WorldWideScience

Sample records for annular fuel elements

  1. FRANCO, Finite Element Method (FEM) Fuel Rod Analysis for Solid and Annular Configurations

    International Nuclear Information System (INIS)

    1 - Description of program or function: The FRANCO code is a quasi- static two-dimensional fuel rod analysis code, that calculates the fuel temperature and material deformation as a function of heat generation rate. Both solid and annular fuel configurations are modeled. 2 - Method of solution: FRANCO uses two-dimensional finite element theory and applications for mechanical deformation and heat conduction, and determines the temperature distribution from the fuel center to the coolant adjacent to the clad at a position along the fuel rod axis. FRANCO calculates the average temperature of each radial division, the nodal displacement, and strain and stress within the fuel pellet and clad. The principal stresses, which represent maximum and minimum stresses within an element, result from Mohr's circle relationship between normal stresses. FRANCO is capable of predicting the thermo-mechanical behavior in the radial direction of a single fuel rod for both boiling water reactors (BWR's) and pressurized water reactors (PWR's). The cross sectional plane geometry of fuel rod is modeled using three-node constant strain triangular finite elements, and both thermal and mechanical solutions are computed with the same finite element configurations. The local linear heat generation rate is modeled as a uniform heat source in a fuel pellet, and the coolant temperature and heat transfer coefficient are applied as known boundary conditions at the boundary of the cladding surface. The total load to form the global force vector consists of the thermal load that results from thermal expansion of the material and the mechanical load exerted by pressure. FRANCO assumes the fuel-cladding gap region to be conductive material in order to simplify the analysis, and this gap is simulated by either an open gap or a closed gap model. A time- dependent problem can be simulated by FRANCO using quasi-static analysis when time-dependent parameters are provided. FRANCO can treat a steady-state or

  2. KNK II third core: design report for the annular fuel elements on the central position to accommodate material test inserts NZ 402 and NZ 403

    International Nuclear Information System (INIS)

    Since August 1984 irradiation experiments with temperature controlled pressure tube probes are being performed in the central position of KNK II. This is part of a long-term experimental program for the development of irradiation resistant reactor materials, which shall also be continued in the third core. The necessary irradiation channel is provided by a special annular fuel element. The present report describes the annular fuel elements for the third core. Aspects of the subassembly design are considered on the basis of the annular element design for the second core and the standard elements of the third core. Two annular elements NZ 402 and NZ 403 (as reserve) are available. It is demonstrated that the expected loadings will allow an unperturbed operation of the annular elements on the central position of the third core

  3. Thermal-hydraulic design calculations for the annular fuel element with replaceable test bundles (TOAST) on the test zone position 205 of KNK II/3

    International Nuclear Information System (INIS)

    Annular fuel elements are foreseen in KNK II as carrier elements for irradiation inserts and test bundles. For the third core a reloadable annular element on position 205 is foreseen, in which replaceable 19-pin test bundles (TOAST) shall be irradiated. The present report deals with the thermal-hydraulic design of the annular carrier element and the test bundle, whereby the test bundle required additional optimization. The code CIA has been used for the calculations. Start of irradiation of the subassembly is planned at the beginning of the third core operation. After optimization of the pin-spacer geometry in the test bundle, design calculations for both bundles were performed, whereby thermal coupling between both was taken into account. The calculated mass-flows and temperature distributions are given for the nominal and the eccentric element configuration. The calculated bundle pressure losses have been corrected according to experimental results

  4. Fluidic Analysis in an Annular Centrifugal Contactor for Fuel Reprocessing

    International Nuclear Information System (INIS)

    An annular centrifugal contactor (ACC) is a promising device for fuel reprocessing process, because it offers several advantages—a smaller size, a smaller holdup volume, and a higher separation performance—over conventional contactors such as a mixer-settler and a pulse column. Fluid dynamics and dispersion in an ACC, which has a combined mixer/centrifuge structure, are closely related to its separation performance and capacity, and this information is useful in improving equipment design. In this paper, experimental and computational fluid dynamics (CFD) studies were conducted to analyze fluidic and dispersion behavior in ACCs. Multiphase mixing (water/TBP-dodecane/air) in the annular zone was observed by Particle Imaging Velocimetry, and the change in the fluidic and dispersion behavior was ascertained under several operational conditions. The results of the CFD studies, which considered multiphase turbulent flow in the annular and rotor interior zones, were in a good agreement with the experimental data. (author)

  5. An Evaluation of the Annular Fuel and Bottle-Shaped Fuel Concepts for Sodium Fast Reactors

    OpenAIRE

    Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

    2010-01-01

    Two innovative fuel concepts, the internally and externally cooled annular fuel and the bottle-shaped fuel, were investigated with the goal of increasing the power density and reduce the pressure drop in the sodium-cooled fast reactor, respectively. The concepts were explored for both high- and low-conversion core configurations, and metal and oxide fuels. The annular fuel concept is best suited for low-conversion metal-fuelled cores, where it can enable a power uprate of ~20%; the magnitude ...

  6. Fuel element design handbook

    Energy Technology Data Exchange (ETDEWEB)

    Merckx, K.R.

    1958-09-01

    The economic development of nuclear reactors depends upon the integrated progress in the fields of reactor design, fuel element design, reactor operation, and fuel production and separation. Broad criteria, which restrict the fuel element design, are determined by the mutual consideration of the problems encountered in all the above fields. Hence, no stage of reactor design or operation is independent of the fuel element problem, nor can the fuel element designer disregard the interest of any one field. As an introduction to the fuel element design problem, this chapter describes how the general criteria for a fuel element are determined.

  7. Comparison of the Characteristics of Solid Type and Annular Type Nuclear Fuels Using Thermoelastic-Plastic-Creep FEM

    Directory of Open Access Journals (Sweden)

    Young-Doo Kwon

    2016-01-01

    Full Text Available The purpose of this study is to compare the characteristics of two types of nuclear fuel using the finite element program of thermoelastic-plastic-creep analysis. The analyzed fuel rods are of two types, solid and annular ones, and their thermomechanical characteristics are compared. Thermoelastic-plastic-creep analyses were made using an in-house finite element analysis program that adopts the “effective-stress-function” algorithm. The temperature-dependent material properties, which were obtained from the experiments for actual nuclear reactors, are adopted. The effects of type of fuel systems are revealed in both stresses and temperature distributions. The maximum tensile and compressive hoop stress of pellet and cladding are monitored to evaluate the mechanical behavior, and the maximum temperature is used to evaluate the thermal behavior. Although the annular type of fuel has certain disadvantage, it would be used very effectively or safely in future nuclear power plants.

  8. Fabrication of Annular Pellet for HANARO Irradiation Test of Dual Cooled Fuel

    International Nuclear Information System (INIS)

    One of the most important components in a Pressurized Water Reactor affecting its safety and economy is a nuclear fuel. The traditional PWR fuel pellet has a shape of cylindrical tablets of about 8 mm in diameter with a chamfer and dishes. A significant reduction in its failure rate has resulted from the improvements in the fuel and cladding quality. Enhanced fuel assembly design allowed appreciable power density increases. However, it is difficult to achieve a significant increase of a power density under the current fuel pin design. An internally and externally cooled annular fuel has been considered seriously as a promising solution for an extended power uprate of a PWR fuel assembly. A dual cooled annular fuel shows a lot of advantages from the point of a fuel safety and its economy due to its unique configurational merit such as an increased heat transfer area and a thin pellet thickness. There must be a lot of considerations in the various fields to introduce an annular internally and externally cooled fuel to commercial PWR reactors. The dimensional changes of the annular fuel pellets during the early irradiation stage are very important, because they have an influence on the size of the gap between the pellet and the inner/outer claddings. In order to gain an insight to how the annular pellets deform, a HANARO irradiation test is planned for annular pellets with 5 different types. The detailed specification of the annular pellet was shown in Table 1. It is noted that Type C has the same pore structure as a commercial PWR pellet. The purpose of this paper is to report on the manufacturing process of an annular fuel pellet for a HANARO irradiation test

  9. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  10. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    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)

  11. NEUTRONIC REACTOR FUEL ELEMENT

    Science.gov (United States)

    Picklesimer, M.L.; Thurber, W.C.

    1961-01-01

    A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

  12. Fuel Element Technical Manual

    Energy Technology Data Exchange (ETDEWEB)

    Burley, H.H. [ed.

    1956-08-01

    It is the purpose of the Fuel Element Technical Manual to Provide a single document describing the fabrication processes used in the manufacture of the fuel element as well as the technical bases for these processes. The manual will be instrumental in the indoctrination of personnel new to the field and will provide a single data reference for all personnel involved in the design or manufacture of the fuel element. The material contained in this manual was assembled by members of the Engineering Department and the Manufacturing Department at the Hanford Atomic Products Operation between the dates October, 1955 and June, 1956. Arrangement of the manual. The manual is divided into six parts: Part I--introduction; Part II--technical bases; Part III--process; Part IV--plant and equipment; Part V--process control and improvement; and VI--safety.

  13. Nuclear fuel element

    Science.gov (United States)

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

    A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

  14. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    The grid-shaped spacer for PWR fuel elements consists of flat, upright metal bars at right angles to the fuel rods. In one corner of a grid mesh it has a spring with two end parts for the fuel rod. The cut-outs for the end parts start from an end edge of the metal bar parallel to the fuel rods. The transverse metal bar is one of four outer metal bars. Both end parts of the spring have an extension parallel to this outer metal arm, which grips a grid mesh adjacent to this grid mesh at the side in one corner of the spacer and forms an end part of a spring for the fuel rod there on the inside of the outer metal bar. (HP)

  15. Diametric Tolerance Control of Dual Cooled Annular Fuel Pellet without Inner Surface Grinding

    International Nuclear Information System (INIS)

    A dual cooled fuel consists of internal and external cladding tubes in which annular pellets are stacked and cooling water flows in both internal and external coolant passages. It is recently being reconsidered as a promising option for a power up-rate of a pressurized water reactor fuel assembly because an annular fuel shows a lot of advantages from the point of a fuel safety and its economy due to an increased heat transfer area and a thin pellet thickness. Many technical issues might cause a serious problem to adopt the dual cooled annular fuel to the commercial PWR reactors. One of the most important issues is a heat flux split toward an internal cladding and an external cladding due to the gap conductance asymmetry which results from a preferential expansion of a fuel pellet toward the outside during an irradiation. Gap conductance is directly related to the inner and outer gap thicknesses. Initial gap thicknesses can vary with a pellet's dimensions which are affected by a reactor operation condition. Recently, it is suggested that a fuel rod with a smaller inner gap and a larger outer gap can reduce this gap conductance asymmetry. This approach can be effective only after precise tolerance technology is achieved. Because of an inhomogeneous green density distribution along the compact height, an hour-glassing usually occurred in a sintered cylindrical PWR fuel pellet fabricated by a conventional double-acting press. Thus, a sintered pellet usually undergoes a center-less grinding process in order to secure a pellet's specifications. In the case of an annular pellet fabrication using a conventional double-acting press, the same hour-glass shape would probably occur. The outer diameter tolerance of an annular pellet can be controlled easily similar to that of a conventional cylindrical PWR pellet through a center-less grinding. However, it appears not to be simple in the case of an inner surface grinding. It would be the best way to satisfy the specifications

  16. Upgraded HFIR Fuel Element Welding System

    Energy Technology Data Exchange (ETDEWEB)

    Sease, John D [ORNL

    2010-02-01

    The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. In recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.

  17. ThO{sub 2}-UO{sub 2} annular pins for high burnup fuels

    Energy Technology Data Exchange (ETDEWEB)

    Caner, Marc; Dugan, Edward T

    2000-06-01

    The main purpose of this work is to investigate the use of annular fuel pins (particularly pins containing thorium dioxide) for high burnup fuel. The following parameters were evaluated and compared between postulated mixed thorium-uranium dioxide standard and annular (9% void fraction) type fuel assemblies, as a function of burnup: the infinite multiplication factor, the uranium and plutonium isotopic compositions, the fuel temperature coefficient of reactivity and the conversion ratio. We used the SCALE-4.3 code system. The calculation method consisted in obtaining actinide and fission product number densities as functions of assembly burnup, by means of a 1-D transport calculation combined with a 0-D burnup calculation. These number densities were then used in a 3-D Monte Carlo code for obtaining k{sub {infinity}} from two-dimensional-symmetry 'snapshots'.

  18. Fuel distribution measurements in a model low NOx double annular combustor using laser induced fluorescence

    OpenAIRE

    Lockett, R. D.; Greenhalgh, D.A.

    2010-01-01

    Planar laser induced fluorescence (PLIF) was employed in a three sector, low NOx double annular combustor in order to characterise the combusting fuel spray. Naphthalene was employed as a fluorescent agent in odourless kerosene in order to determine the behavior of the light fractions in the fuel vapour, and the light to medium fractions in the fuel spray, while 2,5 di-phenyl oxizol (ppo) was employed to determine the behavior of the heavy fractions in the fuel spray. Counter-swirl air blast ...

  19. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    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)

  20. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    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)

  1. Nuclear reactor fuel element splitter

    International Nuclear Information System (INIS)

    A method and apparatus are disclosed for removing nuclear fuel from a clad fuel element. The fuel element is power driven past laser beams which simultaneously cut the cladding lengthwise into at least two longitudinal pieces. The axially cut lengths of cladding are then separated, causing the nuclear fuel contained therein to drop into a receptacle for later disposition. The cut lengths of cladding comprise nuclear waste which is disposed of in a suitable manner. 6 claims, 10 drawing figures

  2. Spacer for a fuel element

    International Nuclear Information System (INIS)

    Spacers for fuel pins arranged to form congish fuel elements can be shaped as plates with openings in accordance with the fuel pin grid. Such a plate that covers the cross section of a fuel element consists according to the invention of at least two parts that are offset in the fuel element's longitudinal direction and joint hinge-like in at least one grid position. Thus, one has smaller parts that are easier to work on with due accuracy. The invention is designed in particular for breeder reactors and high-conversion reactors. (orig.)

  3. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

    Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

  4. Fuel element for nuclear reactor

    International Nuclear Information System (INIS)

    In order to avoid a can box or an adjacent fuel element sitting on the spacer of a fuel element in the corner during assembly, the top and bottom edges of the outer bars of the spacers are provided with deflector bars, which have projections projecting beyond the outside of the outer bars. (orig.)

  5. Increased burnup of fuel elements

    International Nuclear Information System (INIS)

    The specialists' group for fuel elements of the Kerntechnische Gesellschaft e.V. held a meeting on ''Increased Burnup of Fuel Elements'' on 9th and 10th of November 1982 at the GKSS Research Center Geesthacht. Most papers dealt with the problems of burnup increase of fuel elements for light water reactors with respect to fuel manufacturing, power plant operation and reprocessing. Review papers were given on the burnup limits for high temperature gas cooled reactors and sodium fast breeder reactors. The meeting ended with a presentation of the technical equipment of the hot laboratory of the GKSS and the programs which are in progress there. (orig.)

  6. Testing device for fuel element samples

    International Nuclear Information System (INIS)

    The device described is for testing samples for behavior at high temperature in heavy gamma radiation. The whole device is designed to be maintained in the high neutron flux of a nuclear reactor channel. It comprises two co-axial envelopes with cylindrical side walls and with convex truncated bottom and head walls, these truncated walls being maintained in pairs at a small distance and as constant as possible owing to the inner envelope being designed to accept the fuel element or other sample for testing and to be connected to an intake pipe and a return pipe for a sample environmental gas. The truncated head wall of the outer envelope is joined by a sealed thermal expansion bellows to the cylindrical wall of this same envelope. The restricted annular space between the inner envelope and the outer envelope with its bellows is designed to be coupled to an intake pipe and a return pipe for a variable thermal conductivity gas

  7. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    Energy Technology Data Exchange (ETDEWEB)

    Knight, R.W.; Morin, R.A.

    1999-12-01

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

  8. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    International Nuclear Information System (INIS)

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U3O8 powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated

  9. The Design and Manufacturing Report of Non-Instrumented Rig for Dual-cooled Annular Fuel Irradiation Test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Bang, Je Geon; Lim, Ik Sung; Kim, Sun Ki; Yang, Yong Sik; Song, Kun Woo; Seo, Chul Gyo; Park, Chan Kook

    2008-09-15

    This project is preparing to irradiation test of the developed double cooled annular fuel pellet in HANARO for pursuit advanced performance in High Performance Fuel Technology Development as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented rig designed and manufactured for irradiation test in HANARO OR hole. This non- instrumented rig was confirmed the compatibility of HANARO and the integrity of rig structure, and satisfied the quality assurance requirements. This non- instrumented rig is adopt to the irradiation test for double cooled annular fuel pellet in HANARO.

  10. Fuel element development

    International Nuclear Information System (INIS)

    In capsule irradiation tests the influence was studied which is exerted by high power densities on thin oxide fuel rods. Cladding expansions have been observed which are not attributable to creep but to plastic strains. Power jumps during load cycling resulted in stress to the cladding through fuel pressure due to thermal differential strain. - Changes in geometry of oxide fuel pellets during cycling were investigated theoretically using models. The test group 5b was also studied with a view to plutonium redistribution. A very high plutonium enrichment was found at the central channel, and outer zones nearly free from plutonium soon after the beginning of irradiation, which might be due to the high specific power and central temperature and the high PuO2-content (35%) of the fuel. Two contributions include as subjects the porosity of fuel in the context of structural analyses and creep caused by irradiation. The plutonium content itself does not seem to increase substantially the creep rate. Further results of post-examinations are available from the oxide irradiation tests Mol-7B and DFR-435. The zone of maximum damage of the Mol-7B-rods occurs at the upper end of the fuel column; even here the structure of the rod has essentially remained unchanged. The amount of fuel escaping is not as great as at the damaged points of DFR-435. (orig.)

  11. Annular Flow Distribution test

    International Nuclear Information System (INIS)

    This report documents the Babcock and Wilcox (B ampersand W) Annular Flow Distribution testing for the Savannah River Laboratory (SRL). The objective of the Annular Flow Distribution Test Program is to characterize the flow distribution between annular coolant channels for the Mark-22 fuel assembly with the bottom fitting insert (BFI) in place. Flow rate measurements for each annular channel were obtained by establishing ''hydraulic similarity'' between an instrumented fuel assembly with the BFI removed and a ''reference'' fuel assembly with the BFI installed. Empirical correlations of annular flow rates were generated for a range of boundary conditions

  12. Apparatus for locating defective nuclear fuel elements

    International Nuclear Information System (INIS)

    An ultrasonic search unit for locating defective fuel elements within a fuel assembly used in a water cooled nuclear reactor is presented. The unit is capable of freely traversing the restricted spaces between the fuel elements

  13. Compact Fuel Element Environment Test

    Science.gov (United States)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.; Broadway, J. W.

    2012-01-01

    Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that employ high melting point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high-temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

  14. Auto-thermal combustion of lean gaseous fuels utilizing a recuperative annular double-layer catalytic converter

    Energy Technology Data Exchange (ETDEWEB)

    Budzianowski, W.M. [Wroclaw Univ. of Technology, Wroclaw (Poland). Div. of Chemical and Biochemical Processes, Faculty of Chemistry; Miller, R. [Wroclaw Univ. of Technology, Wroclaw (Poland). Inst. of Power Engineering and Fluid Mechanics, Faculty of Mechanical and Power Engineering

    2008-08-15

    This study investigated the auto-thermal combustion of lean gaseous fuels in a recuperative annular double-layer catalytic converter. An analysis of the stationary and transient performance of annular converters was presented. The feasibility of lean gaseous mixture combustion in auto-thermally operated recuperative annular double-layer catalytic converters was investigated. The aim of the study was to build a process model using mass, energy, and momentum differential balances. The model was used to study the static behaviour of a recuperative annular double layer converter; an annular converter operating in transient conditions; and energy accumulation and recuperation interactions. The effects of fuel temperature, external cooling, and fuel concentration were examined. Results of the study showed that a substantial reduction of the inter- and intra-phase resistances to mass and heat transfer was obtained. It was demonstrated that the use of a low value for the substrate's thermal conductivity accelerated ignition and retarded extinction. The recuperative converter was able to transfer short-time inlet disturbances of various parameters due to the energy accumulation and temporal reversed recuperation which counteracted destructive overheating of the catalysts. The stability analyses showed stable and unstable branches of solutions for the different parameters of the recuperative converter. 21 refs., 1 tab., 23 figs.

  15. Experimental and numerical investigation of an entrance blockage of an inner channel in dual-cooled annular nuclear fuel

    International Nuclear Information System (INIS)

    A dual-cooled annular nuclear fuel for a Pressurized Water Reactor (PWR) has been introduced for a significant increase in reactor power. The Korea Atomic Energy Research Institute (KAERI) has been researching the development of a dual-cooled annular fuel for a power increase in an optimized PWR in Korea, OPR-1000. The main advantage of a dual-cooled annular fuel is an increased heat transfer area and a reduction in the fuel temperature, which would result in reduced fission gas release and increased fuel melting margin and Departure from Nucleate Boiling (DNB) margin. The annular fuel rod is configured to allow the coolant flow through the inner channel as well as outer channel. Since the inner channel is isolated from the neighbor channels unlike the outer channels, an inner channel blockage is one of the principal technical issues of a dual-cooled annular fuel. Due to a partial blockage, the inner channel may be faced with a DNB accident. A conceptual design used to complement the entrance blockage of an inner channel was suggested by KAERI. The through holes in this design are formed on a cylindrical wall of the lower end plug. When the inner channel is blocked by debris, coolant for the inner channel will be supplied through the side holes. But due to very unusual shape of the lower end plug, it is difficult to estimate the flow resistance of the side flow holes using empirical correlations available in the open literatures. Experimental and Computational Fluid Dynamics (CFD) studies were performed to investigate the bypass flow through the side holes of the lower end plug to complement the entrance blockage of an inner channel. The form loss coefficient in the side holes was also estimated by using the pressure drop along the bypass flow path and DNB Ratio (DNBR) margin was estimated by a subchannel analysis code. (author)

  16. Power pulse tests on CANDU type fuel elements in TRIGA reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensor for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (author)

  17. Transportation of irradiated fuel elements

    International Nuclear Information System (INIS)

    The report falls under the headings: introduction (explaining the special interest of the London Borough of Brent, as forming part of the route for transportation of irradiated fuel elements); nuclear power (with special reference to transport of spent fuel and radioactive wastes); the flask aspect (design, safety regulations, criticisms, tests, etc.); the accident aspect (working manual for rail staff, train formation, responsibility, postulated accident situations); the emergency arrangements aspect; the monitoring aspect (health and safety reports); legislation; contingency plans; radiation - relevant background information. (U.K.)

  18. A high temperature fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Sekido, A.; Nakai, M.; Ninomiya, Y.

    1982-12-21

    A solid electrolyte which conducts electricity with heating by oxygen ions and operates at a temperature of 1,000C is used in the element. The cathode, besides the ionic conductivity in oxygen, has an electron conductivity. The anode has electron conductivity. Substances such as Bi203, into which oxides of alkaline earth metals are added, are used for making the cathode. The electrolyte consists of ZrO2 and Y2O3, to which CaO is added. WC, to which an H2 type fuel is fed, serves as the anode. The element has a long service life.

  19. Effect of moderator density distribution of annular flow on fuel assembly neutronic characteristics in boiling water reactor cores

    International Nuclear Information System (INIS)

    The effect of the moderator density distribution of annular flow on the fuel assembly neutronic characteristics in a boiling water nuclear reactor was investigated using the SRAC95 code system. For the investigation, a model of annular flow for fuel assembly calculation was utilized. The results of the assembly calculation with the model (Method 1) and those of the fuel assembly calculation with the uniform void fraction distribution (Method 2) were compared. It was found that Method 2 underestimates the infinite multiplication factor in the fuel assembly including the gadolinia rod (type 1 assembly). This phenomenon is explained by the fact that the capture rate in the thermal energy region in gadolinia fuel is estimated to be smaller when the liquid film of annular flow at the fuel rod surface is considered. A burnup calculation was performed under the condition of a void fraction of 65% and a volumetric fraction of the liquid film in liquid phase of 1. It is found that Method 2 underestimates the infinite multiplication factor in comparison to Method 1 in the early stage of burnup, and that Method 2 becomes to overestimate the factor after a certain degree of burnup. This is because Method 2 overestimates the depletion rate of the gadolinia. (author)

  20. Fuel elements of thermionic converters

    Energy Technology Data Exchange (ETDEWEB)

    Hunter, R.L. [ed.] [Sandia National Labs., Albuquerque, NM (United States). Environmental Systems Assessment Dept.; Gontar, A.S.; Nelidov, M.V.; Nikolaev, Yu.V.; Schulepov, L.N. [RI SIA Lutch, Podolsk (Russian Federation)

    1997-01-01

    Work on thermionic nuclear power systems has been performed in Russia within the framework of the TOPAZ reactor program since the early 1960s. In the TOPAZ in-core thermionic convertor reactor design, the fuel element`s cladding is also the thermionic convertor`s emitter. Deformation of the emitter can lead to short-circuiting and is the primary cause of premature TRC failure. Such deformation can be the result of fuel swelling, thermocycling, or increased unilateral pressure on the emitter due to the release of gaseous fission products. Much of the work on TRCs has concentrated on preventing or mitigating emitter deformation by improving the following materials and structures: nuclear fuel; emitter materials; electrical insulators; moderator and reflector materials; and gas-exhaust device. In addition, considerable effort has been directed toward the development of experimental techniques that accurately mimic operational conditions and toward the creation of analytical and numerical models that allow operational conditions and behavior to be predicted without the expense and time demands of in-pile tests. New and modified materials and structures for the cores of thermionic NPSs and new fabrication processes for the materials have ensured the possibility of creating thermionic NPSs for a wide range of powers, from tens to several hundreds of kilowatts, with life spans of 5 to 10 years.

  1. Nuclear reactor fuel element. Kernreaktorbrennelement

    Energy Technology Data Exchange (ETDEWEB)

    Lippert, H.J.

    1985-03-28

    The fuel element box for a BWR is situated with a corner bolt on the inside in one corner of its top on the top side of the top plate. This corner bolt is screwed down with a bolt with a corner part which is provided with leaf springs outside on two sides, where the bolt has a smaller diameter and an expansion shank. The bolt is held captive to the bolt head on the top and the holder on the bottom of the corner part. The holder is a locknut. If the expansion forces are too great, the bolt can only break at the expansion shank.

  2. Direct imaging of light elements by annular dark-field aberration-corrected scanning transmission electron microscopy

    Energy Technology Data Exchange (ETDEWEB)

    Lotnyk, Andriy, E-mail: andriy.lotnyk@iom-leipzig.de; Poppitz, David; Gerlach, Jürgen W.; Rauschenbach, Bernd [Leibniz Institute of Surface Modification, Permoserstr. 15, D-04318 Leipzig (Germany)

    2014-02-17

    In this report, we show that an annular dark-field detector in an aberration-corrected scanning transmission electron microscope allows the direct observation of light element columns in crystalline lattices. At specific imaging conditions, an enhancement of the intensities of light element columns in the presence of heavy element columns is observed. Experimental results are presented for imaging the nitrogen and carbon atomic columns at the GaN-SiC interface and within the GaN and SiC compounds. The crystal polarity of GaN at the interface is identified. The obtained findings are discussed and are well supported by image simulations.

  3. Development of dual cooled annular fuel and its possibility to enhance both economy and safety of light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yanghyun; Kim, Keonsik; Park, Jeongyong; Yang, Yongsik; Kim, Hyungkyu; In, Wangkee; Song, Kunwoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-03-15

    Over the past few decades, extensive studies have been performed to improve the reliability and safety of light water reactor (LWR) fuel. In recent years, power updating of about 10% is being achieved by modifying safety analysis methodology and subsequent increase in safety margin. But departure from nucleate boiling (DNB) and loss of coolant accident (LOCA) are still two of the most important limiting factors which would restrict power updating more than 10%. Duel cooled annular fuel, cooled in both internal and external cooling channel, has advantages of considerably lower heat flux and lower fuel temperature than conventional solid fuel. While lower heat flus gives higher DNB margin for the same power retie, lower temperature reduces the stored energy of fuel. However, there are many technical issues that should be addressed before any new type of fuel can be considered for application to LWR. This paper describes the key technologies that Korea Atomic Energy Research Institute (KAERI) has developed for dual cooled annular fuel and discusses the feasibility of its application to LWR.

  4. Development of dual cooled annular fuel and its possibility to enhance both economy and safety of light water reactor

    International Nuclear Information System (INIS)

    Over the past few decades, extensive studies have been performed to improve the reliability and safety of light water reactor (LWR) fuel. In recent years, power updating of about 10% is being achieved by modifying safety analysis methodology and subsequent increase in safety margin. But departure from nucleate boiling (DNB) and loss of coolant accident (LOCA) are still two of the most important limiting factors which would restrict power updating more than 10%. Duel cooled annular fuel, cooled in both internal and external cooling channel, has advantages of considerably lower heat flux and lower fuel temperature than conventional solid fuel. While lower heat flus gives higher DNB margin for the same power retie, lower temperature reduces the stored energy of fuel. However, there are many technical issues that should be addressed before any new type of fuel can be considered for application to LWR. This paper describes the key technologies that Korea Atomic Energy Research Institute (KAERI) has developed for dual cooled annular fuel and discusses the feasibility of its application to LWR

  5. Automated Fuel Element Closure Welding System

    International Nuclear Information System (INIS)

    The Automated Fuel Element Closure Welding System is a robotic device that will load and weld top end plugs onto nuclear fuel elements in a highly radioactive and inert gas environment. The system was developed at Argonne National Laboratory-West as part of the Fuel Cycle Demonstration. The welding system performs four main functions, it (1) injects a small amount of a xenon/krypton gas mixture into specific fuel elements, and (2) loads tiny end plugs into the tops of fuel element jackets, and (3) welds the end plugs to the element jackets, and (4) performs a dimensional inspection of the pre- and post-welded fuel elements. The system components are modular to facilitate remote replacement of failed parts. The entire system can be operated remotely in manual, semi-automatic, or fully automatic modes using a computer control system. The welding system is currently undergoing software testing and functional checkout

  6. Spacer for supporting fuel element boxes

    International Nuclear Information System (INIS)

    A spacer plate unit arranged externally on each side and at a predetermined level of a polygonal fuel element box for mutually supporting, with respect to one another, a plurality of the fuel element boxes forming a fuel element bundle, is formed of a first and a second spacer plate part each having the same length and the same width and being constituted of unlike first and second materials, respectively. The first and second spacer plate parts of the several spacer plate units situated at the predetermined level are arranged in an alternating continuous series when viewed in the peripheral direction of the fuel element box, so that any two spacer plate units belonging to face-to-face oriented sides of two adjoining fuel element boxes in the fuel element bundle define interfaces of unlike materials

  7. Gamma spectrometry of TRIGA fuel elements

    International Nuclear Information System (INIS)

    The burnupt of 19 TRIGA fuel elements was determined by gamma spectrometry using a special fuel element holder developed and constructed at the Atom Institute, Vienna. The investigated fuel element is kept in a horizontal position about 4 m below the reactor pool water surface. A collimator tube extends to the reactor platform where an intrinsic Ge-detector is located. With this system each fuel element was investigated at eight equidistant points along its active zone and the Cs 137 activity was evaluated. (orig.)

  8. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  9. International experience in conditioning spent fuel elements

    International Nuclear Information System (INIS)

    The purpose of this report is to compile and present in a clear form international experience (USA, Canada, Sweden, FRG, UK, Japan, Switzerland) gained to date in conditioning spent fuel elements. The term conditioning is here taken to mean the handling and packaging of spent fuel elements for short- or long-term storage or final disposal. Plants of a varying nature fall within this scope, both in terms of the type of fuel element treated and the plant purpose eg. experimental or production plant. Emphasis is given to plants which bear some similarity to the concept developed in Germany for direct disposal of spent fuel elements. Worldwide, however, relatively few conditioning plants are in existence or have been conceived. Hence additional plants have been included where aspects of the experience gained are also of relevance eg. plants developed for the consolidation of spent fuel elements. (orig./HP)

  10. Combustion of liquid fuel in the counter-swirled jets of a gas turbine plant annular combustion chamber

    Science.gov (United States)

    Tumanovskii, A. G.; Semichastnyi, N. N.; Sokolov, K. Iu.

    1986-03-01

    Tests were carried out on an annular combustion chamber rig with a stabilizer of the type used in the GTN-25 gas turbine plant to determine the feasibility of burning a liquid fuel (diesel fuel, GOST 4749-73) in a combustion chamber of this type. Very high performance was obtained for a number of important characteristics of the microflame combustion process in counterswirled jets where all the air was supplied through the front unit of the chamber. However, the tests did not make it possible to solve some of the problems which arise when operating under full-scale conditions, such as the required high combustion efficiency under variable operating conditions of a gas turbine plant; elimination of soot formation at the walls of the stabilizer and the internal surfaces of the pipes supplying fuel to the atomizers; and a decrease in smoking under conditions of excess air factor.

  11. Nondestructive examination of TRIGA reactor fuel elements

    International Nuclear Information System (INIS)

    Neutron radiography has proved to be a very useful method for nondestructive examination of used and nonused reactor elements. The method can be used for determination of homogenity and burn-up of fuel and burnable poisons, for detection of fuel and full clad damage and taking into account the capability to perform accurate geometrical measurements it is also possible to assess mechanical deformations of fuel elements. Active fuel elements of TRIGA reactor have been examined for deformations and fuel clad damage. In the course of these investigations the following methods were tested and compared: - transfer neutronradiographic techniques using In and Dy converter screens, - direct neutrongraphic method using solid state track detectors, - X-ray radiography employing lead shielding masks and highly selective photographic material. Considerable information on the burn-up of reactor fuel elements can be obtained from measuring the distribution of radioactive isotopes in the fuel element by gamma ray spectroscopy. For a used TRIGA fuel element the axial distribution of the isotope Cs-137 has been measured and the burn-up determined. We compare the experimental results with a crude estimate of burn-up

  12. Modelling the oxidation of defected fuel elements

    International Nuclear Information System (INIS)

    Interim dry storage of used fuel is an economical alternative to storage in water pools. The fuel must remain intact during the dry-storage period, otherwise future handling of the fuel will be expensive. Oxidation of defected fuel elements can lead to fuel disintegration. Thus it is important to be able to predict the extent of oxidation of defected fuel elements in a dry-storage facility. In this report, a model is developed for predicting the extent or rate of oxidation of defected fuel elements stored at temperatures up to 170 C. The model employs equivalent porous medium representation of the fuel and described the oxygen concentration in the fuel element using a reaction-diffusion equation. The one- and two-dimensional reaction-diffusion equations are solved on the assumption that the oxygen-fuel reaction is either zeroth or first order in the oxygen concentration. Dimensional analysis of the model equations shows that the solution depends explicitly on a single parameter p. The value of p can be calculated using data from the literature, or it can be estimated from the results of the CEX-1 experiments being carried out at Whiteshell Laboratories. The value of p, estimated from the CEX-1 results, is more than two orders of magnitude larger than the value of p calculated from literature data. Although some reasons for this large difference are suggested, further work is needed to resolve this discrepancy. (author). 16 refs., 2 tabs., 11 figs

  13. MRT fuel element inspection at Dounreay

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, J.

    1997-08-01

    To ensure that their production and inspection processes are performed in an acceptable manner, ie. auditable and traceable, the MTR Fuel Element Fabrication Plant at Dounreay operates to a documented quality system. This quality system, together with the fuel element manufacturing and inspection operations, has been independently certified to ISO9002-1987, EN29002-1987 and BS5750:Pt2:1987 by Lloyd`s Register Quality Assurance Limited (LRQA). This certification also provides dual accreditation to the relevant German, Dutch and Australian certification bodies. This paper briefly describes the quality system, together with the various inspection stages involved in the manufacture of MTR fuel elements at Dounreay.

  14. Method for inspecting nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A technique for disassembling a nuclear reactor fuel element without destroying the individual fuel pins and other structural components from which the element is assembled is described. A traveling bridge and trolley span a water-filled spent fuel storage pool and support a strongback. The strongback is under water and provides a working surface on which the spent fuel element is placed for inspection and for the manipulation that is associated with disassembly and assembly. To remove, in a non-destructive manner, the grids that hold the fuel pins in the proper relative positions within the element, bars are inserted through apertures in the grids with the aid of special tools. These bars are rotated to flex the adjacent grid walls and, in this way relax the physical engagement between protruding portions of the grid walls and the associated fuel pins. With the grid structure so flexed to relax the physical grip on the individual fuel pins, these pins can be withdrawn for inspection or replacement as necessary without imposing a need to destroy fuel element components

  15. Fundamental aspects of nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Olander, D.R.

    1976-01-01

    The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO/sub 2/, fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO/sub 2/, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies. (DG)

  16. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

    Some time ago the CCHEN authorities decided to produce a set of 50 low enrichment fuel elements. These elements were produced in the PEC (Fuel Elements Plant), located at CCHEN offices in Lo Aguirre. These new fuel elements have basically the same geometrical characteristics of previous ones, which were British and made with raw material from the U.S. The principal differences between our fuel elements and the British ones is the density of fissile material, U-235, which was increased to compensate the reduction in enrichment. Last year, the Fuel Elements Plant (PEC) delivered the shipment's first four (4) fuel elements, called leaders, to the RECH1. A test element was delivered too, and the complete set was introduced into the reactor's nucleus, following the normal routine, but performing a special follow-up on their behavior inside the nucleus. This experimental element has only one outside fuel plate, and the remaining (15) structural plates are aluminum. In order to study the burnup, the test element was taken out of the nucleus, in mid- November 1999, and left to decay until June 2000, when it was moved to the laboratory (High Activity Cell), to start the burnup measurements, with a gamma spectroscopy system. This work aims to show the results of these measurements and in addition to meet the following objectives: (a) Visual test of the plate's general condition; (b) Sipping test of fission products; (c) Study of burn-up distribution in the plate; (d) Check and improve the calculus algorithm; (e) Comparison of the results obtained from the spectroscopy with the ones from neutron calculus

  17. Apparatus and method for assembling fuel elements

    International Nuclear Information System (INIS)

    A nuclear fuel element assembling method and apparatus is preferably operable under programmed control unit to receive fuel rods from storage, arrange them into axially aligned stacks of closely monitored length, and transfer the stacks of fuel rods to a loading device for insertion into longitudinal passages in the fuel elements. In order to handle large numbers of one or more classifications of fuel rods or other cylindrical parts, the assembling apparatus includes at least two feed troughs each formed by a pair of screw members with a movable table having a plurality of stacking troughs for alignment with the feed troughs and with a conveyor for delivering the stacks to the loading device, the fuel rods being moved along the stacking troughs upon a fluid cushion. 23 claims, 6 figures

  18. Fuel cladding tubes and fuel elements

    International Nuclear Information System (INIS)

    Purpose: To enable non-destructive measurement for the thickness of zirconium barriers. Constitution: Regions capable of non-destructive inspection are provided at the boundary between a fuel cladding tube made of zirconium alloy and the zirconium barrier lined to the inner circumference surface of the tube. As the regions being capable of distinguishing by ultrasonic wave reflection, solid materials, for example, non-metal materials different from that for the tube and the barrier are placed or gaps are provided at the boundary between the zirconium alloy cladding tube and the zirconium barrier. Since ultrasonic waves are reflected at each of the boundaries by the presence of these regions, thickness of the zirconium barrier can be measured in a non-destructive manner from either the inner or the outer surface of the tube. (Yoshino, Y.)

  19. Spring packed particle bed fuel element

    International Nuclear Information System (INIS)

    This patent describes a gas cooled particle bed nuclear fuel element. It comprises: a porous inner frit; a porous outer frit attached to the inner frit by an end cap t a first end and radially guided by a shoulder at a second end, forming an annulus between the frits; a fuel particle bed in the annulus; a first compressive device at each end of the annulus; and a second compressive device positioned in the annulus within the fuel particle bed

  20. HTGR fuel element size reduction system

    Energy Technology Data Exchange (ETDEWEB)

    Strand, J.B.; Cramer, G.T.

    1978-06-01

    Reprocessing of high-temperature gas-cooled reactor fuel requires development of a fuel element size reduction system. This report describes pilot plant testing of crushing equipment designed for this purpose. The test program, the test results, the compatibility of the components, and the requirements for hot reprocessing are discussed.

  1. HTGR fuel element size reduction system

    International Nuclear Information System (INIS)

    Reprocessing of high-temperature gas-cooled reactor fuel requires development of a fuel element size reduction system. This report describes pilot plant testing of crushing equipment designed for this purpose. The test program, the test results, the compatibility of the components, and the requirements for hot reprocessing are discussed

  2. Stability analysis of a liquid fuel annular combustion chamber. M.S. Thesis

    Science.gov (United States)

    Mcdonald, G. H.

    1979-01-01

    The problems of combustion instability in an annular combustion chamber are investigated. A modified Galerkin method was used to produce a set of modal amplitude equations from the general nonlinear partial differential acoustic wave equation. From these modal amplitude equations, the two variable perturbation method was used to develop a set of approximate equations of a given order of magnitude. These equations were modeled to show the effects of velocity sensitive combustion instabilities by evaluating the effects of certain parameters in the given set of equations. By evaluating these effects, parameters which cause instabilities to occur in the combustion chamber can be ascertained. It is assumed that in the annular combustion chamber, the liquid propellants are injected uniformly across the injector face, the combustion processes are distributed throughout the combustion chamber, and that no time delay occurs in the combustion processes.

  3. Developments in fabrication of annular MOX fuel pellet for Indian fast reactor

    International Nuclear Information System (INIS)

    Mechanical rotary presses along with adoption of core rod feature were inducted for fabrication of intricate annular Mixed Oxide (MOX) pellets for Prototype Fast Breeder Reactor (PFBR). In the existing tooling, bottom plungers contain core rod whereas top plungers contain a central hole for the entry of core rod during compaction. Frequent manual clean up of top plungers after few operations were required due to settling of powder in the annular hole of top plungers during compaction. Delay in cleaning can also result in breakage of tooling apart from increase in the dose to extremities of personnel. New design of tooling has been introduced to clean up the top plungers online during the operation of rotary press. It leads to increase in the productivity, reduces the spillage of valuable nuclear material and also reduces man-rem to operators significantly. The present paper describes the modification in tooling design and compaction sequence established for online cleaning of top plungers. (author)

  4. Grids for nuclear fuel elements

    International Nuclear Information System (INIS)

    This invention relates to grids for nuclear fuel assemblies with the object of providing an improved grid, tending to have greater strength and tending to offer better location of the fuel pins. It comprises sets of generally parallel strips arranged to intersect to define a structure of cellular form, at least some of the intersections including a strip which is keyed to another strip at more than one point. One type of strip may be dimpled along its length and another type of strip may have slots for keying with the dimples. (Auth.)

  5. Annular pancreas

    Science.gov (United States)

    ... page: //medlineplus.gov/ency/article/001142.htm Annular pancreas To use the sharing features on this page, please enable JavaScript. An annular pancreas is a ring of pancreatic tissue that encircles ...

  6. Hydraulic modelling of the CARA Fuel element

    International Nuclear Information System (INIS)

    The CARA fuel element is been developing by the National Atomic Energy Commission for both Argentinean PHWRs. In order to keep the hydraulic restriction in their fuel channels, one of CARA's goals is to keep its similarity with both present fuel elements. In this paper is presented pressure drop test performed at a low-pressure facility (Reynolds numbers between 5x104 and 1,5x105) and rational base models for their spacer grid and rod assembly. Using these models, we could estimate the CARA hydraulic performance in reactor conditions that have shown to be satisfactory. (author)

  7. Numerical analyses and experiment investigations of an annular micro gas turbine power system using fuels with low heating values

    Institute of Scientific and Technical Information of China (English)

    YANG ChunHsiang; LEE ChengChia; HSIAO JenHao; CHEN ChiunHsun

    2009-01-01

    This study investigates the effects of using fuels with low heating values on the performance of an annular micro gas turbine(MGT)experimentally and numerically.The MGT used in this study is MW-54,whose original fuel is liquid(Jet al).Its fuel supply system is re-designed to use biogas fuel with low heating value(LHV).The purpose is to reduce the size of a biogas distributed power supply system and to enhance its popularization.This study assesses the practicability of using fuels with LHVs by using various mixing ratios of methane(CH_4)and carbon dioxide(CO_2).Prior to experiments,the corresponding simulations,aided by the commercial code CFD-ACE+,were carried out to investigate the cooling effect in a perforated combustion chamber and combustion behavior in an annular MGT when LHV gas was used.The main purposes are to confirm that there are no hot spots occurring in the liners and the exhaust temperatures of combustor are lower than 700℃ when MGT is operated under different conditions,in experiments,fuel pressure and mass flow rate,turbine rotational speed,generator power output,and temperature distribution were measured to analyze MGT performance.Experimental results indicate that the presented MGT system operates successfully under each tested condition when the minimum heating value of the simulated fuel is approximately 50%of pure methane.The power output is around 170 W at 85000 r/min as 90%CH_4 with 10%CO_2 is used and 70 W at 60000 r/min as 70%CH_4 with 30%CO_2 is used.When a critical limit of 60%CH_4 is used,the power output is extremely low.Furthermore,the best theoretical Brayton cycle efficiency for such MGT is calculated as 23%according to the experimental data while LHV fuel is used.Finally,the numerical results and experiment results reveal that MGT performance can be improved further and the possible solutions for performance improvement are suggested for the future studies.

  8. Numerical analyses and experiment investigations of an annular micro gas turbine power system using fuels with low heating values

    Institute of Scientific and Technical Information of China (English)

    YANG; ChunHsiang; LEE; ChengChia; HSIAO; JenHao; CHEN; ChiunHsun

    2009-01-01

    This study investigates the effects of using fuels with low heating values on the performance of an annular micro gas turbine(MGT)experimentally and numerically.The MGT used in this study is MW-54, whose original fuel is liquid(Jet A1).Its fuel supply system is re-designed to use biogas fuel with low heating value(LHV).The purpose is to reduce the size of a biogas distributed power supply system and to enhance its popularization.This study assesses the practicability of using fuels with LHVs by using various mixing ratios of methane(CH4)and carbon dioxide(CO2).Prior to experiments,the corresponding simulations,aided by the commercial code CFD-ACE+,were carried out to investigate the cooling effect in a perforated combustion chamber and combustion behavior in an annular MGT when LHV gas was used.The main purposes are to confirm that there are no hot spots occurring in the liners and the exhaust temperatures of combustor are lower than 700°C when MGT is operated under different conditions.In experiments,fuel pressure and mass flow rate,turbine rotational speed,generator power output,and temperature distribution were measured to analyze MGT performance.Experimental results indicate that the presented MGT system operates successfully under each tested condition when the minimum heating value of the simulated fuel is approximately 50%of pure methane.The power output is around 170 W at 85000 r/min as 90%CH4 with 10%CO2 is used and 70 W at 60000 r/min as 70%CH4 with 30%CO2 is used.When a critical limit of 60%CH4 is used,the power output is extremely low. Furthermore,the best theoretical Brayton cycle efficiency for such MGT is calculated as 23%according to the experimental data while LHV fuel is used.Finally,the numerical results and experiment results reveal that MGT performance can be improved further and the possible solutions for performance im- provement are suggested for the future studies.

  9. TRIGA - LEU cluster with 36 fuel elements

    International Nuclear Information System (INIS)

    Designing the TRIGA - LEU fuel cluster is part of the mechanical design of TRIGA reactor core. The latter is supported by a square frame (11 x 12 132 meshes) accommodating the 35 fuel clusters. The TRIGA fuel cluster is designed to incorporate 36 fuel elements with 3/8 inch diameter allowing the pins to be arranged into a 6 x 6 matrix. The final mechanical design of reactor zone resulted into a cluster of squared cross section with 87.5 mm side and 88.9 mm separation between the centers of the clusters. This cluster was designed by preserving the dimensions and configuration of fuel clusters with 25 elements. By the positioning of the pins inside the cluster one obtains: - a fuel element protection by reducing the failure risks; - delimitation of fixed channel of the cooling flow for each cluster; - a convenient means of manipulation; - a correct water flow for cooling the pins in a fixed channel by preserving the surface of cooling channels from the 25 fuel element cluster. The cluster has the following principal components: - casing; - bottom plug or adapter; - upper plug for maneuvering; - spacer for fuel elements. The cluster casing is made of aluminium with square cross section of 87.5 mm side and is provided at the lower part with an aluminium adapter allowing its insertion in the reactor core frame. This piece is designed to support the ends of the 36 fuel elements in a blocked position. The fuel elements are subject to asymmetric temperature distribution flux conditions, hence an asymmetric temperature distribution results concomitantly with a symmetrical (about 0.8 mm) swelling of the Incoloy 800 can. Also bending of the fuel element occurs which will be limited by the intermediate spacer. At the casing upper part an aluminium upper plug or handle is mounted allowing cluster maneuvering by means of a special tool. The cluster is provided with lateral holes in its upper part ensuring the necessary cooling water flow in case the upper part of the cluster

  10. Spacer for fuel rods in nuclear fuel elements

    International Nuclear Information System (INIS)

    Spacers for fuel rods in nuclear reactor fuel elements are described, especially for use aboard ships. Spacers are used in a grid formed by web plates orthogonally intersecting and assembled together in a tooth-comb fashion forming a plurality of channels. The web plates are joined together and each of the web plates includes apertures through which resilient and separator members are joined. The resilient and separator members are joined. The resilient and separator members are in adjacent channels and with other similar members in the same channel, contact a fuel rod in the channel. The contact pressure between the members and fuel rod is radially directed

  11. Comparisons of physical experiment and discrete element simulations of sheared granular materials in an annular shear cell

    Science.gov (United States)

    Ji, S.; Hanes, D.M.; Shen, H.H.

