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Sample records for nuclear fuel pellets

  1. Nuclear fuel pellet loading apparatus

    International Nuclear Information System (INIS)

    Gerkey, K.S.

    1979-01-01

    An automatic apparatus for loading a predetermined amount of nuclear fuel pellets into a nuclear fuel element to be used in a nuclear reactor is described. The apparatus consists of a vibratory bed capable of supporting corrugated trays containing rows of nuclear fuel pellets and arranged in alignment with the open ends of several nuclear fuel elements. A sweep mechanism is arranged above the trays and serves to sweep the rows of fuel pellets onto the vibratory bed and into the fuel element. A length detecting system, in conjunction with a pellet stopping mechanism, is also provided to assure that a predetermined amount of nuclear fuel pellets are loaded into each fuel element

  2. Method of manufacturing nuclear fuel pellet

    International Nuclear Information System (INIS)

    Oguma, Masaomi; Masuda, Hiroshi; Hirai, Mutsumi; Tanabe, Isami; Yuda, Ryoichi.

    1989-01-01

    In a method of manufacturing nuclear fuel pellets by compression molding an oxide powder of nuclear fuel material followed by sintering, a metal nuclear material is mixed with an oxide powder of the nuclear fuel material. As the metal nuclear fuel material, whisker or wire-like fine wire or granules of metal uranium can be used effectively. As a result, a fuel pellet in which the metal nuclear fuel is disposed in a network-like manner can be obtained. The pellet shows a great effect of preventing thermal stress destruction of pellets upon increase of fuel rod power as compared with conventional pellets. Further, the metal nuclear fuel material acts as an oxygen getter to suppress the increase of O/M ratio of the pellets. Further, it is possible to reduce the swelling of pellet at high burn-up degree. (T.M.)

  3. Nuclear fuel pellet charging device

    International Nuclear Information System (INIS)

    Komuro, Kojiro.

    1990-01-01

    The present invention concerns a nuclear fuel pellet loading device, in which nuclear fuel pellets are successively charged from an open end of a fuel can while rotating the can. That is, a fuel can sealed at one end with an end plug and opened at the other end is rotated around its pipe axis as the center on a rotationally diriving table. During rotation of the fuel can, nuclear fuel pellets are successively charged by means of a feed rod of a feeding device to the inside of the fuel can. The fuel can is rotated while being supported horizontally and the fuel pellets are charged from the open end thereof. Alternatively, the fuel can is rotated while being supported obliquely and the fuel pellets are charged gravitationally into the fuel can. In this way, the damages to the barrier of the fuel can can be reduce. Further, since the fuel pellets can be charged gravitationally by rotating the fuel can while being supported obliquely, the damages to the barrier can be reduced remarkably. (I.S.)

  4. Axially alignable nuclear fuel pellets

    International Nuclear Information System (INIS)

    Johansson, E.B.; Klahn, D.H.; Marlowe, M.O.

    1978-01-01

    An axially alignable nuclear fuel pellet of the type stacked in end-to-end relationship within a tubular cladding is described. Fuel cladding failures can occur at pellet interface locations due to mechanical interaction between misaligned fuel pellets and the cladding. Mechanical interaction between the cladding and the fuel pellets loads the cladding and causes increased cladding stresses. Nuclear fuel pellets are provided with an end structure that increases plastic deformation of the pellets at the interface between pellets so that lower alignment forces are required to straighten axially misaligned pellets. Plastic deformation of the pellet ends results in less interactions beween the cladding and the fuel pellets and significantly lowers cladding stresses. The geometry of pellets constructed according to the invention also reduces alignment forces required to straighten fuel pellets that are tilted within the cladding. Plastic deformation of the pellets at the pellet interfaces is increased by providing pellets with at least one end face having a centrally-disposed raised area of convex shape so that the mean temperature and shear stress of the contact area is higher than that of prior art pellets

  5. Nuclear fuel pellet loading machine

    International Nuclear Information System (INIS)

    Dazen, J.R.; Denero, J.V.

    1976-01-01

    A nuclear fuel pellet loading machine is described including an inclined rack mounted on a base and having parallel spaced grooves on its upper surface arranged to support fuel rods. A fuel pellet tray is adapted to be placed on a table spaced from the rack, the tray having columns of fuel pellets which are in alignment with the open ends of fuel rods located in the rack grooves. A transition plate is mounted between the fuel rod rack and the fuel pellet tray to receive and guide the pellets into the open ends of the fuel rods. The pellets are pushed into the fuel rods by a number of mechanical fingers mounted on a motor operated block which is moved along the pellet tray length by a drive screw driven by the motor. To facilitate movement of the pellets in the fuel rods the rack is mounted on a number of spaced vibrators which vibrate the fuel rods during fuel pellet insertion. A pellet sensing device movable into an end of each fuel rod indicates to an operator when each rod has been charged with the correct number of pellets

  6. Nuclear fuel pellet production method and nuclear fuel pellet

    International Nuclear Information System (INIS)

    Yuda, Ryoichi; Ito, Ken-ichi; Masuda, Hiroshi.

    1993-01-01

    In a method of manufacturing nuclear fuel pellets by compression-molding UO 2 powders followed by sintering, a sintering agent having a composition of about 40 to 80 wt% of SiO 2 and the balance of Al 2 O 3 , a sintering agent at a ratio of 10 to 500 ppm based on the total amount of UO 2 and UO 2 powders are mixed, compression molded and then sintered at a sintering temperature of about 1500 of 1800degC. The UO 2 particles have an average grain size of about 20 to 60μm, most of the crystal grain boundary thereof is coated with a glassy or crystalline alumina silicate phase, and the porosity is about 1 to 4 vol%. With such a constitution, the sintering agent forms a single liquid phase eutectic mixture during sintering, to promote a surface reaction between nuclear fuel powders by a liquid phase sintering mechanism, increase their density and promote the crystal growth. Accordingly, it is possible to lower the softening temperature, improve the creep velocity of the pellets and improve the resistance against pellet-clad interaction. (T.M.)

  7. Intelligent Automated Nuclear Fuel Pellet Inspection System

    International Nuclear Information System (INIS)

    Keyvan, S.

    1999-01-01

    At the present time, nuclear pellet inspection is performed manually using naked eyes for judgment and decisionmaking on accepting or rejecting pellets. This current practice of pellet inspection is tedious and subject to inconsistencies and error. Furthermore, unnecessary re-fabrication of pellets is costly and the presence of low quality pellets in a fuel assembly is unacceptable. To improve the quality control in nuclear fuel fabrication plants, an automated pellet inspection system based on advanced techniques is needed. Such a system addresses the following concerns of the current manual inspection method: (1) the reliability of inspection due to typical human errors, (2) radiation exposure to the workers, and (3) speed of inspection and its economical impact. The goal of this research is to develop an automated nuclear fuel pellet inspection system which is based on pellet video (photographic) images and uses artificial intelligence techniques

  8. Sintering method for nuclear fuel pellet

    International Nuclear Information System (INIS)

    Omuta, Hirofumi; Nakabayashi, Shigetoshi.

    1997-01-01

    When sintering a compressed nuclear fuel powder in an atmosphere of a mixed gas comprising hydrogen and nitrogen, steams are added to the mixed gas to suppress the nitrogen content in sintered nuclear fuel pellets. In addition, the content of nitrogen impurities in the nuclear fuel pellets can be controlled by controlling the amount of steams to be added to the mixed gas, namely, by controlling the dew point as an index thereof. If the addition amount of steams to the mixed gas is determined by controlling the dew point as an index, the content of nitrogen impurities in the sintered nuclear fuel pellets can be controlled reliably to a specified value of 0.0075% or less. If ammonolyzed gas is used as the mixed gas, a more economical mixed gas can be obtained than in the case of forming mixed gas by mixing the hydrogen gas and the nitrogen gas. (N.H.)

  9. Fluid pressure method for recovering fuel pellets from nuclear fuel elements

    International Nuclear Information System (INIS)

    John, C.D. Jr.

    1979-01-01

    A method is described for removing fuel pellets from a nuclear fuel element without damaging the fuel pellets or fuel element sheath so that both may be reused. The method comprises holding the fuel element while a high pressure stream internally pressurizes the fuel element to expand the fuel element sheath away from the fuel pellets therein so that the fuel pellets may be easily removed

  10. Nuclear fuel pellet transfer escalator

    International Nuclear Information System (INIS)

    Huggins, T.B. Sr.; Roberts, E.; Edmunds, M.O.

    1991-01-01

    This patent describes a nuclear fuel pellet escalator for loading nuclear fuel pellets into a sintering boat. It comprises a generally horizontally-disposed pellet transfer conveyor for moving pellets in single file fashion from a receiving end to a discharge end thereof, the conveyor being mounted about an axis at its receiving end for pivotal movement to generally vertically move its discharge end toward and away from a sintering boat when placed below the discharge end of the conveyor, the conveyor including an elongated arm swingable vertically about the axis and having an elongated channel recessed below an upper side of the arm and extending between the receiving and discharge ends of the conveyor; a pellet dispensing chute mounted to the arm of the conveyor at the discharge end thereof and extending therebelow such that the chute is carried at the discharge end of the conveyor for generally vertical movement therewith toward and away from the sintering boat

  11. Nuclear fuel pellet loading machine

    International Nuclear Information System (INIS)

    Kee, R.W.; Denero, J.V.

    1975-01-01

    An apparatus for loading nuclear fuel pellets on trays for transfer in a system is described. A conveyor supplies pellets from a source to a loading station. When the pellets reach a predetermined position at the loading station, a manual or automatically operated arm pushes the pellets into slots on a tray and this process is repeated until pellet sensing switches detect that the tray is full. Thereupon, the tray is lowered onto a belt or other type conveyor and transferred to other apparatus in the system, such as a furnace for sintering, and in some cases, reduction of UO 2 . 2 to UO 2 . The pellets are retained on the tray and subsequently loaded directly into fuel rods to be used in the reactor core. (auth)

  12. Nuclear fuel pellets

    International Nuclear Information System (INIS)

    Larson, R.I.; Brassfield, H.C.

    1981-01-01

    Increased strength and physical durability in green bodies or pellets formed of particulate nuclear fuel oxides is achieved by inclusion of a fugitive binder which is ammonium bicarbonate, bicarbonate carbomate, carbomate, sesquicarbonate or mixtures thereof. Ammonium oxadate may be included as pore former. (author)

  13. Nuclear fuel pellet collating system and method

    International Nuclear Information System (INIS)

    Rieben, S.L.; Kugler, R.W.; Scherpenberg, J.J.; Wiersema, D.T.

    1990-01-01

    This patent describes a method of collating nuclear fuel pellets. It comprises: supporting a plurality of pellet supply trays and a plurality of pellet storage trays at a tray positioning station. Each of the supply trays containing in at least one row thereon a plurality of nuclear fuel pellets of an enrichment different from the enrichment pellets on at least some other of the supply trays; transferring one pellet supply tray from the tray positioning station and disposing the same at an input station of a pellet collating line; transferring one pellet storage tray from the tray positioning station and disposing the same at an output station of the pellet collating line; sweeping pellets in the at least one row thereof from the one pellet supply tray onto a work station of the pellet collating line located between the input and output stations thereof; measuring a desired length of pellets in the at least one row on the work station and separating the measured desired length of pellets from the remaining pellets, if any, in the row thereof; sweeping the remaining pellets, if any, in the row from the work station back onto the one pellet supply tray; transferring the one pellet supply tray and remaining pellets, if any, back to the tray positioning station; sweeping the measured desired length of pellets from the work station onto the one pellet storage tray; and transferring the one pellet storage tray and measured desired length of pellets back to the tray positioning station

  14. Apparatus for feeding nuclear fuel pellets to a loading tray

    International Nuclear Information System (INIS)

    Huggins, T.B.

    1979-01-01

    Apparatus for feeding nuclear fuel pellets at a uniform predetermined rate between pellet centering and grinding apparatus and a tray for loading pellets into nuclear fuel rod. Pellets discharged from the grinding apparatus are conveyed by a belt to a drive wheel forcing the pellets in engagement with the belt. The pellets under the drive wheel are capable of pushing a line of about 36 pellets onto a pellet dumping mechanism. As the dumping mechanism is actuated to dump the pellets on to a loading tray, the pellets moving toward the mechanism are stopped and the drive wheel is simultaneously lifted off the pellets until the pellet dumping process is completed. (U.K.)

  15. Inspecting fuel pellets for nuclear reactor

    International Nuclear Information System (INIS)

    Wilks, R.S.; Sternheim, E.; Breakey, G.A.; Sturges, R.H.; Taleff, A.; Castner, R.P.

    1982-01-01

    An improved method of controlling the inspection, sorting and classifying of nuclear reactor fuel pellets, including a mechanical handling system and a computer controlled data processing system, is described. Having investigated the diameter, length, surface flaws and weights of the pellets, they are sorted accordingly and the relevant data are stored. (U.K.)

  16. Nuclear fuel rod with burnable plate and pellet-clad interaction fix

    International Nuclear Information System (INIS)

    Boyle, R.F.

    1987-01-01

    This patent describes a nuclear fuel rod comprising a metallic tubular cladding containing nuclear fuel pellets, the pellets containing enriched uranium-235. The improvement described here comprises: ceramic wafers, each wafter comprising a sintered mixture of gadolinium oxide and uranium dioxide, the uranium oxide having no more uranium-235 than is present in natural uranium dioxide. Each of the wafers is axially disposed between a major portion of adjacent the nuclear fuel pellets, whereby the wafers freeze out volatile fission products produced by the nuclear fuel and prevent interaction of the fission products with the metallic tubing cladding

  17. Fuel pellet loading apparatus

    International Nuclear Information System (INIS)

    1980-01-01

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

  18. Method of manufacturing nuclear fuel pellet

    International Nuclear Information System (INIS)

    Oguma, Masaomi; Masuda, Hiroshi.

    1988-01-01

    Purpose: To prevent pellet destruction due to thermal stresses and reduce the swelling or issue of corrosive gaseous fission products. Method: Raw material powder for nuclear fuel pellets constitute so-called secondary particles in which a plurality of primary particles are coagulated. The degree of coagulation of the secondary particles can be determined as the bulk density of the powder. In view of the above, when pellets are sintered by using a powder mixture comprising a powder having the same constitution and different bulk density from the main raw powder as the sub-raw material powder incorporated to the main raw material powder, the pellet tissue provides such a fine porous structure that fine gaps are present a the periphery of high density secondary particles, since there is a difference in the shrinkage factor (sintering-shrinkage degree) between powders of different secondary particle densities in the course of the sintering. Thus, pellets can be prevented from thermal impact destruction and cause no destructive cracks. (Takahashi, M.)

  19. Production of pellets for nuclear fuel elements

    International Nuclear Information System (INIS)

    Butler, G.G.

    1982-01-01

    A method for producing nuclear fuel pellets each made up of a central portion and an outer annular portion surrounding the central portion, the two portions differing in composition. Such pellets are termed annular-layered pellets. The method comprises the steps of pressing powdered refractory material which has been granulated to form separately a central portion and an outer annular portion, assembling the portions together, compacting the assembly and sintering the compact. The portions are bonded together during sintering. The difference in composition may include a difference in density or isotopic enrichment as well as a chemical difference. (author)

  20. Nuclear fuel pellet sintering boat unloading apparatus and method

    International Nuclear Information System (INIS)

    Huggins, T.B.; Widener, W.H.; Klapper, K.K.

    1990-01-01

    This patent describes a method for unloading nuclear fuel pellets from a sintering boat having an open top. It comprises: pivoting a transfer housing loaded with the boat filled with nuclear fuel pellets about a generally horizontal axis from an upright position remote from a pellet deposit surface to an inverted position adjacent to the deposit surface to move the boat from an upright to inverted orientation with the pellets retained within the boat by a latched lid in a closed condition on the housing; unlatching the lid of the housing as the housing reaches its inverted position but engaging the unlatched lid with the deposit surface to retain it in its closed condition; and reverse pivoting the housing from its inverted position back toward its upright position to permit the unlatched lid to pivot from the closed condition to an opened condition thereby allowing pellets to slide out of the open top of the inverted boat and down the opened lid of the housing to the deposit site

  1. Pellet-press-to-sintering-boat nuclear fuel pellet loading system

    International Nuclear Information System (INIS)

    Bucher, G.D.

    1988-01-01

    This patent describes a system for loading nuclear fuel pellets into a sintering boat from a pellet press which ejects newly made the pellets from a pellet press die table surface. The system consists of: (a) a bowl having an inner surface, a longitudinal axis, an open and generally circular top of larger diameter, and an open and generally circular bottom of smaller diameter; (b) means for supporting the bowl in a generally upright position such that the bowl is rotatable about its longitudinal axis; (c) means for receiving the ejected pellets proximate the die table surface of the pellet press and for discharging the received pellets into the bowl at a location proximate the inner surface towards the top of the bowl with a pellet velocity having a horizontal component which is generally tangent to the inner surface of the bowl proximate the location; (d) means for rotating the bowl about the longitudinal axis such that the bowl proximate the location has a velocity generally equal, in magnitude and direction, to the horizontal component of the pellet velocity at the location; and (e) means for moving the sintering boat generally horizontally beneath and proximate the bottom of the bowl

  2. Total and occluded residual gas content inside the nuclear fuel pellets

    International Nuclear Information System (INIS)

    Moura, Sergio C.; Fernandes, Carlos E.; Oliveira, Justine R.; Machado, Joyce F.; Guglielmo, Luisa M.; Bustillos, Oscar V.

    2009-01-01

    This work describes three techniques available to measure total and occluded residual gases inside the UO 2 nuclear fuel pellets. Hydrogen is the major gas compound inside these pellets, due to sintering fabrication process but Nitrogen is present as well, due to storage atmosphere fuel. The total and occluded residual gas content inside these pellets is a mandatory requirement in a quality control to assure the well function of the pellets inside the nuclear reactor. This work describes the Gas Extractor System coupled with mass spectrometry GES/MS, the Gas Extractor System coupled with gas chromatography GES/GC and the total Hydrogen / Nitrogen H/N analyzer as well. In the GES, occlude gases in the UO 2 pellets is determinate using a high temperature vacuum extraction system, in which the minimum limit of detection is in the range 0.002 cc/g. The qualitative and quantitative determination of the amount of gaseous components employs a mass spectrometry or a gas chromatography technique. The total Hydrogen / Nitrogen analyzer employ a thermal conductivity gas detector linked to a gaseous extractor furnace which has a detection limit is in the range 0.005 cc/g. The specification for the residual gas analyses in the nuclear fuel pellets is 0.03 cc/g, all techniques satisfy the requirement but not the nature of the gases due to reaction with the reactor cladding. The present work details the chemical reaction among Hydrogen / Nitrogen and nuclear reactor cladding. (author)

  3. Vibratory-compacted (vipac/sphere-pac) nuclear fuels - a comparison with pelletized nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Chidester, K.; Rubin, J. [Los Alamos National Lab., NM (United States); Thompson, M

    2001-07-01

    In order to achieve the packing densities required for nuclear fuel stability, economy and performance, the fuel material must be densified. This has traditionally been performed by high-temperature sintering. (At one time, fuel densification was investigated using cold/hot swaging. However, this fabrication method has become uncommon.) Alternatively, fuel can be densified by vibratory compaction (VIPAC). During the late 1950's and into the 1970's, in the U.S., vibratory compaction fuel was fabricated and test irradiated to evaluate its applicability compared to the more traditional pelletized fuel for nuclear reactors. These activities were primarily focused on light water reactors (LWR) but some work was performed for fast reactors. This paper attempts to summarize these evaluations and proposes to reconsider VIPAC fuel for future use. (author)

  4. Vibratory-compacted (vipac/sphere-pac) nuclear fuels - a comparison with pelletized nuclear fuels

    International Nuclear Information System (INIS)

    Chidester, K.; Rubin, J.; Thompson, M.

    2001-01-01

    In order to achieve the packing densities required for nuclear fuel stability, economy and performance, the fuel material must be densified. This has traditionally been performed by high-temperature sintering. (At one time, fuel densification was investigated using cold/hot swaging. However, this fabrication method has become uncommon.) Alternatively, fuel can be densified by vibratory compaction (VIPAC). During the late 1950's and into the 1970's, in the U.S., vibratory compaction fuel was fabricated and test irradiated to evaluate its applicability compared to the more traditional pelletized fuel for nuclear reactors. These activities were primarily focused on light water reactors (LWR) but some work was performed for fast reactors. This paper attempts to summarize these evaluations and proposes to reconsider VIPAC fuel for future use. (author)

  5. Apparatus for transferring nuclear fuel pellets to a plate loader

    International Nuclear Information System (INIS)

    Huggins, T.B.

    1978-01-01

    An apparatus is described for transferring nuclear fuel pellets from a grinding machine to a plate loader. It includes a frame, an endless belt fitted to the frame, a control system provided on it for actuating the belt at a preset speed, a V shaped vessel fitted directly above the belt and extending along its length to guide the pellets on the belt and a device to receive the pellets coming from the belt [fr

  6. Fuel rod pellet loading head

    International Nuclear Information System (INIS)

    Howell, T.E.

    1975-01-01

    An assembly for loading nuclear fuel pellets into a fuel rod comprising a loading head for feeding pellets into the open end of the rod is described. The pellets rest in a perforated substantially V-shaped seat through which air may be drawn for removal of chips and dust. The rod is held in place in an adjustable notched locator which permits alignment with the pellets

  7. Comparative analysis of thermal behavior in hollow nuclear fuel pellets

    International Nuclear Information System (INIS)

    Santos, Beatriz M. dos; Alvim, Antonio C.M.

    2017-01-01

    The increase in energy demand in Brazil and in the world is a real problem and several solutions are being considered to mitigate it. Maximization of energy generation, within the safety standards of fuel resources already known, is one of them. In this respect, nuclear energy is a crucial technology to sustain energy demand on several countries. Performances of a solid cylindrical and an annular rod have been verified and compared; where it has been proven that the annular rod can reach a higher nominal power in relation to the solid one. In this paper, the temperature profiles of two distinct nuclear fuel pellets, one of them annular and the other in the shape of a hollow biconcave disc (like the cross section of a red blood cell), were compared to analyze the efficiency and safety of both. The finite differences method allowed the evaluation of the thermal behavior of these pellets, where one specific physical condition was analyzed, regarding convection and conduction at the lateral edges. The results show that the temperature profile of the hollow biconcave disc pellet is lower, about 70 deg C below, when compared to the temperature profile of the annular pellet, considering the same simulation parameters for both pellets. (author)

  8. Comparative analysis of thermal behavior in hollow nuclear fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Beatriz M. dos; Alvim, Antonio C.M., E-mail: bmachado@nuclear.ufrj.br, E-mail: aalvim@gmail.com [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-11-01

    The increase in energy demand in Brazil and in the world is a real problem and several solutions are being considered to mitigate it. Maximization of energy generation, within the safety standards of fuel resources already known, is one of them. In this respect, nuclear energy is a crucial technology to sustain energy demand on several countries. Performances of a solid cylindrical and an annular rod have been verified and compared; where it has been proven that the annular rod can reach a higher nominal power in relation to the solid one. In this paper, the temperature profiles of two distinct nuclear fuel pellets, one of them annular and the other in the shape of a hollow biconcave disc (like the cross section of a red blood cell), were compared to analyze the efficiency and safety of both. The finite differences method allowed the evaluation of the thermal behavior of these pellets, where one specific physical condition was analyzed, regarding convection and conduction at the lateral edges. The results show that the temperature profile of the hollow biconcave disc pellet is lower, about 70 deg C below, when compared to the temperature profile of the annular pellet, considering the same simulation parameters for both pellets. (author)

  9. Remote visual inspection of nuclear fuel pellets with fiber optics and video image processing

    International Nuclear Information System (INIS)

    Moore, F.W.

    1987-01-01

    Westinghouse Hanford Company has designed and constructed a nuclear fuel fabrication process line for the U.S. Department of Energy. This process line includes a system that remotely inspects the cylindrical surface of nuclear fuel pellets for surface spots, flaws, or discoloration. The pellets are inspected on a 100% basis after pellet sintering. A feeder delivers the pellets directly to a fiber optic inspection head, which views one pellet surface at a time and images it to a closed-circuit color television camera (CCTV). The output signal of the CCTV is input to a digital imaging processor that stores approximately 25 pellet images at a time. A human operator visually examines the images of the pellet surfaces on a high resolution monitor and accepts or rejects the pellets based on visual standards. The operator uses a digitizing tablet to record the location of rejected pellets, which are then automatically removed from the product stream. The system is expandable to automated disposition of the pellet surface image

  10. Remote visual inspection of nuclear fuel pellets with fiber optics and video image processing

    International Nuclear Information System (INIS)

    Moore, F.W.

    1986-01-01

    Westinghouse Hanford Company has designed and is constructing a nuclear fuel fabrication process line for the Department of Energy. This process line includes a pellet surface inspection system that remotely inspects the cylindrical surface of nuclear fuel pellets for surface spots, flaws, or discoloration. The pellets are inspected on a 100 percent basis after pellet sintering. A feeder will deliver the pellets directly to a fiber optic inspection head. The inspection head will view one pellet surface at a time. The surface image of the pellet will be imaged to a closed-circuit color television camera (CCTV). The output signal of the CCTV will be input to a digital imaging processor that stores approximately 25 pellet images at a time. A human operator will visually examine the images of the pellet surfaces on a high resolution monitor and accept or reject the pellets based on visual standards. The operator will use a digitizing tablet to record the location of rejected pellets, which will then be automatically removed from the product stream. The system is expandable to automated disposition of the pellet surface image

  11. Remote visual inspection of nuclear fuel pellets with fiber optics and video image processing

    International Nuclear Information System (INIS)

    Moore, F.W.

    1985-01-01

    Westinghouse Hanford Company has designed and is constructing a nuclear fuel fabrication process line for the Department of Energy. This process line includes a pellet surface inspection system that remotely inspects the cylindrical surface of nuclear fuel pellets for surface spots, flaws, or discoloration. The pellets are inspected on a 100% basis after pellet sintering. A feeder will deliver the pellets directly to fiber optic inspection head. The inspection head will view one pellet surface at a time. The surface image of the pellet will be imaged to a closed-circuit color television camera (CCTV). The output signal of the CCTV will be input to a digital imaging processor that stores approximately 25 pellet images at a time. A human operator will visually examine the images of the pellet surfaces on a high resolution monitor and accept or reject the pellets based on visual standards. The operator will use a digitizing tablet to record the location of rejected pellets, which will then be automatically removed from the product stream. The system is expandable to automated disposition of the pellet surface image

  12. Apparatus for loading fuel pellets in fuel rods

    International Nuclear Information System (INIS)

    Tedesco, R.J.

    1976-01-01

    An apparatus is disclosed for loading fuel pellets into fuel rods for a nuclear reactor including a base supporting a table having grooves therein for holding a multiplicity of pellets. Multiple fuel rods are placed in alignment with grooves in the pellet table and a guide member channels pellets from the table into the corresponding fuel rods. To effect movement of pellets inside the fuel rods without jamming, a number of electromechanical devices mounted on the base have arms connected to the lower surface of the fuel rod table which cyclically imparts a reciprocating arc motion to the table for moving the fuel pellets longitudinally of and inside the fuel rods. These electromechanical devices include a solenoid having a plunger therein connected to a leaf type spring, the arrangement being such that upon energization of the solenoid coil, the leaf spring moves the fuel rod table rearwardly and downwardly, and upon deenergization of the coil, the spring imparts an upward-forward movement to the table which results in physical displacement of fuel pellets in the fuel rods clamped to the table surface. 8 claims, 6 drawing figures

  13. Apparatus for unloading more particularly for nuclear fuel pellets, and to fill tubes with these pellets

    International Nuclear Information System (INIS)

    Fort, C.; Masson, S.

