of a post-LOCA gas-cooled fast reactor (GFR). The major finding is that natural circulation cooling of the GFR is feasible under certain circumstances. Both helium and CO 2 cooled system components are found to operate in the mixed convection regime, the effects of which are noticeable as heat transfer enhancement or degradation. It is found that CO 2 outdoes helium under identical natural circulation conditions. Decay heat removal is found to have a quadratic dependence on pressure in the laminar flow regime and linear dependence in the turbulent flow regime. Other parametric studies have been performed as well. In conclusion, convection cooling loops are a credible means for GFR decay heat removal and LOCA-COLA is an effective tool for steady state analysis of cooling loops. (authors)

  3. A comparative study on recycling spent fuels in gas-cooled fast reactors

    International Nuclear Information System (INIS)

    Choi, Hangbok; Baxter, Alan

    2010-01-01

    This study evaluates advanced Gas-cooled Fast Reactor (GFR) fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. A 600 MWth GFR was used for the fuel cycle analysis, and the equilibrium core was searched with different fuel-to-matrix volume ratios such as 70/30 and 60/40. Two fuel cycle scenarios, i.e., a one-tier case combining a Light Water Reactor (LWR) and a GFR, and a two-tier case using an LWR, a Very High Temperature Reactor (VHTR), and a GFR, were evaluated for mass flow and fuel cycle cost, and the results were compared to those of LWR once-through fuel cycle. The mass flow calculations showed that the natural uranium consumption can be reduced by more than 57% and 27% for the one-tier and two-tier cycles, respectively, when compared to the once-through fuel cycle. The transuranics (TRU) which pose a long-term problem in a high-level waste repository, can be significantly reduced in the multiple recycle operation of these options, resulting in more than 110 and 220 times reduction of TRU inventory to be geologically disposed for the one-tier and two-tier fuel cycles, respectively. The fuel cycle costs were estimated to be 9.4 and 8.6 USD/MWh for the one-tier fuel cycle when the GFR fuel-to-matrix volume ratio was 70/30 and 60/40, respectively. However the fuel cycle cost is reduced to 7.3 and 7.1 USD/MWh for the two-tier fuel cycle, which is even smaller than that of the once-through fuel cycle. In conclusion the GFR can provide alternative fuel cycle options to the once-through and other fast reactor fuel cycle options, by increasing the natural uranium utilization and reducing the fuel cycle cost.

  4. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

    Gietzen, A.

    1993-01-01

    Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with General Atomic's standard commercial warranty

  5. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

    Gietzen, A.

    1993-01-01

    Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium-zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched-uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with GA's standard commercial warranty

  6. Donor-estimated GFR as an appropriate criterion for allocation of ECD kidneys into single or dual kidney transplantation.

    Science.gov (United States)

    Snanoudj, R; Rabant, M; Timsit, M O; Karras, A; Savoye, E; Tricot, L; Loupy, A; Hiesse, C; Zuber, J; Kreis, H; Martinez, F; Thervet, E; Méjean, A; Lebret, T; Legendre, C; Delahousse, M

    2009-11-01

    It has been suggested that dual kidney transplantation (DKT) improves outcomes for expanded criteria donor (ECD) kidneys. However, no criteria for allocation to single or dual transplantation have been assessed prospectively. The strategy of DKT remains underused and potentially eligible kidneys are frequently discarded. We prospectively compared 81 DKT and 70 single kidney transplant (SKT) receiving grafts from ECD donors aged >65 years, allocated according to donor estimated glomerular filtration rate (eGFR): DKT if eGFR between 30 and 60 mL/min, SKT if eGFR greater than 60 mL/min. Patient and graft survival were similar in the two groups. In the DKT group, 13/81 patients lost one of their two kidneys due to hemorrhage, arterial or venous thrombosis. Mean eGFR at month 12 was similar in the DKT and SKT groups (47.8 mL/min and 46.4 mL/min, respectively). Simulated allocation of kidneys according to criteria based on day 0 donor parameters such as those described by Remuzzi et al., Andres et al. and UNOS, did not indicate an improvement in 12-month eGFR compared to our allocation based on donor eGFR.

  7. Serum bilirubin concentration is associated with eGFR and urinary albumin excretion in patients with type 1 diabetes mellitus.

    Science.gov (United States)

    Nishimura, Takeshi; Tanaka, Masami; Sekioka, Risa; Itoh, Hiroshi

    2015-01-01

    Although relationships of serum bilirubin concentration with estimated glomerular filtration rate (eGFR) and urinary albumin excretion (UAE) in patients with type 2 diabetes have been reported, whether such relationships exist in patients with type 1 diabetes is unknown. A total of 123 patients with type 1 diabetes were investigated in this cross-sectional study. The relationship between bilirubin (total and indirect) concentrations and log(UAE) as well as eGFR was examined by Pearson's correlation analyses. Multivariate regression analyses were used to assess the association of bilirubin (total and indirect) with eGFR as well as log(UAE). A positive correlation was found between serum bilirubin concentration and eGFR; total bilirubin (r=0.223, p=0.013), indirect bilirubin (r=0.244, p=0.007). A negative correlation was found between serum bilirubin concentration and log(UAE); total bilirubin (r=-0.258, p=0.005), indirect bilirubin (r=-0.271, p=0.003). Multivariate regression analyses showed that indirect bilirubin concentration was an independent determinant of eGFR and log(UAE). Bilirubin concentration is associated with both eGFR and log(UAE) in patients with type 1 diabetes. Bilirubin might have a protective role in the progression of type 1 diabetic nephropathy. Copyright © 2015 Elsevier Inc. All rights reserved.

  8. Estimating GFR in children with 99mTc-DTPA renography

    DEFF Research Database (Denmark)

    Gutte, Henrik; Møller, Michael L; Pfeifer, Andreas K

    2010-01-01

    study was to evaluate the accuracy of this non-invasive method in children. We calculated GFR from (99m)Tc-diethylene triamine pentaacetic acid (DTPA) renography and compared with (51)Cr-EDTA plasma clearance of 29 children between the age of 1 month and 12 years (mean 4.7 years). The correlation...... between (99m)Tc-DTPA renography and (51)Cr-EDTA plasma clearance was for all children R = 0.96 (n = 29, PTc-DTPA renography is reliable...

  9. STUDI ANALISA KOORDINASI RELAI GFR INCOMING BUSBAR 20 KV DAN GFR SALURAN UNTUK MENGAMANKAN GANGGUAN SATU PHASA KETANAH DI TRANSFORMATOR 3 GARDU INDUK KAPAL

    Directory of Open Access Journals (Sweden)

    I Gede Krisnayoga Kusuma

    2017-08-01

    Full Text Available Disruption of the distribution system is generally caused by a short circuit that causes overcurrent, one single phase short circuit to ground. Single phase to ground disturbances that occurred at the feeder should be secured by a ground fault relays in feeders. However, due to an error of coordination, interference with the feeder perceived also to the incoming side of the transformer so that ground fault relays on the incoming 20 kV ordered PMT 20 kV at the transformer side open. Working time ground fault relays in feeders Peguyangan is for 0.25 seconds. While working time ground fault relays in the incoming 20 kV is for 0.5 seconds. This indicates that differences in work time relay by 0.25 seconds is considered selective. The event of disruption of the ground phase, GFR relay coordination in securing the area of ??interference must be gradual. Where the value of working time ground fault relays in the base should be longer than the time ground fault relays working on a feeder. It aims to work ground fault relays working order, so that errors can be avoided in the security and do not lead to the spread of disturbed areas.

  10. Study on two-phase flow in a coolant channel of a plate-type fuel with use of neutron radiography technique

    International Nuclear Information System (INIS)

    Mishima, K.; Hibiki, T.; Nishihara, H.

    1992-01-01

    Two-phase flow in a narrow rectangular duct is important related to abnormal cooling conditions of a MTR type research reactor. In view of this, flow regime, void fraction, slug bubble velocity and pressure loss were measured for rectangular ducts with a narrow gap. The neutron radiography technique was used to visualize the flow and the void fraction was obtained by image processing. The void fraction was correlated well by the drift flux model with existing correlation for the distribution parameter which was about 1.35. Similar results were obtained for slug bubble velocity, however the distribution parameter was in the range from 1.0 to 1.2. The frictional pressure loss was correlated well by the Chisholm-Laird correlation. In collaboration with previously obtained data, it was found that the Chisholm's parameter C, however, changed from 21 to zero as the gap decreased. (author)

  11. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  12. Prediction of renal function (GFR) from cystatin C and creatinine in children: Body cell mass increases accuracy of the estimate

    DEFF Research Database (Denmark)

    Andersen, Trine Borup; Jødal, Lars; Bøgsted, Martin

    ) aged 2-14 years (mean 8.8 years). GFR was 14-147 mL/min/1.73m2 (mean 97 mL/min/1.73m2). BCM was estimated using bioimpedance spectroscopy (Xitron Hydra 4200). Log-transformed data on BCM/CysC, serum creatinine (SCr), body-surface-area (BSA), height x BSA/SCr, serum CysC, weight, sex, age, height, serum....... The present equation also had the highest R2 and the narrowest 95% limits of agreement. CONCLUSION: The new equation predicts GFR with higher accuracy than other equations. Endogenous methods are, however, still not accurate enough to replace exogenous markers when GFR must be determined with high accuracy...

  13. GFR prediction from cystatin C and creatinine in children: body cell mass increases accuracy of the estimate

    DEFF Research Database (Denmark)

    Andersen, Trine Borup; Jødal, Lars; Bøgsted, Martin

    ) aged 2-14 years (mean 8.8 years). GFR was 14-147 mL/min/1.73m2 (mean 97 mL/min/1.73m2). BCM was estimated using bioimpedance spectroscopy (Xitron Hydra 4200). Log-transformed data on BCM/CysC, serum creatinine (SCr), body-surface-area (BSA), height x BSA/SCr, serum CysC, weight, sex, age, height, serum....... The present equation also had the highest R2 and the narrowest 95% limits of agreement. CONCLUSION: The new equation predicts GFR with higher accuracy than other equations. Endogenous methods are, however, still not accurate enough to replace exogenous markers when GFR must be determined with high accuracy...

  14. Status of research reactor fuel development in KAERI

    International Nuclear Information System (INIS)

    Kim, Chang-Kyu; Ryu, Woo-Seok; Park, Jong-Man; Lee, Don-Bae; Kim, Ki-Hwan; Kuk, Il-Hyun

    1996-01-01

    The development of uranium silicide dispersion fuel fabrication technology has been carried out in KAERI. LEU fuel bundle was prepared for irradiation test. In order to compare the performance of atomized and comminuted U 3 Si dispersed fuels, the bundle of two kinds of fuel elements were prepared. Irradiation test will be performed in the OR-hole of HANARO in the near future. U 3 Si 2 atomization technology has been improved by using ceramic crucible and nozzle. Irradiation test for atomized U 3 Si 2 plate type fuel will be carried out in cooperation with ANL by using HANARO in connection with RERTR advanced fuel development. (author)

  15. Risk-informed analysis as a support to the preliminary design of the CEA GFR2400

    International Nuclear Information System (INIS)

    Bertrand, F.; Bassi, C.; Azria, P.; Bentivoglio, F.; Messie, A.; Balmain, M.

    2012-01-01

    The integration of safety issues in the early phase of the design of a 4. generation reactor of the concepts is expected. For this purpose, probabilistic insights are increasingly employed in the safety demonstration in combination with the deterministic approach in the frame of a so-called risk informed approach. The present paper deals with the safety assessment of the preliminary design of the GFR2400 developed by CEA and how it has been improved in order to fulfil deterministic criteria as well as to reach a risk level comparable to the generation III reactors. GFR2400 is a 2400 MWth, 3-loops, helium-cooled fast reactor developed at a pre-conceptual design stage whose secondary circuit is filled with a mixture of helium and nitrogen, the ternary circuit being filled with water vaporized in 3 steam generators according to a classical Rankine cycle. The resulting cycle efficiency is very close to 45 %. Considering the results obtained with a preliminary level 1 PSA (L1PSA) model, it emerged that an increased reliability of the DHR (Decay Heat Removal) function in high pressure conditions (not corresponding to a LOCA) was suitable to reduce the overall core damage frequency. On the other hand, some small break LOCA situations were not adequately mitigated according to the line of protection deterministic method. Both issues have been solved by design improvements. In addition, this final L1PSA model, characterized by success criteria based on transient calculations performed with the CATHARE2 code and performed in a perimeter extended to all representative internal initiating events at full operating power, permitted to propose design evolutions that did not increase significantly the CDF. In the same time, those evolutions enabled the DHR system to increase its redundancy level as required in the deterministic approach. Finally, a modified design has been reached implying a more extended covering of various accidental situations by means of a progressive DHR

  16. Creatinine, eGFR and association with myocardial infarction, ischemic heart disease and early death in the general population

    DEFF Research Database (Denmark)

    Sibilitz, Kirstine L; Benn, Marianne; Nordestgaard, Børge G

    2014-01-01

    OBJECTIVE: We tested the hypothesis that moderately elevated plasma creatinine levels and decreased levels of estimated glomerular filtration rate (eGFR) are associated with increased risk of myocardial infarction, ischemic heart disease, and early death in the general population. METHODS: We...... studied 10,489 individuals with a plasma creatinine measurement and calculated eGFR from the Danish general population, of which 1498 developed myocardial infarction, 3001 ischemic heart disease, and 7573 died during 32 years follow-up. RESULTS: Cumulative incidences of myocardial infarction and ischemic...... heart disease as a function of age increased with increasing levels of creatinine, and survival decreased (log-rank trends: creatinine levels

  17. Thin-plate-type embedded ultrasonic transducer based on magnetostriction for the thickness monitoring of the secondary piping system of a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Tae Hoon; Cho, Seung Hyun [Center for Safety Measurement, Korea Research Institute of Standards and Science, Daejeon (Korea, Republic of)

    2016-12-15

    Pipe wall thinning in the secondary piping system of a nuclear power plant is currently a major problem that typically affects the safety and reliability of the nuclear power plant directly. Regular in-service inspections are carried out to manage the piping system only during the overhaul. Online thickness monitoring is necessary to avoid abrupt breakage due to wall thinning. To this end, a transducer that can withstand a high-temperature environment and should be installed under the insulation layer. We propose a thin plate type of embedded ultrasonic transducer based on magnetostriction. The transducer was designed and fabricated to measure the thickness of a pipe under a high-temperature condition. A number of experimental results confirmed the validity of the present transducer.

  18. An anodic alumina supported Ni-Pt bimetallic plate-type catalysts for multi-reforming of methane, kerosene and ethanol

    KAUST Repository

    Zhou, Lu

    2014-05-01

    An anodic alumina supported Ni-Pt bimetallic plate-type catalyst was prepared by a two-step impregnation method. The trace amount 0.08 wt% of Pt doping efficiently suppressed the nickel particle sintering and improved the nickel oxides reducibility. The prepared Ni-Pt catalyst showed excellent performance during steam reforming of methane, kerosene and ethanol under both 3000 h stationary and 500-time daily start-up and shut-down operation modes. Self-activation ability of this catalyst was evidenced, which was considered to be resulted from the hydrogen spillover effect over Ni-Pt alloy. In addition, an integrated combustion-reforming reactor was proposed in this study. However, the sintering of the alumina support is still a critical issue for the industrialization of Ni-Pt catalyst. Copyright © 2014, Hydrogen Energy Publications, LLC. Published by Elsevier Ltd. All rights reserved.

  19. An application of time-frequency signal analysis technique to estimate the location of an impact source on a plate type structure

    International Nuclear Information System (INIS)

    Park, Jin Ho; Lee, Jeong Han; Choi, Young Chul; Kim, Chan Joong; Seong, Poong Hyun

    2005-01-01

    It has been reviewed whether it would be suitable that the application of the time-frequency signal analysis techniques to estimate the location of the impact source in plate structure. The STFT(Short Time Fourier Transform), WVD(Wigner-Ville distribution) and CWT(Continuous Wavelet Transform) methods are introduced and the advantages and disadvantages of those methods are described by using a simulated signal component. The essential of the above proposed techniques is to separate the traveling waves in both time and frequency domains using the dispersion characteristics of the structural waves. These time-frequency methods are expected to be more useful than the conventional time domain analyses for the impact localization problem on a plate type structure. Also it has been concluded that the smoothed WVD can give more reliable means than the other methodologies for the location estimation in a noisy environment

  20. 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.

  1. 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

  2. A Meta-analysis of the Association of Estimated GFR, Albuminuria, Diabetes Mellitus, and Hypertension With Acute Kidney Injury

    NARCIS (Netherlands)

    James, Matthew T.; Grams, Morgan E.; Woodward, Mark; Elley, C. Raina; Green, Jamie A.; Wheeler, David C.; de Jong, Paul; Gansevoort, Ron T.; Levey, Andrew S.; Warnock, David G.; Sarnak, Mark J.; de Zeeuw, Dick; Bakker, Stephan J. L.; van der Harst, Pim; Heerspink, Hiddo J.

    2015-01-01

    Background: Diabetes mellitus and hypertension are risk factors for acute kidney injury (AKI). Whether estimated glomerular filtration rate (eGFR) and urine albumin-creatinine ratio (ACR) remain risk factors for AKI in the presence and absence of these conditions is uncertain. Study Design:

  3. Prediction of renal function (GFR) from cystatin C and creatinine in children: Body cell mass increases accuracy of the estimate

    DEFF Research Database (Denmark)

    Andersen, Trine Borup; Jødal, Lars; Bøgsted, Martin

    using robust regression in a forward, stepwise procedure. GFR (mL/min) was the dependent variable. The accuracy and precision of the prediction model were compared to other prediction models from the literature, using k-fold cross-validation. Local constants and coefficients were calculated for all...

  4. Preliminary Findings of Serum Creatinine and Estimated Glomerular Filtration Rate (eGFR) in Adolescents with Intellectual Disabilities

    Science.gov (United States)

    Lin, Jin-Ding; Lin, Lan-Ping; Hsieh, Molly; Lin, Pei-Ying

    2010-01-01

    The present study aimed to describe the kidney function profile--serum creatinine and estimated glomerular filtration rate (eGFR), and to examine the relationships of predisposing factors to abnormal serum creatinine in people with intellectual disabilities (ID). Data were collected by a cross-sectional study of 827 aged 15-18 years adolescents…

  5. Can baseline serum creatinine and e-GFR predict renal function outcome after augmentation cystoplasty in children?

    Science.gov (United States)

    Singh, Prempal; Bansal, Ankur; Sekhon, Virender; Nunia, Sandeep; Ansari, M S

    2018-01-01

    To assess cut-off value of creatinine and glomerular filtration rate for augmentation cystoplasty (AC) in paediatric age-group. Data of all paediatric-patients (Creatinine and e-GFR were assessed at the time of surgery, at 6 months and at last follow-up. Renal function deterioration was defined as increase in creatinine by ≥25% from baseline value or new-onset stage-3 CKD or worsening of CKD stage with pre-operative-CKD stage-3. ROCs were plotted using creatinine and e-GFR for AC. A total of 94 patients with mean-age 8.9 years were included. The mean creatinine and e-GFR were 1.33mg/dL and 57.68mL/min respectively. Out of 94 patients, AC was performed in 45 patients and in the remaining 49 patients AC was not done (control-group), as they were not willing for the same. Baseline patient's characteristics were comparable in both Groups. 22 underwent gastro-cystoplasty (GC) and 25 underwent ileo-cystoplasty (IC). Decline in renal function was observed in 15 (33.3%) patients of AC-group and in 31 (63.3%) patients of control-group. Patients having creatinine ≥1.54mg/dL (P=0.004, sensitivity (S) 63.6% and specificity (s) 90.5%) at baseline and e-GFR ≤46mL/min (P=0.000, S=100% and s=85.7%) at the time of surgery had significantly increased probability of renal function deterioration on follow-up after AC. e-GFR ≤46mL/min and creatinine ≥1.54mg/dL at time of surgery could serve as a predictor of renal function deterioration in AC in paediatric patients. Copyright® by the International Brazilian Journal of Urology.

  6. Morphometry Predicts Early GFR Change in Primary Proteinuric Glomerulopathies: A Longitudinal Cohort Study Using Generalized Estimating Equations.

    Directory of Open Access Journals (Sweden)

    Kevin V Lemley

    Full Text Available Most predictive models of kidney disease progression have not incorporated structural data. If structural variables have been used in models, they have generally been only semi-quantitative.We examined the predictive utility of quantitative structural parameters measured on the digital images of baseline kidney biopsies from the NEPTUNE study of primary proteinuric glomerulopathies. These variables were included in longitudinal statistical models predicting the change in estimated glomerular filtration rate (eGFR over up to 55 months of follow-up.The participants were fifty-six pediatric and adult subjects from the NEPTUNE longitudinal cohort study who had measurements made on their digital biopsy images; 25% were African-American, 70% were male and 39% were children; 25 had focal segmental glomerular sclerosis, 19 had minimal change disease, and 12 had membranous nephropathy. We considered four different sets of candidate predictors, each including four quantitative structural variables (for example, mean glomerular tuft area, cortical density of patent glomeruli and two of the principal components from the correlation matrix of six fractional cortical areas-interstitium, atrophic tubule, intact tubule, blood vessel, sclerotic glomerulus, and patent glomerulus along with 13 potentially confounding demographic and clinical variables (such as race, age, diagnosis, and baseline eGFR, quantitative proteinuria and BMI. We used longitudinal linear models based on these 17 variables to predict the change in eGFR over up to 55 months. All 4 models had a leave-one-out cross-validated R2 of about 62%.Several combinations of quantitative structural variables were significantly and strongly associated with changes in eGFR. The structural variables were generally stronger than any of the confounding variables, other than baseline eGFR. Our findings suggest that quantitative assessment of diagnostic renal biopsies may play a role in estimating the baseline

  7. International Collaboration for the Epidemiology of eGFR in Low and Middle Income Populations - Rationale and core protocol for the Disadvantaged Populations eGFR Epidemiology Study (DEGREE).

    Science.gov (United States)

    Caplin, Ben; Jakobsson, Kristina; Glaser, Jason; Nitsch, Dorothea; Jha, Vivekanand; Singh, Ajay; Correa-Rotter, Ricardo; Pearce, Neil

    2017-01-03

    There is an increasing recognition of epidemics of primarily tubular-interstitial chronic kidney disease (CKD) clustering in agricultural communities in low- and middle-income countries (LMICs). Although it is currently unclear whether there is a unified underlying aetiology, these conditions have been collectively termed CKD of undetermined cause (CKDu). CKDu is estimated to have led to the premature deaths of tens to hundreds of thousands of young men and women over the last 2 decades. Thus, there is an urgent need to understand the aetiology and pathophysiology of these condition (s). International comparisons have provided the first steps in understanding many chronic diseases, but such comparisons rely on the availability of standardised tools to estimate disease prevalence. This is a particular problem with CKD, since the disease is asymptomatic until the late stages, and the biases inherent in the methods used to estimate the glomerular filtration rate (GFR) in population studies are highly variable across populations. We therefore propose a simple standardised protocol to estimate the distribution of GFR in LMIC populations - The Disadvantaged Populations eGFR Epidemiology (DEGREE) Study. This involves the quantification of renal function in a representative adult population-based sample and a requirement for standardisation of serum creatinine measurements, along with storage of samples for future measurements of cystatin C and ascertainment of estimates of body composition, in order to obtain valid comparisons of estimated GFR (eGFR) within and between populations. The methodology we present is potentially applicable anywhere, but our particular focus is on disadvantaged populations in LMICs, since these appear to be most susceptible to CKDu. Although the protocol could also be used in specific groups (e.g. occupational groups, thought to be at excess risk of CKDu) the primary aim of the DEGREE project is characterise the population distribution of eGFR

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  9. High rates of albuminuria but not of low eGFR in Urban Indigenous Australians: the DRUID Study

    Directory of Open Access Journals (Sweden)

    Zimmet Paul Z

    2011-05-01

    Full Text Available Abstract Background Indigenous Australians have an incidence of end stage kidney disease 8-10 times higher than non-Indigenous Australians. The majority of research studies concerning Indigenous Australians have been performed in rural or remote regions, whilst the majority of Indigenous Australians actually live in urban settings. We studied prevalence and factors associated with markers of kidney disease in an urban Indigenous Australian cohort, and compared results with those for the general Australian population. Methods 860 Indigenous adult participants of the Darwin Region Urban Indigenous Diabetes (DRUID Study were assessed for albuminuria (urine albumin-creatinine ratio≥2.5 mg/mmol males, ≥3.5 mg/mmol females and low eGFR (estimated glomular filtration rate 2. Associations between risk factors and kidney disease markers were explored. Comparison was made with the AusDiab cohort (n = 8,936 aged 25-64 years, representative of the general Australian adult population. Results A high prevalence of albuminuria (14.8% was found in DRUID, whilst prevalence of low eGFR was 2.4%. Older age, higher HbA1c, hypertension, higher C-reactive protein and current smoking were independently associated with albuminuria on multiple regression. Low eGFR was independently associated with older age, hypertension, albuminuria and higher triglycerides. Compared to AusDiab participants, DRUID participants had a 3-fold higher adjusted risk of albuminuria but not of low eGFR. Conclusions Given the significant excess of ESKD observed in Indigenous versus non-Indigenous Australians, these findings could suggest either: albuminuria may be a better prognostic marker of kidney disease than low eGFR; that eGFR equations may be inaccurate in the Indigenous population; a less marked differential between Indigenous and non-Indigenous Australians for ESKD rates in urban compared to remote regions; or that differences in the pathophysiology of chronic kidney disease exist

  10. Modification of a two blood sample method used for measurement of GFR with 99mTc-DTPA.

    Science.gov (United States)

    Surma, Marian J; Płachcińska, Anna; Kuśmierek, Jacek

    2018-01-01

    Measurements of GFR may be performed with a slope/intercept method (S/I), using only two blood samples taken in strictly defined time points. The aim of the study was to modify this method in order to extend time intervals suitable for blood sampling. Modification was based on a variation of a Russel et al. model parameter, selection of time intervals suitable for blood sampling and assessment of uncertainty of calculated results. Archived values of GFR measurements of 169 patients with different renal function, from 5.5 to 179 mL/min, calculated with a multiple blood sample method were used. Concentrations of a radiopharmaceutical in consecutive minutes, from 60th to 190th after injection, were calculated theoretically, using archived parameters of biexponential functions describing a decrease in 99mTc-DTPA concentration in blood plasma with time. These values, together with injected activities, were treated as measurements and used for S/I clearance calculations. Next, values of S/I clearance were compared with the multiple blood sample method in order to calculate suitable values of exponent present in a Russel's model, for every combination of two blood sampling time points. A model was considered accurately fitted to measured values when SEE ≤ 3.6 mL/min. Assessments of uncertainty of obtained results were based on law of error superposition, taking into account mean square prediction error and also errors introduced by pipetting, time measurement and stochastic radioactive decay. The accepted criteria resulted in extension of time intervals suitable for blood sampling to: between 60 and 90 minutes after injection for the first sample and between 150 and 180 minutes for the second sample. Uncertainty of results was assessed as between 4 mL/min for GFR = 5-10 mL/min and 8 mL/min for GFR = 180 mL/min. Time intervals accepted for blood sampling fully satisfy nuclear medicine staff and ensure proper determination of GFR. Uncertainty of results is entirely

  11. Estimating renal function in children: a new GFR-model based on serum cystatin C and body cell mass.

    Science.gov (United States)

    Andersen, Trine Borup

    2012-07-01

    This PhD thesis is based on four individual studies including 131 children aged 2-14 years with nephro-urologic disorders. The majority (72%) of children had a normal renal function (GFR > 82 ml/min/1.73 square metres), and only 8% had a renal function thesis´ main aims were: 1) to develop a more accurate GFR model based on a novel theory of body cell mass (BCM) and cystatin C (CysC); 2) to investigate the diagnostic performance in comparison to other models as well as serum CysC and creatinine; 3) to validate the new models precision and validity. The model´s diagnostic performance was investigated in study I as the ability to detect changes in renal function (total day-to-day variation), and in study IV as the ability to discriminate between normal and reduced function. The model´s precision and validity were indirectly evaluated in study II and III, and in study I accuracy was estimated by comparison to reference GFR. Several prediction models based on CysC or a combination of CysC and serum creatinine have been developed for predicting GFR in children. Despite these efforts to improve GFR estimates, no alternative to exogenous methods has been found and the Schwartz´s formula based on height, creatinine and an empirically derived constant is still recommended for GFR estimation in children. However, the inclusion of BCM as a possible variable in a CysC-based prediction model has not yet been explored. As CysC is produced at a constant rate from all nucleated cells we hypothesize that including BCM in a new prediction model will increase accuracy of the GFR estimate. Study I aimed at deriving the new GFR-prediction model based on the novel theory of CysC and BCM and comparing the performance to previously published models. The BCM-model took the form GFR (mL/min) = 10.2 × (BCM/CysC)E 0.40 × (height × body surface area/Crea)E 0.65. The model predicted 99% within ± 30% of reference GFR, and 67% within ±10%. This was higher than any other model. The

  12. No-contact method of determining average working-surface temperature of plate-type radiation-absorbing thermal exchange panels of flat solar collectors for heating heat-transfer fluid

    International Nuclear Information System (INIS)

    Avezova, N.R.; Avezov, R.R.

    2015-01-01

    A brand new no-contact method of determining the average working-surface temperature of plate-type radiation-absorbing thermal exchange panels (RATEPs) of flat solar collectors (FSCs) for heating a heat-transfer fluid (HTF) is suggested on the basis of the results of thermal tests in full-scale quasistationary conditions. (authors)

  13. Physical model of the nuclear fuel cycle simulation code SITON

    International Nuclear Information System (INIS)

    Brolly, Á.; Halász, M.; Szieberth, M.; Nagy, L.; Fehér, S.

    2017-01-01

    Finding answers to main challenges of nuclear energy, like resource utilisation or waste minimisation, calls for transient fuel cycle modelling. This motivation led to the development of SITON v2.0 a dynamic, discrete facilities/discrete materials and also discrete events fuel cycle simulation code. The physical model of the code includes the most important fuel cycle facilities. Facilities can be connected flexibly; their number is not limited. Material transfer between facilities is tracked by taking into account 52 nuclides. Composition of discharged fuel is determined using burnup tables except for the 2400 MW thermal power design of the Gas-Cooled Fast Reactor (GFR2400). For the GFR2400 the FITXS method is used, which fits one-group microscopic cross-sections as polynomial functions of the fuel composition. This method is accurate and fast enough to be used in fuel cycle simulations. Operation of the fuel cycle, i.e. material requests and transfers, is described by discrete events. In advance of the simulation reactors and plants formulate their requests as events; triggered requests are tracked. After that, the events are simulated, i.e. the requests are fulfilled and composition of the material flow between facilities is calculated. To demonstrate capabilities of SITON v2.0, a hypothetical transient fuel cycle is presented in which a 4-unit VVER-440 reactor park was replaced by one GFR2400 that recycled its own spent fuel. It is found that the GFR2400 can be started if the cooling time of its spent fuel is 2 years. However, if the cooling time is 5 years it needs an additional plutonium feed, which can be covered from the spent fuel of a Generation III light water reactor.

  14. STUDI ANALISIS KOORDINASI OVER CURRENT RELAY (OCR DAN GROUND FAULT RELAY (GFR PADA RECLOSER DI SALURAN PENYULANG PENEBEL

    Directory of Open Access Journals (Sweden)

    I Dewa Gde Agung Budhi Udiana

    2017-08-01

    Full Text Available Short circuit causing over current problem and can might causing interference of the equipment performance such as distribution transformers also causing widespread disruption occurred. In resolving such interference is required as protection system on the distribution system. Seeing all above is needed coordination between the supporting component of the protection system which is consisted of Over Current Relay (OCR and Ground Fault Relay (GFR. The research was conducted at PT. PLN (Persero South Bali Area Network, INDONESIA on recloser in the feeder line of Penebel. OCR setting between the Relay feeder of Penebel, Recloser Celagi, Recloser Bakisan, and Recloser Benana still less selective, with time value coordination between average security was still less than 0,2 second. Then OCR setting and GFR relay feeder of Penebel, Recloser Celagi, Recloser Bakisan, and Recloser Benana was recommended for re-setting in order to minimize disruption and electric power distribution system to be reliable.

  15. The association between creatinine versus cystatin C-based eGFR and cardiovascular risk in children with chronic kidney disease using a modified PDAY risk score.