    2009-01-01

    In this study, we report a direct comparison between a physical test and a computer simulation of rapidly sheared granular materials. An annular shear cell experiment was conducted. All parameters were kept the same between the physical and the computational systems to the extent possible. Artificially softened particles were used in the simulation to reduce the computational time to a manageable level. Sensitivity study on the particle stiffness ensured such artificial modification was acceptable. In the experiment, a range of normal stress was applied to a given amount of particles sheared in an annular trough with a range of controlled shear speed. Two types of particles, glass and Delrin, were used in the experiment. Qualitatively, the required torque to shear the materials under different rotational speed compared well with those in the physical experiments for both the glass and the Delrin particles. However, the quantitative discrepancies between the measured and simulated shear stresses were nearly a factor of two. Boundary conditions, particle size distribution, particle damping and friction, including a sliding and rolling, contact force model, were examined to determine their effects on the computational results. It was found that of the above, the rolling friction between particles had the most significant effect on the macro stress level. This study shows that discrete element simulation is a viable method for engineering design for granular material systems. Particle level information is needed to properly conduct these simulations. However, not all particle level information is equally important in the study regime. Rolling friction, which is not commonly considered in many discrete element models, appears to play an important role. ?? 2009 Elsevier Ltd.

  12. Effect of fuel zoning and fuel nozzle design on pollution emissions at ground idle conditions for a double-annular ram-induction combustor

    Science.gov (United States)

    Clements, T. R.

    1973-01-01

    An exhaust emission survey was conducted on a double-annular ram induction combustor at simulated ground idle conditions. The combustor was designed for a large augmented turbofan engine capable of sustained flight speeds up to Mach 3.0. The emission levels of total hydrocarbon (THC), carbon monoxide, carbon dioxide, and nitric oxide were measured. The effects of fuel zoning, fuel nozzle design, and operating conditions (inlet temperature and reference Mach number) on the level of these emissions were determined. At an overall combustor fuel/air ratio of 0.007, fuel zoning reduced THC emissions by a factor of 5 to 1. The reduction in THC emissions is attributed to the increase in local fuel/air ratio provided by the fuel zoning. An alternative method of increasing fuel/air ratio would be to operate with larger-than-normal compressor overboard bleed; however, analysis on this method indicated an increase in idle fuel consumption of 20 percent. The use of air-atomizing nozzles reduced the THC emissions by 2 to 1.

  13. Annular dilatation and loss of sino-tubular junction in aneurysmatic aorta: Implications on leaflet quality at the time of surgery. A finite element study

    NARCIS (Netherlands)

    Weltert, L.; De Tullio, M.D.; Afferante, L.; Salica, A.; Scaffa, R.; Maselli, D.; Verzicco, R.; De Paulis, R.

    2013-01-01

    OBJECTIVES In the belief that stress is the main determinant of leaflet quality deterioration, we sought to evaluate the effect of annular and/or sino-tubular junction dilatation on leaflet stress. A finite element computer-assisted stress analysis was used to model four different anatomic condition

  14. Structural analysis of reactor fuel elements

    International Nuclear Information System (INIS)

    An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design

  15. Structural analysis of reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Weeks, R.W.

    1977-01-01

    An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design.

  16. Improved fuel element for fast breeder reactor

    International Nuclear Information System (INIS)

    The invention, in which the United States Department of Energy has participated as co-inventor, relates to breeder reactor fuel elements, and specifically to such elements incorporating 'getters', hereafter designated as fission product traps. The main object of the invention is the construction of a fast breeder reactor fuel pin, free from local stresses induced in the cladding by reactions with cesium. According to the invention, the fast breeder fuel element includes a cladding tube, sealed at both ends by a plug, and containing a fissile stack and a fertile stack, characterized by the interposition of a cesium trap between the fissile and fertile stacks. The trap is effective at reactor operating temperatures in retaining and separating the cesium generated in the fissile material and preventing cesium reaction with the fertile stack. Depending on the construction method adopted, the trap may consists of a low density titanium oxide or niobium oxide pellet

  17. Nuclear fuel element with a bond coating

    International Nuclear Information System (INIS)

    The possibility of undesired interactions between the pellets (of UO2 or a mixture of UO2 + PuO2) and the cladding which can cause stress crack corrosion, are to be excluded in particular in the proposed fuel element. The container enclosing the fuel consists according to the invention of a zirconium alloy having a zirconium oxide diffusion barrier on the side facing the fuel and a metal coating on top of this. Cu is best suited, but Ni, Fe or their alloys are named. The treatment of the surfaces to simplify the coating of the individual layers is described. (UWI) 891 HP/UWI 892 CKA

  18. Performance comparisons of Navy jet mix and MIL-F-5624A (JP-3) fuels in tubular and annular combustors / Richard J. McCafferty

    Science.gov (United States)

    Mccafferty, Richard J

    1954-01-01

    The performances of Navy Jet Mix and MIL-F-5624A (JP-3) fuels were compared in J33, J47, and NACA experimental annular combustion chambers. Combustion efficiencies, altitude operational limits, and carbon-forming tendencies wer determined and discussed. The results indicate that Jet Mix fuel can be utilized satisfactorily in a number of current turbojet engines over the normal operating range. Small differences in combustion efficiences and altitude operational limits existed between the two fuels but these differences depended on the particular combustor and simulated altitude and rotor speed condition. Excessive carbon deposition is not predicted for Jet Mix fuel, although this property may be marginal.

  19. Catalogue of fuel elements - 1. addendum October 1958

    International Nuclear Information System (INIS)

    This document contains sheets presenting various characteristics of nuclear fuel elements which are distinguished with respect to their shape: cylinder bar, plate, tube. Each sheet comprises an indication of the atomic pile in which the fuel element is used, dimensions, cartridge data, data related to cooling, to combustion rate, and to fuel handling. A drawing of the fuel element is also given

  20. Laser assisted decontamination of nuclear fuel elements

    International Nuclear Information System (INIS)

    Laser assisted removal of loosely bound fuel particulates from the clad surface following the process of pellet loading has decided advantages over conventional methods. It is a dry and noncontact process that generates very little secondary waste and can occur inside a glove box without any manual interference minimizing the possibility of exposure to personnel. The rapid rise of the substrate/ particulate temperature owing to the absorption of energy from the incident laser pulse results in a variety of processes that may lead to the expulsion of the particulates. As a precursor to the cleaning of the fuel elements, initial experiments were carried out on contamination simulated on commonly used clad surfaces to gain a first hand experience on the various laser parameters for which as efficient cleaning can be obtained without altering the properties of the clad surface. The cleaning of a dummy fuel element was subsequently achieved in the laboratory by integrating the laser with a work station that imparted simultaneous rotational and linear motion to the fuel element. (author)

  1. Stuck fuel element experience at the Oregon State TRIGA reactor

    International Nuclear Information System (INIS)

    A stuck fuel element was found in June 1975 during the annual fuel element measuring assignment. When an attempt was made to remove the fuel element from position D-6, it was found the element would start to bind after being withdrawn about 10'', and it would not pass through the upper grid plate. A plan was devised to extract the stuck fuel element without having to remove the upper grid plate. An inhouse inquiry is in process to determine the reasons for the fuel element deformation. When the element cools sufficiently, we plan to obtain neutron radiographs that may help determine the answer. (author)

  2. Nuclear criticality assessment of LEU and HEU fuel element storage

    International Nuclear Information System (INIS)

    Criticality aspects of storing LEU (20%) and HEU (93%) fuel elements have been evaluated as a function of 235U loading, element geometry, and fuel type. Silicide, oxide, and aluminide fuel types have been evaluated ranging in 235U loading from 180 to 620 g per element and from 16 to 23 plates per element. Storage geometry considerations have been evaluated for fuel element separations ranging from closely packed formations to spacings of several centimeters between elements. Data are presented in a form in which interpolations may be made to estimate the eigenvalue of any fuel element storage configuration that is within the range of the data. (author)

  3. Nuclear fuel element having oxidation resistant cladding

    International Nuclear Information System (INIS)

    This patent describes an improved nuclear fuel element of the type including a zirconium alloy tube, a zirconium barrier layer metallurgically bonded to the inside surface of the alloy tube, and a central core of nuclear fuel material partially filling the inside of the tube so as to leave a gap between the sponge zirconium barrier and the nuclear fuel material. The improvement comprising an alloy layer formed on the inside surface of the zirconium barrier layer. The alloy layer being composed of one or more impurities present in a thin layer region of the zirconium barrier in amounts less than 1% by weight but sufficient to inhibit the oxidation of the inside surface of the zirconium barrier layer without substantially affecting the plastic properties of the barrier layer, wherein the impurities are selected from the group consisting of iron, chromium, copper, nitrogen, and niobium

  4. Packaging of spent fuel elements into special containers

    International Nuclear Information System (INIS)

    This report contains detailed description of the procedure for packaging the spent fuel elements from the fuel channels into the special steel containers. The previously cooled fuel elements are packaged into containers by the existing crane and transported later into the spen fuel storage. Instructions for crane operation are included

  5. Fuel element situation and performance data TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Electronic data acquisition of the position and movement of Triga fuel elements (FE) in the TRIGA II Vienna reactor was the objective of this project. Using one month power data and the Fuel element position in core it is possible to calculate their burnup. Fuel element performance data during 1962 to 2003 are provided. (nevyjel)

  6. RIA and LOCA simulating tests on experimental fuel elements in TRIGA MT reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Full text: One of the main objectives of Institute for Nuclear Research (INR), Pitesti R and D Program is to investigate thermal and mechanical behaviour of fuel elements, thresholds and mechanisms of cladding failure during RIA and LOCA tests. Dual core TRIGA Material Testing Reactor of INR Pitesti (TRIGA SS MTR and TRIGA ACPR) is utilized extensively for studies of fuel behaviour under normal and postulated accident condition. A total of 39 test fuel elements have been irradiated in the TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti under RIA conditions. The ACPR tests program is still in progress and new experiments are foreseen to be performed in the following period. The test fuel elements are instrumented with CrAl thermocouples for cladding surface temperature measurement and every test fuel element has a pressure sensor for the internal pressure measurement. An experimental database of fuel behaviour parameters including fission - gas release, sheath strain, power - burnup history, etc. has been obtained using in-pile measurements and PIE results of test fuel elements irradiated in the TRIGA Steady State Material Testing Reactor (TRIGA SS MTR) of INR Pitesti. More than 100 test fuel elements have been irradiated in TRIGA SS MTR in different power history conditions. LOCA simulating tests are planned to be performed in C2 LOCA tests capsule and in Loop A of TRIGA SS MTR of INR Pitesti. The LOCA tests in capsule C2 are instrumented to measure fuel, sheath and coolant temperature, internal element and coolant pressure during the entire irradiation period. In the second phase of the experiment the C2 capsule will be connected to the sweep gas system with the on-line gamma ray spectrometer included. RIA type tests are planned in C6 capsule of TRIGA ACPR on test fuel elements with pre-hydrided claddings in order to investigate the influence of the precipitated hydride on fuel element cladding failure at high burnups in RIA conditions. This paper

  7. Searching for a possible fuel element leak

    International Nuclear Information System (INIS)

    A gamma spectrum analysis of a filter paper from an Oregon State University TRIGA Reactor (OSTR) continuous air monitor (CAM) which routinely monitors the air directly over the reactor tank revealed just-detectable levels of several short-lived particulate fission products typically associated with a fuel cladding failure. This prompted an intensive.search to determine the origin of these radionuclides. A number of methods were used, including a fuel element rotation program designed to ultimately remove all of the fuel elements from the core in groups of three, and a scheme to selectively sample bubbles from different parts of the core during operation. Determination of the source was made very difficult by the fact that its presence was erratic in nature and because radioactivity levels found on filter papers were on the border of detectability even when the reactor was operated at the maximum allowable power level of 1MW. The origin and source of the fission product activity was not found, no other abnormality was identified and the reactor was therefore returned to normal operation. In addition to continuing the routine operation of the reactor-top CAM, further surveillance designed to detect a positive reappearance of the source was also implemented and currently involves a complete gamma spectrum analysis of a CAM filter paper each week after a standard (controlled) 3 hour reactor run at 1 MW. (author)

  8. Neutron induced activity in fuel element components

    International Nuclear Information System (INIS)

    A thorough investigation of the importance of various nuclides in neutron-induced radioactivity from fuel element construction materials has been carried out for both BWR and PWR fuel assemblies. The calculations were performed with the ORIGEN computer code. The investigation was directed towards the final storage of the assembly components and special emphasis was put to the examination of the sources of carbon-14, cobalt-60, nickel-59, nickel-63 and zirconium-93/niobium-93m. It is demonstrated that the nuclides nickel-59, in Inconel and stainless steel, and zirconium-93/niobium-93m, in Zircaloy, are the ones which constitute the very long term radiotoxic hazard of the irradiated materials. (author)

  9. Thermionic fuel element verification program—overview

    Science.gov (United States)

    Bohl, Richard J.; Dutt, Dale S.; Dahlberg, Richard C.; Wood, John T.

    1991-01-01

    TFE Verification Program is in the sixth year of a program to demonstrate the performance and lifetime of thermionic fuel elements for high power space applications. It is jointly funded by SIDO and DOE. Data from accelerated tests in FFTF and EBR-II show component lifetimes longer than 7 years. Alumina insulators have shown good performance at high fast fluence. Graphite-cesium reservoirs based on isotropic graphite also meet requirements. Three TFEs are current operating in the TRIGA reactor, the oldest having accumulated 15,000 hours of irradiation as of 1 October 1990.

  10. Thermionic fuel element Verification Program - Overview

    Science.gov (United States)

    Bohl, Richard J.; Dahlberg, Richard C.; Dutt, Dale S.; Wood, John T.

    The TFE Verification Program is in the sixth year of a program to demonstrate the performance and lifetime of thermionic fuel elements for high power space applications. Data from accelerated tests in FETF and EBR-II show component lifetimes longer than 7 yr. Alumina insulators have shown good performance at high fast fluence. Graphite-cesium reservoirs based on isotropic graphite also meet requirements. Three TFEs are currently operating in the TRIGA reactor, the oldest having accumulated 15,000 hr of irradiation as of 1 October 1990.

  11. Fuel element storage pond for nuclear installations

    International Nuclear Information System (INIS)

    In a fuel element storage pond for nuclear installations, with different water levels, radioactive particles are deposited at the points of contact of the water surface with the pond wall. So that this deposition will not occur, a metal apron is provided in the area of the points of contact of the water surface with the bond wall. The metal apron consists of individual sheets of metal which are suspended by claws in wall hooks. To clean the sheets, these are moved to a position below the water level. The sheets are suspended from the wall hooks during this process. (orig.)

  12. Facility modernization Annular Core Research Reactor

    International Nuclear Information System (INIS)

    The Annular Core Research Reactor (ACRR) has undergone numerous modifications since its conception in response to program needs. The original reactor fuel, which was special U-ZrH TRIGA fuel designed primarily for pulsing, has been replaced with a higher pulsing capacity BeO fuel. Other advanced operating modes which use this increased capability, in addition to the pulse and steady state, have been incorporated to tailor power histories and fluences to the experiments. Various experimental facilities have been developed that range from a radiography facility to a 50 cm diameter External Fuel Ring Cavity (FREC) using 180 of the original ZrH fuel elements. Currently a digital reactor console is being produced with GA, which will give enhanced monitoring capabilities of the reactor parameters while leaving the safety-related shutdown functions with analog technology. (author)

  13. Fabrication of MOX Fuel elements for irradiation in Fast Breeder Test Reactor (FBTR)

    International Nuclear Information System (INIS)

    Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur is fabricating Uranium - Plutonium Mixed Oxide Fuel (MOX) for different types of reactors. Recently MOX fuel pins for an experimental fuel subassembly of 37 pins has been fabricated for irradiation in Fast Breeder Test Reactor (FBTR) at Kalpakkam near Chennai. MOX fuel pins containing 44% PuO2 have also been also made for the hybrid core of FBTR. The experimental sub-assembly for irradiation testing in FBTR consisted of 37 short length Prototype Fast Breeder Reactor (PFBR) MOX fuel elements. The composition of the fuel was (0.71 U - 0.29 Pu) O2 with U233 O2 content of 53.5% of total UO2. Uranium enriched with U233 was used to simulate the heat flux of PFBR in FBTR neutron spectrum. MOX fuel pellets were made by powder metallurgy process consisting of pre-compaction, granulation, final compaction and sintering at high temperature. Initially U3233 O8 / U233 O3 powder was subjected to heat treatment. The pellets were sintered at reducing atmosphere at 1650oC for 4 hours to obtain acceptable quality pellets. Over sized pellets were centrelessly ground.without using a liquid coolant. During the fabrication of pins for experimental subassembly, technology was developed and conditions were optimized for making annular pellets, TIG welding of D9 tubes with SS 316 end plugs and wire wrapping. Quality control procedures and process control procedures at different stages of fabrication were developed. The hybrid core of FBTR consists of Mixed Carbide (MC) sub-assemblies containing (0.70 Pu - 0.30 U) C pellets and MOX fuel sub-assemblies containing (0.44 Pu - 0.56 U) O2. Studies were made to fabricate fuel containing higher percentage of Plutonium and the conditions were established. This paper describes the development of flowsheet for making annular MOX fuel pellets containing plutonium and U233, the technology for welding of D-9 clad tubes, wire wrapping and inspection. The paper also

  14. Cold-neutron tomography of annular flow and functional spacer performance in a model of a boiling water reactor fuel rod bundle

    International Nuclear Information System (INIS)

    Highlights: → Annular flows w/wo functional spacers are investigated by cold neutron imaging. → Liquid film thickness distribution on fuel pins and on spacer vanes is measured. → The influence of the spacers on the liquid film distributions has been quantified. → The cross-sectional averaged liquid hold-up significantly affected by the spacers. → The sapers affect the fraction of the entrained liquid hold up in the gas core. - Abstract: Dryout of the coolant liquid film at the upper part of the fuel pins of a boiling water reactor (BWR) core constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is both a safety concern and a limiting factor in the thermal power and thus for the economy of BWRs. We have investigated adiabatic, air-water annular flows in a scaled-up model of two neighboring subchannels as found in BWR fuel assemblies using cold-neutron tomography. The imaging of the double suchannel has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institute, Switzerland. Cold-neutron tomography is shown here to be an excellent tool for investigating air-water annular flows and the influence of functional spacers of different geometries on such flows. The high-resolution, high-contrast measurements provide the spatial distributions of the coolant liquid film thickness on the fuel pin surfaces as well as on the surfaces of the spacer vanes. The axial variations of the cross-section averaged liquid hold-up and its fraction in the gas core shows the effect of the spacers on the redistribution of the two phases.

  15. UNIFRAME interim design report. [Fuel element size reduction plant

    Energy Technology Data Exchange (ETDEWEB)

    Strand, J.B.; Baer, J.W.; Cook, E.J.

    1977-12-01

    A fuel element size reduction system has been designed for the ''cold'' pilot-scale plant for an HTGR Fuel Reference Recycle Facility. This report describes in detail the present design.

  16. Fabrication technology of spherical fuel element for HTR-10

    International Nuclear Information System (INIS)

    R and D on the fabrication technology of the spherical fuel elements for the 10 MW HTR Test Module (HTR-10) began from 1986. Cold quasi-isostatic molding with a silicon rubber die is used for manufacturing the spherical fuel elements.The fabrication technology and the graphite matrix materials were investigated and optimized. Twenty five batches of fuel elements, about 11000 of the fuel elements, have been produced. The cold properties of the graphite matrix materials satisfied the design specifications. The mean free uranium fraction of 25 batches was 5 x 10-5

  17. Three-dimensional numerical simulation of a continuously rotating detonation in the annular combustion chamber with a wide gap and separate delivery of fuel and oxidizer

    Science.gov (United States)

    Frolov, S. M.; Dubrovskii, A. V.; Ivanov, V. S.

    2016-07-01

    The possibility of integrating the Continuous Detonation Chamber (CDC) in a gas turbine engine (GTE) is demonstrated by means of three-dimensional (3D) numerical simulations, i. e., the feasibility of the operation process in the annular combustion chamber with a wide gap and with separate feeding of fuel (hydrogen) and oxidizer (air) is proved computationally. The CDC with an upstream isolator damping pressure disturbances propagating towards the compressor is shown to exhibit a gain in the total pressure of 15% as compared with the same combustion chamber operating in the deflagration mode.

  18. Fuel Element Transfer Cask Modelling Using MCNP Technique

    Science.gov (United States)

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-01

    After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

  19. Reproduction of the RA reactor fuel element fabrication

    International Nuclear Information System (INIS)

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

  20. Determination of oxygen to uranium plus plutonium atom ratio in high density annular mixed oxide fuel pellets for fast reactor

    International Nuclear Information System (INIS)

    This paper highlights the encountered difficulties and applied modifications in the analytical steps for the determination of [O/(U+Pu)] in high density annular (NatU0.335233U0.37 Pu0.295)O2 pellets, manufactured for irradiation in FBTR and discusses the results. (author)

  1. Legal questions concerning the termination of spent fuel element reprocessing

    International Nuclear Information System (INIS)

    The thesis on legal aspects of the terminated spent fuel reprocessing in Germany is based on the legislation, jurisdiction and literature until January 2004. The five chapters cover the following topics: description of the problem; reprocessing of spent fuel elements in foreign countries - practical and legal aspects; operators' responsibilities according to the atomic law with respect to the reprocessing of Geman spent fuel elements in foreign countries; compatibility of the prohibition of Geman spent fuel element reprocessing in foreign countries with international law, European law and German constitutional law; results of the evaluation

  2. Attempt to produce silicide fuel elements in Indonesia

    International Nuclear Information System (INIS)

    After the successful experiment to produce U3Si2 powder and U3Si2-Al fuel plates using depleted U and Si of semiconductor quality, silicide fuel was synthesized using x-Al available at the Fuel Element Production Installation (FEPI) at Serpong, Indonesia. Two full-size U3Si2-Al fuel elements, having similar specifications to the ones of U3O8-Al for the RSG-GAS (formerly known as MPR-30), have been produced at the FEPI. All quality controls required have been imposed to the feeds, intermediate, as well as final products throughout the production processes of the two fuel elements. The current results show that these fuel elements are qualified from fabrication point of view, therefore it is expected that they will be permitted to be tested in the RSG-GAS, sometime by the end of 1989, for normal (∝50%) and above normal burn-up. (orig.)

  3. Demonstration tests for HTGR fuel elements and core components with test sections in HENDEL

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Yoshiaki; Hino, Ryutaro; Inagaki, Yoshiyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1995-03-01

    In the fuel stack test section (T{sub 1}) of the Helium Engineering Demonstration Loop (HENDEL), thermal and hydraulic performances of helium gas flows through a fuel rod channel and a fuel stack have been investigated for the High-Temperature Engineering Test Reactor (HTTR) core thermal design. The test data showed that the turbulent characteristics appearing in the Reynolds number above 2000: no typical behavior in the transition zone, and friction factors and heat transfer coefficients in the fuel channel were found to be higher than those in a smooth annular channel. Heat transfer behavior of gas flow in a fuel element channel with blockage and cross-flow through a gap between upper and lower fuel elements stacked was revealed using the mock-up models. On the other hand, demonstration tests have been performed to verify thermal and hydraulic characteristics and structural integrity related to the core bottom structure using a full-scale test facility named as the in-core structure test section (T{sub 2}). The sealing performance test revealed that the leakage of low-temperature helium gas through gaps between the permanent reflector blocks to the core was very low level compared with the HTTR design value and no change of the leakage flow rate were observed after a long term operation. The heat transfer tests including thermal transient at shutdown of gas circulators verified good insulating performance of core insulation structures in the core bottom structure and the hot gas duct; the temperature of the metal portion of these structure was below the design value. Examination of the thermal mixing characteristics indicated that the mixing of the hot helium gas started at a hot plenum and finished completely at downstream of the outlet hot gas duct. The present results obtained from these demonstration tests have been practically applied to the detailed design works and licensing procedures of the HTTR. (J.P.N.) 92 refs.

  4. CHF Enhancement of Advanced 37-Element Fuel Bundles

    Directory of Open Access Journals (Sweden)

    Joo Hwan Park

    2015-01-01

    Full Text Available A standard 37-element fuel bundle (37S fuel bundle has been used in commercial CANDU reactors for over 40 years as a reference fuel bundle. Most CHF of a 37S fuel bundle have occurred at the elements arranged in the inner pitch circle for high flows and at the elements arranged in the outer pitch circle for low flows. It should be noted that a 37S fuel bundle has a relatively small flow area and high flow resistance at the peripheral subchannels of its center element compared to the other subchannels. The configuration of a fuel bundle is one of the important factors affecting the local CHF occurrence. Considering the CHF characteristics of a 37S fuel bundle in terms of CHF enhancement, there can be two approaches to enlarge the flow areas of the peripheral subchannels of a center element in order to enhance CHF of a 37S fuel bundle. To increase the center subchannel areas, one approach is the reduction of the diameter of a center element, and the other is an increase of the inner pitch circle. The former can increase the total flow area of a fuel bundle and redistributes the power density of all fuel elements as well as the CHF. On the other hand, the latter can reduce the gap between the elements located in the middle and inner pitch circles owing to the increasing inner pitch circle. This can also affect the enthalpy redistribution of the fuel bundle and finally enhance CHF or dry-out power. In this study, the above two approaches, which are proposed to enlarge the flow areas of the center subchannels, were considered to investigate the impact of the flow area changes of the center subchannels on the CHF enhancement as well as the thermal characteristics by applying a subchannel analysis method.

  5. 生物质固体成型燃料环模成型技术研究进展%The research progress in biomass annular mould forming for fuel technology

    Institute of Scientific and Technical Information of China (English)

    欧阳双平; 侯书林; 赵立欣; 田宜水; 孟海波

    2011-01-01

    综合分析了国内外生物质固体成型燃料环模成型技术、成型设备及产业发展现状.比较了生物质环模颗粒成型机和生物质环模压块成型机的性能和产品,指出了生物质固体成型燃料环模技术及设备存在着基础理论薄弱、原料适应差、易损件寿命短等问题;提出了我国生物质固体成型燃料环模成型技术的发展方向.%The research status of annular mould forming for fuel technology, forming equipment and the industrial development has been comprehensively analyzed, the performance and the product of annular mould biomass pellet machine and annular mould biomass briquette machine have been compared, the problems of biomass annular mould forming for fuel technology and equipment, such as weak theory foundation, poor feedstock suitability, short life span of wearing part, were pointed out, the development orientation of biomass annular mould forming for fuel technology in China was proposed.

  6. Nonlinear transient deformation of LMFBR fuel elements under impulsive loading

    International Nuclear Information System (INIS)

    Hypothetical reactor accidents are characterized by a sudden release of substantial thermal energy in one fuel element. Presently it cannot be excluded that for instance pressure pulses due to a fuel coolant interaction may have such time scales and impulses as to deform neighboring subassemblies permanently. Additionally coherent fuel element motion may limit control rod scram action and possibly cause untolerable reactivity increases. Therefore LMFBR safety requires to analyse the complex mechanical response of the core structure under typical loading conditions. An important contribution to this problem is to examine the nonlinear structural dynamics of an individual fuel element under prescribed loading and boundary conditions. The subject of this paper is the elastoplastic transient behaviour of one subassembly under given space-and-time dependent pressure loading. The interaction of several colliding fuel elements including coolant dynamics is briefly discussed. (Auth.)

  7. The International Marketing Target of Fuel Element for Research Rectors

    International Nuclear Information System (INIS)

    The International marketing efforts of PT BATAN Teknologi's fuel element for research reactors are out line. These efforts intensively started in third year marketing time since it is commenced on 24 May 1996. The market segmentation told that there are 269 research reactors in the world, I.e. 65 in USA, 27 in Russia, 18 in Japan, and the remaining are in many Countries. Many of those are 78 swimming fool type reactors, and 17 of them, I.e. 4 in Japan, 4 in USA, and each Austria, Germany, Argentina, Iran, Pakistan, Peru, Brazil, Algeria and Indonesia have the similar fuel element specifications with are close related with PT BATAN Teknologi's. It can be predicated that around 38 fuel elements and 84 fuel control can be marketed. The first feasibility study told that for countries such as Peru, Pakistan, Iran, Algeria, became the potential marketing target of the BATAN Teknologi's fuel element, because for those countries the competitors in producing such fuel elements could be minimal. The fuel elements and fuel control which could be presumably marketed in those countries are 83 and 19 respectively. The problem will be facing in near future such as packaging design and nuclear fuel transportation have to be firstly solved by collaborating with foreign companies abroad. Non technical problems including political situation have to be completely studied in order the uranium, transfer to many countries for exporting purpose could easily take place in the future. The government of the Republic of Indonesia (in this case BATAN) and the International Atomic Energy Agency (IAEA) could assist to solve the non technical problems which might be appear in the future as the chance of the exporting the fuel elements and the fuel controls come true. (author)

  8. Experience with TRIGA aluminum-clad fuel elements

    International Nuclear Information System (INIS)

    During 8 years of operation the cumulative heat energy produced in the steady-state TRIGA Mark II 250 kW reactor at Ljubljana reached 4683 MWh. The initial core had Al-clad fuel elements only. The reactivity loss due to the burnup has been compensated by fresh fuel elements with SS-cladding and, lately, by FLIP fuel elements, moving the most irradiated Al-clad fuel elements from B and C rings to the F ring and, lately, to the storage rack. The inspection of the fuel elements during the summer of 1973 revealed excessive elongations of some Al-clad fuel elements, up to 36.8 mm. By the neutronography, performed by indirect methods (In, Dy), and also by direct methods (track detector CA 80-15 B) and by special radiographic procedures on the element, the activity of which decayed sufficiently, it has been demonstrated that the growth is due to the elongation of aluminum cladding only. No growth and/or swelling of the ZrH--U fuel or the graphite plugs has been observed within the accuracy of detection. (U.S.)

  9. [Granuloma annulare].

    Science.gov (United States)

    Butsch, F; Weidenthaler-Barth, B; von Stebut, E

    2015-11-01

    Granuloma annulare is a benign, chronic inflammatory skin disease. Its pathogenesis is still unclear, but reports on infections as a trigger can be found. In addition, some authors reported an association with other systemic disease, e.g., cancer, trauma, and diabetes mellitus; however, these have not been verified. The clinical picture of granuloma annulare ranges from the localized form predominantly at the extremities to disseminated, subcutaneous, or perforating forms. Diagnosis is based on the typical clinical presentation which may be confirmed by a biopsy. Histologically, necrobiotic areas within granulomatous inflammation are typical. The prognosis of the disease is good with spontaneous resolution being frequently observed, especially in localized forms. Disseminated manifestations tend to persist longer, and recurrences are reported. When choosing between different therapeutic options, the benign disease character versus the individual degree of suffering and the potential therapy side effects must be considered. For local treatment, topical application of corticosteroids is most common. Disseminated forms can be treated systemically with corticosteroids for several weeks; alternatively, dapsone, hydroxychloroquine, retinoids, fumaric acid, cyclosporine, and anti-TNFα appear to be effective. PMID:26487494

  10. Hermetic seal process for nuclear fuel element

    International Nuclear Information System (INIS)

    The welding of the end plug onto the sheath of the fuel rod is made inside an enclosure filled with inert gas under the same pressure at that needed inside the fuel rod. The welding can be a tungsten arc welding, a laser welding or a micro plasma welding

  11. Repurposing an irradiated instrumented TRIGA fuel element for regular use

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Paulo F.; Souza, Luiz C.A., E-mail: pfo@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    TRIGA IPR-R1 is a research reactor also used for training and radioisotope production, located at the Centro de Desenvolvimento da Tecnologia Nuclear da Comissao Nacional de Energia Nuclear (Nuclear Technology Development Centre, Brazilian National Nuclear Energy Commission - CDTN/CNEN). Its first criticality occurred in November 1960. All original fuel elements were aluminum-clad. In 1971 nine new fuel elements, stainless steel-clad were acquired. One of them was an instrumented fuel element (IFE), equipped with 3 thermocouples. The IFE was introduced into the core only on August 2004, and remained there until July 2007. It was removed from the core after the severing of contacts between the thermocouples and their extension cables. After an unsuccessful attempt to recover electrical access to the thermocouples the IFE was transferred from the reactor pool to an auxiliary spent fuel storage well, with water, in the reactor room. In December 2011 the IFE was transferred to an identical well, dry, where it remains so far. This work is a proposal for recovery of this instrumented fuel element, by removing the cable guide rod and adaptation of a superior terminal plug similar to conventional fuel elements. This will enable its handling through the same tool used for regular fuel elements and its return to the reactor core. This is a delicate intervention in terms of radiological protection, and will require special care to minimize the exposure of operators. (author)

  12. Container for transport of radioactive fuel elements

    International Nuclear Information System (INIS)

    Five or six fuel assemblies may directly be inserted into the bearing cage placed in the storage pool. Later, after decay, it will be possible to put the bearing cage containing the fuel assemblies into the shipping cask for the reprocessing plant. The shipping cask has got a cover filled up for the transport with a sealing compound consisting of salt, a mixture of salt, or bitumen. The wall of the shipping cask has got a sandwich structure. (DG)

  13. Design and Testing of Prototypic Elements Containing Monolithic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    N.E. Woolstenhulme; M.K. Meyer; D.M. Wachs

    2011-10-01

    The US fuel development team has performed numerous irradiation tests on small to medium sized specimens containing low enriched uranium fuel designs. The team is now focused on qualification and demonstration of the uranium-molybdenum Base Monolithic Design and has entered the next generation of testing with the design and irradiation of prototypic elements which contain this fuel. The designs of fuel elements containing monolithic fuel, such as AFIP-7 (which is currently under irradiation) and RERTR-FE (which is currently under fabrication), are appropriate progressions relative to the technology life cycle. The culmination of this testing program will occur with the design, fabrication, and irradiation of demonstration products to include the base fuel demonstration and design demonstration experiments. Future plans show that design, fabrication, and testing activities will apply the rigor needed for a demonstration campaign.

  14. Thermomechanical analysis of nuclear fuel elements

    International Nuclear Information System (INIS)

    This work presents development of a code to obtain the thermomechanical analysis of fuel rods in the fuel assemblies inserted in the core of BWR reactors. The code uses experimental correlations developed in several laboratories. The development of the code is divided in two parts: a) the thermal part and b) the mechanical part, extending both the fuel and the cladding materials. The thermal part consists of finding the radial distribution of temperatures in the pellet, from the fuel centerline up to the coolant, along the total active length, considering one and two phase flow in the coolant, as a result of the pressure drop in the system. The mechanical part analyzes the effects of temperature gradients, pressure and irradiation, to which the fuel rod is subjected. The strains produced by swelling, creep and thermal stress in the fuel material are analyzed. In the same way the strains in the cladding are analyzed, considering the effects produced by the pressure exerted on the cladding by pellet swelling, by the pressure caused by fission gas release toward the cavities, and by the strain produced on the cladding by the pressure changes of the system. (Author)

  15. Manufacture of nuclear fuel elements for commercial PWR in China

    International Nuclear Information System (INIS)

    Yibin Nuclear Fuel Element Plant (YFP) under the leadership of China National Nuclear Corporation is sole manufacturer in China to specialize in the production of fuel assemblies and associated core components for commercial PWR nuclear power plant. At the early of 1980's, it began to manufacture fuel assemblies and associated core components for the first core of QINSHAN 300 MW nuclear power plant designed and built by China itself. With the development of nuclear power industry in China and the demand for localization of nuclear fuel elements in the early 1990's, YFP cooperated with FRAMATOME France in technology transfer for design and manufacturing of AFA 2G fuel assembly and successfully supplied the qualified fuel assemblies for the reloads of two units of GUANGDONG Da Ya Bay 900 MW nuclear power plant (Da Ya Bay NPP), and has achieved the localization of fuel assemblies and nuclear power plants. Meanwhile, it supplied fuel assemblies and associated core components for the first core and further reloads of Pakistan CHASHMA 300 MW nuclear power plant which was designed and built by China, and now it is manufacturing AFA 2G fuel assemblies and associated core components for the first core of two units of NPQJVC 600 MW nuclear power plant. From 2001 on, YFP will be able to supply Da Ya Bay NPP with the third generation of fuel assembly-AFA 3G which is to realize a strategy to develop the fuel assembly being of long cycle reload and high burn-up

  16. Behavior analysis of U3Si-Al fuel in MP type fuel elements under irradiation

    International Nuclear Information System (INIS)

    Uranium silicide U3Si is considered as perspective nuclear fuel for Russian research reactors. In order to resolve the problem of enrichment reduction this nuclear fuel is the most real alternative for the Uranium dioxide which is currently used for these purposes. Within RERTR program two MP type fuel element models with the core consisting of U3Si nuclear fuel dispersed in an aluminium matrix were tested in MP reactor. The tests confirmed that the use of U3Si + Al fuel composition is a perspective solution to reduce fuel element enrichment in research reactors. This report represents analysis of post-irradiation tests of the fuel element models. The goal of the analysis being to establish the value and the appropriateness of swelling for the Uranium silicide. The fuel element represents a cylinder tube with four ribs on the outer surface. The claddings are produced of CAB-6 alloy. The contents of nuclear fuel in the core constitute 34% by volume, technological pores constitute 4.5% and the rest is aluminium matrix. The nuclear fuel was produced in ARSRIIM, the fuel elements was produced by ARSRIIM specialists with equipment of NZKH. (author)

  17. Research on Measuring Technology for In-pile Fuel Element Testing

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The tested fuel assembly for In-pile test for PWR fuel element with instrumentation consisted of 4instrumented fuel elements and total 12 sets of transducers. Double claddings are adopted to raise fueltemperature. Two fuel elements each have 2 thermocouples for measuring separately the fuel centerlinetemperature and the cladding surface temperature. The other two elements have membrane type oressure

  18. Inspection of state of spent fuel elements stored in RA reactor spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Bulkin, S.Yu.; Sokolov, A.V. [Research and Development Institute of Power Engineering, Moscow (Russian Federation); Matausek, M.V.; Vukadin, Z. [VINCA Institute of Nuclear Science, Belgrade (Yugoslavia)

    1999-07-01

    About five thousand spent fuel elements from RA reactor have been stored for over 30 years in sealed aluminum barrels in the spent fuel storage pool. This way of storage does not provide complete information about the state of spent fuel elements or the medium inside the barrels, like pressure or radioactivity. The technology has recently been developed and the equipment has been manufactured to inspect the state of the spent fuel and to reduce eventual internal pressure inside the aluminum barrels. Based on the results of this inspection, a procedure will be proposed for transferring spent fuel to a more reliable storage facility. (author)

  19. Failed MTR Fuel Element Detect in a Sipping Tests

    International Nuclear Information System (INIS)

    This work describes sipping tests performed on Material Testing Reactor (MTR) fuel elements of the IEA-R1 research reactor, in order to find out which one failed in the core during a routine operation. Radioactive iodine isotopes 131I and 133I, employed as failure monitors, were detected in samples corresponding to the failed fuel element. The specific activity of each sample, as well as the average leaking rate, were measured for 137Cs. The nuclear fuels U3O8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137Cs

  20. Transferencia de calor incrementada en espacios anulares con elementos helicoidales insertados//Review of augmentation techniques for heat transfer coefficient in annular spaces using helical elements

    Directory of Open Access Journals (Sweden)

    Josué Imbert‐González

    2014-08-01

    Full Text Available La transferencia de calor incrementada por métodos pasivos se emplea en diversosintercambiadores de calor de alta efectividad. El objetivo del trabajo presentado fue la evaluación del estado de las investigaciones en el campo de la transferencia de calor mejorada en espacios anulares, a partir del empleo de elementos turbulizadores helicoidales como técnicas pasivas. La revisión se centró en el empleo de láminas helicoidales y espirales, la obtención de ecuaciones de correlación del coeficiente de transferencia de calor incrementado, el coeficiente de fricción y la evaluación que se realiza de este proceso por parte de diferentes autores. El análisis crítico permitió realizar valoraciones integradas y recomendar sobre los aspectos que podrían ser analizados en el futuro en esta temática.Palabras claves: transferencia de calor incrementada, láminas helicoidales, espirales, espacios anulares, métodos pasivos._______________________________________________________________________________AbstractThe transfer enhancement by passive methods is used in several heat exchanger of high effectiveness. The objective of the presented work was the evaluation of the state of the investigations in heat transfer enhancement in annular spaces, from the employment of elements helical. The revision was centered in the employment of twisted tape and wire coil in spiral, the equations of correlation obtained of the coefficient of transfer of increased heat, the coefficient of friction and the evaluation that was carried out of this process on the part of different authors. From the critical analysis of the published results, the authors recommend on the topics that can be analyzed in the future in this area.Key words: heat transfer enhancement, twisted tape, helical springs, annular spaces, passive methods.

  1. In-pile steam oxidation of model HTGR fuel elements

    International Nuclear Information System (INIS)

    Model HTGR fuel elements were exposed to various concentrations of steam while being irradiated under several sets of temperature conditions in the Oak Ridge Research Reactor. In one test, catalysis by iron impurities in the graphite casing of the fuel element caused a highly localized attack on the graphite by the steam; this resulted in the formation of deep pits in the casing. Furthermore, the iron impurities were sufficiently mobile to cause pitting attack on the pyrolytic carbon coatings of the fuel particles as well. The presence of steam induced a rapid increase in the release of gaseous fission products. However, the cessation of steam ingress in the primary system resulted in a pronounced, but correspondingly smaller, reduction in the level of gaseous release. The incidence of fuel failure was greater than anticipated; however, even though the coatings of greater than 30% of the fuel had failed, the release of fission products beyond the fuel element itself was largely confined to iodine and the noble gases. A novel mode of fuel failure was observed under the rather severe conditions of the tests; this involved the attack of the pyrolytic carbon coatings on intact particles by uncoated fragments of uranium fuel kernel material from failed particles

  2. Uranium density reduction on fuel element side plates assessment

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  3. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    International Nuclear Information System (INIS)

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code

  4. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  5. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Science.gov (United States)

    Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

    2016-01-01

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  6. Use of silicide fuel in the Ford Nuclear Reactor - to lengthen fuel element lifetimes

    Energy Technology Data Exchange (ETDEWEB)

    Bretscher, M.M.; Snelgrove, J.L. [Argonne National Lab., IL (United States); Burn, R.R.; Lee, J.C. [Univ. of Michigan, Ann Arbor, MI (United States). Phoenix Memorial Lab.

    1995-12-31

    Based on economic considerations, it has been proposed to increase the lifetime of LEU fuel elements in the Ford Nuclear Reactor by raising the {sup 235}U plate loading from 9.3 grams in aluminide (UAl{sub x}) fuel to 12.5 grams in silicide (U{sub 3}Si{sub 2}) fuel. For a representative core configuration, preliminary neutronic depletion and steady state thermal hydraulic calculations have been performed to investigate core characteristics during the transition from an all-aluminide to an all-silicide core. This paper discusses motivations for this fuel element upgrade, results from the calculations, and conclusions.

  7. Finite element simulation of thermal, elastic and plastic phenomena in fuel elements

    International Nuclear Information System (INIS)

    Taking as starting point an irradiation experiment of the first Argentine MOX fuel prototype, performed at the HFR reactor of Petten, Holland, the deformation suffered by the fuel element materials during burning has been numerically studied. Analysis of the pellet-cladding interaction is made by the finite element method. The code determines the temperature distribution and analyzes elastic and creep deformations, taking into account the dependency of the physical parameters of the problem on temperature. (author)

  8. Performance and management of IPR-R1 fuel elements

    International Nuclear Information System (INIS)

    The performance of fuel elements during the 23 years of the reactor operation, is presented aiming to introduce improvements in the fuel load distribution and consequent increase of the reactivity. A computer code CORE was developed aiming to calculate the individual burnup of the fuel elements and the value of the reactivity for several core configurations, establishing a routine to control the nuclear material in the IPR-R1. The values calculated were compared with the experimental results. Some alternatives to augment the reactivity of the present core are presented foreseeing the fuel load availability for operation with 100Km and, for angmenting the power reaction in a next stage. (E.G.)

  9. Detection of fuel element vibration at KNK II

    International Nuclear Information System (INIS)

    The reactivity signal of the KNK-II-plant shows almost harmonic oscillations of δrho <= 0.5 c. Very sensitive correlation measurements, made during the regular plant operation with the normal plant instrumentation, revealed, that these oscillations are associated with individual fuel elements. Auxiliary measurements under various operational conditions and theoretical considerations show, that this phenomenon is probably caused by flow-induced mechanical vibration. Similar characteristics with respect to the frequencies have obviously not yet been observed for fuel element vibration during tests in out-of-core loops and in other reactors. Therefore efforts have been made in order to classify the flow-induced vibration and to identify the particular excitation mechanism. Most likely seems a flow-induced vibration of whole fuel elements by vortex shedding or jet switching. This model can explain all observations without exception. (orig.)

  10. Research Progress About Gas-Exhaust-Device for Fuel Element

    Institute of Scientific and Technical Information of China (English)

    ZHONG; Wu-ye

    2012-01-01

    <正>UO2-x stack applied in the fuel element has a form of a cylinder with a central hole, where temperature field characterized by high temperature and high gradient is formed due to irradiation. Then nearly all of the gaseous fission products (GFPs) can release into central cavity. However, uranium oxide will evaporate form the fuel stack’s inner surface because of its high temperature (about 1 800-2 000 ℃),

  11. The manufacture of LEU fuel elements at Dounreay

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, J.

    1997-08-01

    Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.

  12. CONDOR: neutronic code for fuel elements calculation with rods

    International Nuclear Information System (INIS)

    CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author)

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

    International Nuclear Information System (INIS)

    The properties of U3Si2 fuel and U3Si2-Al dispersion fuel element are introduced, which include U-loading; the banding quality, U-homogeneity and 'dog-bone' phenomenon, the minimum thickness of cladding and the corrosion performances. The fabrication technique of fuel elements, NDT for fuel plates, assemble technique of fuel elements and the application of U3Si2-Al dispersion fuel elements in the world are introduced

  14. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    Science.gov (United States)

    Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  15. Liquid films and droplet deposition in a BWR fuel element

    International Nuclear Information System (INIS)

    In the upper part of boiling water reactors (BWR) the flow regime is dominated by a steam-water droplet flow with liquid films on the nuclear fuel rod, the so called (wispy) annular flow regime. The film thickness and liquid flow rate distribution around the fuel rod play an important role especially in regard to so called dryout, which is the main phenomenon limiting the thermal power of a fuel assembly. The deposition of droplets in the liquid film is important, because this process sustains the liquid film and delays dryout. Functional spacers with different vane shapes have been used in recent decades to enhance droplet deposition and thus create more favorable conditions for heat removal. In this thesis the behavior of liquid films and droplet deposition in the annular flow regime in BWR bundles is addressed by experiments in an adiabatic flow at nearly ambient pressure. The experimental setup consists of a vertical channel with the cross-section resembling a pair of neighboring subchannels of a fuel rod bundle. Within this double subchannel an annular flow is established with a gas-water mixture. The impact of functional spacers on the annular flow behavior is studied closely. Parameter variations comprise gas and liquid flow rates, gas density and spacer shape. The setup is instrumented with a newly developed liquid film sensor that measures the electrical conductance between electrodes flush to the wall with high temporal and spatial resolution. Advanced post-processing methods are used to investigate the dynamic behavior of liquid films and droplet deposition. The topic is also assessed numerically by means of single-phase Reynolds-Averaged-Navier-Stokes CFD simulations of the flow in the gas core. For this the commercial code STAR-CCM+ is used coupled with additional models for the liquid film distribution and droplet motion. The results of the experiments show that the liquid film is quite evenly distributed around the circumference of the fuel rods. The

  16. Fuel element reshuffling and fuel follower control rods (FFCR) replacement for PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    The PUSPATI TRIGA Reactor has been utilized for more than 25 years using the same fuel elements and control rods. Generally, there are four control rods being used to control the neutron production inside the reactor core. A maintenance program has been developed to ensure its integrity, capability and safety of the reactor and it has been maintained twice a year since the first operation in 1982. The activities involve during the maintenance period including fuel elements and control rods inspections, electronics and mechanical systems, and others related works. During the maintenance in August 2008, there are some irregularities found on the fuel follower control rods and needed to be replaced. Even though the irregularities was not contributed into any unwanted incident, it were decided to replace with new control rods to avoid any potential hazards and unsafe condition occurred during operation later. Replacing any of the control rods would involved in imbalance of neutron flux and power distribution inside the core. Therefore, a number of fuel elements need to be reshuffled in order to compensate the neutron flux and power distribution as well as to balance the fuel elements burn-up in the core. This paper will described the fuel elements reshuffling and fuel follower control rods (FFCR) replacement for PUSPATI TRIGA Reactor. (Author)

  17. Fission product release from defected nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    The release of gaseous (krypton and xenon) and iodine radioactive fission products from defective fuel elements is described with a semi-empirical model. The model assumes precursor-corrected 'Booth diffusional release' in the UO2 and subsequent holdup in the fuel-to-sheath gap. Transport in the gap is separately modelled with a phenomenological rate constant (assuming release from the gap is a first order rate process), and a diffusivity constant (assuming transport in the gap is dominated by a diffusional process). Measured release data from possessing various states of defection are use in this analysis. One element (irradiated in an earlier experiment by MacDonald) was defected with a small drilled hole. A second element was machined with 23 slits while a third element (fabricated with a porous end plug) displayed through-wall sheath hydriding. Comparison of measured release data with calculated values from the model yields estimates of empirical diffusion coefficients for the radioactive species in the UO2 (1.56 x 10-10 to 7.30 x 10-9 s-1), as well as escape rate constants (7.85 x 10-7 to 3.44 x 10-5 s-1) and diffusion coefficients (3.39 x 10-5 to 4.88 x 10-2 cm2/s) for these in the fuel-to-sheath gap. Analyses also enable identification of the various rate-controlling processes operative in each element. For the noble gas and iodine species, the rate-determining process in the multi-slit element is 'Booth diffusion'; however, for the hydrided element an additional delay results from diffusional transport in the fuel-to-heath gap. Furthermore, the iodine species exhibit an additional holdup in the drilled element because of significant trapping on the fuel and/or sheath surfaces. Using experimental release data and applying the theoretical results of this work, a systematic procedure is proposed to characterize fuel failures in commercial power reactors (i.e., the number of fuel failures and average leak size)

  18. Core analysis during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculations were carried out with the RFSP code, provided by cell averaged fuel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift at a time. The simulation results show that the maximum channel and bundle powers were maintained below the license limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period. 3 refs., 2 figs., 1 tab. (Author)

  19. Automation in inspection of PHWR fuel elements & bundles at Nuclear Fuel Complex

    International Nuclear Information System (INIS)

    Nuclear Fuel Complex (NFC), Hyderabad, a constituent of Department of Atomic Energy, India manufactures fuel for all Indian nuclear power reactors. Currently NFC manufactures both 19 element & 37 element bundles for catering to the requirement of 220 MWe & 540 MWe PHWRs. In order to meet the growing needs for the Nuclear Fuel, NFC engaged in expansion of the production facilities. This calls for enhanced throughput at various inspection stages keeping in tandem with the production & for achieving this objective, NFC has chosen automation. This paper deals with automation of the inspection line at NFC. (author)

  20. Design evaluation of the HTGR fuel element size reduction system

    Energy Technology Data Exchange (ETDEWEB)

    Strand, J.B.