    1985-01-01

    The device allows to discharge the nuclear fuel pellets arranged in trays, and to introduce them to form stacks of pellets of determined length in storage tubes of associated diameter. It comprises a carriage to make the pellets slip from each tray on a guide vibrating bowl to a shute and then on a conveyor which loads the pellets into an intermediate tube to form a stack of the said length. A lift moves the intermediate tube transversally to its length between a loading position and a transfer position. Means allow to move a storage tube bundle to put each tube in its turn face to the transfer position. The stack of pellets contained in the intermediate tube which is in the transfer position is thus sent back to the storage tube facing it. The invention applies to pellets which have been sintered in the trays in inert atmosphere. These pellets have to be stored before several examinations and grinding, and finally loading into the cans to constitute fuel rods. These sintered pellets have a cylindrical shape and the invention spares them hard handling which would damage them [fr

  14. Automatic failure identification of the nuclear power plant pellet fuel

    International Nuclear Information System (INIS)

    Oliveira, Adriano Fortunato de

    2010-01-01

    This paper proposed the development of an automatic technique for evaluating defects to help in the stage of fabrication of fuel elements. Was produced an intelligent image analysis for automatic recognition of defects in uranium pellets. Therefore, an Artificial Neural Network (ANN) was trained using segments of histograms of pellets, containing examples of both normal (no fault) and of defectives pellets (with major defects normally found). The images of the pellets were segmented into 11 shares. Histograms were made of these segments and trained the ANN. Besides automating the process, the system was able to obtain this classification accuracy of 98.33%. Although this percentage represents a significant advance ever in the quality control process, the use of more advanced techniques of photography and lighting will reduce it to insignificant levels with low cost. Technologically, the method developed, should it ever be implemented, will add substantial value in terms of process quality control and production outages in relation to domestic manufacturing of nuclear fuel. (author)

  15. Handling system for nuclear fuel pellet inspection

    International Nuclear Information System (INIS)

    Nyman, D.H.; McLemore, D.R.; Sturges, R.H.

    1978-11-01

    HEDL is developing automated fabrication equipment for fast reactor fuel. A major inspection operation in the process is the gaging of fuel pellets. A key element in the system has been the development of a handling system that reliably moves pellets at the rate of three per second without product damage or excessive equipment wear

  16. Nuclear reactor fuel element containing an end piece for maintaining the column of fuel pellets

    International Nuclear Information System (INIS)

    Pajot, Jacques; Rabellino, Jacques.

    1974-01-01

    The nuclear reactor fuel element described has an end piece for maintaining the column of fuel pellets in position inside the element cladding. This end piece has a central compression spring one end of which presses against the pellets and the other against a plug shaped piece fitted with a seat for the spring, a conical piece with an elastic ring around it diverging towards the end in contact with the spring and a head at the opposite end. The connection between the compression spring and the pellets is through an application piece. A central bore provided in the end piece helps balance the pressure inside the element. This element is particularly intended for liquid metal cooled fast neutron reactors [fr

  17. Process for the fabrication of nuclear fuel oxide pellets

    International Nuclear Information System (INIS)

    Francois, Bernard; Paradis, Yves.

    1977-01-01

    Process for the fabrication of nuclear fuel oxide pellets of the type for which particles charged with an organic binder -selected from the group that includes polyvinyl alcohol, carboxymethyl cellulose, polyvinyl compounds and methyl cellulose- are prepared from a powder of such an oxide, for instance uranium dioxide. These particles are then compressed into pellets which are then sintered. Under this process the binder charged particles are prepared by stirring the powder with a gas, spraying on to the stirred powder a solution or a suspension in a liquid of this organic binder in order to obtain these particles and then drying the particles so obtained with this gas [fr

  18. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Nakai, Keiichi

    1983-01-01

    Purpose: To decrease the tensile stresses resulted in a fuel can as well as prevent decladding of fuel pellets into the bore holes by decreasing the inner pressure within the nuclear fuel element. Constitution: A fuel can is filled with hollow fuel pellets, inserted with a spring for retaining the hollow fuel pellets with an appropriate force and, thereafter, closely sealed at the both ends with end plugs. A cylindrical body is disposed into the bore holes of the hollow fuel pellets. Since initial sealing gases and/or gaseous nuclear fission products can thus be excluded from the bore holes where the temperature is at the highest level, the inner pressure of the nuclear fuel element can be reduced to decrease the tensile strength resulted to the fuel can. Furthermore, decladding of fuel pellets into the bore holes can be prevented. (Moriyama, K.)

  19. Measurement of nuclear reaction rates and spectral indices along the radius of fuel pellets from IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Mura, Luis Felipe Liambos

    2010-01-01

    This work presents the measurements of the nuclear reaction rates along the radial direction of the fuel pellet by irradiation and posterior gamma spectrometry of a thin slice of fuel pellet of UO 2 with 4,3% enrichment. From its irradiation the rate of radioactive capture and fission have been measured as a function of the radius of the pellet disk using a HPGe detector. Lead collimators has been used for this purpose. Simulating the fuel pellet in the pin fuel of the IPEN/MB-01 reactor, a thin UO 2 disk is used. This disk is inserted in the interior of a dismountable fuel rod. This fuel rod is then placed in the central position of the IPEN/MB-01 reactor core and irradiated during 1 hour under a neutron flux of around 9 x 10 8 n/cm 2 s. For gamma spectrometry 10 collimators with different diameters have been used, consequently, the nuclear reactions of radioactive capture that occurs in atoms of 238 U and fissions that occur on both 235 U and 238 U are measured in function of 10 different region (diameter of collimator) of the fuel pellet disk. Corrections in the geometric efficiency due to introduction of collimators on HPGe detection system were estimated using photon transport of MCNP-4C code. Some calculated values of nuclear reaction rate of radioactive capture and fission along of the radial direction of the fuel pellet obtained by Monte Carlo methodology, using the MCNP-4C code, are presented and compared to the experimental data showing very good agreement. Besides nuclear reaction rates, the spectral indices 28 ρ and 25 δ have been obtained at each different radius of the fuel pellet disk. (author)

  20. Fuel pellet

    International Nuclear Information System (INIS)

    Hayashi, K.

    1980-01-01

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

  1. Nondestructive characterization of mixed oxide pellets in welded nuclear fuel pins by neutron radiography and gamma-autoradiography

    International Nuclear Information System (INIS)

    Panakkal, J.P.; Ghosh, J.K.; Roy, P.R.

    1989-01-01

    Nondestructive evaluation of nuclear fuel pellets after the welding of fuel pins plays a vital role in assuring a safe and reliable operation of reactors. Some of the important characteristics to be monitored in low plutonium enriched mixed oxide fuel pellets are plutonium enrichment, size of plutonium dioxide agglomerates, incorrect loading and geometric shape. Experiments were carried out at Bhabha Atomic Research Centre, Bombay on experimental fuel pins containing mixed oxide pellets of different geometry (solid and annular), of different plutonium enrichment (0-6 w% of plutonium dioxide) and containing PuO 2 agglomerates of size 125-2000 microns to evaluate these characteristics nondestructively. Neutron radiography of these fuel pins was carried out using a swimming pool type reactor 'APSARA'. Results of quantitative evaluation of the neutron radiographs and a simple model correlating neutron interaction probability and the optical density are presented. Gamma autoradiography of these fuel pins showed that these parameters could be evaluated with a few limitations. This paper presents the experimental details, quantitative analysis of the radiographs by microdensitometry and merits and demerits of neutron radiography and gamma autoradiography for nondestructive charcterisation of nuclear fuel pellets. (orig.)

  2. Thermal conductivity thermal diffusivity of UO{sub 2}-BeO nuclear fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Mansur, Fábio A.; Camarano, Denise M.; Santos, Ana M. M.; Ferraz, Wilmar B.; Silva, Mayra A.; Ferreira, Ricardo A.N., E-mail: fam@cdtn.br, E-mail: dmc@cdtn.br, E-mail: amms@cdtn.br, E-mail: ferrazw@cdtn.br, E-mail: mayra.silva@cdtn.br, E-mail: ricardoanf@yahoo.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The temperature distribution in nuclear fuel pellets is of vital importance for the performance of the reactor, as it affects the heat transfer, the mechanical behavior and the release of fission gas during irradiation, reducing safety margins in possible accident scenarios. One of the main limitation for the current uranium dioxide nuclear fuel (UO{sub 2}) is its low thermal conductivity, responsible for the higher temperature of the pellet center and, consequently, for a higher radial temperature gradient. Thus, the addition of another material to increase the UO{sub 2} fuel thermal conductivity has been considered. Among the additives that are being investigated, beryllium oxide (BeO) has been chosen due to its high thermal conductivity, with potential to optimize power generation in pressurized light water reactors (PWR). In this work, UO{sub 2}-BeO pellets were obtained by the physical mixing of the powders with additions of 2wt% and 3wt% of BeO. The thermal diffusivity and conductivity of the pellets were determined from room temperature up to 500 °C. The results were normalized to 95% of the theoretical density (TD) of the pellets and varied according to the BeO content. The range of the values of thermal diffusivity and conductivity were 1.22 mm{sup 2}∙s{sup -1} to 3.69 mm{sup 2}∙s{sup -1} and 3.80 W∙m{sup -}'1∙K{sup -1} to 9.36 W∙m{sup -1}∙K{sup -1}, respectively. (author)

  3. Fabrication of Cr-doped UO2 Fuel Pellet using Liquid Phase Sintering

    International Nuclear Information System (INIS)

    Kim, Dong Joo; Yang, Jae Ho; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Oh, Jang Soo; Koo, Yang Hyun

    2013-01-01

    An enhancement of the thermal conductivity of a pellet can be obtained by the addition of a higher thermal conductive material in the pellet. In addition, the resistance to the PCI can be increased through a plasticity increase of the pellet. Thermal conductivity of ceramic materials is generally lower than that of metallic materials. The thermal conductivity of uranium oxide which is a typical ceramic material is low as well. The steep temperature gradient in the fuel pellet results from the low thermal conductivity. Therefore, the thermal conductivity improvement of a nuclear fuel pellet can enhance the fuel performance in various aspects. The lower centerline temperature of a fuel pellet affects the enhancement of fuel safety as well as fuel pellet integrity during nuclear reactor operation. Besides, the nuclear reactor power can be uprated due to the higher safety margin. So, many researches to enhance the thermal conductivity of nuclear fuel pellet have been performed in various ways. To improve the thermal conductivity of UO 2 pellet, an appropriate arrangement of the high thermal conductive material in UO 2 matrix is one of the various methods. We intended to control a placement of chromium as the high thermal conductive material. The metallic chromium and chromium oxide were arranged in a grain boundary of UO 2 using a liquid phase sintering method. The liquid phase sintering of Cr-doped UO 2 pellet could be adjusted using a control of an oxygen potential in sintering atmosphere

  4. Fabrication and characterization of CeO{sub 2} pellets for simulation of nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    García-Ostos, C.; Rodríguez-Ortiz, J.A. [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain); Arévalo, C., E-mail: carevalo@us.es [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain); Cobos, J. [CIEMAT, Avenida Complutense, 40, Madrid (Spain); Gotor, F.J. [Materials Science Institute of Seville (CSIC-US), Av. Américo Vespucio, 49, 41092 Seville (Spain); Torres, Y. [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain)

    2016-03-15

    Highlights: • CeO{sub 2} is presented as a surrogate material for UO{sub 2} to study nuclear fuel. • Powder-metallurgy methods are applied to fabricate CeO{sub 2} pellets with controlled porosity. • An optimization of the fabrication parameters is established. • Microstructural and tribo-mechanical characterizations are performed. • Properties are compared to those of the nuclear fuel. - Abstract: Cerium Oxide, CeO{sub 2}, has been shown as a surrogate material to understand irradiated Mixed Oxide (MOX) based matrix fuel for nuclear power plants due to its similar structure, chemical and mechanical properties. In this work, CeO{sub 2} pellets with controlled porosity have been developed through conventional powder-metallurgy process. Influence of the main processing parameters (binder content, compaction pressure, sintering temperature and sintering time) on porosity and volumetric contraction values has been studied. Microstructure and physical properties of sintered compacts have also been characterized through several techniques. Mechanical properties such as dynamic Young's modulus, hardness and fracture toughness have been determined and connected to powder-metallurgy parameters. Simulation of nuclear fuel after reactor utilization with radial gradient porosity is proposed.

  5. Fuel rod with axial regions of annular and standard fuel pellets

    International Nuclear Information System (INIS)

    Freeman, T.R.

    1991-01-01

    This patent describes a fuel rod for use in a nuclear reactor fuel assembly. It comprises: an elongated hollow cladding tube; a pair of end plugs connected to and sealing the cladding tube at opposite ends of thereof; and an axial stack of fuel pellets contained in and extending between the end plugs at the opposite ends of the tube, all of the fuel pellets contained in the tube being composed of fissile material being enriched above the level of natural enrichment; the fuel pellets in the stack thereof being provided in an arrangement of axial regions. The arrangement of axial regions including a pair of first axial regions defined respectively at the opposite ends of the pellet stack adjacent to the respective end plugs. The pellets in the first axial regions being identical in number and having annular configurations with an annulus of a first void size. The arrangement of axial regions also including another axial region defined between the first axial regions, some of the pellets in the another axial region having solid configurations

  6. Shock and vibration tests of uranium mononitride fuel pellets for a space power nuclear reactor

    Science.gov (United States)

    Adams, D. W.

    1972-01-01

    Shock and vibration tests were conducted on cylindrically shaped, depleted, uranium mononitride (UN) fuel pellets. The structural capabilities of the pellets were determined under exposure to shock and vibration loading which a nuclear reactor may encounter during launching into space. Various combinations of diametral and axial clearances between the pellets and their enclosing structures were tested. The results of these tests indicate that for present fabrication of UN pellets, a diametral clearance of 0.254 millimeter and an axial clearance of 0.025 millimeter are tolerable when subjected to launch-induced loads.

  7. Contribution to numerical and mechanical modelling of pellet-cladding interaction in nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    Retel, V.

    2002-12-01

    Pressurised water reactor fuel rods (PWR) are the place of nuclear fission, resulting in unstable and radioactive elements. Today, the mechanical loading on the cladding is harder and harder and is partly due to the fuel pellet movement. Then, the mechanical behaviour of the cladding needs to be simulated with models allowing to assess realistic stress and strain fields for all the running conditions. Besides, the mechanical treatment of the fuel pellet needs to be improved. The study is part of a global way of improving the treatment of pellet-cladding interaction (PCI) in the 1D finite elements EDF code named CYRANO3. Non-axisymmetrical multidirectional effects have to be accounted for in a context of unidirectional axisymmetrical finite elements. The aim of this work is double. Firstly a model simulating the effect of stress concentration on the cladding, due to the opening of the radial cracks of fuel, had been added in the code. Then, the fragmented state of fuel material has been taken into account in the thermomechanical calculation, through a model which led the strain and stress relaxation in the pellet due to the fragmentation, be simulated. This model has been implemented in the code for two types of fuel behaviour: elastic and viscoplastic. (author)

  8. Hybrid pellets: an improved concept for fabrication of nuclear fuel

    International Nuclear Information System (INIS)

    Matthews, R.B.; Hart, P.E.

    1979-09-01

    The feasibility of fabricating fuel pellets using gel-derived microspheres as press feed was evaluated. By using gel-derived microspheres as press feed, the potential exists for eliminating dusty operations like milling, slugging, and granulation, from the pelleting process. The free-flowing character of the spheres should also result in limited dust generation during powder transport and pressing operations. The results of this study clearly demonstrate that fuel pellets can be successfully fabricated on a laboratory scale using UO 2 gel microspheres as press feed. Under moderate sintering conditions, 1,500 0 C for 4 h in Ar-4% H 2 , UO 2 pellets with densities up to 96% TD were fabricated. A range of pellet microstructures and densities were achieved depending on sphere forming and calcining conditions. Based on these results, a set of necessary sphere properties are suggested: O/U less than 2.20, crystallite size less than 500 A, specific surface area greater than 8 m 2 /g, and sphere size 200 and 400 μm. Preliminary attempts to fabricate ThO 2 and ThO 2 -UO 2 pellets using microspheres were unsuccessful because the requisite sphere properties were not achieved. Areas requiring additional development include: demonstration of the process on scaled-up equipment suitable for use in a remote fuel fabrication facility and evaluation of the irradiation performance of pellet fuels from gel-spheres

  9. Measurements of the nuclear reaction rates and spectral indices along the radius of the fuel pellets of the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Bitelli, Ulysses d'Utra; Mura, Luis Felipe L.; Fanaro, Leda C.C.B.

    2009-01-01

    This work presents the measures of the nuclear reaction rates along of the radial direction of the fuel pellet by irradiation and posterior gamma spectrometry of a thin slice of fuel pellet of UO 2 at 4.3% enrichment. From its irradiation the rate of radioactive capture and fission are measures as a function of the radius of the pellet disk using a HPGe detector. Diverse lead collimators of changeable diameters have been used for this purpose. Simulating the fuel pellet in the pin fuel of the IPEN/MB-01 reactor, a thin disk is used, being inserted in the interior of a dismountable fuel rod. This fuel rod is then placed in the central position of the IPEN/MB-01 reactor core and irradiated during 1 hour under a neutron flux of 5.10 8 n/cm 2 s. The nuclear reaction of radioactive capture occurs in the atoms of U- 238 that when absorbs a neutron transmutes into U- 239 of half-life of only 23 minutes. Thus, it is opted for the detection of the Np- 239 , radionuclide derivative of the radioactive decay of the U- 239 and that has a measurable half-life (2.335 days). In gamma spectrometry 11 collimators with different diameters have been used, consequently, the gamma spectrometry is made in function of the diameter (radius) of the irradiated UO 2 fuel pellet disk, thus is possible to get the average value of the counting for each collimator in function of the specific pellet radius. These values are directly proportional to the radioactive capture nuclear reaction rates. The same way the nuclear fission rate occurs in the atoms of the U- 235 that produce different fission products such as Ce- 143 with a yield fission of 5.9% and applying the same procedure the fission nuclear reaction rate is obtained. This work presents some calculated values of nuclear reaction rate of radioactive capture and fission along of the radial direction of the fuel pellet obtained by Monte Carlo methodology using the MCNP-4C code. The relative values obtained are compared with experimental

  10. Hot impact densification (HID) - a new method of producing ceramic nuclear fuel pellets with tight dimensional tolerances

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.; Muehling, G.; Vollath, D.; Zimmermann, H.

    1984-01-01

    The hot impact densification (HID) is a new powerful method for producing ceramic fuel pellets for nuclear reactors. Green ceramic bodies are directly processed to pellets by high speed shaping in the plastic temperature region of ceramic material. Opposed to the well established press sintering procedure it can be heated, densified, and cooled by orders of magnitude faster. Therefore, at high throughputs, small equipment dimensions become possible. The fuel pellets produced meet all requirements, particular the dimensional tolerances achieved are very closed, consequently circular grinding is omitted. Furthermore, the relatively high temperature level of the impact pressing favors the mixed crystal formation of uranium and plutonium oxide. This improves the solubility of the fuel in nitric acid, an essential point at reprocessing. A prototype facility is designed so that automatic fabrication in continuous operation will be possible. The target working cycle for a fuel pellet is in the range of some seconds. (orig.)

  11. Temperature Calculation of Annular Fuel Pellet by Finite Difference Method

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yong Sik; Bang, Je Geon; Kim, Dae Ho; Kim, Sun Ki; Lim, Ik Sung; Song, Kun Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    KAERI has started an innovative fuel development project for applying dual-cooled annular fuel to existing PWR reactor. In fuel design, fuel temperature is the most important factor which can affect nuclear fuel integrity and safety. Many models and methodologies, which can calculate temperature distribution in a fuel pellet have been proposed. However, due to the geometrical characteristics and cooling condition differences between existing solid type fuel and dual-cooled annular fuel, current fuel temperature calculation models can not be applied directly. Therefore, the new heat conduction model of fuel pellet was established. In general, fuel pellet temperature is calculated by FDM(Finite Difference Method) or FEM(Finite Element Method), because, temperature dependency of fuel thermal conductivity and spatial dependency heat generation in the pellet due to the self-shielding should be considered. In our study, FDM is adopted due to high exactness and short calculation time.

  12. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.; Macivergan, R.; Mckenzie, G.W.

    1980-01-01

    An apparatus incorporating a microprocessor control is provided for automatically loading nuclear fuel pellets into fuel rods commonly used in nuclear reactor cores. The apparatus comprises a split ''v'' trough for assembling segments of fuel pellets in rows and a shuttle to receive the fuel pellets from the split ''v'' trough when the two sides of the split ''v'' trough are opened. The pellets are weighed while in the shuttle, and the shuttle then moves the pellets into alignment with a fuel rod. A guide bushing is provided to assist the transfer of the pellets into the fuel rod. A rod carousel which holds a plurality of fuel rods presents the proper rod to the guide bushing at the appropriate stage in the loading sequence. The bushing advances to engage the fuel rod, and the shuttle advances to engage the guide bushing. The pellets are then loaded into the fuel rod by a motor operated push rod. The guide bushing includes a photocell utilized in conjunction with the push rod to measure the length of the row of fuel pellets inserted in the fuel rod

  13. Increase of thermal conductivity of uranium dioxide nuclear fuel pellets with beryllium oxide addition

    International Nuclear Information System (INIS)

    Camarano, D.M.; Mansur, F.A.; Santos, A.M.M. dos; Ferraz, W.B.

    2016-01-01

    The UO_2 fuel is one of the most used nuclear fuel in thermal reactors and has many advantages such as high melting point, chemical compatibility with cladding, etc. However, its thermal conductivity is relatively low, which leads to a premature degradation of the fuel pellets due to a high radial temperature gradient during reactor operation. An alternative to avoid this problem is to increase the thermal conductivity of the fuel pellets, by adding beryllium oxide (BeO). Pellets of UO_2 and UO_2-BeO were obtained from a homogenized mixture of powders of UO_2 and BeO, containing 2% and 3% by weight of BeO and sintering at 1750 °C for 3 h under H_2 atmosphere after uniaxial pressing at 400 MPa. The pellet densities were obtained by xylol penetration-immersion method and the thermal diffusivity, specific heat and thermal conductivity were determined according to ASTM E-1461 at room temperature (25 deg C) and 100 deg C. The thermal diffusivity measurements were carried out employing the laser flash method. The thermal conductivity obtained at 25 deg C showed an increase with the addition of 2% and 3% of BeO corresponding to 19% and 28%, respectively. As for the measurements carried out at 100 deg C, there was an increase in the thermal conductivity for the same BeO contents of 20% and 31%. These values as a percentage of increased conductivity were obtained in relation to the UO_2 pellets. (author)

  14. The effects of off-center pellets on the temperature distribution and the heat flux distribution of fuel rods in nuclear reactors

    International Nuclear Information System (INIS)

    Peng Muzhang; Xing Jianhua

    1986-01-01

    This paper analyzes the effects of off-center pellets on the steady state temperature distribution and heat flux distribution of fuel rods in the nuclear reactors, and derives the dimensionless temperature distribution relationships and the dimensionless heat flux distribution relationship from the fuel rods with off-center pellets. The calculated results show that the effects of off-center will result in not only deviations of the highest temperature placement in the fuel pellets, but also the circumferentially nonuniform distributions of the temperatures and heat fluxes of the fuel rod surfaces

  15. Apparatus and method for classifying fuel pellets for nuclear reactor

    International Nuclear Information System (INIS)

    Wilks, R.S.; Breakey, G.A.; Castner, R.P.; Sternheim, E.; Sturges, R.H. Jr.; Taleff, A.

    1984-01-01

    Control for the operation of a mechanical handling and gauging system for nuclear fuel pellets is claimed. The pellets are inspected for diameters, lengths, surface flaws and weights in successive stations. The control includes, a computer for commanding the operation of the system and its electronics and for storing and processing the complex data derived at the required high rate. In measuring the diameter, the computer enables the measurement of a calibration pellet, stores that calibration data and computes and stores diameter-correction factors and their addresses along a pellet. To each diameter measurement a correction factor is applied at the appropriate address. The computer commands verification that all critical parts of the system and control are set for inspection and that each pellet is positioned for inspection. During each cycle of inspection, the measurement operation proceeds normally irrespective of whether or not a pellet is present in each station. If a pellet is not positioned in a station, a measurement is recorded, but the recorded measurement indicates maloperation. In measuring diameter and length a light pattern including successive shadows of slices transverse for diameter or longitudinal for length are projected on a photodiode array. The light pattern is scanned electronically by a train of pulses. The pulses are counted during the scan of the lighted diodes. For evaluation of diameter the maximum diameter count and the number of slices for which the diameter exceeds a predetermined minimum is determined. For acceptance, the maximum must be less than a maximum level and the minimum must exceed a set number. For evaluation of length, the maximum length is determined. For acceptance, the length must be within maximum and minimum limits

  16. Nuclear fuel element

    International Nuclear Information System (INIS)

    Hirayama, Satoshi; Kawada, Toshiyuki; Matsuzaki, Masayoshi.

    1980-01-01

    Purpose: To provide a fuel element for reducing the mechanical interactions between a fuel-cladding tube and the fuel element and for alleviating the limits of the operating conditions of a reactor. Constitution: A fuel element having mainly uranium dioxide consists of a cylindrical outer pellet and cylindrical inner pellet inserted into the outer pellet. The outer pellet contains two or more additives selected from aluminium oxide, beryllium oxide, magnesium oxide, silicon oxide, sodium oxide, phosphorus oxide, calcium oxide and iron oxide, and the inner pellet contains nuclear fuel substance solely or one additive selected from calcium oxide, silicon oxide, aluminium oxide, magnesium oxide, zirconium oxide and iron oxide. The outer pellet of the fuel thus constituted is reduced in mechanical strength and also in the mechanical interactions with the cladding tube, and the plastic fluidity of the entire pellet is prevented by the inner pellet increased in the mechanical strength. (Kamimura, M.)

  17. Innovative microstructures in nuclear fuels

    International Nuclear Information System (INIS)

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

    2009-01-01

    For cleaner and safe nuclear power, new processes are required to design better nuclear fuels and make more efficient reactors to generate nuclear power. Therefore, one must understand how the microstructure changes during reactor operation. Accordingly, the materials scientists and engineers can then design and fabricate fuels with higher reliability and performance. Microstructure and its evolution are big unknowns in nuclear fuel. The basic requirements for the high performance of a fuel are: a) Soft pellets - To reduce Pellet clad mechanical interaction (PCMI) b) Large grain size - To reduce fission gas release (FGR). The strength of the pellet at room temperature is related to grain size by the Hall-Petch relation. Accordingly, the lower grain sized pellets will have high strength. But at high temperature (above equicohesive temperature) the grain boundaries becomes weaker than grain matrix. Since the small grain sized pellets have more grain boundary areas, these pellet become softer than pellet that have large grain sizes. Also as grain size decreases, creep rate of the fuel increases. Therefore, pellets with small grain size have higher creep rate and better plasticity. Therefore, these pellets will be useful to reduce the PCMI. On the other hand, pellet with large grain size is beneficial to reduce the fission gas release. In developing thermal reactor fuels for high burn-up, this factor should be taken into consideration. The question being asked is whether the microstructure can be tailored for irradiation hardening, fracture resistance, fission-gas release. This paper deals with the role played by microstructure for better irradiation performance. (author)

  18. Optimization of Additive-Powder Characteristics for Metallic Micro-Cell UO{sub 2} Fuel Pellet Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong-Joo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Oh, Jang Soo; Yang, Jae Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The improvement in the thermal conductivity of the UO{sub 2} fuel pellet can enhance the fuel performance in various aspects. The mobility of the fission gases is reduced by the lower temperature gradient in the UO{sub 2} fuel pellet. That is to say, the capability of the fission gas retention of the fuel pellet can increase. In addition, the lower centerline temperature of the fuel pellet affects the accident tolerance for nuclear fuel as well as the enhancement of fuel safety and fuel pellet integrity under normal operation conditions. The nuclear reactor power can be uprated owing to the higher safety margin. Thus, many researches on enhancing the thermal conductivity of a nuclear fuel pellet for LWRs have been performed. Typically, an enhancement of the thermal conductivity of the UO{sub 2} fuel pellet can be obtained by the addition of a higher thermal conductive material in the fuel pellet. To maximize the effect of the thermal conductivity enhancement, a continuous and uniform channel of the thermal conductive material in the UO{sub 2} matrix must be formed. To enhance the thermal conductivity of a UO{sub 2} fuel pellet, the development of fabrication process of a Cr metallic micro-cell UO{sub 2} pellet with a continuous and uniform channel of the Cr metallic phase was carried out. The formation of the Cr-oxide phases was prevented and the uniformity of the Cr-metal phase distribution was enhanced simultaneously, through the optimization of the additive-powder characteristics. In the results, the Cr metallic micro-cell pellet with continuous and uniform Cr metallic channel could be obtained.