    Science.gov (United States)

    Sharma, Sheena; Denburg, Michelle R; Furth, Susan L

    2017-08-01

    Children with chronic kidney disease (CKD) have a high prevalence of cardiovascular disease (CVD) risk factors which may contribute to the development of cardiovascular events in adulthood. Among adults with CKD, cystatin C-based estimates of glomerular filtration rate (eGFR) demonstrate a stronger predictive value for cardiovascular events than creatinine-based eGFR. The PDAY (Pathobiological Determinants of Atherosclerosis in Youth) risk score is a validated tool used to estimate the probability of advanced coronary atherosclerotic lesions in young adults. To assess the association between cystatin C-based versus creatinine-based eGFR (eGFR cystatin C and eGFR creatinine, respectively) and cardiovascular risk using a modified PDAY risk score as a proxy for CVD in children and young adults. We performed a cross-sectional study of 71 participants with CKD [median age 15.5 years; inter-quartile range (IQR) 13, 17], and 33 healthy controls (median age 15.1 years; IQR 13, 17). eGFR was calculated using age-appropriate creatinine- and cystatin C-based formulas. Median eGFR creatinine and eGFR cystatin C for CKD participants were 50 (IQR 30, 75) and 53 (32, 74) mL/min/1.73 m 2 , respectively. For the healthy controls, median eGFR creatinine and eGFR cystatin were 112 (IQR 85, 128) and 106 mL/min/1.73m 2 (95, 123) mL/min/1.73 m 2 , respectively. A modified PDAY risk score was calculated based on sex, age, serum lipoprotein concentrations, obesity, smoking status, hypertension, and hyperglycemia. Modified PDAY scores ranged from -2 to 20. The Spearman's correlations of eGFR creatinine and eGFR cystatin C with coronary artery PDAY scores were -0.23 (p = 0.02) and -0.28 (p = 0.004), respectively. Ordinal logistic regression also showed a similar association of higher eGFR creatinine and higher eGFR cystatin C with lower PDAY scores. When stratified by age creatinine and eGFR cystatin C with PDAY score were modest and similar in children [-0.29 (p = 0.008) vs. -0.32 (p = 0

  16. Incidence of Hypoglycemia in Patients With Low eGFR Treated With Insulin and Dextrose for Hyperkalemia.

    Science.gov (United States)

    Pierce, Dwayne A; Russell, Greg; Pirkle, James L

    2015-12-01

    Hyperkalemia is a potentially life-threatening condition that is common in kidney disease patients. Insulin is used to treat hyperkalemia, but may cause hypoglycemia, especially in kidney disease when insulin may be metabolized more slowly. We compared the rates of hypoglycemia in patients with low estimated glomerular filtration rate (eGFR) using high versus low doses of insulin for hyperkalemia to determine if lower doses of insulin would decrease the incidence of hypoglycemia. This was a retrospective study of hospitalized patients receiving intravenous insulin for hyperkalemia during a 6-month period. Patients with low eGFR were analyzed based on how much insulin they received: high dose (10 units, n = 78) versus low dose (5 units, n = 71). Postdose nadir blood glucose values were examined for up to 8 hours after the dose. The percentage of hypoglycemia (blood glucose ≤70 mg/dl) and a subset of severe hypoglycemia (blood glucose <50 mg/dl) were then reported for each dose group. A total of 149 doses were identified in patients with low eGFR. The rates of hypoglycemia were 16.7% and 19.7% (P = 0.79), respectively, among high-dose (n = 78) and low-dose (n = 71) groups. Rates of severe hypoglycemia were 8.9% and 7.0%, respectively (P = 0.90). More than 28% of hypoglycemic episodes with high doses occurred after 4 hours (median = 2.5 hours) compared with 14.3% with low doses (median = 2.38 hours). There was no difference in the rate of hypoglycemia or severe hypoglycemia between high or low doses of insulin in patients with low eGFR. We recommend monitoring up to 6 hours after insulin use in hyperkalemia. © The Author(s) 2015.

  17. STUDI ANALISIS KOORDINASI OVER CURRENT RELAY (OCR) DAN GROUND FAULT RELAY (GFR) PADA RECLOSER DI SALURAN PENYULANG PENEBEL

    OpenAIRE

    I Dewa Gde Agung Budhi Udiana; I G Dyana Arjana; Tjok Gede Indra Partha

    2017-01-01

    Short circuit causing over current problem and can might causing interference of the equipment performance such as distribution transformers also causing widespread disruption occurred. In resolving such interference is required as protection system on the distribution system. Seeing all above is needed coordination between the supporting component of the protection system which is consisted of Over Current Relay (OCR) and Ground Fault Relay (GFR). The research was conducted at PT. PLN (Perse...

  18. Accuracy of GFR estimation formula in determination of glomerular filtration rate in kidney donors: Comparison with 24 h urine creatinine clearance

    Directory of Open Access Journals (Sweden)

    Abdul Rauf Hafeez

    2016-01-01

    Full Text Available To determine the accuracy of estimated glomerular filtration rate (eGFR using the modification of diet in renal disease (MDRD, Cockcroft-Gault (CG, and chronic kidney disease epidemiology (CKD-EPI formulas in potential kidney donors compared with 24-h urine creatinine clearance, we studied 207 potential live kidney donors in our center. There were 126 (60.9% males and 81 (39.1% females. Male:female ratio was 1.6:1. The age of the donors ranged from 18-58 years, with mean age of 35.30 ± 9.23 years and most of the individuals were below 40 years of age. The body mass index (BMI was calculated and venous blood samples were obtained for the measurement of serum creatinine and every study participant was instructed to collect 24-h urine. GFR was calculated based on 24-h urine creatinine clearance and the formulas. The accuracy of GFR estimation formula was taken as positive if the GFR calculated by the formulas and urine creatinine clearance fell between 90-120 mL/min/1.73 m 2 . The accuracy of the MDRD formula was 48.8% and the CG formula was 41.5% whereas the accuracy of the CKD-EPI formula was 78.2%. The accuracy of the eGFR using the MDRD formula was significantly higher in males than females (57.9% vs. 33.3% P = 0.001, while there was no statistically significant difference in the eGFR between them in case of the use of the CG and the CKD-EPI formulas. BMI and obesity had no effect on the accuracy of eGFR by the use of the different formulas. The performance of GFR estimation formulas was sub optimal and these either underestimated and/or over-estimated the GFR in healthy subjects. CKD-EPI is closer to 24 -h urinary creatinine clearance in the calculation of eGFR. However, none of the eGFR formulas can be used in renal transplant donors because of their low accuracy, and 24-h urine creatinine clearance should be used for evaluation of the GFR in this population.

  19. Current design efforts for the gas-cooled fast reactor (GFR)

    International Nuclear Information System (INIS)

    Weaver, K.D.

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GCFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFC I) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GCFR: a helium-cooled, direct Brayton cycle power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GCFR. These are EURATOM (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, EURATOM (including the United Kingdom), France, Japan, and Switzerland have active research activities with respect to the GCFR. The research includes GCFR design and safety, and fuels/in-core materials/fuel cycle projects. This paper outlines the current design status of the GCFR, and includes work done in the areas mentioned above. (Author)

  20. A protein diet score, including plant and animal protein, investigating the association with HbA1c and eGFR - the PREVIEW project

    DEFF Research Database (Denmark)

    Møller, Grith; Sluik, Diewertje; Ritz, Christian

    2017-01-01

    with glycated haemoglobin (HbA1c) and estimated glomerular filtration rate (eGFR). Analyses were based on three population studies included in the PREVIEW project (PREVention of diabetes through lifestyle Intervention and population studies in Europe and around the World): NQplus, Lifelines, and the Young Finns.......02 ± 0.01 mmol/mol, p eGFR in Lifelines (slope 0.17 ± 0.02 mL/min/1.73 m², p

  1. Development and application of Siton, a new fuel cycle simulation code

    International Nuclear Information System (INIS)

    Brolly, Aron; Szieberth, Mate; Halasz, Mate; Nagy, Lajos; Feher, Sandor

    2015-01-01

    As the result of the co-operation between the Centre for Energy Research (EK) and the Institute of Nuclear Techniques (NTI) a new fuel cycle simulation code called SITON was developed. Physical model of the code takes into account six facilities of the nuclear fuel cycle namely material stocks, spent fuel interim storages, plants for uranium enrichment, fuel fabrication, spent fuel reprocessing and reactors. Facilities can be linked in a flexible manner and their number is not limited. Lag time of the facilities and cooling time of the spent fuel, which are the two main parameters to introduce lag time into the fuel cycle, are taken into account. Material transfer between the facilities is modelled in a discrete manner tracking 52 nuclides and their short-lived decay daughters. Composition of the discharged fuel is determined by means of burn-up tables except for the 2400 MWth design of gas cooled fast reactor (GFR2400) which has a separate burn-up module developed at the NTI. To demonstrate the capabilities of SITON introduction of a GFR2400 into the Hungarian reactor park using the legacy spent fuel of the four presently operating VVER-440 units was simulated. 2040 was assumed as the commissioning date of the GFR2400 and recycling of its fuel was started as soon as possible. It was found that the plutonium content of the legacy spent fuel is sufficient to the start-up of only one GFR2400. There is an intermediate period between the commissioning of the reactor and the recycling of its first discharged fuel. Plutonium need of this period can be covered by the legacy spent fuel if the cooling time of the spent GFR2400 fuel is 2 years. If the cooling time is 5 years there will be a lack of plutonium in this period. To counterbalance this lack an EPR was started before the GFR2400 and its spent fuel was accumulated and reprocessed. Cooling time of the spent EPR fuel was also varied. Finally, an EPR only scenario is presented using two EPRs as a reference case

  2. A Meta-analysis of the Association of Estimated GFR, Albuminuria, Diabetes Mellitus, and Hypertension With Acute Kidney Injury.

    Science.gov (United States)

    James, Matthew T; Grams, Morgan E; Woodward, Mark; Elley, C Raina; Green, Jamie A; Wheeler, David C; de Jong, Paul; Gansevoort, Ron T; Levey, Andrew S; Warnock, David G; Sarnak, Mark J

    2015-10-01

    Diabetes mellitus and hypertension are risk factors for acute kidney injury (AKI). Whether estimated glomerular filtration rate (eGFR) and urine albumin-creatinine ratio (ACR) remain risk factors for AKI in the presence and absence of these conditions is uncertain. Meta-analysis of cohort studies. 8 general-population (1,285,045 participants) and 5 chronic kidney disease (CKD; 79,519 participants) cohorts. Cohorts participating in the CKD Prognosis Consortium. Diabetes and hypertension status, eGFR by the 2009 CKD Epidemiology Collaboration creatinine equation, urine ACR, and interactions. Hospitalization with AKI, using Cox proportional hazards models to estimate HRs of AKI and random-effects meta-analysis to pool results. During a mean follow-up of 4 years, there were 16,480 episodes of AKI in the general-population and 2,087 episodes in the CKD cohorts. Low eGFRs and high ACRs were associated with higher risks of AKI in individuals with or without diabetes and with or without hypertension. When compared to a common reference of eGFR of 80mL/min/1.73m(2) in nondiabetic patients, HRs for AKI were generally higher in diabetic patients at any level of eGFR. The same was true for diabetic patients at all levels of ACR compared with nondiabetic patients. The risk gradient for AKI with lower eGFRs was greater in those without diabetes than with diabetes, but similar with higher ACRs in those without versus with diabetes. Those with hypertension had a higher risk of AKI at eGFRs>60mL/min/1.73m(2) than those without hypertension. However, risk gradients for AKI with both lower eGFRs and higher ACRs were greater for those without than with hypertension. AKI identified by diagnostic code. Lower eGFRs and higher ACRs are associated with higher risks of AKI among individuals with or without either diabetes or hypertension. Copyright © 2015 National Kidney Foundation, Inc. Published by Elsevier Inc. All rights reserved.

  3. Development of quality assurance methods for low enriched fuel assemblies

    International Nuclear Information System (INIS)

    Woolstenhulme, N.E.; Moore, G.A.; Perez, D.M.; Wachs, D.M.

    2010-01-01

    As the Reduced Enrichment for Research and Test Reactors (RERTR) fuel development program has furthered the technology of low enriched uranium fuels, much effort has been expended to specify requirements, perform appropriate inspections, and to qualify experimental fuel plates and assemblies for irradiation. A great deal of consideration has been given to generate examinations and criteria that are both applicable to the unique fuel types being developed and consistent with industry practices for inspecting plate-type reactor fuel. Recent developments in quality assurance (QA) methodologies have given a heightened confidence in satisfactory fuel plate performance. At the same time, recommendations are given to further develop a system suitable for the testing and acceptance of production fuel elements containing low enriched uranium fuels. (author)

  4. A genome-wide search for linkage of estimated glomerular filtration rate (eGFR in the Family Investigation of Nephropathy and Diabetes (FIND.

    Directory of Open Access Journals (Sweden)

    Farook Thameem

    Full Text Available Estimated glomerular filtration rate (eGFR, a measure of kidney function, is heritable, suggesting that genes influence renal function. Genes that influence eGFR have been identified through genome-wide association studies. However, family-based linkage approaches may identify loci that explain a larger proportion of the heritability. This study used genome-wide linkage and association scans to identify quantitative trait loci (QTL that influence eGFR.Genome-wide linkage and sparse association scans of eGFR were performed in families ascertained by probands with advanced diabetic nephropathy (DN from the multi-ethnic Family Investigation of Nephropathy and Diabetes (FIND study. This study included 954 African Americans (AA, 781 American Indians (AI, 614 European Americans (EA and 1,611 Mexican Americans (MA. A total of 3,960 FIND participants were genotyped for 6,000 single nucleotide polymorphisms (SNPs using the Illumina Linkage IVb panel. GFR was estimated by the Modification of Diet in Renal Disease (MDRD formula.The non-parametric linkage analysis, accounting for the effects of diabetes duration and BMI, identified the strongest evidence for linkage of eGFR on chromosome 20q11 (log of the odds [LOD] = 3.34; P = 4.4 × 10(-5 in MA and chromosome 15q12 (LOD = 2.84; P = 1.5 × 10(-4 in EA. In all subjects, the strongest linkage signal for eGFR was detected on chromosome 10p12 (P = 5.5 × 10(-4 at 44 cM near marker rs1339048. A subsequent association scan in both ancestry-specific groups and the entire population identified several SNPs significantly associated with eGFR across the genome.The present study describes the localization of QTL influencing eGFR on 20q11 in MA, 15q21 in EA and 10p12 in the combined ethnic groups participating in the FIND study. Identification of causal genes/variants influencing eGFR, within these linkage and association loci, will open new avenues for functional analyses and development of novel diagnostic markers

  5. A genome-wide search for linkage of estimated glomerular filtration rate (eGFR) in the Family Investigation of Nephropathy and Diabetes (FIND).

    Science.gov (United States)

    Thameem, Farook; Igo, Robert P; Freedman, Barry I; Langefeld, Carl; Hanson, Robert L; Schelling, Jeffrey R; Elston, Robert C; Duggirala, Ravindranath; Nicholas, Susanne B; Goddard, Katrina A B; Divers, Jasmin; Guo, Xiuqing; Ipp, Eli; Kimmel, Paul L; Meoni, Lucy A; Shah, Vallabh O; Smith, Michael W; Winkler, Cheryl A; Zager, Philip G; Knowler, William C; Nelson, Robert G; Pahl, Madeline V; Parekh, Rulan S; Kao, W H Linda; Rasooly, Rebekah S; Adler, Sharon G; Abboud, Hanna E; Iyengar, Sudha K; Sedor, John R

    2013-01-01

    Estimated glomerular filtration rate (eGFR), a measure of kidney function, is heritable, suggesting that genes influence renal function. Genes that influence eGFR have been identified through genome-wide association studies. However, family-based linkage approaches may identify loci that explain a larger proportion of the heritability. This study used genome-wide linkage and association scans to identify quantitative trait loci (QTL) that influence eGFR. Genome-wide linkage and sparse association scans of eGFR were performed in families ascertained by probands with advanced diabetic nephropathy (DN) from the multi-ethnic Family Investigation of Nephropathy and Diabetes (FIND) study. This study included 954 African Americans (AA), 781 American Indians (AI), 614 European Americans (EA) and 1,611 Mexican Americans (MA). A total of 3,960 FIND participants were genotyped for 6,000 single nucleotide polymorphisms (SNPs) using the Illumina Linkage IVb panel. GFR was estimated by the Modification of Diet in Renal Disease (MDRD) formula. The non-parametric linkage analysis, accounting for the effects of diabetes duration and BMI, identified the strongest evidence for linkage of eGFR on chromosome 20q11 (log of the odds [LOD] = 3.34; P = 4.4 × 10(-5)) in MA and chromosome 15q12 (LOD = 2.84; P = 1.5 × 10(-4)) in EA. In all subjects, the strongest linkage signal for eGFR was detected on chromosome 10p12 (P = 5.5 × 10(-4)) at 44 cM near marker rs1339048. A subsequent association scan in both ancestry-specific groups and the entire population identified several SNPs significantly associated with eGFR across the genome. The present study describes the localization of QTL influencing eGFR on 20q11 in MA, 15q21 in EA and 10p12 in the combined ethnic groups participating in the FIND study. Identification of causal genes/variants influencing eGFR, within these linkage and association loci, will open new avenues for functional analyses and development of novel diagnostic markers

  6. Impact of urine concentration adjustment method on associations between urine metals and estimated glomerular filtration rates (eGFR) in adolescents

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Virginia M., E-mail: vweaver@jhsph.edu [Department of Environmental Health Sciences, Johns Hopkins Bloomberg School of Public Health, Johns Hopkins University, Baltimore, MD (United States); Johns Hopkins University School of Medicine, Baltimore, MD (United States); Welch Center for Prevention, Epidemiology, and Clinical Research, Johns Hopkins Bloomberg School of Public Health, Johns Hopkins University, Baltimore, MD (United States); Vargas, Gonzalo García [Faculty of Medicine, University of Juárez of Durango State, Durango (Mexico); Secretaría de Salud del Estado de Coahuila, Coahuila, México (Mexico); Silbergeld, Ellen K. [Department of Environmental Health Sciences, Johns Hopkins Bloomberg School of Public Health, Johns Hopkins University, Baltimore, MD (United States); Rothenberg, Stephen J. [Instituto Nacional de Salud Publica, Centro de Investigacion en Salud Poblacional, Cuernavaca, Morelos (Mexico); Fadrowski, Jeffrey J. [Johns Hopkins University School of Medicine, Baltimore, MD (United States); Welch Center for Prevention, Epidemiology, and Clinical Research, Johns Hopkins Bloomberg School of Public Health, Johns Hopkins University, Baltimore, MD (United States); Rubio-Andrade, Marisela [Faculty of Medicine, University of Juárez of Durango State, Durango (Mexico); Parsons, Patrick J. [Laboratory of Inorganic and Nuclear Chemistry, Wadsworth Center, New York State Department of Health, Albany, NY (United States); Department of Environmental Health Sciences, School of Public Health, University at Albany, Albany, NY (United States); Steuerwald, Amy J. [Laboratory of Inorganic and Nuclear Chemistry, Wadsworth Center, New York State Department of Health, Albany, NY (United States); and others

    2014-07-15

    Positive associations between urine toxicant levels and measures of glomerular filtration rate (GFR) have been reported recently in a range of populations. The explanation for these associations, in a direction opposite that of traditional nephrotoxicity, is uncertain. Variation in associations by urine concentration adjustment approach has also been observed. Associations of urine cadmium, thallium and uranium in models of serum creatinine- and cystatin-C-based estimated GFR (eGFR) were examined using multiple linear regression in a cross-sectional study of adolescents residing near a lead smelter complex. Urine concentration adjustment approaches compared included urine creatinine, urine osmolality and no adjustment. Median age, blood lead and urine cadmium, thallium and uranium were 13.9 years, 4.0 μg/dL, 0.22, 0.27 and 0.04 g/g creatinine, respectively, in 512 adolescents. Urine cadmium and thallium were positively associated with serum creatinine-based eGFR only when urine creatinine was used to adjust for urine concentration (β coefficient=3.1 mL/min/1.73 m{sup 2}; 95% confidence interval=1.4, 4.8 per each doubling of urine cadmium). Weaker positive associations, also only with urine creatinine adjustment, were observed between these metals and serum cystatin-C-based eGFR and between urine uranium and serum creatinine-based eGFR. Additional research using non-creatinine-based methods of adjustment for urine concentration is necessary. - Highlights: • Positive associations between urine metals and creatinine-based eGFR are unexpected. • Optimal approach to urine concentration adjustment for urine biomarkers uncertain. • We compared urine concentration adjustment methods. • Positive associations observed only with urine creatinine adjustment. • Additional research using non-creatinine-based methods of adjustment needed.

  7. Impact of urine concentration adjustment method on associations between urine metals and estimated glomerular filtration rates (eGFR) in adolescents

    International Nuclear Information System (INIS)

    Weaver, Virginia M.; Vargas, Gonzalo García; Silbergeld, Ellen K.; Rothenberg, Stephen J.; Fadrowski, Jeffrey J.; Rubio-Andrade, Marisela; Parsons, Patrick J.; Steuerwald, Amy J.

    2014-01-01

    Positive associations between urine toxicant levels and measures of glomerular filtration rate (GFR) have been reported recently in a range of populations. The explanation for these associations, in a direction opposite that of traditional nephrotoxicity, is uncertain. Variation in associations by urine concentration adjustment approach has also been observed. Associations of urine cadmium, thallium and uranium in models of serum creatinine- and cystatin-C-based estimated GFR (eGFR) were examined using multiple linear regression in a cross-sectional study of adolescents residing near a lead smelter complex. Urine concentration adjustment approaches compared included urine creatinine, urine osmolality and no adjustment. Median age, blood lead and urine cadmium, thallium and uranium were 13.9 years, 4.0 μg/dL, 0.22, 0.27 and 0.04 g/g creatinine, respectively, in 512 adolescents. Urine cadmium and thallium were positively associated with serum creatinine-based eGFR only when urine creatinine was used to adjust for urine concentration (β coefficient=3.1 mL/min/1.73 m 2 ; 95% confidence interval=1.4, 4.8 per each doubling of urine cadmium). Weaker positive associations, also only with urine creatinine adjustment, were observed between these metals and serum cystatin-C-based eGFR and between urine uranium and serum creatinine-based eGFR. Additional research using non-creatinine-based methods of adjustment for urine concentration is necessary. - Highlights: • Positive associations between urine metals and creatinine-based eGFR are unexpected. • Optimal approach to urine concentration adjustment for urine biomarkers uncertain. • We compared urine concentration adjustment methods. • Positive associations observed only with urine creatinine adjustment. • Additional research using non-creatinine-based methods of adjustment needed

  8. Routine reporting of estimated glomerular filtration rate (eGFR) in African laboratories and the need for its increased utilisation in clinical practice.

    Science.gov (United States)

    Adebisi, Simeon A

    2013-03-01

    Chronic Kidney Disease (CKD) is defined as the presence of markers of kidney damage or of estimated glomerular filtration rate (eGFR)clinical practice. Current guidelines advocate the use of prediction equations, such as the Cockcroft-Gault (CG) formula and the Modification of Diet in Renal Disease (MDRD) study-derived equations. Laboratories in African should commence routine reporting of eGFR for a number of reasons; 1. The sensitivity of serum creatinine (Scr) in identifying CKD is low.2. In Nigeria, a representative country; screening for Chronic Kidney Disease (CKD) is hardly considered in the routine practice of the primary and secondary care medical officers.3 Studies have shown that routine reporting of eGFR improved the documentation and identification of CKD by almost 50%.4 There is the possibility of reversing CKD if picked earlier.5. The high cost of treating CKD patients in advanced stages and the low per capital income status of the populace in Sub-Saharan Africa.6. Poor health infrastructure to manage advanced CKD patients in the continent.7. Several studies, now show lack of awareness of CKD among non-nephrologists that is related, at least in part, to difficulty in interpreting serum creatinine concentrations (the reciprocal, non-linear relationship between GFR and serum creatinine).8 Mathematical estimates of GFR [ as in eGFR] that incorporate creatinine concentration, as well as factors affecting creatinine production rates, such as size, gender, age and ethnic background, are more sensitive to changes in renal function than serum creatinine value alone.9 Recent guidelines define "action plans" for CKD according to the GFR, including referral to nephrologists at GFRs<30 mL.min(-1).(1.73 m2).

  9. Timely Diagnosis of Acute Kidney Injury Using Kinetic eGFR and the Creatinine Excretion to Production Ratio, E/eG - Creatinine Can Be Useful!

    Science.gov (United States)

    Endre, Zoltán H; Pianta, Timothy J; Pickering, John W

    2016-01-01

    Post transplant repeated measurements of urine volume and serum creatinine (sCr) are used to assess kidney function. Under non-steady state conditions, repeated measurement of sCr allows calculation of the kinetic estimated GFR (KeGFR). Additional measurement of urinary creatinine allows the calculation of the creatinine excretion to (estimated) production ratio (E/eG). We hypothesized that post-transplant KeGFR and E/eG would predict delayed graft function (DGF), as early as 4 h and outperform a validated clinical model at 12 h. This was a retrospective analysis of prospectively acquired data in a study of 56 recipients of deceased-donor kidney transplant. We assessed predictive performance with the area under the receiver operator characteristic curve (AUC) and the added value to a clinical model with integrated discrimination improvement analysis. At 4 h, the AUC for E/eG was 0.87 (95% CI 0.77-0.96) and for KeGFR 0.69 (95% CI 0.56-0.83). Both E/eG and KeGFR improved the risk prediction of a clinical model for DGF by 32 and 18%, and for non-DGF by 17 and 10%, respectively. While E/eG had better predictive performance of DGF than KeGFR, KeGFR might also facilitate perioperative management including drug dosing after kidney transplantation. Together these measurements may facilitate the possibility of conducting trials of early intervention to ameliorate the adverse effects of ischaemia-reperfusion injury on long-term DGF. © 2016 S. Karger AG, Basel.

  10. Comparative performance of the CKD Epidemiology Collaboration (CKD-EPI) and the Modification of Diet in Renal Disease (MDRD) Study equations for estimating GFR levels above 60 mL/min/1.73 m2.

    Science.gov (United States)

    Stevens, Lesley A; Schmid, Christopher H; Greene, Tom; Zhang, Yaping Lucy; Beck, Gerald J; Froissart, Marc; Hamm, Lee L; Lewis, Julia B; Mauer, Michael; Navis, Gerjan J; Steffes, Michael W; Eggers, Paul W; Coresh, Josef; Levey, Andrew S

    2010-09-01

    The Modification of Diet in Renal Disease (MDRD) Study equation underestimates measured glomerular filtration rate (GFR) at levels>60 mL/min/1.73 m2, with variable accuracy among subgroups; consequently, estimated GFR (eGFR)>or=60 mL/min/1.73 m2 is not reported by clinical laboratories. Here, performance of a more accurate GFR-estimating equation, the Chronic Kidney Disease Epidemiology Collaboration (CKD-EPI) equation, is reported by level of GFR and clinical characteristics. Test of diagnostic accuracy. Pooled data set of 3,896 people from 16 studies with measured GFR (not used for the development of either equation). Subgroups were defined by eGFR, age, sex, race, diabetes, prior solid-organ transplant, and body mass index. eGFR from the CKD-EPI and MDRD Study equations and standardized serum creatinine. Measured GFR using urinary or plasma clearance of exogenous filtration markers. Mean measured GFR was 68+/-36 (SD) mL/min/1.73 m2. For eGFR73 m2, both equations have similar bias (median difference compared with measured GFR). For eGFR of 30-59 mL/min/1.73 m2, bias was decreased from 4.9 to 2.1 mL/min/1.73 m2 (57% improvement). For eGFR of 60-89 mL/min/1.73 m2, bias was decreased from 11.9 to 4.2 mL/min/1.73 m2 (61% improvement). For eGFR of 90-119 mL/min/1.73 m2, bias was decreased from 10.0 to 1.9 mL/min/1.73 m2 (75% improvement). Similar or improved performance was noted for most subgroups with eGFR73 m2, other than body mass indexor=90 mL/min/1.73 m2. Limited number of elderly people and racial and ethnic minorities with measured GFR. The CKD-EPI equation is more accurate than the MDRD Study equation overall and across most subgroups. In contrast to the MDRD Study equation, eGFR>or=60 mL/min/1.73 m2 can be reported using the CKD-EPI equation. Copyright (c) 2010 National Kidney Foundation, Inc. All rights reserved.

  11. Development of long-life low enrichment fuel

    International Nuclear Information System (INIS)

    Gietzen, A.J.; West, G.B.

    1978-01-01

    With only a few exceptions, TRIGA reactors have always used low-enriched-uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on non-proliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U. S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of this year, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with GA's standard commercial warranty

  12. Reduced estimated glomerular filtration rate (eGFR 73 m2 ) at first transurethral resection of bladder tumour is a significant predictor of subsequent recurrence and progression.

    Science.gov (United States)

    Blute, Michael L; Kucherov, Victor; Rushmer, Timothy J; Damodaran, Shivashankar; Shi, Fangfang; Abel, E Jason; Jarrard, David F; Richards, Kyle A; Messing, Edward M; Downs, Tracy M

    2017-09-01

    To evaluate if moderate chronic kidney disease [CKD; estimated glomerular filtration rate (eGFR) 73 m 2 ] is associated with high rates of non-muscle-invasive bladder cancer (NMIBC) recurrence or progression. A multi-institutional database identified patients with serum creatinine values prior to first transurethral resection of bladder tumour (TURBT). The CKD-epidemiology collaboration formula calculated patient eGFR. Cox proportional hazards models evaluated associations with recurrence-free (RFS) and progression-free survival (PFS). In all, 727 patients were identified with a median (interquartile range [IQR]) patient age of 69.8 (60.1-77.6) years. Data for eGFR were available for 632 patients. During a median (IQR) follow-up of 3.7 (1.5-6.5) years, 400 (55%) patients had recurrence and 145 (19.9%) patients had progression of tumour stage or grade. Moderate or severe CKD was identified in 183 patients according to eGFR. Multivariable analysis identified an eGFR of 73 m 2 (hazard ratio [HR] 1.5, 95% confidence interval [CI]: 1.2-1.9; P = 0.002) as a predictor of tumour recurrence. The 5-year RFS rate was 46% for patients with an eGFR of ≥60 mL/min/1.73 m 2 and 27% for patients with an eGFR of 73 m 2 (P 73 m 2 (HR 3.7, 95% CI: 1.75-7.94; P = 0.001) was associated with progression to muscle-invasive disease. The 5-year PFS rate was 83% for patients with an eGFR of ≥60 mL/min/1.73 m 2 and 71% for patients with an eGFR of 73 m 2 (P = 0.01). Moderate CKD at first TURBT is associated with reduced RFS and PFS. Patients with reduced renal function should be considered for increased surveillance. © 2017 The Authors BJU International © 2017 BJU International Published by John Wiley & Sons Ltd.

  13. Full core operation in JRR-3 with LEU fuels

    International Nuclear Information System (INIS)

    Murayama, Y.; Issiki, M.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Snelgrove, J.L.

    1987-01-01

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

  15. Complete Flow Blockage of a Fuel Channel for Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Park, Suki

    2015-01-01

    The CHF correlation suitable for narrow rectangular channels are implemented in RELAP5/MOD3.3 code for the analyses, and the behavior of fuel temperatures and MCHFR(minimum critical heat flux ratio) are compared between the original and modified codes. The complete flow blockage of fuel channel for research reactor is analyzed using original and modified RELAP5/MOD3.3 and the results are compared each other. The Sudo-Kaminaga CHF correlation is implemented into RELAP5/MOD3.3 for analyzing the behavior of fuel adjacent to the blocked channel. A flow blockage of fuel channels can be postulated by a foreign object blocking cooling channels of fuels. Since a research reactor with plate type fuel has isolated fuel channels, a complete flow blockage of one fuel channel can cause a failure of adjacent fuel plates by the loss of cooling capability. Although research reactor systems are designed to prevent foreign materials from entering into the core, partial flow blockage accidents and following fuel failures are reported in some old research reactors. In this report, an analysis of complete flow blockage accident is presented for a 15MW pool-type research reactor with plate type fuels. The fuel surface experience different heat transfer regime in the results from original and modified RELAP5/MOD3.3. By the discrepancy in heat transfer mode of two cases, a fuel melting is expected by the modified RELAP5/MOD3.3, whereas the fuel integrity is ensured by the original code

  16. 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

  17. Measurement of GFR by Tc-99m DTPA: Comparison of 5 plasma sample and 2 plasma sample methods in North Indian population

    International Nuclear Information System (INIS)

    Mittal, B.R.; Bhattacharya, A.; Singh, B.; Jha, V.; Sarika, Kumar R.

    2007-01-01

    Assessment of Glomerular filtration rate (GFR) has significant impact on both prognosis and treatment of patients with renal disease. In this study we compared the two-plasma-sample method (G2S) using a MS excel spreadsheet based program, with a manual five-plasmasample method (GS) used to measure GFR by determining Tc-99m-diethylenetriamine penta-acetic acid (Tc-99m DTPA) clearance in patients with chronic kidney disease (CKD) and healthy renal donors. The study was conducted in 148 subjects (64 men and 84 women; age range 14 to 70 yr); 59 patients of CKD and 89 prospective healthy kidney donors. Tc-99m DTPA (74-100 MBq) was injected intravenously and thereafter blood samples were obtained at 60, 90, 120, 150 and 180 min via the patent venflon. Radioactivity in the injection syringe and plasma was measured by means of a multi-well gamma counter. The correlation coefficient between the 2 methods was 0.9453, with a slope of 0.90 and an intercept of 14.72 mL/min. Bland Altman plot of disagreement showed that G2S underestimated the GFR values by 9.0 ml/min, 11.3 ml/min and 6.9 ml/min, in the entire study, CKD and healthy donor groups respectively. Our results indicate that in spite of good correlation between GS and G2S method, the G2S method constantly underestimated GFR in our study population. However, regression equation may be applied to the GFR values estimated by G2S method to match the GFR determined by GS method. (author)

  18. Association of blood lead and mercury with estimated GFR in herbalists after the ban of herbs containing aristolochic acids in Taiwan.