    1978-06-01

    A fuel element size reduction system for the ''cold'' pilot plant of the General Atomic HTGR Reference Recycle Facility has been designed and tested. This report is both an evaluation of the design based on results of initial tests and a description of those designs which require completion or modification for hot cell use. 11 figures.

  1. Design evaluation of the HTGR fuel element size reduction system

    International Nuclear Information System (INIS)

    A fuel element size reduction system for the ''cold'' pilot plant of the General Atomic HTGR Reference Recycle Facility has been designed and tested. This report is both an evaluation of the design based on results of initial tests and a description of those designs which require completion or modification for hot cell use. 11 figures

  2. Fuel burnup calculation of a research reactor plate element

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: nadiasam@gmail.com, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This work consists in simulating the burnup of two different plate type fuel elements, where one is the benchmark MTR of the IAEA, which is made of an alloy of uranium and aluminum, while the other belonging to a typical multipurpose reactor is composed of an alloy of uranium and silicon. The simulation is performed using the WIMSD-5B computer code, which makes use of deterministic methods for solving neutron transport. In developing this task, fuel element equivalent cells were calculated representing each of the reactors to obtain the initial concentrations of each isotope constituent element of the fuel cell and the thicknesses corresponding to each region of the cell, since this information is part of the input data. The compared values of the k∞ showed a similar behavior for the case of the MTR calculated with the WIMSD-5B and EPRI-CELL codes. Relating the graphs of the concentrations in the burnup of both reactors, there are aspects very similar to each isotope selected. The application WIMSD-5B code to calculate isotopic concentrations and burnup of the fuel element, proved to be satisfactory for the fulfillment of the objective of this work. (author)

  3. Method to fabricate block fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    The fabrication of block fuel elements for gas-cooled high temperature reactors can be improved upon by adding 0.2 to 2 wt.% of a hydrocarbon compound to the lubricating mixture prior to pressing. Hexanol or octanol are named as substances. The dimensional accuracy of the block is thus improved. 2 examples illustrate the method. (RW)

  4. Experimental analysis of heat flow in simulated fuel elements

    International Nuclear Information System (INIS)

    Since the experimental point of view it has been developed so much thermic simulations of nuclear reactors fuel elements in the laboratory. It is treating to isolate the problem of heat transfer of the complexity of the radioactive materials handling. The simulations starting of electric warming of similar geometric bodies to the real fuel elements. In the Thermo fluids Laboratory of National Institute of Nuclear Research it has been carried out heat transfer experiments in simulated fuel elements using in a first step concentric cylinders, for later to pass to posterior step of direct warming. The purpose of this work is to determine the convective parameters in the refrigerating under the typical prevailing conditions in the experimental reactors. It has been planned to work with isolated bars and groups of bars in convection with water. These works will allow to stablish the infrastructure of laboratory where it can be simulated thermically fuel elements of diverse types of experimental reactors. And specially to observe the solid-fluid effects in vertical surfaces subjected to intense heat fluxes. (Author)

  5. Experimental study of water flow in nuclear fuel elements

    International Nuclear Information System (INIS)

    This work aims to develop an experimental methodology for investigating the water flow through rod bundles after spacer grids of nuclear fuel elements of PWR type reactors. Speed profiles, with the device LDV (Laser Doppler Velocimetry), and the pressure drop between two sockets located before and after the spacer grid, using pressure transducers were measured

  6. Poolside inspection, repair and reconstitution of LWR fuel elements

    International Nuclear Information System (INIS)

    The purpose of the meeting was to review the state of the art in the area of poolside inspection, repair and reconstitution of light water fuel elements. In the present publication it appears that techniques of inspection, repair and reconstitution of fuel elements have been developed by fuel suppliers and are now routinely and successfully applied in many countries. For the first time, the subject of control rod poolside examination was dealt with, poolside inspection and repair of a MOX assembly were reported and the inspection and repair of WWER assemblies were examined. Compared to the results of the previous meeting, present developments in the area aim now at reaching better economics, better reliability, reduction of personal doses and waste volume. Thirty-six participants representing twelve countries attended the meeting. Fifteen papers were presented in two sessions. An abstract was prepared for each of these papers. Refs, figs, tabs, diagrams, pictures and photos

  7. The technical concept of a temporary store for fuel elements

    International Nuclear Information System (INIS)

    In the German federal government's opinion, interim storage on the sites of nuclear power plants of spent fuel elements is to minimize the number of transports within Germany. As a span of approximately five years must be bridged until the interim stores now planned and filed for will be commissioned and, at the same time, transport activities are to be reduced, a kind of anticipated interim storage, or temporary storage, on power plant sites is unavoidable. The concept for the temporary storage of spent fuel elements is described in the article. On the basis of this concept, the Neckar Joint Nuclear Power Station recently was awarded a storage permit for nuclear fuels under Sec. 6 of the German Atomic Energy Act. Temporary stores following the same concept have been filed for, and are now in the licensing procedure, for another four sites (Philippsburg, Biblis, Kruemmel, Brunsbuettel). (orig.)

  8. Some parametric flow analyses of a particle bed fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, D.

    1993-05-01

    Parametric calculations are performed, using the SAFSIM computer program, to investigate the fluid mechanics and heat transfer performance of a particle bed fuel element. Both steady-state and transient calculations are included, addressing such issues as flow stability, reduced thrust operation, transpiration drag, coolant conductivity enhancement, flow maldistributions, decay heat removal, flow perturbations, and pulse cooling. The calculations demonstrate the dependence of the predicted results on the modeling assumptions and thus provide guidance as to where further experimental and computational investigations are needed. The calculations also demonstrate that both flow instability and flow maldistribution in the fuel element are important phenomena. Furthermore, results are encouraging that geometric design changes to the element can significantly reduce problems related to these phenomena, allowing improved performance over a wide range of element power densities and flow rates. Such design changes will help to maximize the operational efficiency of space propulsion reactors employing particle bed fuel element technology. Finally, the results demonstrate that SAFSIM is a valuable engineering tool for performing quick and inexpensive parametric simulations addressing complex flow problems.

  9. Gamma scanning of full scale HTR fuel elements

    International Nuclear Information System (INIS)

    Gamma scanning for the determination of burn-up and fission product inventory has been developed at the Dragon Project, suitable for measurements on fuel elements and segments from full-sized integral block elements. This involved the design and construction of a new lead flask with sophisticated collimator design. State-of-the art gamma spectrometric equipment was set up to cope with strong variations of count-rate and high data throughput. Software efforts concentrated on the calculation of the self absorption and absorption corrections in the complicated geometry of multi-hole graphite block segments with a corrugated circumference. The techniques described here are applicable to the non-destructive examination of a wide range of fuel element designs. (author)

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

    International Nuclear Information System (INIS)

    IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm3 for U3Si2-Al dispersion-based and 2.3 gU/cm3 for U3O8-Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm3 in U3Si2-Al dispersion and 3.2 gU/cm3 U3O8-Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U3Si2-Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U3O8-Al dispersion fuel plates with 3.2 gU/cm3 showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U3Si2 production at 4.8 gU/cm3, with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)

  11. On-line elemental analysis of fossil fuel process streams by inductively coupled plasma spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Chisholm, W.P.

    1995-06-01

    METC is continuing development of a real-time, multi-element plasma based spectrometer system for application to high temperature and high pressure fossil fuel process streams. Two versions are under consideration for development. One is an Inductively Coupled Plasma system that has been described previously, and the other is a high power microwave system. The ICP torch operates on a mixture of argon and helium with a conventional annular swirl flow plasma gas, no auxiliary gas, and a conventional sample stream injection through the base of the plasma plume. A new, demountable torch design comprising three ceramic sections allows bolts passing the length of the torch to compress a double O-ring seal. This improves the reliability of the torch. The microwave system will use the same data acquisition and reduction components as the ICP system; only the plasma source itself is different. It will operate with a 750-Watt, 2.45 gigahertz microwave generator. The plasma discharge will be contained within a narrow quartz tube one quarter wavelength from a shorted waveguide termination. The plasma source will be observed via fiber optics and a battery of computer controlled monochromators. To extract more information from the raw spectral data, a neural net computer program is being developed. This program will calculate analyte concentrations from data that includes analyte and interferant spectral emission intensity. Matrix effects and spectral overlaps can be treated more effectively by this method than by conventional spectral analysis.

  12. MAW and HTR fuel element test disposal in boreholes

    International Nuclear Information System (INIS)

    The Kernforschungsanlage Juelich, KFA, (Nuclear Research Center Juelich) has been handling a project since 1983 on 'Further Development of the Borehole Technology for the Disposal of Radioactive Wastes in Salt, with the Examples of Dissolver Sludge, Fuel Element Claddings, Fuel Hardware und HTR Fuel Elements'. The project is sponsored by the Bundesminister fuer Forschung und Technologie, BMFT, (Federal Ministry of Research and Technology) under the identification number KWA 5302 3 and bears the short title 'MAW and HTR Fuel Element Test Disposal in Boreholes'. The major objective of the project is to develop a technique for the disposal of the above mentioned wastes in unlined boreholes in salt and to test this technique in the Asse salt mine. The Institut fuer Chemische Technologie der Nuklearen Entsorgung, ICT (Institute of Chemical Technology) at the KFA is responsible for the scientific and organizational management of the project. The Institut fuer Tieflagerung, IfT, (Institute for Underground Disposal) of the Gesellschaft fuer Strahlen- und Umweltforschung mbH, GSF, (Society for Radiological and Environmental Research) is responsible for the geomechanical and mining activities in the project. It supervises the in-situ experiments, and as the owner of the Asse salt mine, it submits applications for the experiments to the licensing authorities. Geomechanical calculations are being carried out by the Bundesanstalt fuer Geowissenschaften und Rohstoffe, BGR, (Federal Institute for Geological Sciences and Natural Resources). (orig./RB)

  13. HTGR spent fuel element decay heat and source term analysis

    International Nuclear Information System (INIS)

    Decay heat, gamma dose rates, and neutron source strengths were determined for spent fuel elements from a High-Temperature Gas-Cooled Reactor (HTGR). The calculations were based on curie values reported in General Atomic Report GA-A13886 for the earlier commercial version of a 3000-MW(t) HTGR utilizing the thorium-uranium four-year fuel cycle. The reactor core was designed for an average thermal power density of 8.5 watts per cm3 and a carbon-to-thorium atom ratio which varies between 210:1 and 240:1. Calculations of decay heat, gamma dose rates, and neutron source strengths were made for spent fuel elements from the initial core and from representative nonrecycle and recycle reloads. The study was performed for decay times from 180 days to 10 years. Tables of the isotopic results are given for both the fertile and fissile particles in the fuel elements. In addition, ordered tables of the important isotopic contributors are presented. Graphical presentations of the results are shown and discussed; in addition, comparisons are made with previous determinations

  14. Method of locating a leaking fuel element in a fast breeder power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Honekamp, John R. (Downers Grove, IL); Fryer, Richard M. (Idaho Falls, ID)

    1978-01-01

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  15. Method of locating a leaking fuel element in a fast breeder power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Honekamp, J.R.; Fryer, R.M.

    1978-03-21

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of /sup 134/Xe to /sup 133/Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  16. Fine lattice stochastic modeling of particle fuels in HTGR fuel elements

    International Nuclear Information System (INIS)

    There is growing interest worldwide in high temperature gas-cooled reactors (HTGRs) as candidates for next generation reactor systems. Either in a pebble type or in a prismatic type HTGR, coated particle fuel (TRISO fuel) appears to be the most promising fuel candidate to be used. For design and analysis of such a reactor, transport models, in particular, stochastic models that permit the simulation of neutron transport through the stochastic mixture of fuel and moderator materials, are becoming essential and gaining importance. Naturally, the Monte Carlo methods have been used for this situation. However, the methods reported in the literature all have their own deficiencies. In this thesis, we propose a new Monte Carlo method named fine lattice stochastic (FLS) modeling that is distinct from others. This method is based on fine lattice system in which a lattice circumscribes a fuel particle. Once the problem is given, an interface Fortran code gives out the TRISO particle fuel configurations (a set of lattice center points only) for MCNP input. The number of available lattice center points is far larger than the number of fuel particles according to packing fraction of the fuel element. We apply discrete random sampling here to choose a certain number of lattices to fill with fuel particles. In this aspect, FLS modeling allows more realistic fuel particle distributions. In this thesis, only simple cube (SC) structure is used in cubic lattice. However, FLS model can be easily extended to BCC, FCC structures or hexagonal prism type lattice. The criticality calculations for our FLS modeling were first tested on a small cube problem and compared with other models. The results indicate that the new stochastic model is an accurate and efficient approach to analyze TRISO particle fuel configurations. Then the FLS modeling was performed to analyze HTGR fuel elements for both pebble type and prismatic type and the results were also good as expected

  17. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    Science.gov (United States)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  18. Parametric study of thermo-mechanical behaviour of 19-element PHWR fuel bundle having AHWR fuel material

    International Nuclear Information System (INIS)

    AHWR Th-LEU of 4.3 weight % 235U enrichment is a fuel design option for its trial irradiation in Indian PHWRs. The important component of this option is the large enhancement in the average discharge burn-up from the core. A parametric study of the 19-element fuel bundle, with natural uranium currently is being used in all operating 220 MWe PHWRs, has been carried out for AHWR Th-LEU fuel material by computer code FUDA MOD2. The important fuel parameters such as fuel temperature, fission gas release, fuel swelling and sheath strain have been analyzed for required fuel performance. With Th-LEU, average discharge burnups of about 25,000 MW-d/TeHE can be achieved. The FUDA code (Fuel Design Analysis code) MOD2 version has been used in the fuel element analysis. The code takes into account the inter-dependence of different parameters like fuel pellet temperatures, pellet expansions, fuel-sheath gap heat transfer, sheath strain and stresses, fission gas release and gas pressures, fuel densification etc. Thermo-mechanical analysis of fuel element having AHWR material is carried out for the bundle power histories reaching up to design burn-up 40000 MWd/TeHE. The resultant parameters such as fuel temperature, sheath plastic strain and fission gas pressure for AHWR fuel element were compared with respective thermo-mechanical parameters for similar fuel bundle element with natural uranium as fuel material. (author)

  19. Block fuel element for gas-cooled high temperature reactors

    International Nuclear Information System (INIS)

    The invention concerns a block fuel element consisting of only one carbon matrix which is almost isotropic of high crystallinity into which the coated particles are incorporated by a pressing process. This block element is produced under isostatic pressure from graphite matrix powder and coated particles in a rubber die and is subsequently subjected to heat treatment. The main component of the graphite matrix powder consists of natural graphite powder to which artificial graphite powder and a small amount of a phenol resin binding agent are added

  20. Recent operating experience with 28 element fuel at Pickering NGS

    International Nuclear Information System (INIS)

    A review of 28-element fuel operating experience at Pickering NGS is presented. The following topics are discussed: 1. Recent experience with in-core defects and 131I releases; 2. Operating strategies to minimize defect potential or to mitigate 131I releases to the primary heat transport system; 3. Impact of reduced regulatory limits as well as higher corporate expectations on operating strategies. 3 refs., 3 figs., 2 tabs

  1. Dry store for spent fuel elements from nuclear reactors

    International Nuclear Information System (INIS)

    In the dry store for spent fuel elements from nuclear reactors which are enclosed in storage tubes and cooled with air, the storage tubes being arranged in shafts of a storage building, a loading device is provided underneath the shafts and in a cooling air shaft designed for transporting. The loading device therefore requires only a small lifting height and the chances of storage tubes falling from great heights are excluded. This invention is applicable in particular for intermediate stores. (orig./RW)

  2. CARA CVN: inherently safe fuel element for PHWR power plants

    International Nuclear Information System (INIS)

    This paper presents design alternatives of the CARA fuel element with negative void reactivity coefficient (CVN) enhancing the PHWR safety for L-LOCA sequences. This design enhances the safety and the operation performance in Atucha and Embalse without changes in the operation conditions. This new design balances wide performance margins of CARA SEU 0.9% previous design, with new intrinsic safety requirements without economic penalties. (author)

  3. METHOD AND APPARATUS FOR EXAMINING FUEL ELEMENTS FOR LEAKAGE

    Science.gov (United States)

    Smith, R.R.; Echo, M.W.; Doe, C.B.

    1963-12-31

    A process and a device for the continuous monitoring of fuel elements while in use in a liquid-metal-cooled, argonblanketed nuclear reactor are presented. A fraction of the argon gas is withdrawn, contacted with a negative electrical charge for attraction of any alkali metal formed from argon by neutron reaction, and recycled into the reactor. The electrical charge is introduced into water, and the water is examined for radioactive alkali metals. (AEC)

  4. Convective parameters in fuel elements for research nuclear reactors

    International Nuclear Information System (INIS)

    The study of a prototype for the simulation of fuel elements for research nuclear reactors by natural convection in water is presented in this paper. This project is carry out in the thermofluids laboratory of National Institute of Nuclear Research. The fuel prototype has already been test for natural convection in air, and the first results in water are presented in this work. In chapter I, a general description of Triga Mark III is made, paying special atention to fuel-moderator components. In chapter II and III an approach to convection subject in its global aspects is made, since the intention is to give a general idea of the events occuring around fuel elements in a nuclear reactor. In chapter II, where an emphasis on forced convection is made, some basic concepts for forced convection as well as for natural convection are included. The subject of flow through cylinders is annotated only as a comparative reference with natural convection in vertical cylinders, noting the difference between used correlations and the involved variables. In chapter III a compilation of correlation found in the bibliography about natural convection in vertical cylinders is presented, since its geometry is the more suitable in the analysis of a fuel rod. Finally, in chapter IV performed experiments in the test bench are detailed, and the results are presented in form of tables and graphs, showing the used equations for the calculations and the restrictions used in each case. For the analysis of the prototypes used in the test bench, a constant and uniform flow of heat in the whole length of the fuel rod is considered. At the end of this chapter, the work conclusions and a brief explanation of the results are presented (Author)

  5. The element technology of clean fuel alcohol plant construction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D.S; Lee, D.S. [Sam-Sung Engineering Technical Institute (Korea, Republic of); Choi, C.Y [Seoul National University, Seoul (Korea, Republic of)] [and others

    1996-02-01

    The fuel alcohol has been highlighted as a clean energy among new renewable energy sources. However, the production of the fuel alcohol has following problems; (i)bulk distillate remains is generated and (ii) benzene to be used as a entertainer in the azeotropic distillation causes the environmental problem. Thus, we started this research on the ground of preserving the cleanness in the production of fuel alcohol, a clean energy. We examined the schemes of replacing the azotropic distillation column which causes the problems with MSDP(Molecular Sieve Dehydration Process) system using adsorption technology and of treating the bulk distillate remains to be generated as by-products. In addition, we need to develop the continuous yea station technology for the continuous operation of fuel alcohol plant as a side goal. Thus, we try to develop a continuous ethanol fermentation process by high-density cell culture from tapioca, a industrial substrate, using cohesive yeast. For this purpose, we intend to examine the problem of tapioca, a industrial substrate, where a solid is existed and develop a new process which can solve the problem. Ultimately, the object of this project is to develop each element technology for the construction of fuel alcohol plant and obtain the ability to design the whole plant. (author) 54 refs., 143 figs., 34 tabs.

  6. Fabrication of spherical fuel element for 10 MW high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Cold quasi-isostatic molding with a silicon rubber die was used for manufacturing the spherical fuel elements of 10 MW high temperature gas-cooled reactor. 44 batches of fuel elements, about 20540 of the fuel elements, were produced. The cold properties of the graphite matrix materials satisfies the design specifications. The mean free uranium fraction in spherical fuel element from 44 batches is 4.57 x 10-5, certified products is 99%

  7. Gamma-ray spectroscopy on irradiated MTR fuel elements

    Science.gov (United States)

    Terremoto, L. A. A.; Zeituni, C. A.; Perrotta, J. A.; da Silva, J. E. R.

    2000-08-01

    The availability of burnup data is an important requirement in any systematic approach to the enhancement of safety, economics and performance of a nuclear research reactor. This work presents the theory and experimental techniques applied to determine, by means of nondestructive gamma-ray spectroscopy, the burnup of Material Testing Reactor (MTR) fuel elements irradiated in the IEA-R1 research reactor. Burnup measurements, based on analysis of spectra that result from collimation and detection of gamma-rays emitted in the decay of radioactive fission products, were performed at the reactor pool area. The measuring system consists of a high-purity germanium (HPGe) detector together with suitable fast electronics and an on-line microcomputer data acquisition module. In order to achieve absolute burnup values, the detection set (collimator tube+HPGe detector) was previously calibrated in efficiency. The obtained burnup values are compared with ones provided by reactor physics calculations, for three kinds of MTR fuel elements with different cooling times, initial enrichment grades and total number of fuel plates. Both values show good agreement within the experimental error limits.

  8. Subcutaneous granuloma annulare

    Directory of Open Access Journals (Sweden)

    Dhar Sandipan

    1994-01-01

    Full Text Available Two cases of subcutaneos granuloma annulare are reported. Clinical presentation was in the form of hard subcutaneous nodules; histopathology confirmed the clinical diagnosis. The cases were unique because of onset in adult hood, occurrence over unusual sites and absence of classical lesions of granuloma annulare elsewhere.

  9. Subcutaneous granuloma annulare

    Directory of Open Access Journals (Sweden)

    Dhar Sandipan

    1993-01-01

    Full Text Available Two cases of subcutaneous granuloma annulare are reported. Clinical presentation was in the form of hard subcutaneous nodules, histopathology confirmed the clinical diagnosis. The cases were unique because of onset in adult age, occurrence over unusual sites and absence of classical lesions of granuloma annulare elsewhere.

  10. The fuel element situation at the TRIGA mark II reactor Vienna

    International Nuclear Information System (INIS)

    The fuel history, spent fuel storage situation and recent problems covering the period from 1962 until 1.6.2001 were reviewed. After almost 40 years of TRIGA MARK II reactor Vienna operation, it must be mentioned that the experience with TRIGA fuel elements was and is excellent. During this period only 9 fuel elements had to be permanently be removed from the core and 57 fuel elements from the initial start-up are still used in the core. A careful fuel management and a frequent fuel inspection is of most importance, fuel elements should be moved at least two-times a year from their core position to check free movement and a 180 deg. rotation of the fuel element is also recommended (nevyjel)

  11. Polarly anisotropic thermoelasticity of cylindrical and spherical fuel elements

    International Nuclear Information System (INIS)

    This paper deals with the solution for principal thermally induced stress in log solid and hollow rods and balls, taking onto account not only surface pressure load and internal heat generation, but also non equal elastic parameters and thermal strain. Closed form solutions obtained for circumferentially reinforced bodies are complementary to recently published formulae for transversely isotropic thermoelasticity of cylinders and spheres. Numerical test is gi ven using typical data for the ceramic fuel of roll and ball shape elements for high temperature gas cooled nuclear reactors, but similar values develop in the pressurised light water reactors. (author)

  12. Charging machine for the transport of fuel elements

    International Nuclear Information System (INIS)

    Charging machines for the transport of fuel elements for nuclear reactors have got a bridge body supported by two parallel rails via wheels. According to the invention the wheels are fixed to the bridge body by means of guide rods in such a way that at least relative movements in direction of the wheels and transversal to it are possible. Parallel to the guide rods springs and movement attenuators are force-locking by connected. Therefore a stabilizing effect with respect to the transversal forces occurring during earthquakes is achieved. (orig.)

  13. Storage system and method for spent fuel elements

    International Nuclear Information System (INIS)

    The proposal concerns an additional protection against leakage of a FE-transport container for interim storage of spent fuel elements. The gastight container has a second cover placed at a short distance from the first cover. The intermediate hollow space can be connected with a measuring system which indicates if part of the trace gas (mostly helium) added as indicator has escaped from the container due to leakage. The description explains the method and the assembly of required lines and measuring points etc. (UWI)

  14. Bending of fuel fast reactor fuel elements under action of non-uniform temperature gradients and radiation-induced swelling

    International Nuclear Information System (INIS)

    The bending of rod fuel elements in gas-cooled fast reactors under the action of temperature gradients radiation-induced swelling non-uniform over the perimeter of fuel cans is evaluated. It is pointed out that the radiation-induced swelling gives the main contribution to the bending of fuel elements. Calculated data on the bending of the corner fuel element in the assembly of the fast reactor with dissociating gas coolant are given. With the growth of temperature difference over the perimeter, the bending moment and deformation increase, resulting in the increase of axial stresses. The obtained data give the basis for accounting the stresses connected with thermal and radiation bending when estimating serviceability of fuel elements in gas cooled fast reactors. Fuel element bending must be also taken into account when estimating the thermal hydrualic properties

  15. Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Trammell, Michael P [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Qualls, A L [ORNL; Harrison, Thomas J [ORNL

    2013-01-01

    ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

  16. Comparative analysis of C A R A fuel element in argentinean PHWR Argentinas

    International Nuclear Information System (INIS)

    This paper presents an analysis of the thermal mechanical behaviour, fuel consumption and economical estimations of the CARA fuel element in the Atucha and Embalse nuclear power plants, compared with the present fuel performance.The present results show that the expect profit by the use of the CARA fuel element in our reactor guaranties the recovery of fund for its development. Likewise it reduces the number of spent fuel to be storage and treated

  17. Study on Unigraphics Drawing Modeling Method for 37-Element and CANFLEX Fuel Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Yu Mi; Park, Joo Hwan

    2010-03-15

    The CANFLEX bundle contains 43 elements of two different diameters. It has two rings of small diameter elements on the outside, and eight elements (with diameter slightly larger than those in the standard 37-Element bundle) in the center. This larger number of small diameter elements on the outside of the CANFLEX bundle enhances thermo-hydraulic capability, resulting in a higher power capability and an improvement in operating safety margins. As a Result of advanced fuel design for CANFLEX fuel bundles, components consisting of fuel bundles are more complicated. Hence, the detailed modeling of components is inevitable in order to analyze the fuel performance by computational fluid dynamics. In this report, the basic design of the advanced fuel for CANDU reactors was carried out and the methodology for the modeling of fuel bundle were described. Firstly, the components consisting of fuel bundles were separately modeled and saved with different file names. The final feature of fuel bundle was accomplished by an assembling process of components. Since this report developed the modeling methodology based on the Unigraphics program, the basic explanations for the software were given first, and the complete modeling of 37-elements and CANFLEX fuel bundles were provided. The components of CANFLEX fuel bundles were also compared with that of 37-elements fuel bundles. Although, in this report, the modeling methodology is applied only to 37-elements and CANFLEX fuel bundles, this methodology may be applicable to the newly designed fuel bundles which are to be developed in the future

  18. Annular core for modular high temperature gas-cooled reactor (MHTGR)

    International Nuclear Information System (INIS)

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40 % greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93 m tall. Fuel elements contain TRISO-coated microspheres of 19.8 % enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above. (author)

  19. Graphitic matrix materials for spherical HTR fuel elements

    International Nuclear Information System (INIS)

    The present report comprises the essential results of material development and irradiation testing of graphitic matrix materials for spherical HTR fuel elements and completes the documentation of the irradiation data for 20 matrix materials (Juel-1702). The main emphasis is given to the matrices A3-3 (standard matrix) and A3-27 (matrix synthesized resin), both of which are being used as structural materials for the fuel elements of the AVR and the THTR respectively. In addition, comparisons are made between 18 A3-variants and the standard matrix A3-3. It is shown that three of the variants come into question as a potential for use. The results described were obtained in the framework of the HTR project 'Hochtemperaturreaktor-Brennstoffkreislauf' (HBK), in which are involved the Gesellschaft fuer Hochtemperaturreaktor-Technik mbH, Hochtemperaturreaktor-Brennelemente GmbH, Hochtemperatur-Reaktorbau GmbH, Kernforschungsanlage Juelich GmbH, NUKEM GmbH, and Sigri Elektrographit GmbH/Ringsdorff-Werke GmbH. The project is sponsored by the 'Bundesministerium fuer Forschung und Technologie' and by the state of 'Nordrhein-Westfalen'. (orig.)

  20. Quality control of CANDU6 fuel element in fabrication process

    International Nuclear Information System (INIS)

    To enhance the fine control over all aspects of the production process, improve product quality, fuel element fabrication process for CANDU6 quality process control activities carried out by professional technical and management technology combined mode, the quality of the fuel elements formed around CANDU6 weak links - - end plug , and brazing processes and procedures associated with this aspect of strict control, in improving staff quality consciousness, strengthening equipment maintenance, improved tooling, fixtures, optimization process test, strengthen supervision, fine inspection operations, timely delivery carry out aspects of the quality of information and concerns the production environment, etc., to find the problem from the improvement of product quality and factors affecting the source, and resolved to form the active control, comprehensive and systematic analysis of the problem of the quality management concepts, effectively reducing the end plug weld microstructure after the failure times and number of defects zirconium alloys brazed, improved product quality, and created economic benefits expressly provided, while staff quality consciousness and attention to detail, collaboration department, communication has been greatly improved and achieved very good management effectiveness. (authors)

  1. Triaxial Swirl Injector Element for Liquid-Fueled Engines

    Science.gov (United States)

    Muss, Jeff

    2010-01-01

    A triaxial injector is a single bi-propellant injection element located at the center of the injector body. The injector element consists of three nested, hydraulic swirl injectors. A small portion of the total fuel is injected through the central hydraulic injector, all of the oxidizer is injected through the middle concentric hydraulic swirl injector, and the balance of the fuel is injected through an outer concentric injection system. The configuration has been shown to provide good flame stabilization and the desired fuel-rich wall boundary condition. The injector design is well suited for preburner applications. Preburner injectors operate at extreme oxygen-to-fuel mass ratios, either very rich or very lean. The goal of a preburner is to create a uniform drive gas for the turbomachinery, while carefully controlling the temperature so as not to stress or damage turbine blades. The triaxial injector concept permits the lean propellant to be sandwiched between two layers of the rich propellant, while the hydraulic atomization characteristics of the swirl injectors promote interpropellant mixing and, ultimately, good combustion efficiency. This innovation is suited to a wide range of liquid oxidizer and liquid fuels, including hydrogen, methane, and kerosene. Prototype testing with the triaxial swirl injector demonstrated excellent injector and combustion chamber thermal compatibility and good combustion performance, both at levels far superior to a pintle injector. Initial testing with the prototype injector demonstrated over 96-percent combustion efficiency. The design showed excellent high -frequency combustion stability characteristics with oxygen and kerosene propellants. Unlike the more conventional pintle injector, there is not a large bluff body that must be cooled. The absence of a protruding center body enhances the thermal durability of the triaxial swirl injector. The hydraulic atomization characteristics of the innovation allow the design to be

  2. Neutron spectrum and radial power distribution measurements in a TRIGA reactor fuel element

    International Nuclear Information System (INIS)

    The neutron spectrum in the Illinois Advanced TRIGA Reactor was measured by a crystal spectrometer utilizing an LiF(1, 1, 1) crystal monochromator whose reflectivity was determined experimentally. The fission heat source distribution in a fuel element was also determined as a function of the fuel element temperature. These two measurements were used to investigate the effects of fuel element temperature and the local core loading on the thermal diffusion length in a fuel element. Changes in the thermal diffusion lengths during a reactor pulse underlie the proposed temperature feedback mechanism for the ZrH fuel material. The results of the measurements confirm, in part, this proposed temperature feedback mechanism

  3. Corrosion studies in fuel element reprocessing environments containing nitric acid

    International Nuclear Information System (INIS)

    Nitric acid is universally used in aqueous fuel element reprocessing plants; however, in the processing scheme being developed by the Consolidated Fuel Reprocessing Program, some of the equipment will be exposed to nitric acid under conditions not previously encountered in fuel element reprocessing plants. A previous report presented corrosion data obtained in hyperazeotropic nitric acid and in concentrated magnesium nitrate solutions used in its preparation. The results presented in this report are concerned with the following: (1) corrosion of titanium in nitric acid; (2) corrosion of nickel-base alloys in a nitric acid-hydrofluoric acid solution; (3) the formation of Cr(VI), which enhances corrosion, in nitric acid solutions; and (4) corrosion of mechanical pipe connectors in nitric acid. The results show that the corrosion rate of titanium increased with the refreshment rate of boiling nitric acid, but the effect diminished rapidly as the temperature decreased. The addition of iodic acid inhibited attack. Also, up to 200 ppM of fluoride in 70% HNO3 had no major effect on the corrosion of either titanium or tantalum. In boiling 8 M HNO3-0.05 M HF, Inconel 671 was more resistant than Inconel 690, but both alloys experienced end-grain attack. In the case of Inconel 671, heat treatment was very important; annealed and quenched material was much more resistant than furnace-cooled material.The rate of oxidation of Cr(III) to Cr(VI) increased significantly as the nitric acid concentration increased, and certain forms of ruthenium in the solution seemed to accelerate the rate of formation. Mechanical connectors of T-304L stainless steel experienced end-grain attack on the exposed pipe ends, and seal rings of both stainless steel and a titanium alloy (6% Al-4% V) underwent heavy attack in boiling 8 M HNO3

  4. Specifications for high flux isotope reactor fuel elements HFIR-FE-3

    International Nuclear Information System (INIS)

    This specification covers requirements for two types of aluminum-base fuel elements which together will be used as the fuel assembly in the High Flux Isotope Reactor (HFIR). Requirements are included for materials of construction, fabrication, assembly, inspection, and quality control to produce fuel elements in accordance with Company drawings

  5. BOTHER: a steady-state code that predicts margin to burnout heat flux for N-Reactor fuel elements

    International Nuclear Information System (INIS)

    In order to operate a nuclear reactor safely, some method must be available which can adequately describe the thermal-hydraulics of the reactor core. Further, some method must be available which can be used to predict the effects of changes in system operation. For example it is often necessary to know or be able to predict the effects of reduced coolant flow, front or rear peaked power distribution, etc., on the overall safe operation of the reactor. Because of the uniqueness of the N Reactor (horizontal pressure tubes with no crossflow between tubes or annular subchannels) the commonly available thermal-hydraulics codes are generally not directly applicable. For these reasons the BOTHER (BurnOut THErmal Ratio) computer code has been developed at UNI. Using experimental results for N Reactor flow splits and heat splits as well as enthalpy imbalance and critical heat flux data, BOTHER computes the steady state margin to burnout for N Reactor fuel elements. The equations used by BOTHER to perform the burnout calculations are described. A sample problem for MARK-IV fuel with input and output listings is also included

  6. Annular pancreas (image)

    Science.gov (United States)

    Annular pancreas is an abnormal ring or collar of pancreatic tissue that encircles the duodenum (the part of the ... intestine that connects to stomach). This portion of pancreas can constrict the duodenum and block or impair ...

  7. A combined wet/dry sipping cell for TRIGA fuel element tests

    International Nuclear Information System (INIS)

    A combined wet/dry sipping cell for the investigation of research reactor fuel elements was developed and tested. It is capable of detecting temperature-dependent cladding failures through the release of gaseous fission products. Several TRIGA fuel elements were tested both in the wet in the dry sipping mode. Some elements released fission gases only above 75deg C. (orig.)

  8. Determine the homogeneity of UO2 distribution in HTGR fuel element by X-ray tomography

    International Nuclear Information System (INIS)

    The homogeneity of UO2 distribution in HTGR fuel elements is one of the important properties of a fuel element. The X-ray tomography and image processing technology can nondestructively determine the distribution of UO2 particles in a spherical fuel element. A statement was specifically made on tomography ambiguity, otherwise, the technology and point of tomography to a spherical fuel element was described too. Through the computerized image processing, tomography film shows directly homogeneity of UO2 particles. The hardware for image processing and the special software packing were briefly introduced. As the results of measurement, the colour cross-images drawn by a typing machine or taken from from CRT by a camera indicated that determination homogeneity of UO2 particles in a fuel element homogeneity of UO2 particles in a fuel element sphere by tomography was very successful

  9. Burn-up measurements at TRIGA fuel elements containing strong burnable poison

    International Nuclear Information System (INIS)

    The reactivity method of determining the burn-up of research reactor fuel elements is applied to the highly enriched FLIP elements of TRIGA reactors. In contrast to other TRIGA fuel element types, the reactivity of FLIP elements increases with burn-up due to consumption of burnable poison. 33 fuel elements with burn-up values between 3% and 14% were investigated. The experiments showed that variations in the initial fuel composition significantly influence the reactivity and, consequently, increase the inaccuracy of the burn-up measurements. Particularly important are variations in the initial concentration of erbium, which is used as burnable poison in FLIP fuel. A method for reducing the effects of the material composition variations on the measured reactivity is presented. If it is applied, the accuracy of the reactivity method for highly poisoned fuel elements becomes comparable to the accuracy of other methods for burn-up determination. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-05-01

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

  11. An analytical assessment of the longitudinal ridging of CANDU type fuel element

    International Nuclear Information System (INIS)

    There are 380 fuel channels in a CANDU-6 reactor, and twelve fuel bundles are loaded into each fuel channel. High-pressure, heavy water coolant passes through the fuel bundle string to remove heat generated from the fuel. Fuel sheath collapses down around the uranium dioxide pellet due to the coolant pressure when the fuel is loaded into the reactor. Longitudinal ridges may form in CANDU fuel element sheaths as a result of sheath collapse onto the pellets. A static analysis, finite-element (FE) model was developed to simulate the longitudinal ridging of the fuel element with use of the structural analysis computer code ABAQUS. Collapse pressures were calculated for the fifty-one cases for which test results of WCL in 1973 and 1975 are available. Calculation results under-predicted the critical collapse pressure but it showed significant relationship against test results

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

    International Nuclear Information System (INIS)

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

  13. Subchannel analysis of CANDU 37-element fuel bundles

    International Nuclear Information System (INIS)

    The subchannel analysis codes COBRA-IV and ASSERT-4 have been used to predict the mass and enthalpy imbalance within a CANDU 37-element fuel channel under various system conditions. The objective of this study was to assess the various capabilities of the ASSERT code and highlight areas where further validation or development may be needed. The investigation indicated that the ASSERT code has all the basic models required to accurately predict the flow and enthalpy imbalance for complex rod bundles. The study also showed that the code modelling of void drift and diffusion requires refinement to some coefficients and that further validation is needed at high flow rate and high void fraction conditions, where ASSERT and COBRA are shown to predict significantly different trends. The results of a recent refinement of ASSERT modelling are also discussed

  14. Process for finding defective fuel element cans using ultrasonics

    International Nuclear Information System (INIS)

    As the distance between the test heads can change due to the gassing of fuel elements and as direct ultrasonic echos are assessed as rotary echos, the invention proposes that the expected time range of each can to be tested is determined afresh, before this can is placed between the test heads. The running time of the ultrasonic signals is therefore measured in the intermediate space between the last tested can and the can to be tested next. A constant value is subtracted from this value of running time, which is selected so that the ultrasonic signal received for the running time measurement just does not fall within the expected range. (orig./HP)

  15. Nondestructive testing in fabrication of zirconium alloy tubes and PHWR fuel elements in India

    International Nuclear Information System (INIS)

    The methods and technical means for nondestructive testing, applied at the Nuclear Fuel Complex (Hyderabad, India) in the process of fabricating channel, colander and shell tubes from zirconium alloy and fuel elements with the UO2 fuel for reactor cores of the PHWR-Candu power reactors are described in the review. The significant works on improving the methodology and equipment for ultrasonic quality control of the contact joint welding of fuel elements are noted

  16. FUEL ASSAY REACTOR

    Science.gov (United States)

    Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

    1962-12-25

    A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

  17. Radiation and reflection acoustical fields of an annular phased array

    Institute of Scientific and Technical Information of China (English)

    LAI Puxiang; ZHANG Bixing; WANG Chenghao

    2007-01-01

    The characteristics of the radiation and reflection acoustical fields of an annular phased array are investigated. The effects of the element number, element radius, interelement spacing, centre frequency, focus position, and other parameters on the radiation acoustical field of the annular phased array is theoretically studied. In experiment, an annular transducer with 8 equal-area elements is designed and fabricated, and a series of experimental measurements are conducted. The radiation acoustical field and its reflection on a liquid-solid interface are theoretically and experimentally studied. The experimental result is in good agreement with the theoretical one.

  18. Thermionic Fuel Element performance: TFE Verification Program. Final test report

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The program objective is to demonstrate the technology readiness of a Thermionic Fuel Element (TFE) suitable for use as the basic element in a thermionic reactor with electric power output in the 0.5 to 5.0 MW(e) range, and a full power life of 7 years. A TFE was designed that met the reliability and lifetime requirements for a 2 MW(e) conceptual reactor design. Analysis showed that this TFE could be used over the range of 0.5 to 5 megawatts. This was used as the basis for designing components for test and evaluation. The demonstration of a 7-year component lifetime capability was through the combined use of analytical models and accelerated, confirmatory tests in a fast test reactor. Iterative testing was performed in which the results of one test series led to evolutionary improvements in the next test specimens. The TFE components underwent screening and initial development testing in ex-reactor tests. Several design and materials options were considered for each component. As screening tests permitted, down selection occurred to very specific designs and materials. In parallel with ex-reactor testing, and fast reactor component testing, components were integrated into a TFE and tested in the TRIGA test reactor at GA. Realtime testing of partial length TFEs was used to test support, alignment and interconnective TFE components, and to verify TFE performance in-reactor with integral cesium reservoirs. Realtime testing was also used to verify the relation between TFE performance and fueled emitter swelling, to test the durability of intercell insulation, to check temperature distributions, and to verify the adequacy over time of the fission gas venting channels. Predictions of TFE lifetime rested primarily on the accelerated component testing results, as correlated and extended to realtime by the use of analytical models.

  19. The behaviour of fission products in fuel elements of the AVR-core

    International Nuclear Information System (INIS)

    The fuel elements of the THTR-1-type-, THTR-2-type-, and CFB-2-type reactor have an extraordinary retention capacity for all fission products. Pressed-carbide fuel elements, however, at a temperature above 3000C release considerable Sr- and Eu-activities as a result of diffusion through intact PyC-layers. For GLE-1 and GFB-1 fuel elements a considerable release of cesium has been observed which is caused by defective coated particles. (DG)

  20. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  1. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition as part of a fuel meat thickness optimization effort for reactor performance other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  2. Compaction test of fuel element claddings (hulls) and fuel structures waste. 1

    International Nuclear Information System (INIS)

    The PNC (Power and Nuclear Fuel Development Corporation) plans a hull (fuel element cladding) and endpiece hardware compaction facility in Tokai Works. These tests are carried out to confirm the compaction method to be applied. A Series of tests consist of selection of 'simulated hull material', 'design of capsule', 'correlation between pressure and volume reduction ratio', 'estimation of the disk condition' and 'release of zircalloy-fines from the disk during the compressing process'. The results of these tests are as follows. 1) Non-annealing zircalloy for simulated hull material. 2) Optimization of the capsule design. 3) About 80 wt % of the theoretical zircalloy density at 390 MPa pressure. 4) No large void in the disk without cutting the endpiece. 5) The scattering zircalloy fines volume is about 30ppb by the pressing treatment. This test confirm the compaction to be applied. (author)

  3. Economical analysis to utilize MTR fuel elements using silicides in research reactors

    International Nuclear Information System (INIS)

    According to international programs on reducing enrichment in research reactors and the necessity to maintain their operation, new fuel elements have been developed in order to meet both objectives. Thus, U-Si alloy fuel elements for research reactors are becoming of greater interest for the international markets. It became necessary to make an economic study about the convenience of introducing this type of fuel elements in the RA-3 reactor and to know the potentiality of this fuel. The economical behavior of the reactor operation has been evaluated comparing the actual U3O8 nuclear fuel cycle with U3Si2 nuclear fuels. Results obtained show that the main economical factor to determine the change of fuels is the cost of fabrication, and the change is advisable up to an 80% difference. The other factors related to the cost of nuclear fuel cycle are not relevant or have real minor impacts. (author)

  4. Drying results of K-Basin fuel element 0309M (Run 3)

    International Nuclear Information System (INIS)

    An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-West Basin was subjected to a combination of low- and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of spent nuclear fuel elements removed from both the K-West and K-East Basins. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from K-West canister 0309M during the second fuel selection campaign, conducted in 1996, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. The fuel element was broken in two pieces, with a relatively clean fracture, and the larger piece was tested. A gray/white coating was observed. This was the first test of a damaged fuel element in the furnace. K-West canisters can hold up to seven complete fuel assemblies, but, for purposes of this report, the element tested here is designated as Element 0309M. Element 0309M was subjected to drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step

  5. Instrument for measuring a burnup degree of reactor fuel elements by gamma-scanning method

    Energy Technology Data Exchange (ETDEWEB)

    Jozefowicz, E.T.; Kilim, S.; Lewicki, K.; Rusinowski, Z. (Institute of Nuclear Research, Warsaw (Poland))

    1979-01-01

    The principle of gamma scanning of fuel elements is presented. The measuring instrument built in the Institute of Nuclear Research at Swierk and used for gamma scanning of EK-10, WWR-SM, MR and WWER fuel elements is described. The technical drawing of the instrument is given.

  6. Study of Tower Reactor Fuel Elements Based on Sintered Uranium Dioxide

    International Nuclear Information System (INIS)

    The paper gives the results of loop tests on a large batch of experimental fuel elements based on sintered uranium dioxide. Generalized data on the operation of fuel elements used in the reactors of the icebreaker ''Lenin'' are also included. (author)

  7. Nonuniform Oxidation on the Surface of Fuel Element in HTR

    Directory of Open Access Journals (Sweden)

    Peng Liu

    2016-01-01

    Full Text Available The graphite oxidation of fuel element has obtained high attention in air ingress accident analysis of high temperature gas-cooled reactor (HTR. The shape function, defined as the relationship between the maximum and the average of the oxidation, is an important factor to estimate the consequence of the accident. There are no detailed studies on the shape function currently except two experiments several decades ago. With the development of computer technology, CFD method is used in the numerical experiment about graphite oxidation in pebble bed of HTR in this paper. Structured packed beds are used in the calculation instead of random packed beds. The result shows the nonuniform distribution of oxidation on the sphere surface and the shape function in the condition of air ingress accident. Furthermore, the sensitive factors of shape function, such as temperature and Re number, are discussed in detail and the relationship between the shape function and sensitive factors is explained. According to the results in this paper, the shape function ranges from 1.05 to 4.7 under the condition of temperature varying from 600°C to 1200°C and Re varying from 16 to 1600.

  8. Experimental approach and modelling of the mechanical behaviour of graphite fuel elements subjected to compression pulses

    OpenAIRE

    Forquin P.

    2010-01-01

    Among the activities led by the Generation IV International Forum (GIF) relative to the future nuclear systems, the improvement of recycling of fuel elements and their components is a major issue. One of the studied systems by the GIF is the graphite-moderated high-temperature gas cooled reactor (HTGR). The fuel elements are composed of fuel roads half-inch in diameter named compacts. The compacts contain spherical particles made of actinide kernels about 500 m in diameter coated with t...

  9. Calculation of fuel element temperature TRIGA 2000 reactor in sipping test tubes using CFD

    International Nuclear Information System (INIS)

    It has been calculated the fuel element temperature in the sipping test of Bandung TRIGA 2000 reactor. The calculation needs to be done to ascertain that the fuel element temperatures are below or at the limit of the allowable temperature fuel elements during reactor operation. ensuring that the implementation of the test by using this device, the temperature is still within safety limits. The calculation is done by making a model sipping test tubes containing a fuel element surrounded by 9 fuel elements. according to the position sipping test tubes in the reactor core. by using Gambit. Dimensional model adapted to the dimensions of the tube and the fuel element in the reactor core of Bandung TRIGA 2000 reactor. Sipping test Operation for each fuel element performed for 30 minutes at 300 kW power. Calculations were performed using CFD software and as input adjusted parameters of TRIGA 2000 reactor. Simulations carried out on the operation of the 30, 60, 90, 120, 150, 180 and 210 minutes. The calculation result shows that the temperature of the fuel in tubes sipping test of 236.06 °C, while the temperature of the wall is 87.58 °C. The maximum temperature in the fuel center of TRIGA 2000 reactor in normal operation is 650 °C. and the boiling is not allowed in the reactor. So it can be concluded that the operation of the sipping test device are is very safe because the fuel center temperature is below the temperature limits the allowable fuel under normal operating conditions as well as the fuel element wall temperature is below the boiling temperature of water. (author)

  10. Detonation Initiation by Annular Jets and Shock Waves

    OpenAIRE

    Shepherd, Joseph E.