  19. Progress in researches on MOX fuel pellet producing technology in China

    International Nuclear Information System (INIS)

    Hu Xiaodan

    2010-01-01

    Being the key section of nuclear-fuel cycle, the producing technology of MOX(UO 2 -PuO 2 ) fuel had driven to maturity in France, England, Russia, Belgium, etc. MOX fuel had been applied in FBR and LWR successfully in those countries. With the rapidly developing of nuclear-generated power, the MOX fuel for FBR and LWR was active demanded in China. However, the producing technology of MOX fuel developed slowly. During the period of 'the seventh five year's project', MOX fuel pellet was produced by mechanically mixed method and oxalate deposited method, respectively. Parts of cool performance of MOX fuel pellet produced by oxalate deposited method reached the qualification of fuel for FBR. During the period of 'the ninth five year's project' and 'the tenth five year's project', the technical route of producing MOX fuel was determined, and the test line of producing MOX fuel was built preliminarily. In the same time, the producing technology and analyzing technology of MOX fuel pellet by mechanically mixed was studied roundly, and the representative analogue pellet(UO 2 -CeO 2 ) was produced. That settled the supporting technology for the commercial process and research of MOX fuel rod and MOX fuel module. (authors)

  20. Theory of the frictional interaction between nuclear fuel cladding and a cracked ceramic pellet

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1976-02-01

    A summary is presented of the outcome of theoretical work detailed in five publications, reproduced as appendices, which is concerned with the tendency for the cladding tube of nuclear fuel elements to fracture as the result of power cycling or after a sudden upward power excursion. The relationship is shown between the properties of the clad, those of UO 2 pellets, and the tendency of the clad to fail during upward power excursions. The role of interfacial friction is explored and the benefit to be obtained by reducing it is calculated for cases where a soft metal interlayer is present. It is shown that the experimentally-confirmed magnitude of the strain-concentration in the arc of cladding over a radial pellet crack could not arise if there were interfaceons present. Accordingly, these defects, although they do occur in some sliding situations, are thought to be absent from the pellet clas interface in fuel pins. (author)

  1. Development of railgun pellet injector for nuclear fusion fueling

    International Nuclear Information System (INIS)

    Azuma, Kingo; Oda, Yasushi; Onozuka, Masanori.

    1996-01-01

    Recent fusion plasmas have become larger as fusion research progresses. This requires high-velocity solid-hydrogen pellet injection that is the most efficient fueling method. The application of the electro-magnetic railgun system for pellet injection is one of the most feasible technologies for accelerating a pellet to a high speed. The system consists of a pneumatic pre-accelerator for the first acceleration stage and a railgun for the second stage. The railgun is operated by a low voltage discharged from a pulse-forming-network power supply to accelerate a plasma armature between the rail electrodes. The plasma is induced by high-power laser beam irradiation. The highest velocity of a solid-hydrogen pellet obtained using the railgun was 2.6 km/s. This velocity is higher than the maximum pellet velocity of 2.3 km/s achieved by MHI's pneumatic pellet injector. It was also found that the pellet velocity could be controlled easily using railgun pellet injection. (author)

  2. Development of railgun pellet injector for nuclear fusion fueling

    Energy Technology Data Exchange (ETDEWEB)

    Azuma, Kingo [Mitsubishi Heavy Industries Ltd., Takasago, Hyogo Takasago Research and Development Center (Japan); Oda, Yasushi; Onozuka, Masanori

    1996-03-01

    Recent fusion plasmas have become larger as fusion research progresses. This requires high-velocity solid-hydrogen pellet injection that is the most efficient fueling method. The application of the electro-magnetic railgun system for pellet injection is one of the most feasible technologies for accelerating a pellet to a high speed. The system consists of a pneumatic pre-accelerator for the first acceleration stage and a railgun for the second stage. The railgun is operated by a low voltage discharged from a pulse-forming-network power supply to accelerate a plasma armature between the rail electrodes. The plasma is induced by high-power laser beam irradiation. The highest velocity of a solid-hydrogen pellet obtained using the railgun was 2.6 km/s. This velocity is higher than the maximum pellet velocity of 2.3 km/s achieved by MHI`s pneumatic pellet injector. It was also found that the pellet velocity could be controlled easily using railgun pellet injection. (author).

  3. Development of machine vision system for PHWR fuel pellet inspection

    Energy Technology Data Exchange (ETDEWEB)

    Kamalesh Kumar, B.; Reddy, K.S.; Lakshminarayana, A.; Sastry, V.S.; Ramana Rao, A.V. [Nuclear Fuel Complex, Hyderabad, Andhra Pradesh (India); Joshi, M.; Deshpande, P.; Navathe, C.P.; Jayaraj, R.N. [Raja Ramanna Centre for Advanced Technology, Indore, Madhya Pradesh (India)

    2008-07-01

    Nuclear Fuel Complex, a constituent of Department of Atomic Energy; India is responsible for manufacturing nuclear fuel in India . Over a million Uranium-di-oxide pellets fabricated per annum need visual inspection . In order to overcome the limitations of human based visual inspection, NFC has undertaken the development of machine vision system. The development involved designing various subsystems viz. mechanical and control subsystem for handling and rotation of fuel pellets, lighting subsystem for illumination, image acquisition system, and image processing system and integration. This paper brings out details of various subsystems and results obtained from the trials conducted. (author)

  4. Boosting nuclear fuels

    International Nuclear Information System (INIS)

    Demarthon, F.; Donnars, O.; Dupuy-Maury, F.

    2002-01-01

    This dossier gives a broad overview of the present day status of the nuclear fuel cycle in France: 1 - the revival of nuclear power as a solution to the global warming and to the increase of worldwide energy needs; 2 - the security of uranium supplies thanks to the reuse of weapon grade highly enriched uranium; 3 - the fabrication of nuclear fuels from the mining extraction to the enrichment processes, the fabrication of fuel pellets and the assembly of fuel rods; 4 - the new composition of present day fuels (UO x and chromium-doped pellets); 5 - the consumption of plutonium stocks and the Corail and Apa fuel assemblies for the reduction of plutonium stocks and the preservation of uranium resources. (J.S.)

  5. DUPIC fuel irradiation test and performance evaluation; the performance analysis of pellet-cladding contact fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ho, K. I.; Kim, H. M.; Yang, K. B.; Choi, S. J. [Suwon University, Whasung (Korea)

    2002-04-01

    Thermal and mechanical models were reviewed, and selected for the analysis of nuclear fuel performance in reactor. 2 dimensional FEM software was developed. Thermal models-gap conductances, thermal conductivity of pellets, fission gas release, temperature distribution-were set and packaged into a software. Both thermal and mechanical models were interrelated to each other, and the final results, fuel performance during irradiation is obtained by iteration calculation. Also, the contact phenomena between pellet and cladding was analysed by mechanical computer software which was developed during this work. dimensional FEM program was developed which estimate the mechanical behavior and the thermal behaviors of nuclear fuel during irradiation. Since there is a importance during the mechanical deformation analysis in describing pellet-cladding contact phenomena, simplified 2 dimensional calculation method is used after the contact. The estimation of thermal fuel behavior during irradiation was compared with the results of other. 8 refs., 17 figs. (Author)

  6. Remote nuclear green pellet processing system

    International Nuclear Information System (INIS)

    Cellier, Francis.

    1980-01-01

    An automated system for manufacturing nuclear fuel pellets for use in nuclear fuel elements of nuclear power reactors is described. The system comprises process components arranged vertically but not directly under each other within a single enclosure. The vertical-lateral arrangement provides for gravity flow of the product from one component to the next and for removal of each component without interference with the other components. The single enclosure eliminates time consuming transfer between separate enclosures of each component while providing three-sided access to the component through glove ports. (auth)

  7. Modelling the role of pellet crack motion in the (r-θ) plane upon pellet-clad interaction in advanced gas reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Haynes, T.A. [Centre for Nuclear Engineering & Department of Materials, Imperial College London, Exhibition Rd., London SW7 2AZ (United Kingdom); Ball, J.A. [EDF Energy, Barnett Way, Gloucester GL4 3RS (United Kingdom); Wenman, M.R., E-mail: m.wenman@imperial.ac.uk [Centre for Nuclear Engineering & Department of Materials, Imperial College London, Exhibition Rd., London SW7 2AZ (United Kingdom)

    2017-04-01

    Highlights: • Finite element modelling of pellet relocation in the (r-θ) plane of nuclear fuel. • ‘Soft’ and ‘hard’ PCI have been predicted in a cracked nuclear fuel pellet. • Stress concentration in the cladding ahead of radial pellet cracks is predicted. • The model is very sensitive to the coefficient of friction and power ramp duration. • The model is less sensitive to the number of cracks assumed. - Abstract: A finite element model of pellet fragment relocation in the r-θ plane of advanced gas-cooled reactor (AGR) fuel is presented under conditions of both ‘hard’ and ‘soft’ pellet-clad interaction. The model was able to predict the additional radial displacement of fuel fragments towards the cladding as well as the stress concentration on the inner surface resulting from the azimuthal motion of pellet fragments. The model was subjected to a severe ramp in power from both full power and after a period of reduced power operation; in the former, the maximum hoop stress in the cladding was found to be increased by a factor of 1.6 as a result of modelling the pellet fragment motion. The pellet-clad interaction was found to be relatively insensitive to the number of radial pellet crack. However, it was very sensitive to both the coefficient of friction used between the clad and pellet fragments and power ramp duration.

  8. Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

    Science.gov (United States)

    El-Genk, Mohamed S.

    1991-01-01

    The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept.

  9. Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

    International Nuclear Information System (INIS)

    El-genk, M.S.

    1991-01-01

    The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept

  10. An evaluation of UO2-CNT composites made by SPS as an accident tolerant nuclear fuel pellet and the feasibility of SPS as an economical fabrication process for the nuclear fuel cycle

    Science.gov (United States)

    Cartas, Andrew R.

    The innovative and advanced purpose of this study is to understand and establish proper sintering procedures for Spark Plasma Sintering process in order to fabricate high density, high thermal conductivity UO2 -CNT pellets. Mixing quality and chemical reactions have been investigated by field emission scanning electron microscopy (FESEM), wavelength dispersive spectroscopy (WDS), and X-ray diffraction (XRD). The effect of various types of CNTs on the mixing and sintering quality of UO2-CNT pellets with SPS processing have been examined. The Archimedes Immersion Method, laser flash method, and FE-SEM will be used to investigate the density, thermal conductivity, grain size, pinning effects, and CNT dispersion of fabricated UO2-CNT pellets. Pre-fabricated CNT's were added to UO 2 powder and dispersed via sonication and/or ball milling and then made into composite nuclear pellets. An investigation of the economic impact of SPS on the nuclear fuel cycle for producing pure and composite UO2 fuels was conducted.

  11. Nuclear fuel pellet inspection system

    International Nuclear Information System (INIS)

    Ahmed, H.J.; Beatty, J.M.; Kugler, R.W.

    1992-01-01

    At least one axially extending linear portion of the peripheral surface of the pellet is optically sensed, a set of digital values representative of the pellet surface is generated, and the set is compared to a predetermined standard. Groups of adjacent locations on the surface of the pellet having values greater or less than the predetermined standard are identified, and the pellet is rejected, when a flawed area exceeds a predetermined size. During inspection, the pellet is moved axially through an inspection station by parallel support rolls, spaced by a distance less than the pellet diameter. The rolls are rotated upward and outward from each other, rotating the pellet, and chain dogs are positioned between the spaced rolls for engaging a pellet and moving it along the rolls. The pellet is rejected if its peripheral surface area is too great, and a reference pellet may be used. (author)

  12. A computerised automatic pellet inspection unit for FBTR fuel

    International Nuclear Information System (INIS)

    Ramakumar, M.S.; Mahule, K.N.; Ghosh, J.K.; Venkatesh, D.

    1984-01-01

    Physical inspection and certification of nuclear reactor fuel element components is an activity demanding utmost imagination and skill in devising accurate measuring systems. There is also need for remote handling, automation, rapid processing and inspection data print out when dealing with reactor fuel material. This report deals with an automatic computerised fuel pellet inspection system that has been developed in Radiometallurgy Division, B.A.R.C. to carry out dimensional and weight measurements on fuel pellets for the Fast Breeder Test Reactor (FBTR) at Kalpakkam near Madras. The system consists of several subsystems each developed especially for a specific purpose and as such items are not available off the shelf from manufacturers in India. If a general approach is adopted towards the report, there are many innovations and ideas that can be used in the automatic inspection of a variety of products in industry. As the system is fairly involved the report does not attempt to deal with detailed description of the equipment. The function of the system is to accept a certain quantity of fuel pellets in a bowl feeder, separate the pellets rejected owing to their exceeding dimensional and weight limits and form columns of accepted pellets. Dimensional and weight limits can be set as required and all inspection data are presented in a printed format. The system processes pellets at the rate of 15 per minute. The entire system can be run by operators with no special skills. The unit is currently in use for the inspection of mixed carbide fuel pellets for FBTR. (author)

  13. Study on the development of coating technology for UO{sub 2} nuclear fuel pellet and the microstructural observation of the coated layer

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong; Song, Moon Sup; Cho, In Sik; Kim Yu Sin; Lim Young Kyun [Sunmoon University, Asan (Korea)

    1998-04-01

    In order to enhance inherent safety of UO{sub 2} nuclear fuel pellet and develop future nuclear fuel technology, a coating method for the preparation multi-layers of pyrolytic carbon and silicon carbide on the fuel was developed. Inner pyrolytic carbon layer and outer silicon layer were prepared by thermal decomposition of propane in a fluidized bed type CVD unit and silane in ECR PECVD, respectively. Combustion reaction between two layers resulted in forming silicon carbide layer. The morphology depended on the initial carbon shape. Phase identification and microstructural analysis of the combustion product with XRD, AES, SEM and TEM showed that final products of inner layer and outer layer were pyrolytic carbon with isotropic structure and fine crystalline {beta}-SiC, respectively. This coating process is very useful for the fabrication of coated UO{sub 2} nuclear fuel pellet an future nuclear fuel fabrication technology. (author). 45 refs., 47 figs., 5 tabs.

  14. Technical Issues in the development of high burnup and long cycle fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Joo; Yang, Jae Ho; Oh, Jang Soo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Nam, Ik Hui [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Over the last half century, a nuclear fuel cycle, a fuel discharged burnup and a uranium enrichment of the LWR (Light Water Reactor) fuel have continuously increased. It was the efforts to reduce the LWR fuel cycle cost, and to make reactor operation more efficiently. Improved fuel and reactor performance contribute further to the reduction and management efficiency of spent fuels. The primary incentive for operating nuclear reactor fuel to higher burnup and longer cycle is the economic benefits. The fuel cycle costs could be reduced by extending fuel discharged burnup and fuel cycle length. The higher discharged burnup can increase the energy production per unit fuel mass or fuel assembly. The longer fuel cycle can increase reactor operation flexibility and reduce the fuel changing operation and the spent fuel management burden. The margin to storage capacity limits would be also increased because high burnup and long cycle fuel reduces the mass of spent fuels. However, increment of fuel burnup and cycle length might result in the acceleration of material aging consisting fuel assembly. Then, the safety and integrity of nuclear fuel will be degraded. Therefore, to simultaneously enhance the safety and economics of the LWR fuel through the fuel burnup and cycle extension, it is indispensable to develop the innovative nuclear fuel material concepts and technologies which can overcome degradation of fuel safety. New fuel research project to extend fuel discharged burnup and cycle length has been launched in KAERI. Main subject is to develop innovative LWR fuel pellets which can provide required fuel performance and safety at extended fuel burnup and cycle length. In order to achieve the mission, we need to know that what the impediments are and how to break through current limit of fuel pellet properties. In this study, the technical issues related to fuel pellets at high burnup were surveyed and summarized. We have collected the technical issues in the literatures

  15. Technical Issues in the development of high burnup and long cycle fuel pellets

    International Nuclear Information System (INIS)

    Kim, Dong Joo; Yang, Jae Ho; Oh, Jang Soo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Nam, Ik Hui

    2012-01-01

    Over the last half century, a nuclear fuel cycle, a fuel discharged burnup and a uranium enrichment of the LWR (Light Water Reactor) fuel have continuously increased. It was the efforts to reduce the LWR fuel cycle cost, and to make reactor operation more efficiently. Improved fuel and reactor performance contribute further to the reduction and management efficiency of spent fuels. The primary incentive for operating nuclear reactor fuel to higher burnup and longer cycle is the economic benefits. The fuel cycle costs could be reduced by extending fuel discharged burnup and fuel cycle length. The higher discharged burnup can increase the energy production per unit fuel mass or fuel assembly. The longer fuel cycle can increase reactor operation flexibility and reduce the fuel changing operation and the spent fuel management burden. The margin to storage capacity limits would be also increased because high burnup and long cycle fuel reduces the mass of spent fuels. However, increment of fuel burnup and cycle length might result in the acceleration of material aging consisting fuel assembly. Then, the safety and integrity of nuclear fuel will be degraded. Therefore, to simultaneously enhance the safety and economics of the LWR fuel through the fuel burnup and cycle extension, it is indispensable to develop the innovative nuclear fuel material concepts and technologies which can overcome degradation of fuel safety. New fuel research project to extend fuel discharged burnup and cycle length has been launched in KAERI. Main subject is to develop innovative LWR fuel pellets which can provide required fuel performance and safety at extended fuel burnup and cycle length. In order to achieve the mission, we need to know that what the impediments are and how to break through current limit of fuel pellet properties. In this study, the technical issues related to fuel pellets at high burnup were surveyed and summarized. We have collected the technical issues in the literatures

  16. The study on the methods for improving the gredibility of NDT equipment for the gap of pellets of nuclear fuel rods

    International Nuclear Information System (INIS)

    Zhang Lei; Liu Ming; Wang Changhong; Ma Jinbo

    2014-01-01

    In order to improve the credibility of the new generation of automatic online non-destructive testing equipment for the gap of the pellets of nuclear fuel rods the researchers have done a lot of work in the development of the device. Such measures as multi-thread synchronization, precise timing, upper and lower computer communication control, antijamming processing are adopted such that the detecting device can accurately detecte the size of the gap between pellets, the position and length of the spring cavity at the front end of the nuclear fuel rods at a detection rate of 8 m/min. The detection credibility for the 0.5 mm gap is over 95%, reaching the international advanced level. At present, the device is put into use in the nuclear fuel element production line. (authors)

  17. Review of pellet fueling

    International Nuclear Information System (INIS)

    Turnbull, R.J.

    1978-01-01

    Fusion reactors based on the Tokamak concept (possibly mirrors, too) will require a low energy method of fueling. Refueling by using solid pellets of hydrogen isotopes appears to be the most promising low energy technique. The main issue in assessing the feasibility of pellet fueling is the ability of the pellet to penetrate into the central region of the reactor. A review is presented of the various theories predicting the lifetime of the pellet and their regions of applicability. Among the phenomena considered are neutral ablation of the solid, ionized ablation of the solid, shielding of the pellet by neutral molecules and electrons and ions, flow of the ablation cloud, distortion of the magnetic field by the flow of an ionized ablation cloud, and charging and electrostatic shielding of the pellet. A brief summary of results of experiments done by the University of Illinois-Oak Ridge and Riso groups is presented. The results of these experiments indicate that, at least at the low temperatures and densities used, a neutral ablation-neutral shielding model is correct. Finally, since all indications are that in order for pellet fueling to be successful, high velocity pellets will be needed, a brief discussion of possible acceleration techniques is presented

  18. Nuclear fuel element

    International Nuclear Information System (INIS)

    Yamamoto, Seigoro.

    1994-01-01

    Ultrafine particles of a thermal neutron absorber showing ultraplasticity is dispersed in oxide ceramic fuels by more than 1% to 10% or lower. The ultrafine particles of the thermal neutron absorber showing ultrafine plasticity is selected from any one of ZrGd, HfEu, HfY, HfGd, ZrEu, and ZrY. The thermal neutron absorber is converted into ultrafine particles and solid-solubilized in a nuclear fuel pellet, so that the dispersion thereof into nuclear fuels is made uniform and an absorbing performance of the thermal neutrons is also made uniform. Moreover, the characteristics thereof, for example, physical properties such as expansion coefficient and thermal conductivity of the nuclear fuels are also improved. The neutron absorber, such as ZrGd or the like, can provide plasticity of nuclear fuels, if it is mixed into the nuclear fuels for showing the plasticity. The nuclear fuel pellets are deformed like an hour glass as burning, but, since the end portion thereof is deformed plastically within a range of a repulsive force of the cladding tube, there is no worry of damaging a portion of the cladding tube. (N.H.)

  19. High performance reliability fuel pellet

    International Nuclear Information System (INIS)

    Beuchel, P.H.; Lee, Y.C.

    1989-01-01

    A fuel pellet for a nuclear reactor fuel rod is described comprising: a substantially cylindrical central section; a convex first end section smoothly joined to one axial end of the central section at a first junction, the first junction approximating a smooth and continuous curved surface; a concave second end section joined to the central section at a second junction, the second junction approximating a smooth and continuous curved surface, wherein the curvature of the concave second end section is conformed to the curvature of the convex first end section

  20. Introducing wood pellet fuel to the UK

    Energy Technology Data Exchange (ETDEWEB)

    Cotton, R A; Giffard, A

    2001-07-01

    Technical and non-technical issues affecting the introduction of wood pellet-fired heating to the UK were investigated with the aim of helping to establish a wood pellet industry in the UK. The project examined the growth and status of the industry in continental Europe and North America, reviewed relevant UK standards and legislation, identified markets for pellet heating in the UK, organised workshops and seminars to demonstrate pellet burning appliances, carried out a trial pelletisation of a range of biomass fuels, helped to set up demonstration installations of pellet-fired appliances, undertook a promotional campaign for wood pellet fuel and compiled resource directories for pellet fuel and pellet burning appliances in the UK. The work was completed in three phases - review, identification and commercialisation. Project outputs include UK voluntary standards for wood pellet fuel and combustion appliances, and a database of individuals with an interest in wood pellet fuel.

  1. Pelletizing of NaF granules as adsorbent for fluorides of nuclear fuel materials, (1)

    International Nuclear Information System (INIS)

    Kimura, Syojiro; Tsutsui, Tenson; Kanagawa, Akira.

    1976-01-01

    The pelletizing of NaF granules on laboratory scale used for recovery and purification of fluorides of nuclear fuel materials are investigated, particularity experimental studies on how to pelletize, fluidity of granules and flow rate of the granules in mortar of tablet machine are carried out. As result, added quantity of H 2 O as binder at granulation process of spherical NaF granule influence to yield of NaF granules. The proper quantity of H 2 O to NaF powder is about 18--19%. Friction coefficient of unfixed shape NaF granule is higher than that of spherical granule, in opposition, density of the former granule is lower. So fluidity of unfixed shape NaF granule is low, flow rate of the granules through feeder of tablet machine is slow, and arching on mortar of the tablet machine occurs. Then, some further operation for getting better fluidity needs for pelletizing of NaF granules using unfixed shape granule. The other hand, using spherical NaF granule, continuous pelletizing by commercial tablet machine can be carried out. (auth.)

  2. Development of a manufacturing process of (Th,U)O2 sintered pellets to be used as nuclear fuel

    International Nuclear Information System (INIS)

    Neto Ferreira, R.A.; Santos, A.M. dos; Lameiras, F.S.; Cardoso, P.E.

    1989-01-01

    The R and D result of a reliable manufacturing process of sintered (Th,U)O 2 pellets meeting the operational requirements of pressurized light water nuclear reactors is presented. Available technologies were used as much as possible. The R and D effort was directed to perform the required adaptations. The gel precipitation process was adapted successfully to the specific requirements of direct pressing and sintering. This was done mainly by adjusting the composition of the feed solution. The direct pressing and sintering parameters could be kept almost unchanged in relation to the manufacturing of UO 2 pellets. The design criteria of the (Th,U)O 2 nuclear fuel for pressurized light water reactors were identified and settled in the specification for this fuel. This R and D work was made jointly with the Kernforschungsanlage - Juelich, NUKEM and SIEMENS, Group KWU [pt

  3. Optimal rate of power increase in nuclear fuel. Pellet behaviour under dynamic conditions

    International Nuclear Information System (INIS)

    Karlsson, B.G.

    1976-05-01

    A mathematical model has been worked out for the determination of the optium power escalation rate for nuclear power plants from the view-pint of fuel integrity. The model calculates the stress and strain transients in the pellet-cladding system with rapid power increase. No burnup effects are included due to the short time scale involved. An elastic solution has been transposed to a linear viscoelastic one using the correspondence principle. The cladding has however been treated under the programme assumptions as purely elastic. The fuel material has been assumed to be completely relaxed prior to the power transient. Radial cracking is included. The UO 2 -material distortion has been assumed to be linear viscoelastic, while the dilation is assumed as elastic. The system has been treated assuming plane strain since friction between the pellet and the cladding is large with practical burnsups, and the pellet column can be regarded as infinitely long, compared to the diameter of the pellet. The results of the calculations made show that under the above assumptions the clad stress is independent of the rate of power increase in the pellet. Scince this result is in opposition to general opinion an experimental programme was performed in order to test the results of the model. These results were confirmed. The occurance of clad failures in practice is not determined purely by clad straining. Current thought pays attention to the influence of e.g. stress-corrosion phenomena as significant. The programme reported here pays no attention such-like effects, or the effects of clad creep which could be of considerable significance with local deformations. These later effects are receiving attention in work now being initiated at the Department.(author)

  4. Wood pellets : a worldwide fuel commodity

    International Nuclear Information System (INIS)

    Melin, S.

    2005-01-01

    Aspects of the wood pellet industry were discussed in this PowerPoint presentation. Details of wood pellets specifications were presented, and the wood pellet manufacturing process was outlined. An overview of research and development activities for wood pellets was presented, and issues concerning quality control were discussed. A chart of the effective calorific value of various fuels was provided. Data for wood pellet mill production in Canada, the United States and the European Union were provided, and various markets for Canadian wood pellets were evaluated. Residential sales as well as Canadian overseas exports were reviewed. Production revenues for British Columbia and Alberta were provided. Wood pellet heat and electricity production were discussed with reference to prefabricated boilers, stoves and fireplaces. Consumption rates, greenhouse gas (GHG) emissions, and fuel ratios for wood pellets and fossil fuels were compared. Price regulating policies for electricity and fossil fuels have prevented the domestic expansion of the wood pellet industry. There are currently no incentives for advanced biomass combustion to enter British Columbia markets, and this has led to the export of wood pellets. It was concluded that climate change mitigation policies will be a driving force behind market expansion for wood pellets. tabs., figs

  5. Numerical analysis of the influence of the fuel pellet shape on the pellet-cladding contact condition

    International Nuclear Information System (INIS)

    Marajofsky, Adolfo; Denis, Alicia C.; Soba, Alejandro

    2004-01-01

    One of the problems of greater concern in nuclear fuels operation is that of pellet-cladding interaction (PCI), since it may be cause of fuel failure. In unfailed claddings, the occurrence of contact with the pellet is generally evidenced by a typical deformation pattern known as bamboo effect. In the present work different pellets' shapes are proposed, all of them with a chamfer next to the top and bottom surfaces. The performance of these pellets design is simulated with a numerical code, DIONISIO, previously developed in this working group, which makes use of the finite elements method. It provides the temperature, stress and strain distribution and the inventory of fission gases by analyzing phenomena like thermal expansion, elasticity, plasticity, creep, irradiation growth, PCI, swelling and densification. The pellets' design tested are grouped into two types: those with a straight chamfer running from the central pellet plane to both extremes (R-type pellets) and those with the chamfer occupying one quarter of the pellet's height leaving a central ring of the standard, cylindrical shape (M-type pellets). Different chamfer depths were numerically tested. It was found that the gap increase associated with the introduction of a deep chamfer is responsible for a significant temperature increment. But chamfers which leave a gap of 110 to 150 μm (assuming a normal fuel element with a gap 90 μm thick) gave place to pellets with an adequate thermal response and, moreover, the disappearance of the bamboo effect or even the appearance of an inverse effect, that is, pellets which make contact with the cladding in the region around its middle plane. (author) [es

  6. Acoustic emission from fuel pellets in a simulated reactor environment

    International Nuclear Information System (INIS)

    Kupperman, D.S.; Kennedy, C.R.; Reimann, K.J.