    Science.gov (United States)

    Lin, Hsing-Hua; Chou, Shan-An; Yang, Hsiao-Yu; Hwang, Yaw-Huei; Kuo, Ching-Hua; Kao, Tze-Wah; Lo, Tsai-Chang; Chen, Pau-Chung

    2013-08-01

    This study was undertaken to explore the association of estimated glomerular filtration rate (GFR) with exposure to aristolochic acids (ALAs) and nephrotoxic metals in herbalists after the ban of herbs containing ALAs in Taiwan. This cross-sectional study recruited a total of 138 herbalists without end-stage renal disease or urothelial carcinoma from the Occupational Union of Chinese Herbalists in Taiwan in 2007. Aristolochic acid I (ALA-I) was measured by ultra-high-pressure liquid chromatography/ tandem mass spectrometry (UHPLC-MS/MS) and heavy metals in blood samples were analysed by Agilent 7500C inductively coupled plasma-mass spectrometry. Renal function was assessed by using a simplified Modification of Diet in Renal Disease Study equation to estimate GFR. Blood lead was higher in herbal dispensing procedures (p=0.053) and in subjects who self-prescribe herbal medicine (p=0.057); mercury was also higher in subjects living in the workplace (p=0.03). Lower estimated GFR was significantly associated with lead (β=-10.66, 95% CI -18.7 to -2.6) and mercury (β=-12.52, 95% CI -24.3 to -0.8) with a significant interaction (p=0.01) between mercury and lead; however, estimated GFR was not significantly associated with high ALA-I level groups, arsenic and cadmium after adjusting for other confounding factors. We found that lower estimated GFR was associated with blood lead and mercury in herbalists after the ban of herbs containing ALAs in Taiwan. The ALA-I exposure did not show a significant negative association of estimated GFR, which might due to herbalists having known how to distinguish ALA herbs after the banning policy. Rigorous monitoring is still needed to protect herbalists and the general population who take herbs.

  19. Urine Proteomics Revealed a Significant Correlation Between Urine-Fibronectin Abundance and Estimated-GFR Decline in Patients with Bardet-Biedl Syndrome

    Directory of Open Access Journals (Sweden)

    Marianna Caterino

    2018-03-01

    Full Text Available Background:/Aims: Renal disease is a common cause of morbidity in patients with Bardet-Biedl syndrome (BBS, however the severity of kidney dysfunction is highly variable. To date, there is little information on the pathogenesis, the risk and predictor factors for poor renal outcome in this setting. The present study aims to analyze the spectrum of urinary proteins in BBS patients, in order to potentially identify 1 disease-specific proteomic profiles that may differentiate the patients from normal subjects; 2 urinary markers of renal dysfunction. Methods: Fourteen individuals (7 males and 7 females with a clinical diagnosis of BBS have been selected in this study. A pool of 10 aged-matched males and 10 aged-matched females have been used as controls for proteomic analysis. The glomerular filtration rate (eGFR has been estimated using the CKD-EPI formula. Variability of eGFR has been retrospectively assessed calculating average annual eGFR decline (ΔeGFR in a mean follow-up period of 4 years (3-7. Results: 42 proteins were significantly over- or under-represented in BBS patients compared with controls; the majority of these proteins are involved in fibrosis, cell adhesion and extracellular matrix organization. Statistic studies revealed a significant correlation between urine fibronectin (u-FN (r2=0.28; p<0.05, CD44 antigen (r2 =0.35; p<0.03 and lysosomal alfa glucosidase ( r20.27; p<0.05 abundance with the eGFR. In addition, u-FN (r2 =0.2389; p<0.05 was significantly correlated with ΔeGFR. Conclusion: The present study demonstrates that urine proteome of BBS patients differs from that of normal subjects; in addition, kidney dysfunction correlated with urine abundance of known markers of renal fibrosis.

  20. Longitudinal change in estimated GFR among CKD patients: A 10-year follow-up study of an integrated kidney disease care program in Taiwan.

    Directory of Open Access Journals (Sweden)

    Ching-Wei Tsai

    Full Text Available This study examined the progression of chronic kidney disease (CKD by using average annual decline in estimated GFR (eGFR and its risk factors in a 10-year follow-up CKD cohort.A prospective, observational cohort study, 4600 individuals fulfilled the definition of CKD, with or without proteinuria, were followed for 10 years. The eGFR was estimated by the MDRD equation. Linear regression was used to estimate participants' annual decline rate in eGFR. We defined subjects with annual eGFR decline rate <1 ml/min/1.73 m2 as non-progression and the decline rate over 3 ml/min/1.73 m2 as rapid progression.During the follow-up period, 2870 (62.4% individuals had annual eGFR decline rate greater than 1 ml/min/1.73 m2. The eGFR decline rate was slower in individuals with CKD diagnosed over the age of 60 years than those with onset at a younger age. Comparing to subjects with decline rate <1 ml/min/1.73 m2/year, the odds ratio (OR of developing rapid CKD progression for diabetes, proteinuria and late onset of CKD was 1.72 (95% CI: 1.48-2.00, 1.89(1.63-2.20 and 0.68 (0.56-0.81, respectively. When the model was adjusted for the latest CKD stage, comparing to those with CKD stage 1, patients with stage 4 and stage 5 have significantly higher risks for rapid progression (OR, 5.17 (2.60-10.25, 19.83 (10.05-39.10, respectively. However, such risk was not observed among patients with the latest CKD stage 2 and 3. The risk for incident ESRD was 17% higher for each 1 ml/min/1.73 m2 increasing in annual decline rate.Not everyone with CKD develops ESRD after a 10-year follow-up. Absolute annual eGFR decline rate can help clinicians to better predict the progression of CKD. Individuals with renal function decline rate over 3 ml/min/1.73 m2/year require intensive CKD care.

  1. Fuel Fraction Analysis of 500 MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel without Refueling

    Science.gov (United States)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-01-01

    Nuclear Power Plant (NPP) is one of candidates which can support electricity demand in the world. The Generation IV NPP has fourth main objective, i.e. sustainability, economics competitiveness, safety and reliability, and proliferation and physical protection. One of Gen-IV reactor type is Gas Cooled Fast Reactor (GFR). In this study, the analysis of fuel fraction in small GFR with nitride fuel has been done. The calculation was performed by SRAC code, both Pij and CITATION calculation. SRAC2002 system is a code system applicable to analyze the neutronics of variety reactor type. And for the data library used JENDL-3.2. The step of SRAC calculation is fuel pin calculated by Pij calculation until the data homogenized, after it homogenized we calculate core reactor. The variation of fuel fraction is 40% up to 65%. The optimum design of 500MWth GFR without refueling with 10 years burn up time reach when radius F1:F2:F3 = 50cm:30cm:30cm and height F1:F2:F3 = 50cm:40cm:30cm, variation percentage Plutonium in F1:F2:F3 = 7%:10%:13%. The optimum fuel fraction is 41% with addition 2% Plutonium weapon grade mix in the fuel. The excess reactivity value in this case 1.848% and the k-eff value is 1.01883. The high burn up reached when the fuel fraction is low. In this study 41% fuel fraction produce faster fissile fuel, so it has highest burn-up level than the other fuel fraction.

  2. A decreased soluble Klotho level with normal eGFR, FGF23, serum phosphate, and FEP in an ADPKD patient with enlarged kidneys due to multiple cysts.

    Science.gov (United States)

    Kanai, Takahiro; Shiizaki, Kazuhiro; Betsui, Hiroyuki; Aoyagi, Jun; Yamagata, Takanori

    2018-05-16

    Autosomal dominant polycystic kidney disease (ADPKD) is the most common hereditary renal disorder. ADPKD is characterized clinically by the presence of multiple bilateral renal cysts that lead to chronic renal failure. The cysts evolve from renal tubular epithelial cells that express the Klotho gene. Notably, Klotho acts as a co-receptor for fibroblast growth factor 23 (FGF23); in this context, it induces phosphaturia and maintains serum phosphate at a normal level. Many reports have shown that decreases in the soluble Klotho level and increases in the FGF23 level are associated with glomerular filtration rate (GFR) decline, but a recent study observed these changes in patient with normal eGFR. It remains unclear whether the decrease in the Klotho level precedes the increase in FGF23. Here, we present an ADPKD patient with enlarged kidneys due to multiple cysts who had a decreased soluble Klotho level but a normal eGFR and a normal FGF23 level. The patient's serum phosphate level was normal, as was the fractional excretion of phosphate (FEP). This appears to be the first reported case to show a decreased soluble Klotho level plus normal eGFR, FGF23, and FEP. These results suggest that Klotho decreases before FGF23 increases and further suggest that Klotho is not required to maintain normal serum phosphate levels in ADPKD if the FEP and serum phosphate levels are normal.

  3. GFR Decline as an Alternative End Point to Kidney Failure in Clinical Trials : A Meta-analysis of Treatment Effects From 37 Randomized Trials

    NARCIS (Netherlands)

    Inker, Lesley A.; Lambers Heerspink, Hiddo J.; Mondal, Hasi; Schmid, Christopher H.; Tighiouart, Hocine; Noubary, Farzad; Coresh, Josef; Greene, Tom; Levey, Andrew S.

    2014-01-01

    Background: There is increased interest in using alternative end points for trials of kidney disease progression. The currently established end points of end-stage renal disease and doubling of serum creatinine level, equivalent to a 57% decline in estimated glomerular filtration rate (eGFR), are

  4. Impact of Gate 99mTc DTPA GFR, Serum Creatinine and Urea in Diagnosis of Patients with Chronic Kidney Failure

    Science.gov (United States)

    Miftari, Rame; Nura, Adem; Topçiu-Shufta, Valdete; Miftari, Valon; Murseli, Arbenita; Haxhibeqiri, Valdete

    2017-01-01

    Aim: The aim of this study was determination of validity of 99mTcDTPA estimation of GFR for early detection of chronic kidney failure Material and methods: There were 110 patients (54 males and 56 females) with kidney disease referred for evaluation of renal function at UCC of Kosovo. All patients were included in two groups. In the first group were included 30 patients confirmed with renal failure, whereas in the second group were included 80 patients with other renal disease. In study were included only patients with ready results of creatinine, urea and glucose in the blood serum. For estimation of GFR we have used the Gate GFR DTPA method. The statistical data processing was conducted using statistical methods such as arithmetic average, the student t-test, percentage or rate, sensitivity, specificity and accuracy of the test. Results: The average age of all patients was 36 years old. The average age of female was 37 whereas of male 35. Patients with renal failure was significantly older than patients with other renal disease (p<0.005). Renal failure was found in 30 patients (27.27%). The concentration of urea and creatinine in blood serum of patients with renal failure were significantly higher than in patients with other renal disease (P< 0.00001). GFR in patients with renal failure were significantly lower than in patients with other renal disease, 51.75 ml/min (p<0.00001). Sensitivity of uremia and creatininemia for detection of renal failure were 83.33%, whereas sensitivity of 99mTcDTPA GFR was 100%. Specificity of uraemia and creatininemia were 63% whereas specificity of 99mTcDTPA GFR was 47.5%. Diagnostic accuracy of blood urea and creatinine in detecting of renal failure were 69%, whereas diagnostic accuracy of 99mTcDTPA GFR was 61.8%. Conclusion: Gate 99mTc DTPA scintigraphy in collaboration with biochemical tests are very sensitive methods for early detection of patients with chronic renal failure. PMID:28883673

  5. Comparison of renal function assessment by cystatin c and creatinine based equations for e-gfr in type 2 diabetics in different stages of albuminuria

    International Nuclear Information System (INIS)

    Qamar, A.; Ahmad, T.M.; Hayat, A.; Khan, M.A.; Rehman, S. Z.

    2017-01-01

    To compare e-GFR estimated by creatinine or cystatin C based and combined creatinine and cystatin C based equations in type 2 diabetics in different stages of albuminuria. Study Design: Comparative cross-sectional study. Place and Duration of Study: Department of Chemical Pathology, Army Medical College Rawalpindi in collaboration with endocrinology outpatient department Military Hospital Rawalpindi, from Nov 2015 to Nov 2016. Material and Methods: A total of 119 type 2 diabetic subjects of either gender, aged 30- 60 years were enrolled in the study with duration of diabetes less than 15 years and were divided into further sub groups on the basis of degree of albuminuria determined by spot urine albumin to creatinine ratio (uACR). Fifty age matched disease free controls with no history of any systemic disease were also included in the study. Known patients of type 1 diabetes, chronic inflammatory disorders, uncontrolled hypertension, thyroid disease, chronic kidney disease, on lipid lowering drugs, steroids, ACE inhibitors and pregnant ladies were excluded from the study. Serum creatinine serum cystatin C were assessed on fully automated chemistry analyzer selectra. E-GFR was calculated by online GFR calculator by National Kidney Foundation. Comparison of means of e-GFR calculated by various equations was carried out by one way ANOVA and post-hoc Tukey tests. Degree of agreement between various equations for the estimation of GFR was assessed by kappa statistics. A p-value less than 0.05 were considered statistically significant. Results: Mean e-GFR (ml/min/1.73m2) was lowest in cystatin C based CKD-EPI equation (89.56 +- 39.84) followed by combined cystatin C and creatinine based CKD-EPI (92.34 +- 37.88). Values of e-GFR by creatinine based CKD-EPI equation (95.84 +- 27.24), and by creatinine based MDRD equation (105.37 +- 64.98) were both higher. In creatinine based MDRD, equation normo albuminuria and micro albuminuria groups did not show statistically

  6. Gfr estimation using 99mTc DTPA gates method for assessment of early diabetic nephropathy - a comparison with 24-hour creatinine clearance

    International Nuclear Information System (INIS)

    Ghafoor, S.; Ali, M.K.; Khan, G.

    2014-01-01

    To correlate Gates glomerular filtration rate (GGFR) using technetium-99m diethylene triaminepentacetic acid (99mTc DTPA) with 24-hour creatinine clearance (CRCL) and to establish relationship with duration of diabetes in patients with early diabetic nephropathy. Study Design: A cross-sectional comparative study carried out in Nuclear Medical Centre from Aug 2009 to Jan 2010 at Armed Forces Institute of Pathology (AFIP), Rawalpindi, Pakistan. Patients and Methods: A total of eighty three subjects were enrolled, who were divided into three groups; group 1 comprised 31 normotensive diabetics, group 2 had 37 hypertensive diabetics while group 3 had 15 normal subjects. The DTPA GFR and creatinine clearance in healthy subjects as well as diabetic patients were compared using the unpaired student's t-test. The linear association between GFR, creatinine clearance and disease duration was expressed by Pearson's correlation coefficient 'r' along with their significance levels. Results: Gates GFR showed hyperfiltration in normotensive diabetics (96.6 +- 3.3 ml/min/1.73 m/sub 2/), significantly (p<0.05) higher than controls (85.5 +- 5 ml/min/1.73 m/sub 2/), whereas hypertensive diabetics had a significantly lower (p<0.05) Gates GFR (76.8 +- 3.7) than that of controls. Significant degree of correlation existed between GGFR and CRCL in hypertensive diabetics (p<0.05, r=0.716) and controls (r=0.546). Gates GFR also showed good correlation with duration of diabetes in both diabetic groups as compared to that of CRCL. GGFR also correlated well with duration of hypertension 0.37 (0.31-0.43) as compared to CRCL 0.155 (0.15-0.16) in all groups. Conclusions: The 99mTc-DTPA clearance correlates significantly with 24-hour creatinine clearance as well as with disease duration and can provide a simple and convenient index of kidney function in patients of early diabetic nephropathy. (author)

  7. Application of vacuum technology during nuclear fuel fabrication, inspection and characterization

    International Nuclear Information System (INIS)

    Majumdar, S.

    2003-01-01

    Full text: Vacuum technology plays very important role during various stages of fabrication, inspection and characterization of U, Pu based nuclear fuels. Controlled vacuum is needed for melting and casting of U, Pu based alloys, picture framing of the fuel meat for plate type fuel fabrication, carbothermic reduction for synthesis of (U-Pu) mixed carbide powder, dewaxing of green ceramic fuel pellets, degassing of sintered pellets and encapsulation of fuel pellets inside clad tube. Application of vacuum technology is also important during inspection and characterization of fuel materials and fuel pins by way of XRF and XRD analysis, Mass spectrometer Helium leak detection etc. A novel method of low temperature sintering of UO 2 developed at BARC using controlled vacuum as sintering atmosphere has undergone successful irradiation testing in Cirus. The paper will describe various fuel fabrication flow sheets highlighting the stages where vacuum applications are needed

  8. Applications of high-Tc-superconductors to power engineering. Manufacture of YBCO plate-type conductors and construction of a HTS current limiter model up to 1 MVA nominal power. Final report

    International Nuclear Information System (INIS)

    Utz, B.; Schmidt, W.; Schilling, W.; Fischperer, I.; Kraemer, H.P.; Wacker, B.; Gromoll, B.; Neumueller, H.W.; Arndt, A.; Karras, B.; Krueger, U.; Pyritz, U.; Schiewe, H.; Schiller, H.P.; Volkmar, R.R.; Hering, U.; Roessler, R.; Freyhardt, H.C.; Sievers, S.; Hoffmann, J.; Dzich, J.; Kinder, H.; Hoffmann, C.; Lindmayer, M.; Grundmann, J.; Woerdenweber, R.; Hollmann, E.; Kutzner, R.; Klein, W.; Bunte, S.; Kuhn, M.

    2002-06-01

    In terms of materials, the main focus of the work was on the manufacture of large-area YBCO plate-type conductors with homogeneous properties and maximum current densities of j c >1 MA/cm 2 . j c values of better than 3 MA/cm 2 were achieved reproducibly on sapphire substrates of 100 mm diameter and 10 x 20 cm 2 in size with a homogeneity of 10%; on polycrystalline substrates of 10 x 20 cm 2 in size, homogeneous j c values of up to 2 MA/cm 2 were also successfully demonstrated. Of the total of four methods of coating available at the start of the project, thermal co-evaporation (TCE) proved best for YBCO thin films and the IBAD method best for quasi single-crystal buffer films. The latter are necessary to achieve high j c on polycrystalline substrates such as ZrO 2 (Y), glass and Al 2 O 3 . Polycrystalline substrates are essential in order to make the HTS current limiter as a future product commercially feasible. The favoured solutions ZrO 2 (Y) and glass have not come up to expectations, because present investigations into quench propagation are showing that, with this approach, the high values of power density required for the switching process (1600 VA/cm 2 ) cannot be achieved. Towards the end of the project, polycrystalline Al 2 O 3 began to be seen as a successful alternative; the work is being pursued further within the context of a follow-on project. The coating processes were stabilized successfully and, when combined with strict quality control, allowed the yield of tested, ready-to-use plate-type conductors to be improved to 85%. This success was an essential prerequisite for the building of a 3-phase, 1.2 MVA model (7.2 kV) comprising a total of sixty-three 100 mm plate-type conductors. At the Berlin factory the model has been successfully tested up to a prospective short-circuit current of 5 kV. This has demonstrated the basic suitability of HTS thin-film technology for use in current limiters. So far the model has been switched a total of 43 times

  9. eGFR

    Science.gov (United States)

    ... Diabetes Diarrhea Disseminated Intravascular Coagulation (DIC) Down Syndrome Ebola Virus Infection Endocrine System and Syndromes Epilepsy Excessive ... See More See Less Related Images View More × Structure of a kidney. Image credit: Alan Hoofring, National ...

  10. Effect of tolvaptan on renal handling of water and sodium, GFR and central hemodynamics in autosomal dominant polycystic kidney disease during inhibition of the nitric oxide system

    DEFF Research Database (Denmark)

    Therwani, Safa; Malmberg, My Emma Sofie; Rosenbaek, Jeppe Bakkestroem

    2017-01-01

    -dependent mechanism. U-AQP2 was not changed by tolvaptan, presumeably due to a counteracting effect of elevated p-AVP. The reduced GFR during tolvaptan most likely is caused by the reduction in extracellular fluid volume and blood pressure. Trial registration: Clinical Trial no: NCT02527863. Registered 18 February...... received tolvaptan 60 mg or placebo in a randomized, placebo-controlled, double blind, crossover study. L-NMMA (L-NG-monomethyl-arginine) was given as a bolus followed by continuous infusion during 60 min. We measured: GFR, urine output (UO), free water clearance (CH2O), fractional excretion of sodium...... (FENa), urinary excretion of aquaporin-2 channels (u-AQP2) and epithelial sodium channels (u-ENaCγ), plasma concentrations of vasopressin (p-AVP), renin (PRC), angiotensinII (p-AngII), aldosterone (p-Aldo), and central blood pressure (cBP). Results: During tolvaptan with NO-inhibition, a more pronounced...

  11. Diagnostic reference range of κ/λ free light chain ratio to screen for Bence Jones proteinuria is not significantly influenced by GFR.

    Science.gov (United States)

    Schmidt-Hieltjes, Yvonne; Elshof, Clemens; Roovers, Lian; Ruinemans-Koerts, Janneke

    2016-05-01

    The aim of our study was to analyse whether the κ/λ free light chain ratio reference range for screening for Bence Jones proteinuria should be dependent on the estimated glomerular filtration rate (eGFR). The serum κ/λ free light chain ratio, eGFR, serum M-protein and Bence Jones protein were measured in 544 patients for whom Bence Jones protein analysis was ordered. In the population of patients without Bence Jones proteinuria or a M-protein (n = 402), there is no gradual increase in κ/λ free light chain ratio with diminishing eGFR. The κ/λ free light chain ratio in this group was 0.56-1.86 (95% interval). With this diagnostic reference range of the κ/λ ratio, 105 of the 110 patients with Bence Jones protein could be identified correctly. Only five patients with Bence Jones proteinuria (free light chain ratio was measured without the presence of Bence Jones proteinuria. A κ/λ free light chain ratio in serum can be used safely and efficiently to select urine samples which should be analysed for Bence Jones proteinuria with an electrophoresis/immunofixation technique. Using this diagnostic reference range, the number of urine samples which should be analysed by electrophoresis/immunofixation could be reduced by 74%. The diagnostic reference interval can be determined best in a group of patients for whom Bence Jones analysis is indicated. For calculation of this reference range, the eGFR value does not need to be taken into account. © 2015 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

  12. The relationship of plasma creatinine (as eGFR) and high-sensitivity cardiac troponin and NT-proBNP concentrations in a hospital and community outpatient population.

    Science.gov (United States)

    Potter, Julia M; Simpson, Aaron J; Kerrigan, Jennifer; Southcott, Emma; Salib, Marie M; Koerbin, Gus; Hickman, Peter E

    2017-10-01

    While persons with overt renal failure have a well-described rise in troponin and NT-proBNP, it is less well described what the relationship is between cardiac markers and persons with impaired renal function, not requiring dialysis. We have collected ALL samples referred to our pathology practice over a 24h period and measured hs-cTnI, hs-cTnT, NT-proBNP, calculated the eGFR, and related our measurements to clinical outcomes. For both men and women, for all of hs-cTnI, hs-cTnT and NT-proBNP, there was a graded response, as renal function worsened, the concentration of the cardiac marker increased. There is a graded inverse relationship between eGFR and the concentrations of hs-cTnI, hs-cTnT and NT-proBNP. For women only there appeared to be an increase in mortality at lowest eGFR. Copyright © 2017 The Canadian Society of Clinical Chemists. Published by Elsevier Inc. All rights reserved.

  13. Interest and limits of glomerular filtration rate (GFR) estimation with formulae using creatinine or cystatin C in the malnourished elderly population.

    Science.gov (United States)

    Fabre, Emmanuelle E; Raynaud-Simon, Agathe; Golmard, Jean-Louis; Gourgouillon, Nadège; Beaudeux, Jean-Louis; Nivet-Antoine, Valérie

    2010-01-01

    Renal function is often altered in elderly patients. A lot of formulae are proposed to estimate GFR to adjust drug posology. French guidelines recommend the Cockcroft-Gault formula corrected with the body surface area (cCG), but the initially described unadjusted Cockcroft-Gault equation (CG) is mainly used in geriatric clinical practice. International recommendations have proposed the modification of diet in renal disease (MDRD) formula, since several authors recommended the Rule formula using cystatin C (cystC) in particular population. To appreciate the most accurate GFR estimation for posology adaptation in an elderly polypathological population, a cross-sectional study with prospective inclusion was carried out in Charles Foix Hospital. Plasma glucose levels (PGL), creatinine (CREA) levels and serum cystC, albumin (ALB), transthyretin (TTR), C-reactive protein (CRP), orosomucoid (ORO) total cholesterol (tCHOL) levels were determined among 193 elderly patients aged 70 and older. The results showed that in a malnourished, inflamed old population, CG, MDRD and Rule formulae resulted in different estimations of GFR, depending on nutritional and inflammatory parameters. Only cCG estimation was shown to be independent from these parameters. To conclude, cCG seems to be the most accurate and appropriate formula in a polypathological elderly population to evaluate renal function in order to adapt drug posology. Copyright (c) 2009 Elsevier Ireland Ltd. All rights reserved.

  14. Application of powder metallurgy in production of nuclear fuels for research and power reactors

    International Nuclear Information System (INIS)

    Fukuda, Kosaku

    2000-01-01

    Powder metallurgy has been applied in many of the processes of nuclear fuel fabrication, which has contributed, to a great progress of the nuclear technology to date. Evolution of nuclear fuels still continues to meet various emerging demands in terms of enhanced safety, economical effectiveness, non-proliferation and environmental mitigation. This paper reviews recent progress of nuclear fuels of research and power reactors, in particular, focusing on the powder metallurgy application. First, the review is made on plate type fuels for research reactors, inter alia, silicide fuel which is prevailing worldwide from the viewpoint of non-proliferation. The relation between fabrication and irradiation behavior is also discussed. Next, oxide fuels including MOX are reviewed. Recent interests of UO 2 are directed toward large grain pellets and burnable absorber pellets, both of which arise from requirement of extended burnup. Finally, the MOX fuel for thermal reactors is reviewed. (author)

  15. Spent fuel and fuel pool component integrity. Annual report, FY 1979

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

    1980-05-01

    International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-μm) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report

  16. Fuel Exhaling Fuel Cell.

    Science.gov (United States)

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  17. Nuclear Fuels & Materials Spotlight Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    I. J. van Rooyen,; T. M. Lillo; Y. Q. WU; P.A. Demkowicz; L. Scott; D.M. Scates; E. L. Reber; J. H. Jackson; J. A. Smith; D.L. Cottle; B.H. Rabin; M.R. Tonks; S.B. Biner; Y. Zhang; R.L. Williamson; S.R. Novascone; B.W. Spencer; J.D. Hales; D.R. Gaston; C.J. Permann; D. Anders; S.L. Hayes; P.C. Millett; D. Andersson; C. Stanek; R. Ali; S.L. Garrett; J.E. Daw; J.L. Rempe; J. Palmer; B. Tittmann; B. Reinhardt; G. Kohse; P. Ramuhali; H.T. Chien; T. Unruh; B.M. Chase; D.W. Nigg; G. Imel; J. T. Harris

    2014-04-01

    As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • The first identification of silver and palladium migrating through the SiC layer in TRISO fuel • A description of irradiation assisted stress corrosion testing capabilities that support commercial light water reactor life extension • Results of high-temperature safety testing on coated particle fuels irradiated in the ATR • New methods for testing the integrity of irradiated plate-type reactor fuel • Description of a 'Smart Fuel' concept that wirelessly provides real time information about changes in nuclear fuel properties and operating conditions • Development and testing of ultrasonic transducers and real-time flux sensors for use inside reactor cores, and • An example of a capsule irradiation test. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps to spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at INL, and hope that you find this issue informative.

  18. Development of an innovative plate-type SG for fast breeder reactor. Proposal of the concept and the evaluation of the fabricating method by the test fabrication of the partial model

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Kinoshita, Izumi

    2006-01-01

    The concept of an innovative plate type SG for the fast reactor fabricated by using the HIP (Hot Isostatic Pressing) method was proposed. The heat transfer plate, which is assembled with rectangular tubes and is fabricated by HIP method, is surrounded by leakage detection spaces. It is possible to apply it to both the pool-type and the loop-type LMFR. In this report, the fabrication technique was studied about the concept for the loop-type LMFR, and the following results were obtained. Hip tests, tensile tests, and structure observation were performed to clarify the suitable HIP condition for the modified 9Cr-1Mo steel. As a result, the optimum condition of 1150 deg C x 1200 kgf/cm 2 x 3 hr was found. Nickel-type solder (BNi-5) and gold-type solder (BAu-4) were selected as a joining material to laminate the heat transfer tube plates. Through the comparison of tensile tests, BAu-4 that showed a more excellent joining performance was selected on the assumption of the margin of 5 mm from the welding line. After buckling load had been clarified, the BAu-4 brazing of the heat transfer tube plates was performed using a hot pressing method. Problems were not observed in the welding of simulated header, and in the fabricating of the partial model of SG. (author)

  19. 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)

  20. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  1. A Development of Technical Specification of a Research Reactor with Plate Fuels Cooled by Upward Flow

    International Nuclear Information System (INIS)

    Park, Sujin; Kim, Jeongeun; Kim, Hyeonil

    2016-01-01

    The contents of the TS(Technical Specifications) are definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls. TS for Nuclear Power Plants (NPPs) have been developed since many years until now. On the other hands, there are no applicable modernized references of TS for research reactors with many differences from NPPs in purpose and characteristics. Fuel temperature and Departure from Nuclear Boiling Ratio (DNBR) are being used as references from the thermal-hydraulic analysis point of view for determining whether the design of research reactors satisfies acceptance criteria for the nuclear safety or not. Especially for research reactors using plate-type fuels, fuel temperature and critical heat flux, however, are very difficult to measure during the reactor operation. This paper described the outline of main contents of a TS for open-pool research reactor with plate-type fuels using core cooling through passive systems, where acceptance criteria for nuclear safety such as CHF and fuel temperature cannot be directly measured, different from circumstances in NPPs. Thus, three independent variables instead of non-measurable acceptance criteria: fuel temperature and CHF are considered as safety limits, i.e., power, flow, and flow temperature

  2. Impact of Gate 99mTc DTPA GFR, Serum Creatinine and Urea in Diagnosis of Patients with Chronic Kidney Failure

    OpenAIRE

    Miftari, Rame; Nura, Adem; Top?iu-Shufta, Valdete; Miftari, Valon; Murseli, Arbenita; Haxhibeqiri, Valdete

    2017-01-01

    Aim: The aim of this study was determination of validity of 99mTcDTPA estimation of GFR for early detection of chronic kidney failure Material and methods: There were 110 patients (54 males and 56 females) with kidney disease referred for evaluation of renal function at UCC of Kosovo. All patients were included in two groups. In the first group were included 30 patients confirmed with renal failure, whereas in the second group were included 80 patients with other renal disease. In study were ...

  3. 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

  4. The continuous fuel cycle model and the gas cooled fast reactor

    International Nuclear Information System (INIS)

    Christie, Stuart; Lathouwers, Danny; Kloosterman, Jan Leen; Hagen, Tim van der

    2011-01-01

    The gas cooled fast reactor (GFR) is one of the generation IV designs currently being evaluated for future use. It is intended to behave as an isobreeder, producing the same amount of fuel as it consumes during operation. The actinides in the fuel will be recycled repeatedly in order to minimise the waste output to fission products only. Striking the balance of the fissioning of various actinides against transmutation and decay to achieve these goals is a complex problem. This is compounded by the time required for burn-up modelling, which can be considerable for a single cycle, and even longer for studies of fuel evolution over many cycles. The continuous fuel cycle model approximates the discrete steps of loading, operating and unloading a reactor as continuous processes. This simplifies the calculations involved in simulating the behaviour of the fuel, reducing the time needed to model the changes to the fuel composition over many cycles. This method is used to study the behaviour of GFR fuel over many cycles and compared to results obtained from direct calculations. The effects of varying fuel cycle properties such as feed material, recycling of additional actinides and reprocessing losses are also investigated. (author)

  5. 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.