    2005-01-01

    The objective of this research is to experimentally determine the feasibility of initiating detonation in fuel-air mixtures using only the energy in hot, compressed air. The existing 6-inch shock tube at Caltech was used to create hot, high-pressure air behind a reflected shock wave. The hot air created an imploding annular shock wave when it jetted through an annular orifice into a 76-mm-diameter, 1-m-long tube attached to the end of the shock tube. A special test section with an annular ...

  11. Generalized granuloma annulare

    Directory of Open Access Journals (Sweden)

    Khatri M

    1995-01-01

    Full Text Available A 35-years-old female patient had generalized pruritic papular lesions, distributed like dermatitis herpetiformis for last 4 years. Histopathologic changes were typical of granuloma annulare with negative results of direct immunofluorescence. The patient did not have association of diabetes mellitus or any other systemic disease. She failed to respond to dapsone therapy and 13-cis-retinoic acid.

  12. Annular Planar Monopole Antennas

    OpenAIRE

    Chen, Z. N.; Ammann, Max; Chia, W.Y. W.; See, T.S. P.

    2002-01-01

    A type of annular planar monopole antenna is presented. The impedance and radiation characteristics of the monopole with different holes and feed gaps are experimentally examined. The measured results demonstrate that the proposed antenna is capable of providing significantly broad impedance bandwidth with acceptable radiation performance.

  13. Drying Results of K-Basin Fuel Element 2660M (Run 7)

    International Nuclear Information System (INIS)

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the seventh of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 2660M. This element (referred to as Element 2660M) was stored underwater in the K-West Basin from 1983 until 1996. Element 2660M was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

  14. Drying Results of K-Basin Fuel Element 6513U (Run 8)

    International Nuclear Information System (INIS)

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL)on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the eighth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 6513U. This element (referred to as Element 6513U) was stored underwater in the K-West Basin from 1983 until 1996. Element 6513U was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0 and discussed in Section 6.0

  15. Effect of annular secondary conductor in a linear electromagnetic stirrer

    Indian Academy of Sciences (India)

    R Madhavan; V Ramanarayanan

    2008-10-01

    This paper presents the variation of average axial force density in the annular secondary conductor of a linear electromagnetic stirrer. Different geometries of secondaries are considered for numerical and experimental validation namely, 1. hollow annular ring, 2. annular ring with a solid cylinder and 3. solid cylinder. Experimental and numerical simulations are performed for a 2-pole in house built 15 kW linear electromagnetic stirrer (EMS). It is observed for a supply current of 200 A at 30 Hz the force densities in the hollow annular ring is 67% higher than the equivalent solid cylinder. The same values are 33% for annular ring with a solid cylinder. Force density variation with supply frequency and current are also reported. Numerical simulations using finite element model are validated with experimental results.

  16. Preliminary Nuclear Analysis for the HANARO Fuel Element with Burnable Absorber

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Kim, So Young; In, Won Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Burnable absorber is used for reducing reactivity swing and power peaking in high performance research reactors. Development of the HANARO fuel element with burnable absorber was started in the U-Mo fuel development program at HANARO, but detailed full core analysis was not performed because the current HANARO fuel management system is uncertain to analysis the HANARO core with burnable absorber. A sophisticated reactor physics system is required to analysis the core. The McCARD code was selected and the detailed McCARD core models, in which the basic HANARO core model was developed by one of the McCARD developers, are used in this study. The development of nuclear fuel requires a long time and correct developing direction especially by the nuclear analysis. This paper presents a preliminary nuclear analysis to promote the fuel development. Based on the developed fuel, the further nuclear analysis will improve reactor performance and safety. Basic nuclear analysis for the HANARO and the AHR were performed for getting the proper fuel elements with burnable absorber. Addition of 0.3 - 0.4% Cd to the fuel meat is promising for the current HANARO fuel element. Small addition of burnable absorber may not change any fuel characteristics of the HANARO fuel element, but various basic tests and irradiation tests at the HANARO core are required.

  17. Fate of Cu, Cr, As and some other trace elements during combustion of recovered waste fuels

    OpenAIRE

    Lundholm, Karin

    2007-01-01

    The increased use of biomass and recovered waste fuels in favor of fossil fuels for heat and power production is an important step towards a sustainable future. Combustion of waste fuels also offers several advantages over traditional landfilling, such as substantial volume reduction, detoxification of pathological wastes, and reduction of toxic leaches and greenhouse gas (methane) formation from landfills. However, combustion of recovered waste fuels emits more harmful trace elements than co...

  18. Pellet-clad interaction observations in boiling water reactor fuel elements

    International Nuclear Information System (INIS)

    Under a programme to assess the performance of fuel elements of Tarapur Atomic Power Station, post-irradiation examination has been carried out on 18 fuel elements in the first phase. Pellet-clad mechanical interaction behaviour in 14 elements with varying burnup and irradiation history has been studied using eddy current testing technique. The data has been analysed to evaluate the role of pellet-clad mechanical interaction in PCI/SCC failure in power reactor operating conditions. (author)

  19. Investigations of the behaviour of coated fuel particles and spherical fuel elements at accident temperatures

    International Nuclear Information System (INIS)

    A post irradiation annealing test apparature was constructed for the measurement of fission gas release at temperatures similar to those to be reached in a HTR during a hypothetical accident. From examinations with existing apparatures up to temperatures of 18000C results were available about the load capacity of coated particles as well as knowledges about fission gas release and defect behaviour. These results were used to plan a series of annealing tests with spherical fuel elements up to 25000C. It could be shown that the (U,Th)O2-particles with high burn up will fail during maximum core heat up of a HTR only after some hours at temperatures above 24000C. (orig.)

  20. Proceedings of the specialist meeting on the safety of water reactors fuel elements

    International Nuclear Information System (INIS)

    This specialist meeting on the safety of water reactors fuel elements was held in Saclay (France) in October 1973, and was organized by CSNI and CEA. It attracted specialists from 14 countries. Session I was devoted to normal operating conditions (coolant-cladding and fuel-cladding interactions, fission product release, effects of cladding deformation on fuel element performances and reactor operating limits); Session II was devoted to operating reactor accidents and failures, anomalous transients and handling accidents; Session III was devoted to modifications to be applied to fuel elements in order to enhance their safety and reliability; Session IV was devoted to Loss-of-Coolant Accidents (LOCA)(cladding behaviour during the accident, assembly behaviour during the accident, criteria to be considered for the study of fuel element behaviour during a LOCA)

  1. Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Nine PHWR type fuel elements with reduced length were irradiated in loop A of the TRIGA Research Reactor of INR Pitesti. The primary objective of the test was to determine the performance of nuclear fuel fabricated at INR Pitesti at high linear powers in pressurized water conditions. Six fuel elements were irradiated with a ramp power history, achieving a maximum power of 45 kW/m during pre-ramp and of 64 kW/m in the ramp. The maximum discharge burnup was of 216 MWh/kgU. Another three fuel elements with reduced length were irradiated with declining power history. At the beginning of irradiation the fuel elements achieved a maximum linear power of 66 kW/m. The maximum fuel power was about 1.3 times the maximum expected in PHWR. The maximum discharge burnup was 205 MWh/kgU. The elements were destructively examined in the hot cells of INR Pitesti. Temperature-sensitive parameters such as UO2 grain growth, fission-gas release and sheath deformations were examined. The tests proved the feasibility of irradiating PHWR type fuel elements at linear powers up to 66 kW/m under pressurized water conditions and demonstrated the possibility of more flexible operation of this fuel in power reactors. This paper presents the results of the investigation. (orig.)

  2. Neutron and gamma radiography of UO2 and TRIGA fuel elements

    International Nuclear Information System (INIS)

    The Oregon State TRIGA Reactor neutron radiography facility has been used to produce both neutron and gamma radiographs of reactor fuel. In this paper a comparison of the applicability of neutron and gamma radiography to both UO2 fuel pins and TRIGA fuel elements is made. In the case of UO2 fuel, conventional thermal neutron radiography produces excellent quality radiographs. These radiographs may be used to detect various defects in the fuel such as enrichment differences, cracks, end-capping, inclusions, etc. For TRIGA fuel elements, conventional thermal neutron radiography will not show the internal structure. This is due to the high hydrogen content of the fuel. These elements are typically 8.5 w/o uranium in Zr-H1.7; the density of hydrogen in the fuel being about 80 percent that of water. Further, while epithermal radiography significantly improves the radiographs, defects may go undetected. As an alternative to neutron radiography, high energy gamma radiographs of TRIGA fuel elements have been taken using the same facility. The gamma spectrum emitted by the reactor core is sufficiently high in energy that very good radiographs may be obtained with this technique. These radiographs show excellent detail for the internal structure of the TRIGA fuel. (author)

  3. Neutron and gamma radiography of UO2 and TRIGA fuel elements

    International Nuclear Information System (INIS)

    The Oregon State TRIGA Reactor neutron radiography facility has been used to produce both neutron and gamma radiographs of reactor fuel. In this paper a comparison of the applicability of neutron and gamma radiography to both UO2 fuel pins and TRIGA fuel elements is made. In the case of UO2 fuel, conventional thermal neutron radiography produces excellent quality radiographs. These radiographs may be used to detect various defects in the fuel such as enrichment differences, cracks, end-capping, inclusions, etc. For TRIGA fuel elements, conventional thermal neutron radiography will not show the internal structure. This is due to the high hydrogen content of the fuel. These elements are typically 8.5 w/o uranium in Zr-Hsub(1.7); the density of hydrogen in the fuel being about 80 percent that of water. Further, while epithermal radiography significantly improves the radiographs, defects may go undetected. As an alternative to neutron radiography, high energy gamma radiographs of TRIGA fuel elements have been taken using the same facility. The gamma spectrum emitted by the reactor core is sufficiently high in energy that very good radiographs may be obtained with this technique. These radiographs show excellent detail for the internal structure of the TRIGA fuel. (Auth.)

  4. Assessment of Welding System Modification of The Candu and PWR Fuel Element Types end Plug

    International Nuclear Information System (INIS)

    To anticipate future possibility of a nuclear fuel element industry in Indonesia, research on other types of nuclear fuel element beside Cirene type has to be done. It can be accomplished, one of them, by modifying the already available equipment. Based on the sheath material, the sheath dimension and the welding process parameters such as welding current and welding cycles, the available Magnetic Force Welding can be used for welding end plug of Candu nuclear fuel element by modifying some of its components (tube clamp, plug clamp, etc). The available Pellet drying and element filling furnace with its supporting system with includes helium gas filling, welding chamber, argon gas supply, vacuum system, sheath clamp and sheath driving system can be used for welding end plug with sheath of PWR nuclear fuel element by adding og Tungsten inert Gas (TIG) welding machine in the welding chamber and modifying a few components (seal clamp, sheath clamp)

  5. Single-element coaxial injector for rocket fuel

    Science.gov (United States)

    Larson, L. L.

    1969-01-01

    Improved injector for oxygen difluoride and diborane has better mixing characteristics and is able to project fuel onto the wall of the combustion chamber for better cooling. It produces an essentially conical, diverging, continuous sheet of propellant mixture formed by similarly shaped and continuously impinging sheets of fuel and oxidant.

  6. Thermal-hydraulic transient analysis of a packed particle bed reactor fuel element

    OpenAIRE

    Casey, William Emerson

    1990-01-01

    Title as it appears in the M.I.T. Graduate List, Jun. 4, 1990: Transient thermal-hydraulic analysis of a packed particle bed reactor fuel element A model which describes the thermal-hydraulic behavior of a packed particle bed reactor fuel element is developed and compared to a reference standard. The model represents a step toward a thermal-hydraulic module for a real-time, autonomous reactor powder controller. The general configuration of the fuel element is a bed of small (diameter about...

  7. Long-term testing of HTR fuel elements in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    The extensive results from irradiation experiments carried out on coated particles, on graphitic matrices of different composition and on integral fuel elements have shown that the spherical fuel elements with high-enriched uranium/thorium mixed-oxide particles and optimized graphitic matrix are available for use in the planned HTR facilities. A concentrated qualification programme is on the way in order to bring the fuel elements with particles from low-enriched uranium dioxide (LEU) and TRISO coating to a comparable level of experience and knowledge, i.e. to make them licensable for the planned HTR facilities. (orig.)

  8. Application of neutron radiography for non-destructive testing nuclear fuel elements

    International Nuclear Information System (INIS)

    This paper describes the experimental procedures, testing information and application advantages when neutron radiography is used for non-destructive inspections and quantitative analysis of fuel elements from nuclear power plants. Both the 235U enrichment and the material distribution inside the pellets can be determined by neutron radiography methods for the non-irradiated fuel elements. Both the structural integrity of fuel elements for different reactors such as PWR, BWR, FBTR and the hydrogen accumulation in the cladding material can be inspected for the irradiated samples. (authors)

  9. Testing and implementation program for the modified Darlington 37-element fuel bundle

    International Nuclear Information System (INIS)

    To mitigate the effects of reactor ageing, a design modification to the 37-element fuel is proposed in which the diameter of the centre element will be reduced to 11.5 mm from 13.1 mm. The testing and implementation phase for the 37-element fuel bundle modification is discussed in this paper. The initial plan for testing is to perform a set of out-reactor tests to assess the endurance, acoustic response and cross-flow behaviour of the revised fuel bundle design. The initial schedule outlines activities that will enable OPG to implement full core fuelling of the modified bundle within the next three to four years. (author)

  10. Design criteria -- Modification of fuel element test facilities. 1706-KER Project CGI-839

    Energy Technology Data Exchange (ETDEWEB)

    Rudock, E.R.

    1959-08-27

    The following criteria outlines the basis, objectives, and fundamental methods that shall govern the preparation of final design for ``Project CGI-839, Modification to Fuel Element Test Facilities -- 1706 KER.`` These modifications will provide the equipment to test NPR size fuel elements in the KER recirculating loops. The 1706-KER Recirculation Test Facility of KE Reactor is operated to obtain experimental data regarding high temperature reactor coolant technology and high temperature fuel element testing. The facility consists of four in-pile recirculating loops. These loops will permit testing of fuel elements with the existing process tubes of 2.1 inches I.D. To provide adequate in-reactor fuel element test facilities to support the development of NPR fuel, two KER loops, {number_sign}3 and {number_sign}4 will be converted to provide a process tube of 2.7 inches ID that will be operated at typical NPR irradiation conditions. The remaining loops No. 1 and 2, will be modified to provide additional flow and heat transfer capacity for greater flexibility in the testing of high temperature fuel elements smaller than the NPR size. New pumps, heat exchangers, and minor piping modifications will be required in all loops.

  11. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

    Science.gov (United States)

    Makmal, T.; Aviv, O.; Gilad, E.

    2016-10-01

    A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections.

  12. Fuel element gap irregularities determined from infrared scanning

    International Nuclear Information System (INIS)

    It has been hypothesized that the fuel pellet column in a nuclear fuel rod will not be concentric in the cladding. The pellets will assume random offsets, and frequent contact with the tube, resulting in a series of broken spirals. Evidence for the existence and stability of spiralling pellet columns has been observed from crud patterns on fuel rods irradiated for one and two cycles in commercial reactors. Such behavior has important implications on the pellet to cladding thermal conductivity under both normal and accident conditions and on the swelling and burst of cladding under LOCA conditions. An experimental program to verify and quantify the pellet offset behavior in an unirradiated commercial fuel rod was undertaken at ORNL under the sponsorship of the Central Electricity Generating Board and Westinghouse Electric Co. The results of the experimental program were used to develop a model to predict the burst strain behavior of the NRU test rods and were compared with the empirical correlation of the gap conductance inferred from instrumented Halden rods. The tests were performed on a Westinghouse 17 x 17 fuel rod made to the same specifications as commercial fuel except that the fuel pellets were made from depleted UO2

  13. Transactions of 2. international seminars on the mathematical/mechanical modelling of reactor fuel elements

    International Nuclear Information System (INIS)

    Fuel element modelling is a wide field of activity that spans decades of research and code development for different reactor systems and very different situations such as normal operation, off-normal situations and severe accidents. Modern computer technology helps to take the full advantage of detailed model development performed over the past for daily design analyses, safety analyses, conception of new experiments and investigation of an improved nuclear fuel utilization and fuel element performance. The basic development of the concepts of fuel element modelling can be considered as finished. The future trends are the development of refined models based on a deeper understanding of the physical and mechanical basis. Areas of interest are transient phenomena especially the fission product behaviour, burnup-enhanced phenomena, PCI and fuel reliability, severe core damage and chemical aspects. The seminar presentations reflect this variety

  14. Burn-Up Calculations for the Brookhaven Graphite Research Reactor Fuel Elements

    International Nuclear Information System (INIS)

    Fuel bum-up calculations for the Brookhaven Graphite Research Reactor involve a distribution of the thermal megawatt days of operations to the fuel elements in proportion to the average thermal neutron flux at their location in the reactor. The megawatt days so assigned can be converted to equivalent uranium-235 consumption when needed. The original fuel loading for the BGRR was neutral uranium and a single calculation was performed on each fuel element upon discharge from the reactor. A subsequent change to a fully enriched uranium-235 fuel element, however, introduced complications. The average loading of enriched uranium involves about 4800 individual elements, each occupying four different reactor positions during its term in the reactor. The total term for a central channel element is about one year as against six to eight years for an element in a peripheral channel. With the large number of individual fuel elements involved and the approximately monthly small changes needed for operation, it was necessary to resort to a computer programme to follow the burn-up of all the elements on the reactor continuously. Both this and other functions of the computer programme are discussed in the paper. To date, uranium has been recovered from two batches of spent fuel. On the first, involving 3674 elements discharged from the reactor over a period of 4.9 years, the recovery figures were 5.5% higher than the calculated total of 32.3 kg uranium-235. On the second batch, involving 1296 elements discharged from the reactor over a period of one year, the recovery figures were 2.3% higher than the calculated figures of 10.8 kg uranium-235. This relatively close agreement seems to indicate that the assumptions made to simplify the programme are acceptable and that the results of the programme are satisfactory for our particular accounting and operating requirements. (author)

  15. Current status of U3Si2 fuel element fabrication in Brazil

    International Nuclear Information System (INIS)

    IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U3O8-Al dispersion fuel plates with 2.3 gU/cm3. To support the reactor power increase, higher uranium density in the fuel plate meat had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicide was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and its qualification. (author)

  16. Non-destructive control of cladding thickness of fuel elements for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karlov, Y.; Zhukov, Y.; Chashchin, S

    1997-07-01

    The control method of fuel elements for research reactors by means of measuring beta particles back scattering made it possible to perform complete automatic non-destructive control of internal and external claddings at our plant. This control gives high guarantees of the fuel element correspondence to the requirements. The method can be used to control the three-layer items of different geometry, including plates. (author)

  17. Non-destructive control of cladding thickness of fuel elements for research reactors

    International Nuclear Information System (INIS)

    The control method of fuel elements for research reactors by means of measuring beta particles back scattering made it possible to perform complete automatic non-destructive control of internal and external claddings at our plant. This control gives high guarantees of the fuel element correspondence to the requirements. The method can be used to control the three-layer items of different geometry, including plates. (author)

  18. Specialists' meeting on fuel element performance computer modelling, Preston, United Kingdom, 15-19 March 1982

    International Nuclear Information System (INIS)

    The 46 papers of the meeting concerned with computer models of Water Reactor fuel elements cover practically all aspects of behavior of fuel elements in normal operation and in accident condition. Each session of the meeting produced a critical evaluation of one of the 5 topics into which the subject area had been divided. The sessions' report summarize the papers and make recommendations for further work. Separate abstracts were prepared for all the papers presented at this meeting

  19. Distribution of fission products in Peach Bottom HTGR fuel element E11-07

    Energy Technology Data Exchange (ETDEWEB)

    Wichner, R.P.; Dyer, F.F.; Martin, W.J.; Bate, L.C.

    1977-04-01

    This is the second in a projected series of six post-irradiation examinations of Peach Bottom High-Temperature Gas-Cooled Reactor driver fuel elements. Element E11-07, the subject of this report, received an equivalent of 701 full-power days of irradiation prior to scheduled withdrawal. The examination procedures emphasized the determination of fission product distributions in the graphite portions of the fuel element. Continuous axial scans indicated a /sup 137/Cs inventory of 17 Ci in the graphite sleeve and 8.3 Ci in the spine at the time of element withdrawal from the core. In addition, the nuclides /sup 134/Cs, /sup 110m/Ag, /sup 60/Co, and /sup 154/Eu were found in the graphite portions of the fuel element in significant amounts. Radial distributions of these nuclides plus the distribution of the beta emitters /sup 3/H, /sup 14/C, and /sup 90/Sr were obtained at six axial locations, four within the fueled region and one each above and below. The radial dissection was accomplished by use of a manipulator-operated lathe in a hot cell. These profiles reveal an increased degree of penetration of /sup 134/Cs, relative to /sup 137/Cs, evidently due to a longer time spent as xenon precursor. In addition to fission product distribution, the appearance of the element components was recorded photographically, fuel compact and graphite dimensions were recorded at numerous locations, and metallographic examinations of the fuel were performed.

  20. Measurement of after-heat production and dose rates of spent AVR fuel elements

    International Nuclear Information System (INIS)

    Data on the afterheat production and dose rate of spent AVR fuel elements prepared by the ORIGEN computer program are verified by measurements. Individual measurements of afterheat and dose rate were implemented on 17 AVR fuel elements with decay periods of 150 days and more than four years, and burnups between 4.1 and 16.4% fima were implemented in the HOT CELLS at the Juelich Nuclear Research Centre. The radiation energy absorbed in the fuel elements and converted into heat was measured with a calorimeter, whereas the emitted radiation fraction was determined via dose rate measurements. The measured results for fuel elements with decay periods of more than one year are in good agreement with the data from ORIGEN. In the case of fuel elements with shorter decay periods (approx. 150 days) in part considerably lower values were measured which can be explained by the fact that the power gradient in time of the fuel elements in the reactor can vary considerably whereas mean are included in the ORIGEN computations assuming full-load operation. (orig./HP))

  1. Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element

    Science.gov (United States)

    Kim, Tony; Emrich, William J., Jr.; Barkett, Laura A.; Mathias, Adam D.; Cassibry, Jason T.

    2013-01-01

    In the past, fuel rods have been used in nuclear propulsion applications. A new fuel element concept that reduces weight and increases efficiency uses a stack of grooved discs. Each fuel element is a flat disc with a hole on the interior and grooves across the top. Many grooved ring fuel elements for use in nuclear thermal propulsion systems have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel element with a higher surface-area-to-volume ratio is ideal. When grooves are shallower, i.e., they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of turbulence with those of heat transfer, the effects on the cooler gas flowing through the grooves of the thermally excited solid can be predicted. Parametric studies were done to show how a pressure drop across the axial length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the behaviors that result from the manipulation of various parameters. Temperature profiles of the solid and gas showed that more structural optimization is needed to produce the desired results. Keywords: Nuclear Thermal Propulsion, Fuel Element, Heat Transfer, Computational Fluid Dynamics, Coupled Physics Computations, Finite Element Analysis

  2. Radiological measurements during decontamination of PFBR MOX fuel elements using ultrasonic decontamination technique

    International Nuclear Information System (INIS)

    In a fuel fabrication facility fabrication of MOX fuel elements involving various metallurgical processes is carried out in leak tight glove boxes because of high radio toxicity associated with plutonium,. A fuel pin consists of a thin walled tube loaded with cylindrical fuel pellets with plugs welded on both ends. The pellet loading and welding processes result in cross contamination on the tube surface near the edges. It is important that finished fuel pins should not contain any transferable contamination on the surface beyond safe limits applicable for unrestricted release before subjecting the pins to manual handling for quality control checks. Hence it is imperative that thorough decontamination of fuel pins is essential for safe handling. Conventional decontamination methods result in undue personal exposures and generation of solid waste. Though there are number of techniques available for decontamination of non-fuel elements in the nuclear industry, very few of them can be used for decontamination of fuel elements because of possible damage to fuel clad, Ultrasonic cleaning process, using dc-mineralized water as medium does not affect the properties of the clad and is simple to implement and fast to carry out. This paper brings out radiological measurements carried out to study the effectiveness of ultrasonic decontamination technique and the factors involved in achieving required degree of decontamination with reduced individual exposure

  3. Distribution of fission products in Peach Bottom HTGR fuel element E14-01

    Energy Technology Data Exchange (ETDEWEB)

    Wichner, R.P.; Dyer, F.F.; Martin, W.J.; Fairchild, L.L.

    1977-08-01

    The third in a projected series of six postirradiation examinations of Peach Bottom High-Temperature Gas-Cooled Reactor driver fuel elements is presented. Element E14-01, the subject of the report, was one of the 60 driver elements (out of a total of 804) that contained zirconium boride pellets within a hollow spine. It was also one of the few predimensioned elements, which therefore allowed accurate determination of dimensional change due to irradiation service. The element received an equivalent of 897 full-power days irradiation prior to scheduled termination of Core 2 operation. The examination procedures emphasized the determination of fission product distributions in the graphite portions of the fuel element. Continuous axial scans indicated a /sup 137/Cs inventory of 0.24 Ci in the graphite sleeve and 0.047 Ci in the spine at the time of element withdrawal from the core. In addition, the nuclides /sup 134/Cs, /sup 110m/Ag, /sup 60/Co, and /sup 154/Eu were found in the graphite portions of the fuel element in significant amounts. Radial distributions of these nuclides plus the beta-emitters /sup 3/H, /sup 14/C, and /sup 90/Sr were obtained at three axial locations within the fueled region of the element. The radial dissection was accomplished by use of a manipulator-operated lathe in a hot cell. In addition to fission product distributions, the appearance of the component parts of the element was recorded photographically, fuel compact and graphite dimensions were recorded at numerous locations, and metallographic examinations of the fuel were performed.

  4. Design and production process of bushing-type fuel elements for channel research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Afanasiev, V.L.; Aleksandrov, A.B.; Enin, A.A. [NZHK, Novosibirsk (Russian Federation)

    1998-07-01

    The design of bushing-type fuel elements (FEs) based on the dioxide fuel composition UO{sub 2}+Al for channel research reactors is described. Commercial technological process for bushing-type FEs with up to 0.8 g/cm{sup 3} uranium concentration in the fuel core is presented. This technology is based on fuel core production using powder metallurgy with subsequent chemical treatment of its surface and enclosing into the finished cladding. Commercial technological process for bushing-type FEs with 0.8-3.8 g/cm{sup 3} uranium concentration in the fuel composition is considered. This process is based on fuel core production by means of extrusion technology followed by fuel core enclosing into the cladding. (author)

  5. Dimensional changes in operating UO2 fuel elements: effect of pellet density, burnup and ramp rate

    International Nuclear Information System (INIS)

    We have used an in-reactor diameter measuring rig, in combination with a He-3 coil associated with the X-6 loop of the NRX reactor, to determine the dimensional response of CANDU-type UO2 fuel elements as a function of power, prior irradiation, fuel density and ramp rate. The maximum diametral increase accompanying a ramp was about 1% for high density fuel. No elements failed, despite, in the most severe case, a 1.5 h hold at 56 kW/m, after a ramp from 30 kW/m in less than three minutes. (author)

  6. Development of a new technique for experimental evaluation of the fuel element's subchannel mixing

    International Nuclear Information System (INIS)

    In this work, the development of a new experimental method for the measurement of mixing between the cooling subchannels of nuclear fuel elements by using thermal traces, is presented.The method has been proved on a reduced test section with very positive results, having demonstrated its simplicity and low cost.Because it is suitable for heterogeneous and compact subchannels (asArgentinean fuels) with high water flows in simple and affordable tests at atmospheric pressure, this new method is specially well suited for the design of fuel elements, while it offers advantages over other methods of mixing measurement

  7. [Disseminated granuloma annulare].

    Science.gov (United States)

    Kansky, A

    1975-09-01

    A case of generalized granuloma annulare in a 55 year old man is reported. The disease appeared five years before the first admission to the hospital. A large number of bluish-red or skin-colour papules were scattered mainly around the earlobes, buttocks and on the extremities. Some of the lesions were lined up in rings or plaques. Small depigmented and brownish scars were present. Two biopsies revealed characteristic foci of complete collagen degeneration accompanied by a palisading infiltrate in the upper dermis. Treatment with tuberculostatics and antimalarics was without improvement. The lesions cleared after a course of prednison, but reappeared when the drug was discontinued.

  8. Annular recuperator design

    Science.gov (United States)

    Kang, Yungmo

    2005-10-04

    An annular heat recuperator is formed with alternating hot and cold cells to separate counter-flowing hot and cold fluid streams. Each cold cell has a fluid inlet formed in the inner diameter of the recuperator near one axial end, and a fluid outlet formed in the outer diameter of the recuperator near the other axial end to evenly distribute fluid mass flow throughout the cell. Cold cells may be joined with the outlet of one cell fluidly connected to the inlet of an adjacent downstream cell to form multi-stage cells.

  9. The clad collapse modelling of Indian PHWR fuel element, an FEM approach

    International Nuclear Information System (INIS)

    The fuel elements for PHWR use a thin, collapsible zircaloy clad design. This design is consistent with essential neutron economy in PHWRs, and also results in better heat transfer between fuel and clad. However, thin clad may give rise to problem of permanent clad collapse under coolant pressure in axial gap and radial gap available during the initial stay of fuel inside the reactor. Present work explores the problem of longitudinal ridges, formed due to permanent circumferential collapse of clad on fuel. The tip of these ridges has the potential to become the site for crack initiation under subsequent cyclic thermal/pressure loading. The collapse behavior of fuel element is studied using FEM modeling of pellet, clad and their contact. This study considers the effects of clad thickness, clad yield strength, clad initial ovality, anisotropy in clad yield strength, and radial gap of fuel element on the collapse behavior. The verification of present model is done for the results of critical buckling pressure required for the longitudinal ridge formation by the available CANDU experimental data, which matched satisfactorily for the yield strength ratio (circumferential YS to longitudinal YS) of 1.5. In addition the longitudinal ridge height and increase in ovality were calculated for the collapse experiments done on the 220 MWe PHWR fuel elements

  10. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pope, M. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, M. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morrell, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jamison, R. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nef, E. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nigg, D. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

  11. Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes

    Directory of Open Access Journals (Sweden)

    Park Joo Hwan

    2016-01-01

    Full Text Available A CANDU-6 reactor, which has 380 fuel channels of a pressure tube type, is suffering from aging or creep of the pressure tubes. Most of the aging effects for the CANDU primary heat transport system were originated from the horizontal crept pressure tubes. As the operating years of a CANDU reactor proceed, a pressure tube experiences high neutron irradiation damage under high temperature and pressure. The crept pressure tube can deteriorate the Critical Heat Flux (CHF of a fuel channel and finally worsen the reactor operating performance and thermal margin. Recently, the modification of the central subchannel area with increasing inner pitch length of a standard 37-element fuel bundle was proposed and studied in terms of the dryout power enhancement for the uncrept pressure tube since a standard 37-element fuel bundle has a relatively small flow area and high flow resistance at the central region. This study introduced a subchannel analysis for the crept pressure tubes loaded with the inner pitch length modification of a standard 37-element fuel bundle. In addition, the subchannel characteristics were investigated according to the flow area change of the center subchannels for the crept pressure tubes. Also, it was discussed how much the crept pressure tubes affected the thermalhydraulic characteristics of the fuel channel as well as the dryout power for the modification of a standard 37-element fuel bundle.

  12. Non-destructive-Testing of Nuclear Fuel Element by Means of Neutron Imaging Technique

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Nuclear fuel element is the key component of nuclear reactor. People have to make strictly testing of the element to make sure the reactor operating safely. Neutron imaging is one of Non-destructive-Testing (NDT) techniques, which are very important techniques for

  13. Fabrication and Properties of Cylindrical and Tubular UO2-Stainless Steel Fuel Elements

    International Nuclear Information System (INIS)

    The fabrication by extrusion sintering of canned tubes and rods containing a dispersed fuel nucleus of UO2-stainless steel was investigated. The fuel elements may be up to 2 m long; the diameter of the rods is less than 30 mm; the total thickness of the tubes may be as low as 0.8 mm. The fuel cermet contains 20 to 60% UO2 by weight in the form of spherical particles having a diameter greater than 50 μm. The grade of steel for the cermet can be chosen relative to the conditions in which it is to be used. The fuel elements are checked by gammagraphy, macrography and micrography. The thermal conductivity and mechanical tensile properties of the cermets were investigated with respect to UO2 content and particle granulometry. Thermal shock tests and lengthy annealing of the tubular elements caused no cracking or lessening of the can. By this method of extrusion sintering canned fuel elements can be obtained directly; handling of the UO2 powder is reduced to a minimum, so that the losses of enriched uranium in the industrial fabrication of the fuel elements can also be reduced.

  14. A study on end closure of fuel element cladding of PHWRs by GTA welding

    International Nuclear Information System (INIS)

    Zirconium alloys are extensively used in thermal reactors as cladding materials. Fuel material is enclosed in cladding tube and welding with plugs close ends of tube. There are various methods for end closure welding of these fuel elements. Resistance welding (RW) process has been widely used for PHWRs fuel elements. Resistance welds are made rapidly and it is suitable for large-scale production. This process has certain limitations. To explore alternate avenues for superior quality end closure welds of PHWR fuel elements, a study is undertaken of different welding processes. GTAW was chosen for further work since it has well-established practice. New generation GTAW power sources enhance the quality and reliability of welds. A major advantage of GTAW is that the process, equipment and operators can be deployed for any type of fuel elements e.g. BWRs, PWRs, PHWRs as well as AHWRs. Experiments were conducted and the process was studied for Φ15.3 mm OD and 0.4 mm wall thickness tubes. Welding parameters have been optimised. Joint quality has been assessed with respect to specifications of PHWRs fuel elements. Welds were evaluated by NDT as well as destructive methods. The paper describes the GTAW process, selection of power source, design of welding chamber, parameters, problems faced, weld defects observed during standardisation of parameters and their elimination. (author)

  15. Application of a quality control program for developing fuel elements for research reactors

    International Nuclear Information System (INIS)

    The development of nuclear fuel elements for the IEAR-1 research reactor is a task that is being pursued by IPEN/CNEN-SP for several years. The studies included the development of U3O8-Al Nuclear cermets, rolling of U3O8-Al brickets using the picture frame technique for the obtension of Nuclear fuel plates as well as the fabrication of components and the final assembling of the fuel elements. The prototypes are made to conform to stringent quality control specifications. These specifications cover various aspects such as the metallurgical, ceramical, and mechanical properties of the materials involved as well as non-destructive tests and dimensional and visual requirements of the various components. In this context, an extensive specification of the materials and components used have been compiled and are periodically reviewed and revised. An extensive quality control program was planned and is being tested in practice simultaneously to the fuel element development. During the elaboration of the procedures for the characterization tests, special attention has been devoted to the storage of data that could be used for the analysis of the irradiation behaviour of the fuel element. The various procedures used during implementation of the system required for the quality control of the nuclear fuel elements for the IEAR-1 Nuclear reactor are presented. (Author)

  16. Automation of remote handling in uranium and mixed oxide fuel element fabrication plants

    International Nuclear Information System (INIS)

    The subject of the analyses are plants for the fabrication or uranium oxide and uranium-plutonium mixed oxide fuel elements. The reference basis of the paper is an overview of the state-of-the-art of manufacturing technologies with regard to automation and remote handling during fuel element fabrication in national and foreign plants, and in comparabel sectors of conventional technologies. Proceeding from ambient dose rates, residence times, and technical conditions or individual doses at typical work-places during fuel element fabrication, work processes are pointed out which, taking into account technical possibilities, should be given priority when automating, and technical solutions for it are sought. Advantages and disadvantages of such measures are outlined, and reduction of radiation exposure is shown (example: mixed oxide fuel fabrication plant at Hanau). (orig./HP)

  17. Early encapsulation of spent fuel elements. Separation of requirements from operation and disposal

    International Nuclear Information System (INIS)

    The design requirements to be met by fuel elements in the light of the trend towards higher burnups are aggravated by the need to take into account the logistics of decommissioning. The criteria applying to operation should be separated from those applying to disposal this can be achieved by early encapsulation of spent fuel elements. This creates the flexibility now for decisions to be taken decades from now. A few relatively simple process step will hermetically seal spent fuel elements in the storage pools of nuclear power plants. Only the capsules are handled until the decision about the spent fuel management pathway is taken. Should direct disposal be chosen, only capsules will be handled right to the repository. This process of early encapsulation not only provides more flexibility, but also offers major economic advantages. (orig.)

  18. Application of replication techniques to fractography of irradiated uranium fuel elements

    International Nuclear Information System (INIS)

    Radiation effects in nuclear fuel elements were studied by fractography. Fuel elements were experimentally irradiated at various doses at the IRR-2. Irradiated, as well as unirradiated fuel elements, were bent till fracture with a bending press. The fractured surfaces were examined by standard replication techniques. The structures reveal the same morphological zones as in standard tensile fractures. The crack origin zone (fibrous zone) appears near the fuel indent. The radial marks zone starts near the origin and covers about 2/3 of the cross-sectional area. The third zone (shear rupture zone) has a bump, sometimes 1 cm high. The crack origin zone is mainly brittle. As the crack propagates in the radial marks zone it becomes more ductile until its entirely ductile at the bump top. Beyond the bump it turns brittle again. (author)

  19. Manufacturing of 37-element fuel bundles for PHWR 540 - new approach

    Energy Technology Data Exchange (ETDEWEB)

    Arora, U.K.; Sastry, V.S.; Banerjee, P.K.; Rao, G.V.S.H.; Jayaraj, R.N. [Nuclear Fuel Complex, Dept. Atomic Energy, Government of India, Hyderabad (India)

    2003-07-01

    Nuclear Fuel Complex (NFC), established in early seventies, is a major industrial unit of Department of Atomic Energy. NFC is responsible for the supply of fuel bundles to all the 220 MWe PHWRs presently in operation. For supplying fuel bundles for the forthcoming 540 MWe PHWRs, NEC is dovetailing 37-element fuel bundle manufacturing facilities in the existing plants. In tune with the philosophy of self-reliance, emphasis is given to technology upgradation, higher customer satisfaction and application of modern quality control techniques. With the experience gained over the years in manufacturing 19-element fuel bundles, NEC has introduced resistance welding of appendages on fuel tubes prior to loading of UO{sub 2} pellets, use of bio-degradable cleaning agents, simple diagnostic tools for checking the equipment condition, on line monitoring of variables, built-in process control methods and total productive maintenance concepts in the new manufacturing facility. Simple material handling systems have been contemplated for handling of the fuel bundles. This paper highlights the flow-sheet adopted for the process, design features of critical equipment and the methodology for fabricating the 37-element fuel bundles, 'RIGHT FIRST TIME'. (author)

  20. Manufacturing of 37-element fuel bundles for PHWR 540 - new approach

    International Nuclear Information System (INIS)

    Nuclear Fuel Complex (NFC), established in early seventies, is a major industrial unit of Department of Atomic Energy. NFC is responsible for the supply of fuel bundles to all the 220 MWe PHWRs presently in operation. For supplying fuel bundles for the forthcoming 540 MWe PHWRs, NEC is dovetailing 37-element fuel bundle manufacturing facilities in the existing plants. In tune with the philosophy of self-reliance, emphasis is given to technology upgradation, higher customer satisfaction and application of modern quality control techniques. With the experience gained over the years in manufacturing 19-element fuel bundles, NEC has introduced resistance welding of appendages on fuel tubes prior to loading of UO2 pellets, use of bio-degradable cleaning agents, simple diagnostic tools for checking the equipment condition, on line monitoring of variables, built-in process control methods and total productive maintenance concepts in the new manufacturing facility. Simple material handling systems have been contemplated for handling of the fuel bundles. This paper highlights the flow-sheet adopted for the process, design features of critical equipment and the methodology for fabricating the 37-element fuel bundles, 'RIGHT FIRST TIME'. (author)

  1. Special equipment for the fabrication and quality control of nuclear fuel elements

    International Nuclear Information System (INIS)

    For the fabrication of LWR fuel elements, columns are packed of up to 4 m in length, consisting of fuel pellets with different uranium enrichment, their weight and total length to be measured prior to further processing to fuel rods. An automated column packing device has been developed for this purpose. The packing jobs and other tasks are computer-controlled, measured data are stored and are printed out for quality documentation. The forces in the springs of fuel spacers of LWR fuel elements are to be measured and compared with the standard requirements, deviations to be documented. For this task, another computer-controlled, automated device has been developed for measuring the spring forces at all required positions after positioning and fixation of the spacers. (orig./DG)

  2. HTGR fuel elements operating conditions during accidents with abrupt power raise

    International Nuclear Information System (INIS)

    The necessity of the investigations for developing of HTGR fuel elements operability criteria, connected with the specific energy release values and the rates of its change in fuel is demonstrated in the paper on the example of the accident with positive reactivity increase at VGM reactor pebble bed compression as a result of seismic impact. It is shown, that the average fuel enthalpy over the core in this accident with the emergency protection failure may reach ∼24 Kj/g U02, and the maximum rate of its increase is about 0.14 Kj/g.s. It considerably exceeds the established limit of fuel enthalpy for LWR fuel elements. (author). 5 refs, 2 figs

  3. Fabrication of U3 Si2 based fuel elements in Brazil

    International Nuclear Information System (INIS)

    IPEN - Brazilian Institute of Nuclear and Energy Research, has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has manufacturing its own fuel element, initially based on U3O8-Al dispersion fuel plates with 2.3 gU/cm3. To support the reactor power increase, higher uranium density had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicide was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and qualification. (author)

  4. Infinite fuel element simulation of pin power distributions and control blade history in a BWR fuel assembly

    International Nuclear Information System (INIS)

    Pellet-Cladding Interaction (PCI) is a well known effect in fuel pins. One possible reason for PCI-effects could be local power excursions in the fuel pins, which can led to a rupture of the fuel cladding tube. From a reactor safety point of view this has to be considered as a violence of the barrier principal in order to retain fission products in the fuel pins. This paper focuses on the pin power distributions in a 2D infinite lattice of a BWR fuel element. Lots of studies related PCI effect can be found in the literature. In this compact, coupled neutronic depletion calculations taking the control history effect into account are described. Depletion calculations of an infinite fuel element of a BWR were carried out with controlled, uncontrolled and temporarily controlled scenarios. Later ones are needed to describe the control blade history (CBH) effect. A Monte-Carlo approach is mandatory to simulate the neutron physics. The VESTA code was applied to couple the Monte-Carlo-Code MCNP(X) with the burnup code ORIGEN. Additionally, CASMO-4 is also employed to verify the method of simulation results from VESTA. The cross sections for Monte Carlo and burn-up calculations are derived from ENDF/B-VII.0. (orig.)

  5. Burn-up and Operation Time of Fuel Elements Produced in IPEN

    Science.gov (United States)

    Tondin, Julio Benedito Marin; Filho, Tufic Madi

    2011-08-01

    The aim of this paper is to present the developed work along the operational and reliability tests of fuel elements produced in the Institute of Energetic and Nuclear Research, IPEN-CNEN/SP, from the 1980's. The study analyzed the U-235 burn evolution and the element remain in the research reactor IEA-R1. The fuel elements are of the type MTR (Material Testing Reactor), the standard with 18 plates and a 12-plate control, with a nominal mean enrichment of 20%.

  6. Implementation of the utilization program for the fuel elements of the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    The programming operation for the use of the fuel elements in the Atucha-1 nuclear power plant was initially under the responsibility of the KWU Company, as part of the services rendered due for the manufacturing of said elements. This job was done with the help of the TRISIC program, developed in the early seventies by CNEA and SIEMENS staff. From april 21, 1979 on, CNEA took over the responsibility and strategy of the interchange of fuel elements. The several stages carried out for the implementation of this service are detailed. (M.E.L.)

  7. A modelling study for the determination of temperature distribution in a WWER-1000 type fuel element

    International Nuclear Information System (INIS)

    A short description is presented of the fuel element of the WWER-1000 type nuclear reactor. A physical model of the fuel element was developed in order to determine the temperature distribution within the element. The numerical solution of the mathematical model based on the physical model was performed by using the one-dimensional modular program HOVID. The influence of simplifications introduced in the mathematical model on the calculated temperature distribution is discussed. Finally, the results of steady-state calculations are analyzed. (R.P.) 11 refs.; 7 figs.; 4 tabs

  8. Consideration of subchannel area of a 37-element fuel to enhance CHF

    International Nuclear Information System (INIS)

    CANDU-6 reactor has 380 fuel channels of a pressure tube type, which provides an independent flow passage, and each pressure tube contains 12 fuel bundles horizontally. The CHF of a CANDU fuel bundle in a horizontal fuel channel is one of the important parameters determining the thermalhydraulic safety margin as well as the trip set point of the Regional Overpower Protection (ROP) system. Hence, the CHF enhancement of a CANDU fuel bundle has been an issue for a long time and can be affected by the geometric configuration of the fuel elements as well as several appendages such as the end-plates, bearing pads, and spacers attached to the fuel elements. This paper considers the modification of the inner ring radius of a standard 37-element fuel bundle to enhance the CHF, since the CHFs of a standard 37-element fuel bundle preferably occur at the peripheral subchannels of the center rod, owing to the relative small flow area or high flow resistance under high flow conditions or the normal operating conditions of a CANDU reactor. Subchannel analysis techniques using the ASSERT-PV code were applied to investigate the local CHF characteristics according to the inner ring radius variation for the original diameter of the pressure tube. It was found that the modification of the inner ring radius is very effective in enhancing the dryout power of the fuel bundle under the reactor operating conditions through an enthalpy re-distribution of the subchannels and change in the local locations of the first CHF occurrences. (author)

  9. A CFD numerical model for the flow distribution in a MTR fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Andrade, Delvonei Alves de; Santos, Pedro Henrique Di Giovanni; Oliveira, Fabio Branco Vaz de; Torres, Walmir Maximo; Umbehaun, Pedro Ernesto; Souza, Jose Antonio Batista de; Belchior Junior, Antonio; Sabundjian, Gaiane; Prado, Adelk de Carvalho, E-mail: acprado@ipen.br, E-mail: delvonei@ipen.br, E-mail: dpedro_digiovanni_s@hotmail.com, E-mail: fabio@ipen.br, E-mail: wmtorres@ipen.br, E-mail: umbehaun@ipen.br, E-mail: jasouza@ipen.br, E-mail: abelchior@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Angelo, Edvaldo, E-mail: eangelo@mackenzie.br [Universidade Presbiteriana Mackenzie, Sao Paulo, SP (Brazil); Angelo, Gabriel, E-mail: gangelo@fei.edu.br [Fundacao Educacional Inaciana (FEI), Sao Bernardo do Campo, SP (Brazil)

    2015-07-01

    Previously, an instrumented dummy fuel element (DMPV-01), with the same geometric characteristics of a MTR fuel element, was designed and constructed for pressure drop and flow distribution measurement experiments at the IEA-R1 reactor core. This dummy element was also used to measure the flow distribution among the rectangular flow channels formed by element fuel plates. A CFD numerical model was developed to complement the studies. This work presents the proposed CFD model as well as a comparison between numerical and experimental results of flow rate distribution among the internal flow channels. Numerical results show that the model reproduces the experiments very well and can be used for the studies as a more convenient and complementary tool. (author)

  10. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R., E-mail: rrr@cdtn.br, E-mail: amir@cdtn.br, E-mail: edson@cdtn.br, E-mail: maritzargual@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  11. AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

    International Nuclear Information System (INIS)

    Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700 C. This temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelength dispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700 C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction

  12. An Evaluation of Potential Liner Materials for Eliminating FCCI in Irradiated Metallic Nuclear Fuel Elements

    International Nuclear Information System (INIS)

    Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting long live transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700 deg. C. This temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelength dispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700 deg. C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction. (authors)

  13. Fast neutron reactor fuel elements and power grid duty cycling

    International Nuclear Information System (INIS)

    The PHENIX power grid cycling operation in 1982-1983 will allow verification of the models and criteria developed in the interim. It will provide indispensible statistical data and will open the way to power grid duty for Super PHENIX beginning in 1986. Although at the present time it is impossible to resolve the question of weekly or daily load variations, it is felt that fast neutron reactor fuel subassemblies should provide satisfactory performance for primary and secondary frequency adjustments

  14. Oxide fuel element and blanket element development programs. Quarterly progress report, April-June 1978

    International Nuclear Information System (INIS)

    Approval-in-principle has been granted for run beyond breach experiment XY-2, which will incorporate an F11A series rod. Fuel microstructures and operating parameters have been tabulated for 118 specimens from the F20 power to melt experiment. Retained gas measurements have been compiled indicating 36-50 μl/gm in this high power fuel. Topical report GEFR-00367 was prepared describing F20 results. Preparation of the Test Design Description for axial blanket experiment AB-1 is proceeding on schedule (for Cycle 2 irradiation). The safety analysis calculations, showing no fuel melting nor sodium boiling in design-basis upsets, have been completed

  15. Mechanical separation process for decladding of LWR fuel elements

    International Nuclear Information System (INIS)

    A comparison of the advantages and disadvantages of known methods of decladding led to cavitation erosion being used as a decladding mechanism. This process attacks not the jacket of the fuel rod but the fuel itself. Cavitation erosion is the consequence of imploding vapour bubbles entailing dynamic stress of a high frequency and high amplitude. The separation effect is due to the different material properties. Ductile materials as a rule are much more resistant to dynamic stress than brittle materials. Systematic experiments at varying pressures, volume flow, nozzle geometries and distances between nozzle and sample led to optimized parameters. There was a conspicuous rise in the relations pressure to depth of erosion and volume flow to depth of erosion. This considered, p=700 bar and d=1.6 mm were found to be useful parameters. The relation of the distance from nozzle to sample and the erosion obtained also has an optimum at s=50 mm. This distance can be shortened in the course of the operation. A great entrance angle combined with a nozzle outlet channel of the length l=1/2 D improves the erosion result considerably. The attack of the cavitating water jet on the jacket of the fuel rod causes a weight loss of <=2per mille. (orig./HP)

  16. IFPE/EFE-RO, Experimental Fuel Elements RO89 and RO51 in TRIGA 14 MW Reactor (INR-Pitesti)

    International Nuclear Information System (INIS)

    Description of program or function: Romanian irradiation tests concerned with Candu type fuel elements behavior and with the limits of the design parameters. A particular feature of the Candu fuel project is the small plenum (void volume) added for relaxation of the fission gases, which are inherently released during the fuel irradiation. Two irradiation tests in the C2 device from the TRIGA 14 MW reactor were performed between the years 1985-1987. The tests were done to evaluate the effect of the fuel density on the time-evolution of the fission gas pressure. Experimental fuel elements were adequately instrumented with pressure transducers to follow the fission gas pressure changes during fuel irradiation. The first irradiation test was conducted on the fuel element coded No.89 whose main characteristics were the nominal values of the main fuel design parameters. The second one was conducted on the fuel element coded No.51. Because of the axial flux asymmetry inside the TRIGA reactor core, the experimental elements are shorter in length than the Candu fuel design. The irradiation tests consisted in evaluation of the time-evolution of the internal pressure from two experimental fuel elements having the main design characteristics as the Romanian Candu type fuel element design and to follow the dependence of the internal pressure of the fission gas on the fuel density

  17. Choices of canisters and elements for the first fuel shipment from K West Basin

    Energy Technology Data Exchange (ETDEWEB)

    Makenas, B.J.