    1977-01-01

    Thermal-shock damage of nuclear reactor fuel pellets in a simulated reactor environment has been correlated with acoustic-emission data obtained from sensors placed on extensions of the electrical feedthroughs. Ringdown counts, rms output data, and event-location data has been acquired for experiments carried out with single pellets as well as multiple pellet stacks. These tests have shown that acoustic-emission monitoring can provide information indicating the onset and the extent of cracking

  7. Model for the behaviour of thorium and uranium fuels at pelletization

    International Nuclear Information System (INIS)

    Ferreira Neto, Ricardo Alberto

    2000-11-01

    In this work, a model for the behaviour of thorium-uranium-mixed oxide microspheres in the pelletizing process is presented. This model was developed in a program whose objective was to demonstrate the viability of producing fissile material through the utilization of thorium in pressurized water reactors. This is important because it allows the saving of the strategic uranium reserves, and makes it possible the nuclear utilization of the large brazilian thorium reserves. The objective was to develop a model for optimizing physical properties of the microspheres, such as density, fracture strength and specific surface, so as to produce fuel pellets with microstructure, density, open porosity and impurity content, in accordance with the fuel specification. And, therefore, to adjust the sol-gel processing parameters in order to obtain these properties, and produce pellets with an optimized microstructure, adequate to a stable behaviour under irradiation. The model made it clear that to achieve this objective, it is necessary to produce microspheres with density and specific surface as small as possible. By changing the sol-gel processing parameters, microspheres with the desired properties were produced, and the model was experimentally verified by manufacturing fuel pellets with optimized microstructures, density, open porosity and impurity content, meeting the specifications for this new nuclear fuel for pressurized water reactors. Furthermore it was possible to obtain mathematical expressions that enables to calculate from the microspheres properties and the utilized compaction pressure, the sinter density that will be obtained in the sintered pellet and the necessary compaction pressure to reach the sintered density specified for the fuel. (author)

  8. Influence of pellet-clad-gap-size on LWR fuel rod performance

    International Nuclear Information System (INIS)

    Brzoska, B.; Fuchs, H.P.; Garzarolli, F.; Manzel, R.

    1979-01-01

    The as-fabricated pellet-clad-gap size varies due to fabricational tolerances of the cladding inner diameter and the pellet outer diameter. The consequences of these variations on the fuel rod behaviour are analyzed using the KWU fuel rod code CARO. The code predictions are compared with experimental results of special pathfinder test fuel rods irradiated in the OBRIGHEIM nuclear power plant. These test fuel rods include gap sizer in the range of 140 μm to 270 μm, prepressurization between 13 bar to 36 bar and Helium and Argon fill gases irradiated up to a local burnup of 35 MWd/kg(U). Post irradiation examination were performed at different burnups. CARC calculations have been performed with special emphasis in cladding creep down, fission gas release and pellet clad gap closure. (orig.)

  9. Evaluation of the in pile performance of boron containing fuel pellets

    International Nuclear Information System (INIS)

    Jeong, Gwanyoon; Sohn, Dongseong

    2012-01-01

    The world rare earth resource are heavily concentrated in certain area and if these natural resources are weaponized by a country, we may confront serious difficulty because rare earth element gadolinium(Gd) is used as burnable poison material in some nuclear power plants (NPP) in Korea. Gd is used as a neutron absorbing material in Gd 2 O 3 form and mixed with UO 2 When boron is used as burnable poison in nuclear fuel, in fuel pellets. The burnable poison mixed in the fuel pellets is called integral burnable absorber (BA) design which differentiates it from the old separate BA design. In the old separate BA design, boron(B) was used in borosilicate glass (PYREX) form and placed in guide tubes. With the development of the concern over the availability of rare earth material Gd, B is considered as a candidate material replacing Gd for the case when the rare earth material is weaponized. However the idea for new boron BA design is integral type because the integral type BA design has several benefits over the separate BA design, such as reduction of radioactive waste, more positions for BA location, etc. 10 B absorbs a neutron and produces helium by the following reaction: 10 B + n → 7 Li + 4 He The helium produced by the nuclear reaction may cause the increase of rod internal pressure and change the gap conductivity if the significant amount of helium gas is released to the gap between the pellet and the cladding. Thus, it is necessary to investigate the in-pile behaviors of B containing pellet. However, few experiment have been carried out so far on the behavior of in-pile produced helium in UO 2 fuel pellets, especially for the cases boron compound is mixed with UO 2 In this paper, we will evaluate the production and the release of helium depending on fuel. 10 B concentration in the fuel

  10. Improvement of the center boring device for the irradiated fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Usami, Koji; Onozawa, Atsushi; Kimura, Yasuhiko; Sakuraba, Naotoshi; Shiina, Hidenori; Harada, Akito; Nakata, Masahito [Japan Atomic Energy Agency, Nuclear Science Research Inst., Tokai, Ibaraki (Japan)

    2012-03-15

    The power ramp tests performed at JMTR in Oarai R and D Center are objected to study the safety margin of the high burnup fuels. One of the important parameters measured during this test is the center temperature of the fuel pellet. For this measurement, a thermocouple is installed into the hole bored at the pellet center by the center boring device, which can fix the fuel pellet with the frozen CO{sub 2} gas during its boring process. At the Reactor Fuel Examination Facility (RFEF) in Tokai R and D Center, several improvements were applied for the previous boring device to gain its performance and reliability. The major improvements are the change of the drill bit, modification of the boring process and the optimization of the remote operability. The mock-up test will be performed with the irradiated fuel pellet to confirm the benefit of improvement. This study was conducted under a contract with the Nuclear and Industrial Safety Agency (NISA) of the Ministry of Economy, Trade and Industry (METI). (author)

  11. Fuel pellet fracture and relocation

    International Nuclear Information System (INIS)

    Walton, L.A.; Husser, D.L.

    1983-01-01

    The model used to describe fuel pellet fracture and relocation is an important feature of a fuel performance computer code. This model becomes especially important if the computer code is principally to be used for the evaluation of pellet clad interaction. The fracture and relocation model being developed for the B and W fuel performance code FUMAC was derived from an extensive data base. Cross sections of irradiated fuel rods were photographically magnified and measured to determine the configuration of the fragments of the fractured fuel pellets. Data, representing a wide range of LWR fuel designs and as-manufactured mechanical configurations, were catalogued and systematically reduced and then correlated as a function of the likely independent variables. These correlations define the key phenomenological behavior patterns which the relocation model must duplicate and indicate which mechanistic approaches are viable explanations of this behavior. The data base covers the burnup range from approximately one to 35 GWd/mtU and linear heat rates from less than 100 to nearly 700 W/Cm. This paper presents the correlated data base and the methods used to derive and interpret it. It was determined from this data base that pellet cracking is initially both power level and burnup dependent but tends to saturate eventually with continued steady irradiation. Fuel pellet relocation was found to be much more extensive than would be deduced from thermal considerations alone. Even at very low burnups fuel fragments were found to move outward until restrained by the cladding. The results also suggest that changes in internal resistance to heat flow within the pellets due to the opening of cracks may be as important as peripheral gap changes to the thermal modeler. The transient response and thermal implications of this model are recommended as primary areas for future investigation

  12. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.

    1981-01-01

    A nuclear fuel loading apparatus, incorporating a microprocessor control unit, is described which automatically loads nuclear fuel pellets into dual fuel rods with a minimum of manual involvement and in a manner and sequence to ensure quality control and accuracy. (U.K.)

  13. Development of uranium dioxide fuel pellets with addition of beryllium oxide for increasing of thermal conductivity

    International Nuclear Information System (INIS)

    Queiroz, Carolinne Mol; Ferreira, Ricardo Alberto Neto

    2011-01-01

    The CDTN - Centro de Desenvolvimento de Tecnologia Nuclear presents a project named 'Beryllium Project' viewing to increasing the thermal conductivity of UO 2 fuel pellets, increasing the lifetime of those pellets in the reactor, generating a greater economy. This increase of conductivity is obtained by means of Be O addition to the UO 2 fuel pellets, which is very used for the production of nuclear energy. The UO 2 pellets however present a thermal conductivity relatively low, generating a high temperature gradient between the center and his side surface. The addition of beryllium oxide, with higher thermal conductivity gives pellets which will present lower temperature gradient and, consequently, more durability and better utilization of energy potential of the pellet in the reactor. (author)

  14. Developments in MOX fuel pellet fabrication technology: Indian experience

    International Nuclear Information System (INIS)

    Kamath, H.S.; Majumdar, S.; Purusthotham, D.S.C.

    1998-01-01

    India is interested in mixed oxide (MOX) fuel technology for better utilisation of its nuclear fuel resources. In view of this, a programme involving MOX fuel design, fabrication and irradiation in research and power reactors has been taken up. A number of experimental irradiations in research reactors have been carried out and a few MOX assemblies of ''All Pu'' type have been loaded in our commercial BWRs at Tarapur. An island type of MOX fuel design is under study for use in PHWRs which can increase the burn-up of the fuel by more than 30% compared to natural UO 2 fuel. The MOX fuel pellet fabrication technology for the above purpose and R and D efforts in progress for achieving better fuel performance are described in the paper. The standard MOX fuel fabrication route involves mechanical mixing and milling of UO 2 and PuO 2 powders. After detailed investigations with several types of mixing and milling equipments, dry attritor milling has been found to be the most suitable for this operation. Neutron Coincident Counting (NCC) technique was found to be the most convenient and appropriate technique for quick analysis of Pu content in milled MOX powder and to know Pu mixing is homogenous or not. Both mechanical and hydraulic presses have been used for powder compaction for green pellet production although the latter has been preferred for better reproducibility. Low residue admixed lubricants have been used to facilitate easy compaction. The normal sintering temperature used in Nitrogen-Hydrogen atmosphere is between 1600 deg. C to 1700 deg. C. Low temperature sintering (LTS) using oxidative atmospheres such as carbon dioxide, Nitrogen and coarse vacuum have also been investigated on UO 2 and MOX on experimental scale and irradiation behaviour of such MOX pellets is under study. Ceramic fibre lined batch furnaces have been found to be the most suitable for MOX pellet production as they offer very good flexibility in sintering cycle, and ease of maintainability

  15. DUPIC nuclear fuel manufacturing and process technology development

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Park, J. J.; Lee, J. W.

    2000-05-01

    In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated

  16. DUPIC nuclear fuel manufacturing and process technology development

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Park, J. J.; Lee, J. W. [and others

    2000-05-01

    In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated.

  17. Nuclear fuel element

    International Nuclear Information System (INIS)

    Hirama, H.

    1978-01-01

    A nuclear fuel element comprises an elongated tube having upper and lower end plugs fixed to both ends thereof and nuclear fuel pellets contained within the tube. The fuel pellets are held against the lower end plug by a spring which is supported by a setting structure. The setting structure is maintained at a proper position at the middle of the tube by a wedge effect caused by spring force exerted by the spring against a set of balls coacting with a tapered member of the setting structure thereby wedging the balls against the inner wall of the tube, and the setting structure is moved free by pushing with a push bar against the spring force so as to release the wedge effect

  18. 3D modeling of missing pellet surface defects in BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, B.W., E-mail: Benjamin.Spencer@inl.gov; Williamson, R.L.; Stafford, D.S.; Novascone, S.R.; Hales, J.D.; Pastore, G.

    2016-10-15

    Highlights: • A global/local analysis procedure for missing pellet surface defects is proposed. • This is applied to defective BWR fuel under blade withdrawal and high power ramp conditions. • Sensitivity of the cladding response to key model parameters is studied. - Abstract: One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can be used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed here. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding

  19. UO2 Fuel pellet impurities, pellet surface roughness and n(18O)/n(16O) ratios, applied to nuclear forensic science

    International Nuclear Information System (INIS)

    Pajo, L.

    2001-01-01

    In the last decade, law enforcement has faced the problem of illicit trafficking of nuclear materials. Nuclear forensic science is a new branch of science that enables the identification of seized nuclear material. The identification is not based on a fixed scheme, but further identification parameters are decided based on previous identification results. The analysis is carried out by using traditional analysis methods and applying modern measurement technology. The parameters are generally not unambiguous and not self-explanatory. In order to have a full picture about the origin of seized samples, several identification parameters should be used together and the measured data should be compared to corresponding data from known sources. A nuclear material database containing data from several fabrication plants is installed for the purpose. In this thesis the use of UO 2 fabrication plant specific parameters, fuel impurities, fuel pellet surface roughness and oxygen isotopic ratio in UO 2 were investigated for identification purposes in nuclear forensic science. The potential use of these parameters as 'fingerprints' is discussed for identification purposes of seized nuclear materials. Impurities of the fuel material vary slightly according to the fabrication method employed and a plant environment. Here the impurities of the seized UO 2 were used in order to have some clues about the origin of the fuel material by comparing a measured data to nuclear database information. More certainty in the identification was gained by surface roughness of the UO 2 fuel pellets, measured by mechanical surface profilometry. Categories in surface roughness between a different fuel element type and a producer were observed. For the time oxygen isotopic ratios were determined by Thermal Ionisation Mass Speckometry (TIMS). Thus a TIMS measurement method, using U 16 O + and U 18 0 + ions, was developed and optimised to achieve precise oxygen isotope ratio measurements for the

  20. Recycling of nuclear fuel swarf at the fabrication of UO sub(2)-pellets and its influence on the irradiation behavior

    International Nuclear Information System (INIS)

    Dias, M.S.; Lameiras, F.S.; Santos, A.M.M. dos

    1991-01-01

    From the fabrication of UO sub(2) pellets for light water reactor fuel rods, nuclear fuel scraps results in form of UO sub(2) grinding swarf and UO sub(2) sinter scraps oxidized to U sub(3)O sub(8) powder. Detailed investigations on five types of UO sub(2) pellets fabricated with different portions of this scrap kinds added to the UO sub(2) press powder showed that there is only a small influence of such scrap additions on the irradiation behavior, especially for the fission gas release. This allows to recycle the fabrication scrap in a simple and economic way. (author)

  1. Present status of laser fusion fuel pellet

    International Nuclear Information System (INIS)

    Nakai, Sadao; Mima, Kunioki; Norimatsu, Takayoshi; Takagi, Masaru.

    1986-01-01

    Accompanying the advance of pellet implosion experiment, the data base required for fuel pellet design has been steadily accumulated. The clarification of the physics related to the process of absorbing laser beam, energy transport, the generation of ablative pressure, the hydrodynamic mechanism of implosion, the energy transmission to fuel core and so on progressed, and the design data supported by these results are prepared. Based on the data base like this, the design of fuel pellets taking the optimization of implosion in consideration is carried out. The various fuel pellets designed in this way are tested for their effectiveness by implosion experiment. For this purpose, the high performance measurement of implosion and the high accuracy manufacture of fuel pellets become very important. In this paper, the present state of the research on the method of laser implosion, the example of pellet design and the law of proportion, the manufacturing techniques of the fuel pellets having various structures, the techniques dealing with tritium and so on is summarized, and the direction of future research and development is ascertained. At present, implosion experiment is carried out mostly by hanging a pellet target with a fiber of several μm diameter, but the fiber impairs the symmetry of implosion. The levitation techniques without contact is required. (Kako, I.)

  2. Investigation and recovery of unrecovered fuel pellets and cladding tube pieces

    International Nuclear Information System (INIS)

    Kobayashi, Keiji

    1980-01-01

    The total weight of the fuel pellets lost due to break was about 1206 g, and cladding tube pieces were about 217 g. Among these, the pellets of about 527 g and the cladding tube pieces of about 152 g were recovered when broken fuel rods were discovered. It is not desirable to leave these broken pieces as unrecovered in view of safety and the management of nuclear fuel materials. Kansai Electric Power Co., Inc., investigated the position and the amount of these pellets and cladding tube pieces for about a year, and recovered a part of them. The results were written in two reports. The objects of the investigation and recovery, and the method of recovery are explained. The UO 2 and zirconium recovered were 58.52 g and 369.58 g, respectively. The solid pellets were recovered from the reactor, fuel assemblies, a spent fuel pit and canals, and the content in sludge was recovered from other installations. The amounts of unrecovered pellets and cladding tube pieces in primary cooling water, coolant filters, sealing water filters, primary cooling pipes, waste resins and fuel assemblies were estimated. The problems concerning the recovery and estimation are pointed out. The results of estimating the amount of uranium in coolant filters and sealing water filters are useful to know the time of the occurrence of accident. (Kako, I.)

  3. Mixed oxide fuel pellet and manufacturing method thereof

    International Nuclear Information System (INIS)

    Yuda, Ryoichi; Ito, Ken-ichi; Masuda, Hiroshi.

    1993-01-01

    In a method of manufacturing nuclear fuel pellets which comprises compression molding a mixed oxide powder containing UO 2 and PuO 2 followed by sintering, a sintering agent having a composition comprising about 40 to 80 wt% of SiO 2 and the balance of Al 2 O 3 is mixed to a mixed oxide at a ratio of about 40ppm to about 0.5 wt% based on the total amount of the mixed oxide and the sintering agent, to prepare a mixture. The mixture is molded into a compression product and then sintered at a weakly acidic atmosphere at a temperature of about 1500degC to 1800degC. With such procedures, the sintering agent forms an eutectic product of a single liquid phase, PuO 2 is dispersed over the entire region of the pellet by way of the liquid phase, formation of a solid solution phase is promoted to annihilate a free PuO 2 phase. Further, growth of crystal grains is promoted. Accordingly, since the MOX fuel pellets prepared according to the present invention have a uniform solid solution state, and no free PuO 2 phase remains, increase of FP gas emission due to local nuclear fission of Pu can be avoided. (T.M.)

  4. Measurement of nuclear reaction rates and spectral indices along the radius of fuel pellets from IPEN/MB-01 reactor; Medidas de taxas de reacao nuclear e de indices espectrais ao longo do raio das pastilhas combustiveis do reator IPEN/MB-01

    Energy Technology Data Exchange (ETDEWEB)

    Mura, Luis Felipe Liambos

    2010-07-01

    This work presents the measurements of the nuclear reaction rates along the radial direction of the fuel pellet by irradiation and posterior gamma spectrometry of a thin slice of fuel pellet of UO{sub 2} with 4,3% enrichment. From its irradiation the rate of radioactive capture and fission have been measured as a function of the radius of the pellet disk using a HPGe detector. Lead collimators has been used for this purpose. Simulating the fuel pellet in the pin fuel of the IPEN/MB-01 reactor, a thin UO{sub 2} disk is used. This disk is inserted in the interior of a dismountable fuel rod. This fuel rod is then placed in the central position of the IPEN/MB-01 reactor core and irradiated during 1 hour under a neutron flux of around 9 x 10{sup 8} n/cm{sup 2}s. For gamma spectrometry 10 collimators with different diameters have been used, consequently, the nuclear reactions of radioactive capture that occurs in atoms of {sup 238}U and fissions that occur on both {sup 235}U and {sup 238}U are measured in function of 10 different region (diameter of collimator) of the fuel pellet disk. Corrections in the geometric efficiency due to introduction of collimators on HPGe detection system were estimated using photon transport of MCNP-4C code. Some calculated values of nuclear reaction rate of radioactive capture and fission along of the radial direction of the fuel pellet obtained by Monte Carlo methodology, using the MCNP-4C code, are presented and compared to the experimental data showing very good agreement. Besides nuclear reaction rates, the spectral indices {sup 28{rho}} and {sup 25{delta}} have been obtained at each different radius of the fuel pellet disk. (author)

  5. Apparatus for unloading nuclear fuel pellets from a sintering boat

    International Nuclear Information System (INIS)

    Bucher, G.D.; Raymond, T.E.

    1987-01-01

    An apparatus is described for unloading nuclear fuel pellets from a loaded sintering boat having an open top, comprising: (a) means for receiving the boat in an upright position with the pellets contained therein, the boat receiving means including a platform for supporting the loaded boat in the upright position, the boat supporting platform having first and second portions; (b) means for clamping the boat including a pair of plates disposed at lateral sides of the boat and being movable in a first direction relative to one another for applying clamping forces to the boat on the platform and in a second direction relative to one another for releasing the clamping forces from the boat. The pair of plates have inner surfaces facing toward one another, the first and second platform portions of the boat supporting platform being mounted to the plates on the respective facing surfaces thereof and disposed in a common plane. One of the plates and one of the platform portions mounted thereto are disposed in a stationary position and the other of the plates and the other of the platform portions mounted thereto are movable relative thereto in the first and second directions for applying and releasing clamping forces to and from the boat while the boat is supported in the upright position by the platform portions; (c) means for transferring the clamped boat from the upright position to an inverted position and then back to the upright position; and (d) means of receiving the pellets from the clamped boat as the boat is being transferred from the upright position to the inverted position

  6. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1984-01-01

    The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  7. Spin-polarized fuel in ICF pellets

    International Nuclear Information System (INIS)

    Wakuta, Yoshihisa; Emoto, Nobuya; Nakao, Yasuyuki; Honda, Takuro; Honda, Yoshinori; Nakashima, Hideki.

    1990-01-01

    The use of parallel spin-polarized DT or D 3 He fuel increases the fusion cross-section by 50%. By implosion-burn simulation for inertially confined fusion (ICF) pellets of the spin-polarized fuels, we found that the input energy requirement could be reduced by nearly a fact of two. These pellets taken up here include large-high-aspect-ratio DT target proposed in ILE Osaka University and DT ignitor/D 3 He fuel pellet proposed by our group. We also found that the polarized state could survive during the implosion-burn phase. (author)

  8. Recent activities of the nuclear fuel technology department of Cekmece Nuclear Research and Training Center

    International Nuclear Information System (INIS)

    Can, S.; Aybers, T.; Kopuz, B.

    1997-01-01

    The Nuclear Fuel Technology Department (NFTD) in CNRTC is a unique unit in Turkey in charge of performing all activities in nuclear fuel field. It has a pilot plant on uranium refining and conversion to UO 2 since 1986. Presently, its R and D activities are focused on pellet manufacturing and characterization: UO 2 , ThO 2 and (Th,U)O 2 . The studies on thorium dioxide fuel include to obtain ThO 2 pellets from thorium nitrate and mixed (Th,U)O 2 pellets. A study on evaluation of different fuel cycle options in accordance with nuclear energy planning in Turkey is also going on. (author)

  9. Fuel compliance model for pellet-cladding mechanical interaction

    International Nuclear Information System (INIS)

    Shah, V.N.; Carlson, E.R.

    1985-01-01

    This paper describes two aspects of fuel pellet deformation that play significant roles in determining maximum cladding hoop strains during pellet-cladding mechanical interaction: compliance of fragmented fuel pellets and influence of the pellet end-face design on the transmission of axial compressive force in the fuel stack. The latter aspect affects cladding ridge formation and explains several related observations that cannot be explained by the hourglassing model. An empirical model, called the fuel compliance model and representing the above aspects of fuel deformation, has been developed using the results from two Halden experiments and incorporated into the FRAP-T6 fuel performance code

  10. Water chemistry control in thermal and nuclear power plants. 9. Nuclear fuel management

    International Nuclear Information System (INIS)

    2008-01-01

    The chemical management of fuels in nuclear power plants aims at maintenance of the soundness of nuclear fuels and at reduction of the radiation exposure of the working employees. With regard to the former, particular attention should be paid to the fabrication process of fuel assembly, mainly for chemical management for fuel cladding tubes together with fuel pellet-clad chemical interactions, and to the outer tubes in the power plants. With regard to the latter, the fabrication process should be carefully controlled to prevent radioactive impurity increase in primary cooling water systems by maintaining cleaning level and decreasing surface contamination. Reactions of zircalloy with water or hydrogen forming ZrH 2 , sintered density of UO 2 pellet controlling water content, pellet-clad interactions, stress corrosion cracking, crud induced fuel failure, behaviors of such fission products as I, Xe, Kr, and Cs in plants are also important to water and chemical management of nuclear fuels. (S. Ohno)

  11. Method for distinguishing fuel pellets

    International Nuclear Information System (INIS)

    Sagami, Masaharu; Kurihara, Kunitoshi.

    1978-01-01

    Purpose: To distinguish correctly and efficiently the kind of fuel substance enclosed in a cladding tube. Method: Elements such as manganess 55, copper 65, vanadium 51, zinc 64, scandium 45 and the like, each having a large neutron absorption cross section and discharging gamma rays of inherent bright line spectra are applied to or mixed in fuel pellets of different kinds in uranium enrichment degree, plutonium concentration, burnable poison concentration or the like. These fuel rods are irradiated with neutron beams, and energy spectra of gamma rays discharged upon this occasion are observed to carry out distinguishing of fuel pellets. (Aizawa, K.)

  12. Detailed Reaction Kinetics for CFD Modeling of Nuclear Fuel Pellet Coating for High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Battaglia, Francine

    2008-01-01

    The research project was related to the Advanced Fuel Cycle Initiative and was in direct alignment with advancing knowledge in the area of Nuclear Fuel Development related to the use of TRISO fuels for high-temperature reactors. The importance of properly coating nuclear fuel pellets received a renewed interest for the safe production of nuclear power to help meet the energy requirements of the United States. High-temperature gas-cooled nuclear reactors use fuel in the form of coated uranium particles, and it is the coating process that was of importance to this project. The coating process requires four coating layers to retain radioactive fission products from escaping into the environment. The first layer consists of porous carbon and serves as a buffer layer to attenuate the fission and accommodate the fuel kernel swelling. The second (inner) layer is of pyrocarbon and provides protection from fission products and supports the third layer, which is silicon carbide. The final (outer) layer is also pyrocarbon and provides a bonding surface and protective barrier for the entire pellet. The coating procedures for the silicon carbide and the outer pyrocarbon layers require knowledge of the detailed kinetics of the reaction processes in the gas phase and at the surfaces where the particles interact with the reactor walls. The intent of this project was to acquire detailed information on the reaction kinetics for the chemical vapor deposition (CVD) of carbon and silicon carbine on uranium fuel pellets, including the location of transition state structures, evaluation of the associated activation energies, and the use of these activation energies in the prediction of reaction rate constants. After the detailed reaction kinetics were determined, the reactions were implemented and tested in a computational fluid dynamics model, MFIX. The intention was to find a reduced mechanism set to reduce the computational time for a simulation, while still providing accurate results

  13. Pellet dimension checker

    International Nuclear Information System (INIS)

    Marmo, A.R.

    1980-01-01

    A pellet dimension checker was developed for use in making nuclear-fuel pellets. This checker eliminates operator handling of the pellet but permits remote-monitoring of the operation, and is thus suitable for mass production of green fuel pellets particularly in reprocessing plants handling irradiated uranium or plutonium. It comprises a rotatable arm for transferring a pellet from a conveyor to several dimensional measuring stations and back to the conveyor if the dimensions of the pellet are within predetermined limits. If the pellet is not within the limits, the arm removes the pellet from the process stream. (DN)

  14. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1981-01-01

    An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  15. The microstructure of fuel pellets as object of quality characterization on base of FMEA analysis

    International Nuclear Information System (INIS)

    Goncharov, U.V.; Matveev, A.A.; Strucov, A.V.; Loktev, I.I.

    2012-01-01

    It is difficult to find new effective reserves in nuclear fuel production as its experience of production and operation become more and more. FMEA method can help it on base of the system analysis. The state corporation Rosatom, consistently pursuing a policy of economical manufacture, make all efforts for identification of deep dependences between conditions of manufacture, characteristics of fuel materials and features of their operational behaviour. This report continues earlier discussion of the important feature of produced nuclear fuel pellets grain size distribution. This distribution defines gas release in reactor and has not appropriate method of characterization. There are descriptions of optimal microstructure of fuel pellets with large grain size literature

  16. Method of manufacturing nuclear fuel elements

    International Nuclear Information System (INIS)

    Ishida, Masao; Oguma, Masaomi.