  6. Fuel assemblies

    International Nuclear Information System (INIS)

    Mukai, Hideyuki

    1987-01-01

    Purpose: To prevent bending of fuel rods caused by the difference of irradiation growth between coupling fuel rods and standards fuel rods thereby maintain the fuel rod integrity. Constitution: The f value for a fuel can (the ratio of pole of zirconium crystals in the entire crystals along the axial direction of the fuel can) of a coupling fuel rod secured by upper and lower tie plates is made smaller than the f value for the fuel can of a standard fuel rod not secured by the upper and the lower tie plates. This can make the irradiation growth of the fuel can of the coupling fuel rod greater than the irradiation growth of the fuel can of the standard fuel rod and, accordingly, since the elongation of the standard fuel rod can always by made greater, bending of the standard fuel rod can be prevented. (Yoshihara, M.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-08-01

    for dimensional changes, blisters, or any other visible defect. After inspection the experiment is reassembled and reinserted into the reactor for further irradiation. At completion of the specified in reactor exposure of the miniplates, the experiment will be disassembled, and the irradiated samples will be returned to the respective fuel plate fabricators for detailed postirradiation examination. Ultimately, fuel plate types with suitable characteristics will be manufactured into full-sized plate-type fuel elemental suitable for testing in the ORR. (author)

  8. The Role of Friction Stir Welding in Nuclear Fuel Plate Fabrication

    International Nuclear Information System (INIS)

    Burkes, D.; Medvedev, P.; Chapple, M.; Amritkar, A.; Wells, P.; Charit, I

    2009-01-01

    The friction bonding process combines desirable attributes of both friction stir welding and friction stir processing. The development of the process is spurred on by the need to fabricate thin, high density, reduced enrichment fuel plates for nuclear research reactors. The work seeks to convert research and test reactors currently operating on highly enriched uranium fuel to operate on low enriched uranium fuel without significant loss in reactor performance, safety characteristics, or significant increase in cost. In doing so, the threat of global nuclear material proliferation will be reduced. Feasibility studies performed on the process show that this is a viable option for mass production of plate-type nuclear fuel. Adapting the friction stir weld process for nuclear fuel fabrication has resulted in the development of several unique ideas and observations. Preliminary results of this adaptation and process model development are discussed

  9. Renal toxicity of 2 cycles of peptide receptor radionuclide therapy as determined by serial measurements of the Glomerular Filtration Rate (GFR): comparison between Y-90 DOTA-TATE and Lu-177 DOTA-TATE

    International Nuclear Information System (INIS)

    Baum, R.P.; Prasad, V.

    2007-01-01

    Full text: Aim: To determine the effect of peptide receptor radionuclide therapy on GFR after 2 cycles of Y-90-DTATATE as compared to Lu-177-DOTA-TATE. Methods: Group A (Y-90), Group B (Lu-177). Group A1: 24 pts (age 60.5±11a), injected with 4.00± 0.72 GBq of Y-90 (1st cycle). Group A2: 16 pts (age 62.2±9.4a, followed up after 7.8 GBq±0.82 GBq of Y-90 (after 2nd cycle). Group B1: 14 pts (age 62.2±10.6a) 4.8± 0.8 GBq Lu-177 (1st cycle). Group B2: 6 patients (age 58.5±12a) after 9.57±1.5 GBq of Lu-177 (after 2nd cycle). GFR was determined using 110- 185 MBq of Tc-99m DTPA before and 3-4 months after therapy. Absolute/normalized values for GFR pre/post PRRT were compared (paired T-test).The effect of total amount of radioactivity administered, pre-existing diabetes, hypertension and age on renal function post 2 cycles of PRRT were also evaluated. Results: In group A1 normalized GFR dropped by 2% (absol. GFR fall: 2 ml/min) as compared to 16% in group A2 (absol. GFR fall: 7 ml/min). Baseline normalized/absol. GFR values were 1.02/87.5 ml/min in subgroup A1 and 1.02 / 86.5 ml/min in A2. The fall in both, the absolute and normalized GFR values was not significant after the 1st cycle (p=0.555), but was significant (p= 0.007) after the 2nd PRRT. In group B1 there was a fall of normalized GFR value by 10% vs. 8% in group B2. The fall in absol. GFR value was 7.7 ml/min in group B1 and 9.7 ml/min in group B2. Baseline normalized GFR values was 0.86 and 0.89 in subgroups B1 and B2, respectively. Baseline absol. GFR value was 78.3 ml/min and 81 ml/min in subgroups B1 and B2. The fall in the absol. GFR values was significant after the 1st cycle (p=.009), but was not significant (p=0.486) after the 2nd cycle of PRRT. The fall in normalized GFR value was not significant in both subgroups (p=0.07 after 1st cycle and p=0.49 after 2nd cycle). No correlation between the activity administered and the percentage change in the GFR values in both the groups (Pearson's correlation

  10. The Comparative Study of the Measurement of Serum Cystatin C and 99mTc-DTPA for the Measurement of Glomerular Filtration Rate (GFR) in Type 2 Diabetic Nephropathy

    International Nuclear Information System (INIS)

    Kong Linghua; Wu Junyuan; Gao Bin; Wang Xueqin; Gu Kaikai

    2010-01-01

    To explore the clinical value between measurement of serum cystatin C (Cys C) and determination of plasma with 99m Tc-DTPA glomerular filtration rate (GFR) clearance for type 2 diabetes (T2DM) in early detection of renal damage, the serum cystatin C levels in 87 patients with type 2 diabetes were determined by immune turbidimetry Cys C. The patients were also carried out dynamic imaging to measure renal GFR. The result showed that the serum levels of Cys C in patients with Type 2 diabetic was 1.68 ± 0.52 mg / L, the normal group was 0.72±0.26 mg/L; The GFR in patients with T2DM was 93.8 ± 30.2 ml/min/1.73m 2 , the control group was 107.48±15.23 ml/min/1.73m 2 , there was significant difference between patients and controls (P 99m Tc-DTPA renal dynamic imaging. The serum Cys C could reflect the damage in early diabetic patients. The method is simple, accurate and easy to spread. The renal dynamic imaging method of GFR determination is slightly complex and requires specialized equipment. It can reflect the sub-renal function. The combined measurement of serum Cys C and GFR are very important in the early detection of type 2 diabetic patients with early renal function. (authors)

  11. 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.

  12. Fibroblast Growth Factor-23 and Vitamin D Metabolism in Subjects with eGFR ≥60 ml/min/1.73 m².

    Science.gov (United States)

    Nakatani, Shinya; Nakatani, Ayumi; Tsugawa, Naoko; Yamada, Shinsuke; Mori, Katsuhito; Imanishi, Yasuo; Ishimura, Eiji; Okano, Toshio; Inaba, Masaaki

    2015-01-01

    Fibroblast growth factor (FGF)-23 and parathyroid hormone (PTH) are both potent phosphaturic hormones. Since they exert opposite effects on vitamin D metabolism, the measurement of 3 vitamin D metabolites; 25-hydroxyvitamin D (25-OH-D), 1,25-dihydroxyvitamin D (1,25(OH)2D), and 24,25-dihydroxyvitamin D (24,25(OH)2D), allows the distinction of the effects of FGF-23 from those of PTH. The aim of this study was to elucidate which factor, FGF-23 or PTH, plays a more important role in the regulation of vitamin D metabolites in subjects with estimated glomerular filtration (eGFR) ≥60 ml/min/1.73 m(2). Subjects with eGFR ≥60 ml/min/1.73 m(2) (n = 20) were enrolled and their serum levels of FGF-23, intact PTH, and vitamin D metabolites were determined. Serum FGF-23 correlated inversely with 1,25(OH)2D (r = -0.717, p = 0.0004) and the 1,25(OH)2D/25-OH-D ratio (r = -0.518, p = 0.019), compared with a significant positive correlation between serum intact PTH and the 1,25(OH)2D/25-OH-D ratio (r = 0.562, p = 0.010). Multiple regression analyses revealed serum FGF-23 as a significant factor that was associated with serum 1,25(OH)2D (β = -0.593, p = 0.018), 1,25(OH)2D/25-OH-D ratio (β = -0.521, p = 0.025), and the 24,25(OH)2D/1,25(OH)2D ratio (β = 0.632, p = 0.008), and intact PTH as a significant factor associated with the 1,25(OH)2D/25-OH-D ratio (β = 0.445, p = 0.028). This study demonstrated that, even in subjects with eGFR ≥60 ml/min/1.73 m(2), FGF-23 might play an important role in the regulation of vitamin D metabolism. In addition to the established role of PTH, the association between FGF-23 and indices of vitamin D metabolism suggested the potential role of FGF-23 on phosphate metabolism in such patients. © 2015 S. Karger AG, Basel.

  13. Replacement of highly enriched uranium by medium or low-enriched uranium in fuels for research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    To exclude the possibility of an explosive use of the uranium obtained from an elementary chemical process, one needs to use a fuel less enriched than 20 weight percent in U 235 . This goal can be reached by two ways: 1. The low density fuels, i.e. U or U 3 O 8 /Al fuels. One has to increase their U content from 1.3 g U/cm 3 presently qualified under normal operation conditions. Several manufacturers such as CERCA in France developed these fuels with a near-term objective of about 2 g U/cm 3 and a long-term objective of 3 g U/cm 3 . 2. The high density fuels. They are the UO 2 Caramel plate type fuels now under consideration, and U 3 Si and UMo as a long-term potential

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  15. A precise evaluation of glomerular filtration rate (GFR) in two plasma samples following a single administration of 57Co-B12 vitamin

    International Nuclear Information System (INIS)

    Camargo, E.E.; Rockmann, R.L.; Barreto, T.M.; Eston, T.E.; Papaleo Netto, M.; Carvalho, N.

    1974-01-01

    Through a logarithmic regression performed with the contings of 4 plasma samples withdrawn at 20,40,60 and 80 minutes after a venous injection of vitamin B 12 - 57 Co, the glomerular filtration-rate(GFR) in 11 patients, performing simultaneously the same study with EDTA- 51 Cr in 3 of them, is evaluated. The values obtained through the regression straight line are compared with those given by only 2 points, in the 6 possible combinations: 20 and 40 minutes, 20 and 60 minutes, 20 and 80 minutes, 40 and 60 minutes, 40 and 80 minutes, 60 and 80 minutes. The pair of points obtained at 20 and 80 minutes determined the straight line most similar to the logarithmic regression and as a simplification of the method, the withdraw of only 2 plasma samples, at and 80 minutes after a single injection of vitamin B 12 -57 Co is proposed [pt

  16. Alternative Fuels

    Science.gov (United States)

    Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

  17. Fuel assembly

    International Nuclear Information System (INIS)

    Chaki, Masao; Nishida, Koji; Karasawa, Hidetoshi; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Kashiwai, Shin-ichi; Masuhara, Yasuhiro

    1998-01-01

    The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ≥ 173m, Y ≤ - 9.7X + 292, Y ≤ - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

  18. Fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Hajime.

    1995-01-01

    In a fuel assembly having fuel rods of different length, fuel pellets of mixed oxides of uranium and plutonium are loaded to a short fuel rod. The volume ratio of a pellet-loaded portion to a plenum portion of the short fuel rod is made greater than the volume ratio of a fuel rod to which uranium fuel pellets are loaded. In addition, the volume of the plenum portion of the short fuel rod is set greater depending on the plutonium content in the loaded fuel pellets. MOX fuel pellets are loaded on the short fuel rods having a greater degree of freedom relevant to the setting for the volume of the plenum portion compared with that of a long rod fuel, and the volume of the plenum portion is ensured greater depending on the plutonium content. Even if a large amount of FP gas and He gas are discharged from the MOX fuels compared with that from the uranium fuels, the internal pressure of the MOX fuel rod during operation is maintained substantially identical with that of the uranium fuel rod, so that a risk of generating excess stresses applied to the fuel cladding tubes and rupture of fuels are greatly reduced. (N.H.)

  19. 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.

  20. Nuclear fuels

    International Nuclear Information System (INIS)

    Gangwani, Saloni; Chakrabortty, Sumita

    2011-01-01

    Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

  1. Definition of breeding gain for the closed fuel cycle and application to a gas cooled fast reactor

    International Nuclear Information System (INIS)

    Van Rooijen, W. F. G.; Kloosterman, J. L.; Van Der Hagen, T. H. J. J.; Van Dam, H.

    2006-01-01

    In this paper a definition is given for the Breeding Gain (BG) of a nuclear reactor, taking into account compositional changes of the fuel during irradiation, cool down and reprocessing. A definition is given for the reactivity weights required to calculate BG. To calculate the effects of changes in the initial fuel composition on BG, first order nuclide perturbation theory is used. The theory is applied to the fuel cycle of GFR600, a 600 MWth Generation IV Gas Cooled Fast Reactor. This reactor should have a closed fuel cycle, with a BG equal to zero, breeding just enough new fuel during irradiation to allow refueling by only adding fertile material. All Heavy Metal is recycled in the closed fuel cycle. The result is that a closed fuel cycle is possible if the reprocessing has low losses ( 238 U, 15% Pu, and low amounts of the Minor Actinides. (authors)

  2. Fuel and nuclear fuel cycle

    International Nuclear Information System (INIS)

    Prunier, C.

    1998-01-01

    The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Sakuyama, Tadashi; Mukai, Hideyuki.

    1988-01-01

    Purpose: To prevent the bending of a fuel rod caused by the difference in the elongation between a joined fuel rod and a standard fuel rod thereby maintain the fuel rod integrity. Constitution: A joined fuel rod is in a thread engagement at its lower end plug thereof with a lower plate, while passed through at its upper end plug into an upper tie plate and secured with a nut. Further, a standard fuel rod is engaged at its upper end plug and lower end plug with the upper tie plate and the lower tie plate respectively. Expansion springs are mounted to the upper end plugs of these bonded fuel rods and the standard fuel rods for preventing this lifting. Each of the fuel rods comprises a plurality of sintered pellets of nuclear fuel materials laminated in a zircaloy fuel can. The content of the alloy ingredient in the fuel can of the bonded fuel rod is made greater than that of the alloy ingredient of the standard fuel rod. this can increase the elongation for the bonded fuel rod, and the spring of the standard fuel rod is tightly bonded to prevent the bending. (Yoshino, Y.)

  4. Gd-EOB-DTPA enhanced MRI of the liver: Correlation of relative hepatic enhancement, relative renal enhancement, and liver to kidneys enhancement ratio with serum hepatic enzyme levels and eGFR

    Energy Technology Data Exchange (ETDEWEB)

    Talakic, Emina; Steiner, Jürgen; Kalmar, Peter; Lutfi, Andre [Division of General Radiology, Department of Radiology, Medical University of Graz, Auenbruggerplatz 9, 8036 Graz (Austria); Quehenberger, Franz [Institute for Medical Informatics, Statistics and Documentation, Medical University of Graz, Auenbruggerplatz 2, 8036 Graz (Austria); Reiter, Ursula; Fuchsjäger, Michael [Division of General Radiology, Department of Radiology, Medical University of Graz, Auenbruggerplatz 9, 8036 Graz (Austria); Schöllnast, Helmut, E-mail: helmut.schoellnast@medunigraz.at [Division of General Radiology, Department of Radiology, Medical University of Graz, Auenbruggerplatz 9, 8036 Graz (Austria)

    2014-04-15

    Objectives: To assess the correlation of relative hepatic enhancement (RHE), relative renal enhancement (RRE) and liver to kidneys enhancement ratio (LKR) with serum hepatic enzyme levels and eGFR in Gd-EOB-DTPA enhanced MRI of the liver and to assess threshold levels for predicting enhancement of the liver parenchyma. Methods: Data of 75 patients who underwent Gd-EOB-DTPA enhanced MRI of the liver were collected. Images were obtained before contrast injection, during the early arterial phase, late arterial phase, venous phase, delayed phase, and hepatobiliary phase which was 20 min after Gd-EOB-DTPA administration. Signal intensity of the liver and the kidneys in all phases was defined using region-of-interest measurements for relative enhancement calculation. Serum hepatic enzyme levels and eGFR were available in all patients. Spearman correlation test was used to test the correlation of RHE, RRE and LKR with serum hepatic enzyme levels and eGFR. Results: In the hepatobiliary phase all serum hepatic enzymes were significantly correlated with RHE; total bilirubin (TBIL) and cholin esterase (CHE) showed strongest correlations. TBIL and CHE were significantly correlated with RRE in the arterial phases. TBIL and CHE were significantly correlated with LKR in the arterial phase and hepatobiliary phase. eGFR showed no correlation. Conclusions: In Gd-EOB-DTPA enhanced MRI, TBIL and CHE levels may predict RHE, RRE and LKR.

  5. Fuel processing

    International Nuclear Information System (INIS)

    Allardice, R.H.

    1990-01-01

    The technical and economic viability of the fast breeder reactor as an electricity generating system depends not only upon the reactor performance but also on a capability to recycle plutonium efficiently, reliably and economically through the reactor and fuel cycle facilities. Thus the fuel cycle is an integral and essential part of the system. Fuel cycle research and development has focused on demonstrating that the challenging technical requirements of processing plutonium fuel could be met and that the sometimes conflicting requirements of the fuel developer, fuel fabricator and fuel reprocessor could be reconciled. Pilot plant operation and development and design studies have established both the technical and economic feasibility of the fuel cycle but scope for further improvement exists through process intensification and flowsheet optimization. These objectives and the increasing processing demands made by the continuing improvement to fuel design and irradiation performance provide an incentive for continuing fuel cycle development work. (author)

  6. Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

    International Nuclear Information System (INIS)

    Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae

    2016-01-01

    Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed

  7. Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

    2016-10-15

    Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed.

  8. Prevention of criticality accidents. Fuel elements storage; Prevencion de accidentes de criticidad. Almacenamiento de elementos combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Canavese, S I; Capadona, N M

    1991-12-31

    Before the need to store fuel elements of the plate type MTR (Materials Testing Reactors), produced with enriched uranium at 20% in U235 for research reactors, it requires the design of a deposit for this purpose, which will give intrinsic security at a great extent and no complaints regarding its construction, is required. (Author). [Espanol] Partiendo de la necesidad de almacenar elementos combustibles tipo placa MTR (Materials Testing Reactors), producidos con uranio enriquecido al 20% en U235 para reactores de investigacion, se requiere el diseno de un deposito para tal fin que brinde esencialmente un alto grado de seguridad intrinseca y que no ofrezca complicaciones en cuanto a su construccion. (Autor).

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

    International Nuclear Information System (INIS)

    Martin, M.M.

    1993-01-01

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

  10. Fuel assembly

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1970-01-01

    Herein disclosed is a fuel assembly in which a fuel rod bundle is easily detachable by rotating a fuel rod fastener rotatably mounted to the upper surface of an upper tie-plate supporting a fuel bundle therebelow. A locking portion at the leading end of each fuel rod protrudes through the upper tie-plate and is engaged with or separated from the tie-plate by the rotation of the fastener. The removal of a desired fuel rod can therefore be remotely accomplished without the necessity of handling pawls, locking washers and nuts. (Owens, K.J.)

  11. Nuclear fuel

    International Nuclear Information System (INIS)

    D Hondt, P.

    1998-01-01

    The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

  12. Fuel management

    International Nuclear Information System (INIS)

    Schwarz, E.R.

    1975-01-01

    Description of the operation of power plants and the respective procurement of fuel to fulfil the needs of the grid. The operation of the plants shall be optimised with respect to the fuel cost. (orig./RW) [de

  13. Fuel gases

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    This paper gives a brief presentation of the context, perspectives of production, specificities, and the conditions required for the development of NGV (Natural Gas for Vehicle) and LPG-f (Liquefied Petroleum Gas fuel) alternative fuels. After an historical presentation of 80 years of LPG evolution in vehicle fuels, a first part describes the economical and environmental advantages of gaseous alternative fuels (cleaner combustion, longer engines life, reduced noise pollution, greater natural gas reserves, lower political-economical petroleum dependence..). The second part gives a comparative cost and environmental evaluation between the available alternative fuels: bio-fuels, electric power and fuel gases, taking into account the processes and constraints involved in the production of these fuels. (J.S.)

  14. Fuel cycles

    International Nuclear Information System (INIS)

    Hawley, N.J.

    1983-05-01

    AECL publications, from the open literature, on fuels and fuel cycles used in CANDU reactors are listed in this bibliography. The accompanying index is by subject. The bibliography will be brought up to date periodically

  15. Nuclear fuels

    International Nuclear Information System (INIS)

    2008-01-01

    The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

  16. French LEU fuel for research reactor with emphasis on the Osiris experience of core conversion and reactor operation with the new fuel

    International Nuclear Information System (INIS)

    Cerles, J.-M.

    1981-09-01

    One of the various activities carried out in France concerned with the design, fabrication and development of nuclear fuels was the development by the CEA of a plate type fuel (Caramel fuel). A Caramel fuel element is in the form of a plate consisting of two tight covering zircaloy sheets in which the UO 2 platelets are confined themselves within the network of a zircaloy grid. The plane geometry provides an effective means of overcoming the drawback of poor uranium oxide conductivity, and makes it possible to combine high specific power with low fuel temperature. The chief advantages of this fuel are the following: it is a very low enriched fuel. It can be used in research reactors demanding high volumetric powers and neutron fluxes, with a required enrichment significantly lower than 20% 235 U. The difference between the densities of UO 2 matrix and U-Al, 10.3 and 1.6 g/cm respectively, leads to a higher uranium charge, making it possible to reduce the enrichment to between 3 and 10%. Owing to fuel dispersion, any loss of tightness only puts a small amount of fissile material in contact with the coolant, thus limiting any contamination of the primary circuit. Another safety factor is the operating temperature, which is considerably lower than the temperature at which fission gases are liberated

  17. Fuel pellet

    International Nuclear Information System (INIS)

    Hayashi, K.

    1980-01-01

    Fuel pellet for insertion into a cladding tube in order to form a fuel element or a fuel rod. The fuel pellet has got a belt-like projection around its essentially cylindrical lateral circumferential surface. The upper and lower edges in vertical direction of this belt-like projection are wave-shaped. The projection is made of the same material as the bulk pellet. Both are made in one piece. (orig.) [de

  18. 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.

  19. A comparative study to investigate burnup in research reactor fuel using two independent experimental methods

    International Nuclear Information System (INIS)

    Iqbal, M.; Mehmood, T.; Ayazuddin, S.K.; Salahuddin, A.; Pervez, S.

    2001-01-01

    Two independent experimental methods have been used for the comparative study of fuel burnup measurement in low enriched uranium, plate type research reactor. In the first method a gamma ray activity ratio method was employed. An experimental setup was established for gamma ray scanning using prior calibrated high purity germanium detector. The computer software KORIGEN gave the theoretical support. In the second method reactivity difference technique was used. At the same location in the same core configuration the fresh and burned fuel element's reactivity worth was estimated. For theoretical estimated curve, group cross-sections were generated using computer code WIMS-D/4, and three dimensional modeling was made by computer code CITATION. The measured burnup of different fuel elements using these methods were found to be in good agreement

  20. RERTR program activities related to the development and application of new LEU fuels

    International Nuclear Information System (INIS)

    Travelli, A.

    1983-01-01

    The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U 3 Si 2 -Al and U 3 Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm 3 each year, from the current 1.7 g U/cm 3 to the 7.0 g U/cm 3 which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years

  1. Fossil Fuels.

    Science.gov (United States)

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  2. Fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A new fuel can with a loose bottom and head is described. The fuel bar is attached to the loose bottom and head with two grid poles keeping the distance between bottom and head. A bow-shaped handle is attached to the head so that the fuel bar can be lifted from the can

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-12-01

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

  4. LPG fuel

    International Nuclear Information System (INIS)

    Dagnas, F.X.; Jeuland, N.; Fouquet, J.P.; Lauraire, S.; Coroller, P.

    2005-01-01

    LPG fuel has become frequently used through a distribution network with 2 000 service stations over the French territory. LPG fuel ranks number 3 world-wide given that it can be used on individual vehicles, professional fleets, or public transport. What is the environmental benefit of LPG fuel? What is the technology used for these engines? What is the current regulation? Government commitment and dedication on support to promote LPG fuel? Car makers projects? Actions to favour the use of LPG fuel? This article gathers 5 presentations about this topic given at the gas conference

  5. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto; Ogiya, Shunsuke.

    1989-01-01

    For improving the economy of a BWR type reactor by making the operation cycle longer, the fuel enrichment degree has to be increased further. However, this makes the subcriticality shallower in the upper portion of the reactor core, to bring about a possibility that the reactor shutdown becomes impossible. In the present invention, a portion of fuel rod is constituted as partial length fuel rods (P-fuel rods) in which the entire stack length in the effective portion is made shorter by reducing the concentration of fissionable materials in the axial portion. A plurality of moderator rods are disposed at least on one diagonal line of a fuel assembly and P-fuel rods are arranged at a position put between the moderator rods. This makes it possible to reactor shutdown and makes the axial power distribution satisfactory even if the fuel enrichment degree is increased. (T.M.)

  6. Fuel Services

    International Nuclear Information System (INIS)

    Silberstein, A.

    1982-09-01

    FRAGEMA has developed most types of inspection equipments to work on irradiated fuel assemblies and on single fuel rods during reactor outages with an efficiency compatible with the utilities operating priorities. In order to illustrate this statement, two specific examples of inspection equipments are shortly described: the on-site removable fuel rod assembly examination stand, and the fuel assembly multiple examination device. FRAGEMA has developed techniques for the identifiction of the leaking fuel rods in the fuel assembly and the tooling necessary to perform the replacement of the faulted element. These examples of methods, techniques and equipments described and the experience accumulated through their use allow FRAGEMA to qualify for offering the supply of the corresponding software, hardware or both whenever an accurate understanding of the fuel behaviour is necessary and whenever direct intervention on the assembly and associated components is necessary due to safety, operating or economical reasons

  7. Fuel assembly

    International Nuclear Information System (INIS)

    Watanabe, Shoichi; Hirano, Yasushi.

    1998-01-01

    A one-half or more of entire fuel rods in a fuel assembly comprises MOX fuel rods containing less than 1wt% of burnable poisons, and at least a portion of the burnable poisons comprises gadolinium. Then, surplus reactivity at an initial stage of operation cycle is controlled to eliminate burnable poisons remained unburnt at a final stage, as well as increase thermal reactivity. In addition, the content of fission plutonium is determined to greater than the content of uranium 235, and fuel rods at corner portions are made not to incorporate burnable poisons. Fuel rods not containing burnable poisons are disposed at positions in adjacent with fuel rods facing to a water rod at one or two directions. Local power at radial center of the fuel assembly is increased to flatten the distortion of radial power distribution. (N.H.)

  8. Bilirubin concentration is positively associated with haemoglobin concentration and inversely associated with albumin to creatinine ratio among Indigenous Australians: eGFR Study.

    Science.gov (United States)

    Hughes, J T; Barzi, F; Hoy, W E; Jones, G R D; Rathnayake, G; Majoni, S W; Thomas, M A B; Sinha, A; Cass, A; MacIsaac, R J; O'Dea, K; Maple-Brown, L J

    2017-12-01

    Low serum bilirubin concentrations are reported to be strongly associated with cardio-metabolic disease, but this relationship has not been reported among Indigenous Australian people who are known to be at high risk for diabetes and chronic kidney disease (CKD). serum bilirubin will be negatively associated with markers of chronic disease, including CKD and anaemia among Indigenous Australians. A cross-sectional analysis of 594 adult Aboriginal and Torres Strait Islander (TSI) people in good health or with diabetes and markers of CKD. Measures included urine albumin: creatinine ratio (ACR), estimated glomerular filtration rate (eGFR), haemoglobin (Hb) and glycated haemoglobin (HbA1c). Diabetes was defined by medical history, medications or HbA1c≥6.5% or ≥48mmol/mol. Anaemia was defined as Hbbilirubin was performed. Participants mean (SD) age was 45.1 (14.5) years, and included 62.5% females, 71.7% Aboriginal, 41.1% with diabetes, 16.7% with anaemia, 41% with ACR>3mg/mmol and 18.2% with eGFRbilirubin concentration was lower in females than males (6 v 8μmol/L, pbilirubin; Hb and cholesterol (both positively related) and ACR, triglycerides, Aboriginal ethnicity and female gender (all inversely related). Serum bilirubin concentrations were positively associated with Hb and total cholesterol, and inversely associated with ACR. Further research to determine reasons explaining lower bilirubin concentrations among Aboriginal compared with TSI participants are needed. Copyright © 2017 The Canadian Society of Clinical Chemists. Published by Elsevier Inc. All rights reserved.

  9. Caramel fuel for research reactors: experience acquired in the fabrication, monitoring and irradiation of Osiris core

    International Nuclear Information System (INIS)

    Contenson, Ghislain de; Foulquier, Henri; Trotabas, Maria; Vignesoult, Nicole; Cerles, J.-M.; Delafosse, Jacques.

    1981-06-01

    A plate type nuclear fuel (Caramel fuel) has been developed in France in the framework of the various activities pursued in the design, fabrication and development of nuclear fuels by the CEA. This fuel can be adapted to various different categories of water cooled reactor (power reactors, marine propulsion reactors, urbain heating reactors, research reactors). The successful work conducted in this field led the realization of a complete core and reloads for the high performance research reactor, Osiris, at Saclay. The existing highly enriched U-Al alloy fuel was replaced by a non-proliferating low enrichment (7%) caramel fuel. This new core has been operating successfully since january 1980. A brief description of Caramel and its main advantages is given. The way in which it is fabricated is described together with the quality controls to which it is subjected. The qualification program and the main results deduced from it are also presented. The program used to monitor its in-pile behavior is described. The essential purpose of this program is to ensure the high performance of the fuel under irradiation. The successful operation of Osiris, which terminated 11 irradiation cycles on the 21st of April 1981 confirmed the correctness of the decisions made and the excellent performance that could be achieved with these fuel elements under the severe conditions encountered in a high performance research reactor [fr

  10. Fuel assembly

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi; Matsuzuka, Ryuji.

    1976-01-01

    Object: To provide a fuel assembly which can decrease pressure loss of coolant to uniform temperature. Structure: A sectional area of a flow passage in the vicinity of an inner peripheral surface of a wrapper tube is limited over the entire length to prevent the temperature of a fuel element in the outermost peripheral portion from being excessively decreased to thereby flatten temperature distribution. To this end, a plurality of pincture-frame-like sheet metals constituting a spacer for supporting a fuel assembly, which has a plurality of fuel elements planted lengthwise and in given spaced relation within the wrapper tube, is disposed in longitudinal grooves and in stacked fashion to form a substantially honeycomb-like space in cross section. The fuel elements are inserted and supported in the space to form a fuel assembly. (Kamimura, M.)

  11. Fuel assemblies

    International Nuclear Information System (INIS)

    Nagano, Mamoru; Yoshioka, Ritsuo

    1983-01-01

    Purpose: To effectively utilize nuclear fuels by increasing the reactivity of a fuel assembly and reduce the concentration at the central region thereof upon completion of the burning. Constitution: A fuel assembly is bisected into a central region and a peripheral region by disposing an inner channel box within a channel box. The flow rate of coolants passing through the central region is made greater than that in the peripheral region. The concentration of uranium 235 of the fuel rods in the central region is made higher. In such a structure, since the moderating effect in the central region is improved, the reactivity of the fuel assembly is increased and the uranium concentration in the central region upon completion of the burning can be reduced, fuel economy and effective utilization of uranium can be attained. (Kamimura, M.)

  12. Fuel assembly

    International Nuclear Information System (INIS)

    Bando, Masaru.

    1993-01-01

    As neutron irradiation progresses on a fuel assembly of an FBR type reactor, a strong force is exerted to cause ruptures if the arrangement of fuel elements is not displaced, whereas the fuel elements may be brought into direct contact with each other not by way of spacers to cause burning damages if the arrangement is displaced. In the present invention, the circumference of fuel elements arranged in a normal triangle lattice is surrounded by a wrapper tube having a hexagonal cross section, wire spacers are wound therearound, and deformable spacers are distributed to optional positions for fuel elements in the wrapper tube. Interaction between the fuel elements caused by irradiation is effectively absorbed, thereby enabling to delay the occurrence of the rupture and burning damages of the elements. (N.H.)

  13. Fuel assembly

    International Nuclear Information System (INIS)

    Yokota, Tokunobu.

    1990-01-01

    A fuel assembly used in a FBR type nuclear reactor comprises a plurality of fuel rods and a moderator guide member (water rod). A moderator exit opening/closing mechanism is formed at the upper portion of the moderator guide member for opening and closing a moderator exit. In the initial fuel charging operation cycle to the reactor, the moderator exit is closed by the moderator exit opening/closing mechanism. Then, voids are accumulated at the inner upper portion of the moderator guide member to harden spectrum and a great amount of plutonium is generated and accumulated in the fuel assembly. Further, in the fuel re-charging operation cycle, the moderator guide member is used having the moderator exit opened. In this case, voids are discharged from the moderator guide member to decrease the ratio, and the plutonium accumulated in the initial charging operation cycle is burnt. In this way, the fuel economy can be improved. (I.N.)

  14. Fuel spacer

    International Nuclear Information System (INIS)

    Nishida, Koji; Yokomizo, Osamu; Kanazawa, Toru; Kashiwai, Shin-ichi; Orii, Akihito.

    1992-01-01

    The present invention concerns a fuel spacer for a fuel assembly of a BWR type reactor and a PTR type reactor. Springs each having a vane are disposed on the side surface of a circular cell which supports a fuel rods. A vortex streams having a vertical component are formed by the vanes in the flowing direction of a flowing channel between adjacent cylindrical cells. Liquid droplets carried by streams are deposited on liquid membrane streams flowing along the fuel rod at the downstream of the spacer by the vortex streams. In view of the above, the liquid droplets can be deposited to the fuel rod without increasing the amount of metal of the spacer. Accordingly, the thermal margin of the fuel assembly can be improved without losing neutron economy. (I.N.)

  15. Fuel Cells

    DEFF Research Database (Denmark)

    Smith, Anders; Pedersen, Allan Schrøder

    2014-01-01

    Fuel cells have been the subject of intense research and development efforts for the past decades. Even so, the technology has not had its commercial breakthrough yet. This entry gives an overview of the technological challenges and status of fuel cells and discusses the most promising applications...... of the different types of fuel cells. Finally, their role in a future energy supply with a large share of fluctuating sustainable power sources, e.g., solar or wind, is surveyed....