    1995-03-01

    Twenty-two canisters (10 prime and 12 backup candidates) in the K West Basin have been identified as containing fuel which, when examined, will satisfy the Data Quality Objectives for the first fuel shipment from this basin. These were chosen as meeting criteria such as containing relatively long fuel elements, locking bar integrity, and the availability of gas/liquid interface level measurements for associated canister gas traps. Two canisters were identified as having reported broken fuel on initial loading. Usage and interpretation of canister cesium concentration measurements have also been established and levels of maximum and minimum acceptable cesium concentration (from a data optimization point of view) for decapping have been determined although other operational cesium limits may also apply. Criteria for picking particular elements, once a canister is opened, are reviewed in this document. A pristine, a slightly damaged, and a badly damaged element are desired. The latter includes elements with end caps removed but does not include elements which have large amounts of swelling or split cladding that might interfere with handling tools. Finally, operational scenarios have been suggested to aid in the selections of canisters and elements in a way that utilizes anticipated canister gas sampling and leads to a correct and quick choice of elements which will supply the desired data.

  18. Trace elements retained in washed nuclear fuel reprocessing solvents

    International Nuclear Information System (INIS)

    Analysis of purified TBP extractant from solvent extraction processes at Savannah River Plant showed several stable elements and several long-lived radioisotopes. Stable elements Al, Na, Br, Ce, Hg, and Sm are found in trace quantities in the solvent. The only stable metallic element consistently found in the solvent was Al, with a concentration which varies from about 30 ppM to about 10 ppM. The halogens Br and Cl appear to be found in the solvent systems as organo halides. Radionuclides found were principally 106Ru, 129I, 3H, 235U, and 239Pu. The 129I concentration was about 1 ppM in the first solvent extraction cycle of each facility. In the other cycles, 129I concentration varied from about 0.1 to 0.5 ppM. Both 129I and 3H appear to be in the organic solvent as a result of exchange with hydrogen

  19. Oxide fuel element and blanket element development programs. Quarterly progress report, January-February-March, 1979

    International Nuclear Information System (INIS)

    Fuel pin profilometry of some 9% burnup F20-F5 pins showed small diameter increases at the fuel-insulator interface at the top of the core. Neither these secondary peaks nor the larger diameter increases near the core midplane exhibited any relationship to the local presence of once-molten fuel in any F20 fuel pin. Augmented safety analysis computations for experiment AB-1 (additional transients suggested by HEDL) showed that cumulative damage fractions from the additional transients were in every case less than 10-4. Mechanical tests have been performed that confirm previous computations for the removal end plugs to be used in a characterizer subassembly for AB-1. The resulting pin removal forces are well within the design envelope

  20. Fabrication and Testing of Prototype APM-Clad UO2 Fuel Elements

    International Nuclear Information System (INIS)

    In support of the 50-MW(e) Prototype Organic Power Reactor Programme (POPR), extensive development work has been performed on aluminium powder metallurgy (ARM) products, toward their use as cladding for UO2 fuel. As part of this development work, eutectic bonding, flash butt welding, and cold-pressure welding were investigated as methods for making end closures in die fuel element cladding. Vibratory packing was studied as a means of filling APM tubes with UO2. Out-of-pile tests were conducted to obtain information on APM-UO2 compatibility. This work revealed that, under present conditions, eutectic bonding was the most suitable method for making end closures; vibratory packing produced fuel densities in the range of 80 to 88% of theoretical density; and no APM-UO2 reaction took place in the range of POPR operating temperatures (850oF maximum fuel-cladding interface temperature). As a result o f this development work, five APM-clad UO2 prototype fuel elements have been fabricated for testing in the Organic Moderated Reactor Experiment (OMRE). Each element consisted of 24 or 25 APM-clad fuel rods, arranged in a 5 x 5 array in a nickel-plated steel or an APM fuel box. To increase surface area, the extruded APM cladding had eight fins which were spiralled to a pitch of 45 or 90e/ ft to further improve heat transfer. The fuel rod end closures were made by eutectic bonding of silver-plated aluminium end plugs to the APM tubing. The elements were instrumented to: (1) Measure cladding surface and coolant temperatures, (2) Detect fuel rod failure, (3) Change coolant velocity (means of achieving peak cladding surface temperature of 850oF), (4) Measure coolant velocity, and (5) Measure fission gas build-up. These elements have been installed in the OMRE with target fuel burn-ups of 25000 to 30000 MWd/t of uranium. As of 1 April 1963, they had achieved accumulated burn-ups ranging from 7700 to 12 000 MWd/t of uranium. Two of the elements had been removed from the reactor as a

  1. Load following tests on CANDU-type fuel elements in TRIGA research reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Two load following (LF) tests on CANDU-type fuel elements were performed in the TRIGA Research Reactor of INR Pitesti, where the tests were designed to represent fuel in a CANDU reactor operating in a load following regime. In the first LF test the designated '78R' fuel element successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, developed under INR-AECL co-operation, the fuel element designated as 'ME01' withstood 200 power cycles from 27 to 54 kW/m average linear power, as well as additional ramps due to reactor trips and restarts during the test period. This experimental program is ongoing at INR Pitesti. Both LF tests were simulated with finite element computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the cladding arising from expansion and contraction of the pellets. New LF tests are planned to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in the INR TRIGA Research Reactor compared with the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (author)

  2. Destructive Examination of Experimental Candu Fuel Elements Irradiated in TRIGA-SSR Reactor

    International Nuclear Information System (INIS)

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW(th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (INR) Pitesti. zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post- irradiation Examination Laboratory (PIEL) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Microstructural characterization by metallographic analyses; (iii) Determination of mechanical properties; (iv) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

  3. Post Irradiation Examination of Experomental CANDU Fuel Elements Irradiated in TRIGA-SSR Reactor

    International Nuclear Information System (INIS)

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW (th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (INR) Pitesti. Zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory (PIEL) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Gamma scanning and tomography; (iii) Measurement of pressure, volume and isotopic composition of fission gas; (iv) Microstructural characterization by metallographic analyses; (v) Determination of mechanical properties; amd (vi) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

  4. The nuclear fuel elements' world market and the position of the Argentine Republic as producer

    International Nuclear Information System (INIS)

    The development of the nuclear fuel elements' industry is analyzed, both in the present and projected world market, up to the year 2000, in the light of the situation affecting the nucleoelectric industry. By means of the offer/demand function, an analysis is made of the behaviour of the fuel elements' market throughout the fuel cycle structure. The regional unbalances between availability and demand of uranium resources are considered, as well as the factors having an unfavorable incidence on the fuel cycle's economic equation. The economic structure to be used for the calculation of the nucleoelectric generating cost is presented, in order to situate, within said nuclear economy, the component corresponding to the fuel cycle cost. Emphasis is placed on the 'front end' stages of the fuel cycle, but also considering those stages belonging to the 'back end'. Argentina's fuel elements market and its present and projected nucleoelectric park are analyzed, indicating their relative position in the world market. (R.J.S.)

  5. Theoretical and experimental studies of non-linear structural dynamics of fast breeder reactor fuel elements

    International Nuclear Information System (INIS)

    Descriptions are presented of theoretical and experimental studies of the deformation behaviour of fast-breeder fuel elements as a consequence of extreme impulsive stresses produced by an incident. The starting point for the studies is the assumption that local disturbances in a fuel element have resulted in a thermal interaction between fuel and sodium and in a corresponding increase in pressure. On the basis of the current state of knowledge, the possibility cannot be ruled out that this pressure build-up may lead to the bursting of the fuel-element wrapper, to the propagation of pressure in the core, and to coherent structural movements and deformations. A physical model is established for the calculation of the dynamic response of elastic-plastic beam systems, and the differential equations of p motion for the discrete equivalent system are derived with the aid of D'Alembert's principle. On this basis and with the aid of a semi-empirical pin-bundle model, an appropriate computer program allows a static and dynamic analysis to be obtained for a complete fuel element. In the experimental part of the study, a description is given of static and impulsive loading tests on 1:1 SNR-like fuel-element models. Making use of measured impact forces and of known material characteristics, it was possible to a large extent for the experiments to be reproduced by calculations. In agreement with existing experience from explosion experiments on 1:1 core models, the results (of relevance for fast-breeder safety and in particular the SNR-300) show that only local limited deformations occur and that the compact fuel-element and core structure constitutes an effective inherent barrier in the presence of extreme incident stresses. (author)

  6. Welding device for nuclear fuel assembly structure elements

    International Nuclear Information System (INIS)

    The device has two parallel assembling positions next to each other. The welding robot is carried by a carriage with displacement parallel to the guide tubes and has enough degrees of freedom to move from one assembling position to the other and have access to the structural elements

  7. Automatic X-ray inspection for the HTR-PM spherical fuel elements

    International Nuclear Information System (INIS)

    Highlights: • An automatic X-ray inspection method is established to characterize HTR pebbles. • The method provides physical characterization and the inner structure of pebbles. • The method can be conducted non-destructively, quickly and automatically. • Sample pebbles were measured with this AXI method for validation. • The method shows the potential to be applied in situ. - Abstract: Inefficient quality assessment and control (QA and C) of spherical fuel elements for high temperature reactor-pebblebed modules (HTR-PM) has been a long-term problem, since conventional methods are labor intensive and cannot reveal the inside information nondestructively. Herein, we proposed a nondestructive, automated X-ray inspection (AXI) method to characterize spherical fuel elements including their inner structures based on X-ray digital radiography (DR). Briefly, DR images at different angles are first obtained and then the chosen important parameters such as spherical diameters, geometric and mass centers, can be automatically extracted and calculated via image processing techniques. Via evaluating sample spherical fuel elements, we proved that this AXI method can be conducted non-destructively, quickly and automatically. This method not only provides accurate physical characterization of spherical fuel elements but also reveals their inner structure with good resolution, showing great potentials to facilitate fast QA and C in HTM-PM spherical fuel element development and production

  8. Standard laboratory hydraulic pressure drop characteristics of various solid and I&E fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Waters, E.D.; Horn, G.R.

    1958-01-20

    The purpose of this report is to present a set of standard pressure-drop curves for various fuel elements in process tubes of Hanford reactors. The flow and pressures within a process tube assembly under normal conditions are dependent to a large extent on the magnitude of the pressure drop across the fuel elements. The knowledge of this pressure drop is important in determination of existing thermal conditions within the process tubes and in predicting conditions for new fuel element designs or changes in operating conditions. The pressure-flow relations for the different Hanford fuel element-process tube assemblies have all been determined at one time or another in the 189-D Hydraulics Laboratory but the data had never been collected into a single report. Such a report is presented now in the interest of establishing a set of ``standard curves`` as determined by laboratory investigations. It must be recognized that the pressure drops of fuel elements in actual process tubes in the reactors may be slightly different than those reported here. The data presented here were obtained in new process tubes while reactor process tubes are usually either corroded or filmed, depending on their past history.

  9. Latest development on the disposal of Research Reactor Fuel and Triga Fuel elements

    International Nuclear Information System (INIS)

    The MTR spent nuclear fuel reprocessing by the UKAEA, the intermediate storage+direct disposal for research reactors, the research reactor spent nuclear fuel return to the U.S., shipments and ports of entry, management sites, fees, storage technologies, contracts, actual shipments, legal processes, and NUKEM activities are listed. (HSI)

  10. Analytical Solution of Fick's Law of the TRISO-Coated Fuel Particles and Fuel Elements in Pebble-Bed High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Two kinds of approaches are built to solve the fission products diffusion models (Fick's equation) based on sphere fuel particles and sphere fuel elements exactly. Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented, respectively. The analytical solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation. In the fuel element system, a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element. Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations. (general)

  11. Analytical solution of Fick's law of the TRISO-coated fuel particles and fuel elements in pebble-bed high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Two kinds of approaches are built to solve the fission products diffusion models (Fick's equation) based on sphere fuel particles and sphere fuel elements exactly. Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented, respectively. The analytical solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation. In the fuel element system, a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element. Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations. (authors)

  12. Irradiation behaviour of advanced fuel elements for the helium-cooled high temperature reactor (HTR)

    International Nuclear Information System (INIS)

    The design of modern HTRs is based on high quality fuel. A research and development programme has demonstrated the satisfactory performance in fuel manufacturing, irradiation testing and accident condition testing of irradiated fuel elements. This report describes the fuel particles with their low-enriched UO2 kernels and TRISO coating, i.e. a sequence of pyrocarbon, silicon carbide, and pyrocarbon coating layers, as well as the spherical fuel element. Testing was performed in a generic programme satisfying the requirements of both the HTR-MODUL and the HTR 500. With a coating failure fraction less than 2x10-5 at the 95% confidence level, the results of the irradiation experiments surpassed the design targets. Maximum accident temperatures in small, modular HTRs remain below 1600deg C, even in the case of unrestricted core heatup after depressurization. Here, it was demonstrated that modern TRISO fuels retain all safety-relevant fission products and that the fuel does not suffer irreversible changes. Isothermal heating tests have been extended to 1800deg C to show performance margins. Ramp tests to 2500deg C demonstrate the limits of present fuel materials. A long-term programm is planned to improve the statistical significance of presently available results and to narrow remaining uncertainty limits. (orig.)

  13. FEM analysis of the rolled bottom nozzle of 16NGF fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Schettino, Carlos Frederico Mattos; Brittes, Luiz Henrique Alves; Silva, Marcio Adriano Coelho da [Industrias Nucleares do Brasil (INB), Resende, RJ (Brazil)]. E-mails: carlosschettino@inb.gov.br; brittes@inb.gov.br; marcio.adriano@inb.gov.br

    2007-07-01

    The present work aims to evaluate structurally the Rolled Bottom Nozzle (previously casting) used in the fuel element type 16 x 16 used in the Nuclear Power Plant Angra I. The full solid model of this component was generated with SOLIDWORKS program and later imported to ANSYS 10.0 program. For the Finite Element model were used, elements SOLID-92 and BEAM4. The analysis covered specific loads simulating the conditions found during the shipping and handling of the Fuel Element (static loads corresponding to 4g - four times the Fuel Element weight) as well as simulating the conditions found during the operation of the nuclear power plant (Conditions I, II, III and IV). The structural integrity of the Bottom Nozzle is assured when the design criteria defined in the ASME Boiler and Pressure Vessel Code Section III is satisfied. The results of these analyses were used to prove that the Rolled Bottom Nozzle is capable to keep the dimensional stability for which it was designed when submitted to loads that correspond to required stresses for its use in Nuclear Fuel Elements. The performed analysis provided INB to get more information of extreme importance for the continuity of the development of the welded Top Nozzle and its later production. (author)

  14. Axisymmetric annular curtain stability

    International Nuclear Information System (INIS)

    A temporal stability analysis was carried out to investigate the stability of an axially moving viscous annular liquid jet subject to axisymmetric disturbances in surrounding co-flowing viscous gas media. We investigated in this study the effects of inertia, surface tension, the gas-to-liquid density ratio, the inner-to-outer radius ratio and the gas-to-liquid viscosity ratio on the stability of the jet. With an increase in inertia, the growth rate of the unstable disturbances is found to increase. The dominant (or most unstable) wavenumber decreases with increasing Reynolds number for larger values of the gas-to-liquid viscosity ratio. However, an opposite tendency for the most unstable wavenumber is predicted for small viscosity ratio in the same inertia range. The surrounding gas density, in the presence of viscosity, always reduces the growth rate, hence stabilizing the flow. There exists a critical value of the density ratio above which the flow becomes stable for very small viscosity ratio, whereas for large viscosity ratio, no stable flow appears in the same range of the density ratio. The curvature has a significant destabilizing effect on the thin annular jet, whereas for a relatively thick jet, the maximum growth rate decreases as the inner radius increases, irrespective of the surrounding gas viscosity. The degree of instability increases with Weber number for a relatively large viscosity ratio. In contrast, for small viscosity ratio, the growth rate exhibits a dramatic dependence on the surface tension. There is a small Weber number range, which depends on the viscosity ratio, where the flow is stable. The viscosity ratio always stabilizes the flow. However, the dominant wavenumber increases with increasing viscosity ratio. The range of unstable wavenumbers is affected only by the curvature effect. (paper)

  15. Analytical Solution of Fick's Law of the TRISO-Coated Fuel Particles and Fuel Elements in Pebble-Bed High Temperature Gas-Cooled Reactors

    Institute of Scientific and Technical Information of China (English)

    CAO Jian-Zhu; FANG Chao; SUN Li-Feng

    2011-01-01

    T wo kinds of approaches are built to solve the fission products diffusion models (Fick's equation) based on sphere fuel particles and sphere fuel elements exactly. Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented, respectively. The analytica,solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation.In the fuel element system, a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element. Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations.%@@ Two kinds of approaches are built to solve the fission products diffusion models(Fick's equation) based on sphere fuel particles and sphere fuel elements exactly.Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented,respectively.The analytical solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation.In the fuel element system,a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element.Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations.

  16. Reactivity measurements of the IPR-R1 TRIGA reactor fuel elements

    International Nuclear Information System (INIS)

    The thermal power of the IPR-R1 TRIGA reactor, belonging to the Centro de Desenvolvimento da Tecnologia Nuclear, will be upgraded from 100 k W to 250 k W. To attain this objective, mew additional fuel elements will be inserted in the reactor core. In order to provide information to the calculations of the new core arrangement, some fuel rods reactivity measurements were carried out as well as the determination of the reactivity increase due to the substitution of the present fuel by a new one. A first estimate indicates that the addition of 5 new fuel elements might be sufficient to reach the desired value of 3$ ρ excess. (author). 5 refs., 1 fig., 2 tabs

  17. RA3: Application of a calculation model for fuel management with SEFE (Slightly Enriched Fuel Elements)

    International Nuclear Information System (INIS)

    The RA-3 (5 MW, MTR) reactor is mainly utilized to produce radioisotopes (Mo-99, I-131, etc.). It started operating with Low Enrichment Uranium (LEU) in 1990, and spends around 12 fuels per year. Although this consumption is small compared to a nuclear power station. It is important to do a good management of them. The present report describes: - A reactor model to perform the Fuel Shuffling. - Results of fuel management simulations for 2 and a half years of operation. Some features of the calculations can be summarized as follows: 1) A 3D calculation model is used with the code PUMA. It does not have experimental adjustments, except for some approximations in the reflector representation and predicts: power, flux distributions and reactivity of the core in an acceptable way. 2) Comparisons have been made with the measurements done in the commissioning with LEU fuels, and it has also been compared with the empirical method (the previous one) which had been used in the former times of operation with LEU fuel. 3) The number of points of the model is approximately 13500, an it can be run in 80386 personal computer. The present method has been verified as a good tool to perform the simulations for the fuel management of RA-3 reactor. It is expected to produce some economic advantages in: - Achieving a better utilization of the fuels. - Leaving more time of operation for radioisotopes production. The activation measurements through the whole core required by the previous method can be significantly reduced. (author)

  18. Determination of the minimum number of fuel elements of the RP-10 research reactor

    International Nuclear Information System (INIS)

    The peruvian research reactor RP-10 is composed of a compound nucleus of boxes containing fuel plates which are cooled with light water in order to remove heat produced by fission of uranium atoms. However from a certainty viewpoint, it exists certain restrictions to design the cooling system. The most admissible caloric flux of 90.3 watts/cm2 is deflux of 90.3 watts/cm2 is determined on the basis of these thermic restrictions when cooling speed is 409 cm/sec permitting at least 24 fuel elements(boxes) within the nucleus. On the basis of restrictions of load loss in the nucleus, it would be permitted at least 18 fuel elements, but this quantity breaks thermic restrictions for this reason, 24 boxes in the nucleus will be the minimum number of elements

  19. Determination of the fuel element burn-up for mixed TRIGA core by measurement and calculation with new TRIGLAV code

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, T.; Ravnik, M.; Persic, A. (J.Stefan Institute, Ljubljana (Slovenia))

    1999-12-15

    Results of fuel element burn-up determination by measurement and calculation are given. Fuel element burn-up was calculated with two different programs TRIGLAV and TRIGAC using different models. New TRIGLAV code is based on cylindrical, two-dimensional geometry with four group diffusion approximation. TRIGAC program uses one-dimensional cylindrical geometry with twogroup diffusion approximation. Fuel element burn-up was measured with reactivity method. In this paper comparison and analysis of these three methods is presented. Results calculated with TRIGLAV show considerably better alignment with measured values than results calculated with TRIGAC. Some two-dimensional effects in fuel element burn-up can be observed, for instance smaller standard fuel element burn-up in mixed core rings and control rod influence on nearby fuel elements. (orig.)

  20. Accident simulations and post irradiation examinations on spherical fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    An important aspect of the safety of high temperature reactors is the quality of the nuclear fuel and its ability to remain intact even at high temperatures and to safely contain the radioactive fission products. In combination with a suitable reactor an inherent safety against large release of fission products can be achieved. In this work experimental simulations of severe accidents were conducted on spherical fuel elements for high temperature reactors with TRISO-coated particles and fission product release was measured. The fuel elements originated from various irradiation experiments conducted at high temperatures with high burn-up. The experiments were performed using the cold finger apparatus, a test apparatus which was already used in the past in a former version at the Research Center Juelich. The new cold finger apparatus is installed since 2005 in the Hot Cells of the European Institute for Transuranium Elements. The cold finger apparatus at the Institute for Transuranium enabled incident simulations on irradiated high temperature reactor fuel elements in a helium atmosphere at ambient pressure, at temperatures up to 1800 C and for periods of several hundred hours. Here, both the release of fission gases and the release of solid fission products were measured. In addition, in the context of the present study, the mechanical behavior of the fuel particles and the transport mechanisms of the main fission products were analyzed and the expected release was computed. For a better understanding of the processes post irradiation examinations were conducted on the available fuel elements. It was finally made an assessment of the test results which were compared with results in the existing literature. A key objective of the work was the extension of the existing data base for modern HTR-fuel towards higher burn-up and higher fluences of fast neutrons, higher operating temperatures and extended accident temperatures.

  1. Vibration analysis of annular-like plates

    Science.gov (United States)

    Cheng, L.; Li, Y. Y.; Yam, L. H.

    2003-05-01

    The existence of eccentricity of the central hole for an annular plate results in a significant change in the natural frequencies and mode shapes of the structure. In this paper, the vibration analysis of annular-like plates is presented based on numerical and experimental approaches. Using the finite element analysis code Nastran, the effects of the eccentricity, hole size and boundary condition on vibration modes are investigated systematically through both global and local analyses. The results show that analyses for perfect symmetric conditions can still roughly predict the mode shapes of "recessive" modes of the plate with a slightly eccentric hole. They will, however, lead to erroneous results for "dominant" modes. In addition, the residual displacement mode shape is verified as an effective parameter for identifying damage occurring in plate-like structures. Experimental modal analysis on a clamped-free annular-like plate is performed, and the results obtained reveal good agreement with those obtained by numerical analysis. This study provides guidance on modal analysis, vibration measurement and damage detection of plate-like structures.

  2. Irradiation of Superheater Test Fuel Elements in the Steam Loop of the R2 Reactor

    International Nuclear Information System (INIS)

    The design, fabrication, irradiation results, and post-irradiation examination for three superheater test fuel elements are described. During the spring of 1966 these clusters, each consisting of six fuel rods, were successfully exposed in the superheater loop No. 5 in the R2 reactor for a maximum of 24 days at a maximum outer cladding surface temperature of ∼ 650 deg C. During irradiation the linear heat rating of the rods was in the range 400-535 W/cm. The diameter of the UO2 pellets was 11.5 and 13.0 mm; the wall thickness of the 20/25 Nb and 20/35 cladding was in every case 0.4 mm. The diametrical gap between fuel and cladding was one of the main parameters and was chosen to be 0.05, 0.07 and 0.10 mm. These experiments, to be followed by one high cladding temperature irradiation (∼ 750 deg C) and one long time irradiation (∼ 6000 MWd/tU), were carried out to demonstrate the operational capability of short superheater test fuel rods at steady and transient operational environments for the Marviken superheater fuel elements and also to provide confirmation of design criteria for the same fuel elements

  3. New Nuclear Materials Including Non Metallic Fuel Elements. Vol. I. Proceedings of the Conference on New Nuclear Materials Technology, Including Non Metallic Fuel Elements

    International Nuclear Information System (INIS)

    One of the major aims of the International Atomic Energy Agency in furthering the peaceful uses of atomic energy is to encourage the development of economical nuclear power. Certainly, one of the more obvious methods of producing economical nuclear power is the development of economical fuels that can be used at high temperatures for long periods of time, and which have sufficient strength and integrity to operate under these conditions without permitting the release of fission products. In addition it is desirable that after irradiation these new fuels be economically reprocessed to reduce further the cost of the fuel cycle. As nuclear power becomes more and more competitive with conventional power the interest in new and more efficient higher-temperature fuels naturally increases rapidly. For these reasons, the Agency organized a Conference on New Nuclear Materials Technology, Including Non-Metallic Fuel Elements, which was held from 1 to 5 July 1963 at the International Hotel, Prague, with the assistance and co-operation of the Government of the Czechoslovak Socialist Republic. A total of 151 scientists attended, from 23 countries and 4 international organizations. The participants heard and discussed more than 60 scientific papers

  4. Process for extracting and preparing fuel elements from a reactor for transport and device for carrying out the process

    International Nuclear Information System (INIS)

    The fuel element is lifted from the reactor using the lifting device of a bridge crane, is taken over the opening of a transport container, and is lowered into this. During transfer the fuel element is situated in a shielding hood at the end of the crane chain. The transport container contains liquid salt or a slat mixture, which solidifies after one or several fuel elements are inserted in it. (DG)

  5. Burnup measurements on spent fuel elements of the RP-10 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  6. Grain growth and temperature distribution in the irradiated fuel elements designed for Qinshan Power Plant

    International Nuclear Information System (INIS)

    The micro-analysis results of UO2 fuel core of irradiated test elements designed for Qinshan Power Plant are presented. The temperature of element centers is quantitatively calculated by using the grain size measurement results according to the principle of grain growth dynamics, and the results are in accordance with that of the FRAP-CON calculation. Based on the post-irradiation examination data, the dynamic parameters, K and Q which are particularly suitable to description of irradiation behaviours of UO2 fuel made in China are deduced

  7. Reactivity change measurements on plutonium-uranium fuel elements in hector experimental techniques and results

    International Nuclear Information System (INIS)

    The techniques used in making reactivity change measurements on HECTOR are described and discussed. Pile period measurements were used in the majority of oases, though the pile oscillator technique was used occasionally. These two methods are compared. Flux determinations were made in the vicinity of the fuel element samples using manganese foils, and the techniques used are described and an error assessment made. Results of both reactivity change and flux measurements on 1.2 in. diameter uranium and plutonium-uranium alloy fuel elements are presented, these measurements being carried out in a variety of graphite moderated lattices at temperatures up to 450 deg. C. (author)

  8. Safety analysis report for packaging: the ORNL HFIR spent-fuel-element shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.H.; Chipley, K.K.; Eversole, R.E.; Just, R.A.; Llewellyn, G.H.

    1977-11-01

    The Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) spent-fuel-element shipping cask is used to transport HFIR, Oak Ridge Research Reactor (ORR), and other reactor fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures and tests were used to determine behavior of the cask relative to the general standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations.

  9. Development of eddy current test system for fuel element based on LabVIEW

    International Nuclear Information System (INIS)

    Fuel element plays an important role in high temperature gas-cooled reactor-pebble-bed module. For adjusting the fuel element precisely, an eddy current test system based on LabVIEW was developed to count the plumbago ball precisely and select the defective balls. The system was composed of the hardware circuit, the computer, the data acquisition card and relative software. The design of the excitation source, head amplifier circuit and the phase-sensitive detector was introduced in detail. The plumbago balls were counted and defects were tested by this system , and the results showed that the system is with good test capability. (authors)

  10. Determination of the burn-up of TRIGA fuel elements by calculation and reactivity experiments

    International Nuclear Information System (INIS)

    The burnup of 17 fuel elements of the TRIGA Mark-II reactor in Vienna was measured. Different types of fuel elements had been simultaneously used for several years. The measured burnup values are compared with those calculated on the basis of core configuration and reactor operation history records since the beginning of operation. A one-dimensional, two-group diffusion computer code TRIGAP was used for the calculations. Comparison with burnup values determined by γ-scanning is also made. (orig./HP)

  11. Study of causes of can leakage in original WWER and RBMK fuel elements

    International Nuclear Information System (INIS)

    The post-irradiation examination is accomplished for four WWER-1000 spent fuel assemblies (SFA) and three RBMK-1000 SFAs accepted to be failed in the course of inspection. From the investigation results and in accordance with the literature data the following explanation of the reasons for clad leakage is given: local mechanical damages (primary defects) of fuel cans arise from the interaction with debris entered the coolant circuit to the extent that through holes can be formed. The penetration of the coolant into a fuel element results in hydrogenation and embrittlement of various sections of fuel cans (secondary defects). Primary defects are noted to be of frequent occurrence near spacer grids. To reveal defective WWER-1000 fuel assemblies in operation it is necessary that the data on the primary circuit coolant activity should be taken into account

  12. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2010-02-23

    Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

  13. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  14. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  15. Feasibility study of modeling a CANDU fuel element using a multiphysics object-oriented simulation environment

    International Nuclear Information System (INIS)

    The first phase of the feasibility study of using a Multiphysics Object-Oriented Simulation Environment (MOOSE) for modeling a CANDU fuel element is presented. A two-dimensional model of a fuel pellet sheath was created to examine the contact algorithm within MOOSE. The results obtained show the expected behaviour of contact pressure and penetration in 2D. Preliminary results for a 3D model of a quarter fuel pellet and sheath are provided but at present contain anomalies currently being investigated. The next steps in the feasibility study are outlined. (author)

  16. Development of Non-Metallic Fuel Elements for a High-Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    In connection with fuel element development work for the high-temperature gas-coolcd reactor of the Brown-Boveri/Krupp Reaktorbau G.m.b.H., two different fuel element concepts were considered and developed. In both cases the fuel element consists of a graphite ball of 6 cm in diam. which contains the fuel insert, a cylindrical pellet of about 20 mm in diam. and 16 mm in height. The two concepts differ in the type of the.fuel insert as well as in the preparation of the graphite ball. In the first concept the fuel insert consists of a mixture of UC2 and graphite which is prepared by blending U3O8 and graphite, pressing them into pellets and reacting the two components in a vacuum furnace at 1800oC. The atomic ratio of U : C is 1:45. Since this type of fuel pellet does not retain the fission products completely the surrounding graphite sphere had to be made impervious to fission products by impregnation in order to obtain a fission-product retaining element. Permeabilities of the order of 10-6cm2/s could be achieved. In the second concept the fuel insert consists of a solid solution of UC in ZrC and is coated with a layer of ZrC. The molar ratio of UC to ZrC is 1 : 20. The fuel pellet preparation was accomplished by the following procedure: UO2, ZrO2, and graphite were mixed and pressed into pellets. The pellets were reacted to the carbides. Ball milling of the carbides was followed by hot pressing at temperatures o f 2000oC. Densities of more than 95% of the theoretical density could be achieved. A full description of the preparation and of some physical properties of the fuel pellets is given in the paper. A sufficient fission gas retention behaviour of this type of fuel insert which allows it to be put into unimpregnated graphite balls is expected. Other advantages of this kind of fuel are discussed. (author)

  17. Feasibility study of modeling a CANDU fuel element using a multiphysics object-oriented simulation environment

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, K., E-mail: Kyle.Gamble@rmc.ca [Royal Military College of Ontario, Kingston, Ontario (Canada); Williams, A. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Chan, P.K. [Royal Military College of Ontario, Kingston, Ontario (Canada)

    2013-07-01

    The first phase of the feasibility study of using a Multiphysics Object-Oriented Simulation Environment (MOOSE) for modeling a CANDU fuel element is presented. A two-dimensional model of a fuel pellet sheath was created to examine the contact algorithm within MOOSE. The results obtained show the expected behaviour of contact pressure and penetration in 2D. Preliminary results for a 3D model of a quarter fuel pellet and sheath are provided but at present contain anomalies currently being investigated. The next steps in the feasibility study are outlined. (author)

  18. Final disposal of dissolver sludges, claddings, fuel hardware, and spent HTGR fuel elements in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    In the Federal Republic of Germany, radioactive waste of marked decay heat production shall be disposed of in vertical 300-m-deep boreholes in a salt dome formation. The disposal technique for heat generating intermediate-level wastes arising from LWR fuel reprocessing and for spent HTGR pebble bed fuel is referred to as the ILW borehole technique. This technique is under development since 1983 in a R and D project for intermediate-level waste and spent HTGR fuel element test disposal in boreholes. This paper reports on the main objectives of this project, to establish the ILW borehole technique by developing the emplacement and plugging concept up to design and testing of prototype components and testing of the multicomponent technique under real conditions in the ASSE salt mine

  19. Current research projects at the Austrian TRIGA Mark II. Location of failed fuel elements in Austrian TRIGA Mark II

    International Nuclear Information System (INIS)

    The system developed at the Atominstitut monitors the radioactive Krypton- and Xenon nuclides in the primary water circuitry and allows selective control of any fuel element for its fission gas release. A suspected fuel element is enclosed in an underwater capsule attached in the reactor tank. Water is pumped along the fuel element to a vacuum degasser where the gases are separated from the tank water. The degassed water is returned to the reactor pool while the gases are pumped to a very sensitive proportional counter. The fuel elements of the TRIGA core were checked by the described procedure

  20. Trace elements in co-combustion of solid recovered fuel and coal

    DEFF Research Database (Denmark)

    Wu, Hao; Glarborg, Peter; Jappe Frandsen, Flemming;

    2013-01-01

    Trace element partitioning in co-combustion of a bituminous coal and a solid recovered fuel (SRF) was studied in an entrained flow reactor. The experiments were carried out at conditions similar to pulverized coal combustion, with SRF shares of 7.9 wt.% (wet basis), 14.8 wt.% and 25.0 wt.%. In ad......Trace element partitioning in co-combustion of a bituminous coal and a solid recovered fuel (SRF) was studied in an entrained flow reactor. The experiments were carried out at conditions similar to pulverized coal combustion, with SRF shares of 7.9 wt.% (wet basis), 14.8 wt.% and 25.0 wt......-based additives increased the volatility of Cd, Pb and As, whereas addition of ammonium sulphate generally decreased the volatility of trace elements. Addition of kaolinite reduced the volatility of Pb, while the influence on other trace elements was insignificant. The results from the present work imply...

  1. An organization of the thorium fuel cycle start on the basis of fast reactors with spherical fuel elements of the small size

    International Nuclear Information System (INIS)

    The possibility of the organization of thorium fuel cycle start by means of conversion of high background plutonium into isotopically pure Uranium 233 into the highly stressed breeders with the fuel in the form of spherical fuel elements has been studied. A high efficiency of usage of compact plutonium fuel in the form of spherical fuel elements for its transmutation into low background Uranium 233 has been shown as a result of the revealed temporary regularities in the main characteristic behaviour of the reactors of such a type. (authors). 7 refs., 2 figs., 1 tab

  2. Annular Hybrid Rocket Motor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Engineers at SpaceDev have conducted a preliminary design and analysis of a proprietary annular design concept for a hybrid motor. A U.S. Patent application has...

  3. Manufacture of annular cermet articles

    Science.gov (United States)

    Forsberg, Charles W.; Sikka, Vinod K.

    2004-11-02

    A method to produce annular-shaped, metal-clad cermet components directly produces the form and avoids multiple fabrication steps such as rolling and welding. The method includes the steps of: providing an annular hollow form with inner and outer side walls; filling the form with a particulate mixture of ceramic and metal; closing, evacuating, and hermetically sealing the form; heating the form to an appropriate temperature; and applying force to consolidate the particulate mixture into solid cermet.

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

    International Nuclear Information System (INIS)

    The first low-enriched U3Si2-Al dispersion plate-type fuel element produced at the Nuclear Fuel Element Center, BATAN, Indonesia, was irradiated to a peak 235U burnup of 62%. Postirradiation examinations performed to data shows the irradiation behavior of this element to be similar to that of U3Si2-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

  5. An estimate of the reactivity of assemblies of NRX fuel elements in light water

    International Nuclear Information System (INIS)

    This report contains calculations on the criticality of assemblies of NRX fuel elements in light water. The elements are dealt with in three sections, 'X rods' of natural uranium, enriched elements of U235/A1 alloy and enriched elements of Pu/Al alloy. Values of k∞ and B2 are provided for two fuel concentrations for each of the two enriched types and for a range of irradiations of the X rods. The calculations for the X rods provide maximum and minimum values of k∞. The maximum values for some lattices are a few per cent above unity. Unfortunately, the present experimental evidence does not prove that it is impossible to achieve values of k∞ greater than unity in lattices of natural uranium in light water. Hence for safety predictions maximum values have been used. The resulting restrictions are not very severe. It is possible to make critical assemblies of the enriched elements, Part (5) contains a set of recommended minimum spacings such that elements of all kinds may safely be mixed in a stack together. There are also predictions of the minimum critical numbers of complete elements or elements cut into slugs. (author)

  6. Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E.R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs

  7. New Nuclear Materials Including Non Metallic Fuel Elements. Vol. II. Proceedings of the Conference on New Nuclear Materials Technology, Including Non Metallic Fuel Elements

    International Nuclear Information System (INIS)

    One of the major aims of the International Atomic Energy Agency in furthering the peaceful uses of atomic energy is to encourage the development of economical nuclear power. Certainly, one of the more obvious methods of producing economical nuclear power is the development of economical fuels that can be used at high temperatures for long periods of time, and which have sufficient strength and integrity to operate under these conditions without permitting the release of fission products. In addition it is desirable that after irradiation these new fuels be economically reprocessed to reduce further the cost of the fuel cycle. As nuclear power becomes more and more competitive with conventional power the interest in new and more efficient higher-temperature fuels naturally increases rapidly. For these reasons, the Agency organized a Conference on New Nuclear Materials Technology, Including Non-Metallic Fuel Elements, which was held from 1 to 5 July 1963 at the International Hotel, Prague, with the assistance and co-operation of the Government of the Czechoslovak Socialist Republic. A total of 151 scientists attended, from 23 countries and 4 international organizations. The participants heard and discussed more than 60 scientific papers. The Agency wishes to thank the scientists who attended this Conference for their papers and for many spirited discussions that truly mark a successful meeting. The Agency wishes also to record its gratitude for the assistance and generous hospitality accorded the Conference, the participants and the Agency's staff by the Government of the Czechoslovak Socialist Republic and by the people of Prague. The scientific information contained in these Proceedings should help to quicken the pace of progress in the fabrication of new and m ore economical fuels, and it is hoped that these proceedings will be found useful to all workers in this and related fields

  8. Thermomechanical evaluation of BWR fuel elements for procedures of preconditioned with FEMAXI-V

    International Nuclear Information System (INIS)

    The limitations in the burnt of the nuclear fuel usually are fixed by the one limit in the efforts to that undergo them the components of a nuclear fuel assembly. The limits defined its provide the direction to the fuel designer to reduce to the minimum the fuel failure during the operation, and they also prevent against some thermomechanical phenomena that could happen during the evolution of transitory events. Particularly, a limit value of LHGR is fixed to consider those physical phenomena that could lead to the interaction of the pellet-shirt (Pellet Cladding Interaction, PCI). This limit value it is related directly with an PCI limit that can be fixed based on experimental tests of power ramps. This way, to avoid to violate the PCI limit, the conditioning procedures of the fuel are still required for fuel elements with and without barrier. Those simulation procedures of the power ramp are carried out for the reactor operator during the starting maneuvers or of power increase like preventive measure of possible consequences in the thermomechanical behavior of the fuel. In this work, the thermomechanical behavior of two different types of fuel rods of the boiling water reactor is analyzed during the pursuit of the procedures of fuel preconditioning. Five diverse preconditioning calculations were carried out, each one with three diverse linear ramps of power increments. The starting point of the ramps was taken of the data of the cycle 8 of the unit 1 of the Laguna Verde Nucleo electric Central. The superior limit superior of the ramps it was the threshold of the lineal power in which a fuel failure could be presented by PCI, in function of the fuel burnt. The analysis was carried out with the FEMAXI-V code. (Author)

  9. Thermal simulations and tests in the development of a helmet transport spent fuel elements Research Reactor

    International Nuclear Information System (INIS)

    A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half-scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled. (author)

  10. Finite difference analysis of the transient temperature profile within GHARR-1 fuel element

    International Nuclear Information System (INIS)

    Highlights: • Transient heat conduction for GHARR-1 fuel was developed and simulated by MATLAB. • The temperature profile after shutdown showed parabolic decay pattern. • The recorded temperature of about 411.6 K was below the melting point of the clad. • The fuel is stable and no radioactivity will be released into the coolant. - Abstract: Mathematical model of the transient heat distribution within Ghana Research Reactor-1 (GHARR-1) fuel element and related shutdown heat generation rates have been developed. The shutdown heats considered were residual fission and fission product decay heat. A finite difference scheme for the discretization by implicit method was used. Solution algorithms were developed and MATLAB program implemented to determine the temperature distributions within the fuel element after shutdown due to reactivity insertion accident. The simulations showed a steady state temperature of about 341.3 K which deviated from that reported in the GHARR-1 safety analysis report by 2% error margin. The average temperature obtained under transient condition was found to be approximately 444 K which was lower than the melting point of 913 K for the aluminium cladding. Thus, the GHARR-1 fuel element was stable and there would be no release of radioactivity in the coolant during accident conditions

  11. Mass transport in annular spherical system

    Science.gov (United States)

    Bauer, Helmut F.

    The mass transport between two concentric spheres with inlet and outlet at the poles was determined for ideal liquid flow (plug flow) and laminar flow for constant concentration at the spherical walls and constant concentration at the inlet. Velocity distribution and local concentration profiles were determined analytically for various widths of the annular spherical conduit and various diffusive flow parameters. It is found that with the increase of this parameter, the decay becomes quite rapid and that the same effect occurs for increasing diameter ratio of the spheres. This configuration may possibly be used as a basic element of an artificial kidney.

  12. Apparatus for the storage of transport- and storage-containers containing radioactive fuel elements

    International Nuclear Information System (INIS)

    The invention concerns an apparatus for the storage of transport and storage containers containing radioactive fuel elements. For each transport or storage container there is a separate silo-type container of steel, concrete, prestressed concrete or suchlike breakproof and fireproof material, to be placed in the open, that can be opened for removal and placing of the transport or storage container respectively. (orig.)

  13. Welding procedures used in the fabrication of fuel elements for the DON Reactor exponential experiment

    International Nuclear Information System (INIS)

    This exponential experiment required 74 units (37 loaded with UO2 and 37 with UC) to simulate the Reactor fuel channels. Each unit was enclosed in a tube similar to the calandria ones. It contained the pressure tube, the shroud and the 19 rods cluster. Within the pressure tube, in touch with the elements, was the organic liquid. (Author)

  14. Eddy current examination of the nuclear fuel elements of IPR-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Roger F.; Frade, Rangel T.; Oliveira, Paulo F.; Silva, Marlucio A.; Silva Junior, Silverio F., E-mail: rfs@cdtn.br, E-mail: rtf@cdtn.br, E-mail: pfo@cdtn.br, E-mail: mas@cdtn.br, E-mail: silvasf@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    Tubes of AISI 304 stainless steel as well as tubes of Aluminum 1100-F are used as cladding of the fuel elements of TRIGA MARK 1 nuclear research reactor. Usually, these tubes are periodically inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements in which the cladding has failed, but it is not able to determine the place where the discontinuity is located. In turn, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In this paper, a study about the use of eddy current testing for detection and characterization of discontinuities in the fuel elements cladding is proposed. The study involves the development of probes able to operate in underwater inspections, the design and manufacture of reference standards and the development of a test methodology to perform the evaluations. (author)

  15. Eddy current examination of the nuclear fuel elements of IPR-R1 research reactor

    International Nuclear Information System (INIS)

    Tubes of AISI 304 stainless steel as well as tubes of Aluminum 1100-F are used as cladding of the fuel elements of TRIGA MARK 1 nuclear research reactor. Usually, these tubes are periodically inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements in which the cladding has failed, but it is not able to determine the place where the discontinuity is located. In turn, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In this paper, a study about the use of eddy current testing for detection and characterization of discontinuities in the fuel elements cladding is proposed. The study involves the development of probes able to operate in underwater inspections, the design and manufacture of reference standards and the development of a test methodology to perform the evaluations. (author)

  16. End plug welding of FBTR fuel element by Nd-YAG CW laser

    International Nuclear Information System (INIS)

    Currently thin walled SS-316 FBTR fuel element is closed at both ends with SS solid plug using TIG welding at Radiometallurgy Division, Bhabha Atomic Research Centre. This welding is involved with material of non uniform thickness. The tube is thinner whereas solid plug is thicker material. Precise control of welding parameters and positioning of torch is necessary during welding to get satisfactory penetration without any major defect. In order to meet the objective of achieving good quality weld heat input during welding has to be kept minimal. Of late in Radiometallurgy Division, Bhabha Atomic Research Centre Nd-YAG CW laser welding has been introduced for end plug welding of the fuel element. Necessary experiments have been done to evaluate the performance of laser welding process for the welding of FBTR fuel elements. Laser welding of fuel element has been done at different power level and speeds. The depth of penetration, fusion zone shape, overall weld profile and weld bead thickness have been evaluated. The amount of liquid metal formed during welding has been calculated theoretically by fitting the overall weld profile in mathematical function. The weld profile area has been estimated by image analysis technique. From these data, the heat absorption by the weld pool have been calculated and thermal efficiency has been estimated. Attempt has been made to optimize laser welding parameters. (author)

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

    International Nuclear Information System (INIS)

    An attempt to obtain high uranium-loading in silicide dispersion fuel element using the fabrication technology applicable nowadays can reach Uranium-loading slightly above 5 gU/cm3. It is difficult to achieve a higher uranium-loading than that because of fabricability constraints. To overcome those difficulties, the use of uranium silicide U3Si based is considered. The excess of U is obtained by synthesising U3Si2 in Si-hypostoichiometric stage, without applying heat treatment to the ingot as it can generate undesired U3Si. The U U will react with the matrix to form U alx compound, that its pressure is tolerable. This experiment is to consider possibilities of employing the U3Si2 as nuclear fuel element which have been performed by synthesising U3Si2-U with the composition of 3.7 % weigh and 3 % weigh U. The ingot was obtained and converted into powder form which then was fabricated into experimental plate nuclear fuel element. The interaction between free U and Al-matrix during heat-treatment is the rolling phase of the fuel element was observed. The study of the next phase will be conducted later

  18. System for characterization of objects, particularly of nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    In order to be able to unmistakably identify fissionable materials of single fuel elements within the framework of monitoring, it is suggested to make impurity inclusions in defined regions consisting of materials which can be easily recognized by an ultrasonic sensor device. These 'sealing samples' enable perfect identification and are safe against falsing. (HP)

  19. R and D activities for ultimate disposal of HTR spent fuel elements

    International Nuclear Information System (INIS)

    For the direct ultimate disposal of HTR spent fuel, the LWR MAW storage method is examined, with a view to the following aspects: Activity released under rock pressure, and data of the design basis accidents: drop of waste package, underground fire, and inflow of lye in the post-operational phase of the repository. Strength tests with model packages containing HTR spent fuel in loose packing, or backfilled, have shown that backfilling with cement mortar or sand prevents damage to the spent fuel caused by rock pressure. Particle cracking in damaged, not backfilled fuel elements was at a maximum of 1 p.c. In leaching experiments with spent, complete fuel elements (900C - 2000C and 130/300 bar) only contaminations of the graphite matrix were released. Nuclide release after some 100 days of leaching time is determined by diffusion processes, but even after 1250 days of leaching, only trace amounts of Cs, Sr, Ce, Ba, Eu, Co, actinides or heavy metals could be found. High burn-up UO2 fuel kernels without coating and graphite matrix under the effects of leaching at 900C, 130 bar already after 60 days exhibited strong corrosion damage, and after about 100 days, cesium was released almost completely. (orig.)

  20. An Expert System to Analyze Homogeneity in Fuel Element Plates for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tolosa, S.C.; Marajofsky, A.