    1980-01-01

    Purpose: To effectively prevent the bending of nuclear fuel elements in the reactor by grinding the end faces of pellets due to their mutual sliding. Method: In the manufacturing process of nuclear fuel elements, a plurality of pellets whose sides have been polished are fed one by one by way of a feeding mechanism through the central aperture in an electric motor into movable arms and retained horizontally with the central axis by being held on the side. Then, the pellet held by one of the arms is urged to another pellet held by the other of the arms by way of a pressing mechanism and the mating end faces of both of the pellets are polished by mutual sliding. Thereafter, the grinding dusts resulted are eliminated by drawing pressurized air and then the pellets are enforced into a cladding tube. Thus, the pellets are charged into the cladding tube with both polished end faces being contacted to each other, whereby the axial force is uniformly transmitted within the end faces to prevent the bending of the cladding tube. (Kawakami, Y.)

  17. Nuclear Fuel elements

    International Nuclear Information System (INIS)

    Hirakawa, Hiromasa.

    1979-01-01

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

  18. Sphere-pac versus pellet UO2 fuel in de Dodewaard BWR

    International Nuclear Information System (INIS)

    Linde, A. van der.

    1989-04-01

    Comparative testing of UO 2 sphere-pac and pellet fuel rods under LWR conditions has been jointly performed by the Netherlands Utilities Research Centre (KEMA) in Arnhem, the Netherlands Energy Research Foundation (ECN) at Petten and the Netherlands Joint Nuclear Power Utility (GKN) at Dodewaard. This final report summarizes the highlights of this 1968-1988 program with strong emphasis on the fuel rods irradiated in the Dodewaard BWR. The conclusion reached is that under normal LWR conditions sphere-pac UO 2 in LWR fuel rods offers better resistance against stress corrosion cracking of the cladding, but that under fast, single step, power ramping conditions pellet UO 2 in LWR fuel rods has a better resistance against hoop stress failure of the cladding. 128 figs., 36 refs., 19 tabs

  19. The high temperature out-of-pile test of LVDT for elongation measurement of fuel pellet

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Jo, M. S.; Joo, K. N.; Park, S. J.; Gang, Y. H.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the elongation measurement technique of the fuel pellet is being developed using LVDT(Linear Variable Differential Transformer). The well qualified out-of-pile test were needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation instrumented capsule, because LVDT is very sensitive to variation of temperature. Therefore, the high temperature out-of-pile test system for fuel pellet elongation was developed, and this test was performed under the temperature condition between room temperature and 300 .deg. C with increasing the elongation from 0 to 5 mm. The LVDT's high temperature characteristics and temperature sensitivity of LVDT were analyzed through this experiment. Based on the result of this test, the method for the application of LVDT and elongation detector at high temperature was introduced. It is known that the results will be used to predict accurately the elongation of fuel pellet during irradiation test.

  20. Fuel Pellets Production from Biodiesel Waste

    Directory of Open Access Journals (Sweden)

    Kawalin Chaiyaomporn

    2010-01-01

    Full Text Available This research palm fiber and palm shell were used as raw materials to produce pelletised fuel, and waste glycerol were used as adhesive to reduce biodiesel production waste. The aim of this research is to find optimum ratio of raw material (ratio of palm fiber and palm shell, raw material size distribution, adhesive temperature, and ratio of ingredients (ratio of raw material, waste glycerol, and water. The optimum ratio of pelletized fuel made only by palm fiber was 50:10:40; palm fiber, water, and waste glycerol respectively. In the best practice condition; particle size was smaller than 2 mm, adhesive glycerol was heated. From the explained optimum ratio and ingredient, pelletizing ratio was 62.6%, specific density was 982.2 kg/m3, heating value was 22.5 MJ/kg, moisture content was 5.9194%, volatile matter was 88.2573%, fix carbon content was 1.5894%, and ash content was 4.2339% which was higher than the standard. Mixing palm shell into palm fiber raw material reduced ash content of the pellets. The optimum raw material ratio, which minimizes ash content, was 80 to 20 palm fiber and palm shell respectively. Adding palm shell reduced ash content to be 2.5247% which was higher than pelletized fuel standard but followed cubed fuel standard. At this raw material ratio, pelletizing ratio was 70.5%, specific density was 774.8 kg/m3, heating value was 19.71 MJ/kg, moisture content was 9.8137%, volatile matter was 86.2259%, fix carbon content was 1.4356%, and compressive force was 4.83 N. Pelletized fuel cost at optimum condition was 1.14 baht/kg.

  1. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Hayashi, Hiroshi; Watari, Yoshio; Hizahara, Hiroshi; Masuoka, Ryuzo.

    1970-01-01

    When exchanging nuclear fuel assemblies during the operation of a nuclear reactor, melting of fuel bodies, and severence of tubular claddings is halted at the time of insertion by furnishing a neutron absorbing material such as B 10 , Cd, Gd or the like at the forward end of the fuel assembly to thereby lower the power peak at the forward ends of the fuel elements to within tolerable levels and thus prevent both fuel liquification and excessive expansion. The neutron absorbing material may be attached in the form of a plate to the fuel assembly forward tie plate, or may be inserted as a pellet into the front end of the tubular cladding. (Owens, K.J.)

  2. Effects of variations in fuel pellet composition and size on mixed-oxide fuel pin performance

    International Nuclear Information System (INIS)

    Makenas, B.J.; Jensen, B.W.; Baker, R.B.

    1980-10-01

    Experiments have been conducted which assess the effects on fuel pin performance of specific minor variations from nominal in both fuel pellet size and pellet composition. Such pellets are generally referred to in the literature as rogue pellets. The effect of these rogue pellets on fuel pin and reactor performance is shown to be minimal

  3. Fabrication of nano-structured UO2 fuel pellets

    International Nuclear Information System (INIS)

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

    2007-01-01

    Nano-structured materials have received much attention for their possibility for various functional materials. Ceramics with a nano-structured grain have some special properties such as super plasticity and a low sintering temperature. To reduce the fuel cycle costs and the total mass of spent LWR fuels, it is necessary to extend the fuel discharged burn-up. In order to increase the fuel burn-up, it is important to understand the fuel property of a highly irradiated fuel pellet. Especially, research has focused on the formation of a porous and small grained microstructure in the rim area of the fuel, called High Burn-up Structure (HBS). The average grain size of HBS is about 300nm. This paper deals with the feasibility study on the fabrication of nano-structured UO 2 pellets. The nano sized UO 2 particles are prepared by a combined process of a oxidation-reducing and a mechanical milling of UO 2 powder. Nano-structured UO 2 pellets (∼300nm) with a density of ∼93%TD can be obtained by sintering nano-sized UO 2 compacts. The SEM study reveals that the microstructure of the fabricated nano-structure UO 2 pellet is similar to that of HBS. Therefore, this bulk nano-structured UO 2 pellet can be used as a reference pellet for a measurement of the physical properties of HBS

  4. Particle fueling experiments with a series of pellets in LHD

    Science.gov (United States)

    Baldzuhn, J.; Damm, H.; Dinklage, A.; Sakamoto, R.; Motojima, G.; Yasuhara, R.; Ida, K.; Yamada, H.; LHD Experiment Group; Wendelstein 7-X Team

    2018-03-01

    Ice pellet injection is performed in the heliotron Large Helical Device (LHD). The pellets are injected in short series, with up to eight individual pellets. Parameter variations are performed for the pellet ice isotopes, the LHD magnetic configurations, the heating scenario, and some others. These experiments are performed in order to find out whether deeper fueling can be achieved with a series of pellets compared to single pellets. An increase of the fueling efficiency is expected since pre-cooling of the plasma by the first pellets within a series could aid deeper penetration of later pellets in the same series. In addition, these experiments show which boundary conditions must be fulfilled to optimize the technique. The high-field side injection of pellets, as proposed for deep fueling in a tokamak, will not be feasible with the same efficiency in a stellarator or heliotron because there the magnetic field gradient is smaller than in a tokamak of comparable size. Hence, too shallow pellet fueling, in particular in a large device or a fusion reactor, will be an issue that can be overcome only by extremely high pellet velocities, or other techniques that will have to be developed in the future. It turned out by our investigations that the fueling efficiency can be enhanced by the injection of a series of pellets to some extent. However, further investigations will be needed in order to optimize this approach for deep particle fueling.

  5. Uranium dioxide and beryllium oxide enhanced thermal conductivity nuclear fuel development

    International Nuclear Information System (INIS)

    Andrade, Antonio Santos; Ferreira, Ricardo Alberto Neto

    2007-01-01

    The uranium dioxide is the most used substance as nuclear reactor fuel for presenting many advantages such as: high stability even when it is in contact with water in high temperatures, high fusion point, and high capacity to retain fission products. The conventional fuel is made with ceramic sintered pellets of uranium dioxide stacked inside fuel rods, and presents disadvantages because its low thermal conductivity causes large and dangerous temperature gradients. Besides, the thermal conductivity decreases further as the fuel burns, what limits a pellet operational lifetime. This research developed a new kind of fuel pellets fabricated with uranium dioxide kernels and beryllium oxide filling the empty spaces between them. This fuel has a great advantage because of its higher thermal conductivity in relation to the conventional fuel. Pellets of this kind were produced, and had their thermophysical properties measured by the flash laser method, to compare with the thermal conductivity of the conventional uranium dioxide nuclear fuel. (author) (author)

  6. Expandable device for a nuclear fuel rod

    International Nuclear Information System (INIS)

    Gesinski, L.T.

    1978-01-01

    A nuclear fuel rod and a device for use within the rod cladding to maintain the axial position of the fuel pellets stacked one atop another within the cladding are described. The device is initially of a smaller external cross-section than the fuel rod cladding internal cross-section so as to accommodate loading into the rod at preselected locations. During power operation the device responds to a rise in temperature, so as to permanently maintain its position and restrain any axial motion of the fuel pellets

  7. Study of pellet clad interaction defects in Dresden-3 fuel rods

    International Nuclear Information System (INIS)

    Pasupathi, V.; Perrin, J.S.

    1979-01-01

    During Cycle-3 operation of Dresden-3, fuel rod failures occurred following a transient power increase. Ten fuel rods from five of the leaking fuel assemblies were examined at Battelle's Columbus Laboratory and General Electric-Vallecitos Nuclear Center. Examinations consisted of nondestructive and destructive methods including metallography and scanning electron microscopy (SEM). Results showed the cause of fuel rod failure to be pellet clad interaction involving stress corrosion cracking. Results of SEM studies of the cladding crack surfaces and deposits on clad inner surfaces were in agreement with those reported by other investigators

  8. Pellets - A fuel with a future

    International Nuclear Information System (INIS)

    2004-01-01

    This special brochure presents a series of articles on the topic of wood pellets as a fuel of the future. Dr. Walter Steinmann, director of the Swiss Federal Office of Energy (SFOE) introduces the topic, stressing that the Swiss Confederation and the Cantons are supporting efforts to increase the sustainable use of wood fuels. Further articles take a closer look at pellets and their form. Pellets-fired heating units are introduced as a viable alternative to traditional oil-fired units. Tips are presented on the various ways of storing pellets. Quality-assurance aspects are examined and manufacturers and distributors of wood pellets are listed. A further article takes a closer look at a large Swiss manufacturer of pellets and describes the production process used as well as the logistics necessary for the transportation of raw materials and finished products. The brochure also presents a selection of pellet ovens and heating systems from various manufacturers. A further article illustrates the use of pellets as a means of heating apartment blocks built to the MINERGIE low-energy-consumption standard. In the example quoted, the classic combination of solar energy for the pre-heating of hot water and pellets for the central heating and hot water supply is used. The use of a buried spherical tank to store pellets - and thus the saving of space inside the building - is described in a further article that takes a look at the refurbishment of the heating system in a single-family home. Finally, various contributions presented at the Pellets Forum held in Berne in November 2003 are summarised in a short article

  9. U.S. Department Of Energy's nuclear engineering education research: highlights of recent and current research-II. 5. Automation of Nuclear Fuel Pellet Quality Control

    International Nuclear Information System (INIS)

    Keyvan, Shahla; Song, Xiaolong

    2001-01-01

    At the present time, nuclear fuel pellet inspection is performed by humans using the naked eye for judgment and decision making as to whether to accept or reject the pellet. Unnecessary re-fabrication of pellets will be costly, and having too many low-quality pellets in a fuel assembly is unacceptable. The current practice of pellet inspection by humans is tedious and subject to inconsistencies and error. In addition, manual inspection is cumbersome since the inspector must keep the pellet at arm's length and must wear glasses to protect the lenses of his or her eyes. The pellets are taken from a pellet sizing machine, dumped onto a rack, and shaken into rows; they are then viewed as a group. The entire group is rotated 90 deg four times to provide the inspector with a 360-deg view of each pellet. The pellets are examined for certain types of cracks, chips, and unusual markings, i.e., water stains and machine banding. These defects appear at any location on the pellet surface image with different intensity, size, shape, and background noise. Figure 1 shows typical defective fuel pellets with chip, banded, and end defects. The goal of this work is to automate the pellet inspection process. A prototype of such an inspection system is developed. The system examines photographic images of pellets using various artificial intelligence techniques for image analysis and defect classification. Figure 2 shows the user interface of this inspection system, which is built using Java programming language. A total of 252 pellets with various defects was available for this research. Each pellet was photographed four times at rotations of 90 deg. The resultant black-and-white negatives were scanned into the computer in 256 gray scale mode. The inspection of a fuel pellet by image analysis involves several steps, as described in Fig. 3 and as follows: Step 1-On-line image conversion: This process involves on-line digitization of the input image. Step 2-Reference model: The second

  10. Study on dynamic measurement of fuel pellet length during loading into cladding tube

    International Nuclear Information System (INIS)

    Zhang Kai

    1993-09-01

    Various methods are presented for measuring the pellet length in the cladding tube (zirconium tube) during the loading process of the preparation of single rod of nuclear fuel assembly. These methods are used in former Soviet Union, west European countries and China in the manufacturing of nuclear power plant element. Different methods of dynamic measurement by using mechanics, optics and electricity and their special features are analysed and discussed. The structure and measuring principle of a developed measuring device,and its measuring precision and system deviation are also introduced. Finally, the length of loaded pellets is checked with analog pellets. The results are as expected and show that the method and principle used in the measuring device are feasible. It is an ideal and advanced method for the pellet loading of single cladding tube. The principle mentioned above can also be used in other industries

  11. Process for automatic filling of nuclear fuel rod cans

    International Nuclear Information System (INIS)

    Bezold, H.

    1977-01-01

    A drying section is inserted in the production line for the automation of the filling process for fuel rods with nuclear fuel pellets. The pellets are taken in a drum magazine to a drying furnace and then pushed out one after the other into the can to be filled. (TK) [de

  12. Pellet fueling development at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Foster, C.A.; Milora, S.L.; Schuresko, D.D.; Combs, S.K.; Lunsford, R.V.

    1982-01-01

    A pellet injector development program has been under way at the Oak Ridge National Laboratory (ORNL) since 1976 with the goals of developing D 2 , T 2 pellet fuel injectors capable of reliable repetitive fueling of reactors and of continued experimentation on contemporary plasma devices. The development has focused primarily on two types of injectors that show promise. One of these injectors is the centrifuge-type injector, which accelerates pellets in a high speed rotating track. The other is the gas or pneumatic gun, which accelerates pellets in a gun barrel using compressed helium of H 2 gas

  13. Method and apparatus for sizing nuclear fuel rod cladding tubes

    International Nuclear Information System (INIS)

    Koehler, L.

    1976-01-01

    Nuclear fuel rod cladding tubes are sized internally to diameters precisely fitting nuclear fuel pellets with which the tubes are charged by externally applying hydraulic pressure to short lengths of each tube. The pressure is applied while the tube is stationary. The tube is then moved to bring a new length within the hydraulic pressure zone. The volume of the hydraulic liquid used and the pressure applied to this liquid is such that the liquid is compressed slightly so that the length being sized yields, the expansion of the liquid then completing the sizing. The lengths being sized step-by-step are internally supported by either the fuel pellets or a mandrel having the same diameter as the pellets

  14. Steam-treated wood pellets: Environmental and financial implications relative to fossil fuels and conventional pellets for electricity generation

    International Nuclear Information System (INIS)

    McKechnie, Jon; Saville, Brad; MacLean, Heather L.

    2016-01-01

    Highlights: • Steam-treated pellets can greatly reduce greenhouse gas emissions relative to coal. • Cost advantage is seen relative to conventional pellets. • Higher pellet cost is more than balanced by reduced retrofit capital requirements. • Low capacity factors further favour steam-treated pellets over conventional pellets. - Abstract: Steam-treated pellets can help to address technical barriers that limit the uptake of pellets as a fuel for electricity generation, but there is limited understanding of the cost and environmental impacts of their production and use. This study investigates life cycle environmental (greenhouse gas (GHG) and air pollutant emissions) and financial implications of electricity generation from steam-treated pellets, including fuel cycle activities (biomass supply, pellet production, and combustion) and retrofit infrastructure to enable 100% pellet firing at a generating station that previously used coal. Models are informed by operating experience of pellet manufacturers and generating stations utilising coal, steam-treated and conventional pellets. Results are compared with conventional pellets and fossil fuels in a case study of electricity generation in northwestern Ontario, Canada. Steam-treated pellet production has similar GHG impacts to conventional pellets as their higher biomass feedstock requirement is balanced by reduced process electricity consumption. GHG reductions of more than 90% relative to coal and ∼85% relative to natural gas (excluding retrofit infrastructure) could be obtained with both pellet options. Pellets can also reduce fuel cycle air pollutant emissions relative to coal by 30% (NOx), 97% (SOx), and 75% (PM 10 ). Lesser retrofit requirements for steam-treated pellets more than compensate for marginally higher pellet production costs, resulting in lower electricity production cost compared to conventional pellets ($0.14/kW h vs. $0.16/kW h). Impacts of retrofit infrastructure become increasingly

  15. Computerized x-ray radiographic system for fuel pellet measurements

    International Nuclear Information System (INIS)

    Green, D.R.; Karnesky, R.A.; Bromley, C.

    1977-01-01

    The development and operation of a computerized system for determination of fuel pellet diameters from x-ray radiography is described. Actual fuel pellet diameter measurements made with the system are compared to micrometer measurements on the same pellets, and statistically evaluated. The advantages and limitations of the system are discussed, and recommendations are made for further development

  16. Burnable absorber coated nuclear fuel

    International Nuclear Information System (INIS)

    Chubb, W.; Radford, K.C.; Parks, B.H.

    1984-01-01

    A nuclear fuel body which is at least partially covered by a burnable neutron absorber layer is provided with a hydrophobic overcoat generally covering the burnable absorber layer and bonded directly to it. In a method for providing a UO 2 fuel pellet with a zirconium diboride burnable poison layer, the fuel body is provided with an intermediate niobium layer. (author)

  17. Pellet-clad interaction in water reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

  18. Pellet-clad interaction in water reactor fuels

    International Nuclear Information System (INIS)

    2004-01-01

    The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

  19. Radiological surveillance in the nuclear fuel fabrication in Mexico

    International Nuclear Information System (INIS)

    Garcia A, J.; Reynoso V, R.; Delgado A, G.

    1996-01-01

    The objective of this report is to present the obtained results related to the application of the radiological safety programme established at the Nuclear Fuel Fabrication Pilot Plant (NFFPF) in Mexico, such as: surveillance methods, radiological protection criteria and regulations, radiation control and records and the application of ALARA recommendation. During the starting period from April 1994 to April 1995, at the NFFPF were made two nuclear fuel bundles a Dummy and other to be burned up in a BWR the mainly process activities are: UO 2 powder receiving, powder pressing for the pellets formation, pellets grinding, cleaning and drying, loading into a rod, Quality Control testing, nuclear fuel bundles assembly. The NFFPF is divided into an unsealed source area (pellets manufacturing plant) and into a sealed source area (rods fabrication plant). The control followed have helped to detect failures and to improve the safety programme and operation. (authors). 1 ref., 3 figs

  20. Nuclear fuels

    International Nuclear Information System (INIS)

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

    2009-01-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

  1. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  2. Structure change of fuel pellets

    International Nuclear Information System (INIS)

    Imanaka, Tetsuji

    1980-01-01

    The investigation of the broken pieces of fuel rods in Mihama No. 1 reactor was carried out in the Japan Atomic Energy Research Institute, and unexpectedly led to the post-irradiation tests. The investigation group of the Kyoto University Research Institute considers that the pursuit of the causes of accident by the government was insufficient, and the countermeasures are problematical, as the result of having examined various reports. In this study, the white foreign phase and swelling of cladding tubes were investigated, because these are especially important in view of the soundness of the fuel. Besides, the possibility of the oxidation of UO 2 pellets by cooling water was examined. It was found by metallographic test that the featuring phase different from UO 2 structure existed in the central part of pellets remaining in two broken fuel rod pieces. The report of JAERI judged that it is the product of solid phase reaction above a certain threshold temperature. The change of pellet structure observed in the white foreign phase and the swell of a cladding tube was caused by the oxidation of UO 2 pellets by primary coolant. The result of observation of the white foreign phase showed that it had been liquid phase at the time of the formation. From the thermodynamic examination based on oxygen potential, UO 2 is oxidized above 1100 deg C in the atmosphere of primary coolant. The liquid phase of the oxidized phase of UO 2 is formed above 1600 deg C. (Kako, I.)

  3. UO2 fuel pellets fabrication via Spark Plasma Sintering using non-standard molybdenum die

    Science.gov (United States)

    Papynov, E. K.; Shichalin, O. O.; Mironenko, A. Yu; Tananaev, I. G.; Avramenko, V. A.; Sergienko, V. I.

    2018-02-01

    The article investigates spark plasma sintering (SPS) of commercial uranium dioxide (UO2) powder of ceramic origin into highly dense fuel pellets using non-standard die instead of usual graphite die. An alternative and formerly unknown method has been suggested to fabricate UO2 fuel pellets by SPS for excluding of typical problems related to undesirable carbon diffusion. Influence of SPS parameters on chemical composition and quality of UO2 pellets has been studied. Also main advantages and drawbacks have been revealed for SPS consolidation of UO2 in non-standard molybdenum die. The method is very promising due to high quality of the final product (density 97.5-98.4% from theoretical, absence of carbon traces, mean grain size below 3 μm) and mild sintering conditions (temperature 1100 ºC, pressure 141.5 MPa, sintering time 25 min). The results are interesting for development and probable application of SPS in large-scale production of nuclear ceramic fuel.

  4. Trial production of fuel pellet from Acacia mangium bark waste biomass

    Science.gov (United States)

    Amirta, R.; Anwar, T.; Sudrajat; Yuliansyah; Suwinarti, W.

    2018-04-01

    Fuel pellet is one of the innovation products that can be produced from various sources of biomass such as agricultural residues, forestry and also wood industries including wood bark. Herein this paper, the potential fuel pellet production using Acacia mangium bark that abundant wasted from chip mill industry was studied. Fuel pellet was produced using a modified animal feed pellet press machine equipped with rotating roller-cylinders. The international standards quality of fuel pellet such as ONORM (Austria), SS (Sweden), DIN (Germany), EN (European) and ITEBE (Italy) were used to evaluate the optimum composition of feedstock and additive used. Theresults showed the quality offuel pellet produced were good compared to commercial sawdust pellet. Mixed of Acacia bark (dust) with 10% of tapioca and 20% of glycerol (w/w) was increased the stable form of pellet and the highest heating value to reached 4,383 Kcal/kg (calorific value). Blending of Acacia bark with tapioca and glycerol was positively improved its physical, chemical and combustion properties to met the international standards requirement for export market. Based on this finding, production of fuel pellet from Acacia bark waste biomass was promising to be developed as an alternative substitution of fossil energy in the future.

  5. Non destructive examination of UN / U-Si fuel pellets using neutrons (preliminary assessment)

    Energy Technology Data Exchange (ETDEWEB)

    Bourke, Mark Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Voit, Stewart Lancaster [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Losko, Adrian S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tremsin, Anton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-31

    Tomographic imaging and diffraction measurements were performed on nine pellets; four UN/ U Si composite formulations (two enrichment levels), three pure U3Si5 reference formulations (two enrichment levels) and two reject pellets with visible flaws (to qualify the technique). The U-235 enrichments ranged from 0.2 to 8.8 wt.%. The nitride/silicide composites are candidate compositions for use as Accident Tolerant Fuel (ATF). The monophase U3Si5 material was included as a reference. Pellets from the same fabrication batches will be inserted in the Advanced Test Reactor at Idaho during 2016. The goal of the Advanced Non-destructive Fuel Examination work package is the development and application of non-destructive neutron imaging and scattering techniques to ceramic and metallic nuclear fuels. Data reported in this report were collected in the LANSCE run cycle that started in September 2015 and ended in March 2016. Data analysis is ongoing; thus, this report provides a preliminary review of the measurements and provides an overview of the characterized samples.

  6. Fugitive binder for nuclear fuel materials

    International Nuclear Information System (INIS)

    Gallivan, T.J.

    1980-01-01

    A compound consisting of ammonium cations and carbonate, bicarbonate, or carbamate anions, or a mixture of such compounds, is useful as a binder for uranium dioxide fuel pellets for which it is desired to maintain a certain degree of porosity, uniformity of pore size, a lack of interconnections between the pores, and the shape or configuration of the base material particles in the final article after sintering. Upon heating, these binders decompose into gases and leave substantially no impurities. A process for sintering green nuclear fuel pellets using these binders is provided. (LL)

  7. The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Hao, E-mail: jiangh@ornl.gov; Wang, Jy-An John; Wang, Hong

    2016-12-01

    Highlights: • To investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on its dynamic performance. • Flexural rigidity, EI = M/κ, estimated from FEA results were benchmarked with SNF dynamic experimental results, and used to evaluate interface bonding efficiency. • Interface bonding efficiency can significantly dictate the SNF system rigidity and the associated dynamic performance. • With consideration of interface bonding efficiency and fuel cracking, HBU SNF fuel property was estimated with SNF static and dynamic experimental data. - Abstract: Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets to the clad, which results in a reduction in composite rod system flexural rigidity. Therefore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.

  8. Repeating pneumatic hydrogen pellet injector for plasma fueling

    International Nuclear Information System (INIS)

    Combs, S.K.; Milora, S.L.; Foust, C.R.; Foster, C.A.; Schuresko, D.D.

    1985-01-01

    A repeating pneumatic pellet injector has been developed for plasma fueling applications. The repetitive device extends pneumatic injector operation to steady state. The active mechanism consists of an extruder and a gun assembly that are cooled by flowing liquid-helium refrigerant. The extruder provides a continuous supply of solid hydrogen to the gun assembly, where a reciprocating gun barrel forms and chambers cylindrical pellet from the extrusion; pellets are then accelerated with compressed hydrogen gas (pressures up to 125 bar) to velocities -1 have been obtained with 2.1- , 3.4- , and 4.0-mm-diameter pellets. The present apparatus operates at higher firing rates in short bursts; for example, a rate of 6 s -1 for 2 s with the larger pellets. These pellet parameters are in the range applicable for fueling large present-day fusion devices such as the Tokamak Fusion Test Reactor (TFTR). Experimental results are presented, including effects of propellant pressure and barrel length on gun performance

  9. Dispersion fuel for nuclear research facilities

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  10. Development of high burnup nuclear fuel technology

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Kang, Young Hwan; Jung, Jin Gone; Hwang, Won; Park, Zoo Hwan; Ryu, Woo Seog; Kim, Bong Goo; Kim, Il Gone

    1987-04-01

    The objectives of the project are mainly to develope both design and manufacturing technologies for 600 MWe-CANDU-PHWR-type high burnup nuclear fuel, and secondly to build up the foundation of PWR high burnup nuclear fuel technology on the basis of KAERI technology localized upon the standard 600 MWe-CANDU- PHWR nuclear fuel. So, as in the first stage, the goal of the program in the last one year was set up mainly to establish the concept of the nuclear fuel pellet design and manufacturing. The economic incentives for high burnup nuclear fuel technology development are improvement of fuel utilization, backend costs plant operation, etc. Forming the most important incentives of fuel cycle costs reduction and improvement of power operation, etc., the development of high burnup nuclear fuel technology and also the research on the incore fuel management and safety and technologies are necessary in this country

  11. Manufacturing at industrial level of UO2 pellets for the fuel elements of the Atucha I Nuclear Power Plant

    International Nuclear Information System (INIS)

    Dyment, I.G.; Noguera Rojas, Francisco

    1982-01-01

    The interest to produce fuel elements within a policy of self sufficiency arose with the installation of Atucha I. The first steps towards this goal consisted in processing the uranium oxide, transforming it into fuel pellets of high density. The developments towards the fabrication of said pellets, performed by CNEA since 1968, first at a laboratory level and afterwards on an industrial scale, allowed CNEA to obtain its own technological capability to produce 400 kg of UO 2 per day. The fuel pellets manufacturing method developed by CNEA is a powder-metallurgical process, which, besides conventional equipment, involves the use of special equipment that required the performance of systematic testing programmes, as well as special training at operational level. The developed processes respond to a modern and advanced technology. A general scheme of the process, starting with a directly sinterable UO 2 powder, is described, including compacting of the powder into pellets, sintering, control of the temperature in the sintering and reduction zones and of the time of permanence in both zones, and cylindric rectifying of the pellets. During the whole process, specialized personnel controls the operations, after which the material is released by the Quality Control Department. The national contribution to the manufacturing technology of the pellets for fuel elements of power and research reactors was of 100%. (M.E.L.) [es

  12. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Patarin, L.