  16. Fuel cycle

    International Nuclear Information System (INIS)

    Bahm, W.

    1989-01-01

    The situation of the nuclear fuel cycle for LWR type reactors in France and in the Federal Republic of Germany was presented in 14 lectures with the aim to compare the state-of-the-art in both countries. In addition to the momentarily changing fuilds of fuel element development and fueling strategies, the situation of reprocessing, made interesting by some recent developmnts, was portrayed and differences in ultimate waste disposal elucidated. (orig.) [de

  17. Nuclear fuel

    International Nuclear Information System (INIS)

    Azevedo, J.B.L. de.

    1980-01-01

    All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.) [pt

  18. Fuel assembly

    International Nuclear Information System (INIS)

    Nomata, Terumitsu.

    1993-01-01

    Among fuel pellets to be loaded to fuel cans of a fuel assembly, fuel pellets having a small thermal power are charged in a region from the end of each of spacers up to about 50mm on the upstream of coolants that flow vertically at the periphery of fuel rods. Coolants at the periphery of fuel rods are heated by the heat generation, to result in voids. However, since cooling effect on the upstream of the spacers is low due to influences of the spacers. Further, since the fuel pellets disposed in the upstream region have small thermal power, a void coefficient is not increased. Even if a thermal power exceeding cooling performance should be generated, there is no worry of causing burnout in the upstream region. Even if burnout should be caused, safety margin and reliability relative to burnout are improved, to increase an allowable thermal power, thereby enabling to improve integrity and reliability of fuel rods and fuel assemblies. (N.H.)

  19. Fuel assembly

    International Nuclear Information System (INIS)

    Gjertsen, R.K.; Bassler, E.A.; Huckestein, E.A.; Salton, R.B.; Tower, S.N.

    1988-01-01

    A fuel assembly adapted for use with a pressurized water nuclear reactor having capabilities for fluid moderator spectral shift control is described comprising: parallel arranged elongated nuclear fuel elements; means for providing for axial support of the fuel elements and for arranging the fuel elements in a spaced array; thimbles interspersed among the fuel elements adapted for insertion of a rod control cluster therewithin; means for structurally joining the fuel elements and the guide thimbles; fluid moderator control means for providing a volume of low neutron absorbing fluid within the fuel assembly and for removing a substantially equivalent volume of reactor coolant water therefrom, a first flow manifold at one end of the fuel assembly sealingly connected to a first end of the moderator control tubes whereby the first ends are commonly flow connected; and a second flow manifold, having an inlet passage and an outlet passage therein, sealingly connected to a second end of the moderator control tubes at a second end of the fuel assembly

  20. Development of the control assembly pattern and dynamic analysis of the generation IV large gas-cooled fast reactor (GFR); Developpement du design d'un assemblage de controle et analyse dynamique des reacteurs a neutrons rapides de quatrieme generation refroidis au gaz

    Energy Technology Data Exchange (ETDEWEB)

    Girardin, G.

    2009-07-09

    Among the systems selected by the GIF, the Gas-cooled Fast Reactor (GFR) is a highly innovative system with advanced fuel geometry and materials. It is in the context of the large, 2400 MWth reference GFR design that the present doctoral research has been conducted, the principal aim having been to develop and qualify the control assembly (CA) pattern and corresponding CA implementation scheme for this system. The work has been carried out in three successive and complementary phases: (1) validation of the neutronics tools, (2) the CA pattern development and related static analysis, and (3) dynamic core behavior studies for hypothetical CA driven transients. During the first phase of the thesis, the reference PROTEUS test lattice from these experiments has been analyzed with ERANOS-2.0 and its associated, adjusted nuclear data library ERALIB1. Additionally, benchmark calculations were performed with the Monte Carlo code MCNPX, allowing one to both check the deterministic results and to analyze the sensitivity to different modern data libraries. It has been found that, for the main reaction rate ratios, the new analysis of the GCFR-PROTEUS reference lattice generally yields good agreement - within 1{sigma} measurement uncertainty - with experimental values and with the Monte Carlo simulations. As shown by the analysis, the predictions were in somewhat better agreement in the case of the adjusted ERALIB1 library. The applicability of ERANOS-2.0/ERALIB1 as the reference neutronics tool for the GFR analysis could thus be demonstrated. Furthermore, neutronics aspects related to the novel features of the GFR, for which new experimental investigations are needed, were highlighted. In the second phase of the research, the CA pattern was developed for the GFR, based on iterative neutronics and thermal-hydraulics calculations, 2D and 3D neutronics models for the reactor core having first been set up using the reference ERANOS-2.0/ERALIB1 computational scheme. For the thermal

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

    International Nuclear Information System (INIS)

    Martin, M.M.

    1993-01-01

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

  2. Fuel element

    International Nuclear Information System (INIS)

    Kennedy, S.T.

    1982-01-01

    A nuclear reactor fuel element wherein a stack of nuclear fuel is prevented from displacement within its sheath by a retainer comprising a tube member which is radially expanded into frictional contact with the sheath by means of a captive ball within a tapered bore. (author)

  3. Nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, H [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

    1976-10-01

    It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts.

  4. Fuel cells

    NARCIS (Netherlands)

    Veen, van J.A.R.; Janssen, F.J.J.G.; Santen, van R.A.

    1999-01-01

    The principles and present-day embodiments of fuel cells are discussed. Nearly all cells are hydrogen/oxygen ones, where the hydrogen fuel is usually obtained on-site from the reforming of methane or methanol. There exists a tension between the promise of high efficiency in the conversion of

  5. Fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Akiyoshi; Bessho, Yasunori; Aoyama, Motoo; Koyama, Jun-ichi; Hirakawa, Hiromasa; Yamashita, Jun-ichi; Hayashi, Tatsuo

    1998-01-01

    In a fuel assembly of a BWR type reactor in which a water rod of a large diameter is disposed at the central portion, the cross sectional area perpendicular to the axial direction comprises a region a of a fuel rod group facing to a wide gap water region to which a control rod is inserted, a region b of a fuel rod group disposed on the side of the wide gap water region other than the region a, a region d of a fuel rod group facing to a narrow gap water region and a region c of a fuel rod group disposed on the side of the narrow gap water region other than the region d. When comparing an amount of fission products contained in the four regions relative to that in the entire regions and average enrichment degrees of fuel rods for the four regions, the relative amount and the average enrichment degree of the fuel rod group of the region a is minimized, and the relative amount and the average enrichment degree of the fuel rod group in the region b is maximized. Then, reactor shut down margin during cold operation can be improved while flattening the power in the cross section perpendicular to the axial direction. (N.H.)

  6. Nuclear fuel

    International Nuclear Information System (INIS)

    Quinauk, J.P.

    1990-01-01

    Since 1985, Fragema has been marketing and selling the Advanced Fuel Assemby AFA whose main features are its zircaloy grids and removable top and bottom nozzles. It is this product, which exists for several different fuel assembly arrays and heights, that will be employed in the reactors at Daya Bay. Fragema employs gadolinium as the consumable poison to enable highperformance fuel management. More recently, the company has supplied fuel assemblies of the mixed-oxide(MOX) and enriched reprocessed uranium type. The reliability level of the fuel sold by Fragema is one of the highest in the world, thanks in particular to the excellence of the quality assurance and quality control programs that have been implemented at all stages of its design and manufacture

  7. Fuel assemblies

    International Nuclear Information System (INIS)

    Echigoya, Hironori; Nomata, Terumitsu.

    1983-01-01

    Purpose: To render the axial distribution relatively flat. Constitution: First nuclear element comprises a fuel can made of zircalloy i.e., the metal with less neutron absorption, which is filled with a plurality of UO 2 pellets and sealed by using a lower end plug, a plenum spring and an upper end plug by means of welding. Second fuel element is formed by substituting a part of the UO 2 pellets with a water tube which is sealed with water and has a space for allowing the heat expansion. The nuclear fuel assembly is constituted by using the first and second fuel elements together. In such a structure, since water reflects neutrons and decrease their leakage to increase the temperature, reactivity is added at the upper portion of the fuel assembly to thereby flatten the axial power distribution. Accordingly, stable operation is possible only by means of deep control rods while requiring no shallow control rods. (Sekiya, K.)

  8. Studies on capacity management for factories of nuclear fuel for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Negro, Miguel Luiz Miotto; Durazzo, Michelangelo; Mesquita, Marco Aurélio de; Carvalho, Elita Fontenele Urano de; Andrade, Delvonei Alves de, E-mail: mlnegro@ipen.br, E-mail: mdurazzo@ipen.br, E-mail: elitaucf@ipen.br, E-mail: delvonei@ipen.br, E-mail: mamesqui@usp.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Escola Politécnica. Departamento de Engenharia de Produção

    2017-11-01

    The use and the power of nuclear reactors for research and materials testing is increasing worldwide. That implies the demand for nuclear fuel for this kind of reactors is rising. Thus, the production facilities of this kind of fuel need reliable guidance on how to augment their production in order to meet the increasing demand efficiently, safely and keeping good quality. Focus is given to factories that produce plate type fuel elements loaded with LEU U{sub 3}Si{sub 2}-Al fuel, which are typically used in nuclear research reactors. Of the various production routes for this kind of fuel, we chose the route which uses hydrolysis of uranium hexafluoride. Raising the capacity of this kind of plants faces several problems, especially regarding safety against nuclear criticality. Some of these problems are briefly addressed. The new issue of the paper is the application of knowledge from the area of production administration to the fabrication of nuclear fuel for research reactors. A specific method for the increase in production capacity is proposed. That method was tested by means of discrete event simulation. The data were collected from the nuclear fuel factory at IPEN. The results indicated the proposed method achieved its goal as well as ways of raising production capacity in up to 50%. (author). (author)

  9. Studies on capacity management for factories of nuclear fuel for research reactors

    International Nuclear Information System (INIS)

    Negro, Miguel Luiz Miotto; Durazzo, Michelangelo; Mesquita, Marco Aurélio de; Carvalho, Elita Fontenele Urano de; Andrade, Delvonei Alves de; Universidade de São Paulo

    2017-01-01

    The use and the power of nuclear reactors for research and materials testing is increasing worldwide. That implies the demand for nuclear fuel for this kind of reactors is rising. Thus, the production facilities of this kind of fuel need reliable guidance on how to augment their production in order to meet the increasing demand efficiently, safely and keeping good quality. Focus is given to factories that produce plate type fuel elements loaded with LEU U_3Si_2-Al fuel, which are typically used in nuclear research reactors. Of the various production routes for this kind of fuel, we chose the route which uses hydrolysis of uranium hexafluoride. Raising the capacity of this kind of plants faces several problems, especially regarding safety against nuclear criticality. Some of these problems are briefly addressed. The new issue of the paper is the application of knowledge from the area of production administration to the fabrication of nuclear fuel for research reactors. A specific method for the increase in production capacity is proposed. That method was tested by means of discrete event simulation. The data were collected from the nuclear fuel factory at IPEN. The results indicated the proposed method achieved its goal as well as ways of raising production capacity in up to 50%. (author). (author)

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  11. Development of Coated Particle Fuel Technology

    International Nuclear Information System (INIS)

    Lee, Young Woo; Kim, B. G.; Kim, S. H.

    2007-06-01

    Uranium kernel fabrication technology using a wet chemical so-gel method, a key technology in the coated particle fuel area, is established up to the calcination step and the first sintering of UO2 kernel was attempted. Experiments on the parametric study of the coating process using the surrogate ZrO2 kernel give the optimum conditions for the PyC and SiC coating layer and ZrC coating conditions were obtained for the vaporization of the ZrCl4 precursor and coating condition from ZrC coating experiments using plate-type graphite substrate. In addition, by development of fuel performance analysis code a part of the code system is completed which enables the participation to the benchmark calculation and comparison in the IAEA collaborated research program. The technologies for irradiation and post irradiation examination, which are important in developing the HTGR fuel technology of its first kind in Korea was started to develop and, through a feasibility study and preliminary analysis, the technologies required to be developed are identified for further development as well as the QC-related basic technologies are reviewed, analyzed and identified for the own technology development. Development of kernel fabrication technology can be enhanced for the remaining sintering technology and completed based on the technologies developed in this phase. In the coating technology, the optimum conditions obtained using a surrogate ZrO2 kernel material can be applied for the uranium kernel coating process development. Also, after completion of the code development in the next phase, more extended participation to the international collaboration for benchmark calculation can be anticipated which will enable an improvement of the whole code system. Technology development started in this phase will be more extended and further focused on the detailed technology development to be required for the related technology establishment

  12. Fuel cells:

    DEFF Research Database (Denmark)

    Sørensen, Bent

    2013-01-01

    A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil and nucl......A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil...... and nuclear fuel-based energy technologies....

  13. Fuel assemblies

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi.

    1979-01-01

    Purpose: To prevent scattering of gaseous fission products released from fuel assemblies stored in an fbr type reactor. Constitution; A cap provided with means capable of storing gas is adapted to amount to the assembly handling head, for example, by way of threading in a storage rack of spent fuel assemblies consisting of a bottom plate, a top plate and an assembly support mechanism. By previously eliminating the gas inside of the assembly and the cap in the storage rack, gaseous fission products upon loading, if released from fuel rods during storage, are stored in the cap and do not scatter in the storage rack. (Horiuchi, T.)

  14. GEN IV: Carbide Fuel Elaboration for the 'Futurix Concepts' experiment

    International Nuclear Information System (INIS)

    Vaudez, Stephane; Riglet-Martial, Chantal; Paret, Laurent; Abonneau, Eric

    2008-01-01

    In order to collect information on the behaviour of the future GFR (Gas Fast Reactor) fuel under fast neutron irradiation, an experimental irradiation program, called 'Futurix-concepts' has been launched at the CEA. The considered concept is a composite material made of a fissile fuel embedded in an inert ceramic matrix. Fissile fuel pellets are made of UPuN or UPuC while ceramics are SiC for the carbide fuel and TiN for the nitride fuel. This paper focuses on the description of the carbide composite fabrication. The UPuC pellets are manufactured using a metallurgical powder process. Fabrication and handling of the fuels are carried out in glove boxes under a nitrogen atmosphere. Carbide fuel is synthesized by carbo-thermic reduction under vacuum of a mixture of actinide oxide and graphitic carbon up to 1550 deg. C. After ball milling, the UPuC powder is pressed to create hexagonal or spherical compacts. They are then sintered up to 1750 deg. C in order to obtain a density of 85 % of the theoretical one. The sintered pellets are inserted into an inert and tight capsule of SiC. In order to control the gap between the fuel and the matrix precisely, the pellets are abraded. The inert matrix is then filled with the pellets and the whole system is sealed by a BRASiC R process at high temperature under a helium atmosphere. Fabrication of the sample to be irradiated was done in 2006 and the irradiation began in May 2007 in the Phenix reactor. This presentation will detail and discuss the results obtained during this fabrication phase. (authors)

  15. Is There Association Between Changes in eGFR Value and the Risk of Permanent Type of Atrial Fibrillation? - Analysis of Valvular and Non-Valvular Atrial Fibrillation Population

    Directory of Open Access Journals (Sweden)

    Elzbieta Mlodawska

    2014-12-01

    Full Text Available Background/Aims: There are no data concerning renal function in population with valvular and non-valvular atrial fibrillation (AF. To assess renal function in patients with AF, the association between eGFR and AF perpetuation, in-hospital mortality. Methods: We studied 1523 patients with AF. Patients with chronic kidney disease (CKD were compared to population with preserved renal function. Results: CKD was more frequently observed in patients with valvular AF(p=0.009. In non-valvular AF patients eGFR 2 had more often permanent AF(p2DS2VASc score was 4.1±1.5 and HAS-BLED score was 2.1±1.2 and it was higher as compared to population with preserved renal function (p75 years old(OR=3.70,p=0.01,95%CI1.33-10.28, with CKD (OR=2.61,p=0.03,95%CI1.09-6.23. The type of AF had no significant influence on in-hospital mortality(OR=0.71,p=0.45,95%CI0.30-1.70. Conclusions: CKD is more often observed in patients with valvular AF. In population with non-valvular AF decreased eGFR is associated with permanent type of AF and with higher CHA2DS2VASc and HAS-BLED score. Among valvular AF patients there are no differences in type of AF between patients with and without CKD. There is the correlation between CKD and AF perpetuation but only in non-valvular population.

  16. Fuel behaviour

    International Nuclear Information System (INIS)

    Fodor, M.; Matus, L.; Vigassy, J.

    1987-11-01

    A short summary of the main critical points in fuel performance of nuclear power reactors from chemical and mechanical point of view is given. A schedule for a limited research program is included. (author) 17 refs

  17. Fuel cells

    International Nuclear Information System (INIS)

    Niederdoeckl, J.

    2001-01-01

    Europe has at present big hopes on the fuel cells technology, in comparison with other energy conversion technologies, this technology has important advantages, for example: high efficiency, very low pollution and parallel use of electric and thermal energy. Preliminary works for fuel cells developing and its commercial exploitation are at full speed; until now the European Union has invested approx. 1.7 billion Schillings, 60 relevant projects are being executed. The Austrian industry is interested in applying this technique to drives, thermal power stations and the miniature fuel cells as replacement of batteries in electronic products (Notebooks, mobile telephones, etc.). A general description of the historic development of fuel cells including the main types is given as well as what is the situation in Austria. (nevyjel)

  18. Fuel assembly

    International Nuclear Information System (INIS)

    Ishibashi, Yoko; Aoyama, Motoo; Oyama, Jun-ichi.

    1995-01-01

    Burnable poison-incorporating fuel rods of a first group are disposed in a region in adjacent with a water rod having a large diameter (neutron moderator rod) disposed to the central portion of a fuel assembly. Burnable poison-incorporating fuel rods of a second group are disposed to a region other than peripheral zone in adjacent with a channel box and corners positioned at an inner zone, in adjacent with the channel box. The average concentration of burnable poisons of the burnable poison-incorporating fuel rods of the first group is made greater than that of the second group. With such a constitution, when the burnable poisons of the first group are burnt out, the burnable poisons of the second group are also burnt out at the same time. Accordingly, an amount of burnable poisons left unburnt at the final stage of the operation cycle is reduced, to improve the reactivity. This can improve the economical property. (I.N.)

  19. Fuel element

    International Nuclear Information System (INIS)

    Armijo, J.S.

    1976-01-01

    A fuel element for nuclear reactors is proposed which has a higher corrosion resisting quality in reactor operations. The zirconium alloy coating around the fuel element (uranium or plutonium compound) has on its inside a protection layer of metal which is metallurgically bound to the substance of the coating. As materials are namned: Alluminium, copper, niobium, stainless steel, and iron. This protective metallic layer has another inner layer, also metallurgically bound to its surface, which consists usually of a zirconium alloy. (UWI) [de

  20. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Sei; Ando, Ryohei; Mitsutake, Toru.

    1995-01-01

    The present invention concerns a fuel assembly suitable to a BWR-type reactor and improved especially with the nuclear characteristic, heat performance, hydraulic performance, dismantling or assembling performance and economical property. A part of poison rods are formed as a large-diameter/multi-region poison rods having a larger diameter than a fuel rod. A large number of fuel rods are disposed surrounding a large diameter water rod and a group of the large-diameter/multi-region poison rods in adjacent with the water rod. The large-diameter water rod has a burnable poison at the tube wall portion. At least a portion of the large-diameter poison rods has a coolant circulation portion allowing coolants to circulate therethrough. Since the large-diameter poison rods are disposed at a position of high neutron fluxes, a large neutron multiplication factor suppression effect can be provided, thereby enabling to reduce the number of burnable poison rods relative to fuels. As a result, power peaking in the fuel assembly is moderated and a greater amount of plutonium can be loaded. In addition the flow of cooling water which tends to gather around the large diameter water rod can be controlled to improve cooling performance of fuels. (N.H.)

  1. 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)

  2. 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)

  3. The Analysis of the Effect of Coolant Channel Width on Fuel Loading of the RSG-GAS Core

    International Nuclear Information System (INIS)

    Surbakti; Tukiran

    2004-01-01

    The RGS-GAS using uranium silicide fuel, plate type and 250 g U of loading is planned to increase the fuel loading to 300 g U even to 400 g U. The silicide fuel has advantages when increase the fuel loading in the same volume. Because of that case, it is necessary to analyze the effect of coolant channel width on fuel loading of the RSG-GAS core. Analyzing the effect the work which done is to generate cell and core calculation using WIMSD/4 and Batan-2DIFF codes. The WIMSD/4 code is used to generate cross section of core material and Batan-2DIFF is used to calculate the effective multiplication factor. The model that used in this calculation there are three kind of fuel loading namely, 250 g U, 300 g U and 400 g U. The coolant channel width is simulated from 1.75 mm to 2.55 mm. From that fuel loadings, it is analyzed which coolant channel width gave the best effective multiplication factor. From result of analysis showed that the best effective multiplication factor is on the coolant channel width of 2.55 mm for third of fuel loadings. (author)

  4. High temperature mechanisms and kinetics of SiC oxidation under low partial pressures of oxygen: application to the fuel cladding of gas fast reactors

    International Nuclear Information System (INIS)

    Hun, N.

    2011-01-01

    Gas Fast Reactor (GFR) is one of the different Generation IV concepts under investigation for energy production. SiC/SiC composites are candidates of primary interest for a GFR fuel cladding use, thanks to good corrosion resistance among other properties. The mechanisms and kinetics of SiC oxidation under operating conditions have to be identified and quantified as the corrosion can decrease the mechanical properties of the composite. An experimental device has been developed to study the oxidation of silicon carbide under high temperature and low oxygen partial pressure. The results pointed out that not only parabolic oxidation, but also interfacial reactions and volatilization occur under such conditions. After determining the kinetics of each mechanism, as functions of oxygen partial pressure and temperature, the data are used for the modeling of the composites oxidation. The model will be used to predict the lifetime of the composite in operating conditions. (author) [fr

  5. Fuel rods

    International Nuclear Information System (INIS)

    Adachi, Hajime; Ueda, Makoto

    1985-01-01

    Purpose: To provide a structure capable of measuring, in a non-destructive manner, the releasing amount of nuclear gaseous fission products from spent fuels easily and at a high accuracy. Constitution: In order to confirm the integrity and the design feasibility of a nuclear fuel rod, it is important to accurately determine the amount of gaseous nuclear fission products released from nuclear pellets. In a structure where a plurality of fuel pellets are charged in a fuel cladding tube and retained by an inconel spring, a hollow and no-sealed type spacer tube made of zirconium or the alloy thereof, for example, not containing iron, cobalt, nickel or manganese is formed between the spring and the upper end plug. In the fuel rod of such a structure, by disposing a gamma ray collimator and a gamma ray detector on the extension of the spacer pipe, the gamma rays from the gaseous nuclear fission products accumulated in the spacer pipe can be detected while avoiding the interference with the induction radioactivity from inconel. (Kamimura, M.)

  6. Fuel rods

    International Nuclear Information System (INIS)

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

    Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

  7. Fuel assembly

    International Nuclear Information System (INIS)

    Abe, Hideaki; Sakai, Takao; Ishida, Tomio; Yokota, Norikatsu.

    1992-01-01

    The lower ends of a plurality of plate-like shape memory alloys are secured at the periphery of the upper inside of the handling head of a fuel assembly. As the shape memory alloy, a Cu-Zn alloy, a Ti-Pd alloy or a Fe-Ni alloy is used. When high temperature coolants flow out to the handling head, the shape memory alloy deforms by warping to the outer side more greatly toward the upper portion thereof with the temperature increase of the coolants. As the result, the shape of the flow channel of the coolants is changed so as to enlarge at the exit of the upper end of the fuel assembly. Then, the pressure loss of the coolants in the fuel assembly is decreased by the enlargement. Accordingly, the flow rate of the coolants in the fuel assembly is increased to lower the temperature of the coolants. Further, high temperature coolants and low temperature coolants are mixed sufficiently just above the fuel assembly. This can suppress the temperature fluctuation of the mixed coolants in the upper portion of the reactor core, thereby enabling to decrease a fatigue and failures of the structural components in the upper portion of the reactor core. (I.N.)

  8. Fuel assembly

    International Nuclear Information System (INIS)

    Sano, Hiroki; Fushimi, Atsushi; Tominaga, Kenji; Aoyama, Motoo; Ishii, Kazuya.

    1997-01-01

    In burnable poison-incorporated uranium fuels of a BWR type reactor, the compositional ratio of isotopes of the burnable poisons is changed so as to increase the amount of those having a large neutron absorbing cross sectional area. For example, if the ratio of Gd-157 at the same burnable poison enrichment degree is made greater than the natural ratio, this gives the same effect as the increase of the enrichment degree per one fuel rod, thereby providing an effect of reducing a surplus reactivity. Gadolinium, hafnium and europium as burnable poisons have an absorbing cross sectional area being greater in odd numbered nuclei than in even numbered nuclei, on the contrary, boron has a cross section being greater in even numbered nucleus than odd numbered nuclei. Accordingly, if the ratio of isotopes having greater cross section at the same burnable poison enrichment degree is made greater than the natural ratio, surplus reactivity at the initial stage of the burning can be reduced without greatly increasing the amount of burnable poison-incorporated uranium fuels, fuel loading amount is not reduced and the fuel economy is not worsened. (N.H.)

  9. Canadian power reactor fuel

    International Nuclear Information System (INIS)

    Page, R.D.

    1976-03-01

    The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

  10. CANDU fuel

    International Nuclear Information System (INIS)

    MacEwan, J.R.; Notley, M.J.F.; Wood, J.C.; Gacesa, M.

    1982-09-01

    The direction of CANDU fuel development was set in 1957 with the decision to build pressure tube reactors. Short - 50 cm long - rodded bundles of natural UO 2 clad in Zircaloy were adopted to facilitate on-power fuelling to improve uranium utilization. Progressive improvements were made during 25 years of development, involving 650 man years and 180 million dollars. Today's CANDU bundle is based on the knowledge gained from extensive irradiation testing and experience in power reactors. The main thrust of future development is to demonstrate that the present bundle is suitable, with minor modifications, for thorium fuels

  11. COUPLED SIMULATION OF GAS COOLED FAST REACTOR FUEL ASSEMBLY WITH NESTLE CODE SYSTEM

    Directory of Open Access Journals (Sweden)

    Filip Osusky

    2018-05-01

    Full Text Available The paper is focused on coupled calculation of the Gas Cooled Fast Reactor. The proper modelling of coupled neutronics and thermal-hydraulics is the corner stone for future safety assessment of the control and emergency systems. Nowadays, the system and channel thermal-hydraulic codes are accepted by the national regulatory authorities in European Union for license purposes, therefore the code NESTLE was used for the simulation. The NESTLE code is a coupled multigroup neutron diffusion code with thermal-hydraulic sub-channel code. In the paper, the validation of NESTLE code 5.2.1 installation is presented. The processing of fuel assembly homogeneous parametric cross-section library for NESTLE code simulation is made by the sequence TRITON of SCALE code package system. The simulated case in the NESTLE code is one fuel assembly of GFR2400 concept with reflective boundary condition in radial direction and zero flux boundary condition in axial direction. The results of coupled calculation are presented and are consistent with the GFR2400 study of the GoFastR project.

  12. Fuels characterization studies. [jet fuels

    Science.gov (United States)

    Seng, G. T.; Antoine, A. C.; Flores, F. J.

    1980-01-01

    Current analytical techniques used in the characterization of broadened properties fuels are briefly described. Included are liquid chromatography, gas chromatography, and nuclear magnetic resonance spectroscopy. High performance liquid chromatographic ground-type methods development is being approached from several directions, including aromatic fraction standards development and the elimination of standards through removal or partial removal of the alkene and aromatic fractions or through the use of whole fuel refractive index values. More sensitive methods for alkene determinations using an ultraviolet-visible detector are also being pursued. Some of the more successful gas chromatographic physical property determinations for petroleum derived fuels are the distillation curve (simulated distillation), heat of combustion, hydrogen content, API gravity, viscosity, flash point, and (to a lesser extent) freezing point.

  13. Alternative Fuels Data Center: Ethanol Fueling Stations

    Science.gov (United States)

    ... More in this section... Ethanol Basics Benefits & Considerations Stations Locations Infrastructure fueling stations by location or along a route. Infrastructure Development Learn about ethanol fueling infrastructure; codes, standards, and safety; and ethanol equipment options. Maps & Data E85 Fueling Station

  14. Alternative Fuels Data Center: Biodiesel Fueling Stations

    Science.gov (United States)

    Locations Infrastructure Development Vehicles Laws & Incentives Biodiesel Fueling Stations Photo of a location or along a route. Infrastructure Development Learn about biodiesel fueling infrastructure codes Case Studies California Ramps Up Biofuels Infrastructure Green Fueling Station Powers Fleets in Upstate

  15. Fuels processing for transportation fuel cell systems

    Science.gov (United States)

    Kumar, R.; Ahmed, S.

    Fuel cells primarily use hydrogen as the fuel. This hydrogen must be produced from other fuels such as natural gas or methanol. The fuel processor requirements are affected by the fuel to be converted, the type of fuel cell to be supplied, and the fuel cell application. The conventional fuel processing technology has been reexamined to determine how it must be adapted for use in demanding applications such as transportation. The two major fuel conversion processes are steam reforming and partial oxidation reforming. The former is established practice for stationary applications; the latter offers certain advantages for mobile systems and is presently in various stages of development. This paper discusses these fuel processing technologies and the more recent developments for fuel cell systems used in transportation. The need for new materials in fuels processing, particularly in the area of reforming catalysis and hydrogen purification, is discussed.

  16. Fuel assembly

    International Nuclear Information System (INIS)

    Kurihara, Kunitoshi; Azekura, Kazuo.

    1992-01-01

    In a reactor core of a heavy water moderated light water cooled pressure tube type reactor, no sufficient effects have been obtained for the transfer width to a negative side of void reactivity change in a region of a great void coefficient. Then, a moderation region divided into upper and lower two regions is disposed at the central portion of a fuel assembly. Coolants flown into the lower region can be discharged to the cooling region from an opening disposed at the upper end portion of the lower region. Light water flows from the lower region of the moderator region to the cooling region of the reactor core upper portion, to lower the void coefficient. As a result, the reactivity performance at low void coefficient, i.e., a void reaction rate is transferred to the negative side. Thus, this flattens the power distribution in the fuel assembly, increases the thermal margin and enables rapid operaiton and control of the reactor core, as well as contributes to the increase of fuel burnup ratio and reduction of the fuel cycle cost. (N.H.)

  17. Fuel assembly

    International Nuclear Information System (INIS)

    Fushimi, Atsushi; Shimada, Hidemitsu; Aoyama, Motoo; Nakajima, Junjiro

    1998-01-01

    In a fuel assembly for an n x n lattice-like BWR type reactor, n is determined to 9 or greater, and the enrichment degree of plutonium is determined to 4.4% by weight or less. Alternatively, n is determined to 10 or greater, and the enrichment degree of plutonium is determined to 5.2% by weight or less. An average take-out burnup degree is determined to 39GWd/t or less, and the matrix is determined to 9 x 9 or more, or the average take-out burnup degree is determined to 51GWd/t, and the matrix is determined to 10 x 10 or more and the increase of the margin of the maximum power density obtained thereby is utilized for the compensation of the increase of distortion of power distribution due to decrease of the kinds of plutonium enrichment degree, thereby enabling to reduce the kind of the enrichment degree of MOX fuel rods to one. As a result, the manufacturing step for fuel pellets can be simplified to reduce the manufacturing cost for MOX fuel assemblies. (N.H.)

  18. Fuel rods

    International Nuclear Information System (INIS)

    Fukushima, Kimichika.

    1984-01-01

    Purpose: To reduce the size of the reactor core upper mechanisms and the reactor container, as well as decrease the nuclear power plant construction costs in reactors using liquid metals as the coolants. Constitution: Isotope capturing devices comprising a plurality of pipes are disposed to the gas plenum portion of a nuclear fuel rod main body at the most downstream end in the flowing direction of the coolants. Each of the capturing devices is made of nickel, nickel alloys, stainless steel applied with nickel plating on the surface, nickel alloys applied with nickel plating on the surface or the like. Thus, radioactive nuclides incorporated in the coolants are surely captured by the capturing devices disposed at the most downstream end of the nuclear fuel main body as the coolants flow along the nuclear fuel main body. Accordingly, since discharging of radioactive nuclides to the intermediate fuel exchange system can be prevented, the maintenance or reparing work for the system can be facilitated. (Moriyama, K.)

  19. Transport fuel

    DEFF Research Database (Denmark)

    Ronsse, Frederik; Jørgensen, Henning; Schüßler, Ingmar

    2014-01-01

    Worldwide, the use of transport fuel derived from biomass increased four-fold between 2003 and 2012. Mainly based on food resources, these conventional biofuels did not achieve the expected emission savings and contributed to higher prices for food commod - ities, especially maize and oilseeds...

  20. 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)

  1. Application of reactivity method to MTR fuel burn-up measurement

    International Nuclear Information System (INIS)

    Zuniga, A.; Ravnik, M.; Cuya, R.