    2004-10-06

    In the manufacturing control of Fuel Element Plates for Research Reactors, one of the problems to be addressed is how to determine the U-density homogeneity in a fuel plate and how to obtain qualitative and quantitative information in order to establish acceptance or rejection criteria for such, as well as carrying out the quality follow-up. This paper is aimed at developing computing software which implements an Unsupervised Competitive Learning Neural Network for the acknowledgment of regions belonging to a digitalized gray scale image. This program is applied to x-ray images. These images are generated when the x-ray beams go through a fuel plate of approximately 60 cm x 8 cm x 0.1 cm thick. A Nuclear Fuel Element for Research Reactors usually consists of 18 to 22 of these plates, positioned in parallel, in an arrangement of 8 x 7 cm. Carrying out the inspection of the digitalized x-ray image, the neural network detects regions with different luminous densities corresponding to U-densities in the fuel plate. This is used in quality control to detect failures and verify acceptance criteria depending on the homogeneity of the plate. This modality of inspection is important as it allows the performance of non-destructive measurements and the automatic generation of the map of U-relative densities of the fuel plate.

  1. Fluidized-bed gasification under pressure of fuel element graphite in an industrial-scale reprocessing plant for HTR fuel elements

    International Nuclear Information System (INIS)

    In the head end of nuclear fuel reprocessing, the graphite cladding of spent HTR fuel elements is separated from the fissible material. Fluidized-bed combustion has some advantages over fixed-bed combustion. It is the method of choice in the design of a large-scale plant of 50.000 MWe HTR power. By means of an excess pressure of about 5 bar, a threefold increase in efficiency of a fluidized-bed ractor can be achieved. For an optimum layout of a prototype combustion plant, jacket cooling and internal heat exchangers are required. For an assessment of fluidized-bed combustion under pressure as a process step in the head end of a reprocessing plant, the author presents heat transfer calculations on the basis of a varying specific combustion load and investigations of the necessary peripheral equipment (reactor vessel, dust removal systems, gas supply and distribution, etc.) in several model set-ups. (RB)

  2. A review of trace element emissions from the combustion of refuse-derived fuel with coal

    International Nuclear Information System (INIS)

    The effects of cocombusting refuse-derived fuel (RDF) with coal on stack emissions of trace elements in the ash stream were reviewed. The large number of variables and uncertainties involved precluded drawing definitive conclusions regarding many of the trace elements. However, it is evident that cocombustion resulted in increased emissions of Cd, Cu, Hg, Pb, and Zn. Emissions of As and Ni tended to decrease when RDF was fired with coal. Modeling studies indicated that ambient levels of trace elements during cocombustion should be within acceptable limits. However, periodic monitoring of Cd, Hg, and Pb may be warranted in some instances

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

    International Nuclear Information System (INIS)

    The High-Uranium-Loaded U3O8-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 U3O8-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 U3O8-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 % U3O8). 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 U3O8 and aluminum

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

    Science.gov (United States)

    Govers, K.; Verwerft, M.

    2016-09-01

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

  5. TRISO-Fuel Element Performance Modeling for the Hybrid LIFE Engine with Pu Fuel Blanket

    Energy Technology Data Exchange (ETDEWEB)

    DeMange, P; Marian, J; Caro, M; Caro, A

    2010-02-18

    A TRISO-coated fuel thermo-mechanical performance study is performed for the hybrid LIFE engine to test the viability of TRISO particles to achieve ultra-high burnup of a weapons-grade Pu blanket. Our methodology includes full elastic anisotropy, time and temperature varying material properties for all TRISO layers, and a procedure to remap the elastic solutions in order to achieve fast fluences up to 30 x 10{sup 25} n {center_dot} m{sup -2} (E > 0.18 MeV). In order to model fast fluences in the range of {approx} 7 {approx} 30 x 10{sup 25} n {center_dot} m{sup -2}, for which no data exist, careful scalings and extrapolations of the known TRISO material properties are carried out under a number of potential scenarios. A number of findings can be extracted from our study. First, failure of the internal pyrolytic carbon (PyC) layer occurs within the first two months of operation. Then, the particles behave as BISO-coated particles, with the internal pressure being withstood directly by the SiC layer. Later, after 1.6 years, the remaining PyC crumbles due to void swelling and the fuel particle becomes a single-SiC-layer particle. Unrestrained by the PyC layers, and at the temperatures and fluences in the LIFE engine, the SiC layer maintains reasonably-low tensile stresses until the end-of-life. Second, the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Obtaining more reliable measurements, especially at higher fluences, is an imperative for the fidelity of our models. Finally, varying the geometry of the TRISO-coated fuel particles results in little differences in the scope of fuel performance. The mechanical integrity of 2-cm graphite pebbles that act as fuel matrix has also been studied and it is concluded that they can reliable serve the entire LIFE burnup cycle without failure.

  6. Experimental verification of structural models to analyze the nonlinear dynamics of LMFBR fuel elements

    International Nuclear Information System (INIS)

    Local fault situations in LMFBR cores may produce severe pressure pulses within one fuel element. The fact cannot be ignored that these pressures can have peaks and impulses that may expand and rupture the wrapper around the element. This will impulsively load the surrounding subassemblies and possibly the control rods due to extreme coolant pressure gradients and/or subassembly collision forces. Fast reactor safety requires this mechanical propagation process through the core to be analyzed, and therefore appropriate models and solution methods are needed to simulate the nonlinear structural dynamics of one typical hexagonal fuel element. The aim of this paper is to outline one- and two-dimensional structural models and discuss their capabilities and suitability for multirow core calculations. For this purpose static and impulsive single subassembly loading experiments are described and typical results are reported and compared with numerical predictions. (Auth.)

  7. Hazards review: N-Reactor 1.25% co-producer fuel element test

    Energy Technology Data Exchange (ETDEWEB)

    Miller, N.R.; Nechodom, W.S.

    1964-07-13

    The N-Reactor Hazard Summary Report examines the hazard from operating the N-Reactor with a uniform fuel loading enriched to 0.947% U{sup 235}. Incentives have been developed for reactor testing of a block of 49 tubes loaded with co-producer elements, i.e. elements capable of producing both weapons grade plutonium and tritium. The element utilizes an outer fuel tube enriched to 1.25% U{sup 235} with an inner target lithium-aluminum rod. Criteria have been developed to guide the evaluation of safety aspects of such tests. It is the purpose of this document to review the hazards associated with the proposed test and to set forth special precautions which will be necessary to maintain a high level of safety.

  8. Papers concerning fuel particles, fuel elements and graphitic materials for high temperature reactors presented at the 1980 annual conference Kerntechnik

    International Nuclear Information System (INIS)

    This report is a compilation of the papers presented by staff of the Institute for Reactor Materials, KFA-Juelich, at the 1980 annual conference Rekatortechnik, held in Berlin, 25-27th March 1980. In some cases, there were co-authors from other organisations. Where possible the manuscripts of the presentations have been reproduced, as well as the display cards shown during the poster session on the conference. In the presentations, the questions of the characterization of fuel particles and the retention of fission products are dealt with, and special attention is given to fission product release at very high temperatures. One presentation deals with the disposal of fuel elements from the AVR. Another report presents the results of radiation experiments on the standard matrix material A3-3 of the THTR fuel elements; the changes in dimensions, creep coefficient and thermal conductivity were measured as functions of the fluence and the radiation temperature. Interim results obtained from long-term radiation experiments on reflector graphites for high and very high flux reactors are presented, and models for the calculation of dimensional changes in components subjected to fluctuating temperatures from data obtained in isothermal tests are discussed. (orig.)

  9. 17506 - Order of 20 july 1989 on the storage period for fuel elements for spanish nuclear power plants

    International Nuclear Information System (INIS)

    The Order was made in furtherance of Decree No. 813/1988 amending a Decree of 1985 on the reorganisation of activities in the nuclear fuel cycle. It establishes new requirements regarding fuel elements for PWRs and BWRs, namely by providing that their operators should stock enough fuel elements for one load at least two months prior to the planned loading. Other plants should have the number of fuel elements necessary for their continuous operation for four months at 80 per cent of their nominal power

  10. Legal questions concerning the termination of spent fuel element reprocessing; Rechtsfragen der Beendigung der Wiederaufarbeitung abgebrannter Brennelemente

    Energy Technology Data Exchange (ETDEWEB)

    John, Michele

    2005-07-01

    The thesis on legal aspects of the terminated spent fuel reprocessing in Germany is based on the legislation, jurisdiction and literature until January 2004. The five chapters cover the following topics: description of the problem; reprocessing of spent fuel elements in foreign countries - practical and legal aspects; operators' responsibilities according to the atomic law with respect to the reprocessing of Geman spent fuel elements in foreign countries; compatibility of the prohibition of Geman spent fuel element reprocessing in foreign countries with international law, European law and German constitutional law; results of the evaluation.

  11. Development of TUF-ELOCA - a software tool for integrated single-channel thermal-hydraulic and fuel element analyses

    International Nuclear Information System (INIS)

    The TUF-ELOCA tool couples the TUF and ELOCA codes to enable an integrated thermal-hydraulic and fuel element analysis for a single channel during transient conditions. The coupled architecture is based on TUF as the parent process controlling multiple ELOCA executions that simulate the fuel elements behaviour and is scalable to different fuel channel designs. The coupling ensures a proper feedback between the coolant conditions and fuel elements response, eliminates model duplications, and constitutes an improvement from the prediction accuracy point of view. The communication interfaces are based on PVM and allow parallelization of the fuel element simulations. Developmental testing results are presented showing realistic predictions for the fuel channel behaviour during a transient. (author)

  12. Flow of viscoplastic fluids in eccentric annular geometries

    DEFF Research Database (Denmark)

    Szabo, Peter; Hassager, Ole

    1992-01-01

    A classification of flowfields for the flow of a Bingham fluid in general eccentric annular geometries is presented. Simple arguments show that a singularity can exist in the stress gradient on boundaries between zones with yielded and un-yielded fluid respectively. A Finite Element code is used ...

  13. Non-Destructive Testing of Reactor-Fuel, Target, and Control Elements

    International Nuclear Information System (INIS)

    At the Savannah River Plant (a production facility of the United States Atomic Energy Commission), fuel, target, and control elements are non-destructively tested before and after irradiation in reactors. Design and performance of unique instruments - used for measuring physical soundness, nuclear properties, and dimensions - are described. A nickel thickness gauge, utilizing the Hall effect in a magnetic field, is used to measure the thickness of nickel layers on uncanned uranium cores. A similar instrument is used after cores have been diffusion bonded to aluminium cladding to determine that each core has a layer of residual nickel, and that the end cap is sufficiently thick. Ultrasonic instruments are used (1) to measure uranium grain size to determine whether the cores were properly heat-treated, and (2) to detect unbonded areas between cladding and core. The Nuclear Test Gauge (NTG), a small subcritical assembly of U235-A1 alloy slugs in an H2O-moderated lattice, is used to determine the fuel (U235) or absorber (Li6) content of reactor elements. These determinations, made from changes in neutron multiplication, have a 1-sigma precision of about ± 0.5% for fuel elements containing up to 250 g U235/ 30.5 cm (1 ft), and about ± 1% for target and control elements containing up to 4 g Li6/30.5 cm (1 ft). Compared to the more commonly used large.critical test pile, the NTG costs about 1/20 as much; measures fuel or absorber content in about one minute vs. ten minutes; and measures the axial distribution of fuel or absorber which the test pile cannot do. Irradiated fuel elements are measured under water with (1) differential transformers that can measure diameter and length to an accuracy of ± 0.05 cm (0.002 in), and (2) simple mechanical linkages with dial indicators above water that can measure inside diameter and warp. By keeping the elements submerged in water, personnel are shielded from radiation, and the elements do not undergo the dimensional changes that

  14. Romanian irradiation experiment on AHWR type fuel elements containing mixed oxide of thorium and uranium pellets

    International Nuclear Information System (INIS)

    One of the main objectives of the Institute for Nuclear Research (ICN) - Nuclear Fuel R and D Program is the development of new types of fuel based on: Slightly Enriched Uranium (SEU), Recycled Uranium (RU) and Thorium. Two experimental fuel elements (A23 and A24) were irradiated in TRIGA Research Reactor of ICN-Romania (C1 device) in a power ramp conditions. Element A23, contained mixed oxide of thorium and uranium pellets, achieved a maximum linear power of 51 KW/m and has reached a discharge burn-up around 189.2 MWh/KgHE; element A24, contained dioxide of uranium pellets, achieved a maximum linear power of 63 kW/m and has reached a discharge burn-up of around 207.8 MWh/KgHE. The experiments simulation has been performed using an improved version of ROFEM Code, version developed through the efforts of researchers from ICN - Nuclear Fuel Performance Department. The simulation results are in good agreement with experimental data. (author)

  15. Performance evaluation of two CANDU fuel elements tested in the TRIGA reactor

    International Nuclear Information System (INIS)

    Nuclear Research Institute at Pitesti has a set of facilities, which allow the testing, manipulation and examination of nuclear fuel and structure materials irradiated in CANDU reactors from Cernavoda NPP. These facilities consist of TRIGA materials testing reactor and Post-Irradiation Examination Laboratory (LEPI). The purpose of this work is to describe the post-irradiation examination, of two experimental CANDU fuel elements (EC1 and EC2). The fuel elements were mounted into a pattern port, one in extension of the other in a measuring test for the central temperature evolution. The results of post-irradiation examination are obtained from: Visual inspection and photography of the outer appearance of sheath; Profilometry (diameter, bending, ovalization) and length measuring; Determination of axial and radial distribution of the fission products activity by gamma scanning; Measurement of pressure, volume and isotopic composition of fission gas; Microstructural characterization by metallographic and ceramographic analyzes; Isotopic composition and burn-up determination. The post-irradiation examination results are used, on one hand, to confirm the security, reliability and performance of the irradiated fuel, and on the other hand, for further development of CANDU fuel. (authors)

  16. The isolation and analysis of critical stress areas in graphite fuel elements

    International Nuclear Information System (INIS)

    This paper presents a summary of methods being used at General Atomic for the analysis of irradiation-induced stresses in the graphite fuel elements of a high-temperature gas-cooled reactor (HTGR). The main topic of the paper is a systematic strategy for the isolation and detailed analysis of critical stress areas. The first program in the hierarchy, named SURVEY/STRESS, uses a simplified beam theory model to sort and identify potentially critical fuel elements among the thosands in the core. SURVEY/STRESS employs an idealized geometry consisting of an array of parallel fibers, each one of which is an accurate representation of the symmetrical web between the fuel and coolant holes. The program automatically superimposes the localized web stress field on the average stress fiel of the block. A more refined sorting is done with the aid of a general purpose beam theory code named GBEAM. GBEAM employs an array of triangular elements to accurately model the geometry. The calculated axial stresses are very close to those of a two-dimensional analysis, but they are obtained at a considerable cost saving. The third program in the series is SAFIRE, a two-dimensional, viscoelastic, finite element program. SAFIRE employs a generalized plane strain (GPS) element in which three global bending degrees of freedom are introduced at the element level. The GPS element correctly considers the stiffness and Poisson coupling effects of out-of-plane deformations. This is a significant improvement over earlier programs that superimposed reversed reaction forces on a plane str

  17. Adiabatic Steam-Water Annular Flow in an Annular Geometry

    DEFF Research Database (Denmark)

    Andersen, P. S.; Würtz, J.

    1981-01-01

    Experimental results for fully developed steam-water annular flow in annular geometries are presented. Rod and tube film flow rates and axial pressure gradients were measured for mass fluxes between 500 and 2000 kg/m2s, steam qualities between 20 and 60 per cent and pressures ranging from 3 to 9...... MPa. It was found that the measured tube film flow rate per unit tube perimeter is always many times greater than the corresponding rod film flow rate. Possible explanations for this asymmetry are discussed....

  18. Behavior of CANDU fuel under power pulse conditions at the TRIGA reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensors for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280 cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (orig.)

  19. Conceptual design of a multicell thermionic fuel element for a 40-KWE space nuclear power system

    International Nuclear Information System (INIS)

    This paper addresses the conceptual design of a Russian thermionic fuel element (TFE) in support of the S-PRIME 40 kWe in-core thermionic space power reactor studies sponsored by the U.S. Department of Energy TI-SNPS program. The design responds to requirements specified by the U.S. component of the S-PRIME team and is based on the multicell ''flashlight'' TFE approach. Following a general description of the TFE design, the considerations leading to several key design decisions are discussed. These include nuclear and thermionic performance, cesium management, fission product management, materials selection, fuel system, and lifetime. copyright American Institute of Physics 1995

  20. ENEA TRIGA RC-1 reactor spent fuel elements shipment to the USA

    International Nuclear Information System (INIS)

    TRIGA Mark II reactor of ENEA's Casaccia research Center (in Italy named RC-1) reached first criticality in 1960. In more than thirty years of operation, 1 MW reactor core has been modified many times for fuel elements burn-up optimization. Till now, because of achieved maximum burn-up, 146 fuel elements have been definitively removed from reactor core and transferred to the hot storages in reactor pool (5 racks around reactor vessel) and in the reactor room (pits). The activities planning, the organizing aspect study, the analysis and valuations both nuclear safety and radioprotection have been suitable for the TRIGA RC-1 fuel element shipment. Infact, no operative anomaly is appeared respect the approved procedures. Personnel engagement has been as expectations and the personnel absorbed gamma dose resulted negligible. Finally, the NAC disposable narrow time (only one week at the end of July) has not produced heavy organization problems but it has been a strong goad per all operative structures involved in the TRIGA RC-1 elements shipment

  1. A two-dimensional finite element method for analysis of solid body contact problems in fuel rod mechanics

    International Nuclear Information System (INIS)

    Two computer codes for the analysis of fuel rod behavior have been developed. Fuel rod mechanics is treated by a two-dimensional, axisymmetric finite element method. The program KONTAKT is used for detailed examinations on fuel rod sections, whereas the second program METHOD2D allows instationary calculations of whole fuel rods. The mechanical contact of fuel and cladding during heating of the fuel rod is very important for it's integrity. Both computer codes use a Newton-Raphson iteration for the solution of the nonlinear solid body contact problem. A constitutive equation is applied for the dependency of contact pressure on normal approach of the surfaces which are assumed to be rough. If friction is present on the contacting surfaces, Coulomb's friction law is used. Code validation is done by comparison with known analytical solutions for special problems. Results of the contact algorithm for an elastic ball pressing against a rigid surface are confronted with Hertzian theory. Influences of fuel-pellet geometry as well as influences of discretisation of displacements and stresses of a single fuel pellet are studied. Contact of fuel and cladding is calculated for a fuel rod section with two fuel pellets. The influence of friction forces between fuel and cladding on their axial expansion is demonstrated. By calculation of deformations and temperatures during an instationary fuel rod experiment of the CABRI-series the feasibility of two-dimensional finite element analysis of whole fuel rods is shown. (orig.)

  2. Annular lipoatrophy of the ankles.

    Science.gov (United States)

    Dimson, Otobia G; Esterly, Nancy B

    2006-02-01

    Lipoatrophic panniculitis likely represents a group of disorders characterized by an inflammatory panniculitis followed by lipoatrophy. It occurs locally in a variety of settings and has been reported in the literature under various terms, including annular atrophic connective tissue panniculitis of the ankles, annular and semicircular lipoatrophy, abdominal lipoatrophy, and connective tissue panniculitis. Herein, a case of annular lipoatrophy of the ankles is described in a 6-year-old girl with autoimmune thyroid disease. Histologically, a mixed lobular panniculitis with lipophages was present. This pattern resembles that seen in lipoatrophic panniculitis. After a single, acute episode of an inflammatory process with subsequent lipoatrophy, her skin lesions have stabilized for 2 years requiring no treatment.

  3. Fusion option to dispose of spent nuclear fuel and transuranic elements

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.

    2000-02-10

    The fusion option is examined to solve the disposition problems of the spent nuclear fuel and the transuranic elements. The analysis of this report shows that the top rated solution, the elimination of the transuranic elements and the long-lived fission products, can be achieved in a fusion reactor. A 167 MW of fusion power from a D-T plasma for sixty years with an availability factor of 0.75 can transmute all the transuranic elements and the long-lived fission products of the 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. The operating time can be reduced to thirty years with use of 334 MW of fusion power, a system study is needed to define the optimum time. In addition, the fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future. Fusion blankets with a liquid carrier for the transuranic elements can achieve a transmutation rate for the transuranic elements up to 80 kg/MW.y of fusion power with k{sub eff} of 0.98. In addition, the liquid blankets have several advantages relative to the other blanket options. The energy from this transmutation is utilized to produce revenue for the system. Molten salt (Flibe) and lithium-lead eutectic are identified as the most promising liquids for this application, both materials are under development for future fusion blanket concepts. The Flibe molten salt with transuranic elements was developed and used successfully as nuclear fuel for the molten salt breeder reactor in the 1960's.

  4. Fuel composition optimization in a 78-element fuel bundle for use in a pressure tube type supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    A 78-element fuel bundle containing a plutonium-thorium fuel mixture has been proposed for a Generation IV pressure tube type supercritical water-cooled reactor. In this work, using a lattice cell model created with the code DRAGON,the lattice pitch, fuel composition (fraction of PuO2 in ThO2) and radial enrichment profile of the 78-element bundle is optimized using a merit function and a metaheuristic search algorithm.The merit function is designed such that the optimal fuel maximizes fuel utilization while minimizing peak element ratings and coolant void reactivity. A radial enrichment profile of 10 wt%, 11 wt% and 20 wt% PuO2 (inner to outer ring) with a lattice pitch of 25.0 cm was found to provide the optimal merit score based on the aforementioned criteria. (author)

  5. Fabrication of simulated plate fuel elements: Defining role of stress relief annealing

    Science.gov (United States)

    Kohli, D.; Rakesh, R.; Sinha, V. P.; Prasad, G. J.; Samajdar, I.

    2014-04-01

    This study involved fabrication of simulated plate fuel elements. Uranium silicide of actual fuel elements was replaced with yttria. The fabrication stages were otherwise identical. The final cold rolled and/or straightened plates, without stress relief, showed an inverse relationship between bond strength and out of plane residual shear stress (τ13). Stress relief of τ13 was conducted over a range of temperatures/times (200-500 °C and 15-240 min) and led to corresponding improvements in bond strength. Fastest τ13 relief was obtained through 300 °C annealing. Elimination of microscopic shear bands, through recovery and partial recrystallization, was clearly the most effective mechanism of relieving τ13.

  6. Finite element simulation of fission gas release and swelling in UO2 fuel pellets

    International Nuclear Information System (INIS)

    A fission gas release model is presented, which solves the atomic diffusion problem with xenon and krypton elements tramps produced by uranium fission during UO2 nuclear fuel irradiation. The model considers intra and intergranular precipitation bubbles, its re dissolution owing to highly energetic fission products impact, interconnection of intergranular bubbles and gas sweeping by grain border in movement because of grain growth. In the model, the existence of a thermal gradient in the fuel pellet is considered, as well as temporal variations of fission rate owing to changes in the operation lineal power. The diffusion equation is solved by the finite element method and results of gas release and swelling calculation owing to gas fission are compared with experimental data. (author)

  7. Conceptual Engineering of CARA Fuel Element with Negative Void Coefficient for Atucha II

    Directory of Open Access Journals (Sweden)

    H. Lestani

    2011-01-01

    Full Text Available Experimentally validated void reactivity calculations were used to study the feasibility of a change in the design basis of Atucha II Nuclear Power Plant including the Large LOCA event. The use of CARA fuel element with burnable neutronic absorbers and enriched uranium is proposed instead of the original fuel. The void reactivity, refuelling costs, and power peaking factors are analysed at conceptual level to optimize the burnable neutronic absorber, the enrichment grade, and their distribution inside the fuel. This work concludes that, for the considered plant conditions, either a void reactivity coefficient granting no prompt critical excursion on Large LOCA or negative void reactivity is achievable, with advantages on refuelling cost and linear power density.

  8. Fabrication of simulated plate fuel elements: Defining role of out-of-plane residual shear stress

    Energy Technology Data Exchange (ETDEWEB)

    Rakesh, R., E-mail: rakesh.rad87@gmail.com [DAE Graduate Fellows, IIT Bombay, Powai, Mumbai 400076 (India); Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Kohli, D. [DAE Graduate Fellows, IIT Bombay, Powai, Mumbai 400076 (India); Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Sinha, V.P.; Prasad, G.J. [Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Samajdar, I. [Department of Metallurgical Engineering and Materials Science, IIT Bombay, Powai, Mumbai 400076 (India)

    2014-02-01

    Bond strength and microstructural developments were investigated during fabrication of simulated plate fuel elements. The study involved roll bonding of aluminum–aluminum (case A) and aluminum–aluminum + yttria (Y{sub 2}O{sub 3}) dispersion (case B). Case B approximated aluminum–uranium silicide (U{sub 3}Si{sub 2}) ‘fuel-meat’ in an actual plate fuel. Samples after different stages of fabrication, hot and cold rolling, were investigated through peel and pull tests, micro-hardness, residual stresses, electron and micro-focus X-ray diffraction. Measurements revealed a clear drop in bond strength during cold rolling: an observation unique to case B. This was related to significant increase in ‘out-of-plane’ residual shear stresses near the clad/dispersion interface, and not from visible signatures of microstructural heterogeneities.

  9. Fabrication of simulated plate fuel elements: Defining role of out-of-plane residual shear stress

    Science.gov (United States)

    Rakesh, R.; Kohli, D.; Sinha, V. P.; Prasad, G. J.; Samajdar, I.

    2014-02-01

    Bond strength and microstructural developments were investigated during fabrication of simulated plate fuel elements. The study involved roll bonding of aluminum-aluminum (case A) and aluminum-aluminum + yttria (Y2O3) dispersion (case B). Case B approximated aluminum-uranium silicide (U3Si2) 'fuel-meat' in an actual plate fuel. Samples after different stages of fabrication, hot and cold rolling, were investigated through peel and pull tests, micro-hardness, residual stresses, electron and micro-focus X-ray diffraction. Measurements revealed a clear drop in bond strength during cold rolling: an observation unique to case B. This was related to significant increase in 'out-of-plane' residual shear stresses near the clad/dispersion interface, and not from visible signatures of microstructural heterogeneities.

  10. A study on the laser welding of aluminum for end-capping of KMRR fuel elements

    International Nuclear Information System (INIS)

    The selecting of proper welding process for aluminum end-capping of KMRR nuclear fuel elements is considered important in respect to the soundness of weldments and the improvement of the performance of nuclear fuels during the operation in reactor. The probability of leakage of the fission products is mostly apt to occur at the weldments, and it is connected directly with the safety and life prediction of the nuclear reactor in operation. This study is to investigate the influence of Ti and Zr micro-additions on the grain refinement in Al weld metal and to prevent weld cracks in pure Al. In addition, mechanical properties of laser welded specimen were observed and the technical data for KMRR nuclear fuels manufacturing was provided. In the future, this study will also present the improvement of KMRR nuclear end-caps welding. (Author)

  11. Analysis of gamma ray spectrum of Al-Pu plate fuel elements

    International Nuclear Information System (INIS)

    Gamma ray spectrometry of aluminium (0.5mm thick) clad Al-Pu plate fuel elements containing ∼10 g plutonium per plate, was carried out using a high purity germanium detector (HPGe) and a 4K multichannel analyser. The detector to fuel plate distance was kept 750 mm to have a full view of the plate and to reduce dead time corrections. The data was analysed using peak fitting program SAMPO. Typical spectral shape obtained in this study is presented. Such spectral calibration avoids errors based on extrapolation from point source geometries and could be useful in evaluating (i) plutonium contents in similar plate fuels and in turn checking the final plutonium inventory and (ii)radiation fields at different working distances. (author). 3 refs., 1 fig., 1 tab

  12. Optimization of thorium-uranium content in a 54-element fuel bundle for use in a CANDU-SCWR

    International Nuclear Information System (INIS)

    A new 54-element fuel bundle design has been proposed for use in a pressure-tube supercritical water-cooled reactor, a pre-conceptual evolution of existing CANDU reactors. Pursuant to the goals of the Generation IV International Forum regarding advancement in nuclear fuel cycles, optimization of the thorium and uranium content in each ring of fuel elements has been studied with the objectives of maximizing the achievable fuel utilization (burnup) and total thorium content within the bundle, while simultaneously minimizing the linear element ratings and coolant void reactivity. The bundle was modeled within a reactor lattice cell using WIMS-AECL, and the uranium and thorium content in each ring of fuel elements was optimized using a weighted merit function of the aforementioned criteria and a metaheuristic search algorithm. (author)

  13. Modelling of pressurized water reactor fuel, rod time dependent radial heat flow with boundary element method

    International Nuclear Information System (INIS)

    The basic principles of the boundary element method numerical treatment of the radial flow heat diffusion equation are presented. The algorithm copes the time dependent Dirichlet and Neumann boundary conditions, temperature dependent material properties and regions from different materials in thermal contact. It is verified on the several analytically obtained test cases. The developed method is used for the modelling of unsteady radial heat flow in pressurized water reactor fuel rod. (author)

  14. Heat transfer burnout in tube-type fuel elements of nuclear power reactors

    International Nuclear Information System (INIS)

    The conditions are formulated under which the results of the experimental research of the boilino. water heat transfer burnout carried out on models may be applied to fuel elements of nuclear reactors. Experimental material providing data on the heat transfer burnout was expanded by the results of measurements of the uneven (cosine) longitudinal distribution of heat sources. The results of the effects of helical fins or wires on heat transfer burnout are presented. (F.M.)

  15. Cna 1 spent fuel element interim dry storage system thermal analysis

    International Nuclear Information System (INIS)

    At the moment, the Atucha I Nuclear Power Plant (Cnea-I) located in the city of Lima, has enough room to store its spent fuel (Sf) in their two pools spent fuel until about 2015.In case of life extension a spend fuel element interim dry storage system is needed.Nucleolectrica Argentina S.A. (N A-S A) and the Comision Nacional de Energia Atomica (Cnea), have proposed different interim dry storage systems.These systems have to be evaluated in order to choose one of them.The present work's objective is the thermal analysis of one dry storage alternative for the Sf element of Cna 1.In this work a simple model was developed and used to perform the thermal calculations corresponding to the system proposed by Cnea.This system considers the store of sealed containers with 37 spent fuels in concrete modules.Each one of the containers is filled in the pool houses and transported to the module in a transference cask with lead walls.Fulfill the maximum cladding temperature requirement (<200 degrade C) for a total 2140 W thermal power

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

    Science.gov (United States)

    Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

    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.

  17. Thermal energy management and chemical reaction investigation of micro-proton exchange membrane fuel cell and fuel cell system using finite element modelling

    OpenAIRE

    McGee, Seán

    2015-01-01

    Fuel cell systems are becoming more commonplace as a power generation method and are being researched, developed, and explored for commercial use, including portable fuel cells that appear in laptops, phones, and of course, chargers. This thesis examines a model constructed on inspiration from the myFC PowerTrekk, a portable fuel cell charger, using COMSOL Multiphysics, a finite element analysis software. As an educational tool and in the form of zero-dimensional, two-dimensional, and three-d...

  18. A study of parameters on marking of Prototype Fast Breeder Reactor fuel elements

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor Fuel (PFBR) elements are identified with a permanent unique marking. Identification of the fuel elements is very much necessary for traceability during initial fabrication as well as for post irradiation examination. Marking on fuel element has to be permanent and capable of being identified after irradiation. Laser marking is a relatively new method as compared to other marking technologies such as ink marking, mechanical engraving and electro chemical methods. It is used for the product identification and traceability during its service life. Laser marking has many advantages compared to other conventional marking. In laser marking process, mark quality is a very important factor, which depends on so many variables like input current, pulse frequency, marking speed and number of passes. The influence of the pulse frequency and the speed of travel of the laser beam on the mark depth and width have been studied in this paper. An optical microscope, scanning electron microscope were used to measure the effects of pulse frequency on the mark depth and width. It has been found that the mark depth and width depend on the interaction process of the laser beam and the material, which was influenced by the pulse frequency. Micro hardness testing is carried out to report Heat Affected Zone (HAZ) variation with parameters. Marking speed and input current selected for suitable depth and width were mentioned in the present study. (author)

  19. Phase flow rate measurements of annular flows

    OpenAIRE

    Al-Yarubi, Qahtan

    2010-01-01

    In the international oil and gas industry multiphase annular flow in pipelines and wells is extremely important, but not well understood. This thesis reports the development of an efficient and cheap method for measuring the phase flow rates in two phase annular and annular mist flow, in which the liquid phase is electrically conducting, using ultrasonic and conductance techniques. The method measures changes in the conductance of the liquid film formed during annular flow and uses these to c...

  20. The FPIN2 code - an application of the finite element method to the analysis of the transient response of oxide and metal fuel elements

    International Nuclear Information System (INIS)

    The FPIN2 code simulates the thermal-mechanical response of fast reactor fuel pins to transient events. Temperatures of the fuel pin and coolant are calculated using a simple pin-in-a-pipe geometry. The mechanical analysis uses an implicit finite element formulation with linear shape functions which allow for general material behavior in the fuel and cladding including cracking and melting. This formulation provides a very convenient structure for implementing different models and improvements in algorithms. The paper summarizes the FPIN2 methodology and presents results for the transient response of both oxide and metallic fuel pins under similar overpower transients. (author)

  1. Eosinophilic annular erythema in childhood - Case report.

    Science.gov (United States)

    Abarzúa, Alvaro; Giesen, Laura; Silva, Sergio; González, Sergio

    2016-01-01

    Eosinophilic annular erythema is a rare, benign, recurrent disease, clinically characterized by persistent, annular, erythematous lesions, revealing histopathologically perivascular infiltrates with abundant eosinophils. This report describes an unusual case of eosinophilic annular erythema in a 3-year-old female, requiring sustained doses of hydroxychloroquine to be adequately controlled. PMID:27579748

  2. Eosinophilic annular erythema in childhood - Case report*

    Science.gov (United States)

    Abarzúa, Alvaro; Giesen, Laura; Silva, Sergio; González, Sergio

    2016-01-01

    Eosinophilic annular erythema is a rare, benign, recurrent disease, clinically characterized by persistent, annular, erythematous lesions, revealing histopathologically perivascular infiltrates with abundant eosinophils. This report describes an unusual case of eosinophilic annular erythema in a 3-year-old female, requiring sustained doses of hydroxychloroquine to be adequately controlled. PMID:27579748

  3. 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.

  4. CFD analysis of the 37-element fuel channel for CANDU6 reactor

    International Nuclear Information System (INIS)

    We analyzed the thermal-hydraulic behavior of coolant flow along fuel bundles with appendages of end support plate, spacer pad, and bearing pad, which are the CANDU6 characteristic design. The computer code used is a commercial CFD code, CFX-12. The present CFD analysis model calculates the conjugate heat transfer between the fuel and coolant. Using the same volumetric heat source as the O6 channel, the CFD predictions of the axial temperature distributions of the fuel element are compared with those by the CATHENA (one-dimensional safety analysis code for CANDU6 reactor). It is shown that CFX-12 predictions are in good agreement with those by the CATHENA code for the single liquid convection region (especially before the axial position of the first half of the channel length). However, the CFD analysis at the second half of the fuel channel, where the two-phase flow is expected to occur, over-predicts the fuel temperature, since the wall boiling model is not considered in the present CFD model. (author)

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

    International Nuclear Information System (INIS)

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

  6. Effect of hydrogenation on the anisotropy of mechanical properties of the fuel elements pipes of an alloy Zr-1%Nb

    International Nuclear Information System (INIS)

    Mechanical characteristics of the fuel tube the alloy Zr-1%Nb Ukrainian production tensile in radial (annular samples) and longitudinal direction (''quasi-plane'' samples) in the delivery state and after hydrogenation were defined. It is shown that the fuel tubes in the initial state possess significant anisotropy of mechanical properties: significant difference is observed for the proof stress, which in a tangential direction (with respect to the tube axis) is 20% higher than in the longitudinal direction. Hydrogenation of the fuel tubes increases anisotropy of mechanical properties: yield strength in the tangential direction by 20...40% higher than in the longitudinal direction. Tensile strength of the samples which have been deformed in the tangential direction exceeds the tensile strength in the longitudinal direction by 25...50%. The process of hydrogenation virtually has no effect on the plasticity of the ring samples, but it leads to embrittlement of quasi-plane samples. The cause of the anisotropy of the mechanical properties of the fuel tubes in initial state is the texture which is formed in the material in resulting in mechanical and heat treatments. Features of change of the strength and plastic properties of the hydrogenated material is associated with the formation of oriented zirconium hydride precipitates, which have different effects on the deformation of the tubes in the longitudinal and transverse directions

  7. Benign concentric annular macular dystrophy

    Directory of Open Access Journals (Sweden)

    Luísa Salles de Moura Mendonça

    2015-06-01

    Full Text Available The purpose of the authors is to show clinical findings of a patient with benign concentric annular macular dystrophy, which is an unusual condition, and part of the "bull’s eye" maculopathy differential diagnosis. An ophthalmologic examination with color perception, fluorescein angiography, and ocular electrophysiology was performed.

  8. The Application of Various Nondestructive Testing Methods to Fuel Elements of the Orgel Type

    International Nuclear Information System (INIS)

    The paper describes the various methods employed to detect flaws (dimensional or structural) in fuel-element canning tubes. The authors also describe the final tests on complete fuel elements, in particular radiography of welds and leak-tightness tests. This subject has already been discussed to some extent. The dimensional characteristics of smooth SAP (sintered aluminium powder) canning tubes have been fairly extensively investigated, and in particular: 1. The internal and external diameters have been measured using pneumatic pick-ups and recording the result; 2. The thicknesses have been measured using either ultrasonic resonance methods or y-rays (a Euratom- Istituto Sperimentale Metalli Leggeri contract); 3. Checking the deflection; 4. Tests of finned tubes. Work has also been carried out on detecting flaws in smooth canning tubes, and rejection criteria have been adopted depending on the prospective use of the tubes. (a) The making of artificial flaws corresponding to the harmfulness of actual flaws in the SAP is described. This study revealed high sensitivity to flaws of the longitudinal type generally caused by large inclusions during processing. (b) Ultrasonic tests. Longitudinal flaws: Comparison between the method with two pick-ups and that with one shows the limitations of these two methods. Transverse flaws: The single pick-up method used in investigating these is briefly described. Mechanical drive: A laboratory type mechanical test bench for investigating test criteria and a special semi-industrial bench for the continuous inspection of the tubes and the recording of flaws are mentioned. The difficulties encountered and the steps taken to prevent them are described. (c) Radiographic tests. This method will be discussed in a special paper; here we simply indicate the results obtained on pressure tubes and canning tubes. (d) Various tests. The final tests on complete fuel elements can be summed up in two sections: Helium leak-tests developed by SOGEV

  9. Subcutaneous granuloma annulare: radiologic appearance

    International Nuclear Information System (INIS)

    Objective. Granuloma annulare is an uncommon benign inflammatory dermatosis characterized by the formation of dermal papules with a tendency to form rings. There are several clinically distinct forms. The subcutaneous form is the most frequently encountered by radiologists, with the lesion presenting as a superficial mass. There are only a few scattered reports of the imaging appearance of this entity in the literature. We report the radiologic appearance of five cases of subcutaneous granuloma annulare. Design and patients. The radiologic images of five patients (three male, two female) with subcutaneous granuloma annulare were retrospectively studied. Mean patient age was 6.4 years (range, 2-13 years). The lesions occurred in the lower leg (two), foot, forearm, and hand. MR images were available for all lesions, gadolinium-enhanced imaging in three cases, radiographs in four, and bone scintigraphy in one. Results. Radiographs showed unmineralized nodular masses localized to the subcutaneous adipose tissue. The size range, in greatest dimension on imaging studies, was 1-4 cm. MR images show a mass with relatively decreased signal intensity on all pulse sequences, with variable but generally relatively well defined margins. There was extensive diffuse enhancement following gadolinium administration. Conclusion. The radiologic appearance of subcutaneous granuloma annulare is characteristic, typically demonstrating a nodular soft-tissue mass involving the subcutaneous adipose tissue. MR images show a mass with relatively decreased signal intensity on all pulse sequences and variable but generally well defined margins. There is extensive diffuse enhancement following gadolinium administration. Radiographs show a soft-tissue mass or soft-tissue swelling without evidence of bone involvement or mineralization. This radiologic appearance in a young individual is highly suggestive of subcutaneous granuloma annulare. (orig.)

  10. Large Eddy Simulation of thermoacoustic instabilities in annular combustion chambers

    OpenAIRE

    Wolf, Pierre

    2011-01-01

    Increasingly stringent regulations and the need to tackle rising fuel prices have placed great emphasis on the design of aeronautical gas turbines. This drive towards innovation has resulted sometimes in new concepts being prone to combustion instabilities. Combustion instabilities arise from the coupling of acoustics and combustion. In the particular field of annular combustion chambers, these instabilities often take the form of azimuthal modes. To predict these modes, one must consider the...

  11. Stainless steel sheathed UO2 fuel elements irradiated in the X-7 organic loop

    International Nuclear Information System (INIS)

    The X-7 organic loop of the NRX reactor was successfully commissioned by irradiating nine stainless steel sheathed UO2 fuel elements to a maximum burnup of about 1450 MWd/tonne U. The loop conditions and coolant chemistry for the tests are reported, The sheath thermocouples all failed early in the Commissioning test irradiation, while the bulk coolant thermocouples with enclosed junctions operated satisfactorily. All elements were covered with an organic fouling film whose thickness and appearance seemed to vary with sheath temperature. Dimensions of the elements had not changed from pre-irradiation values, The behavior of the UO2 was consistent with other Chalk River tests at similar heat ratings. (author)

  12. Japanese study on water reactor fuel element materials and method of measurement

    International Nuclear Information System (INIS)

    So many studies have been carried out in Japan on water reactor fuel element materials and the method of measuring their properties. Some topics will be high-lighted in this report to give an idea of what they have been doing in Japan. Studies on the properties of zircaloy, including the development work for modified alloy have been performed since the late 1950s, both in fundamental work at universities and research organizations and development work for zircaloy tube commercial production in metal industry. Among them, the latest work on the creep characteristics of zircaloy tubing is presented. For other material used in fuel element than cladding tube, spacer-spring will be discussed as the second topic. Reduction of spring force was carried out in Japan PWR fuel spacer to reduce rod bow in the early 1970s which could successfully eliminate the problem since 1977. Relaxation of spring force against burn-up has been discussed in the so-called reliability test program of both BWR and PWR on then-standard fuel since 1975 which was reported at Stockholm symposium in September 1986. The data obtained will be presented. As the study for the method of measurement, the author proposed a modified testing procedure for tensile and burst test for zircaloy tubing, at ASTM-B10 Committee in 1976, to emphasize the shape of mandrel in the tensile test which has a significant effect on elongation values. Various measurement techniques in post irradiation examination of water reactor fuel have been developed in Japan, among which the shadow measurement of tube during ballooning to burst will be described. (author)

  13. Current status of research and development of nuclear fuel elements for PWR in Indonesia

    International Nuclear Information System (INIS)

    Full text: The energy need of Indonesia is increasing due to the population growth and for the economic progress. The government of Indonesia intends to apply an optimum energy mix comprising all viable prospective energy sources. The Government Regulation No. 5 year 2006 indicates the target of energy mix until 2025 and the share of nuclear energy is about 2% of primary energy or 4% of electricity (4000 MWe). The first two units of NPP is expected to be operated before 2020 as stated in Act No. 17 year 2007 on National Long Term Development Planning 2005-2025. The first NPP to be operated in Indonesia is PWR type with capacity of 1000 MWe/unit. One of the strategies to strength and increase national capacity in the program for NPP introduction is domestication industry for nuclear fuel. To reach this purpose, the activities of research and development is focus on nuclear fuel production technology for PWR. Currently research and development activity in Indonesia is to produce prototype of nuclear fuel element for PWR in the form of test fuel pin or mini pin. In this paper we will presenting the pelletization and fabrication technology development. The existing facility was designed for PHWR fuel element of CIRENE type. Development of pelletization technology is carried out by modifying the compacting machine. Parameters of compacting and sintering are determined based on both compressibility and compactibility of the pellet as indicated by density and mechanical strength of the UO2 green pellets. The sintering parameters to be determined are temperature, heating rate, and soaking time. Currently, the fabrication process is under experiment. All of the data resulted from the experiment that will be presented in this meeting. (author)

  14. A magnetorheological valve with both annular and radial fluid flow resistance gaps

    International Nuclear Information System (INIS)

    In order to increase the efficiency of magnetorheological (MR) valves, Ai et al (2006) proposed an MR valve simultaneously possessing annular and radial fluid flow resistance channels with the assumption that the magnetic flux densities at the annular and radial fluid flow gaps are identical. In this paper, an MR valve simultaneously possessing annular and radial fluid flow resistance channels is designed, fabricated, modeled and tested. A model for the developed MR valve is produced and its performances are theoretically predicted based on the average magnetic flux densities in the annular and radial fluid flow gaps through finite element analysis. The theoretical results for the developed MR valve are compared with the experimental results. In addition, the performances of the developed MR valve are theoretically and experimentally compared with those of the MR valve with only annular fluid flow gaps. It has been shown that the theoretical results match well with the experimental results. Mainly attributed to the radial fluid flow gaps, the pressure drops across the MR valve with both annular and radial fluid flow gaps are larger than those across the MR valve with only annular fluid flow gaps for varying valve parameters. The radial fluid flow gaps in the MR valve can reach a higher efficiency and larger controllable range than those by annular fluid flow gaps to some extent

  15. Annular-Efficient Triangulations of 3-manifolds

    CERN Document Server

    Jaco, William

    2011-01-01

    A triangulation of a compact 3-manifold is annular-efficient if it is 0-efficient and the only normal, incompressible annuli are thin edge-linking. If a compact 3-manifold has an annular-efficient triangulation, then it is irreducible, boundary-irreducible, and an-annular. Conversely, it is shown that for a compact, irreducible, boundary-irreducible, and an-annular 3-manifold, any triangulation can be modified to an annular-efficient triangulation. It follows that for a manifold satisfying this hypothesis, there are only a finite number of boundary slopes for incompressible and boundary-incompressible surfaces of a bounded Euler characteristic.

  16. Experimental investigations for determination of heat-transfer coefficients and temperature fields in simulated fuel assemblies of BREST reactor with fuel elements spaced by transverse grids

    International Nuclear Information System (INIS)

    The consideration is given to heat transfer and temperature fields in fuel pin bundles with transverse spacer grids (s/d =1.33) equally spaced along energy deposition length. Experimental data are obtained on two simulated 37-rod core assemblies: one assembly is with uniform geometry along the cross-section and in the other there is nonheated rod simulating supporting pipe in fuel assembly of reactor with heavy coolant. Eutectic Na-K alloy is used as coolant. Nusselt numbers and temperature nonuniformity along the perimeter of measurement fuel element simulator obtained in these assemblies are compared as well as available data for finned (wire to wire) fuel rods

  17. Feasibility study of power reactor fuel elements factory development: I. Economical aspects

    International Nuclear Information System (INIS)

    For determining the feasibility study on manufacturing nuclear fuel element from economical aspect point of view, it necessary to fix its capacity which it was found from fuel element reloading requirement for nuclear power plat (PLTN). NEWJEC report which use as a base in this study that is possibly of a complex of NPP as big as 7200 MW in Muria region. If the capacity factor is 80 %, the reload requirement is therefore become from 120 to 142 tons uranium every year. So, its considered to fix the nominal capacity of a fabric for nuclear fuel element manufacturing as much as 200 tons-U per year with economical lifetimes of 20 years. NEWJEC data show, for manufacturing capacity of 200 tons-U per year with, plant have a fixed capital investment of US$ 43.9 million. With working capital as much as 15 % correspond to fixed capital investment (FCI); 10 % of interest rate; US$ 17 million of fixed cost; US$ 106.2/kg-U of variable production cost, its calculated that break even point/BEP is 50 % for price of nuclear fuel is US$ 350/kg-U without uranium cost. On this economic condition, it was found that the return on investment/ROI is 20.2 %; the internal rate of return/IRR is 11.2 % and the benefit cost ration/BCR is 1.22. For all of above, it was assumed that such nuclear fuel element manufacturing service will be operate in the year of 2012. Some of NEWJEC data have been revised, there were the value of FCI; cost of salary; the value in percent of working capital/WC; the cost of non-uranium materials and the price of product service are US$ 68 million; US$ 4.1 million; 30 %; US$ 100/kg-U and US$ 370/kg-U respectively, where the new data appear as higher than old date from NEWJEC, excluding the cost of salary. For all new economical data in the latest, we found that 45 %; 16.73 %; 11.8 % and 1.25 for BEB; IRR and BCR respectively

  18. Economical benefits for the use of slightly enriched fuel elements at the Atucha-I nuclear power plant

    International Nuclear Information System (INIS)

    The fuel represents a very important factor in the operative cost of the Atucha I nuclear power plant. This cost is drastically reduced with the use of fuel elements of slightly enriched uranium. The annual saving is analyzed with actual values for fuel elements with an enrichment of 0.85% by weight of U-235. With the reactor core in equilibrium state the annual saving achieved is approximately 7.5-10 u$s. According to the present irradiation plan, the benefit for the transition period is studied. An analysis of the sensitivity to differential increments in factors determining the cost of fuel elements or to changes in manufacturing losses is also performed, calculating its effect on the waste, the storage of irradiated elements and the amount of UO2 required. (Author)

  19. Annular beam shaping and optical trepanning

    Science.gov (United States)

    Zeng, Danyong

    Percussion drilling and trepanning are two laser drilling methods. Percussion drilling is accomplished by focusing the laser beam to approximately the required diameter of the hole, exposing the material to one or a series of laser pulses at the same spot to melt and vaporize the material. Drilling by trepanning involves cutting a hole by rotating a laser beam with an optical element or an x-y galvo-scanner. Optical trepanning is a new laser drilling method using an annular beam. The annular beams allow numerous irradiance profiles to supply laser energy to the workpiece and thus provide more flexibility in affecting the hole quality than a traditional circular laser beam. Heating depth is important for drilling application. Since there are no good ways to measure the temperature inside substrate during the drilling process, an analytical model for optical trepanning has been developed by considering an axisymmetric, transient heat conduction equation, and the evolutions of the melting temperature isotherm, which is referred to as the melt boundary in this study, are calculated to investigate the influences of the laser pulse shapes and intensity profiles on the hole geometry. This mathematical model provides a means of understanding the thermal effect of laser irradiation with different annular beam shapes. To take account of conduction in the solid, vaporization and convection due to the melt flow caused by an assist gas, an analytical two-dimensional model is developed for optical trepanning. The influences of pulse duration, laser pulse length, pulse repetition rate, intensity profiles and beam radius are investigated to examine their effects on the recast layer thickness, hole depth and taper. The ray tracing technique of geometrical optics is employed to design the necessary optics to transform a Gaussian laser beam into an annular beam of different intensity profiles. Such profiles include half Gaussian with maximum intensities at the inner and outer

  20. Out-of-pile modelling of nuclear fuel elements for MTR type reactors. Pt. 1

    Energy Technology Data Exchange (ETDEWEB)

    Farhadi, Kazem [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Radiations Application Research School

    2014-03-15

    In the first part of the present paper, for a 5 MW thermal pool-type research reactor, the fuel element is modelled for when undergoing both natural circulation of the coolant and forced convection of the coolant operational conditions. First, the required dimensionless groups were identified and then the pertinent similarity criteria were derived accordingly. The derived similitude laws were modified under the conditions of identical pressure, identical temperature difference and identical coolant and fuel cladding in the model and the prototype. These modifications were done for the system under both natural and forced convections. The effect of varying cladding materials under normal operating conditions of the research reactor were observed via coolant channel thickness. Also the effect of a wider coolant channel on the nature of the coolant fluid was observed. The results obtained indicate that it is not possible to conserve all the dimensionless groups between the model and the prototype and hence achieve an errorless outcome. Among all the liquids available, methanol is the only liquid which nearly satisfies the thermal-hydraulic similitude and must be used in place of ordinary coolant water. This in turn necessitates the coolant channel to be wider and as a consequence the traditional Aluminium cladding in research reactors should be replaced by Iron. The derived scale down criteria can be used for the design of fuel element for the out-of-pile testing. (orig.)