    2002-01-01

    This book treats of the different aspects of the industrial operations linked with the nuclear fuel, before and after its use in nuclear reactors. The basis science of this nuclear fuel cycle is chemistry. Thus a recall of the elementary notions of chemistry is given in order to understand the phenomena involved in the ore processing, in the isotope enrichment, in the fabrication of fuel pellets and rods (front-end of the cycle), in the extraction of recyclable materials (residual uranium and plutonium), and in the processing and conditioning of wastes (back-end of the fuel cycle). Nuclear reactors produce about 80% of the French electric power and the Cogema group makes 40% of its turnover at the export. Thus this book contains also some economic and geopolitical data in order to clearly position the stakes. The last part, devoted to the management of wastes, presents the solutions already operational and also the research studies in progress. (J.S.)

  13. Effects of pellet shape on the fuel failure behavior under a RIA condition

    International Nuclear Information System (INIS)

    Hosokawa, Takanori; Hoshi, Tsutao; Yanagihara, Satoshi; Iwamura, Takamichi; Orita, Yoshihiko.

    1980-10-01

    The two types of fuel rods with different pellet shaped, i.e. flat pellets and dished pellets, were tested in the NSRR to investigate the effects of pellet shapes on the fuel failure behavior under an RIA condition and the results were compared with those of the chamfered pellet fuel rods which are used as the reference rod in the NSRR experiments. In addition, the deformation of pellets due to thermal expansion is calculated by using an FEM computer code. Through the above results, following conclusions are obtained. (1) In the experiments, insignificant differences on the cladding surface temperature responses and the appearance of post-irradiated rods are observed in each type of rods. (2) Evident differences on the deformation of fuel pellets have not appeared in the calculation. (3) In the RIA conditions, it is concluded that the fuel failure behavior and threshold energy might not be affected by pellet shape of which size is in the range of the current LWR's fuel rods. (author)

  14. A survey on fuel pellet cracking and healing phenomena in reactor operation

    International Nuclear Information System (INIS)

    Faya, S.C.S.

    1981-10-01

    In normal reactor operation, oxide fuel pellets will crack. The majority of the pellet segments will lie against the cladding. When temperature in the central region of the fuel during irradiation is raised to the plastic region, crack healing occurs. The repetition of cracking-healing-cracking sequence resulting from repeated power cycle has a significant effect on fuel relocation. The fuel pellet relocation must be known since it effects the cladding life time. The fuel pellet cracking and healing phenomeno in reactor operation are described and the pertinant method of analysis is also discussed. (Author) [pt

  15. Modeling the UO2 ex-AUC pellet process and predicting the fuel rod temperature distribution under steady-state operating condition

    Science.gov (United States)

    Hung, Nguyen Trong; Thuan, Le Ba; Thanh, Tran Chi; Nhuan, Hoang; Khoai, Do Van; Tung, Nguyen Van; Lee, Jin-Young; Jyothi, Rajesh Kumar

    2018-06-01

    Modeling uranium dioxide pellet process from ammonium uranyl carbonate - derived uranium dioxide powder (UO2 ex-AUC powder) and predicting fuel rod temperature distribution were reported in the paper. Response surface methodology (RSM) and FRAPCON-4.0 code were used to model the process and to predict the fuel rod temperature under steady-state operating condition. Fuel rod design of AP-1000 designed by Westinghouse Electric Corporation, in these the pellet fabrication parameters are from the study, were input data for the code. The predictive data were suggested the relationship between the fabrication parameters of UO2 pellets and their temperature image in nuclear reactor.

  16. Effect of Granule Size on Diametric Tolerance of Annular Fuel Pellet

    International Nuclear Information System (INIS)

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

    2008-01-01

    A dual cooled annular fuel has been seriously considered as a favorable option for an extended power uprate of a Pressurized Water Reactor fuel assembly. An 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. From the viewpoint of the fuel pellet fabrication, however, the unique shape of annular fuel pellet causes challenging difficulties to satisfy a diametric tolerance. A sintered cylindrical PWR fuel pellet fabricated by a conventional double-acting press has an hour-glass shape due to an inhomogeneous green density distribution in a powder compact. Thus, a sintered pellet usually undergoes a centerless grinding process in order to secure diametric tolerance specifications. In the case of an annular pellet fabrication using a conventional double-acting press, the same hour-glass shape would probably occur. An inhomogeneous green density distribution in a powder compact is attributed to granule-granule frictions and granule to pressing mold wall frictions. Frictions result in an irregular pressing load distribution in a powder compact. In order to mitigate the frictions, a lot of process variables should be considered such as pre-compaction pressure, lubricant content, granule size and compaction pressure. The purpose of this study is to investigate the effect of a granule size on the amount of deformation after sintering, in other words, the amount of an hour-glassing. The granules with classified size ranges were made to green annular pellets with the same height and diameters. The hour-glassing amounts of the sintered annular pellets were measured and compared with that of the annular pellet made by unclassified granule

  17. Thermal expansion of UO2-Gd2O3 fuel pellets

    International Nuclear Information System (INIS)

    Une, Katsumi

    1986-01-01

    In recent years, more consideration has been given to the application of UO 2 -Gd 2 O 3 burnable poison fuel to LWRs in order to improve the core physics and to extend the burnup. It has been known that UO 2 forms a single phase cubic fluorite type solid solution with Gd 2 O 3 up to 20 - 30 wt.% above 1300 K. The addition of Gd 2 O 3 to UO 2 lattices changes the properties of the fuel pellets. The limited data on the thermal expansion of UO 2 -Gd 2 O 3 fuel exist, but those are inconsistent. UO 2 -Gd 2 O 3 fuel pellets were fabricated, and the linear thermal expansion of UO 2 and UO 2 -(5, 8 and 10 wt.%)Gd 2 O 3 fuel pellets was measured with a differential dilatometer over the temperature range of 298 - 1973 K. A sapphire rod of 6 mm diameter and 15.5 mm length was used as the reference material. After the preheating cycle, the measurement was performed in argon atmosphere. The results for UO 2 pellets showed excellent agreement with the data in literatures. The linear thermal expansion of UO 2 -Gd 2 O 3 fuel pellets showed the increase with increasing the Gd 2 O 3 content. Consideration must be given to this excessive expansion in the fuel design of UO 2 -Gd 2 O 3 pellets. The equations for the linear thermal expansion and density of UO 2 -Gd 2 O 3 fuel pellets were derived by the method of least squares. (Kako, I.)

  18. Nuclear fuel deformation phenomena

    International Nuclear Information System (INIS)

    Van Brutzel, L.; Dingreville, R.; Bartel, T.J.

    2015-01-01

    Nuclear fuel encounters severe thermomechanical environments. Its mechanical response is profoundly influenced by an underlying heterogeneous microstructure but also inherently dependent on the temperature and stress level histories. The ability to adequately simulate the response of such microstructures, to elucidate the associated macroscopic response in such extreme environments is crucial for predicting both performance and transient fuel mechanical responses. This chapter discusses key physical phenomena and the status of current modelling techniques to evaluate and predict fuel deformations: creep, swelling, cracking and pellet-clad interaction. This chapter only deals with nuclear fuel; deformations of cladding materials are discussed elsewhere. An obvious need for a multi-physics and multi-scale approach to develop a fundamental understanding of properties of complex nuclear fuel materials is presented. The development of such advanced multi-scale mechanistic frameworks should include either an explicit (domain decomposition, homogenisation, etc.) or implicit (scaling laws, hand-shaking,...) linkage between the different time and length scales involved, in order to accurately predict the fuel thermomechanical response for a wide range of operating conditions and fuel types (including Gen-IV and TRU). (authors)

  19. Ceria-thoria pellet manufacturing in preparation for plutonia-thoria LWR fuel production

    Energy Technology Data Exchange (ETDEWEB)

    Drera, Saleem S., E-mail: saleem.drera@scatec.no [Thor Energy AS, Karenslyst allé 9C, 0278 Oslo (Norway); Björk, Klara Insulander [Thor Energy AS, Karenslyst allé 9C, 0278 Oslo (Norway); Sobieska, Matylda [Institute for Energy Technology (IFE), Nuclear Materials, Os allé 5, NO-1777, Halden (Norway)

    2016-10-15

    Thorium dioxide (thoria) has potential to assist in niche roles as fuel for light water reactors (LWRs). One such application for thoria is its use as the fertile component to burn plutonium in a mixed oxide fuel (MOX). Thor Energy and an international consortium are currently irradiating plutonia-thoria (Th-MOX) fuel in an effort to produce data for its licensing basis. During fuel-manufacturing research and development (R&D), surrogate materials were utilized to highlight procedures and build experience. Cerium dioxide (ceria) provides a good surrogate platform to replicate the chemical nature of plutonium dioxide. The project’s fuel manufacturing R&D focused on powder metallurgical techniques to ensure manufacturability with the current commercial MOX fuel production infrastructure. The following paper highlights basics of the ceria-thoria fuel production including powder milling, pellet pressing and pellet sintering. Green pellets and sintered pellets were manufactured with average densities of 67.0% and 95.5% that of theoretical density respectively. - Highlights: • High quality Ce−Th fuel production can be accomplished by utilizing powder metallurgical procedures. • Powder morphology is key to obtaining high density fuels. • Optimal pellet pressing is obtained when 3.5–4 tons of force is applied by the pellet press for powder compaction. • Pellet sintering is accomplished effectively in an Air oxidizing atmosphere. • Based on this surrogate work, expected (Th,Pu)O{sub 2} fuel density is 95.5% of theoretical density.

  20. Pellet fueling development at ORNL

    International Nuclear Information System (INIS)

    Combs, S.K.; Milora, S.L.; Foster, C.A.; Schuresko, D.D.; Foust, C.R.; Simmons, D.W.; Beard, D.S.

    1986-09-01

    Advanced plasma fueling systems for magnetic confinement devices are being developed at the Oak Ridge National Laboratory (ORNL). The general approach is that of producing and accelerating frozen hydrogenic pellets at speeds in the range of 1-2 km/s and higher. Two specific concepts are under development: (1) high-speed pneumatic acceleration; and (2) mechanical (centrifugal) acceleration. Both approaches are being pursued to meet the projected pellet size and delivery rates for major near-term plasma confinement devices, such as the Tokamak Fusion Test Reactor (TFTR), Tore Supra, the Joint European Torus (JET), JT-60, and Doublet III-D (DIII-D), as well as future applications. In addition to these confinement physics related activities, ORNL is pursuing advanced technologies to achieve pellet velocities significantly in excess of the 2-km/s range already attained with pneumatic injectors and has embarked on a development program designed to explore the feasibility of fabricating and accelerating tritium pellets. This paper describes these ongoing activities

  1. Processing of surrogate nuclear fuel pellets for better dimensional control with dry bag isostatic pressing

    Energy Technology Data Exchange (ETDEWEB)

    Hoggan, Rita E., E-mail: Rita.hoggan@inl.gov; Zuck, Larry D., E-mail: Larry.zuck@inl.gov; Cannon, W. Roger, E-mail: cannon@rutgers.edu; Lessing, Paul A., E-mail: p.a.l.2@hotmail.com

    2016-12-15

    A study of improved methods of processing fuel pellets was undertaken using ceria and zirconia/yttria/alumina as surrogates. Through proper granulation, elimination of fines and vertical vibration (tapping) of the parts bag prior to dry bag isostatic pressing (DBIP), reproducibility of diameter profiles among multiple pellets of ceria was improved by almost an order of magnitude. Reproducibility of sintered pellets in these studies was sufficient to allow pellets to be introduced into the cladding with a gap between the pellet and cladding on the order of 50 μm to 100 μm but not a uniform gap with tolerance of ±12 μm as is currently required. Deviation from the mean diameter along the length of multiple pellets, and deviation from roundness, decreased after sintering. This is not generally observed with dry pressed pellets. Sintered shrinkage was uniform to ±0.05% and thus, as an alternative, pellets may be machined to tolerance before sintering, thus avoiding the waste associated with post-sinter grinding. - Highlights: • Three methods of granule preparation for two different powder sources were outlined and compared using tap density curves. • A dry bag isostatic press was used to fabricate pellets and longer rods. Thus longer pellets could be fabricated by this technique. • Vertical vibrations to pack granules decreased variation in dimensions from pellet to pellet by a factor of nine. • Sintering shrinkage varied by only 0.1% along the length of a rod. Thus green machining prior to sintering could result in tight tolerances.

  2. Development for analysis system of rods enrichment of nuclear fuels

    International Nuclear Information System (INIS)

    Rojas C, E.L.

    1998-01-01

    Nuclear industry is strongly regulated all over the world and quality assurance is important in every nuclear installation or process related with it. Nuclear fuel manufacture is not the exception. ININ was committed to manufacture four nuclear fuel bundles for the CFE nucleo electric station at Laguna Verde, Veracruz, under General Electric specifications and fulfilling all the requirements of this industry. One of the quality control requisites in nuclear fuel manufacture deals with the enrichment of the pellets inside the fuel bundle rods. To achieve the quality demanded in this aspect, the system described in this work was developed. With this system, developed at ININ it is possible to detect enrichment spikes since 0.4 % in a column of pellets with a 95 % confidence interval and to identify enrichment differences greater than 0.2 % e between homogeneous segments, also with a 95 % confidence interval. ININ delivered the four nuclear fuel bundles to CFE and these were introduced in the core of the nuclear reactor of Unit 1 in the fifth cycle. Nowadays they are producing energy and have shown a correct mechanical performance and neutronic behavior. (Author)

  3. Dissolution test for homogeneity of mixed oxide fuel pellets

    International Nuclear Information System (INIS)

    Lerch, R.E.

    1979-08-01

    Experiments were performed to determine the relationship between fuel pellet homogeneity and pellet dissolubility. Although, in general, the amount of pellet residue decreased with increased homogeneity, as measured by the pellet figure of merit, the relationship was not absolute. Thus, all pellets with high figure of merit (excellent homogeneity) do not necessarily dissolve completely and all samples that dissolve completely do not necessarily have excellent homogeneity. It was therefore concluded that pellet dissolubility measurements could not be substituted for figure of merit determinations as a measurement of pellet homogeneity. 8 figures, 3 tables

  4. Fuel pellets from lodge pole pine first thinnings

    Energy Technology Data Exchange (ETDEWEB)

    Hoegqvist, Olof; Larsson, Sylvia H.; Samuelsson, Robert; Lestander, Torbjoern A. [Swedish Univ. of Agricultural Sciences, Unit of Biomass Technology and Chemistry, Umeaa (Sweden)], e-mail: sylvia.larsson@slu.se

    2012-11-01

    Stemwood and whole trees of lodgepole pine (Pinus contorta Dougl. var. latifolia L.) were evaluated as raw materials for fuel pellets in a pilot scale pelletizing study. Pellet and pelletizing properties were measured and modeled in an experimental design where raw material moisture content (%), die channel length (mm), and storage time (days) were varied. Additionally, ash contents (%), extractive contents (%), and ash melting temperatures (deg C) were analyzed. For both assortments, raw material moisture content was positively correlated to pellet bulk density and durability (range 9-13%, wet base). Both assortments had ash contents below 0.7%, and thus, fulfilled the demands for class A1 pellets.

  5. Nuclear fuels and development of nuclear fuel elements

    International Nuclear Information System (INIS)

    Sundaram, C.V.; Mannan, S.L.

    1989-01-01

    Safe, reliable and economic operation of nuclear fission reactors, the source of nuclear power at present, requires judicious choice, careful preparation and specialised fabrication procedures for fuels and fuel element structural materials. These aspects of nuclear fuels (uranium, plutonium and their oxides and carbides), fuel element technology and structural materials (aluminium, zircaloy, stainless steel etc.) are discussed with particular reference to research and power reactors in India, e.g. the DHRUVA research reactor at BARC, Trombay, the pressurised heavy water reactors (PHWR) at Rajasthan and Kalpakkam, and the Fast Breeder Test Reactor (FBTR) at Kalpakkam. Other reactors like the gas-cooled reactors operating in UK are also mentioned. Because of the limited uranium resources, India has opted for a three-stage nuclear power programme aimed at the ultimate utilization of her abundant thorium resources. The first phase consists of natural uranium dioxide-fuelled, heavy water-moderated and cooled PHWR. The second phase was initiated with the attainment of criticality in the FBTR at Kalpakkam. Fast Breeder Reactors (FBR) utilize the plutonium and uranium by-products of phase 1. Moreover, FBR can convert thorium into fissile 233 U. They produce more fuel than is consumed - hence, the name breeders. The fuel parameters of some of the operating or proposed fast reactors in the world are compared. FBTR is unique in the choice of mixed carbides of plutonium and uranium as fuel. Factors affecting the fuel element performance and life in various reactors e.g. hydriding of zircaloys, fuel pellet-cladding interaction etc. in PHWR and void swelling; irradiation creep and helium embrittlement of fuel element structural materials in FBR are discussed along with measures to overcome some of these problems. (author). 15 refs., 9 tabs., 23 figs

  6. Effects of pellet-to-cladding gap design parameters on the reliability of high burnup PWR fuel rods under steady state and transient conditions

    International Nuclear Information System (INIS)

    Tas, Fatma Burcu; Ergun, Sule

    2013-01-01

    Highlights: • Fuel performance of a typical Pressurized Water Reactor rod is analyzed. • Steady state fuel rod behavior is examined to see the effects of pellet to cladding gap thickness and gap gas pressure. • Transient fuel rod behavior is examined to see the effects of pellet to cladding gap thickness and gap gas pressure. • The optimum pellet to cladding gap thickness and gap gas pressure values of the simulated fuel are determined. • The effects of pellet to cladding gap design parameters on nuclear fuel reliability are examined. - Abstract: As an important improvement in the light water nuclear reactor operations, the nuclear fuel burnup rate is increased in recent decades and this increase causes heavier duty for the nuclear fuel. Since the high burnup fuel is exposed to very high thermal and mechanical stresses and since it operates in an environment with high radiation for about 18 month cycles, it carries the risk of losing its integrity. In this study; it is aimed to determine the effects of pellet–cladding gap thickness and gap pressure on reliability of high burnup nuclear fuel in Pressurized Water Reactors (PWRs) under steady state operation conditions and suggest optimum values for the examined parameters only and validate these suggestions for a transient condition. In the presented study, fuel performance was analyzed by examining the effects of pellet–cladding gap thickness and gap pressure on the integrity of high burnup fuels. This work is carried out for a typical Westinghouse type PWR fuel. The steady state conditions were modeled and simulated with FRAPCON-3.4a steady state fuel performance code and the FRAPTRAN-1.4 fuel transient code was used to calculate transient fuel behavior. The analysis included the changes in the important nuclear fuel design limitations such as the centerline temperature, cladding stress, strain and oxidation with the change in pellet–cladding gap thickness and initial pellet–cladding gap gas

  7. Wood pellets. The cost-effective fuel

    International Nuclear Information System (INIS)

    Anon.

    2001-01-01

    The article is based on an interview with Juhani Hakkarainen of Vapo Oy. Wood pellets are used in Finland primarily to heat buildings such as schools and offices and in the home. They are equally suitable for use in larger installations such as district heating plants and power stations. According to him wood pellets are suitable for use in coal-fired units generating heat, power, and steam. Price-wise, wood pellets are a particularly competitive alternative for small coal-fired plants away from the coast. Price is not the only factor on their side, however. Wood pellets also offer a good environmental profile, as they burn cleanly and generate virtually no dust, an important plus in urban locations. The fact that pellets are a domestically produced fuel is an added benefit, as their price does not fluctuate in the same way that the prices of electricity, oil, coal, and natural gas do. The price of pellets is largely based on direct raw material and labour costs, which are much less subject to ups and downs

  8. Advanced fuel pellet materials and designs for water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2004-10-01

    This meeting was the second IAEA meeting on this subject. The first was held in 1996 in Tokyo, Japan. They are all part of a cooperative effort through the Technical Working Group on Water Reactor Fuel Performance and Technology (TWGFPT) of IAEA, with a series of three further meetings organized by CEA, France and co-sponsored by the IAEA and OECD/NEA. In the seven years since the first meeting took place, the demands on fuel duties have increased, with higher burnup, longer fuel cycles and higher temperatures. This places additional demands on fuel performance to comply with safety requirements. Criteria relative to fuel components, i.e. pellets and fuel rod column, require limiting of fission gas release and pellet-cladding interaction (PCI). This means that fuel components should maintain the composite of rather contradictory properties from the beginning until the end of its in-pile operation. Fabrication and design tools are available to influence, and to some extent, to ensure desirable in-pile fuel properties. Discussion of these tools was one of the objectives of the meeting. The second objective was the analysis of fuel characteristics at high burnup and the third and last objective was the discussion of specific feature of MOX and urania gadolinia fuels. Sixty specialists in the field of fuel fabrication technology attended the meeting from 18 countries. Twenty-five papers were presented in five sessions covering all relevant topics from the practices and modelling of fuel fabrication technology to its optimization. Eight papers were presented in session 'Optimization of fuel fabrication technology' which all were devoted to fuel fabrication technology. They mostly treated methods for optimizing fuel manufacturing processes, but gave also a good overview on nuclear fabrication needs and capabilities in different countries. During Session 'UO 2 , MOX and UO 2 -Gd 2 O 3 pellets with additives', six papers were presented in this session, which dealt mainly

  9. Steady-state thermal-hydraulic analysis of the pellet-bed reactor for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Morley, N.J.; Yang, J.Y.

    1992-01-01

    The pellet-bed reactor (PBR) for nuclear thermal propulsion is a hydrogen-cooled, BeO-reflected, fast reactor, consisting of an annular core region filled with randomly packed, spherical fuel pellets. The fuel pellets in the PBR are self-supported, eliminating the need for internal core structure, which simplifies the core design and reduces the size and mass of the reactor. Each spherical fuel pellet is composed of hundreds of fuel microspheres embedded in a zirconium carbide (ZrC) matrix. Each fuel microsphere is composed of a UC-NbC fuel kernel surrounded by two consecutive layers of the NbC and ZrC. Gaseous hydrogen serves both as core coolant and as the propellant for the PBR rocket engine. The cold hydrogen flows axially down the inlet channel situated between the core and the external BeO reflector and radially through the orifices in the cold frit, the core, and the orifices in the hot frit. Finally, the hot hydrogen flows axially out the central channel and exits through converging-diverging nozzle. A thermal-hydraulic analysis of the PBR core was performed with an emphasis on optimizing the size and axial distribution of the orifices in the hot and cold frits to ensure that hot spots would not develop in the core during full-power operation. Also investigated was the validity of the assumptions of neglecting the axial conduction and axial cross flow in the core

  10. Experimental Observation of Densification Behavior of UO2 Annular Pellet

    International Nuclear Information System (INIS)

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

    2007-01-01

    Recently, in the nuclear industry, one of the major issues is the improvement of a fuel economy. And many efforts have been made to develop a nuclear fuel for a high burnup and extended cycle. In the development of a high performance fuel, in-reactor fuel behavior (fission gas release, pellet-clad interaction, stress corrosion cracking, cladding corrosion, etc.) must be seriously reconsidered. Also, fuel fabrication (high enriched UO 2 powder handling, fuel rod and assembly manufacturing, fabricated fuel rod and assembly storage and transport, etc.) and an enrichment process (5 w/o criticality limit, etc.) must be discussed. A modification and an improvement of the nuclear fuel system will be also required. The typical fuel geometry of a PWR (Pressurized Water Reactor) is composed of a cylindrical pellet with a tubular cladding. And the outer surface of the cladding is cooled with water. However, to allow a substantial increase in the power density, an additional cooling is needed. One of the best ways is the application of the new fuel geometry that is of annular shape and has both internal and external cooling. From this point of view, the double cooled fuel is being developed by KAERI (Korea Atomic Energy Research Institute), and as a part of the project, the development of a fabrication process of a UO 2 annular pellet is now in progress. The dimensional behavior of UO 2 fuel is an important parameter in an irradiation performance. Various investigations (resintering test, model calculation, in-pile dimensional change measuring, etc.) had been performed. In designing a double cooled fuel, the importance of the dimensional behavior of a fuel pellet is higher, because the gap distance between a pellet and cladding can considerably affect on the in reactor fuel performance (gap conductance). And the dimensional behavior of an inner/outer gap is different with a cylindrical pellet, when the pellet shrinks (densification), the inner gap distance decreases and the

  11. Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

    International Nuclear Information System (INIS)

    Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

    2011-01-01

    The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens' theory and reported thermal conductivities of unirradiated (U, Pu) O 2 and irradiated UO 2 pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.

  12. Artificial vision in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Dorado, P.

    2007-01-01

    The development of artificial vision techniques opens a door to the optimization of industrial processes which the nuclear industry cannot miss out on. Backing these techniques represents a revolution in security and reliability in the manufacturing of a highly technological products as in nuclear fuel. Enusa Industrias Avanzadas S. A. has successfully developed and implemented the first automatic inspection equipment for pellets by artificial vision in the European nuclear industry which is nowadays qualified and is already developing the second generation of this machine. There are many possible applications for the techniques of artificial vision in the fuel manufacturing processes. Among the practices developed by Enusa Industrias Avanzadas are, besides the pellets inspection, the rod sealing drills detection and positioning in the BWR products and the sealing drills inspection in the PWR fuel. The use of artificial vision in the arduous and precise processes of full inspection will allow the absence of human error, the increase of control in the mentioned procedures, the reduction of doses received by the personnel, a higher reliability of the whole of the operations and an improvement in manufacturing costs. (Author)

  13. Repetitive fueling pellet injection in large helical device

    International Nuclear Information System (INIS)

    Yamada, H.; Sakamoto, R.; Viniar, I.; Oda, Y.; Kikuchi, K.; Lukin, A.; Skoblikov, S.; Umov, A.; Takaura, K.; Onozuka, M.; Kato, S.; Sudo, S.

    2003-01-01

    A repetitive pellet injector has been developed for investigation of fueling issues towards the steady-state operation in Large Helical Device (LHD). The goal of this approach is achievement of the plasma operation for longer than 1000 s. A principal technical element of the pellet injector is solidification of hydrogen and extrusion of a solid hydrogen rod through a cryogenic screw extruder cooled by Giffard-McMahon (GM) cryo-coolers. Continuous operation of more than 10000 pellet launches at 10 Hz has been demonstrated. The reliability of pellet launch exceeds 99%. The pellet mass and velocity, the consumption of propellant gas and quality of pellets have been successfully tested to fit the experimental requirement in LHD

  14. Repetitive fueling pellet injection in large helical device

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, H. E-mail: hyamada@lhd.nifs.ac.jp; Sakamoto, R.; Viniar, I.; Oda, Y.; Kikuchi, K.; Lukin, A.; Skoblikov, S.; Umov, A.; Takaura, K.; Onozuka, M.; Kato, S.; Sudo, S

    2003-09-01

    A repetitive pellet injector has been developed for investigation of fueling issues towards the steady-state operation in Large Helical Device (LHD). The goal of this approach is achievement of the plasma operation for longer than 1000 s. A principal technical element of the pellet injector is solidification of hydrogen and extrusion of a solid hydrogen rod through a cryogenic screw extruder cooled by Giffard-McMahon (GM) cryo-coolers. Continuous operation of more than 10000 pellet launches at 10 Hz has been demonstrated. The reliability of pellet launch exceeds 99%. The pellet mass and velocity, the consumption of propellant gas and quality of pellets have been successfully tested to fit the experimental requirement in LHD.