    2001-01-01

    Fuel element burn-up has been measured for the first time by reactivity method in a MTR reactor. The measurement was performed in RP-10 reactor of Peruvian Institute for Nuclear Energy (IPEN) in Lima. It is a pool type 10MW material testing reactor using standard 20% enriched uranium plate type fuel elements. A fresh element and an element with well defined burn-up were selected as reference elements. Several elements in the core were selected for burn-up measurement. Each of them was replaced in its original position by both reference elements. Change in excess reactivity was measured using control rod calibration curve. The burn-up reactivity worth of fuel elements was plotted as a function of their calculated burnup. Corrected burn-up values of the measured fuel elements were calculated using the fitting function at experimental reactivity for all elements. Good agreement between measured and calculated burn-up values was observed indicating that the reactivity method can be successfully applied also to MTR fuel element burn-up determination.(author)

  2. Thorium fuel cycle management

    International Nuclear Information System (INIS)

    Zajac, R.; Darilek, P.; Breza, J.; Necas, V.

    2010-01-01

    In this presentation author deals with the thorium fuel cycle management. Description of the thorium fuels and thorium fuel cycle benefits and challenges as well as thorium fuel calculations performed by the computer code HELIOS are presented.

  3. Repairing fuel for reinsertion

    International Nuclear Information System (INIS)

    Krukshenk, A.

    1986-01-01

    Eqiupment for nuclear reactor fuel assembly repairing produced by Westinghouse and Brawn Bovery companies is described. Repair of failed fuel assemblies replacement of defect fuel elements gives a noticeable economical effect. Thus if the cost of a new fuel assembly is 450-500 thousand dollars, the replacement of one fuel element in it costs approximately 40-60 thousand dollars. In simple cases repairing includes either removal of failed fuel elements from a fuel assembly and its reinsertion with the rest of fuel elements into the reactor core (reactor refueling), or replacement of unfailed fuel elements from one fuel assembly to a new one (fuel assembly overhaul and reconditioning)

  4. 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

  5. Nuclear power fuel cycle

    International Nuclear Information System (INIS)

    Havelka, S.; Jakesova, L.

    1982-01-01

    Economic problems are discussed of the fuel cycle (cost of the individual parts of the fuel cycle and the share of the fuel cycle in the price of 1 kWh), the technological problems of the fuel cycle (uranium ore mining and processing, uranium isotope enrichment, the manufacture of fuel elements, the building of long-term storage sites for spent fuel, spent fuel reprocessing, liquid and gaseous waste processing), and the ecologic aspects of the fuel cycle. (H.S.)

  6. Trends and Developments for Fast Neutron Reactors and Related Fuel Cycles

    International Nuclear Information System (INIS)

    Carré, Frank

    2013-01-01

    • FR13 – A unique and dedicated framework to share updates on national programs of Fast Reactor developments, projects of new builds and plans for the future: - Near term projects of sodium and lead-alloy Fast Reactors; - Gen-IV visions of sodium-cooled and alternative types of Fast Neutron Reactors (GFR, LFR…). • FR13 – A special emphasis put on Fast Reactor Safety, Sustainability of nuclear fuel cycle and Young Generation perspective. • FR13 – A catalyst for further collaborations and alliances: - To share visions of goals and advisable options for future Fast Reactors and Nuclear Fuel Cycle; - To share cost of R&D and large demonstrations (safety, security, recycling); - To progress towards harmonized international standards; - To integrate national projects into a consistent international roadmap

  7. Fuel trading

    International Nuclear Information System (INIS)

    2015-01-01

    A first part of this report proposes an overview of trends and predictions. After a synthesis on the sector changes and trends, it indicates and comments the most recent predictions for the consumption of refined oil products and for the turnover of the fuel wholesale market, reports the main highlights concerning the sector's life, and gives a dashboard of the sector activity. The second part proposes the annual report on trends and competition. It presents the main operator profiles and fuel categories, the main determining factors of the activity, the evolution of the sector context between 2005 and 2015 (consumptions, prices, temperature evolution). It analyses the evolution of the sector activity and indicators (sales, turnovers, prices, imports). Financial performances of enterprises are presented. The economic structure of the sector is described (evolution of the economic fabric, structural characteristics, French foreign trade). Actors are then presented and ranked in terms of turnover, of added value, and of result

  8. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto.

    1991-01-01

    In a fuel assembly in which spectral shift type moderator guide members are arranged, the moderator guide member has a flow channel resistance member, that provides flow resistance against the moderators, in the upstream of a moderator flowing channel, by which the ratio of removing coolants is set greater at the upstream than downstream. With such a constitution, the void distribution increasing upward in the channel box except for the portion of the moderator guide member is moderated by the increase of the area of the void region that expands downward in the guide member. Accordingly, the axial power distribution is flattened throughout the operation cycle and excess distortion is eliminated to improve the fuel integrity. (T.M.)

  9. Fuel element

    International Nuclear Information System (INIS)

    Hirose, Yasuo.

    1982-01-01

    Purpose: To increase the plenum space in a fuel element used for a liquid metal cooled reactor. Constitution: A fuel pellet is secured at one end with an end plug and at the other with a coil spring in a tubular container. A mechanism for fixing the coil spring composed of a tubular unit is mounted by friction with the inner surface of the tubular container. Accordingly, the recoiling force of the coil spring can be retained by fixing mechanism with a small volume, and since a large amount of plenum space can be obtained, the internal pressure rise in the cladding tube can be suppressed even if large quantities of fission products are discharged. (Kamimura, M.)

  10. Fuel assembly

    International Nuclear Information System (INIS)

    Kawai, Mitsuo.

    1988-01-01

    Purpose: To reduce the corrosion rate and suppress the increase of radioactive corrosion products in reactor water of nuclear fuel assemblies for use in BWR type reactors having spacer springs made of nickel based deposition reinforced type alloys. Constitution: Spacer rings made of nickel based deposition reinforced type alloy are incorporated and used as fuel assemblies after applying treatment of dipping and maintaining at high temperature water followed by heating in steams. Since this can remove the nickel leaching into reactor water at the initial stage, Co-58 as the radioactive corrosion products in the reactor water can be reduced, and the operation at in-service inspection or repairement can be facilitated to improve the working efficiency of the nuclear power plant. The dipping time is desirably more than 10 hours and more desirably more than 30 hours. (Horiuchi, T. )

  11. Fuel assemblies

    International Nuclear Information System (INIS)

    Yoshioka, Ritsuo.

    1983-01-01

    Purpose: To improve the operation performance of a BWR type reactor by improving the distribution of the uranium enrichment and the incorporation amount of burnable poisons in fuel assemblies. Constitution: The average enrichment of uranium 235 is increased in the upper portion as compared with that in the lower portion, while the incorporation amount of burnable poisons is increased in an upper portion as compared with that in the lower portion. The difference in the incorporation amount of the burnable poisons between the upper and lower portions is attained by charging two kinds of fuel rods; the ones incorporated with the burnable poisons over the entire length and the others incorporated with the burnable poisons only in the upper portions. (Seki, T.)

  12. Fuel assembly

    International Nuclear Information System (INIS)

    Hirukawa, Koji; Sakurada, Koichi.

    1992-01-01

    In a fuel assembly for a BWR type reactor, water rods or water crosses are disposed between fuel rods, and a value with a spring is disposed at the top of the coolant flow channel thereof, which opens a discharge port when pressure is increased to greater than a predetermined value. Further, a control element for the amount of coolant flow rate is inserted retractable to a control element guide tube formed at the lower portion of the water rod or the water cross. When the amount of control elements inserted to the control element guide tube is small and the inflown coolant flow rate is great, the void coefficient at the inside of the water rod is less than 5%. On the other hand, when the control elements are inserted, the flow resistance is increased, so that the void coefficient in the water rod is greater than 80%. When the pressure in the water rod is increased, the valve with the spring is raised to escape water or steams. Then, since the variation range of the change of the void coefficient can be controlled reliably by the amount of the control elements inserted, and nuclear fuel materials can be utilized effectively. (N.H.)

  13. Solid TRU fuels and fuel cycle technology

    International Nuclear Information System (INIS)

    Ogawa, Toru; Suzuki, Yasufumi

    1997-01-01

    Alloys and nitrides are candidate solid fuels for transmutation. However, the nitride fuels are preferred to the alloys because they have more favorable thermal properties which allows to apply a cold-fuel concept. The nitride fuel cycle technology is briefly presented

  14. Used fuel packing plant for CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Menzies, I.; Thayer, B.; Bains, N., E-mail: imenzies@atsautomation.com [ATS Automation, Cambridge, ON (Canada); Murchison, A., E-mail: amurchison@nwmo.ca [NWMO, Toronto, ON (Canada)

    2015-07-01

    Large forgings have been selected to containerize Light Water Reactor used nuclear fuel. CANDU fuel, which is significantly smaller in size, allows novel approaches for containerization. For example, by utilizing commercially available extruded ASME pipe a conceptual design of a Used Fuel Packing Plant for containerization of used CANDU fuel in a long lived metallic container has been developed. The design adopts a modular approach with multiple independent work cells to transfer and containerize the used fuel. Based on current technologies and concepts from proven industrial systems, the Used Fuel Packing Plant can assemble twelve used fuel containers per day considering conservative levels of process availability. (author)

  15. What the difference to use LEU and HEU fuel elements separately or together in a research reactor

    International Nuclear Information System (INIS)

    Kaya, S.; Uestuen, G.

    2005-01-01

    Concerning of nuclear material safety, most of the research reactors are advised to shift from HEU (high enriched-%93 U-235) to LEU (low enriched-%20 U-235) fuel elements. When LEU and HEU fuel elements are to be used together in a research reactor, some design and safety problems are encountered. According to use of the reactor, some research reactors such as MTR type may not show any considerable difference for HEU or LEU fuel elements, but the efficiency of radioisotope production generated by thermal neutron interaction may decrease about twenty-thirty percent when LEU fuel elements are used. Here, fine mesh-sized 3D neutronic analysis of TR-2 research reactor is presented to indicate the arising problem when LEU end HEU fuel elements are used together in a research reactor. Partial thermohydraulic analysis of the reactor is also given to show the betterness of the LEU fuel element design. However, there might be some points that should be noticed for safer operation of plate type fuelled research reactors. (author)

  16. Reclamation and reuse of LEU silicide fuel from manufacturing scrap

    International Nuclear Information System (INIS)

    Gale, G.R.; Pace, B.W.; Evans, R.S.

    2004-01-01

    In order to provide an understanding of the organization which is the sole supplier of United States plate type research and test reactor fuel and LEU core conversions, a brief description of the structure and history is presented. Babcock and Wilcox (B and W) is a part of McDermott International, Inc. which is a large diversified corporation employing over 20,000 people primarily in engineering and construction for the off-shore oil and power generation industries throughout the world. B and W provides many energy related products requiring precision machining and high quality systems. This is accomplished by using state-of-the-art equipment, technology and highly skilled people. The RTRFE group within B and W has the ability to produce various complexly shaped fuel elements with a wide variety of fuels and enrichments. B and W RTRFE has fabricated over 200,000 plates since 1981 and gained the diversified experience necessary to satisfy many customer requirements. This accomplishment was possible with the support of McDermott International and all of its resources. B and W has always had a commitment to high quality and integrity. This is apparent by the success and longevity (125 years) of the company. A lower cost to convert cores to LEU provides direct support to RERTR and demonstrates Babcock and Wilcox's commitment to the program. As a supporter of RERTR reactor conversion from HEU to LEU, B and W has contributed a significant amount of R and D money to improve the silicide fuel process which ultimately lowers the LEU core costs. In the most recent R and D project, B and W is constructing a LEU silicide reclamation facility to re-use the unirradiated fuel scrap generated from the production process. Remanufacturing use of this fuel completes the fuel cycle and provides a contribution to LEU cores by reducing scrap inventory and handling costs, lowering initial purchase of fuel due to increasing the process yields, and lowering the replacement costs. This

  17. Nuclear fuel preheating system

    International Nuclear Information System (INIS)

    Andrea, C.

    1975-01-01

    A nuclear reactor new fuel handling system which conveys new fuel from a fuel preparation room into the reactor containment boundary is described. The handling system is provided with a fuel preheating station which is adaptd to heat the new fuel to reactor refueling temperatures in such a way that the fuel is heated from the top down so that fuel element cladding failure due to thermal expansions is avoided. (U.S.)

  18. Fuel element loading system

    International Nuclear Information System (INIS)

    Arya, S.P; s.

    1978-01-01

    A nuclear fuel element loading system is described which conveys a plurality of fuel rods to longitudinal passages in fuel elements. Conveyor means successively position the fuel rods above the longitudinal passages in axial alignment therewith and adapter means guide the fuel rods from the conveyor means into the longitudinal passages. The fuel elements are vibrated to cause the fuel rods to fall into the longitudinal passages through the adapter means

  19. Artificial fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hamon, L L.W.

    1918-08-20

    Lignite, peat, sud, leaf-mold, or shale, or two or more of these raw carbonaceous materials are mixed with cellulose material, such as sawdust, silica, alkali, and tar or pitch, or residues from tar or pitch, or residues from the distillation of oils, and the mixture is molded into blocks. Other carbonaceous materials, such as graphite, anthracite, or coal-dust, coke, breeze, or culm, and mineral substances, such as iron and manganese ores, may be added. A smokeless fuel can be obtained by coking the blocks in the usual way in retorts.

  20. Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

    Energy Technology Data Exchange (ETDEWEB)

    Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

    1985-07-01

    This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

  1. Experiments of JRR-4 low-enriched-uranium-silicied fuel core

    International Nuclear Information System (INIS)

    Hirane, Nobuhiko; Ishikuro, Yasuhiro; Nagadomi, Hideki; Yokoo, Kenji; Horiguchi, Hironori; Nemoto, Takumi; Yamamoto, Kazuyoshi; Yagi, Masahiro; Arai, Nobuyoshi; Watanabe, Shukichi; Kashima, Yoichi

    2006-03-01

    JRR-4, a light-water-moderated and cooled, swimming pool type research reactor using high-enriched uranium plate-type fuels had been operated from 1965 to 1996. In order to convert to low-enriched-uranium-silicied fuels, modification work had been carried out for 2 years, from 1996 to 1998. After the modification, start-up experiments were carried out to obtain characteristics of the low-enriched-uranium-silicied fuel core. The measured excess reactivity, reactor shutdown margin and the maximum reactivity addition rate satisfied the nuclear limitation of the safety report for licensing. It was confirmed that conversion to low-enriched-uranium-silicied fuels was carried out properly. Besides, the necessary data for reactor operation were obtained, such as nuclear, thermal hydraulic and reactor control characteristics. This report describes the results of start-up experiments and burnup experiments. The first criticality of low-enriched-uranium-silicied core was achieved on 14th July 1998, and the operation for joint-use has been carried out since 6th October 1998. (author)

  2. Nuclear Fuel elements

    International Nuclear Information System (INIS)

    Hirakawa, Hiromasa.

    1979-01-01

    Purpose: To reduce the stress gradient resulted in the fuel can in fuel rods adapted to control the axial power distribution by the combination of fuel pellets having different linear power densities. Constitution: In a fuel rod comprising a first fuel pellet of a relatively low linear power density and a second fuel pellet of a relatively high linear power density, the second fuel pellet is cut at its both end faces by an amount corresponding to the heat expansion of the pellet due to the difference in the linear power density to the adjacent first fuel pellet. Thus, the second fuel pellet takes a smaller space than the first fuel pellet in the fuel can. This can reduce the stress produced in the portion of the fuel can corresponding to the boundary between the adjacent fuel pellets. (Kawakami, Y.)

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Hiraiwa, Koji; Ueda, Makoto

    1989-01-01

    In a fuel assembly used for a light water cooled reactor such as a BWR type reactor, a water rod is divided axially into an upper outer tube and a lower outer tube by means of a plug disposed from the lower end of a water rod to a position 1/4 - 1/2 of the entire length for the water rod. Inlet apertures and exit apertures for moderators are respectively perforated for the divided outer tube and upper and lower portions. Further, an upper inner tube with less neutron irradiation growing amount than the outer tube is perforated on the plug in the outer tube, while a lower inner tube with greater neutron irradiation growing amount than the outer tube is suspended from the lower surface of the plug in the outer tube. Then, the opening area for the exit apertures disposed to the upper outer tube and the lower outer tube is controlled depending on the difference of the neutron irradiation growing amount between the upper inner tube and the upper outer tube, and the difference of the neutron irradiation growing amount between the lower inner tube and the lower outer tube. This enables effective spectral shift operation and improve the fuel economy. (T.M.)

  4. Fuel Burn Estimation Model

    Science.gov (United States)

    Chatterji, Gano

    2011-01-01

    Conclusions: Validated the fuel estimation procedure using flight test data. A good fuel model can be created if weight and fuel data are available. Error in assumed takeoff weight results in similar amount of error in the fuel estimate. Fuel estimation error bounds can be determined.

  5. Constant strength fuel-fuel cell

    International Nuclear Information System (INIS)

    Vaseen, V.A.

    1980-01-01

    A fuel cell is an electrochemical apparatus composed of both a nonconsumable anode and cathode; and electrolyte, fuel oxidant and controls. This invention guarantees the constant transfer of hydrogen atoms and their respective electrons, thus a constant flow of power by submergence of the negative electrode in a constant strength hydrogen furnishing fuel; when said fuel is an aqueous absorbed hydrocarbon, such as and similar to ethanol or methnol. The objective is accomplished by recirculation of the liquid fuel, as depleted in the cell through specific type membranes which pass water molecules and reject the fuel molecules; thus concentrating them for recycle use

  6. Estimated GFR Decline as a Surrogate End Point for Kidney Failure : A Post Hoc Analysis From the Reduction of End Points in Non-Insulin-Dependent Diabetes With the Angiotensin II Antagonist Losartan (RENAAL) Study and Irbesartan Diabetic Nephropathy Trial (IDNT)

    NARCIS (Netherlands)

    Lambers Heerspink, Hiddo; Weldegiorgis, Misghina; Inker, Lesley A.; Gansevoort, Ron; Parving, Hans-Henrik; Dwyer, Jamie P.; Mondal, Hasi; Coresh, Josef; Greene, Tom; Levey, Andrew S.; de Zeeuw, Dick

    Background: A doubling of serum creatinine value, corresponding to a 57% decline in estimated glomerular filtration rate (eGFR), is used frequently as a component of a composite kidney end point in clinical trials in type 2 diabetes. The aim of this study was to determine whether alternative end

  7. KMRR fuel design

    International Nuclear Information System (INIS)

    Son, D.S.; Sim, B.S.; Kim, T.R.; Hwang, W.; Kim, B.G.; Ku, Y.H.; Lee, C.B.; Lim, I.C.

    1992-06-01

    KMRR fuel rod design criteria on fuel swelling, blistering and oxide spallation have been reexamined. Fuel centerline temperature limit of 250deg C in normal operation condition and fuel swelling limit of 12 % at the end of life have been proposed to prevent fuel failure due to excessive fuel swelling. Fuel temperature limit of 485deg C has been proposed to exclude the possibility of fuel failures during transients or under accident condition. Further analyses are needed to decide the fuel cladding temperature limit to preclude the oxide spallation. Design changes in fuel assembly structure and their effects on related systems have been reviewed from a structural integrity viewpoint. The remained works in fuel mechanical design area have been identified and further efforts of fuel design group will be focused on these aspects. (Author)

  8. Fuel Property Blend Model

    Energy Technology Data Exchange (ETDEWEB)

    Pitz, William J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mehl, Marco [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wagnon, Scott J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Zhang, Kuiwen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kukkadapu, Goutham [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Westbrook, Charles K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-01-12

    The object of this project is to develop chemical models and associated correlations to predict the blending behavior of bio-derived fuels when mixed with conventional fuels like gasoline and diesel fuels.

  9. Logistic Fuel Processor Development

    National Research Council Canada - National Science Library

    Salavani, Reza

    2004-01-01

    The Air Base Technologies Division of the Air Force Research Laboratory has developed a logistic fuel processor that removes the sulfur content of the fuel and in the process converts logistic fuel...

  10. Fuel pellet loading apparatus

    International Nuclear Information System (INIS)

    1980-01-01

    Apparatus is described for loading a predetermined amount of nuclear fuel pellets into nuclear fuel elements and particularly for the automatic loading of fuel pellets from within a sealed compartment. (author)

  11. Fuel assembly

    International Nuclear Information System (INIS)

    Wataumi, Kazutoshi; Tajiri, Hiroshi.

    1992-01-01

    In a fuel assembly of a BWR type reactor, a pellet to be loaded comprises an external layer of fissile materials containing burnable poisons and an internal layer of fissile materials not containing burnable poison. For example, there is provided a dual type pellet comprising an external layer made of UO 2 incorporated with Gd 2 O 3 at a predetermined concentration as the burnable poisons and an internal layer made of UO 2 not containing Gd 2 O 3 . The amount of the burnable poisons required for predetermined places is controlled by the thickness of the ring of the external layer. This can dissipate an unnecessary poisoning effect at the final stage of the combustion cycle. Further, since only one or a few kinds of powder mixture of the burnable poisons and the fissile materials is necessary, production and product control can be facilitated. (I.N.)

  12. Fuel storage

    International Nuclear Information System (INIS)

    Palacios, C.; Alvarez-Miranda, A.

    2009-01-01

    ENSA is a well known manufacturer of multi-system primary components for the nuclear industry and is totally prepared to satisfy future market requirements in this industry. At the same time that ENSA has been gaining a reputation world wider for the supply of primary components, has been strengthening its commitment and experience in supplying spent fuel components, either pool racks or storage and transportation casks, and offers not only fabrication but also design capabilities for its products. ENSA has supplied Spent Fuel Pool Racks, in spain, Finland, Taiwan, Korea, China, and currently it is in the process of licensing its own rack design in the United States of America for the ESBWR along with Ge-Hitachi. ENSA has supplied racks for 20 pools and 22 different reactors and it has also manufactured racks under all available technologies and developed a design known as Interlock Cell Matrix whose main features are outlined in this article. Another ENSA achievement in rack technology is the use of remote control for re-racking activities instead of using divers, which improves the ALARA requirements. Regarding casks for storage and transportation, ENSA also has al leading worldwide position, with exports prevailing over the Spanish market where ENSA has supplied 16 storage and transportation casks to the Spanish nuclear power Trillo. In some cases, ENSA acts as subcontractor for other clients. Foreign markets are still a major challenge for ENSA. ENSA-is well known for its manufacturing capabilities in the nuclear industry, but has been always involved in design activities through its engineering division, which carries out different tasks: components Design; Tooling Design; Engineering and Documentation; Project Engineering; Calculations, Design and Development Engineering. (Author)

  13. Nuclear fuel replacement device

    International Nuclear Information System (INIS)

    Ritz, W.C.; Robey, R.M.; Wett, J.F.

    1984-01-01

    A fuel handling arrangement for a liquid metal cooled nuclear reactor having a single rotating plug eccentric to the fuel core and a fuel handling machine radially movable along a slot in the plug with a transfer station disposed outside the fuel core but covered by the eccentric plug and within range of movement of said fuel handling machine to permit transfer of fuel assemblies between the core and the transfer station. (author)

  14. CANDU fuel performance

    International Nuclear Information System (INIS)

    Ivanoff, N.V.; Bazeley, E.G.; Hastings, I.J.

    1982-01-01

    CANDU fuel has operated successfully in Ontario Hydro's power reactors since 1962. In the 19 years of experience, about 99.9% of all fuel bundles have performed as designed. Most defects occurred before 1979 and subsequent changes in fuel design, fuel management, reactor control, and manufacturing quality control have reduced the current defect rate to near zero. Loss of power production due to defective fuel has been negligible. The outstanding performance continues while maintaining a low unit energy cost for fuel

  15. Estimated GFR and Subsequent Higher Left Ventricular Mass in Young and Middle-Aged Adults With Normal Kidney Function: The Coronary Artery Risk Development in Young Adults (CARDIA) Study.

    Science.gov (United States)

    Bansal, Nisha; Lin, Feng; Vittinghoff, Eric; Peralta, Carmen; Lima, Joao; Kramer, Holly; Shlipak, Michael; Bibbins-Domingo, Kirsten

    2016-02-01

    Left ventricular hypertrophy is common and is associated with cardiovascular events and death among patients with known chronic kidney disease. However, the link between reduced glomerular filtration rate (GFR) and left ventricular mass index (LVMI) remains poorly explored among young and middle-aged adults with preserved kidney function. In this study, we examined the association of cystatin C-based estimated GFR (eGFRcys) and rapid decline in eGFR with subsequent LVMI. Observational study. We included 2,410 participants from the Coronary Artery Risk Development in Young Adults (CARDIA) cohort with eGFRcys > 60mL/min/1.73m(2) at year 15 and who had an echocardiogram obtained at year 25. eGFRcys at year 15 and rapid decline in eGFRcys (defined as >3% per year over 5 years from years 15 to 20). LVMI measured at year 25. We adjusted for age, sex, race, diabetes, body mass index, low- and high-density lipoprotein cholesterol levels, cumulative systolic blood pressure, and albuminuria. Mean age was 40±4 (SD) years, 58% were women, and 43% were black. After 10 years of follow-up, mean LVMI was 39.6±13.4g/m(2.7). Compared with eGFRcys > 90mL/min/1.73m(2) (n = 2,228), eGFRcys of 60 to 75mL/min/1.73m(2) (n = 29) was associated with 5.63 (95% CI, 0.90-10.36) g/m(2.7) greater LVMI (P = 0.02), but there was no association of eGFRcys of 76 to 90mL/min/1.73m(2) (n = 153) with LVMI after adjustment for confounders. Rapid decline in eGFRcys was associated with higher LVMI compared with participants without a rapid eGFRcys decline (β coefficient, 1.48; 95% CI, 0.11-2.83; P = 0.03) after adjustment for confounders. There were a limited number of participants with eGFRcys of 60 to 90mL/min/1.73m(2). Among young and middle-aged adults with preserved kidney function, eGFRcys of 60 to 75mL/min/1.73m(2) and rapid decline in eGFRcys were significantly associated with subsequently higher LVMI. Further studies are needed to understand the mechanisms that contribute to elevated

  16. Fuels Combustion Research: Supercritical Fuel Pyrolysis

    National Research Council Canada - National Science Library

    Glassman, Irvin

    2001-01-01

    .... The focus during the subject period was directed to understanding the pyrolysis and combustion of endothermic fuels under subcritical conditions and the pyrolysis of these fuels under supercritical conditions...

  17. Fuels Combustion Research: Supercritical Fuel Pyrolysis

    National Research Council Canada - National Science Library

    Glassman, Irvin

    2000-01-01

    .... The focus during the subject period was directed to understanding the pyrolysis and combustion of endothermic fuels under subcritical conditions and the pyrolysis of these fuels under supercritical conditions...

  18. Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

    Science.gov (United States)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-07-01

    Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (Dpower is 300 MWth. First calculation, we calculate survey parameter for UN-PuN fuel with fissile contain from Plutonium waste LWR for each geometry. The minimum power density is around 72 Watt/cc, and maximum power density 114 Watt/cc. After we calculate with various geometry core, when we use the balance geometry, the k-eff value flattest and more stable than the others.

  19. 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.

  20. Influence of single-phase heat transfer correlations on safety analysis of research reactors with narrow rectangular fuel channels

    International Nuclear Information System (INIS)

    Rawashdeh, A.; Altamimi, R.; Lee, B.; Chung, Y. J.; Park, S.

    2013-01-01

    The influence of different single-phase heat transfer correlations on the fuel temperature and minimum critical heat flux ratio (MCHFR) during a typical accident of a 5 MW research reactor is investigated. A reactor uses plate type fuel, of which the cooling channels have a narrow rectangular shape. RELAP5/MOD3.3 tends to over-predict the Nusselt number (Nu) at a low Reynolds number (Re) region, and therefore the correlation set is modified to properly describe the thermal behavior at that region. To demonstrate the effect of Nu at a low-Re region on an accident analysis, a two-pump failure accident was chosen as a sample problem. In the accident, the downward core flow decreases by a pump coast-down, and then reverses upward by natural convection. During the pump coast-down and flow reversal, the flow undergoes a laminar flow regime which has a different Nu with respect to the correlation sets. Compared to the results by the original RELAP5/MOD3.3, the modified correlation set predicts the fuel temperature to be a little higher than the original value, and the MCHFR to be a little lower than the original value. Although the modified correlation set predicts the fuel temperature and the MCHFR to be less conservative than those calculated from the original correlation of RELAP5/MOD3.3, the maximum fuel temperature and the MCHFR still satisfy the safety acceptance criteria

  1. GSPEL - Fuel Cell Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — The Fuel Cell Lab (FCL)Established to investigate, integrate, testand verifyperformance and technology readiness offuel cell systems and fuel reformers for use with...

  2. Fuel performance experience

    International Nuclear Information System (INIS)

    Sofer, G.A.

    1986-01-01

    The history of LWR fuel supply has been characterized by a wide range of design developments and fuel cycle cost improvements. Exxon Nuclear Company, Inc. has pursued an aggressive fuel research and development program aimed at improved fuel performance. Exxon Nuclear has introduced many design innovations which have improved fuel cycle economics and operating flexibility while fuel failures remain at very low levels. The removable upper tie plate feature of Exxon Nuclear assemblies has helped accelerate this development, enabling repeated inspections during successive plant outages. Also, this design feature has made it possible to repair damaged fuel assemblies during refueling outages, thereby minimizing the economic impact of fuel failure from all causes

  3. Catalytic Fuel Conversion Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This facility enables unique catalysis research related to power and energy applications using military jet fuels and alternative fuels. It is equipped with research...

  4. HTGR fuel reprocessing technology

    International Nuclear Information System (INIS)

    Brooks, L.H.; Heath, C.A.; Shefcik, J.J.

    1976-01-01

    The following aspects of HTGR reprocessing technology are discussed: characteristics of HTGR fuels, criteria for a fuel reprocessing flowsheet; selection of a reference reprocessing flowsheet, and waste treatment

  5. Nuclear fuel production

    International Nuclear Information System (INIS)

    Randol, A.G.

    1985-01-01

    The production of new fuel for a power plant reactor and its disposition following discharge from the power plant is usually referred to as the ''nuclear fuel cycle.'' The processing of fuel is cyclic in nature since sometime during a power plant's operation old or ''depleted'' fuel must be removed and new fuel inserted. For light water reactors this step typically occurs once every 12-18 months. Since the time required for mining of the raw ore to recovery of reusable fuel materials from discharged materials can span up to 8 years, the management of fuel to assure continuous power plant operation requires simultaneous handling of various aspects of several fuel cycles, for example, material is being mined for fuel to be inserted in a power plant 2 years into the future at the same time fuel is being reprocessed from a discharge 5 years prior. Important aspects of each step in the fuel production process are discussed

  6. Fuel manufacturing and utilization

    International Nuclear Information System (INIS)

    2005-01-01

    The efficient utilisation of nuclear fuel requires manufacturing facilities capable of making advanced fuel types, with appropriate quality control. Once made, the use of such fuels requires a proper understanding of their behaviour in the reactor environment, so that safe operation for the design life can be achieved. The International Atomic Energy Agency supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle. It provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection. The IAEA supports the development fuel modelling expertise in Member States, covering both normal operation and postulated and severe accident conditions. It provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation. The IAEA supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, it provides information and support research into the basic properties of fuel materials, including UO 2 , MOX and zirconium alloys. It further offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology

  7. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    Durand, J.P.; Fanjas, Y.

    1993-01-01

    When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have led to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

  8. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    Durand, J.P.; Fanjas, Y.

    1994-01-01

    When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have lead to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

  9. GEN IV: Carbide Fuel Elaboration for the 'Futurix Concepts' experiment

    Energy Technology Data Exchange (ETDEWEB)

    Vaudez, Stephane; Riglet-Martial, Chantal; Paret, Laurent; Abonneau, Eric [Commissariat a l' Energie Atomique (C.E.A.), Direction de l' Energie Nucleaire, Centre d' Etudes de Cadarache, 13108 Saint Paul lez Durance Cedex (France)

    2008-07-01

    In order to collect information on the behaviour of the future GFR (Gas Fast Reactor) fuel under fast neutron irradiation, an experimental irradiation program, called 'Futurix-concepts' has been launched at the CEA. The considered concept is a composite material made of a fissile fuel embedded in an inert ceramic matrix. Fissile fuel pellets are made of UPuN or UPuC while ceramics are SiC for the carbide fuel and TiN for the nitride fuel. This paper focuses on the description of the carbide composite fabrication. The UPuC pellets are manufactured using a metallurgical powder process. Fabrication and handling of the fuels are carried out in glove boxes under a nitrogen atmosphere. Carbide fuel is synthesized by carbo-thermic reduction under vacuum of a mixture of actinide oxide and graphitic carbon up to 1550 deg. C. After ball milling, the UPuC powder is pressed to create hexagonal or spherical compacts. They are then sintered up to 1750 deg. C in order to obtain a density of 85 % of the theoretical one. The sintered pellets are inserted into an inert and tight capsule of SiC. In order to control the gap between the fuel and the matrix precisely, the pellets are abraded. The inert matrix is then filled with the pellets and the whole system is sealed by a BRASiC{sup R} process at high temperature under a helium atmosphere. Fabrication of the sample to be irradiated was done in 2006 and the irradiation began in May 2007 in the Phenix reactor. This presentation will detail and discuss the results obtained during this fabrication phase. (authors)

  10. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Coobs, J.H.