  1. Radionuclide Compositions and Total Activity of Spent MTR-HEU Fuel Elements of the IAN-R1 Research Reactor

    Science.gov (United States)

    Sarta, Josè A.; Castiblanco, Luis A.

    2005-05-01

    With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energìas Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer of the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.

  2. Regulatory inspection practices on fuel elements and core lay-out at NPPs

    International Nuclear Information System (INIS)

    The basic description of the reactor core of a nuclear power plant (NPP) is an important part of the Safety Analysis Report in all countries. Due to increased interest by regulatory authorities in the Member countries, in 1996 WGIP proposed looking at inspection aspects on fuel elements and core lay-out at nuclear power plants. The CNRA subsequently approved proceeding with this report. The report deals primarily with inspection practices and inspection requirements during nuclear power plant (NPP) operation with special emphasis on refuelling procedures. All license related topics, such as fuel and core design (mechanical, neutronic, thermal-hydraulic), as well as inspection philosophy and practices on fuel fabrication are included as appropriate serving as background information and may not be completely described. WGIP members describe their country's inspection programme according to the structure of a questionnaire (appendix 1). The individual contributions are contained in the appendix 2 and are compiled within the main chapters (1 through 3). Report Structure: 1. Licensing and Quality Assurance (QA) requirements for nuclear fuel; 2. Regulatory inspection programme during NPP operation and refuelling outages; 3. Procedures for inspection practices and inspection programme. Appendix: Questionnaire and Country specific contributions. Contributions are presented by Belgium, Finland, France, Germany, Hungary, Italy, Japan, Mexico, The Netherlands, Spain, Sweden, Switzerland, United Kingdom, USA

  3. Studies on production planning of IPEN fuel-element plant in order to meet RMB demand

    Energy Technology Data Exchange (ETDEWEB)

    Negro, Miguel L.M.; Saliba-Silva, Adonis M.; Durazzo, Michelangelo, E-mail: mlnegro@ipen.br, E-mail: saliba@ipen.br, E-mail: mdurazzo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The plant of the Nuclear Fuel Center (CCN) will have to change its current laboratorial production level to an industrial level in order to meet the fuel demand of RMB and of IEA-R1. CCN's production process is based on the hydrolysis of UF6, which is not a frequent production route for nuclear fuel. The optimization of the production capacity of such a production route is a new field of studies. Two different approaches from the area of Operations Research (OR) were used in this paper. The first one was the PERT/CPM technique and the second one was the creation of a mathematical linear model for minimization of the production time. PERT/CPM's results reflect the current situation and disclose which production activities may not be critical. The results of the second approach show a new average time of 3.57 days to produce one Fuel Element and set the need of inventory. The mathematical model is dynamic, so that it issues better results if performed monthly. CCN's management team will therefore have a clearer view of the process times and production and inventory levels. That may help to shape the decisions that need to be taken for the enlargement of the plant's production capacity. (author)

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

    International Nuclear Information System (INIS)

    Within the framework of the RERTR Program and supported by the technical cooperation work agreed upon between the USA and Argentina in May 1979, the CNEA Nuclear Fuel Department - Low Enriched Fuel Elements Project (ECBE Project) has carried on its own program for developing fuels with low enrichment for research and test reactors. Up to the present, its main objective has been to replace the highly enriched fuel used in its only reactor (RA-3) for research, development and radioisotopes production. The basic stages of the Argentine Program are shown. The densities reached were within the range of 3.12 to 3.58 g/cm3 for U3O8-Al, 2.99 to 3.09 g/cm3 for UAl2-Al and 5.18 to 6.10 g/cm3 for U3Si-Al. If further miniplates can be irradiated, it is the purpose of the program to research uranium densities of 3.5 g/cm3 in UAl2-Al and 6.5 g/cm3 in U3Si-Al

  5. Fusion solution to dispose of spent nuclear fuel, transuranic elements, and highly enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Yousry E-mail: gohar@anl.gov

    2001-11-01

    The disposal of the nuclear spent fuel, the transuranic elements, and the highly enriched uranium represents a major problem under investigation by the international scientific community to identify the most promising solutions. The investigation of this paper focused on achieving the top rated solution for the problem, the elimination goal, which requires complete elimination for the transuranic elements or the highly enriched uranium, and the long-lived fission products. To achieve this goal, fusion blankets with liquid carrier, molten salts or liquid metal eutectics, for the transuranic elements and the uranium isotopes are utilized. The generated energy from the fusion blankets is used to provide revenue for the system. The long-lived fission products are fabricated into fission product targets for transmutation utilizing the neutron leakage from the fusion blankets. This paper investigated the fusion blanket designs for small fusion devices and the system requirements for such application. The results show that 334 MW of fusion power from D-T plasma for 30 years with an availability factor of 0.75 can dispose of the 70,000 tons of the U.S. inventory of spent nuclear fuel generated up to the year 2015. In addition, this fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future.

  6. Operational health physics safeguards during transfer of 38 used Triga fuel elements

    International Nuclear Information System (INIS)

    During the spring of 1974, the University of California, Irvine acquired 38 used Triga fuel elements. The health physics planning of the operation was designed chiefly by the Reactor Supervisor in conjunction with the Radiation Safety Officer. 13 transfer operations of three fuel elements each in a one-inch shielded transfer cask were required to take the elements out of the shipping cask, 70 feet through air, and lowered into a shielded transport cast. The latter was moved under the reactor compartment crane for final transfer to the reactor poll for storage. Radiation levels up to 2.7 roentgens per hour (r/hr) at a meter from the transfer cask were noted. Both Crane Operators served the entire eight-hour shift. The Reactor Crane Operator was often bathed in a 160 mr/hr field. The outside Crane Operator was often exposed at levels up to 80 mr/hr. Exposures as indicated by both body badges during the operation were well below monthly exposure levels, even when radiation exposures from normal reactor operation were added in. Aspects of the handling operation, dry runs, additional shielding, and field dosimetry played a large part in achieving a relatively low exposure to the very few people utilized for this extensive operation

  7. Water reactor fuel element computer modelling in steady state, transient and accident conditions

    International Nuclear Information System (INIS)

    The present meeting was scheduled by the International Atomic Energy Agency, upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT). This meeting was the fifth in the series of IAEA meetings on the topic of Water Reactor Fuel Element Modelling, previous meetings being held in 1978, 1980, 1982 and 1984. Sixty-seven participants from 21 countries attended the meeting, and 35 papers were presented and discussed. These numbers are almost exactly the same as for the 1984 meeting, which demonstrates a continuing interest in the topic. The papers were presented in five sessions under the following headings: Session I - General Modelling (6 papers); Session II - Thermo-Mechanical Modelling and PCI (7 papers); Session III - Fission Gas Release (7 papers); Session IV - Transient Behaviour (8 papers); Session V - Axial Gas Transport and Thermal Modelling (7 papers). A separate abstract was prepared for each of these 35 papers. Refs, figs and tabs

  8. Fuel element failure detection experiments, evaluation of the experiments at KNK II/1 (Intermediate Report)

    CERN Document Server

    Bruetsch, D

    1983-01-01

    In the frame of the fuel element failure detection experiments at KNK II with its first core the measurement devices of INTERATOM were taken into operation in August 1981 and were in operation almost continuously. Since the start-up until the end of the first KNK II core operation plugs with different fuel test areas were inserted in order to test the efficiency of the different measuring devices. The experimental results determined during this test phase and the gained experiences are described in this report and valuated. All three measuring techniques (Xenon adsorption line XAS, gas-chromatograph GC and precipitator PIT) could fulfil the expectations concerning their susceptibility. For XAS and GC the nuclide specific sensitivities as determined during the preliminary tests could be confirmed. For PIT the influences of different parameters on the signal yield could be determined. The sensitivity of the device could not be measured due to a missing reference measuring point.

  9. Selective Catalytic Oxidation of Hydrogen Sulfide to Elemental Sulfur from Coal-Derived Fuel Gases

    Energy Technology Data Exchange (ETDEWEB)

    Gardner, Todd H.; Berry, David A.; Lyons, K. David; Beer, Stephen K.; Monahan, Michael J.

    2001-11-06

    The development of low cost, highly efficient, desulfurization technology with integrated sulfur recovery remains a principle barrier issue for Vision 21 integrated gasification combined cycle (IGCC) power generation plants. In this plan, the U. S. Department of Energy will construct ultra-clean, modular, co-production IGCC power plants each with chemical products tailored to meet the demands of specific regional markets. The catalysts employed in these co-production modules, for example water-gas-shift and Fischer-Tropsch catalysts, are readily poisoned by hydrogen sulfide (H{sub 2}S), a sulfur contaminant, present in the coal-derived fuel gases. To prevent poisoning of these catalysts, the removal of H{sub 2}S down to the parts-per-billion level is necessary. Historically, research into the purification of coal-derived fuel gases has focused on dry technologies that offer the prospect of higher combined cycle efficiencies as well as improved thermal integration with co-production modules. Primarily, these concepts rely on a highly selective process separation step to remove low concentrations of H{sub 2}S present in the fuel gases and produce a concentrated stream of sulfur bearing effluent. This effluent must then undergo further processing to be converted to its final form, usually elemental sulfur. Ultimately, desulfurization of coal-derived fuel gases may cost as much as 15% of the total fixed capital investment (Chen et al., 1992). It is, therefore, desirable to develop new technology that can accomplish H{sub 2}S separation and direct conversion to elemental sulfur more efficiently and with a lower initial fixed capital investment.

  10. Atomic pile Directorate, Department of Metallurgy, Departments of Technology, Department of Fuel Elements and Structures, Division of Study of Fuel Elements - Semi annual report on the 1968-10-1

    International Nuclear Information System (INIS)

    This document gathers a set of reports of studies performed on nuclear fuel elements. The addressed topics are: creep behaviour of UMo and UMoAl tubes and pellets under the action of an external pressure (creep strength of tubes under external pressure, creep strength of pellets under external pressure, uncertainties on irradiation parameters in Pegase), problems related to centring devices (measurements and tests), irradiations of ring elements in power reactors, uranium/sheath metallurgical relationship for Bugey and influence of irradiation (cartridge behaviour in Pegase, long duration irradiation in power reactors, extrapolation in Bugey of results obtained in G2), theoretical study of kinetic oxidation phenomena in metal fuels, tests of leaking cartridges in EdF2, evolution of pressure in EL4 type irradiated fuel rods with ZrCu liners with respect to the conductivity integral, a focus on irradiations of Z0 type fuel elements in Pegase, cluster safety tests with uranium carbide in pile and out of pile, a review of studies performed on fuel elements with blowhole, and application of neutrography to fuel elements

  11. KUEBEL. A Fortran program for computation of cooling-agent-distribution within reactor fuel-elements

    International Nuclear Information System (INIS)

    KUEBEL is a Fortran-program for computation of cooling-agent-distribution within reactor fuel-elements or -zones of theirs. They may be assembled of max. 40 cooling-channels with laminar up to turbulent type of flow (respecting Reynolds' coefficients up to 2.0E+06) at equal pressure loss. Flow-velocity, dynamic flow-, contraction- and friction-losses will be calculated for each channel and for the total zone. Other computations will present mean heat-up of cooling-agent, mean outlet-temperature of the core, boiling-temperature and absolute pressure at flow-outlet. All characteristic coolant-values, including the factor of safety for flow-instability of the most-loaded cooling gap are computed by 'KUEBEL' too. Absolute pressure at flow-outlet or is-factor may be defined as dependent or independent variables of the program alternatively. In latter case 3 variations of solution will be available: Adapted flow of cooling-agent, inlet-temperature of the core and thermal power. All calculations can be done alternatively with variation of parameters: flow of cooling-agent, inlet-temperature of the core and thermal power, which are managed by the program itself. 'KUEBEL' is able to distinguish light- and heavy-water coolant, flow-direction of coolant and fuel elements with parallel, rectangular, respectively concentric, cylindrical shape of their gaps. Required material specifics are generated by the program. Segments of fuel elements or constructively unconnected gaps can also be computed by means of interposition of S.C. 'phantom channels'. (orig.)

  12. 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

  13. Identification of a leaking TRIGA fuel element at the reactor facility of Pavia

    International Nuclear Information System (INIS)

    On January 28th 2004, during a periodical activity of characterization of the ionic-exchange resins of the demineralizer of the primary cooling circuit of the TRIGA Mark II reactor of the University of Pavia a small but detectable amount of 137Cs contamination was measured. Since the reactor has been running for several hundreds of hours at full power without showing any anomaly in the radiometric and thermo-hydraulic parameters, the reactor was brought at the nominal power of 250 kW for one hour and a sample of water was collected from the reactor tank and analyzed in a low background gamma-ray detector. As a result a small amount of fission products were detected in the reactor pool water (few Bq/g) suggesting the existence of a possible clad defect in one ore more fuel elements. As a consequence of this situation a campaign of gamma-ray spectrometry was implemented in order to evaluate the importance of the release. Analyzes using a HGe detector (1.72 keV FWHM - 31.3 % efficiency - 58.5 Photo Peak/Compton) were performed and the most significant results are presented as well as the identification of the leaking fuel element. The fission products leakage was due to a micro-fissure of a fuel element that released only noble gas when it was heated up to a temperature around 90oC , i.e. at the reactor power of about 100 kW. The oldest SST clad instrumented fuel element in the core was identified as the origin of the release. It was removed from its position and stored in a rack of the reactor pool under 4 m of water shield. The reactor came back in regular operation on March 22nd 2004 and no other fission products leakages were detected. After this situation the reactor pool water is sampled and measured with a low-background gamma-ray detector every month before the reactor start-up and after one hour of operation of the reactor at full nominal power. (nevyjel)

  14. Control of criticality risk in the manufacture of fuel elements for research reactors

    International Nuclear Information System (INIS)

    The control of criticality risk in a chemical plant adopts different forms according to the quantities of fissile material and the type of compounds used. This work presents the treatment of the critical excursion risk adopted in production plants of U3 O8 and manufacturing plants of fuel elements for research reactors, located in Constituyentes Atomic Center. The possible events and accidents related to the fissile material control are analyzed, and the systems of administrative control and intrinsic safety through engineering are described. (Author)

  15. Welding technique research between fuel element Zr-4 end and tantalum tube and its application

    International Nuclear Information System (INIS)

    The author introduces welding technique between fuel element cladding Zr-4 end and tantalum tube which is used to introduce the W/Re thermocouple and water pressure test and corrosion test of the weld. The results show that the welded surface takes form better, lacks metallurgical defect and the weld causes no leakage under 9.0 MPa pressure. There is the corrosion phenomenon in the weld after heating at 450 degree C for 100 h in melting PbBi alloy, but it can retard the corrosion in PbBi alloy after preoxidation film treatment. The adopted measures has been proposed in the design and assemble of irradiation capsule

  16. Remote examination of Fort St. Vrain HTGR fuel and reflector elements

    International Nuclear Information System (INIS)

    A robotic device developed at GA Technologies was used to perform remote metrological and visual examinations on 105 irradiated fuel and reflector elements removed from the High Temperature Gas Cooled Reactor (HTGR) core of the Fort St. Vrain (FSV) Nuclear Generating Station. All examinations were accomplished in a fraction of the time and at a fraction of the cost for manual methods. Theses examinations have been useful in the verification of core design codes, in monitoring the in-pile performance of graphite blocks, and in qualifying improved core materials

  17. Ultrasonic inspection of end cap welds in thin walled fuel elements

    International Nuclear Information System (INIS)

    The quality of end cap welds in thin walled fuel elements can be determined by using high frequency, sharp focussed, long focus transducrs. It was necessary to determine the true angle of refraction, which was different from what theory predicted, and the attenuation of a focussed beam skipping inside a thin wall. It was found possible to penetrate the weld zone, recognize the weld contours and determine the presence of defects smaller than 0.1 mm deep. Case histories show that no weld considered good by ultrasonics was defective and nearly all defects found ultrasonically were also found metallographically

  18. Multi-cell thermionic fuel element for nuclear electric power and propulsion system

    Science.gov (United States)

    Nikolaev, Yuri V.; Gontar, Alexander S.; Eremin, Stanislav A.; Lapochkin, Nikolai V.; Andreev, Pavel V.; Zhabotinsky, Evgeny E.

    1999-01-01

    Conceptual problems of development of two-mode multi-cell thermionic fuel element (TFE) for nuclear electric power and propulsion system are considered. The results of analysis of the design and TFE output parameters are presented. It is shown that application of advanced high effective materials and technologies provides operating of the TFE in two modes: a) in nominal mode of power generation for power supply of spacecraft payload at operational orbit and b) in forced mode of power generation for power supply of electric thrusters under spacecraft orbit transfer from intermediate to operational one.

  19. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    Energy Technology Data Exchange (ETDEWEB)

    SCHWINKENDORF, K.N.

    2006-05-12

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements. The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprising two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with ''green'' (fresh) fuel and one with spent fuel. Both the green and spent fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, 3 green fuel and 4 spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements can

  20. Exhaust emissions of a double annular combustor: Parametric study

    Science.gov (United States)

    Schultz, D. F.

    1974-01-01

    A full scale double-annular ram-induction combustor designed for Mach 3.0 cruise operation was tested. Emissions of oxides of nitrogen, carbon monoxide, unburned hydrocarbons, and smoke were measured over a range of combustor operating variables including reference velocity, inlet air temperature and pressure, and exit average temperature. ASTM Jet-A fuel was used for these tests. An equation is provided relating oxides of nitrogen emissions as a function of the combustor, operating variables. A small effect of radial fuel staging on reducing exhaust emissions (which were originally quite low) is demonstrated.

  1. Performance of low-enriched U3Si2-aluminum dispersion fuel elements in the Oak Ridge Research Reactor

    International Nuclear Information System (INIS)

    Six high-density, low-enriched U3Si2-Al dispersion fuel elements have been tested in the Oak Ridge Research Reactor (ORR). The elements were geometrically identical to standard ORR elements. The uranium density in the fuel meat ranged between 4.6 and 5.2 Mg/m3. The elements were fabricated by B and W, CERCA, and NUKEM using their normal materials and fabrication practices, with minor modifications necessitated by the new fuel. The U3Si2 contained minor amounts of USi, U3Si and/or uranium solid solution. The elements were irradiated to approximately normal ORR burnup, and three elements were irradiated twice as long, to average burnups of ∼80% of the initially contained 235U, well above the burnups normally achieved in research and test reactors. Peak burnups of 98% were achieved. Following suitable cooling periods, the elements were subjected to a series of nondestructive and destructive examinations. The behavior of the fuel was found to be entirely consistent with the known irradiation behavior of the constituent phases. The extremely stable swelling behavior of the U3Si2 phase dominated in all cases. The plates showed small, uniform thickness changes. Blister threshold temperatures were ≥5500C. It is concluded that low-enriched U3Si2-Al dispersion fuel elements will perform at least as well in research and test reactors with power densities up to that of the ORR as the highly enriched UAl/sub x/-Al and U3O8-Al dispersion fuels currently being used. There were no indications that use of this fuel under substantially more stringent conditions might be precluded. 10 refs., 87 figs., 9 tabs

  2. Shut-down margin study for the next generation VVER-1000 reactor including 13 x 13 hexagonal annular assemblies

    International Nuclear Information System (INIS)

    Highlights: → Shut-Down Margin (SDM) for the next generation annular fuel core of typical VVER-1000, 13 x 13 assemblies are calculated. → The MCNP-5 code is run for many cases with different core burn up at various core temperatures. → There is a substantial drop in SDM in the case of annular fuel for the same power level. → SDM for our proposed VVER-1000 annular pins is calculated for specific average fuel burn up values at the BOC, MOC, and EOC. - Abstract: Shut-Down Margin (SDM) for the next generation annular fuel core of typical VVER-1000, 13 x 13 assemblies are calculated as the main aim of the present research. We have applied the MCNP-5 code for many cases with different values of core burn up at various core temperatures, and therefore their corresponding coolant densities and boric acid concentrations. There is a substantial drop in SDM in the case of annular fuel for the same power level. Specifically, SDM for our proposed VVER-1000 annular pins is calculated when the average fuel burn up values at the BOC, MOC, and EOC are 0.531, 11.5, and 43 MW-days/kg-U, respectively.

  3. Design of reactor internals in larger high-temperature reactors with spherical fuel elements

    International Nuclear Information System (INIS)

    In his paper, the author analyzes and summarizes the present state of the art with emphasis on the prototype reactor THTR 300 MWe, because in addition to spherical fuel elements, this type includes other features of future HTR design such as the same flow direction of cooland gas through the core. The paper on hand also elaborates design guidelines for reactor internals applicable with large HTR's of up to 1200 MWe. Proved designs will be altered so as to meet the special requirements of larger cores with spherical elements to be reloaded according to the OTTO principle. This paper is furthermore designed as a starting point for selective and swift development of reactor internals for large HTR's to be refuelled according to the OTTO principle. (orig./GL)

  4. Numerical analysis of nuclear fuel element end cap resistance upset welding using SORPAS

    International Nuclear Information System (INIS)

    Finite element simulation of nuclear fuel element end cap solid state resistance upset welding process is attempted using dedicated resistance welding FE simulation package SORPAS. Basic two-dimensional axsymmetric modeling has been done with temperature dependent electrical, thermal and structural properties for Zircaloy-4. The simulation results of weld bead shape and weld upset are compared with experimental values with same welding parameters. It is observed that simulation package SORPAS is very useful for modeling bulk deformation process such as end cap welding. Using the developed two-dimensional model, the simulation results are obtained with varying pre-heat and post-heat welding current, pre-heat and post-heat cycle duration. The observations are validated using experimental results with same range of varying welding parameter values. However, better modeling and analysis of complicated end cap resistance upset welding is possible in future by simulating real welding environments using SORPAS and suitable experimental arrangements

  5. A finite element method with contact for tensile analysis in fuel rods

    International Nuclear Information System (INIS)

    Elements for mechanical analysis of fuel rod of a PWR type reactor, are presented. The rod, consists basically in a cylindrical coating of zircalloy which contains pilling of UO2 pellets, is submitted to strong internal and external pressures, intense temperature gradients and neutron flux. These conditions lead several phenomena in the pellet (swelling, fracture, densification, creep) and in the cladding (embrittlement, corrosion, creep) which undergo deformations leading them to contact the restriction for the interpenetration is included in the problem without restriction by Lagrange multipliers. Considering a non-linear problem, due to the surface of contact to be not known a priori, the numerical solutions were obtained using the finite element method. (M.C.K.)

  6. Radioactivity calculation and measurement of spent fuel elements of SPRR-300

    International Nuclear Information System (INIS)

    As one of the original preparation works of SPRR-300 decommissioning, radioactivity of spent fuel elements in the reactor should be estimated by both calculation and experiment. ORIGEN2 is chosen to do the calculation. The ACCESS database of reactor operation history is built, whose tables is connected by load ID. The needed input history data is obtained using inquiries. Assistant programs, such as Operation Data Processing Program, Composition Data Processing Program, etc. are developed to change the original data into ORIGEN2 input file. In the output file, there are nuclides data, gamma activity data, and so on. Gamma spectrum of three elements in three distances is measured. radioactivity of three nuclides measured is compared to that calculated. The two kinds of data are at the same order. So the data calculated are believable. The comparison shows that the data calculated are reliable. (authors)

  7. Experimental evaluation of different mixing promoter for nuclear fuel element by means of a new thermal tracing technique

    International Nuclear Information System (INIS)

    In this work a new experimental method is used to experimentally evaluate the performance of different appendages promoting the turbulent mixing between the coupled subchannels of nuclear fuel elements.The method used will be introduced in another presentation and consists in the generation and measurement of small thermal traces in the refrigerating water flow between the fuel rods.Because it is suitable for heterogeneous and compact subchannels (as Argentinean fuels) with high water flows in simple and affordable tests at atmospheric pressure, this new method is specially well suited for the design of fuel elements, while it offers advantages over other methods of mixing measurement.The experiments carried out on a small test section proved that the buttons brazed to the fuel rods (similar to the 'turbulence promoters' of the Canflex fuel) had an excellent thermohydraulic performance as compared to different mixing vane designs studied.The thermal traces method developed has shown its potential as a thermohydraulic design tool for the development of advanced nuclear fuels, that eventually incorporate mixing promoter elements. In the case of CARA, and as it includes spacer grids, it could be possible to use them to incorporate these elements without the need of brazing them to the rods (as is the case in Canflex), and therefore without penalizing its integrity

  8. Theoretical models to predict the transient heat transfer performance of HIFAR fuel elements under non-forced convective conditions

    International Nuclear Information System (INIS)

    Simple theoretical models have been developed which are suitable for predicting the thermal responses of irradiated research fuel elements of markedly different geometries when they are subjected to loss-of-coolant accident conditions. These models have been used to calculate temperature responses corresponding to various non-forced convective conditions. Comparisons between experimentally observed temperatures and calculated values have shown that a suitable value for surface thermal emissivity is 0.35; modelling of the fuel element beyond the region of the fuel plate needs to be included since these areas account for approximately 25 per cent of the thermal power dissipated; general agreement between calculated and experimental temperatures for both transient and steady-state conditions is good - the maximum discrepancy between calculated and experimental temperatures for a HIFAR Mark IV/V fuel element is ∼ 70 deg C, and for an Oak Ridge Reactor (ORR) box-type fuel element ∼ 30 deg C; and axial power distribution does not significantly affect thermal responses for the conditions investigated. Overall, the comparisons have shown that the models evolved can reproduce experimental data to a level of accuracy that provides confidence in the modelling technique and the postulated heat dissipation mechanisms, and that these models can be used to predict thermal responses of fuel elements in accident conditions that are not easily investigated experimentally

  9. Application of gamma scanning and neutron radiography methods to control fuel element state as tested in MR reactor loop channels

    International Nuclear Information System (INIS)

    The gamma-scanning and neutron radiography methods are described used for non-destructive control of fuel elements after their test in the loop channels of MR reactor, and also the equipment utilized. The techniques and the results obtained while studying fuel elements using the above mentioned methods are provided. It is established that gamma-scanning method can only indicate the presence of defect in the continuity of the fuel element core without identifying its type whereas the advantage of neutron radiography method is in obtaining visual results. At the same time gamma-scanning method makes it possible to determine energy release on the length of fuel elements, to find the burn up fraction, to study the phenomenon of fusion products migration which is difficult or impossible with neutron radiography method. A conclusion is drawn that gamma-scanning and neutron radiography methods successfully supplement each other and make it possible to obtain important information on fuel state in the irradiated fuel elements

  10. Dose Rate Calculations of Spent MTR-HEU Fuel Elements of the IAN-R1 Research Reactor

    Science.gov (United States)

    Sarta Fuentes, Jose Antonio

    2005-04-01

    With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safetly, a program was established at the Instituto de Ciencias Nucleares y Energìas Alternativas (INEA). This program included training, acquisition of hardware and sofware, design and construction of a decay pool, transfer of the spent HEU fuel elements into the decay pool and his final transport to Savanah River in United States. In this paper are presented external dose rates which were calculated for a standard spent MTR-HEU fuel element of the IAN-R1 Research Reactor. The calculations take in consideration the activity due to contributions of fission, activation and actinides products for each relevant radionuclide present in a standard spent MTR-HEU fuel. The datas obtained were the base for the respective dosimetric evaluations in the transfering operations of fuel elements into the decay pool and for shielding calculations in designing of the decay pool.

  11. The possibility of prediction of the lifetime of metallic nuclear fuel elements in a radiation field of thermal nuclear reactors

    International Nuclear Information System (INIS)

    An attempt is made to clarify the possible causes of failure of irradiated nuclear fuel cartridges, in order to determine the parameters which govern the lifetime of the fuel and a way to predict it. Measurements of mechanical properties of irradiated uranium metal and cladding, can serve as a basis for failure prediction. Testing irradiated fuel elements by bending till fracture enables to evaluate the integral character of the fuel element, along the cross-section, taking into account the difference in brittleness of several zones. It is likely that the bending test, which indicates the behaviour of a stress-strain function, is a faster and more reliable way to determine the mechanical properties of irradiated nuclear fuel. Since the stresses applied to the cladding during irradiation are locally hydrostatic, its postirradiation blow-up provide information on strength and elasticity variations of the irradiated cladding material. (B.G.)

  12. Elemental characterization of particulate matter emitted from biomass burning: Wind tunnel derived source profiles for herbaceous and wood fuels

    Science.gov (United States)

    Turn, S. Q.; Jenkins, B. M.; Chow, J. C.; Pritchett, L. C.; Campbell, D.; Cahill, T.; Whalen, S. A.

    1997-02-01

    Particulate matter emitted from wind tunnel simulations of biomass burning for five herbaceous crop residues (rice, wheat and barley straws, corn stover, and sugar cane trash) and four wood fuels (walnut and almond prunings and ponderosa pine and Douglas fir slash) was collected and analyzed for major elements and water soluble species. Primary constituents of the particulate matter were C, K, Cl, and S. Carbon accounted for roughly 50% of the herbaceous fuel PM and about 70% for the wood fuels. For the herbaceous fuels, particulate matter from rice straw in the size range below 10 μm aerodynamic diameter (PM10) had the highest concentrations of both K (24%) and Cl, (17%) and barley straw PM10 contained the highest sulfur content (4%). K, Cl, and S were present in the PM of the wood fuels at reduced levels with maximum concentrations of 6.5% (almond prunings), 3% (walnut prunings), and 2% (almond prunings), respectively. Analysis of water soluble species indicated that ionic forms of K, Cl, and S made up the majority of these elements from all fuels. Element balances showed K, Cl, S, and N to have the highest recovery factors (fraction of fuel element found in the particulate matter) in the PM of the elements analyzed. In general, chlorine was the most efficiently recovered element for the herbaceous fuels (10 to 35%), whereas sulfur recovery was greatest for the wood fuels (25 to 45%). Unique potassium to elemental carbon ratios of 0.20 and 0.95 were computed for particulate matter (PM10 K/C(e)) from herbaceous and wood fuels, respectively. Similarly, in the size class below 2.5 μm, high-temperature elemental carbon to bromine (PM2.5 C(eht)/Br) ratios of ˜7.5, 43, and 150 were found for the herbaceous fuels, orchard prunings, and forest slash, respectively. The molar ratios of particulate phase bromine to gas phase CO2 (PM10 Br/CO2) are of the same order of magnitude as gas phase CH3Br/CO2 reported by others.

  13. Coaxial fuel and air premixer for a gas turbine combustor

    Science.gov (United States)

    York, William D; Ziminsky, Willy S; Lacy, Benjamin P

    2013-05-21

    An air/fuel premixer comprising a peripheral wall defining a mixing chamber, a nozzle disposed at least partially within the peripheral wall comprising an outer annular wall spaced from the peripheral wall so as to define an outer air passage between the peripheral wall and the outer annular wall, an inner annular wall disposed at least partially within and spaced from the outer annular wall, so as to define an inner air passage, and at least one fuel gas annulus between the outer annular wall and the inner annular wall, the at least one fuel gas annulus defining at least one fuel gas passage, at least one air inlet for introducing air through the inner air passage and the outer air passage to the mixing chamber, and at least one fuel inlet for injecting fuel through the fuel gas passage to the mixing chamber to form an air/fuel mixture.

  14. Elements of radioprotection culture at Nuclear Fuel Plant at Pitesti, Romania

    International Nuclear Information System (INIS)

    Culture as an element of human existence is characterized primarily by accumulating, skills, attitudes, behavior and promotion to something. Nuclear security culture has developed a desire to prepare to cope with nuclear risk, born with the discovery of radioactivity. Culture of radiation protection is required for the implementation of safety and security in the protection against ionizing radiation. In the activity of producing nuclear fuel for CANDU based on natural and depleted uranium stakeholders are developing a specific radiation protection culture due the existing radiological risks that can lead to exposure to ionizing radiation. For Nuclear Fuel Plant Pitesti (FCN) where nuclear material is both as bulk and as itemized form, individual radiological monitoring as well as the working environment are established in Radiological Security Manual (MSR), in accordance with national legislation and international. All employees FCN are classified as A and B classes of exposure to ionizing radiation. Classification is based on work performed and the degree of exposure and specific training for radiological safety is sized for each category. External workers, contractors and visitors are aware of the entry in the plant on the risks of exposure to ionizing radiation and are properly trained. Basics of radiation protection culture as individual monitoring, communication, rigorous and prudent approach, pro-active action activity are transmitted continuously. In addition the radiation protection culture existed there from the beginning of the commissioning of the plant, but its elements still remain scattered in knowledge, attitudes and behaviors of employees. (authors)

  15. Fission product release profiles from spherical HTR fuel elements at accident temperatures

    International Nuclear Information System (INIS)

    With the construction of the cold finger apparatus, a new method has been developed to determine fission product release profiles during heating tests of irradiated spherical fuel elements. 22 fuel elements with modern TRISO particles have been tested in the temperature range of 1500-25000C. In addition, experiments were done on seven UO2 samples at 1400 to 18000C. For heating times up to 100 hours at the maximum temperature, the following results were obtained: silver is the only fission product to be released at 1200-16000C by diffusion through intact SiC, but is of low significance in accident scenarios; caesium, iodine, strontium and noble gas releases up to 16000C are solely due to various forms of contamination. At 1700-18000C, corrosion-induced SiC defects cause the release of Cs, Sr, I/Xe/Kr. Above 20000C, thermal decomposition of the silicon carbide layer sets in, while pyrocarbons still remain intact. Around 16000C, the accident specific contribution of caesium, strontium, iodine and noble gas release is negligible. (orig./HP)

  16. Burnup determination of silicide MTR fuel elements (20% 235U) in the LFR laboratory

    International Nuclear Information System (INIS)

    The LFR facility is a radiochemical laboratory designed and constructed with a hot-cells line, a glove-box and a fume hood, all of them suited to work radioactive materials. At the beginning of the LFR operation a series of dissolutions of MTR irradiated silicide fuel elements was performed, and determined its isotopic composition of 235U, 239Pu and 148Nd (the last one as burn up monitor), by the thermal ionization mass spectrometry (TIMS). These assays are linked to the IAEA RLA/4/018 Regional Project 'Management of Spent Fuel from Research Reactors'. It is concluded that this technique of burn up measurement is powerful and accurate when properly applied, and permit to validate the calculation codes when isotopic dilution is performed. It is worth noticed the LFR capacity to carry on different research and development programs in the nuclear fuel cycle field, such as the previously mentioned absolute burn up measurements, or the evaluation of radioactive waste immobilization processes and researches on burnable poisons. (author)

  17. Study on the effect of the CANFLEX-NU fuel element bowing on the critical heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Ho Chun; Cho, Moon Sung; Jeon, Ji Su

    2001-01-01

    The effect of the CANFLEX-NU fuel element bowing on the critical heat flux is reviewed and analyzed, which is requested by KINS as the Government design licensing condition for the use of the fuel bundles in CANDU power reactors. The effect of the gap between two adjacent fuel elements on the critical heat flux and onset-of-dryout power is studied. The reduction of the width of a single inter-rod gap from its nominal size to the minimum manufacture allowance of 1 mm has a negligible effects on the thermal-hydraulic performance of the bundle for the given set of boundary conditions applied to the CANFLEX-43 element bundle in an uncrept channel. As expected, the in-reactor irradiation test results show that there are no evidence of the element bow problems on the bundle performance.

  18. New mitral annular force transducer optimized to distinguish annular segments and multi-plane forces.

    Science.gov (United States)

    Skov, Søren Nielsen; Røpcke, Diana Mathilde; Ilkjær, Christine; Rasmussen, Jonas; Tjørnild, Marcell Juan; Jimenez, Jorge H; Yoganathan, Ajit P; Nygaard, Hans; Nielsen, Sten Lyager; Jensen, Morten Olgaard

    2016-03-21

    Limited knowledge exists about the forces acting on mitral valve annuloplasty repair devices. The aim of this study was to develop a new mitral annular force transducer to measure the forces acting on clinically used mitral valve annuloplasty devices. The design of an X-shaped transducer in the present study was optimized for simultaneous in- and out-of-plane force measurements. Each arm was mounted with strain gauges on four circumferential elements to measure out-of-plane forces, and the central parts of the X-arms were mounted with two strain gauges to measure in-plane forces. A dedicated calibration setup was developed to calibrate isolated forces with tension and compression for in- and out-of-plane measurements. With this setup, it was possible with linear equations to isolate and distinguish measured forces between the two planes and minimize transducer arm crosstalk. An in-vitro test was performed to verify the crosstalk elimination method and the assumptions behind it. The force transducer was implanted and evaluated in an 80kg porcine in-vivo model. Following crosstalk elimination, in-plane systolic force accumulation was found to be in average 4.0±0.1N and the out-of-plane annular segments experienced an average force of 1.4±0.4N. Directions of the systolic out-of-plane forces indicated movements towards a saddle shaped annulus, and the transducer was able to measure independent directional forces in individual annular segments. Further measurements with the new transducer coupled with clinical annuloplasty rings will provide a detailed insight into the biomechanical dynamics of these devices. PMID:26903412

  19. New mitral annular force transducer optimized to distinguish annular segments and multi-plane forces.

    Science.gov (United States)

    Skov, Søren Nielsen; Røpcke, Diana Mathilde; Ilkjær, Christine; Rasmussen, Jonas; Tjørnild, Marcell Juan; Jimenez, Jorge H; Yoganathan, Ajit P; Nygaard, Hans; Nielsen, Sten Lyager; Jensen, Morten Olgaard

    2016-03-21

    Limited knowledge exists about the forces acting on mitral valve annuloplasty repair devices. The aim of this study was to develop a new mitral annular force transducer to measure the forces acting on clinically used mitral valve annuloplasty devices. The design of an X-shaped transducer in the present study was optimized for simultaneous in- and out-of-plane force measurements. Each arm was mounted with strain gauges on four circumferential elements to measure out-of-plane forces, and the central parts of the X-arms were mounted with two strain gauges to measure in-plane forces. A dedicated calibration setup was developed to calibrate isolated forces with tension and compression for in- and out-of-plane measurements. With this setup, it was possible with linear equations to isolate and distinguish measured forces between the two planes and minimize transducer arm crosstalk. An in-vitro test was performed to verify the crosstalk elimination method and the assumptions behind it. The force transducer was implanted and evaluated in an 80kg porcine in-vivo model. Following crosstalk elimination, in-plane systolic force accumulation was found to be in average 4.0±0.1N and the out-of-plane annular segments experienced an average force of 1.4±0.4N. Directions of the systolic out-of-plane forces indicated movements towards a saddle shaped annulus, and the transducer was able to measure independent directional forces in individual annular segments. Further measurements with the new transducer coupled with clinical annuloplasty rings will provide a detailed insight into the biomechanical dynamics of these devices.

  20. Fluiddynamic effects in the fuel element top nozzle area during refilling and reflooding

    International Nuclear Information System (INIS)

    During the refilling and reflooding phase following a hypothetical loss of coolant accident in lightwater cooled nuclear reactors, there will be countercurrent flow between discharging steam and the feed of emergency core cooling water. It was the objective of this research project to contribute to a better physical understanding of the fluiddynamic processes in the area of the fuel element top nozzle and so to improve emergency core cooling calculations. Therefore, experimental and theoretical investigations about the entrainment and countercurrent behaviour of gas/liquid flows have been implemented within this project. Fluiddynamic processes in the fuel element top nozzle area were simulated during the reflooding and refilling phase. Based on special internals as single and multiple-hole orifices, basic phenomena of fluidynamics were studied first with air-water. Subsequently, investigations of the system steam/water were conducted. The reactor geometry was approximated step by step, until a complete reactor fuel assembly top nozzle was constituted. The system pressure was 4.8 bars (abs), in accordance with the conditions in the reactor pressure vessel at the end of the blowdown phase. The water was initially fed in at saturation temperature, then, as a second step, fed in at subcooled condition relative to the steam temperature, in order to be able to study condensation effects as well. First, investigations on gas/liquid countercurrent flows in the fluid system air/water are presented. Then one studies countercurrent flow in the system steam/water, including the investigation of condensation effects. Finally, a detailed description of the research on droplet size determination is given

  1. Effect of water-chemical regimes and exploitation parameters on the fuel can corrosion and light-water reactor fuel elements security

    International Nuclear Information System (INIS)

    A physico-chemical model of nodal and uniform corrosion of zirconium fuel cans in reactors is proposed. The model takes account of the following physical parameters: values of neutron and thermal fluxes, burnup, temperature, as well as of the water chemistry parameters. The model developed is based on a correlation between zirconium oxide solubility in the coolant and measured values of fuel can corrosion in reactors. Based on the reliability theory, a mathematical model is proposed, connecting the rate of the can corrosion with the fuel element failure intensity at NPPs with RBMK and WWER type reactors

  2. CFD Simulation of Annular Centrifugal Extractors

    OpenAIRE

    Vedantam, S.; Wardle, K. E.; Tamhane, T. V.; Ranade, V. V.; Joshi, J. B.

    2012-01-01

    Annular centrifugal extractors (ACE), also called annular centrifugal contactors offer several advantages over the other conventional process equipment such as low hold-up, high process throughput, low residence time, low solvent inventory and high turn down ratio. The equipment provides a very high value of mass transfer coefficient and interfacial area in the annular zone because of the high level of power consumption per unit volume and separation inside the rotor due to the high g of cent...

  3. Psoriatic Arthritis with Annular Pustular Psoriasis.

    Science.gov (United States)

    Nagafuchi, Hiroko; Watanabe, Kyoko; Mikage, Hidenori; Ozaki, Shoichi

    2016-01-01

    We herein present the case of a 56-year-old woman who presented with symptoms of psoriatic arthritis (PsA) with erythema that progressed to annular pustular psoriasis. The patient had a 15-year history of polyarthritis. Annular pustular psoriasis is not typically observed in cases of arthritis. This is the first reported case of PsA with annular pustular psoriasis. PMID:26935375

  4. DEVELOPMENT OF LOW-COST MANUFACTURING PROCESSES FOR PLANAR, MULTILAYER SOLID OXIDE FUEL CELL ELEMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Scott Swartz; Matthew Seabaugh; William Dawson; Harlan Anderson; Tim Armstrong; Michael Cobb; Kirby Meacham; James Stephan; Russell Bennett; Bob Remick; Chuck Sishtla; Scott Barnett; John Lannutti

    2004-06-12

    This report summarizes the results of a four-year project, entitled, ''Low-Cost Manufacturing Of Multilayer Ceramic Fuel Cells'', jointly funded by the U.S. Department of Energy, the State of Ohio, and by project participants. The project was led by NexTech Materials, Ltd., with subcontracting support provided by University of Missouri-Rolla, Michael A. Cobb & Co., Advanced Materials Technologies, Inc., Edison Materials Technology Center, Gas Technology Institute, Northwestern University, and The Ohio State University. Oak Ridge National Laboratory, though not formally a subcontractor on the program, supported the effort with separate DOE funding. The objective of the program was to develop advanced manufacturing technologies for making solid oxide fuel cell components that are more economical and reliable for a variety of applications. The program was carried out in three phases. In the Phase I effort, several manufacturing approaches were considered and subjected to detailed assessments of manufacturability and development risk. Estimated manufacturing costs for 5-kW stacks were in the range of $139/kW to $179/kW. The risk assessment identified a number of technical issues that would need to be considered during development. Phase II development work focused on development of planar solid oxide fuel cell elements, using a number of ceramic manufacturing methods, including tape casting, colloidal-spray deposition, screen printing, spin-coating, and sintering. Several processes were successfully established for fabrication of anode-supported, thin-film electrolyte cells, with performance levels at or near the state-of-the-art. The work in Phase III involved scale-up of cell manufacturing methods, development of non-destructive evaluation methods, and comprehensive electrical and electrochemical testing of solid oxide fuel cell materials and components.

  5. Design of Production Test IP-262-A-11-FP -- Evaluation of projection fuel elements for use in ribbed process tubes -- Demonstration loading

    Energy Technology Data Exchange (ETDEWEB)

    Hodgson, W.H.; Hall, R.E.

    1959-06-29

    For several years, a major category of fuel element failures has been the side corrosion type, characterized by localized accelerated fuel element jacket corrosion. Since it has been demonstrated {sup 1} that misalignment of fuel elements in a process tube will produce flow patterns and accelerated corrosion, termed ``hot spots``, failure to align the fuel elements in process tubes is considered a contributing factor in the production of side corrosion failures. Preliminary testing of both self-supporting and ``bumper`` fuel elements is underway. Data on the self-supporting fuel elements have demonstrated that the bridge-rail projections have sufficient support strength, do not of themselves create a corrosion problem and in actuality probably eliminate any hot-spot areas. Although one tube of bumper fuel elements in KW Reactor {sup 3} has been discharged, data are not as yet available. Potentially, the most sever corrosion conditions exist during the summer months when reactor inlet temperatures are high. It is desirable then, provided bumper fuel elements limit hot- spot corrosion, to evaluate the bumper concept for large scale use possibly by the summer of 1960. To accomplish this, a demonstration loading of the bumper type fuel elements must be underway by about July, 1959. The purpose of this report is to present the design of a test to evaluate the fabrication process and irradiation performance of fuel elements having projections, which may prevent misalignment in ribbed process tubes and meet the aforementioned goals.

  6. Analysis of a possible experimental assessment of a prototype fuel element containing burnable poison in the RA-3 reactor

    International Nuclear Information System (INIS)

    The Argentine RA-3 research reactor (5 MW) is presently operated with LEU fuel by the National Atomic Energy Commission (CNEA). It belongs to the group of nuclear installations controlled, from the radiological and nuclear safety point of view, by the Nuclear Regulatory Authority (ARN). A new type of fuel elements containing burnable absorbers, with similar enrichment as the standard fuel elements but greater fissile contents, has recently been proposed for a new Argentine reactor design (RRR). In this framework the ARN considers interesting, if technically possible, the performance of an experiment in the RA-3 reactor. The experiment might enable, for such fuel element containing burnable poison, the verification of its neutronic behaviour under irradiation as well as a validation of the calculation line by comparison to measured values. It should be desirable that such experiment could reproduce as much as possible those conditions estimated for the RRR reactor, still under design in Argentina, having Silicide fuel elements with burnable poison, in the shape of cadmium wires in their structure. We here analyse a possible experiment consisting in the loading of a prototype fuel element with burnable poison in a normally loaded RA-3 core configuration. It would essentially be a standard RA-3 fuel element, having cadmium wires in its frame. This experiment would enable the verification of the prototype behaviour under irradiation, its operation limits and conditions, and particularly, the reactivity safety margins established in Argentine Standards, both calculated and measured. The main part of the experiment would imply some 200 full power days of operation at 5 MW, which would be drastically reduced if the reactor power is increased to 10 MW, as foreseen. We also show that under the proposed conditions, the experiment would not represent a significant penalty to the reactor normal operation. (author)

  7. Systematic comparison of the use of annular and Zernike circle polynomials for annular wavefronts

    NARCIS (Netherlands)

    Mahajan, V.N.; Aftab, M.

    2010-01-01

    The theory of wavefront analysis of a noncircular wavefront is given and applied for a systematic comparison of the use of annular and Zernike circle polynomials for the analysis of an annular wavefront. It is shown that, unlike the annular coefficients, the circle coefficients generally change as t

  8. Review of consequences of uranium hydride formation in N-Reactor fuel elements stored in the K-Basins

    Energy Technology Data Exchange (ETDEWEB)

    Weber, J.W.