  15. Nuclear fuel rods

    International Nuclear Information System (INIS)

    Wada, Toyoji.

    1979-01-01

    Purpose: To remove failures caused from combination of fuel-cladding interactions, hydrogen absorptions, stress corrosions or the likes by setting the quantity ratio of uranium or uranium and plutonium relative to oxygen to a specific range in fuel pellets and forming a specific size of a through hole at the center of the pellets. Constitution: In a fuel rods of a structure wherein fuel pellets prepared by compacting and sintering uranium dioxide, or oxide mixture consisting of oxides of plutonium and uranium are sealed with a zirconium metal can, the ratio of uranium or uranium and plutonium to oxygen is specified as 1 : 2.01 - 1 : 2.05 in the can and a passing hole of a size in the range of 15 - 30% of the outer diameter of the fuel pellet is formed at the center of the pellet. This increases the oxygen partial pressure in the fuel rod, oxidizes and forms a protection layer on the inner surface of the can to control the hydrogen absorption and stress corrosion. Locallized stress due to fuel cladding interaction (PCMI) can also be moderated. (Horiuchi, T.)

  16. Nuclear fuels

    International Nuclear Information System (INIS)

    2008-01-01

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

  17. Measuring method for amount of fissionable gas in spent fuel pellet

    International Nuclear Information System (INIS)

    Kashibe, Shinji.

    1992-01-01

    The method of the present invention separately measures the amount of both of a fission product (FP) gas accumulated in bubbles at the crystal grain boundary of spent fuel pellets and an FP gas accumulated in the crystal grains. That is, in a radial position of the spent fuel pellet, a microfine region is mechanically destroyed. The amount of the FP gas released by the destruction from the crystal grains is measured by using a mass analyzer. Then, when the destroyed pieces formed by the destruction are recovered and dissolved, FP gas accumulated in the crystal grains of the pellet is released. The amount released is measured by the mass analyzer. With such procedures, the amount of FP gas accumulated in the bubbles at the crystal grain boundary and in the crystal grains at the radial position of the spent fuel pellet can be measured discriminately. Accordingly, the integrity of the fuel pellet can be recognized appropriately. (I.S.)

  18. Design and manufacturing of non-instrumented capsule for advanced PWR fuel pellet irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Song, K. W. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This project is preparing to irradiation test of the developed large grain UO{sub 2} fuel pellet in HANARO for pursuit fuel safety and high burn-up in 'Advanced LWR Fuel Technology Development Project' 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 capsule are designed and manufactured for irradiation test in HANARO. This non-instrumented irradiation capsule of Advanced PWR Fuel pellet was referred the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO(DUPIC Rig-001) and 18-element HANARO fuel, was designed to ensure the integrity and the endurance of non-instrumented capsule during the long term(2.5 years) irradiation. To irradiate the UO{sub 2} pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. This non-instrumented irradiation capsule will be based to develope the non-instrumented capsule for the more long term irradiation in HANARO. 22 refs., 13 figs., 5 tabs. (Author)

  19. End plug welding of nuclear fuel elements-AFFF experience

    International Nuclear Information System (INIS)

    Bhatt, R.B.; Singh, S.; Aniruddha Kumar; Amit; Arun Kumar; Panakkal, J.P.; Kamath, H.S.

    2004-01-01

    Advanced Fuel Fabrication Facility is engaged in the fabrication of mixed oxide (U,Pu)O 2 fuel elements of various types of nuclear reactors. Fabrication of fuel elements involves pellet fabrication, stack making, stack loading and end plug welding. The requirement of helium bonding gas inside the fuel elements necessitates the top end plug welding to be carried out with helium as the shielding gas. The severity of the service conditions inside a nuclear reactor imposes strict quality control criteria, which demands for almost defect free welds. The top end plug welding being the last process step in fuel element fabrication, any rejection at this stage would lead to loss of effort prior to this step. Moreover, the job becomes all the more difficult with mixed oxide (MOX) as the entire fabrication work has to be carried out in glove box trains. In the case of weld rejection, accepted pellets are salvaged by cutting the clad tube. This is a difficult task and recovery of pellets is low (requiring scrap recovery operation) and also leads to active metallic waste generation. This paper discusses the experience gained at AFFF, in the past 12 years in the area of end plug welding for different types of MOX fuel elements

  20. Sintering densification of CaO–UO{sub 2}–Gd{sub 2}O{sub 3} nuclear fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yun [Fundamental Science on Radioactive Geology and Exploration Technology Laboratory, East China Institute of Technology, Nanchang, 330013, Jiangxi (China); Sun, Huidong [China Nucle Power Engineering Co., Ltd (China); Wang, Hui, E-mail: yinchanggeng5525@163.com [National Key Laboratory for Nuclear Fuel and Materials, Nuclear Power Institute of China, Chengdu, 610041 (China); Pan, Xiaoqiang; Li, Tongye; Liu, Jinhong; Zhang, Yong; Wang, Xinjie [National Key Laboratory for Nuclear Fuel and Materials, Nuclear Power Institute of China, Chengdu, 610041 (China)

    2015-10-15

    CaO-doped UO{sub 2}-10 wt% Gd{sub 2}O{sub 3} burnable poison fuel was prepared by co-precipitation reaction method. It was found that 0.3 wt% CaO-doping significantly improved the sintered density, grain sizes and crushing strength of UO{sub 2}–Gd{sub 2}O{sub 3} fuel pellets at the sintering temperature of 1650 °C in the sintering atmosphere of hydrogen for 3.5 h. In addition, homogeneous solid solution without precipitation of free phases of CaO and Gd{sub 2}O{sub 3} was successfully achieved. CaO doping in UO{sub 2}–Gd{sub 2}O{sub 3} fuel pellet system accelerated the thermally activated material transport, so the onset temperature of densification as well as the temperature of the maximum densification rate shifted to a lower temperature region. - Highlights: • A small amount of 0.3% doped CaO{sub 2} can significantly improve the sintered density. • Homogeneous solid solution forms without precipitation of free phases. • The pellet has good density, high strength and increasing grain sizes with homogeneity. • The pellet accelerates a thermally activated material transport.

  1. Fuel element for a nuclear reactor

    International Nuclear Information System (INIS)

    Linning, D.L.

    1977-01-01

    An improvement of the fuel element for a fast nuclear reactor described in patent 15 89 010 is proposed which should avoid possible damage due to swelling of the fuel. While the fuel element according to patent 15 89 010 is made in the form of a tube, here a further metal jacket is inserted in the centre of the fuel rod and the intermediate layer (ceramic uranium compound) is provided on both sides, so that the nuclear fuel is situated in the centre of the annular construction. Ceramic uranium or plutonium compounds (preferably carbide) form the fuel zone in the form of circular pellets, which are surrounded by annular gaps, so that gaseous fission products can escape. (UWI) [de

  2. Beryllium Project: developing in CDTN of uranium dioxide fuel pellets with addition of beryllium oxide to increase the thermal conductivity

    International Nuclear Information System (INIS)

    Ferreira, Ricardo Alberto Neto; Camarano, Denise das Merces; Miranda, Odair; Grossi, Pablo Andrade; Andrade, Antonio Santos; Queiroz, Carolinne Mol; Gonzaga, Mariana de Carvalho Leal

    2013-01-01

    Although the nuclear fuel currently based on pellets of uranium dioxide be very safe and stable, the biggest problem is that this material is not a good conductor of heat. This results in an elevated temperature gradient between the center and its lateral surface, which leads to a premature degradation of the fuel, which restricts the performance of the reactor, being necessary to change the fuel before its full utilization. An increase of only 5 to 10 percent in its thermal conductivity, would be a significant increase. An increase of 50 percent would be a great improvement. A project entitled 'Beryllium Project' was developed in CDTN - Centro de Desenvolvimento da Tecnologia Nuclear, which aimed to develop fuel pellets made from a mixture of uranium dioxide microspheres and beryllium oxide powder to obtain a better heat conductor phase, filling the voids between the microspheres to increase the thermal conductivity of the pellet. Increases in the thermal conductivity in the range of 8.6% to 125%, depending on the level of addition employed in the range of 1% to 14% by weight of beryllium oxide, were obtained. This type of fuel promises to be safer than current fuels, improving the performance of the reactor, in addition to last longer, resulting in great savings. (author)

  3. Ultrasonic analysis of UO{sub 2} pellets

    Energy Technology Data Exchange (ETDEWEB)

    Bittencourt, Marcelo de S.Q.; Baroni, Douglas B.; Martorelli, Daniel S., E-mail: bittenc@ien.gov.br, E-mail: douglasbaroni@ien.gov.br, E-mail: daniel@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Ultrassom; Dias, Fabio C.; Silva, Jose W.S. da, E-mail: fabio@ird.gov.br, E-mail: wanderley@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Salvaguardas

    2013-07-01

    Ceramic materials have been widely used for various purposes in many different industries due to certain characteristics, such as high melting point and high resistance to corrosion. In the nuclear area, ceramics are of great importance due to the process of fabrication of fuel pellets for nuclear reactors. Generally, high accuracy destructive techniques are used to characterize nuclear materials for fuel fabrication. These techniques usually require costly equipment and facilities, as well as experienced personnel. This paper aims at presenting an analysis methodology for UO2 pellets using a non-destructive ultrasonic technique for porosity measurement. This technique differs from traditional ultrasonic techniques in the sense it uses ultrasonic pulses in frequency domain instead of time domain. Therefore, specific characteristics of the analyzed material are associated with the obtained frequency spectrum. In the present work, four fuel grade UO2 pellets were analyzed and the corresponding results evaluated. (author)

  4. The nuclear fuel cycle: (2) fuel element manufacture

    International Nuclear Information System (INIS)

    Doran, J.

    1976-01-01

    Large-scale production of nuclear fuel in the United Kingdom is carried out at Springfields Works of British Nuclear Fuels Ltd., a company formed from the United Kingdom Atomic Energy Authority in 1971. The paper describes in some detail the Springfields Works processes for the conversion of uranium ore concentrate to uranium tetrafluoride, then conversion of the tetrafluoride to either uranium metal for cladding in Magnox to form fuel for the British Mk I gas-cooled reactors, or to uranium hexafluoride for enrichment of the fissile 235 U isotope content at the Capenhurst Works of BNFL. Details are given of the reconversion at Springfields Works of this enriched uranium hexafluoride to uranium dioxide, which is pelleted and then clad in either stainless steel or zircaloy containers to form the fuel assemblies for the British Mk II AGR or advanced gas-cooled reactors or for the water reactor fuels. (author)

  5. proximate and ultimate analysis of fuel pellets from oil palm residues

    African Journals Online (AJOL)

    HOD

    Keywords: Oil Palm Residues, Fuel Pellets, Proximate Analysis, Ultimate Analysis. 1. INTRODUCTION ... Pelletizing of this biomass resources into pellets is a way of ensuring a ... demand for pellets [3], and alternative feed-stocks such as palm kernel ... agro-residues, selection of the best pellets has to be made based on ...

  6. Some possibilities for improvement of fuel utilization in nuclear power plants

    International Nuclear Information System (INIS)

    Kocic, A.; Marinkovic, N.

    1983-01-01

    Methods for improving the nuclear fuel utilization with the emphasis on LWRs are being dealt with in this paper. Some basic results concerning tubular fuel pellets of the Krsko nuclear power plants are presented, showing promising possibilities for uranium saving from the neutronics point of view. (author)

  7. Advanced Fuel Pellet Materials and Fuel Rod Design for Water Cooled Reactors. Proceedings of a Technical Committee Meeting

    International Nuclear Information System (INIS)

    2010-10-01

    The economics of current nuclear power plants have improved through increased fuel burnup and longer fuel cycles, i.e. increasing the effective time that fuel remains in the reactor core and the amount of energy it generates. Efficient consumption of fissile material in the fuel element before it is discharged from the reactor means that less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. Better utilization of fissile nuclear materials, as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all, on fuel and cladding materials. It entails increased attention to measures ensuring desired in-pile fuel performance parameters that require adequate improvements in fuel material properties and fuel rod designs. These are the main reasons that motivated the IAEA Technical Working Group on Fuel Performance and Technology (TWG-FPT) to recommend the organization of a Technical Committee Meeting on Advanced Fuel Pellet Materials and Fuel Rod Designs for Power Reactors. The proposal was supported by the IAEA TWGs on Advanced Technologies for Light and Heavy Water-Cooled Reactors (TWG-LWR and TWG-HWR), and the meeting was held at the invitation of the Government of Switzerland at the Paul Scherrer Institute in Villigen, from 23 to 26 November 2009. This was the third IAEA meeting on these subjects (the first was held in 1996 in Tokyo, Japan, and the second in 2003 in Brussels, Belgium), which reflects the continuous interest in the above issues among Member States. The purpose of the meeting was to review the current status in the development of fuel pellet materials and to explore recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting was attended by 45 specialists representing fuel vendors, nuclear utilities, research and development

  8. Fuel assemblies for nuclear reactor

    International Nuclear Information System (INIS)

    Nishi, Akihito.

    1987-01-01

    Purpose: To control power-up rate at the initial burning stage of new fuel assemblies due to fuel exchange in a pressure tube type power reactor. Constitution: Burnable poisons are disposed to a most portion of fuel pellets in a fuel assembly to such a low concentration as the burn-up rate changes with time at the initial stage of the burning. The most portion means substantially more than one-half part of the pellets and gadolinia is used as burn-up poisons to be dispersed and the concentration is set to less than about 0.2 %. Upon elapse of about 15 days after the charging, the burnable poisons are eliminated and the infinite multiplication factors are about at 1.2 to attain a predetermined power state. Since the power-up rate of the nuclear reactor fuel assembly is about 0.1 % power/hour and the power-up rate of the fuel assembly around the exchanged channel is lower than that, it can be lowered sufficiently than the limit for the power-up rate practiced upon reactor start-up thereby enabling to replace fuels during power operation. (Horiuchi, T.)

  9. Fuel fabrication and reprocessing for nuclear fuel cycle with inherent safety demands

    Energy Technology Data Exchange (ETDEWEB)

    Shadrin, Andrey Yurevich; Dvoeglazov, Konstantin Nikolaevich; Ivanov, Valentine Borisovich; Volk, Vladimir Ivanovich; Skupov, Mikhail Vladimirovich; Glushenkov, Alexey Evgenevich [Joint Stock Company ' ' The High Technological Research Institute of Inorganic Materials' ' , Moscow (Russian Federation); Troyanov, Vladimir Mihaylovich; Zherebtsov, Alexander Anatolievich [Innovation and Technology Center of Project ' ' PRORYV' ' , State Atomic Energy Corporation ' ' Rosatom' ' , Moscow (Russian Federation)

    2015-06-01

    The strategies adopted in Russia for a closed nuclear fuel cycle with fast reactors (FR), selection of fuel type and recycling technologies of spent nuclear fuel (SNF) are discussed. It is shown that one of the possible technological solutions for the closing of a fuel cycle could be the combination of pyroelectrochemical and hydrometallurgical methods of recycling of SNF. This combined scheme allows: recycling of SNF from FR with high burn-up and short cooling time; decreasing the volume of stored SNF and the amount of plutonium in a closed fuel cycle in FR; recycling of any type of SNF from FR; obtaining the high pure end uranium-plutonium-neptunium end-product for fuel refabrication using pellet technology.

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

    International Nuclear Information System (INIS)

    Fukuda, Kosaku

    2000-01-01

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

  11. Apparatus for checking the dimensions of nuclear fuel pellets

    International Nuclear Information System (INIS)

    Marmo, A.R.

    1978-01-01

    The description is given of an apparatus for checking the dimensions of pellets comprising a housing, a feeding device near this housing to move a pellet towards the latter and away from it, and a platform with a hole, this platform being fitted to the housing near the feeding system in order to hold the pellet [fr

  12. The current state of development works for manufacturing and methods of controlling the nuclear fuel for NPPs of Ukraine

    International Nuclear Information System (INIS)

    Odeychuk, N.P.; Levenets, V.V.; Krasnorutsky, V.S.

    2000-01-01

    The paper presents the results of NSC KIPT researches on manufacturing the fuel microspheres and pellets based on uranium dioxide. The data on fuel characteristics for different manufacturing stages are given. The problems of improving the fuel quality with changing the structure characteristics of pellets are considered. Demonstrated is the hardware for pellet controlling and presented are the new ways for developing the methods of controlling the nuclear fuel: X-ray fluorescent analysis; complex of nuclear-physical methods on the base of accelerators; laser-excitation energy-mass-spectrometer. (author)

  13. Optimization parametric study of the fuel pellet dimensions

    International Nuclear Information System (INIS)

    Mai, L.A.

    1986-01-01

    A method to determine the dimensions of fuel pellets, is presented, obtaining the maximum core reactivity at the end of cycle. Other unit cell parameters, fixed in a given reactor, are considered constants. It is seen that the cycle length is an important parameter in the determinations of the pellet dimensions. The optimal pellet radius is found as an increasing function of the cycle length. All calculation have been performed using the HAMMER code. (Author) [pt

  14. Pelletizing using forest fuels and Salix as raw materials. A study of the pelletizing properties; Pelletering med skogsbraensle och Salix som raavara. En undersoekning av pelleterbarheten

    Energy Technology Data Exchange (ETDEWEB)

    Martinsson, Lars; Oesterberg, Stefan [Swedish National Testing and Research Inst., Boraas (Sweden)

    2004-08-01

    Three common forest fuels: light thinning material, cull tree and logging residues as well as energy forest fuel (Salix) has been used as fuel pellet materials. Logging residues and Salix were stacked for approximately 6 and 10 months respectively. Parameters varied for each raw material have been the moisture content and the press length of the die. These parameters have been changed to obtain best possible quality, mainly concerning mechanical durability. Pellets were also produced from bark free shavings in order to use as a reference in this study. Physical as well as chemical properties have been compared. It was comparatively easy to press logging residues and Salix into durable pellets and, even with larger press length, the production of pellets was higher than it was for the other raw materials. The density was equal for all pellets while the mechanical durability was better for all tested raw materials compared with the reference material. The fact that all raw materials besides the reference material contains bark which has an improving effect on the degree of hardness. The quality properties were mainly about the same or better for pellets made of light thinning material and cull tree respectively, compared with the reference pellets. However, the ash content was approximately twice as high compared with the reference pellets. The pellets made of logging residues and Salix respectively were of very good quality concerning duration and density but the ash content was approximately 10 times higher than in the reference pellets. Additionally, the nitrogen content was 6-9 times higher compared with the reference pellets.

  15. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    International Nuclear Information System (INIS)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

    2014-01-01

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests

  16. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

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

    International Nuclear Information System (INIS)

    Majumdar, S.

    2003-01-01

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

  18. Measuring deformation of Fuel pin in a Nuclear Fuel Test Rig

    Energy Technology Data Exchange (ETDEWEB)

    Heo, S. H.; Yang, T. H.; Hong, J. T.; Joung, C. Y.; Ahn, S. H.; Jang, S. Y.; Kim, J. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, an LVDT core for measuring the longitudinal displacement of fuel pellets and clad was designed and produced. A signal processing method for the prepared core was investigated. The Nuclear Fuel Test Rig is used to observe changes in the characteristics of the fuel according to the neutron irradiation at HANARO (High-flux Advanced Neutron Application Reactor), which is a research reactor. Which are the strain and internal temperature of the irradiated nuclear fuel and the internal pressure of fuel due to fission gas, the characteristics of the fuel are measured using various sensors such as a thermocouple, SPND and LVDT. In this study, two shaped LVDT (Linear Variable Differential Transformer) cores for displacement measurements were designed and manufactured in order to measure the displacement of a fuel pellet and cladding tube using LVDT sensors for measuring electrical signals by converting the physical variation such as the force and displacement into a linear motion. In addition, signals from the manufactured LVDT sensor were collected and calibrated. Moreover, a method for obtaining the displacement in the core according to the sensing signal was planned. A derived equation can used to predict the change in the position of core. A following study should be conducted to test the output signal and real variation of out-pile system. For further work, a performance verification is required for an in-pile irradiation test.

  19. 3D Finite Element Simulation of Pellet-Cladding Mechanical Interaction

    International Nuclear Information System (INIS)

    Seo, Sang Kyu; Lee, Sung Uk; Lee, Eun Ho; Yang, Dong Yol; Kim, Hyo Chan; Yang, Dong Yol

    2016-01-01

    In a nuclear power plant, the fuel assembly, which is composed of fuel rods, burns, and the high temperature can generate power. The fuel rod consists of pellets and a cladding that covers the pellets. It is important to understand the pellet-cladding mechanical interaction with regard to nuclear safety. This paper proposes simulation of the PCMI. The gap between the pellets and the cladding, and the contact pressure are very important for conducting thermal analysis. Since the gap conductance is not known, it has to be determined by a suitable method. This paper suggests a solution. In this study, finite element (FE) contact analysis is conducted considering thermal expansion of the pellets. As the contact causes plastic deformation, this aspect is considered in the analysis. A 3D FE module is developed to analyze the PCMI using FORTRAN 90. The plastic deformation due to the contact between the pellets and the cladding is the major physical phenomenon. The simple analytical solution of a cylinder is proposed and compared with the fuel rod performance code results

  20. Optimization of UO{sub 2} Granule Characteristics for UO{sub 2}-Mo Pellet Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dongjoo; Rhee, Young Woo; Kim, Jong Hun; Kim, Keon Sik; Oh, Jang Soo; Yang, Jae Ho; Koo, Yanghyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The in-reactor performance, integrity, safety and accident tolerance of the nuclear fuel can be significantly affected by the thermal conductivity of the UO{sub 2} fuel pellet. The improvement in the thermal conductivity of the UO{sub 2} fuel pellet can enhance the fuel performance in various ways. Typically, the FGR (Fission Gas Release) can be reduced by the application of a large-grain fuel pellet because the moving path of the fission gas to the grain boundary is much longer. In addition, the mobility of the fission gases is reduced by the lower temperature gradient in the UO{sub 2} fuel pellet. That is to say, the capacity of the fission gas retention of the fuel pellet can increase. In addition, the lower centerline temperature of the fuel pellet affects the accident tolerance for nuclear fuel as well as the enhancement of fuel safety and fuel pellet integrity under normal operation conditions. In addition, the nuclear reactor power can be uprated owing to the higher safety margin. Thus, many researches on enhancing the thermal conductivity of a nuclear fuel pellet for LWRs have been performed in various ways. From the viewpoint of the development of fuel pellet fabrication technology, an enhancement of the thermal conductivity of a pellet can be obtained by the addition of a higher thermal conductive material in the UO{sub 2} pellet. It is known that a UO{sub 2}-metal composite pellet is one of the most effective concepts. However, to maximize the effect of the metallic phase for thermal conductivity enhancement, a continuous channel of the metallic phase in the UO{sub 2} matrix must be formed. Additionally, if the fabrication process of a UO{sub 2}-metal composite pellet is compatible with a conventional sintering process, the developed technology will be favorable. To enhance the thermal conductivity of a UO{sub 2} pellet, there are the various methods for an appropriate arrangement of the high thermal conductive material in a UO{sub 2} matrix. In this

  1. Finite element method programs to analyze irradiation behavior of fuel pellets

    International Nuclear Information System (INIS)

    Yamada, Rayji; Harayama, Yasuo; Ishibashi, Akihiro; Ono, Masao.

    1979-09-01

    For the safety assessment of reactor fuel, it is important to grasp local changes of fuel pins due to irradiation in a reactor. Such changes of fuel result mostly from irradiation of fuel pellets. Elasto-plastic analysis programs based on the finite element method were developed to analyze these local changes. In the programs, emphasis is placed on the analysis of cracks in pellets; the interaction between cracked-pellets and cladding is not taken into consideration. The two programs developed are FEMF3 based on a two-dimensional axially symmetric model (r-z system) and FREB4 on a two-dimensional plane model (r-theta system). It is discussed in this report how the occurrence and distribution of cracks depend on heat rate of the fuel pin. (author)

  2. Studies on a burner used biomass pellets as fuel. Performance of a spiral vortex pellet burner

    Energy Technology Data Exchange (ETDEWEB)

    Iwao, Toshio

    1987-12-21

    In order to develop a small size burner with high performance using biomass pellets fuel substitute for fuel oil, the burning performance of a spiral vortex pallet burner has been studied. An experimental equipment for the pellet burning is made up of a fuel supply unit, combustion chamber and a furnace. The used woody pellet is made of mixed sawdust and bark; with water content of 10.29%, particle diameter of 5.5-9mm, length of 5-50mm, and, apparent and real specific gravities are 0.59 and 1.334 respectively. The pellets are sent to bottom of the combustion chamber, spiral vortex combustion are carried out with blown air, the ashes and unburnt residues are discharged to out of combustion chamber with spiral vortex hot gases. As the result, it was clarified that the formation of the burning layers in a burner is required to be in order of the layers of ash, oxidation, reduction and carbonization up to the upper layer for high burning performance, and the formation of the layer is influenced by the condition of sedimentation of pellets in the combustion chamber. In the meanwhile the burning performance of the burner is influenced by the quantity of blast, the rate of feeding, and by the time of pre-heating in the combustion chamber. (23 figs, 5 refs)

  3. Measuring method for heat-shrinkage of fuel pellet

    International Nuclear Information System (INIS)

    Komono, Akira; Ishizaki, Jin; Inaki, Kiyohiro.

    1997-01-01

    The present invention concerns a method of determining an amount of heat-shrinkage of UR 2 pellets containing gadolinium oxide (Gd 2 O 2 ) based on the difference of the density thereof before and after heating. In a heat shrinkage test of UO 2 pellets containing from 1.0 to 15.0% by weight of gadolinium oxide, the amount of heat-shrinkage is measured under the condition of heat-retaining temperature: from 1700 to 1750degC, temperature elevation time and lowering time: from 90 to 120mins, heat-retaining time: 24hours, inert gas atmosphere, gas pressure: 0.35kg/cm 2 and gas dew point: from -55 to 40degC without changing O/M. This invention has a feature in the use of the inert gas and the elevation of the dew point of the gas. Then, oxygen dissociation phenomenon from crystal lattices of the fuel pellets is suppressed, and normal densification value is shown. Then, fuel pellets of good quality with less fluctuation of the heat-shrinkage can be obtained. (N.H.)

  4. Westinghouse Advanced Doped Pellet - Characteristics and irradiation behavior

    International Nuclear Information System (INIS)

    Backman, K.; Hallstadius, L.; Roennberg, G.

    2009-01-01

    Full text: There are a number of trends in the nuclear power industry, which put additional requirements on the operational flexibility and reliability of nuclear fuel, for example power uprates and longer cycles in order to increase production, higher burnup levels in order to reduce the backend cost of the fuel cycle, and lower goals for activity release from power plant operation. These additional requirements can be addressed by increasing the fuel density, improving the FG retention, improving the PCI resistance and improving the post-failure performance. In order to achieve that, Westinghouse has developed ADOPT (Advanced Doped Pellet Technology) UO 2 fuel containing additions of chromium and aluminium oxides. The additives facilitate pellet densification during sintering, enlarge the pellet grain size, and increase the creep rate. The final manufactured doped pellets reach about 0.5 % higher density within a shorter sintering time and a five times larger grain size compared with standard UO 2 fuel pellets. Fuel rods with ADOPT pellets have been irradiated in several light water reactors (LWRs) since 1999, including two full SVEA Optima2 reloads in 2005. ADOPT pellets has been investigated in pool-side and hot cell Post Irradiation Examinations (PIEs), as well as in a ramp test and a fuel washout test in the Studsvik R2 test reactor. The investigations have identified three areas of improved operational behaviour: Reduced Fission Gas Release (FGR), improved Pellet Cladding Interaction (PCI) performance thanks to increased pellet plasticity and higher resistance against post-failure degradation. The better FGR behaviour of ADOPT has been verified with a pool side FGR gamma measurement performed at 55 MWd/kgU, as well as transient tests in the Studsvik R2 reactor. Creep measurements performed on fresh pellets show that ADOPT has a higher creep rate which is beneficial for the PCI performance. ADOPT has also been part of a high power Halden test (IFA-677). The

  5. Galvanic cell for processing of used nuclear fuel

    Science.gov (United States)

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2017-02-07

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  6. Electrochemical fluorination for processing of used nuclear fuel

    Science.gov (United States)

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2016-07-05

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  7. Nuclear fuel recycling system

    International Nuclear Information System (INIS)

    Lee, H.R.; Koch, A.K.; Krawczyk, A.