    1976-08-01

    The status of fuel and fuel cycle technology for high-temperature gas-cooled reactors (HTGRs) is reviewed. The all-ceramic core of the HTGRs permits high temperatures compared with other reactors. Core outlet temperatures of 740 0 C are now available for the steam cycle. For advanced HTGRs such as are required for direct-cycle power generation and for high-temperature process heat, coolant temperatures as high as 1000 0 C may be expected. The paper discusses the variations of HTGR fuel designs that meet the performance requirements and the requirements of the isotopes to be used in the fuel cycle. Also discussed are the fuel cycle possibilities, which include the low-enrichment cycle, the Th- 233 U cycle, and plutonium utilization in either cycle. The status of fuel and fuel cycle development is summarized

  11. Estimate of fuel burnup spatial a multipurpose reactor in computer simulation

    International Nuclear Information System (INIS)

    Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes

    2015-01-01

    In previous research, which aimed, through computer simulation, estimate the spatial fuel burnup for the research reactor benchmark, material test research - International Atomic Energy Agency (MTR/IAEA), it was found that the use of the code in FORTRAN language, based on the diffusion theory of neutrons and WIMSD-5B, which makes cell calculation, bespoke be valid to estimate the spatial burnup other nuclear research reactors. That said, this paper aims to present the results of computer simulation to estimate the space fuel burnup of a typical multipurpose reactor, plate type and dispersion. the results were considered satisfactory, being in line with those presented in the literature. for future work is suggested simulations with other core configurations. are also suggested comparisons of WIMSD-5B results with programs often employed in burnup calculations and also test different methods of interpolation values obtained by FORTRAN. Another proposal is to estimate the burning fuel, taking into account the thermohydraulics parameters and the appearance of xenon. (author)

  12. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Homan, F.J.; Balthesen, E.; Turner, R.F.

    1977-01-01

    Significant advances have occurred in the development of HTGR fuel and fuel cycle. These accomplishments permit a wide choice of fuel designs, reactor concepts, and fuel cycles. Fuels capable of providing helium outlet temperatures of 750 0 C are available, and fuels capable of 1000 0 C outlet temperatures may be expected from extension of present technology. Fuels have been developed for two basic HTGR designs, one using a spherical (pebble bed) element and the other a prismatic element. Within each concept a number of variations of geometry, fuel composition, and structural materials are permitted. Potential fuel cycles include both low-enriched and high-enriched Th- 235 U, recycle Th- 233 U, and Th-Pu or U-Pu cycles. This flexibility offered by the HTGR is of great practical benefit considering the rapidly changing economics of power production. The inflation of ore prices has increased optimum conversion ratios, and increased the necessity of fuel recycle at an early date. Fuel element makeup is very similar for prismatic and spherical designs. Both use spherical fissile and fertile particles coated with combinations of pyrolytic carbon and silicon carbide. Both use carbonaceous binder materials, and graphite as the structural material. Weak-acid resin (WAR) UO 2 -UC 2 fissile fuels and sol-gel-derived ThO 2 fertile fuels have been selected for the Th- 233 U cycle in the prismatic design. Sol-gel-derived UO 2 UC 2 is the reference fissile fuel for the low-enriched pebble bed design. Both the United States and Federal Republic of Germany are developing technology for fuel cycle operations including fabrication, reprocessing, refabrication, and waste handling. Feasibility of basic processes has been established and designs developed for full-scale equipment. Fuel and fuel cycle technology provide the basis for a broad range of applications of the HTGR. Extension of the fuels to higher operating temperatures and development and commercial demonstration of fuel

  13. 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

  14. Integrated fuel processor development

    International Nuclear Information System (INIS)

    Ahmed, S.; Pereira, C.; Lee, S. H. D.; Krumpelt, M.

    2001-01-01

    The Department of Energy's Office of Advanced Automotive Technologies has been supporting the development of fuel-flexible fuel processors at Argonne National Laboratory. These fuel processors will enable fuel cell vehicles to operate on fuels available through the existing infrastructure. The constraints of on-board space and weight require that these fuel processors be designed to be compact and lightweight, while meeting the performance targets for efficiency and gas quality needed for the fuel cell. This paper discusses the performance of a prototype fuel processor that has been designed and fabricated to operate with liquid fuels, such as gasoline, ethanol, methanol, etc. Rated for a capacity of 10 kWe (one-fifth of that needed for a car), the prototype fuel processor integrates the unit operations (vaporization, heat exchange, etc.) and processes (reforming, water-gas shift, preferential oxidation reactions, etc.) necessary to produce the hydrogen-rich gas (reformate) that will fuel the polymer electrolyte fuel cell stacks. The fuel processor work is being complemented by analytical and fundamental research. With the ultimate objective of meeting on-board fuel processor goals, these studies include: modeling fuel cell systems to identify design and operating features; evaluating alternative fuel processing options; and developing appropriate catalysts and materials. Issues and outstanding challenges that need to be overcome in order to develop practical, on-board devices are discussed

  15. Methanol Fuel Cell

    Science.gov (United States)

    Voecks, G. E.

    1985-01-01

    In proposed fuel-cell system, methanol converted to hydrogen in two places. External fuel processor converts only part of methanol. Remaining methanol converted in fuel cell itself, in reaction at anode. As result, size of fuel processor reduced, system efficiency increased, and cost lowered.

  16. Reactor fueling system

    International Nuclear Information System (INIS)

    Hattori, Noriaki; Hirano, Haruyoshi.

    1983-01-01

    Purpose: To optimally position a fuel catcher by mounting a television camera to a fuel catching portion and judging video images by the use of a computer or the like. Constitution: A television camera is mounted to the lower end of a fuel catching mechanism for handling nuclear fuels and a fuel assembly disposed within a reactor core or a fuel storage pool is observed directly from above to judge the position for the fuel assembly by means of video signals. Then, the relative deviation between the actual position of the fuel catcher and that set in a memory device is determined and the positional correction is carried out automatically so as to reduce the determined deviation to zero. This enables to catch the fuel assembly without failure and improves the efficiency for the fuel exchange operation. (Moriyama, K.)

  17. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Nakai, Keiichi

    1983-01-01

    Purpose: To decrease the tensile stresses resulted in a fuel can as well as prevent decladding of fuel pellets into the bore holes by decreasing the inner pressure within the nuclear fuel element. Constitution: A fuel can is filled with hollow fuel pellets, inserted with a spring for retaining the hollow fuel pellets with an appropriate force and, thereafter, closely sealed at the both ends with end plugs. A cylindrical body is disposed into the bore holes of the hollow fuel pellets. Since initial sealing gases and/or gaseous nuclear fission products can thus be excluded from the bore holes where the temperature is at the highest level, the inner pressure of the nuclear fuel element can be reduced to decrease the tensile strength resulted to the fuel can. Furthermore, decladding of fuel pellets into the bore holes can be prevented. (Moriyama, K.)

  18. Failed fuel rod detector

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Katsuya; Matsuda, Yasuhiko

    1984-05-02

    The purpose of the project is to enable failed fuel rod detection simply with no requirement for dismantling the fuel assembly. A gamma-ray detection section is arranged so as to attend on the optional fuel rods in the fuel assembly. The fuel assembly is adapted such that a gamma-ray shielding plate is detachably inserted into optional gaps of the fuel rods or, alternatively, the fuel assembly can detachably be inserted to the gamma-ray shielding plate. In this way, amount of gaseous fission products accumulated in all of the plenum portions in the fuel rods as the object of the measurement can be determined without dismantling the fuel assembly. Accordingly, by comparing the amounts of the gaseous fission products, the failed fuel rod can be detected.

  19. 77 FR 13009 - Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel...

    Science.gov (United States)

    2012-03-05

    ... Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel Pathways Under the Renewable Fuel Standard Program AGENCY: Environmental Protection Agency (EPA). ACTION: Withdrawal... Renewable Fuel Standard program regulations. Because EPA received adverse comment, we are withdrawing the...

  20. Materials for fuel cells

    OpenAIRE

    Haile, Sossina M

    2003-01-01

    Because of their potential to reduce the environmental impact and geopolitical consequences of the use of fossil fuels, fuel cells have emerged as tantalizing alternatives to combustion engines. Like a combustion engine, a fuel cell uses some sort of chemical fuel as its energy source but, like a battery, the chemical energy is directly converted to electrical energy, without an often messy and relatively inefficient combustion step. In addition to high efficiency and low emissions, fuel cell...

  1. Advanced fuels safety comparisons

    International Nuclear Information System (INIS)

    Grolmes, M.A.

    1977-01-01

    The safety considerations of advanced fuels are described relative to the present understanding of the safety of oxide fueled Liquid Metal Fast Breeder Reactors (LMFBR). Safety considerations important for the successful implementation of advanced fueled reactors must early on focus on the accident energetics issues of fuel coolant interactions and recriticality associated with core disruptive accidents. It is in these areas where the thermal physical property differences of the advanced fuel have the greatest significance

  2. Nuclear fuel storage facility

    International Nuclear Information System (INIS)

    Matsumoto, Takashi; Isaka, Shinji.

    1987-01-01

    Purpose: To increase the spent fuel storage capacity and reduce the installation cost in a nuclear fuel storage facility. Constitution: Fuels handled in the nuclear fuel storage device of the present invention include the following four types: (1) fresh fuels, (2) 100 % reactor core charged fuels, (3) spent fuels just after taking out and (4) fuels after a certain period (for example one half-year) from taking out of the reactor. Reactivity is high for the fuels (1), and some of fuels (2), while low in the fuels (3) (4), Source intensity is strong for the fuels (3) and some of the fuels (2), while it is low for the fuels (1) and (4). Taking notice of the fact that the reactivity, radioactive source intensity and generated after heat are different in the respective fuels, the size of the pool and the storage capacity are increased by the divided storage control. While on the other hand, since the division is made in one identical pool, the control method becomes important, and the working range is restricted by means of a template, interlock, etc., the operation mode of the handling machine is divided into four, etc. for preventing errors. (Kamimura, M.)

  3. Fuel pattern recognition device

    International Nuclear Information System (INIS)

    Sato, Tomomi.

    1995-01-01

    The device of the present invention monitors normal fuel exchange upon fuel exchanging operation carried out in a reactor of a nuclear power plant. Namely, a fuel exchanger is movably disposed to the upper portion of the reactor and exchanges fuels. An exclusive computer receives operation signals of the fuel exchanger during operation as inputs, and outputs reactor core fuel pattern information signals to a fuel arrangement diagnosis device. An underwater television camera outputs image signals of a fuel pattern in the reactor core to an image processing device. If there is any change in the image signals for the fuel pattern as a result of the fuel exchange operation of the fuel exchanger, the image processing device outputs the change as image signals to the fuel pattern diagnosis device. The fuel pattern diagnosis device compares the pattern information signals from the exclusive computer with the image signals from the image processing device, to diagnose the result of the fuel exchange operation performed by the fuel exchanger and inform the diagnosis by means of an image display. (I.S.)

  4. BWR fuel performance

    International Nuclear Information System (INIS)

    Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

    1979-01-01

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

  5. Dual Tank Fuel System

    Science.gov (United States)

    Wagner, Richard William; Burkhard, James Frank; Dauer, Kenneth John

    1999-11-16

    A dual tank fuel system has primary and secondary fuel tanks, with the primary tank including a filler pipe to receive fuel and a discharge line to deliver fuel to an engine, and with a balance pipe interconnecting the primary tank and the secondary tank. The balance pipe opens close to the bottom of each tank to direct fuel from the primary tank to the secondary tank as the primary tank is filled, and to direct fuel from the secondary tank to the primary tank as fuel is discharged from the primary tank through the discharge line. A vent line has branches connected to each tank to direct fuel vapor from the tanks as the tanks are filled, and to admit air to the tanks as fuel is delivered to the engine.

  6. HTGR Fuel performance basis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-05-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents

  7. Elongated fuel road

    International Nuclear Information System (INIS)

    Williams, A.E.; Linkison, W.S.

    1977-01-01

    A fuel rod is proposed where a reorientation of the fuel in case of a considerable temperature increase, causing the melting of the densified fuel powder, will be avoided. For this purpose, in longitudinal direction of the fuel rod, a number of diameter reductions of the can are applied of certain distances. In the reduction zone the cross-sectional area of the fuel is reduced, as compared to the one of the remaining fuel material in the regions without diameter reduction, but not the density of the fuel. The recess is chosen to that in case of melting of the fuel in the center of the not contracted zone the fuel in the center of the narrowed area will remain solid and keep the molten material in position. (HR) [de

  8. Consolidated fuel reprocessing program

    Science.gov (United States)

    1985-04-01

    A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

  9. Nuclear fuel storage

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1979-01-01

    A method and apparatus for the storage of fuel in a stainless steel egg crate structure within a storage pool are described. Fuel is initially stored in a checkerboard pattern or in each opening if the fuel is of low enrichment. Additional fuel (or fuel of higher enrichment) is later stored by adding stainless steel angled plates within each opening, thereby forming flux traps between the openings. Still higher enrichment fuel is later stored by adding poison plates either with or without the stainless steel angles. 8 claims

  10. BNFL Springfields Fuel Division

    International Nuclear Information System (INIS)

    Tarkiainen, S.; Plit, H.

    1998-01-01

    The Fuel Division of British Nuclear Fuels Ltd (BNFL) manufactures nuclear fuel elements for British Magnox and AGR power plants as well as for LWR plants. The new fuel factory - Oxide Fuel Complex (OFC), located in Springfields, is equipped with modern technology and the automation level of the factory is very high. With their quality products, BNFL aims for the new business areas. A recent example of this expansion was shown, when BNFL signed a contract to design and license new VVER-440 fuel for Finnish Loviisa and Hungarian Paks power plants. (author)

  11. Nuclear fuel activities in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Bairiot, H

    1997-12-01

    In his presentation on nuclear fuel activities in belgium the author considers the following directions of this work: fuel fabrication, NPP operation, fuel performance, research and development programmes.

  12. Advanced fuel development at AECL: What does the future hold for CANDU fuels/fuel cycles?

    Energy Technology Data Exchange (ETDEWEB)

    Kupferschmidt, W.C.H. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    This paper outlines advanced fuel development at AECL. It discusses expanding the limits of fuel utilization, deploy alternate fuel cycles, increase fuel flexibility, employ recycled fuels; increase safety and reliability, decrease environmental impact and develop proliferation resistant fuel and fuel cycle.

  13. 76 FR 37703 - Regulation of Fuels and Fuel Additives: 2012 Renewable Fuel Standards; Public Hearing

    Science.gov (United States)

    2011-06-28

    ... Regulation of Fuels and Fuel Additives: 2012 Renewable Fuel Standards; Public Hearing AGENCY: Environmental... hearing to be held for the proposed rule ``Regulation of Fuels and Fuel Additives: 2012 Renewable Fuel... be proposing amendments to the renewable fuel standard program regulations to establish annual...

  14. DUPIC fuel compatibility assessment

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Rho, G. H.; Park, J. W. [and others

    2000-03-01

    The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition.

  15. DUPIC fuel compatibility assessment

    International Nuclear Information System (INIS)

    Choi, Hang Bok; Rho, G. H.; Park, J. W. and others

    2000-03-01

    The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition

  16. The Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    2011-08-01

    This brochure describes the nuclear fuel cycle, which is an industrial process involving various activities to produce electricity from uranium in nuclear power reactors. The cycle starts with the mining of uranium and ends with the disposal of nuclear waste. The raw material for today's nuclear fuel is uranium. It must be processed through a series of steps to produce an efficient fuel for generating electricity. Used fuel also needs to be taken care of for reuse and disposal. The nuclear fuel cycle includes the 'front end', i.e. preparation of the fuel, the 'service period' in which fuel is used during reactor operation to generate electricity, and the 'back end', i.e. the safe management of spent nuclear fuel including reprocessing and reuse and disposal. If spent fuel is not reprocessed, the fuel cycle is referred to as an 'open' or 'once-through' fuel cycle; if spent fuel is reprocessed, and partly reused, it is referred to as a 'closed' nuclear fuel cycle.

  17. The plutonium fuel cycles

    International Nuclear Information System (INIS)

    Pigford, T.H.; Ang, K.P.

    1975-01-01

    The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000-MW water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium and recycled uranium. The radioactivity quantities of plutonium, americium and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the U.S. nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing ad fuel fabrication to eliminate the off-site transport of separated plutonium. (author)

  18. Fuel Assembly Damping Summary

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kanghee; Kang, Heungseok; Oh, Dongseok; Yoon, Kyungho; Kim, Hyungkyu; Kim, Jaeyong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    This paper summary the fuel assembly damping data in air/in still water/under flow, released from foreign fuel vendors, compared our data with the published data. Some technical issues in fuel assembly damping measurement testing are also briefly discussed. Understanding of each fuel assembly damping mechanisms according to the surrounding medium and flow velocity can support the fuel design improvement in fuel assembly dynamics and structural integrity aspect. Because the upgraded requirements of the newly-developed advanced reactor system will demands to minimize fuel design margin in integrity evaluation, reduction in conservatism of fuel assembly damping can contribute to alleviate the fuel design margin for sure. Damping is an energy dissipation mechanism in a vibrating mechanical structure and prevents a resonant structure from having infinite vibration amplitudes. The sources of fuel assembly damping are various from support friction to flow contribution, and it can be increased by the viscosity or drag of surrounding fluid medium or the average velocity of water flowing. Fuel licensing requires fuel design evaluation in transient or accidental condition. Dynamic response analysis of fuel assembly is to show fuel integrity and requires information on assembly-wise damping in dry condition and under wet or water flowing condition. However, damping measurement test for the full-scale fuel assembly prototype is not easy to carry out because of the scale (fuel prototype, test facility), unsteadiness of test data (scattering, random sampling and processing), instrumentation under water flowing (water-proof response measurement), and noise. LWR fuel technology division in KAERI is preparing the infra structure for damping measurement test of full-scale fuel assembly, to support fuel industries and related research activities. Here is a preliminary summary of fuel assembly damping, published in the literature. Some technical issues in fuel assembly damping

  19. Romanian nuclear fuel program

    International Nuclear Information System (INIS)

    Budan, O.

    1999-01-01

    The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified and authorised CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle

  20. Fuel Cell Electric Bus Evaluations | Hydrogen and Fuel Cells | NREL

    Science.gov (United States)

    Bus Evaluations Fuel Cell Electric Bus Evaluations NREL's technology validation team evaluates fuel cell electric buses (FCEBs) to provide comprehensive, unbiased evaluation results of fuel cell bus early transportation applications for fuel cell technology. Buses operate in congested areas where

  1. Fuel Cell and Hydrogen Technologies Program | Hydrogen and Fuel Cells |

    Science.gov (United States)

    NREL Fuel Cell and Hydrogen Technologies Program Fuel Cell and Hydrogen Technologies Program Through its Fuel Cell and Hydrogen Technologies Program, NREL researches, develops, analyzes, and validates fuel cell and hydrogen production, delivery, and storage technologies for transportation

  2. Oxy-fuel combustion of pulverized fuels

    DEFF Research Database (Denmark)

    Yin, Chungen; Yan, Jinyue

    2016-01-01

    Oxy-fuel combustion of pulverized fuels (PF), as a promising technology for CO2 capture from power plants, has gained a lot of concerns and also advanced considerable research, development and demonstration in the last past years worldwide. The use of CO2 or the mixture of CO2 and H2O vapor as th...

  3. Logistic Fuel Processor Development

    National Research Council Canada - National Science Library

    Salavani, Reza

    2004-01-01

    ... to light gases then steam reform the light gases into hydrogen rich stream. This report documents the efforts in developing a fuel processor capable of providing hydrogen to a 3kW fuel cell stack...

  4. Future automotive fuels

    International Nuclear Information System (INIS)

    Lepik, M.

    1993-01-01

    There are several important factors which are fundamental to the choice of alternative automobile fuels: the chain of energetic efficiency of fuels; costs; environmental friendliness; suitability for usual engines or adapting easiness; existing reserves of crude oil, natural gas or the fossil energy sources; and, alternatively, agricultural potentiality. This paper covers all these factors. The fuels dealt with in this paper are alcohol, vegetable oil, gaseous fuel, hydrogen and ammonia fuels. Renewable fuels are the most valuable forms of renewable energy. In addition to that rank, they can contribute to three other problem areas: agricultural surpluses, environmental degradation, and conservation of natural resources. Due to the competitive utilization of biomass for food energy production, bio-fuels should mainly be produced in those countries where an energy shortage is combined with a food surplus. The fuels arousing the most interest are alcohol and vegetable oil, the latter for diesel engines, even in northern countries. (au)

  5. Fuel cells: Project Volta

    Energy Technology Data Exchange (ETDEWEB)

    Vellone, R.; Di Mario, F.

    1987-09-01

    This paper discusses research and development in the field of fuel cell power plants. Reference is made to the Italian research Project Volta. Problems related to research program financing and fuel cell power plant marketing are discussed.

  6. Fuel transporting device

    International Nuclear Information System (INIS)

    Shiratori, Hirozo.

    1979-01-01

    Purpose: In a liquid-metal cooled reactor, to reduce the waiting time of fuel handling apparatuses and shorten the fuel exchange time. Constitution: A fuel transporting machine is arranged between a reactor vessel and an out-pile storage tank, thereby dividing the transportation line of the pot for contracting fuel and transporting the same. By assuming such a construction, the flow of fuel transportation which has heretofore been carried out through fuel transportation pipes is not limited to one direction but the take-out of fuels from the reactor and the take-in thereof from the storage tank can be carried out constantly, and much time is not required for fuel exchange. (Kamimura, M.)

  7. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1981-01-01

    An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  8. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1984-01-01

    The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  9. Fuel assembly guide tube

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    This invention is directed toward a nuclear fuel assembly guide tube arrangement which restrains spacer grid movement due to coolant flow and which offers secondary means for supporting a fuel assembly during handling and transfer operations

  10. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    This chapter explains the distinction between fissile and fertile materials, examines briefly the processes involved in fuel manufacture and management, describes the alternative nuclear fuel cycles and considers their advantages and disadvantages. Fuel management is usually divided into three stages; the front end stage of production and fabrication, the back end stage which deals with the fuel after it is removed from the reactor (including reprocessing and waste treatment) and the stage in between when the fuel is actually in the reactor. These stages are illustrated and explained in detail. The plutonium fuel cycle and thorium-uranium-233 fuel cycle are explained. The differences between fuels for thermal reactors and fast reactors are explained. (U.K.)

  11. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Tomihiro.

    1970-01-01

    The present invention relates to fuel assemblies employing wire wrap spacers for retaining uniform spatial distribution between fuel elements. Clad fuel elements are helically wound in the oxial direction with a wave-formed wire strand. The strand is therefore provided with spring action which permits the fuel elements to expand freely in the axial and radial directions so as to retain proper spacing and reduce stresses due to thermal deformation. (Ownes, K.J.)

  12. Fuels and auxiliary materials

    International Nuclear Information System (INIS)

    Svab, V.

    A brief survey is given of the problems of fuels, fuel cans, absorption and moderator materials proceeding from the papers presented at the 1971 4th Geneva Conference on the Peaceful Uses of Nuclear Energy and the 1970 IAEA Conference in New York. Attention is focused on the behaviour of fuel and fuel can materials for thermal and fast reactors during irradiation, radiation stability of absorption materials and the effects of radiation on concrete and on moderator materials. (Z.M.)

  13. Fuel management and economics

    Energy Technology Data Exchange (ETDEWEB)

    Vendryes, G

    1972-11-01

    From international conference on nuclear solutions to world energy problems; Washington, District of Columbia, USA (12 Nov The low cost of the fuel cycle is the most attractive feature of the fast neutron breeder reactor. In order to achieve it a good fuel management is essential, with well balanced fixed investment and renewal fuel costs. In addition the designer can optimize the power station as a whole (fuel cycle and thermal characteristics). (auth)

  14. Direct hydrocarbon fuel cells

    Science.gov (United States)

    Barnett, Scott A.; Lai, Tammy; Liu, Jiang

    2010-05-04

    The direct electrochemical oxidation of hydrocarbons in solid oxide fuel cells, to generate greater power densities at lower temperatures without carbon deposition. The performance obtained is comparable to that of fuel cells used for hydrogen, and is achieved by using novel anode composites at low operating temperatures. Such solid oxide fuel cells, regardless of fuel source or operation, can be configured advantageously using the structural geometries of this invention.

  15. Spent fuels program

    International Nuclear Information System (INIS)

    Shappert, L.B.

    1983-01-01

    The goal of this task is to support the Domestic Spent Fuel Storage Program through studies involving the transport of spent fuel. A catalog was developed to provide authoritative, timely, and accessible transportation information for persons involved in the transport of irradiated reactor fuel. The catalog, drafted and submitted to the Transportation Technology Center, Sandia National Laboratories, for their review and approval, covers such topics as federal, state, and local regulations, spent fuel characteristics, cask characteristics, transportation costs, and emergency response information

  16. Fuel vapor pressure (FVAPRS)

    International Nuclear Information System (INIS)

    Mason, R.E.

    1979-04-01

    A subcode (FVAPRS) is described which calculates fuel vapor pressure. This subcode was developed as part of the fuel rod behavior modeling task performed at EG and G Idaho, Inc. The fuel vapor pressure subcode (FVAPRS), is presented and a discussion of literature data, steady state and transient fuel vapor pressure equations and estimates of the standard error of estimate to be expected with the FVAPRS subcode are included

  17. Hydrogen Fuel Cell Vehicles

    OpenAIRE

    Anton Francesch, Judit

    1992-01-01

    Hydrogen is an especially attractive transportation fuel. It is the least polluting fuel available, and can be produced anywhere there is water and a clean source of electricity. A fuel cycle in which hydrogen is produced by solar-electrolysis of water, or by gasification of renewably grown biomass, and then used in a fuel-cell powered electric-motor vehicle (FCEV), would produce little or no local, regional, or global pollution. Hydrogen FCEVs would combine the best features of bat...

  18. Denatured fuel cycles

    International Nuclear Information System (INIS)

    Till, C.E.

    1979-01-01

    This paper traces the history of the denatured fuel concept and discusses the characteristics of fuel cycles based on the concept. The proliferation resistance of denatured fuel cycles, the reactor types they involve, and the limitations they place on energy generation potential are discussed. The paper concludes with some remarks on the outlook for such cycles

  19. Hydrogen and fuel cells

    International Nuclear Information System (INIS)

    2006-06-01

    This road-map proposes by the Group Total aims to inform the public on the hydrogen and fuel cells. It presents the hydrogen technology from the production to the distribution and storage, the issues as motor fuel and fuel cells, the challenge for vehicles applications and the Total commitments in the domain. (A.L.B.)

  20. Nuclear fuel cycle

    International Nuclear Information System (INIS)

    1993-01-01

    Status of different nuclear fuel cycle phases in 1992 is discussed including the following issues: uranium exploration, resources, supply and demand, production, market prices, conversion, enrichment; reactor fuel technology; spent fuel management, as well as trends of these phases development up to the year 2010. 10 refs, 11 figs, 15 tabs

  1. PWR fuel thermomechanics

    International Nuclear Information System (INIS)

    Traccucci, R.; Leclercq, J.

    1986-01-01

    Fuel thermo-mechanics means the studies of mechanical and thermal effects, and more generally, the studies of the behavior of the fuel assembly under stresses including thermal and mechanical loads, hydraulic effects and phenomena induced by materials irradiation. This paper describes the studies dealing with the fuel assembly behavior, first in normal operating conditions, and then in accidental conditions. 43 refs [fr

  2. Plutonium fuel program

    International Nuclear Information System (INIS)

    1979-09-01

    A review is presented of the development of the (UPu)C sphere-pac fuel project during 1978. In particular, the problems encountered in obtaining good fuel quality in the fabrication process and their solution is discussed. The development of a fabrication pilot plant is considered, and the post-irradiation examination of fuel pins is presented. (Auth.)

  3. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sasaki, Y.; Tashima, J.

    1975-01-01

    A description is given of nuclear reactor fuel assemblies arranged in the form of a lattice wherein there is attached to the interface of one of two adjacent fuel assemblies a plate spring having a concave portion curved toward said interface and to the interface of the other fuel assembly a plate spring having a convex portion curved away from said interface

  4. Gelled fuel simulant

    International Nuclear Information System (INIS)

    Christy, J.; Hiser, E.J.; Sippel, N.J.

    1980-01-01

    A relatively stable inert simulant formulation for a hazardous metallized fuel has the density, shear rate and yield stress of the duplicated fuel. This formulation provides inexpensive and safe testing of exploratory hydraulic studies, or testing of the mechanical strength of containers, plumbing, etc., in which the metallized fuels are to be used

  5. Fireplaces and Fireplace Fuels.

    Science.gov (United States)

    Metz, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fireplaces and fuels. Its objective is for the student to be able to discuss the structural design, operation, and efficiency of fireplaces and characteristics of different fireplace fuels. Some topics covered are fuels, elements…

  6. Metallic fuel development

    International Nuclear Information System (INIS)

    Walters, L.C.

    1987-01-01

    Metallic fuels are capable of achieving high burnup as a result of design modifications instituted in the late 1960's. The gap between the fuel slug and the cladding is fixed such that by the time the fuel swells to the cladding the fission gas bubbles interconnect and release the fission gas to an appropriately sized plenum volume. Interconnected porosity thus provides room for the fuel to deform from further swelling rather than stress the cladding. In addition, the interconnected porosity allows the fuel pin to be tolerant to transient events because as stresses are generated during a transient event the fuel flows rather than applying significant stress to the cladding. Until 1969 a number of metallic fuel alloys were under development in the US. At that time the metallic fuel development program in the US was discontinued in favor of ceramic fuels. However, development had proceeded to the point where it was clear that the zirconium addition to uranium-plutonium fuel would yield a ternary fuel with an adequately high solidus temperature and good compatibility with austenitic stainless steel cladding. Furthermore, several U-Pu-Zr fuel pins had achieved about 6 at.% bu by the late 1960's, without failure, and thus the prospect for high burnup was promising

  7. Modeling fuel succession

    Science.gov (United States)

    Brett Davis; Jan van Wagtendonk; Jen Beck; Kent van Wagtendonk

    2009-01-01

    Surface fuels data are of critical importance for supporting fire incident management, risk assessment, and fuel management planning, but the development of surface fuels data can be expensive and time consuming. The data development process is extensive, generally beginning with acquisition of remotely sensed spatial data such as aerial photography or satellite...

  8. Fuel and fuel cycles with high burnup for WWER reactors

    International Nuclear Information System (INIS)

    Chernushev, V.; Sokolov, F.

    2002-01-01

    The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

  9. Integral nuclear fuel element assembly

    International Nuclear Information System (INIS)

    Schluderberg, D. C.

    1985-01-01

    An integral nuclear fuel element assembly utilizes longitudinally finned fuel pins. The continuous or interrupted fins of the fuel pins are brazed to fins of juxtaposed fuel pins or directly to the juxtaposed fuel pins or both. The integrally brazed fuel assembly is designed to satisfy the thermal and hydraulic requirements of a fuel assembly lattice having moderator to fuel atom ratios required to achieve high conversion and breeding ratios

  10. Reactor fuel element and fuel assembly

    International Nuclear Information System (INIS)

    Okada, Seiji; Ishida, Tsuyoshi; Ikeda, Atsuko.

    1997-01-01

    A mixture of fission products and burnable poisons is disposed at least to a portion between MOX pellets to form a burnable poison-incorporated fuel element without mixing burnable poisons to the MOX pellets. Alternatively, a mixture of materials other than the fission products and burnable poisons is formed into disks, a fuel lamination portion is divided into at least to two regions, and the ratio of number of the disks of the mixture relative to the volume of the region is increased toward the lower portion of the fuel lamination portion. With such a constitution, the axial power distribution of fuels can be made flat easily. Alternatively, the thickness of the disk of the mixture is increased toward the lower region of the fuel lamination portion to flatten the axial power distribution of the fuels in the same manner easily. The time and the cost required for the manufacture are reduced, and MOX fuels filled with burnable poisons with easy maintenance and control can be realized. (N.H.)

  11. 77 FR 61313 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard and Diesel...

    Science.gov (United States)

    2012-10-09

    ... transportation fuels, including gasoline and diesel fuel, or renewable fuels such as ethanol and biodiesel, as... that which arose under RFS1 for certain renewable fuels (in particular biodiesel) that were produced...

  12. Fuel transfer machine

    International Nuclear Information System (INIS)

    Bernstein, I.

    1978-01-01

    A nuclear fuel transfer machine for transferring fuel assemblies through the fuel transfer tube of a nuclear power generating plant containment structure is described. A conventional reversible drive cable is attached to the fuel transfer carriage to drive it horizontally through the tube. A shuttle carrying a sheave at each end is arranged in parallel with the carriage to also travel into the tube. The cable cooperating with the sheaves permit driving a relatively short fuel transfer carriage a large distance without manually installing sheaves or drive apparatus in the tunnel. 8 claims, 3 figures

  13. Transportation of nuclear fuel

    International Nuclear Information System (INIS)

    Prowse, D.R.