    1994-09-28

    The 105-K Basins on the Hanford site are used to store uranium fuel elements and assemblies irradiated in and discharged from N Reactor. The storage cylinders in KW Basin are known to have some broken N reactor fuel elements in which the exposed uranium is slowly reacting chemically with water in the cylinder. The products of these reactions are uranium oxide, hydrogen, and potentially some uranium hydride. The purpose of this report is to document the results f the latest review of potential, but highly unlikely accidents postulated to occur as closed cylinders containing N reactor fuel assemblies are opened under water in the KW basin and as a fuel assembly is raised from the basin in a shipping cask for transportation to the 327 Building for examination as part of the SNF Characterization Program. The postulated accidents reviews in this report are considered to bound all potential releases of radioactivity and hydrogen. These postulated accidents are: (1) opening and refill of a cylinder containing significant amounts of hydrogen and uranium hydride; and (2) draining of the single element can be used to keep the fuel element submerged in water after the cask containing the can and element is lifted from the KW Basin. Analysis shows the release of radioactivity to the site boundary is significantly less than that allowed by the K Basin Safety Evaluation. Analysis further shows there would be no damage to the K Basin structure nor would there be injury to personnel for credible events.

  9. Identification of a leaking TRIGA fuel element at the nuclear reactor facility of the University of Pavia

    International Nuclear Information System (INIS)

    During a periodical activity of characterization of the ionic-exchange resins of the demineralizer of the primary cooling circuit of the TRIGA Mark-2 reactor of the University of Pavia a small but detectable amount of Cs137 contamination was measured. Since the reactor has been running for several hundreds of hours at full power without showing any anomaly in the radiometric and thermo-hydraulic parameters, the reactor was brought to the nominal power of 250 kW for one hour and a sample of water was collected from the reactor tank and analysed in a low-background gamma ray detector. As a result a small amount of fission products were detected in the reactor pool water (a few Bq/g) suggesting the existence of a possible clad defect in one or more fuel elements. Since no halogens such as iodine and bromine were detected in the sampled water, the more likely hypothesis, also supported by literature, seemed to be a micro-fissure in the neck of an instrumented fuel element. A dedicated apparatus for reactor pool water sampling and on-line spectroscopy measurements was realized. As expected, the fission product leakage was due to a micro-fissure of a fuel element that released noble gas only when the fuel element was heated up to a temperature around 90 Celsius degrees. The leaking fuel element was identified and removed from its position and the reactor was back in regular operation after 2 months from leakage detection. (authors)

  10. Elemental balance of SRF production process: solid recovered fuel produced from municipal solid waste.

    Science.gov (United States)

    Nasrullah, Muhammad; Vainikka, Pasi; Hannula, Janne; Hurme, Markku; Oinas, Pekka

    2016-01-01

    In the production of solid recovered fuel (SRF), certain waste components have excessive influence on the quality of product. The proportion of rubber, plastic (hard) and certain textiles was found to be critical as to the elemental quality of SRF. The mass flow of rubber, plastic (hard) and textiles (to certain extent, especially synthetic textile) components from input waste stream into the output streams of SRF production was found to play the decisive role in defining the elemental quality of SRF. This paper presents the mass flow of polluting and potentially toxic elements (PTEs) in SRF production. The SRF was produced from municipal solid waste (MSW) through mechanical treatment (MT). The results showed that of the total input chlorine content to process, 55% was found in the SRF and 30% in reject material. Of the total input arsenic content, 30% was found in the SRF and 45% in fine fraction. In case of cadmium, lead and mercury, of their total input content to the process, 62%, 38% and 30%, respectively, was found in the SRF. Among the components of MSW, rubber material was identified as potential source of chlorine, containing 8.0 wt.% of chlorine. Plastic (hard) and textile components contained 1.6 and 1.1. wt.% of chlorine, respectively. Plastic (hard) contained higher lead and cadmium content compared with other waste components, i.e. 500 mg kg(-1) and 9.0 mg kg(-1), respectively.

  11. Time delay controlled annular array transducers for omnidirectional guided wave mode control in plate like structures

    International Nuclear Information System (INIS)

    Guided waves in plate like structures offer several modes with unique characteristics that can be taken advantage for nondestructive inspection applications. Conditions relating to the structure under inspection like the surrounding media, liquid loading, coatings etc require the use of special modes for successful inspection. Therefore, transducers that can excite mode controlled guided waves are essential for defect detection and discrimination in structures. Array transducers with annular elements can generate omnidirectional guided waves in plate like structures. However, the wave modes excited are limited to a particular wavelength governed by the element spacing. This limitation on the annular array transducers can be overcome by controlling the phase at each element relative to one another. In this work, annular array transducer construction techniques are theoretically examined and the optimum phase delays between the annular elements to excite a desired guided wave mode are calculated. A five element comb type annular array transducer is fabricated utilizing 1–3 type piezocomposite material. The mode control capability of the transducer is experimentally verified by selectively exciting the A0 and S0 guided wave modes in an aluminum plate like structure. (paper)

  12. Theoretical study to determine the heat transfer by forced convection coefficient in an empirical correlation in single phase, for annular channels

    International Nuclear Information System (INIS)

    In the heat transfer studies by forced convection, we have few data about behavior of the fluids in an annular channel heated by a concentric pipe, such date is necessary to know the heat transfer coefficient that establish the interchange of energy and the thermic properties of the fluid with the geometry of the flow. In this work the objective, was to compare some empirical correlations that we needed for determinate the heat transfer coefficient for annular channels, where we obtained similar at the theoretical results of an experiment made by Miller and Benforado. It is important to know such coefficients because we can determinate the heat quantity transmitted to a probe zone, in which we simulate a nuclear fuel element that developed huge heat quantity that must be dispersed in short time. We give theoretical data of the heat forced transfer convection and we analyzed the phenomena in annular channels given some empirical correlations employed by some investigators and we analyzed each one. (Author)

  13. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    Energy Technology Data Exchange (ETDEWEB)

    TOFFER, H.

    2006-07-18

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Fuel that had experienced a neutron environment in a reactor is known as spent, exposed, or irradiated fuel. In contrast fuel that has not yet been placed in a reactor is known as green, unexposed, or unirradiated fuel. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled (References 1 and 2) and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements (Reference 3). The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprised of two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with unirradiated fuel and one with irradiated fuel. Both the unirradiated and irradiated fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, three (3) green fuel

  14. Preparation for shipment of spent TRIGA fuel elements from the research reactor of the Medical University of Hannover

    International Nuclear Information System (INIS)

    In the early seventies a research reactor of type TRIGA Mark I was installed in the Department of Nuclear Medicine at the Medical University of Hannover (MHH) for the production of isotopes with short decay times for medical use. Since new production methods have been developed, the reactor has become obsolete and the MHH decided to decommission it. Probably in the second quarter of 1999 all 76 spent TRIGA fuel elements will be shipped to Idaho National Engineering and Environmental Laboratory (INEEL), USA, in one cask of type GNS 16. Due to technical reasons within the MHH a special Mobile Transfer System, which is being developed by the company Noell-KRC, will be used for reloading the fuel elements and transferring them from the reactor to the cask GNS 16. A description of the main components of this system as well as the process for transferring the fuel elements follows. (author)

  15. Preparation for shipment of spent TRIGA fuel elements from the research reactor of the Medical University of Hannover

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, Gabriele; Cordes, Harro [Medical University of Hannover, D-30625 Hannover (Germany); Ebbinghaus, Kurt; Haferkamp, Dirk [NOELL-KRC, D-97064 Wuerzburg (Germany)

    1998-07-01

    In the early seventies a research reactor of type TRIGA Mark I was installed in the Department of Nuclear Medicine at the Medical University of Hannover (MHH) for the production of isotopes with short decay times for medical use. Since new production methods have been developed, the reactor has become obsolete and the MHH decided to decommission it. Probably in the second quarter of 1999 all 76 spent TRIGA fuel elements will be shipped to Idaho National Engineering and Environmental Laboratory (INEEL), USA, in one cask of type GNS 16. Due to technical reasons within the MHH a special Mobile Transfer System, which is being developed by the company Noell-KRC, will be used for reloading the fuel elements and transferring them from the reactor to the cask GNS 16. A description of the main components of this system as well as the process for transferring the fuel elements follows. (author)

  16. Confocal Annular Josephson Tunnel Junctions

    Science.gov (United States)

    Monaco, Roberto

    2016-04-01

    The physics of Josephson tunnel junctions drastically depends on their geometrical configurations and here we show that also tiny geometrical details play a determinant role. More specifically, we develop the theory of short and long annular Josephson tunnel junctions delimited by two confocal ellipses. The behavior of a circular annular Josephson tunnel junction is then seen to be simply a special case of the above result. For junctions having a normalized perimeter less than one, the threshold curves in the presence of an in-plane magnetic field of arbitrary orientations are derived and computed even in the case with trapped Josephson vortices. For longer junctions, a numerical analysis is carried out after the derivation of the appropriate motion equation for the Josephson phase. We found that the system is modeled by a modified and perturbed sine-Gordon equation with a space-dependent effective Josephson penetration length inversely proportional to the local junction width. Both the fluxon statics and dynamics are deeply affected by the non-uniform annulus width. Static zero-field multiple-fluxon solutions exist even in the presence of a large bias current. The tangential velocity of a traveling fluxon is not determined by the balance between the driving and drag forces due to the dissipative losses. Furthermore, the fluxon motion is characterized by a strong radial inward acceleration which causes electromagnetic radiation concentrated at the ellipse equatorial points.

  17. Confocal Annular Josephson Tunnel Junctions

    Science.gov (United States)

    Monaco, Roberto

    2016-09-01

    The physics of Josephson tunnel junctions drastically depends on their geometrical configurations and here we show that also tiny geometrical details play a determinant role. More specifically, we develop the theory of short and long annular Josephson tunnel junctions delimited by two confocal ellipses. The behavior of a circular annular Josephson tunnel junction is then seen to be simply a special case of the above result. For junctions having a normalized perimeter less than one, the threshold curves in the presence of an in-plane magnetic field of arbitrary orientations are derived and computed even in the case with trapped Josephson vortices. For longer junctions, a numerical analysis is carried out after the derivation of the appropriate motion equation for the Josephson phase. We found that the system is modeled by a modified and perturbed sine-Gordon equation with a space-dependent effective Josephson penetration length inversely proportional to the local junction width. Both the fluxon statics and dynamics are deeply affected by the non-uniform annulus width. Static zero-field multiple-fluxon solutions exist even in the presence of a large bias current. The tangential velocity of a traveling fluxon is not determined by the balance between the driving and drag forces due to the dissipative losses. Furthermore, the fluxon motion is characterized by a strong radial inward acceleration which causes electromagnetic radiation concentrated at the ellipse equatorial points.

  18. Production Cycle for Large Scale Fission Mo-99 Separation by the Processing of Irradiated LEU Uranium Silicide Fuel Element Targets

    OpenAIRE

    Abdel-Hadi Ali Sameh

    2013-01-01

    Uranium silicide fuels proved over decades their exceptional qualification for the operation of higher flux material testing reactors with LEU elements. The application of such fuels as target materials, particularly for the large scale fission Mo-99 producers, offers an efficient and economical solution for the related facilities. The realization of such aim demands the introduction of a suitable dissolution process for the applied U3Si2 compound. Excellent results are achieved by the oxidiz...

  19. Characterisation of nuclear fuel element welds of stainless steel and zircaloy made by GTAW, EBW and LBW techniques

    International Nuclear Information System (INIS)

    Fuel elements are fabricated by welding the end plugs to the fuel tubes. GTAW is widely practiced in view of its simplicity and low cost. EBW and LBW techniques are being developed because of better process control and flexibility. In this paper we have studied the welds made by these techniques. The study includes optical and scanning electron microscopy, microhardness measurements and EPMA of these welds

  20. RA-3 reactor core with uranium silicide fuel elements P-07 type

    International Nuclear Information System (INIS)

    Following the studies on the utilization of fuel elements (FE) containing uranium silicide, core of the RA-3 was analyzed with several calculation models. At first, the present situation, i.e. the core charged with normal FE (U3O8), has been analyzed to validate the simulation methodology comparing with experimental results and to establish reference data to 5 and 10 MW able to be compared with future new situations. Also, CITVAP's nuclear data libraries to be used in irradiation experiment planning were completed. The results were satisfactory and were applied to the study of the core containing P-07 FE [U3Si2], in face of a future core change. Comparing with the performance of the U3O8FE, the silicides ones show the following advantages: - average burnup: 45 % greater; -extraction burnup increase 12 %; and, -the residence time [in full power days] could be a 117 % greater. (author)

  1. Comparison of ASSERT-4 CHF predictions with measurements in 28-element simulated fuel string

    International Nuclear Information System (INIS)

    The subchannel analysis code ASSERT-4 was used to model CHF experiments with a 28-element simulated fuel string carried out by Ontario Hydro at Columbia University. ASSERT was run with eleven CHF test conditions to determine dryout power using each of the five CHF models available in the code. The location of first dryout and the spreading of dryout with increasing power were also investigated. The five CHF models used in ASSERT are the Whalley mechanistic model; W-3, Bowring and Biasi correlations; and table lookup. Of these, implementation in ASSERT-4 of the Biasi correlation gives the most accurate dryout power predictions. Predictions using the table lookup method and the Whalley model are almost equally good, while those using the W-3 correlation and, especially, the Bowring correlation are inferior. In predicting dryout spreading patterns, use of the Whalley model comes the closest to adequately identifying the actual dryout location

  2. Internal flow measurements of the SSME fuel preburner injector element using real time neutron radiography

    Science.gov (United States)

    Lindsay, John T.; Elam, Sandy; Koblish, Ted; Lee, Phil; Mcauliffe, Dave

    1990-01-01

    Due to observations of unsteady flow in the Space Shuttle Main Engine fuel preburner injector element, several flow studies have been performed. Real time neutron radiography tests were recently completed. This technique provided real time images of MiL-c-7024 and Freon-22 flow through an aluminum liquid oxygen post model at three back pressures (0, 150, and 545 psig) and pressure drops up to 1000 psid. Separated flow appeared only while operating at back pressures of 0 and 150 psig. The behavior of separated flow was similar to that observed for water in a 3x acrylic model of the LOX post. On the average, separated flow appeared to reattach near the exit of the post when the ratio of pressure drop to supply pressure was about 0.75.

  3. Research on graphite powders used for HTR-PM fuel elements

    Institute of Scientific and Technical Information of China (English)

    ZHAO Hongsheng; LIANG Tongxiang; ZHANG Jie; LI Ziqiang; TANG Chunhe

    2006-01-01

    Different batches of natural graphite powders and electrographite powders were characterized by impurity, degree of graphitization, particle size distribution, specific surface area, and shape characteristics. The graphite balls consist of proper mix-ratio of natural graphite, electrographite and phenolic resin were manufactured and characterized by thermal conductivity, anisotropy of thermal expansion, crush strength, and drop strength. Results show that some types of graphite powders possess very high purity, degree of graphitization, and sound size distribution and apparent density, which can serve for matrix graphite of HTR-PM. The graphite balls manufactured with reasonable mix-ratio of graphite powders and process method show very good properties. It is indicated that the properties of graphite balls can meet the design criterion of HTR-PM. We can provide a powerful candidate material for the future manufacture of HTR-PM fuel elements.

  4. An experimental investigation of the temperature behavior of a CANDU 37-element spent fuel bundle with air backfill

    International Nuclear Information System (INIS)

    As part of the thermal analysis of a CANDU spent fuel dry storage system, a series of experiment has been conducted using a thermal mock-up of a simulated CANDU spent fuel bundle in a dry storage basket. The experimental system was designed to obtain the maximum fuel rod temperature along with the radial and axial temperature distributions within the fuel bundle. The main purpose of these experiments was to characterize the relevant heat transfer mechanisms in a dry, vertically oriented CANDU spent fuel bundle, and to verify the MAXROT code developed for the thermal analysis of a CANDU spent fuel bundle in a dry storage basket. A total of 48 runs were made with 8 different power inputs to the 37-element heater rod bundle ranging from 5 to 40 W, while using 6 different band heaters power inputs from 0 to 250 W to maintain the basket wall at a desired boundary condition temperature at the steady state. The temperature distribution in a heater rod bundle was measured and recorded at the saturated condition for each set of heater rod power and band heaters power. To characterize the heat transfer mechanism involved, the experimental data were corrected analytically for radiation heat transfer and presented as a Nusselt number correlation in terms of the Rayleigh number of the heater rod bundle. The results show that the Nusselt number remains nearly constant and all the experimental dada fall within a conduction regime. The experimental data were compared with the predictions of the MAXROT code to examine the code's accuracy and validity of assumptions used in the code. The MAXROT code explicitly models each representative fuel rod in a CANDU fuel bundle and couples the conductive and radiative heat transfer of the internal gas between rods. Comparisons between the measured and predicted maximum fuel rod temperatures of the simulated CANDU 37-element spent fuel bundle for all 48 tests show that the MAXROT code slightly over-predicts and the agreement is within 2

  5. Study of neutronic effects of water hole in the control fuel element in the Tehran research reactor

    International Nuclear Information System (INIS)

    Due to the existence of some fission products in the core outlet of the Tehran Research Reactor, fuel leakage tests were done on all fuel elements. These experiments showed that the fission products come out mostly from some of the fuel control elements. So the failures of fuel control elements were taken out of the core and reactor started with the new core configuration. Failure of fuel plate in pool type research reactors resulting in a release of fission products to the pool water fall into the following categories: 1. Blistering or swelling due to excessive fission gas or fission products build-up. 2. Failure due to corrosion (chemical-metallurgical, ...). 3. Manufacturing defect such as non-homogeneous fuel distribution. 4. Coolant channel blockage by foreign materials. From neutronic point of view, existance of water hole in the CFE causes increasing thermal flux on the fuel plate adjusted to the water hole. Therefore, there is a probability of blistering or swelling in some parts of the plates. For more studies of this effect, we did neutronic calculations using computer codes. Calculation also showed that the fission rate in the above mentioned plate in the CFE is more than other plates in the core. Therefore blistering or swelling due to excessive fission gas or fission products build-up in some part of the plate is probable. As it was mentioned before, the probability of chemical and metallurgical corrosion on the fuel clad should not be ignored. However, since the release of fission products from some of the CFE were more than other fuel assemblies, the probability of occuring cladding defects by the neutronic effects of the water hole is stronger than other effects. (Author)

  6. Mechanistic modeling of Zircaloy deformation and fracture in fuel element analysis

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, F.A.

    1985-06-01

    A review is given of the comprehensive model developed in the 1960's at the Bettis Atomic Power Laboratory to explain the creep of Zircaloy during neutron irradiation and applied to fuel element analysis and design. The in-pile softening observed at low stresses was hypothesized to be due to a combination of the growth-directed Roberts-Cottrell yielding creep originally proposed for ..cap alpha..-uranium and the formation of point defect loops preferentially on certain planes in response to the applied stress, with the second process being of relatively greater importance. The in-pile hardening observed at high stresses (or strain-rates) was proposed to be due to the cutting by dislocations of radiation-produced obstacles. In this stress (strain-rate) region, in-pile behavior was proposed to be identical to post-irradiation behavior. At intermediate stresses (strain-rates) a mechanism of radiation-enhanced climb around obstacles was suggested as being rate controlling. As the stress is decreased, the climb process becomes easier and the rate was then predicted to be controlled by glide at a flow stress characteristic of unirradiated, annealed material, where radiation-enhanced diffusion enabled climbing around the normal strain-hardening obstacles. At still lower stresses, this glide process became negligibly slow compared with the growth-connected creep mechanism which was presumed to operate independently. The overall scheme was shown to be good agreement with all the in-pile data then available and implemented into the computer analysis of fuel element behavior. 48 refs., 1 fig.

  7. German physical protection concept for the storage of spent fuel elements in transport and storage casks

    International Nuclear Information System (INIS)

    Full text: In Germany, the legal regulations and requirements derived from rules and guidelines for the protection of storage facilities for spent fuel elements from disruptive action or other inference by third parties are structured hierarchically. The Atomic Energy Act constitutes the top level. It is supported by federal ordinances. The next level down is formed by the rules and guidelines. The storage of nuclear fuels may only be authorized, according to the provisions of the Atomic Energy Act, if the required protection from disruptive action or other interference by third parties can be guaranteed as following: it must be possible to prevent any danger to life and health due to a substantial amount of direct radiation or due to the release of a substantial amount of radioactive material; it must be possible to prevent singular or repeated acts of stealing nuclear fuels in such amounts that a critical accumulation can be produced directly without reprocessing and enrichment. Knowing that nuclear installations cannot be protected from every possible interference, physical protection is focused on basic security standards, the so-called design basic threat (DBT), departing from the assumed interference. DBT is regularly reviewed by the competent federal authorities and authorities of the states and are revised on the basis of newly gained knowledge, if necessary, such as in the wake of the terrorist attacks in the U.S. on September 11, 2001. The operator must guarantee and give proof of a sufficient level of physical protection of the plant. The sole physical protection measures implemented by the operator cannot ensure the required protection from other interference by third parties for an unlimited time span. The concept therefore requires additional physical protection measures by the police. (author)

  8. Bistability and hysteresis of annular impinging jets

    Science.gov (United States)

    Tisovsky, Tomas

    2016-06-01

    In present study, the bistability and hysteresis of annular impinging jets is investigated. Annular impinging jets are simulated using open source CFD code - OpenFOAM. Both flow field patterns of interest are obtained and hysteresis is found by means of dynamic mesh simulation. Effect of nozzle exit velocity on resulting hysteresis loop is also illustrated.

  9. Soliton bunching in annular Josephson junctions

    DEFF Research Database (Denmark)

    Vernik, I.V; Lazarides, Nickos; Sørensen, Mads Peter;

    1996-01-01

    By studying soliton (fluxon) motion in long annular Josephson junctions it is possible to avoid the influence of the boundaries and soliton-soliton collisions present in linear junctions. A new experimental design consisting of a niobium coil placed on top of an annular junction has been used...

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

    International Nuclear Information System (INIS)

    According to the present state of development, full size test fuel elements with UAl/sub x/, U3O8, and U3Si2 fuel were fabricated at Nukem in production scale. The maximum uranium densities amount to 1.8 g/cm3 for UAl/sub x/, 2.9 g/cm3 for U3O8, and 4.76 g/cm3 for U3Si2. The irradiation performance of these fuel elements is good: Up to the end of September 1982 the following burnups were achieved: 73% with UAl/sub x/, 60% with U3O8, 39% with U3Si2; no defects could be detected. For an economical fuel element production with reduced 235U enrichment chemical uranium recycling methods were developed allowing immediate scrap recovery at minimum waste generation. In addition test plates with UAl/sub x/ and U3O8 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 UAl/sub x/, and 3.14 g/cc for U3O8 fuel. Apart from plates with standard geometry also plates with increased meat thickness were inserted

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

    International Nuclear Information System (INIS)

    According to the present state of development full size test fuel elements with UAlx, U3O8, and U3Si2 fuel were fabricated at Nukem in production scale. The maximum uranium densities amount to 1.8 g/cc for UAIx, 2.9 g/cc for U3O8, and 4.76 g/cc for U3Si2. The irradiation performance of these fuel elements is good: Up to the end of September 1982 the following burnups were achieved: 73% with UA1x, 60% with U3O8, 39% with U3Si2; 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 UAlx and U3O8 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 U3O8 fuel. Apart from plates with standard geometry also plates with increased meat thickness were inserted. (author)

  12. Formation of intermetallic compound at interface between rare earth elements and ferritic-martensitic steel by fuel cladding chemical interaction

    Institute of Scientific and Technical Information of China (English)

    Jun Hwan Kim; Byoung Oon Lee; Chan Bock Lee; Seung Hyun Jee; Young Soo Yoon

    2012-01-01

    The intermetallic compounds formation at interface between rare earth elements and clad material were investigated to demonstrate the effects of rare earth elements on fuel-cladding chemical interaction (FCCI) behavior.Mischmetal (70Ce-30La) and Nd were prepared as rare earth elements.Diffusion couple testing was performed on the rare earth elements and cladding (9Cr2W steel) near the operation temperature of(sodium-cooled fast reactor) SFR fuel.The performance of a diffusion barrier consisting of Zr and V metallic foil against the rare earth elements was also evaluated.Our results showed that Ce and Nd in the rare earth elements and Fe in the clad material interdiffused and reacted to form intermetallic species according to the parabolic rate law,describing the migration of the rare earth element.The diffusion of Fe limited the reaction progress such that the entire process was governed by the cubic rate law.Rare earth materials could be used as a surrogate for high burnup metallic fuels,and the performance of the barrier material was demonstrated to be effective.

  13. Corrosion product deposition on fuel element surfaces of a boiling water reactor

    International Nuclear Information System (INIS)

    Over the last decade the problem of corrosion products deposition on light water reactor fuel elements has been extensively investigated in relation to the possibility of failures caused by them. The goal of the present study is to understand in a quantitative way the formation of such kind of deposits and to analytically understand the mechanism of formation and deposition with help of the quasi-steady state concentrations of a number of 3d metals in reactor water. Recent investigations on the complex corrosion product deposits on a Boiling Water Reactor (BWR) fuel cladding have shown that the observed layer locally presents unexpected magnetic properties. The buildup of magnetic corrosion product deposits (crud) on the fuel cladding of the BWR, Kernkraftwerk Leibstadt (KKL) Switzerland has hampered the Eddy-current based measurements of ZrO2 layer thickness. The magnetic behavior of this layer and its axial variation on BWR fuel cladding is of interest with respect to non-destructive cladding characterization. Consequently, a cladding from a BWR was cut at elevations of 810 mm, where the layer was observed to be magnetic, and of 1810 mm where it was less magnetic. The samples were subsequently analyzed using electron probe microanalysis (EPMA), magnetic analysis and X-ray techniques (μXRF, μXRD and μXAFS). Both EPMA and μXRF have shown that the observed corrosion deposit layer which is situated on the Zircaloy corrosion layer consists mostly of 3-d elements’ oxides (Fe, Zn, Ni and Mn). The distribution of these elements within the investigated layer is rather complex and not homogeneous. The main components identified by 2D μXRD mapping inside the layer were hematite and spinel phases with the common formula (MxFey)[M(1-x)Fe(2-y)]O4, where M = Zn, Ni, Mn. With μXRD it was clearly shown that the cell parameter of analyzed spinel is different from the one of the pure endmembers (ZnFe2O4, NiFe2O4 and MnFe2O4) proving the existence of solid solutions. These

  14. Release to the Gas Phase of Inorganic Elements during Wood Combustion. Part 2: Influence of Fuel Composition

    DEFF Research Database (Denmark)

    van Lith, Simone Cornelia; Jensen, Peter Arendt; Frandsen, Flemming;

    2008-01-01

    Combustion of wood for heat and power production may cause problems such as ash deposition, corrosion, and harmful emissions of gases and particulate matter. These problems are all directly related to the release of inorganic elements (in particular Cl, S, K, Na, Zn, and Pb) from the fuel...... temperatures in the range of 500–1150 °C in a laboratory-scale tube reactor and by performing mass balance calculations based on the weight measurements and chemical analyses of the wood fuels and the residual ash samples. Four wood fuels with different ash contents and inorganic compositions were investigated...

  15. Investigation of coolant thermal mixing within 28-element CANDU fuel bundles using the ASSERT-PV thermal hydraulics code

    International Nuclear Information System (INIS)

    This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles under steady-state conditions. The study, which is based on simulations performed using the ASSERT-PV thermal hydraulic code, consists of two main parts. In the first part the various physical mechanisms that contribute to coolant mixing are identified and their impact is isolated via ASSERT-PV simulations. The second part is concerned with development of a preliminary model suitable for use in the fuel and fuel channel code FACTAR to predict the thermal mixing that occurs between flow annuli. (author)

  16. Optimum annular focusing by a phase plate

    CERN Document Server

    Arrizón, Victor; Aguirre-Olivas, Dilia; Mellado-Villaseñor, Gabriel

    2015-01-01

    Conventional light focusing, i. e. concentration of an extended optical field within a small area around a point, is a frequently used process in Optics. An important extension to conventional focusing is the generation of the annular focal field of an optical beam. We discuss a simple optical setup that achieves this kind of focusing employing a phase plate as unique optical component. We first establish the class of beams that being transmitted through the phase plate can be focused into an annular field with topological charge of arbitrary integer order q. Then, for each beam in this class we determine the plate transmittance that generates the focal field with the maximum possible peak intensity. In particular, we discuss and implement experimentally the optimum annular focusing of a Gaussian beam. The attributes of optimum annular focal fields, namely the high peak intensity, intensity gradient and narrow annular section, are advantageous for different applications of such structured fields.

  17. Fuel injection and mixing systems having piezoelectric elements and methods of using the same

    Science.gov (United States)

    Mao, Chien-Pei; Short, John; Klemm, Jim; Abbott, Royce; Overman, Nick; Pack, Spencer; Winebrenner, Audra

    2011-12-13

    A fuel injection and mixing system is provided that is suitable for use with various types of fuel reformers. Preferably, the system includes a piezoelectric injector for delivering atomized fuel, a gas swirler, such as a steam swirler and/or an air swirler, a mixing chamber and a flow mixing device. The system utilizes ultrasonic vibrations to achieve fuel atomization. The fuel injection and mixing system can be used with a variety of fuel reformers and fuel cells, such as SOFC fuel cells.

  18. Experimental approach and modelling of the mechanical behaviour of graphite fuel elements subjected to compression pulses

    Science.gov (United States)

    Forquin, P.

    2010-06-01

    Among the activities led by the Generation IV International Forum (GIF) relative to the future nuclear systems, the improvement of recycling of fuel elements and their components is a major issue. One of the studied systems by the GIF is the graphite-moderated high-temperature gas cooled reactor (HTGR). The fuel elements are composed of fuel roads half-inch in diameter named compacts. The compacts contain spherical particles made of actinide kernels about 500 m in diameter coated with three layers of carbon and silicon carbide, each about 50 m thick, dispersed in a graphite matrix. Recycling of compacts requires first a separation of triso-particles from the graphite matrix and secondly, the separation of the triso-coating from the kernels. This aim may be achieved by using pulsed currents: the compacts are placed within a cell filled by water and exposed to high voltage between 200 - 500 kV and discharge currents from 10 to 20 kA during short laps of time (about 2 µs) [1-2]. This repeated treatment leads to a progressive fragmentation of the graphite matrix and a disassembly of the compacts. In order to improve understanding of the fragmentation properties of compacts a series of quasi-static and dynamic experiments have been conducted with similar cylindrical samples containing 10% (volume fraction) of SiC particles coated in a graphite matrix. First, quasi-static compression tests have been performed to identify the mechanical behaviour of the material at low strain-rates (Fig.1). The experiments reveal a complex elasto-visco-plastic behaviour before a brittle failure. The mechanical response is characterised by a low yield stress (about 1 MPa), a strong strain-hardening in the loading phase and marked hysteresis-loops during unloading-reloading stages. Brittle failure is observed for axial stress about 13 MPa. In parallel, a series of flexural tests have been performed with the aim to characterise the quasi-static tensile strength of the particulate

  19. Experimental approach and modelling of the mechanical behaviour of graphite fuel elements subjected to compression pulses

    Directory of Open Access Journals (Sweden)

    Forquin P.

    2010-06-01

    Full Text Available Among the activities led by the Generation IV International Forum (GIF relative to the future nuclear systems, the improvement of recycling of fuel elements and their components is a major issue. One of the studied systems by the GIF is the graphite-moderated high-temperature gas cooled reactor (HTGR. The fuel elements are composed of fuel roads half-inch in diameter named compacts. The compacts contain spherical particles made of actinide kernels about 500 m in diameter coated with three layers of carbon and silicon carbide, each about 50 m thick, dispersed in a graphite matrix. Recycling of compacts requires first a separation of triso-particles from the graphite matrix and secondly, the separation of the triso-coating from the kernels. This aim may be achieved by using pulsed currents: the compacts are placed within a cell filled by water and exposed to high voltage between 200 – 500 kV and discharge currents from 10 to 20 kA during short laps of time (about 2 µs [1-2]. This repeated treatment leads to a progressive fragmentation of the graphite matrix and a disassembly of the compacts. In order to improve understanding of the fragmentation properties of compacts a series of quasi-static and dynamic experiments have been conducted with similar cylindrical samples containing 10% (volume fraction of SiC particles coated in a graphite matrix. First, quasi-static compression tests have been performed to identify the mechanical behaviour of the material at low strain-rates (Fig.1. The experiments reveal a complex elasto-visco-plastic behaviour before a brittle failure. The mechanical response is characterised by a low yield stress (about 1 MPa, a strong strain-hardening in the loading phase and marked hysteresis-loops during unloading-reloading stages. Brittle failure is observed for axial stress about 13 MPa. In parallel, a series of flexural tests have been performed with the aim to characterise the quasi-static tensile strength of the

  20. Annular burnout data from rod bundle experiments

    International Nuclear Information System (INIS)

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident. Level average fluid conditions within the test section were calculated using steady-state mass and energy conservation considerations for the steady-state tests and a transient, homogeneous, equilibrium computer code for the transient tests. Unlike tube dryout, burnout within a rod bundle does not necessarily occur at one distinct axial level. The location of individual rod dryout was determined by scanning rods axially and locating the position where rod superheat increased from approx. =0 to 30 K or greater. Thermocouple instrumentation within the bundle allows the location of dryout to be determined to within approximately +.5 cm for many of the tests

  1. [Generalized granuloma annulare or diffuse dermal histiocytosis?].

    Science.gov (United States)

    Kretzschmar, L; Biel, K; Luger, T A; Goerdt, S

    1995-08-01

    Generalized granuloma annulare is a rare variant of granuloma annulare affecting the trunk and extremities with a multitude of lesions. In contrast to localized granuloma annulare, generalized granuloma annulare occurs in older patients, shows a stronger association with diabetes, and is characteristically chronic. Like our 55-year-old patient, most patients present with papules and annular plaques; less often, macular or non-annular lesions may be encountered. Histology often fails to show necrobiotic or necrotic connective tissue changes demarcated by a palisading granuloma. Instead, there are diffuse dermal, band-like or nodular aggregations of histiocytes intermingled with some multinucleated giant cells and a predominantly lymphocytic infiltrate in the periphery. Because of its special characteristics, it has been suggested that generalized granuloma annulare might constitute a separate disease entity and that it should be classed among the primary cutaneous histiocytoses as a diffuse dermal histiocytosis. Using immunohistochemistry to determine the macrophage phenotype of the lesional histiocytes, we have shown that generalized granuloma annulare is not a cutaneous histiocytosis. Neither MS-1 high-molecular-weight protein, a new specific marker for cutaneous non-Langerhans cell histiocytoses, nor CD1a, the well-known marker for Langerhans cells and Langerhans cell histiocytoses, is expressed by the lesional histiocytes of our patient. In contrast, the antigen expression pattern was diagnostic for non-infectious granulomas and was highly similar to that in localized granuloma annulare. In contrast to the successful treatment of localized granuloma annulare reported with intralesional interferon beta-1, systemic treatment with interferon alpha-2b (9 x 10(6) units three times a week) was ineffective.

  2. 3D laser inspection of fuel assembly grid spacers for nuclear reactors based on diffractive optical elements

    Science.gov (United States)

    Finogenov, L. V.; Lemeshko, Yu A.; Zav'yalov, P. S.; Chugui, Yu V.

    2007-06-01

    Ensuring the safety and high operation reliability of nuclear reactors takes 100% inspection of geometrical parameters of fuel assemblies, which include the grid spacers performed as a cellular structure with fuel elements. The required grid spacer geometry of assembly in the transverse and longitudinal cross sections is extremely important for maintaining the necessary heat regime. A universal method for 3D grid spacer inspection using a diffractive optical element (DOE), which generates as the structural illumination a multiple-ring pattern on the inner surface of a grid spacer cell, is investigated. Using some DOEs one can inspect the nomenclature of all produced grids. A special objective has been developed for forming the inner surface cell image. The problems of diffractive elements synthesis, projecting optics calculation, adjusting methods as well as calibration of the experimental measuring system are considered. The algorithms for image processing for different constructive elements of grids (cell, channel hole, outer grid spacer rim) and the experimental results are presented.

  3. Space Nuclear Thermal Propulsion Nuclear Element Tests at Sandia National Laboratories

    International Nuclear Information System (INIS)

    Nuclear Element Tests (NET) are being performed as part of the U.S. Air Force Space Nuclear Thermal Propulsion (SNTP) Program to evaluate high performance fuel elements intended for use in future nuclear propulsion systems. The NET experiments are to be performed at the Sandia National Laboratories (SNL's) Annular Core Research Reactor (ACRR). Objectives of these experiments are to provide engineering validation and demonstration of critical-fuel-element-related technologies and an experimental data base to support analytical design methods for the SNTP Program. Currently, hardware for the first two fueled NET experiments has been fabricated, and cold flow tests have been accomplished with a representative set of hardware to assure the experimental capability to achieve test objectives in-reactor. Assembly of the first NET experiment to test a representative nuclear fuel element is in progress, and planned operational sequences have been defined

  4. The Influence of Distance and Atmospheric Elements on the Concentration of Odour from Refuse Derived Fuel (RDF) Operations

    OpenAIRE

    Zaini Sakawi; Lukman Ismail; Noor Khafazilah Abdullah; S. A. Sharifah Mastura; Othman Jaafar

    2013-01-01

    Odour is an environmental element that occurs as varieties of aroma, either pleasant or otherwise to its immediate community. The various sources of odour pollution may come from either natural or of human activities. Odour concentration may change due to environmental factors such as atmosphere, topography, distance and mitigation efforts. This study describes a study on the influence of distance and athmospheric elements on concentration of odour generated by the Refuse Derived Fuel (RDF) o...

  5. Process for transporting, intermediate storage and final storage of burnt-up fuel elements, container systems for implementing the process and manufacture of a container for the transport and/or storage of burnt-up fuel elements

    International Nuclear Information System (INIS)

    A container for the transport and/or storage of burnt-up fuel elements is manufactured using a container jacket material having shielding and good thermal conduction properties, which is a hydraulically binding dry mass, which consists of non-corroding components resistant to the acid vapours and solutions occurring in the salt mines used for final storage. (orig./HP)

  6. Post irradiation examination of thermal reactor fuels

    Science.gov (United States)

    Sah, D. N.; Viswanathan, U. K.; Ramadasan, E.; Unnikrishnan, K.; Anantharaman, S.

    2008-12-01

    The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O 2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16 000 MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO 2 fuel bundles from Indian PHWRs, which included two high burnup (˜15 000 MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO 2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ˜11 000 MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29 000 MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO 2, enriched UO 2, MOX, and ThO 2 fuels.

  7. Annular burnout data from rod-bundle experiments

    International Nuclear Information System (INIS)

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident

  8. Evaluation of vacuum-canned and hot-press fuel elements PT IP-44A and IP-45A

    Energy Technology Data Exchange (ETDEWEB)

    Smith, E.A.

    1958-10-15

    As new designs, geometries, and procedures for fabricating experimental fuel elements are developed, it becomes necessary to determine how well each new product fills the need which prompted its development. While the criteria for satisfactoriness vary with the type of fuel element, there is one criterion that is common to all types designed for use in heterogeneous reactors: the elements must withstand irradiation in the reactor for the required period without jacket rupture or excessive change in dimensions. To learn whether a new type of element meets this and other requirements, a statistically significant number of representative samples must be subjected to, first, comprehensive screening tests designed to insure trouble-free performance in the reactor; next, a long-term irradiation exposure equivalent to or exceeding that which will be normal for that type of element; and finally, a thorough examination to determine the extent of quality loss resulting from the irradiation. This report discusses the application of the foregoing philosophy in the evaluation of two lots of experimental fuel elements fabricated by unconventional procedures.

  9. Fuel element development committee: Annual report from the General Electric Company, Hanford

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M.; Minor, J.E.; Stringer, J.T.

    1964-08-14

    A summary of HAPO activities is given to include separate sections on the N-Reactor and other current production reactors. Specific programs and fuel performance for current production reactor fuels is discussed. Also, the production status, fuel performance, development program and process technology for N-Reactor fuels is presented.

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

    International Nuclear Information System (INIS)

    Within the framework of the RERTR Program and supported by the technical cooperation work agreed upon between the U.S.A. and Argentina in May 1979, the CNEA Nuclear Fuel Department - Low Enriched Fuel Elements Project (ECBE Project), has carried on its own program for developing fuels with low enrichment for research and test reactors. Up to the present, its main objective has been to replace the highly enriched fuel used in its only reactor (RA-3) for research, development and radioisotopes production. The basic stages of the Argentine Program are shown in Table 1. At a meeting held in Vienna in March, 1980, the CNEA stated that its development of fuels with low enrichment would be in two fuel lines: UAlx-Al and U3O8-Al, and that its aim would be to reach uranium densities of 18-2.2 g/cm3 for the UAIx-Al line and 2.4-3.0 g/cm3 for the U3O8 line. At the international meeting held at ANL in November, 1980, and after having received depleted uranium and uranium with 20% and 45% enrichment (purchased from the U.S.A. for manufacturing miniplates and possible standard fuels) to carry on the proposed development, CNEA anticipated -- after its first tests -- that the conditions were satisfactory for reaching uranium densities of 2.4-3.0 g/cm3 in U3O8-Al fuel and of 2.4 g/cm3 in UAIx-Al fuel. In February 1981, after Argentina accepted the obligation of paying for the irradiation service, authorization was obtained for irradiating miniplates in the Oak Ridge Reactor within the RERTR Program. In June 1981, the first set of miniplates was sent to Oak Ridge National Laboratory (ORNL). The maximum actual densities reached at that time were 3.12 g/cm3 with U3O8-Al and 2.52 g/cm3 with UAlx-Al. During a visit of the CNEA Project Technical Manager to the Argonne National Laboratory (ANL) in July 1981, and after exchanging ideas with ANL professional staff, the CNEA decided to incorporate a new line of development, that of U3Si-Al. Three months later, in October 1981 a second set of

  11. Methods for analysis of uranium-plutonium mixed fuel and transplutonium elements at RIAR (Preprint no. IT-25)

    International Nuclear Information System (INIS)

    Different methods for analysis of the uranium-plutonium mixed nuclear fuel and transplutonium elements are briefly discussed in this paper: coulometry, radiometric techniques, emission spectrography, mass-spectrometry, chromatography, spectrophotometry. The main analytical characteristics of the methods developed are given. (author). 30 refs., 2 tabs

  12. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Science.gov (United States)

    2010-01-01

    ... Equipment and Components Under NRC's Export Licensing Authority O Appendix O to Part 110 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Pt. 110, App. O... NRC's Export Licensing Authority Note: Nuclear fuel elements are manufactured from source or...

  13. Authorisation to irradiate two U3Si2 prototype fuel elements in the RA-3 research reactor

    International Nuclear Information System (INIS)

    The Argentine RA-3 research reactor (with U3O8-Al LEU MTR fuel type elements and in the process to increase its power from 5 MW to 10 MW since October 2002) is operated by the National Atomic Energy Commission (CNEA). After evaluating the safety studies submitted by CNEA, the Argentine Nuclear Regulatory Authority (ARN) authorised the irradiation of two U3Si2-Al prototype fuel elements named P-06 and P-07, in the RA-3, in September 2000 and September 2001 respectively. These irradiations are part of the qualification programme as manufacturer (CNEA). This report contains a brief description of the regulatory process the ARN carries out in order to approve major modifications or new experiments in research reactors; it also summarises conditions in force for operable core configurations, considers a definition of prototype fuel element and lists the main characteristics of the RA-3 research reactor. Then, the process involved in the authorisation to irradiate the prototypes P-06 and P-07 is particularly analysed. Finally, and in case an authorisation to irradiate a prototype fuel element of a non-qualified material (such as U-Mo) in the RA-3 is requested, the regulatory aspects that should be taken into consideration are analysed. This would be the first time a question of this nature is submitted to the ARN. (author)

  14. Structural dynamics studies for single and clustered SNR-300 fuel elements: a comparison of analytical and experimental results

    International Nuclear Information System (INIS)

    Structural damage may occur in liquid metal cooled fast breeder reactors (LMFBR) when local failures escalate to a fast release of excessive thermal energy within one fuel element. Time scales (5 to 50. msec) and peaks of associated pressure pulses (50. to 500. bar) may exceed the burst pressure of the hexagonal wrapper containing the pin-bundle. Though after fuel rupture the pressure level in the vicinity is markedly reduced, still a significant pressure field is propagating through the core. Therefore surrounding subassemblies or control rods are exposed to impulsive differential pressures causing duct flexure as well as local cross section deformation. A theoretical and experimental research program at GfK Karlsruhe investigates the mechanical effects of conservative pressure transients on LMFBR fuel elements. As part of the program this paper describes two computer codes (BEDYN-2, CORE-1) for non-linear structural dynamics of single as well as clustered subassemblies. (Auth.)

  15. Etizolam-induced superficial erythema annulare centrifugum.

    Science.gov (United States)

    Kuroda, K; Yabunami, H; Hisanaga, Y

    2002-01-01

    Erythema annulare centrifugum (EAC) is characterized by slowly enlarging annular erythematous lesions. Although the origin is not clear in most cases, EAC has been associated with infections, medications, and in rare cases, underlying malignancy. We describe a patient who developed annular erythematous lesions after etizolam administration. The eruptions were typical of the superficial form of EAC, both clinically and histopathologically. The lesions disappeared shortly after discontinuation of the medication. Patch testing with etizolam gave positive results. To our knowledge this is the first reported case of etizolam-induced superficial EAC. PMID:11952667

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

    International Nuclear Information System (INIS)

    The High-Uranium-Loaded U3O8-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 U3O8-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 U3O8-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 % U3O8). 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 U3O8 and aluminum. (author)

  17. Finite-element procedure for calculating the three-dimensional inelastic bowing of fuel rods (AWBA development program)

    Energy Technology Data Exchange (ETDEWEB)

    Martin, S E

    1982-05-01

    An incremental finite element procedure is developed for calculating the in-pile lateral bowing of nuclear fuel rods. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the results of an axisymmetric stress analysis of the fuel rod. The effects which are taken into account in calculating the rod's lateral bowing include: (a) lateral, axial, and rotational motions and forces at the rod supports, (b) transverse gradients of temperature, fast-neutron flux, and fissioning rate, and (c) cladding circumferential wall thickness variation. The procedure developed in this report could be used to form the basis for a computer program to calculate the time-dependent bowing as a function of the fuel rod's operational and environmental history.

  18. Control Methods Used in the Department of Metallurgy for Structure and Fuel Elements

    International Nuclear Information System (INIS)

    Studies of various reactor types undertaken by the Commissariat à l'énergie atomique have led to the use and-development of many non-destructive methods of testing the various materials employed in their construction, and especially methods involving the use of X- and γ-rays, ultrasonic waves and eddy currents. A summary is given below of the tests carried out during the construction of EDF (graphite gas) and EL 4 (heavy water) reactor types and of co-laminated elements. Emphasis is laid on certain typical aspects of these various methods, some of which are well known. EDF type: ultrasonic measurement of the wall thicknesses of uranium or uranium alloy tubes; gamma- radiographic detection of voids in the tubes; ultrasonic inspection following heat treatment of the tubes ultrasonic flaw detection (inclusions, pipes) in billets and rough casts for MgZr cladding; leak-tightness testing of slugs by helium eliquation. EL4 type: ultrasonic and eddy-current measurements of the wall thicknesses of Zircaloy pressure and guide tubes; ultrasonic inspection of Zircaloy pressure tubes and stainless-steel cladding tubes; in vacuo radiography of Be claddings; leak-tightness testing of pencils by helium eliquation. Co-laminated elements; cladding thickness measurements by pulsed eddy currents; determination of core position in tubes and plates, using X-rays, gamma counting and pulsed eddy currents; checking of fuel homogeneity by gamma counting; ultrasonic and gamma-radiographic flaw detection in ZrU billets; detection of unbonded areas in plates, using ultrasonic waves, pulsed eddy currents and resistivity measurements. The methods have been developed for industrial use. Several of the inspections already constitute routine manufacturing procedures, others will be introduced shortly and a further number are likely to come into use at a later date. (author)

  19. Thermionic fuel element optimization studies for the S-PRIME system

    International Nuclear Information System (INIS)

    The S-PRIME reactor is being developed to address the requirements of the Thermionic Space Nuclear Power System (TI-SNPS) program (DOE, 1992). The design features a pumped NaK loop, moderated spectrum, in-core thermionic power system which is fully scalable over a 5 to 40 kWe power range using the same basic thermionic fuel element (TFE) design approach. The TFE design is based on multi-cell technology in which several converters are stacked to establish a series electrical connection similar to that of flashlight batteries. The TFE design incorporates an integral cesium reservoir which is thermally coupled to the emitter and collector of the adjacent thermionic cell. The cesium pressure in the inter electrode gap is controlled by the temperature of the reservoir graphite. This paper will address the optimization studies which led to the choice of a 6-cell uniform converter TFE design for the 40 kWe reactor. Close interaction with reactor nuclear analysis (Chiu, 1993) was maintained to assure that the TFE configuration meets criticality requirements as well as produces the desired electrical power

  20. An experimental and analytical study of fluid flow and critical heat flux in PWR fuel elements

    International Nuclear Information System (INIS)

    This report describes experiments that have been carried out at the Winfrith Establishment of the United Kingdom Atomic Energy Authority to determine the critical heat flux characteristics of pressurized water reactor fuel elements over an unusually wide range of coolant flow conditions that are relevant to both normal and fault conditions of reactor operation. The experiments were carried out in the TITAN loop using an electrically heated bundle of 25 rods of 9.5 mm diameter on a 12.7 mm pitch fitted with plain grids in order to provide a generic base for code validation. The fully tabulated experimental data for critical heat flux, pressure drop and sub-channel mixing are encompassed by ranges of pressure between 20 and 160 Bar, coolant flow between 150 and 3600 Kg/m2s, and coolant inlet temperature between 150 and 3200C. The results of the experiments are compared with predicted data based upon several established critical heat flux correlations. It is concluded that the extrapolation of some correlations to conditions beyond their intended range of application can lead to dangerous over estimates of critical heat flux, but the Winfrith WSC-2 and the EPRI NP-2609 correlations perform well over the whole data range and correlate all data with RMS errors of 9% and 6% respectively. (author)