    1979-01-01

    Shipment of used fuel from nuclear reactors to a central fuel management facility is discussed with particular emphasis on the assessment of the risk to the public due to these shipments. The methods of transporting used fuel in large shipping containers is reviewed. In terms of an accident scenario, it is demonstrated that the primary risk of transport of used fuel is due to injury and death in common road accidents. The radiological nature of the used fuel cargo is, for all practical purposes, an insignificant factor in the total risk to the public. (author)

  14. Nuclear fuel lease accounting

    International Nuclear Information System (INIS)

    Danielson, A.H.

    1986-01-01

    The subject of nuclear fuel lease accounting is a controversial one that has received much attention over the years. This has occurred during a period when increasing numbers of utilities, seeking alternatives to traditional financing methods, have turned to leasing their nuclear fuel inventories. The purpose of this paper is to examine the current accounting treatment of nuclear fuel leases as prescribed by the Financial Accounting Standards Board (FASB) and the Federal Energy Regulatory Commission's (FERC's) Uniform System of Accounts. Cost accounting for leased nuclear fuel during the fuel cycle is also discussed

  15. Nuclear fuel string assembly

    International Nuclear Information System (INIS)

    Ip, A.K.; Koyanagi, K.; Tarasuk, W.R.

    1976-01-01

    A method of fabricating rodded fuels suitable for use in pressure tube type reactors and in pressure vessel type reactors is described. Fuel rods are secured as an inner and an outer sub-assembly, each rod attached between mounting rings secured to the rod ends. The two sub-assemblies are telescoped together and positioned by spaced thimbles located between them to provide precise positioning while permittng differential axial movement between the sub-assemblies. Such sub-assemblies are particularly suited for mounting as bundle strings. The method provides particular advantages in the assembly of annular-section fuel pins, which includes booster fuel containing enriched fuel material. (LL)

  16. Mox fuels recycling

    International Nuclear Information System (INIS)

    Gay, A.

    1998-01-01

    This paper will firstly emphasis that the first recycling of plutonium is already an industrial reality in France thanks to the high degree of performance of La Hague and MELOX COGEMA's plants. Secondly, recycling of spent Mixed OXide fuel, as a complete MOX fuel cycle, will be demonstrated through the ability of the existing plants and services which have been designed to proceed with such fuels. Each step of the MOX fuel cycle concept will be presented: transportation, reception and storage at La Hague and steps of spent MOX fuel reprocessing. (author)

  17. Fuel cell opportunities

    Energy Technology Data Exchange (ETDEWEB)

    Harris, K. [Hydrogenics Corporation, Mississauga, ON (Canada)

    2002-07-01

    The opportunities for fuel cell development are discussed. Fuel cells are highly efficient, reliable and require little maintenance. They also produce virtually zero emissions. The author stated that there are some complicated issues to resolve before fuel cells can be widely used. These include hydrogen availability and infrastructure. While the cost of fuel cells is currently very high, these costs are constantly coming down. The industry is still in the early stages of development. The driving forces for the development of fuel cells are: deregulation of energy markets, growing expectations for distributed power generation, discontinuity between energy supply and demand, and environmental concerns. 12 figs.

  18. Fuel loads and fuel type mapping

    Science.gov (United States)

    Chuvieco, Emilio; Riaño, David; Van Wagtendonk, Jan W.; Morsdof, Felix; Chuvieco, Emilio

    2003-01-01

    Correct description of fuel properties is critical to improve fire danger assessment and fire behaviour modeling, since they guide both fire ignition and fire propagation. This chapter deals with properties of fuel that can be considered static in short periods of time: biomass loads, plant geometry, compactness, etc. Mapping these properties require a detail knowledge of vegetation vertical and horizontal structure. Several systems to classify the great diversity of vegetation characteristics in few fuel types are described, as well as methods for mapping them with special emphasis on those based on remote sensing images.

  19. Fuel cells : a viable fossil fuel alternative

    Energy Technology Data Exchange (ETDEWEB)

    Paduada, M.

    2007-02-15

    This article presented a program initiated by Natural Resources Canada (NRCan) to develop proof-of-concept of underground mining vehicles powered by fuel cells in order to eliminate emissions. Recent studies on American and Canadian underground mines provided the basis for estimating the operational cost savings of switching from diesel to fuel cells. For the Canadian mines evaluated, the estimated ventilation system operating cost reductions ranged from 29 per cent to 75 per cent. In order to demonstrate the viability of a fuel cell-powered vehicle, NRCan has designed a modified Caterpillar R1300 loader with a 160 kW hybrid power plant in which 3 stacks of fuel cells deliver up to 90 kW continuously, and a nickel-metal hydride battery provides up to 70 kW. The battery subsystem transiently boosts output to meet peak power requirements and also accommodates regenerative braking. Traction for the loader is provided by a brushless permanent magnet traction motor. The hydraulic pump motor is capable of a 55 kW load continuously. The loader's hydraulic and traction systems are operated independently. Future fuel cell-powered vehicles designed by the program may include a locomotive and a utility vehicle. Future mines running their operations with hydrogen-fueled equipment may also gain advantages by employing fuel cells in the operation of handheld equipment such as radios, flashlights, and headlamps. However, the proton exchange membrane (PEM) fuel cells used in the project are prohibitively expensive. The catalytic content of a fuel cell can add hundreds of dollars per kW of electric output. Production of catalytic precious metals will be strongly connected to the scale of use and acceptance of fuel cells in vehicles. In addition, the efficiency of hydrogen production and delivery is significantly lower than the well-to-tank efficiency of many conventional fuels. It was concluded that an adequate hydrogen infrastructure will be required for the mining industry

  20. Fuel characteristics pertinent to the design of aircraft fuel systems

    Science.gov (United States)

    Barnett, Henry C; Hibbard, R R

    1953-01-01

    Because of the importance of fuel properties in design of aircraft fuel systems the present report has been prepared to provide information on the characteristics of current jet fuels. In addition to information on fuel properties, discussions are presented on fuel specifications, the variations among fuels supplied under a given specification, fuel composition, and the pertinence of fuel composition and physical properties to fuel system design. In some instances the influence of variables such as pressure and temperature on physical properties is indicated. References are cited to provide fuel system designers with sources of information containing more detail than is practicable in the present report.

  1. Fuel charging machine

    International Nuclear Information System (INIS)

    Uchikawa, Sadao.

    1978-01-01

    Purpose: To enable continuous fuel discharging and charging steps in a bwr type reactor by effecting positioning only for once by providing a plurality of fuel assembly grippers and their drives co-axially on a rotatable surface. Constitution: A plurality of fuel assembly grippers and their drives are provided co-axially on a rotatable surface. For example, a gripper A, a drive B, a gripper C and a drive D are arranged co-axially in symmetric positions on a disk rotated on rails by wheels and rotational drives. A new fuel in a fuel pool is gripped by the gripper A and transported above the reactor core. Then, the disk is positioned so that the gripper C can grip the spent fuel in the core, and the fuel to be discharged is gripped and raised by the gripper C. Then the disk is rotated by 180 0 and the new fuel in the gripper A is charged into the position from which the old fuel has been discharged and, finally, the discharged fuel is sent to the fuel pool for storage. (Seki, T.)

  2. Fuel nozzle assembly

    Science.gov (United States)

    Johnson, Thomas Edward [Greer, SC; Ziminsky, Willy Steve [Simpsonville, SC; Lacey, Benjamin Paul [Greer, SC; York, William David [Greer, SC; Stevenson, Christian Xavier [Inman, SC

    2011-08-30

    A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

  3. Ducted fuel injection

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Charles J.

    2018-03-06

    Various technologies presented herein relate to enhancing mixing inside a combustion chamber to form one or more locally premixed mixtures comprising fuel and charge-gas with low peak fuel to charge-gas ratios to enable minimal, or no, generation of soot and other undesired emissions during ignition and subsequent combustion of the locally premixed mixtures. To enable sufficient mixing of the fuel and charge-gas, a jet of fuel can be directed to pass through a bore of a duct causing charge-gas to be drawn into the bore creating turbulence to mix the fuel and the drawn charge-gas. The duct can be located proximate to an opening in a tip of a fuel injector. The duct can comprise of one or more holes along its length to enable charge-gas to be drawn into the bore, and further, the duct can cool the fuel and/or charge-gas prior to combustion.

  4. Reactor fuel charging equipment

    International Nuclear Information System (INIS)

    Wade, Elman.

    1977-01-01

    In many types of reactor fuel charging equipment, tongs or a grab, attached to a trolley, housed in a guide duct, can be used for withdrawing from the core a selected spent fuel assembly or to place a new fuel assembly in the core. In these facilities, the trolley may have wheels that roll on rails in the guide duct. This ensures the correct alignment of the grab, the trolley and fuel assembly when this fuel assembly is being moved. By raising or lowering such a fuel assembly, the trolley can be immerged in the coolant bath of the reactor, whereas at other times it can be at a certain level above the upper surface of the coolant bath. The main object of the invention is to create a fuel handling apparatus for a sodium cooled reactor with bearings lubricated by the sodium coolant and in which the contamination of these bearings is prevented [fr

  5. Fuel element services

    International Nuclear Information System (INIS)

    Marta, H.; Alvarez, P.; Jimenez, J.

    2006-01-01

    Refuelling outages comprise a number of maintenance tasks scheduled long in advance to assure a reliable operation throughout the next cycle and, in the long run, a safer and more efficient plant. Most of these tasks are routine service of mechanical and electrical system and likewise fuel an be considered a critical component as to handling, inspection, cleaning and repair. ENUSA-ENWESA AIE has been working in this area since 1995 growing from fuel repair to a more integrated service that includes new and spent fuel handling, inserts, failed fuel rod detection systems, ultrasonic fuel cleaning, fuel repair and a comprehensive array of inspection and tests related to the reliability of the mechanical components in the fuel assembly, all this, performed in compliance with quality, safety, health physics and any other nuclear standard. (Author)

  6. Fuel cells 101

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, B.

    2003-06-01

    A capsule history of fuel cells is given, beginning with the first discovery in 1839 by William Grove, a Welsh judge who, when experimenting with electrolysis discovered that by re-combining the two components of electrolysis (water and oxygen) an electric charge was produced. A century later, in 1958, Francis Thomas Bacon, a British scientist demonstrated the first working fuel cell stack, a technology which was licensed and used in the Apollo spacecraft. In Canada, early research on the development of fuel cells was carried out at the University of Toronto, the Defence Research Establishment and the National Research Council. Most of the early work concentrated on alkaline and phosphoric acid fuel cells. In 1983, Ballard Research began the development of the electrolyte membrane fuel cell, which marked the beginning of Canada becoming a world leader in fuel cell technology development. The paper provides a brief account of how fuel cells work, describes the distinguishing characteristics of the various types of fuel cells (alkaline, phosphoric acid, molten-carbonate, solid oxide, and proton exchange membrane types) and their principal benefits. The emphasis is on proton exchange membrane fuel cells because they are the only fuel cell technology that is appropriate for providing primary propulsion power onboard a vehicle. Since vehicles are by far the greatest consumers of fossil fuels, it follows that proton exchange membrane fuel cells will have the greatest potential impact on both environmental matters and on our reliance on oil as our primary fuel. Various on-going and planned fuel cell demonstration projects are also described. 1 fig.

  7. 77 FR 72746 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard and Diesel...

    Science.gov (United States)

    2012-12-06

    ... Fuels and Fuel Additives: Modifications to Renewable Fuel Standard and Diesel Sulfur Programs AGENCY... Fuel Standard (``RFS'') program under section 211(o) of the Clean Air Act. The direct final rule also... marine diesel fuel produced by transmix processors, and the fuel marker requirements for 500 ppm sulfur...

  8. 78 FR 12005 - Regulation of Fuels and Fuel Additives: 2013 Renewable Fuel Standards; Public Hearing

    Science.gov (United States)

    2013-02-21

    ... Regulation of Fuels and Fuel Additives: 2013 Renewable Fuel Standards; Public Hearing AGENCY: Environmental... EPA is announcing a public hearing to be held for the proposed rule ``Regulation of Fuels and Fuel Additives: 2013 Renewable Fuel Standards,'' which was published separately in the Federal Register on...

  9. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  10. Oxy-fuel combustion of solid fuels

    DEFF Research Database (Denmark)

    Toftegaard, Maja Bøg; Brix, Jacob; Jensen, Peter Arendt

    2010-01-01

    Oxy-fuel combustion is suggested as one of the possible, promising technologies for capturing CO2 from power plants. The concept of oxy-fuel combustion is removal of nitrogen from the oxidizer to carry out the combustion process in oxygen and, in most concepts, recycled flue gas to lower the flame...... provide additional options for improvement of process economics are however likewise investigated. Of particular interest is the change of the combustion process induced by the exchange of carbon dioxide and water vapor for nitrogen as diluent. This paper reviews the published knowledge on the oxy......-fuel process and focuses particularly on the combustion fundamentals, i.e. flame temperatures and heat transfer, ignition and burnout, emissions, and fly ash characteristics. Knowledge is currently available regarding both an entire oxy-fuel power plant and the combustion fundamentals. However, several...

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

    Energy Technology Data Exchange (ETDEWEB)

    Samuel J. Miller; Hakan Ozaltun

    2012-11-01

    This article analyzes dimensional changes due to irradiation of monolithic plate-type nuclear fuel and compares results with finite element analysis of the plates during fabrication and irradiation. Monolithic fuel plates tested in the Advanced Test Reactor (ATR) at Idaho National Lab (INL) are being used to benchmark proposed fuel performance for several high power research reactors. Post-irradiation metallographic images of plates sectioned at the midpoint were analyzed to determine dimensional changes of the fuel and the cladding response. A constitutive model of the fabrication process and irradiation behavior of the tested plates was developed using the general purpose commercial finite element analysis package, Abaqus. Using calculated burn-up profiles of irradiated plates to model the power distribution and including irradiation behaviors such as swelling and irradiation enhanced creep, model simulations allow analysis of plate parameters that are either impossible or infeasible in an experimental setting. The development and progression of fabrication induced stress concentrations at the plate edges was of primary interest, as these locations have a unique stress profile during irradiation. Additionally, comparison between 2D and 3D models was performed to optimize analysis methodology. In particular, the ability of 2D and 3D models account for out of plane stresses which result in 3-dimensional creep behavior that is a product of these components. Results show that assumptions made in 2D models for the out-of-plane stresses and strains cannot capture the 3-dimensional physics accurately and thus 2D approximations are not computationally accurate. Stress-strain fields are dependent on plate geometry and irradiation conditions, thus, if stress based criteria is used to predict plate behavior (as opposed to material impurities, fine micro-structural defects, or sharp power gradients), unique 3D finite element formulation for each plate is required.

  12. Nuclear fuel storage

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1981-01-01

    A nuclear fuel storage apparatus for use in a water-filled pool is fabricated of a material such as stainless steel in the form of an egg crate structure having vertically extending openings. Fuel may be stored in this basic structure in a checkerboard pattern with high enrichment fuel, or in all openings when the fuel is of low effective enrichment. Inserts of a material such as stainless steel are adapted to fit within these openings so that a water gap and, therefore, a flux trap is formed between adjacent fuel storage locations. These inserts may be added at a later time and fuel of a higher enrichment may be stored in each opening. When it is desired to store fuel of still greater enrichment, poison plates may be added to the water gap formed by the installed insert plates, or substituted for the insert plates. Alternately, or in addition, fuel may be installed in high neutron absorption poison boxes which surround the fuel assembly. The stainless steel inserts and the poison plates are each not required until the capacity of the basic egg crate structure is approached. Purchase of these items can, therefore, be deferred for many years. Should the fuel to be stored be of higher enrichment than initially forecast, the deferred decision on the poison plates makes it possible to obtain increased poison in the plates to satisfy the newly discovered requirement

  13. Diesel fuel filtration system

    International Nuclear Information System (INIS)

    Schneider, D.

    1996-01-01

    The American nuclear utility industry is subject to tight regulations on the quality of diesel fuel that is stored at nuclear generating stations. This fuel is required to supply safety-related emergency diesel generators--the backup power systems associated with the safe shutdown of reactors. One important parameter being regulated is the level of particulate contamination in the diesel fuel. Carbon particulate is a natural byproduct of aging diesel fuel. Carbon particulate precipitates from the fuel's hydrocarbons, then remains suspended or settles to the bottom of fuel oil storage tanks. If the carbon particulate is not removed, unacceptable levels of particulate contamination will eventually occur. The oil must be discarded or filtered. Having an outside contractor come to the plant to filter the diesel fuel can be costly and time consuming. Time is an even more critical factor if a nuclear plant is in a Limiting Condition of Operation (LCO) situation. A most effective way to reduce both cost and risk is for a utility to build and install its own diesel fuel filtration system. The cost savings associated with designing, fabricating and operating the system inhouse can be significant, and the value of reducing the risk of reactor shutdown because of uncertified diesel fuel may be even higher. This article describes such a fuel filtering system

  14. Fuel safety research 1999

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-07-01

    In April 1999, the Fuel Safety Research Laboratory was newly established as a result of reorganization of the Nuclear Safety Research Center, JAERI. The laboratory was organized by combining three laboratories, the Reactivity Accident Laboratory, the Fuel Reliability Laboratory, and a part of the Sever Accident Research Laboratory. Consequently, the Fuel Safety Research Laboratory is now in charge of all the fuel safety research in JAERI. Various types of experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of five research groups corresponding to each research fields. They are; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). This report summarizes the outline of research activities and major outcomes of the research executed in 1999 in the Fuel Safety Research Laboratory. (author)

  15. Fuel related risks; Braenslerisker

    Energy Technology Data Exchange (ETDEWEB)

    Englund, Jessica; Sernhed, Kerstin; Nystroem, Olle; Graveus, Frank (Grontmij AB, (Sweden))

    2012-02-15

    The project, within which this work report was prepared, aimed to complement the Vaermeforsk publication 'Handbook of fuels' on fuel related risks and measures to reduce the risks. The fuels examined in this project where the fuels included in the first version of the handbook from 2005 plus four additional fuels that will be included in the second and next edition of the handbook. Following fuels were included: woodfuels (sawdust, wood chips, powder, briquettes), slash, recycled wood, salix, bark, hardwood, stumps, straw, reed canary grass, hemp, cereal, cereal waste, olive waste, cocoa beans, citrus waste, shea, sludge, forest industrial sludge, manure, Paper Wood Plastic, tyre, leather waste, cardboard rejects, meat and bone meal, liquid animal and vegetable wastes, tall oil pitch, peat, residues from food industry, biomal (including slaughterhouse waste) and lignin. The report includes two main chapters; a general risk chapter and a chapter of fuel specific risks. The first one deals with the general concept of risk, it highlights laws and rules relevant for risk management and it discuss general risks that are related to the different steps of fuel handling, i.e. unloading, storing, processing the fuel, transportation within the facility, combustion and handling of ashes. The information that was used to produce this chapter was gathered through a literature review, site visits, and the project group's experience from risk management. The other main chapter deals with fuel-specific risks and the measures to reduce the risks for the steps of unloading, storing, processing the fuel, internal transportation, combustion and handling of the ashes. Risks and measures were considered for all the biofuels included in the second version in the handbook of fuels. Information about the risks and risk management was gathered through interviews with people working with different kinds of fuels in electricity and heat plants in Sweden. The information from

  16. Method of decladding spent fuel

    International Nuclear Information System (INIS)

    Fukutome, Kazuyuki; Kitagawa, Kazuo.

    1988-01-01

    Purpose: To enable to safety and easy decladding of nuclear fuels thereby reduce the processing cost. Constitution: Upon dismantling of a spent fuel rod, the fuel rod is heated at least to such a temperature that the ductility of a fuel can is recovered, then transported by using seizing rollers, by which the fuel rod is pressurized from the outer circumference to break the nuclear fuels at the inside thereof. Then, the destructed fuels are recovered from both ends of the fuel can. With such a constitution, since the ductility of the fuel can is recovered by heating, when the fuel rod is passed through the rollers in this state, the fuel can is deformed to destroy the nuclear fuels at the inside thereof. Since the nuclear fuels are destroyed into small pieces, they can be taken out easily from both ends of the fuel can. (Kawakami, Y.)

  17. Segmented fuel and moderator rod

    International Nuclear Information System (INIS)

    Doshi, P.K.

    1987-01-01

    This patent describes a continuous segmented fuel and moderator rod for use with a water cooled and moderated nuclear fuel assembly. The rod comprises: a lower fuel region containing a column of nuclear fuel; a moderator region, disposed axially above the fuel region. The moderator region has means for admitting and passing the water moderator therethrough for moderating an upper portion of the nuclear fuel assembly. The moderator region is separated from the fuel region by a water tight separator

  18. Nuclear fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding, and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

  19. HTPEM Fuel Cell Impedance

    DEFF Research Database (Denmark)

    Vang, Jakob Rabjerg

    As part of the process to create a fossil free Denmark by 2050, there is a need for the development of new energy technologies with higher efficiencies than the current technologies. Fuel cells, that can generate electricity at higher efficiencies than conventional combustion engines, can...... potentially play an important role in the energy system of the future. One of the fuel cell technologies, that receives much attention from the Danish scientific community is high temperature proton exchange membrane (HTPEM) fuel cells based on polybenzimidazole (PBI) with phosphoric acid as proton conductor....... This type of fuel cell operates at higher temperature than comparable fuel cell types and they distinguish themselves by high CO tolerance. Platinum based catalysts have their efficiency reduced by CO and the effect is more pronounced at low temperature. This Ph.D. Thesis investigates this type of fuel...

  20. Boosting nuclear fuels

    International Nuclear Information System (INIS)

    Demarthon, F.; Donnars, O.; Dupuy-Maury, F.

    2002-01-01

    This dossier gives a broad overview of the present day status of the nuclear fuel cycle in France: 1 - the revival of nuclear power as a solution to the global warming and to the increase of worldwide energy needs; 2 - the security of uranium supplies thanks to the reuse of weapon grade highly enriched uranium; 3 - the fabrication of nuclear fuels from the mining extraction to the enrichment processes, the fabrication of fuel pellets and the assembly of fuel rods; 4 - the new composition of present day fuels (UO x and chromium-doped pellets); 5 - the consumption of plutonium stocks and the Corail and Apa fuel assemblies for the reduction of plutonium stocks and the preservation of uranium resources. (J.S.)

  1. Nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Butterfield, R.S.; Garner, D.L.M.

    1977-01-01

    Reference is made to nuclear fuel assemblies designed for cooling on the 'tube-in-shell' principle in which the fuel is contained by a shell and is cooled by coolant passed through tubes extending through the shell. It has been proposed to employ coated particle fuel as a porous bed on the tube side and the bleed coolant from the tubes into direct contact with the fuel particles. In this way heat is extracted both by direct contact with the fuel and by heat transfer through the coolant tube walls. The system described aims to provide an improved structure of tube and shell for a fuel assembly of this kind and is particularly suitable for use in a gas cooled fast reactor, being able to withstand the neutron flux and high temperature conditions in these reactors. Constructional details are given. (U.K.)

  2. Nuclear fuel element

    International Nuclear Information System (INIS)

    Thompson, J.R.; Rowland, T.C.

    1976-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting, fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

  3. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Hayashi, Hiroshi; Watari, Yoshio; Hizahara, Hiroshi; Masuoka, Ryuzo.

    1970-01-01

    When exchanging nuclear fuel assemblies during the operation of a nuclear reactor, melting of fuel bodies, and severence of tubular claddings is halted at the time of insertion by furnishing a neutron absorbing material such as B 10 , Cd, Gd or the like at the forward end of the fuel assembly to thereby lower the power peak at the forward ends of the fuel elements to within tolerable levels and thus prevent both fuel liquification and excessive expansion. The neutron absorbing material may be attached in the form of a plate to the fuel assembly forward tie plate, or may be inserted as a pellet into the front end of the tubular cladding. (Owens, K.J.)

  4. Fuel assembly spacer

    International Nuclear Information System (INIS)

    Shirakawa, Ken-etsu.

    1988-01-01

    Purpose: To reduce the pressure loss of coolants by fuel assembly spacers. Constitution: Spacers for supporting a fuel assembly are attached by means of a plurality of wires to an outer frame. The outer frame is made of shape memory alloy such that the wires are caused to slacken at normal temperature and the slacking of the wires is eliminated in excess of the transition temperature. Since the wires slacken at the normal temperature, fuel rods can be inserted easily. After the insertion of the fuel rods, when the entire portion or the outer frame is heated by water or gas at a predetermined temperature, the outer frame resumes its previously memorized shape to tighten the wires and, accordingly, the fuel rods can be supported firmly. In this way, since the fuel rods are inserted in the slacken state of the wires and, after the assembling, the outer frame resumes its memorized shape, the assembling work can be conducted efficiently. (Kamimura, M.)

  5. Nuclear fuel element

    International Nuclear Information System (INIS)

    Hirayama, Satoshi; Kawada, Toshiyuki; Matsuzaki, Masayoshi.

    1980-01-01

    Purpose: To provide a fuel element for reducing the mechanical interactions between a fuel-cladding tube and the fuel element and for alleviating the limits of the operating conditions of a reactor. Constitution: A fuel element having mainly uranium dioxide consists of a cylindrical outer pellet and cylindrical inner pellet inserted into the outer pellet. The outer pellet contains two or more additives selected from aluminium oxide, beryllium oxide, magnesium oxide, silicon oxide, sodium oxide, phosphorus oxide, calcium oxide and iron oxide, and the inner pellet contains nuclear fuel substance solely or one additive selected from calcium oxide, silicon oxide, aluminium oxide, magnesium oxide, zirconium oxide and iron oxide. The outer pellet of the fuel thus constituted is reduced in mechanical strength and also in the mechanical interactions with the cladding tube, and the plastic fluidity of the entire pellet is prevented by the inner pellet increased in the mechanical strength. (Kamimura, M.)

  6. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Betten, P.R.

    1976-01-01

    Under the invention the fuel assembly is particularly suitable for liquid metal cooled fast neutron breeder reactors. Hence, according to the invention a fuel assembly cladding includes inward corrugations with respect to the remainder of the cladding according to a recurring pattern determined by the pitch of the metal wire helically wound round the fuel rods of the assembly. The parts of the cladding pressed inwards correspond to the areas in which the wire encircling the peripheral fuel rods is generally located apart from the cladding, thereby reducing the play between the cladding and the peripheral fuel rods situated in these areas. The reduction in the play in turn improves the coolant flow in the internal secondary channels of the fuel assembly to the detriment of the flow in the peripheral secondary channels and thereby establishes a better coolant fluid temperature profile [fr

  7. Nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Natori, Hisahide; Kurihara, Kunitoshi.

    1982-01-01

    Purpose: To increase the fuel safety by decreasing the gap conductance between fuels and cladding tubes, as well as improve the reactor core controllability by rendering the void coefficient negative. Constitution: Fuel assemblies in a pressure tube comprise a tie-rod, fuel rods in a central region, and fuel rods with burnable poison in the outer circumference region. Here, B 4 C is used as the burnable poison by 1.17 % by weight ratio. The degrees of enrichment for the fissile plutonium as PuO 2 -UO 2 fuel used in the assemblies are 2.7 %, 2.7 % and 1.5 % respectively in the innermost layer, the intermediate layer and the outermost layer. This increases the burn-up degree to improve the plant utilizability, whereby the void coefficient is rendered negative to improve the reactor core controllability. (Horiuchi, T.)

  8. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sakurai, Shungo; Ogiya, Shunsuke.

    1990-01-01

    In a fuel assembly, if the entire fuels comprise mixed oxide fuels, reactivity change in cold temperature-power operation is increased to worsen the reactor shutdown margin. The reactor shutdown margin has been improved by increasing the burnable poison concentration thereby reducing the reactivity of the fuel assembly. However, since unburnt poisons are present at the completion of the reactor operation, the reactivity can not be utilized effectively to bring about economical disadvantage. In view of the above, the reactivity change between lower temperature-power operations is reduced by providing a non-boiling range with more than 9.1% of cross sectional area at the inside of a channel at the central portion of the fuel assembly. As a result, the amount of the unburnt burnable poisons is decreased, the economy of fuel assembly is improved and the reactor shutdown margin can be increase. (N.H.)

  9. Fuel Cell and Hydrogen Technology Validation | Hydrogen and Fuel Cells |

    Science.gov (United States)

    NREL Fuel Cell and Hydrogen Technology Validation Fuel Cell and Hydrogen Technology Validation The NREL technology validation team works on validating hydrogen fuel cell electric vehicles; hydrogen fueling infrastructure; hydrogen system components; and fuel cell use in early market applications such as

  10. Alternative Fuels Data Center: Krug Energy Opens Natural Gas Fueling

    Science.gov (United States)

    Station in Arkansas Krug Energy Opens Natural Gas Fueling Station in Arkansas to someone by E -mail Share Alternative Fuels Data Center: Krug Energy Opens Natural Gas Fueling Station in Arkansas on Facebook Tweet about Alternative Fuels Data Center: Krug Energy Opens Natural Gas Fueling Station in

  11. Fuel cell generator with fuel electrodes that control on-cell fuel reformation

    Science.gov (United States)

    Ruka, Roswell J [Pittsburgh, PA; Basel, Richard A [Pittsburgh, PA; Zhang, Gong [Murrysville, PA

    2011-10-25

    A fuel cell for a fuel cell generator including a housing including a gas flow path for receiving a fuel from a fuel source and directing the fuel across the fuel cell. The fuel cell includes an elongate member including opposing first and second ends and defining an interior cathode portion and an exterior anode portion. The interior cathode portion includes an electrode in contact with an oxidant flow path. The exterior anode portion includes an electrode in contact with the fuel in the gas flow path. The anode portion includes a catalyst material for effecting fuel reformation along the fuel cell between the opposing ends. A fuel reformation control layer is applied over the catalyst material for reducing a rate of fuel reformation on the fuel cell. The control layer effects a variable reformation rate along the length of the fuel cell.

  12. Fuel safety research 2001

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-11-01

    The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

  13. Transport of MOX fuel

    International Nuclear Information System (INIS)

    Porter, I.R.; Carr, M.

    1997-01-01

    The regulatory framework which governs the transport of MOX fuel is set out, including packages, transport modes and security requirements. Technical requirements for the packages are reviewed and BNFL's experience in plutonium and MOX fuel transport is described. The safety of such operations and the public perception of safety are described and the question of gaining public acceptance for MOX fuel transport is addressed. The paper concludes by emphasising the need for proactive programmes to improve the public acceptance of these operations. (Author)

  14. Fuel assembly storage pool

    International Nuclear Information System (INIS)

    Hiranuma, Hiroshi.

    1976-01-01

    Object: To remove limitation of the number of storage of fuel assemblies to increase the number of storage thereof so as to relatively reduce the water depth required for shielding radioactive rays. Structure: Fuel assembly storage rack containers for receiving a plurality of spent fuel assembly racks are stacked in multi-layer fashion within a storage pool filled with water for shielding radioactive rays and removing heat. (Furukawa, Y.)

  15. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Butterfield, C.E.; Waite, E.

    1982-01-01

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  16. Spent fuel storage

    International Nuclear Information System (INIS)

    Huppert

    1976-01-01

    To begin with, the author explains the reasons for intermediate storage of fuel elements in nuclear power stations and in a reprocessing plant and gives the temperature and radioactivity curves of LWR fuel elements after removal from the reactor. This is followed by a description of the facilities for fuel element storage in a reprocessing plant and of their functions. Futher topics are criticality and activity control, the problem of cooling time and safety systems. (HR) [de

  17. Liquid fuel cells

    Directory of Open Access Journals (Sweden)

    Grigorii L. Soloveichik

    2014-08-01

    Full Text Available The advantages of liquid fuel cells (LFCs over conventional hydrogen–oxygen fuel cells include a higher theoretical energy density and efficiency, a more convenient handling of the streams, and enhanced safety. This review focuses on the use of different types of organic fuels as an anode material for LFCs. An overview of the current state of the art and recent trends in the development of LFC and the challenges of their practical implementation are presented.

  18. Nuclear fuel accounting

    International Nuclear Information System (INIS)

    Aisch, D.E.

    1977-01-01

    After a nuclear power plant has started commercial operation the actual nuclear fuel costs have to be demonstrated in the rate making procedure. For this purpose an accounting system has to be developed which comprises the following features: 1) All costs associated with nuclear fuel shall be correctly recorded; 2) it shall be sufficiently flexible to cover also deviations from proposed core loading patterns; 3) it shall be applicable to different fuel cycle schemes. (orig./RW) [de

  19. Nuclear fuel financing

    International Nuclear Information System (INIS)

    Lurf, G.

    1975-01-01

    Fuel financing is only at its beginning. A logical way of developing financing model is a step by step method starting with the financing of pre-payments. The second step will be financing of natural uranium and enrichment services to the point where the finished fuel elements are delivered to the reactor operator. The third step should be the financing of fuel elements during the time the elements are inserted in the reactor. (orig.) [de

  20. Alternative Fuels (Briefing Charts)

    Science.gov (United States)

    2009-06-19

    feedstock for HRJ, plant cost for F-T) Courtesy AFRL, Dr. Tim Edwards Unclassified • Agricultural crop oils (canola, jatropha, soy, palm , etc...Fuels Focus  Various conversion processes  Upgraded to meet fuel specs Diverse energy sources Petroleum Crude Oil Petroleum based Single Fuel in the...data and resources – Conduct gap analysis – synfuel efforts, expand to biofuels, ID potential joint efforts – Increase visibility outside SCP world