    1981-01-01

    A process is provided for recycling sintered uranium dioxide fuel pellets rejected during fuel manufacture and the swarf from pellet grinding. The scrap material is prepared mechanically by crushing and milling as a high solids content slurry, using scrap sintered UO 2 pellets as the grinding medium under an inert atmosophere

  8. The market for fuel pellets produced from biomass and waste in the Netherlands

    International Nuclear Information System (INIS)

    Koppejan, J.; Meulman, P.D.M.

    2001-12-01

    Several initiatives are currently being developed in the Netherlands for the production of fuel pellets from waste and biomass. This report presents an overview of the current producers and (potential) users of these pellets in the Netherlands. It also outlines the Dutch and European policies and legislations concerned. Furthermore, important barriers to market development of fuel pellets are defined and future expectations are summarized. The study covers fuel pellets made from different feedstock, varying from clean biomass to waste with traces of contaminants. In each project, pellets are produced that are unique as to their product specifications, as they are usually designed for a single application. It is therefore impossible to generalize characteristics and end use. 27 refs

  9. Fuel Pellets from Wheat Straw: The Effect of Lignin Glass Transition and Surface Waxes on Pelletizing Properties

    Science.gov (United States)

    Wolfgang Stelte; Craig Clemons; Jens K. Holm; Jesper Ahrenfeldt; Ulrik B. Henriksen; Anand R. Sanadi

    2012-01-01

    The utilization of wheat straw as a renewable energy resource is limited due to its low bulk density. Pelletizing wheat straw into fuel pellets of high density increases its handling properties but is more challenging compared to pelletizing wood biomass. Straw has a lower lignin content and a high concentration of hydrophobic waxes on its outer surface that may limit...

  10. Modelling of pellet-cladding interaction for PWRs reactors fuel rods

    International Nuclear Information System (INIS)

    Esteves, A.M.

    1991-01-01

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyzes the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. Linear and non-linear material behaviors are allowed. Elastic, plastic and creep behaviors are considered for the cladding materials. The modelling is applied to Angra-II fuel rod design. The results are analyzed and compared. (author)

  11. Fuel-pellet-fabrication experience using direct-denitration-recycle-PuO2-coprecipitated mixed oxide

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Schaus, P.S.

    1980-01-01

    The fuel pellet fabrication experience described in this paper involved three different feed powders: coprecipitated PuO 2 -UO 2 which was flash calcined in a fluidized bed; co-direct denitrated PuO 2 -UO 2 ; and direct denitrated LWR recycle PuO 2 which was mechanically blended with natural UO 2 . The objectives of this paper are twofold; first, to demonstrate that acceptable quality fuel pellets were fabricated using feed powders manufactured by processes other than the conventional oxalate process; and second, to highlight some pellet fabrication difficulties experienced with the direct denitration LWR recycle PuO 2 feed material, which did not produce acceptable pellets. The direct denitration LWR recycle PuO 2 was available as a by-product and was not specifically produced for use in fuel pellet fabrication. Nevertheless, its characteristics and pellet fabrication behavior serve to re-emphasize the importance of continued process development involving both powder suppliers and fuel fabricators to close the fuel cycle in the future

  12. Study of production of fuel pellets for a reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mendes, Luiz F.F.; Conti, Thadeu N., E-mail: luiz.f.f.mendes@gmail.com, E-mail: tnconti@yahoo.com.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    Nowadays the electrical energy was been used much on society. A method for getting electricity is through nuclear power plants, this power plant uses fission that occurs inside the UO{sub 2} pellets to generate thermal energy that will be transform into electric. The pellets production was made from enriched UF{sub 6} uses some techniques of reprocessing UF{sub 6} gas to UO{sub 2} powder. This reprocessing process done by wet route (Ammonium Diuranate ADU or Ammonium Uranium Carbonate AUC) or by dry route (Fluidized bed or GECO). With getting of UO{sub 2} powder is forwarded to metallurgy where this powder is compacted in cylindrical matrix so that powder take the desired shape, this green pellets are full of the empty spaces (porosity) for this it is sent to the sintering. The sintering consists of a joint of these particles of powders by means of the heating of this green pellets, coming arrive the melting temperature, the UO{sub 2} molecules melting each other so decrease the porosity and increase the density. For the production of fuel pellets the process all most used is wed route by means the AUC ,this process arrive created for replace the ADU because the AUC is a process where less rework for the pore geometry is required compared to DUA. The fluidized bed process is more used in small samples however, for a large amount it becomes unfeasible, moreover the dry route process require more robust materials because of the generation of HF that is highly corrosive and cannot used the UNH (uranyl nitrate hexahydrate) used for recycle materials discarded in manufacturing. (author)

  13. Study of production of fuel pellets for a reactor

    International Nuclear Information System (INIS)

    Mendes, Luiz F.F.; Conti, Thadeu N.

    2017-01-01

    Nowadays the electrical energy was been used much on society. A method for getting electricity is through nuclear power plants, this power plant uses fission that occurs inside the UO 2 pellets to generate thermal energy that will be transform into electric. The pellets production was made from enriched UF 6 uses some techniques of reprocessing UF 6 gas to UO 2 powder. This reprocessing process done by wet route (Ammonium Diuranate ADU or Ammonium Uranium Carbonate AUC) or by dry route (Fluidized bed or GECO). With getting of UO 2 powder is forwarded to metallurgy where this powder is compacted in cylindrical matrix so that powder take the desired shape, this green pellets are full of the empty spaces (porosity) for this it is sent to the sintering. The sintering consists of a joint of these particles of powders by means of the heating of this green pellets, coming arrive the melting temperature, the UO 2 molecules melting each other so decrease the porosity and increase the density. For the production of fuel pellets the process all most used is wed route by means the AUC ,this process arrive created for replace the ADU because the AUC is a process where less rework for the pore geometry is required compared to DUA. The fluidized bed process is more used in small samples however, for a large amount it becomes unfeasible, moreover the dry route process require more robust materials because of the generation of HF that is highly corrosive and cannot used the UNH (uranyl nitrate hexahydrate) used for recycle materials discarded in manufacturing. (author)

  14. A system automatic study for the spent fuel rod cutting and simulated fuel pellet extraction device

    International Nuclear Information System (INIS)

    Jeong, J. H.; Yun, J. S.; Hong, D. H.; Kim, Y. H.; Park, K. Y.

    2001-01-01

    A fuel pellet extraction device of the spent fuel rods is described. The device consists of a cutting device of the spent fuel rods and the decladding device of the fuel pellets. The cutting device is to cut a spent fuel rod to n optimal size for fast decladding operation. To design the device, the fuel rod properties are investigated including the dimension and material of fuel rod tubes and pellets. Also, various methods of existing cutting method are investigated. The design concepts accommodate remote operability for the Hot-Cell(radioactive ) area operation. Also, the modularization of the device structure is considered for the easy maintenance. The decladding device is to extract the fuel pellet from the rod cut. To design this device, the existing method is investigated including the chemical and mechanical decladding methods. From the view point of fuel recovery and feasibility of implementation. it is concluded that the chemical decladding method is not appropriate due to the mass production of radioactive liquid wastes, in spite of its high fuel recovery characteristics. Hence, in this paper, the mechanical decladding method is adopted and the device is designed so as to be applicable to various lengths of rod-cuts. As like the cutting device,the concepts of remote operability and maintainability is considered. Both devices are fabricated and the performance is investigated through a series of experiments. From the experimental result, the optimal operational condition of the devices is established

  15. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Obara, Hiroshi.

    1981-01-01

    Purpose: To suppress iodine release thereby prevent stress corrosion cracks in fuel cans by dispersing ferrous oxide at the outer periphery of sintered uranium dioxide pellets filled and sealed within zirconium alloy fuel cans of fuel elements. Constitution: Sintered uranium dioxide pellets to be filled and sealed within a zirconium alloy fuel can are prepared either by mixing ferric oxide powder in uranium dioxide powder, sintering and then reducing at low temperature or by mixing iron powder in uranium dioxide powder, sintering and then oxidizing at low temperature. In this way, ferrous oxide is dispersed on the outer periphery of the sintered uranium dioxide pellets to convert corrosive fission products iodine into iron iodide, whereby the iodine release is suppressed and the stress corrosion cracks can be prevented in the fuel can. (Moriyama, K.)

  16. An infrared technique for on-line detection of orientation of PHWR fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Behere, P G [Bhabha Atomic Research Centre, Tarapur (India). Advanced Fuel Fabrication Facility

    1994-12-31

    The PHWR (Pressurised Heavy Water Reactor) fuel pellets fabricated in a fuel fabrication plant are cylindrical in shape and after sintering acquire a nominal size of 14.3 mm diameter and 17 mm height. These pellets have dish at one end while the other end is flat. The dish is provided to accommodate fission gases and thermal expansion. The sintered pellets are examined for physical damages such as cracks, chippings etc. and these should have one particular orientation while loading. A technique is suggested to solve the problems arising during the fuel pellet loadings. 3 figs.

  17. An infrared technique for on-line detection of orientation of PHWR fuel pellets

    International Nuclear Information System (INIS)

    Behere, P.G.

    1994-01-01

    The PHWR (Pressurised Heavy Water Reactor) fuel pellets fabricated in a fuel fabrication plant are cylindrical in shape and after sintering acquire a nominal size of 14.3 mm diameter and 17 mm height. These pellets have dish at one end while the other end is flat. The dish is provided to accommodate fission gases and thermal expansion. The sintered pellets are examined for physical damages such as cracks, chippings etc. and these should have one particular orientation while loading. A technique is suggested to solve the problems arising during the fuel pellet loadings. 3 figs

  18. Increase of thermal conductivity of uranium dioxide nuclear fuel pellets with beryllium oxide addition; Condutividade termica de pastilhas de combustivel nuclear de UO{sub 2}-BeO nas temperaturas de 25 deg C e 100 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Camarano, D.M.; Mansur, F.A.; Santos, A.M.M. dos; Ferraz, W.B., E-mail: dmc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTM/CNEN-MG), Belo Horizonte, MG (Brazil)

    2016-07-01

    The UO{sub 2} fuel is one of the most used nuclear fuel in thermal reactors and has many advantages such as high melting point, chemical compatibility with cladding, etc. However, its thermal conductivity is relatively low, which leads to a premature degradation of the fuel pellets due to a high radial temperature gradient during reactor operation. An alternative to avoid this problem is to increase the thermal conductivity of the fuel pellets, by adding beryllium oxide (BeO). Pellets of UO{sub 2} and UO{sub 2}-BeO were obtained from a homogenized mixture of powders of UO{sub 2} and BeO, containing 2% and 3% by weight of BeO and sintering at 1750 °C for 3 h under H{sub 2} atmosphere after uniaxial pressing at 400 MPa. The pellet densities were obtained by xylol penetration-immersion method and the thermal diffusivity, specific heat and thermal conductivity were determined according to ASTM E-1461 at room temperature (25 deg C) and 100 deg C. The thermal diffusivity measurements were carried out employing the laser flash method. The thermal conductivity obtained at 25 deg C showed an increase with the addition of 2% and 3% of BeO corresponding to 19% and 28%, respectively. As for the measurements carried out at 100 deg C, there was an increase in the thermal conductivity for the same BeO contents of 20% and 31%. These values as a percentage of increased conductivity were obtained in relation to the UO{sub 2} pellets. (author)

  19. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  20. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Domoto, Noboru; Masuda, Hiroyuki

    1989-01-01

    In a nuclear fuel assembly loaded with a plurality of fuel rods, the inside of a fuel rod disposed at a high neutron flux region is divided into an inner region and an outer region, and more burnable poisons are mixed in the inner region than in the outer region. Alternatively, the central portion of a pellet disposed in a high neutron flux region is made hollow, in which burnable poisons are charged. This can prevent neutron infinite multiplication factor from decreasing extremely at the initial burning stage. Further, the burnable poisons are not rapidly burnt completely and local peaking coefficient can be controlled. Accordingly, in a case of suppressing a predetermined excess reactivity by using a fuel rod incorporated with the burnable poison, the fuel economy can be improved more and the reactor core controllability can also be improved as compared with the usual case. (T.M.)

  1. Thermochemical modeling of nuclear fuel and the effects of oxygen potential buffers

    Energy Technology Data Exchange (ETDEWEB)

    Loukusa, Henri, E-mail: henri.loukusa@vtt.fi; Ikonen, Timo; Valtavirta, Ville; Tulkki, Ville

    2016-12-01

    The elemental and chemical composition of nuclear fuel pellets are key factors influencing the material properties of the pellets. The oxidation state of the fuel is one of the most important chemical properties influencing the material properties of the fuel, and it can only be determined with the knowledge of the chemical composition. A measure of the oxidation state is the oxygen chemical potential of the fuel. It can be buffered by redox pairs, such as the well-known Mo/MoO{sub 2} pair. In this work, the elemental composition of the fuel is obtained from a burnup calculation and the temperature and pressure calculated with a fuel performance code. An estimate of the oxygen potential of fuel is calculated with Gibbs energy minimization. The results are compared against experimental data from the literature. The significance of the UMoO{sub 6} compound and its buffering effect on the oxygen potential is emphasized. - Highlights: • A Gibbs energy minimization routine has been developed for nuclear fuel modeling. • The initial stoichiometry affects the development of the oxygen potential of fuel. • UMoO{sub 6} is found to buffer the oxygen potential of nuclear fuel.

  2. Reference Neutron Radiographs of Nuclear Reactor Fuel

    DEFF Research Database (Denmark)

    Domanus, Joseph Czeslaw

    1986-01-01

    Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts...... of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as fabricated parts are included. The collection contains 158 neutron radiographs, reproduced on photographic paper...

  3. Emissions from small scale combustion of pelletized wood fuels

    International Nuclear Information System (INIS)

    Bachs, A.

    1998-01-01

    Combustion of wood pellets in small scale heating systems with an effect below 20 kW has increased. During the winter season 1995/96 1500 small plants for heating houses are estimated to be in operation. Stack emissions from three pellet burners and two pellet stoves have been studied at laboratory. Different pellet qualities were tested. When the fraction of fines increased also the NO x emissions increased with about 10 %. As reference fuel 8 mm pellets was used. Tests with 6 mm pellets gave, in most cases, significant lower emissions of CO and THC. Eleven stoves, burners and boilers were studied in a field test. The results show that all the plants generally have higher emissions in the field than during conditions when the plants are adjusted with a stack gas monitoring instrument. A conclusion is that it is difficult for the operator to adjust the plant without a monitoring instrument. The emissions from the tested plants give an estimation of stack gas emissions from small scale pellet plants. The difference between the 'best' and 'worst' technologies is big. The span of emissions with the best technology to the worst is given below. The interval is concerning normal combustion . During abnormal conditions the emissions are on a significant higher level: * CO 80-1 000 mg/MJ; * Tar 0,3-19 mg/MJ; * THC (as methane equivalents) 2-100 mg/MJ; * NO x 50-70 mg/W;, and * Dust emissions 20-40 mg/MJ. Emissions from pellets heating are lower than from wood combustion and the best technology is close to the emission from oil burners. Wood and pellets have the same origin but the conditions to burn them in an environmental friendly way differ. Combustion of pellets could be improved through improved control of the air and fuel ratio that will create more stable conditions for the combustion

  4. Operating experience with Exxon nuclear advanced fuel assembly and fuel cycle designs in PWRs

    International Nuclear Information System (INIS)

    Skogen, F.B.; Killgore, M.R.; Holm, J.S.; Brown, C.A.

    1986-01-01

    Exxon Nuclear Company (ENC) has achieved a high standard of performance in its supply of fuel reloads for both BWRs and PWRs, while introducing substantial innovations aimed at realization of improved fuel cycle costs. The ENC experience with advanced design features such as the bi-metallic spacer, the dismountable upper tie plate, natural uranium axial blankets, optimized water-to-fuel designs, annular pellets, gadolinia burnable absorbers, and improved fuel management scenarios, is summarized

  5. Accuracy of dimension measurements from neutron radiographs of nuclear fuel pins

    International Nuclear Information System (INIS)

    Domanus, J.C.

    1976-01-01

    A comparison is given of accuracies obtained with measuring the dimensions (pellet diameter and fuel-clad gap) from neutron and X-ray radiographs of a calibrated nuclear fuel pin performed with a projection microscope, microdensitometers and a video micrometer

  6. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E. D.

    1984-01-01

    An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value

  7. Nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Hindle, E. D.

    1984-10-16

    An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value.

  8. Prototype fuel fabrication for nuclear reactors of Laguna Verde

    International Nuclear Information System (INIS)

    Nocetti, C.; Torres, J.; Medrano, A.

    1996-01-01

    Four prototype fuel bundles for the Laguna Verde Nuclear Power Plant have been fabricated. the type of nuclear fuel produced is described and the process used is commented. As an example of the fabrication criteria adopted, the production model to determine the density of the U O 2 pellets for the different batches of ceramic powder is described. the results are evaluated using the statistical indexes C p and C pk . (author)

  9. Performance of a pellet boiler fired with agricultural fuels

    International Nuclear Information System (INIS)

    Carvalho, Lara; Wopienka, Elisabeth; Pointner, Christian; Lundgren, Joakim; Verma, Vijay Kumar; Haslinger, Walter; Schmidl, Christoph

    2013-01-01

    Highlights: ► Performance evaluation of a pellet boiler operated with different agricultural fuels. ► Agricultural fuels could be burn in the tested boiler for a certain period of time. ► All the fuels (except straw and Sorghum) satisfied the European legal requirements. ► Boilers for burning agricultural fuels should have a flexible control system. - Abstract: The increasing demand for woody biomass increases the price of this limited resource, motivating the growing interest in using woody materials of lower quality as well as non-woody biomass fuels for heat production in Europe. The challenges in using non-woody biomass as fuels are related to the variability of the chemical composition and in certain fuel properties that may induce problems during combustion. The objective of this work has been to evaluate the technical and environmental performance of a 15 kW pellet boiler when operated with different pelletized biomass fuels, namely straw (Triticum aestivum), Miscanthus (Miscanthus × giganteus), maize (Zea mays), wheat bran, vineyard pruning (from Vitis vinifera), hay, Sorghum (Sorghum bicolor) and wood (from Picea abies) with 5% rye flour. The gaseous and dust emissions as well as the boiler efficiency were investigated and compared with the legal requirements defined in the FprEN 303-5 (final draft of the European standard 303-5). It was found that the boiler control should be improved to better adapt the combustion conditions to the different properties of the agricultural fuels. Additionally, there is a need for a frequent cleaning of the heat exchangers in boilers operated with agricultural fuels to avoid efficiency drops after short term operation. All the agricultural fuels satisfied the legal requirements defined in the FprEN 303-5, with the exception of dust emissions during combustion of straw and Sorghum. Miscanthus and vineyard pruning were the best fuels tested showing comparable emission values to wood combustion

  10. Standard specification for nuclear-grade aluminum oxide pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This specification applies to pellets of aluminum oxide that may be ultimately used in a reactor core, for example, as filler or spacers within fuel, burnable poison, or control rods. In order to distinguish between the subject pellets and “burnable poison” pellets, it is established that the subject pellets are not intended to be used as neutron-absorbing material. 1.2 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.

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

    International Nuclear Information System (INIS)

    Sahoo, K.C.; Bahl, J.K.; Sivaramakrishnan, K.S.; Roy, P.R.

    1981-01-01

    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)

  12. Model for the behaviour of thorium and uranium fuels at pelletization; Modelo para o comportamento de microesferas combustiveis de torio e uranio na peletizacao

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira Neto, Ricardo Alberto

    2000-11-15

    In this work, a model for the behaviour of thorium-uranium-mixed oxide microspheres in the pelletizing process is presented. This model was developed in a program whose objective was to demonstrate the viability of producing fissile material through the utilization of thorium in pressurized water reactors. This is important because it allows the saving of the strategic uranium reserves, and makes it possible the nuclear utilization of the large brazilian thorium reserves. The objective was to develop a model for optimizing physical properties of the microspheres, such as density, fracture strength and specific surface, so as to produce fuel pellets with microstructure, density, open porosity and impurity content, in accordance with the fuel specification. And, therefore, to adjust the sol-gel processing parameters in order to obtain these properties, and produce pellets with an optimized microstructure, adequate to a stable behaviour under irradiation. The model made it clear that to achieve this objective, it is necessary to produce microspheres with density and specific surface as small as possible. By changing the sol-gel processing parameters, microspheres with the desired properties were produced, and the model was experimentally verified by manufacturing fuel pellets with optimized microstructures, density, open porosity and impurity content, meeting the specifications for this new nuclear fuel for pressurized water reactors. Furthermore it was possible to obtain mathematical expressions that enables to calculate from the microspheres properties and the utilized compaction pressure, the sinter density that will be obtained in the sintered pellet and the necessary compaction pressure to reach the sintered density specified for the fuel. (author)

  13. Alteration in fuel processing at Tokai Works of Mitsubishi Nuclear Fuel Co., Ltd

    International Nuclear Information System (INIS)

    1977-01-01

    The report of the Committee on Examination of Nuclear Fuel Safety to the Atomic Energy Commission of Japan concerning the alteration is given, which is attached to the reply from the commission to the prime minister, and its safety was confirmed. The alterations are installation of the storage for transport containers containing fuel assemblies, construction of radiation control and other buildings; and improvement and installation of the facilities for chemical-processing, pellet fabrication, fuel assembling, and storage. (Mori, K.)

  14. Fueling of magnetically confined plasmas by single- and two-stage repeating pneumatic pellet injectors

    International Nuclear Information System (INIS)

    Gouge, M.J.; Combs, S.K.; Foust, C.R.; Milora, S.L.

    1990-01-01

    Advanced plasma fueling systems for magnetic fusion confinement experiments are under development at Oak Ridge National Laboratory (ORNL). The general approach is that of producing and accelerating frozen hydrogenic pellets to speeds in the kilometer-per-second range using single shot and repetitive pneumatic (light-gas gun) pellet injectors. The millimeter-to-centimeter size pellets enter the plasma and continuously ablate because of the plasma electron heat flux, depositing fuel atoms along the pellet trajectory. This fueling method allows direct fueling in the interior of the hot plasma and is more efficient than the alternative method of injecting room temperature fuel gas at the wall of the plasma vacuum chamber. Single-stage pneumatic injectors based on the light-gas gun concept have provided hydrogenic fuel pellets in the speed range of 1--2 km/s in single-shot injector designs. Repetition rates up to 5 Hz have been demonstrated in repetitive injector designs. Future fusion reactor-scale devices may need higher pellet velocities because of the larger plasma size and higher plasma temperatures. Repetitive two-stage pneumatic injectors are under development at ORNL to provide long-pulse plasma fueling in the 3--5 km/s speed range. Recently, a repeating, two-stage light-gas gun achieved repetitive operation at 1 Hz with speeds in the range of 2--3 km/s

  15. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    Shoup, R.L.

    1976-07-01

    The fabrication of fuel capsules with refractory metal and alloy clads used in nuclear-powered cardiac pacemakers precludes the expedient dissolution of the clad in inorganic acid solutions. An experiment to measure penetration rates of acids on commonly used fuel pellet clads indicated that it is not impossible, but that it would be very difficult to dissolve the multiple cladding. This work was performed because of a suggestion that a 238 PuO 2 -powered pacemaker could be transformed into a terrorism weapon

  16. High performance nuclear fuel element

    International Nuclear Information System (INIS)

    Mordarski, W.J.; Zegler, S.T.

    1980-01-01

    A fuel-pellet composition is disclosed for use in fast breeder reactors. Uranium carbide particles are mixed with a powder of uraniumplutonium carbides having a stable microstructure. The resulting mixture is formed into fuel pellets. The pellets thus produced exhibit a relatively low propensity to swell while maintaining a high density

  17. A Comparison of Fueling with Deuterium Pellet Injection from Different Locations on the DIII-D Tokamak

    International Nuclear Information System (INIS)

    Baylor, L.R.; Combs, S.K.; Gohil, P.; Houlberg, W.A.; Hsieh, C.; Jernigan, T.C.; Parks, P.B.

    1999-01-01

    Initial pellet injection experiments on DIII-D with high field side (HFS) injection have demonstrated that deeper pellet fuel deposition is possible even with HFS injected pellets that are significantly slower than pellets injected from the low field side (LFS) (outer midplane) location. A radial displacement of the pellet mass shortly after or during the ablation process is consistent with the observed mass deposition profiles measured shortly after injection. Vertical injection inside the magnetic axis shows some improvement in fueling efficiency over LFS injection and may provide an optimal injection location for fueling with high speed pellets

  18. Nuclear fuel element

    International Nuclear Information System (INIS)

    Yamanaka, Tsuneyasu.

    1976-01-01

    Purpose: To provide a mechanism for the prevention of fuel pellet dislocation in fuel can throughout fuel fablication, fuel transportation and reactor operation. Constitution: A plenum spacer as a mechanism for the prevention of fuel pellet dislocation inserted into a cladding tube comprises split bodies bundled by a frame and an expansion body being capable of inserting into the central cavity of the split bodies. The expansion body is, for example, in a conical shape and the split bodies are formed so that they define in the center portion, when disposed along the inner wall of the cladding tube, a gap capable of inserting the conical body. The plenum spacer is assembled by initially inserting the split bodies in a closed state into the cladding tube after the loading of the pellets, pressing their peripheral portions and then inserting the expansion body into the space to urge the split bodies to the inner surface of the cladding tube. (Kawakami, Y.)

  19. A mathematical model of an automatic assembler to stack fuel pellets

    International Nuclear Information System (INIS)

    Jarvis, R.G.; Joynes, R.; Bretzlaff, C.I.

    1980-11-01

    Fuel elements for CANDU reactors are assembled from stacks of cylindrical UO 2 pellets, with close tolerances on lengths and diameters. Present stacking techniques involve extensive manual operations and they can be speeded up and reduced in cost by an automated device. If gamma-active fuel is handled such a device is essential. An automatic fuel pellet assembly process was modelled mathematically. The model indicated a suitable sequence of pellet manipulations to arrive at a stack length that was always within tolerance. This sequence was used as the inital input for the design of mechanical hardware. The mechanical design and the refinement of the mathematical model proceeded simultaneously. Mechanical constraints were allowed for in the model, and its optimized sequence of operations was incorporated in a microcomputer program to control the mechanical hardware. (auth)

  20. Fuel can for a nuclear reactor

    International Nuclear Information System (INIS)

    Shimizu, Shigeo.

    1984-01-01

    Purpose: To decrease the possibility of damages in a fuel can by avoiding the close contact of the outer circumferential surface of a pellet to the entire inner circumference of the fuel can in the case if the pellet undergoes heat expansion. Constitution: The inner circumference of a fuel can includes at least three linear portions each with an equi-angular distance. The center for the circle (radius R2) inscribing each of the linear portions aligns with the axial center of the fuel can. A gap is formed to each inscribing circle with a band-like circular inner wall. The radius R2 for the inscribing circle is made larger than the radius R1 for the pellet and the length of the linear portion and the radius R2 for the inscribing circle are determined to desired values in view of the fuel design. If the fuel pellet expands thermally during reactor operation, since a gap is remained between the outer circumferential surface of the pellet and the inner circumferential surface of the fuel can and the outer circumferential surface of the pellet is not in close contact entirely with the inner circumferential surface of the fuel can, the possibility of damaging the fuel can is decreased. (Seki, T.)

  1. Determination Of Simulated Pellet To Pellet Gap Using Neutron Radiography

    International Nuclear Information System (INIS)

    Kusnowo, A.

    1996-01-01

    The defect on the irradiated fuel element could be detected using neutron radiography. The defect could occurred in pellet to pellet gap, cladding, or even cladding to pellet gap. An investigations has been performed to detect pellet to pellet gap defect that might occur in an irradiated fuel element. An Al foil of 0,1; 0,2; 0,3; und 0,4 mm was inserted between pellets to simulate various pellet to pellet gap. The neutron radiography used had power of 700 kW. The result showed that this simulation represented well enough problems that irradiated fuel element may experience

  2. Synthesis of dense yttrium-stabilised hafnia pellets for nuclear applications by spark plasma sintering

    International Nuclear Information System (INIS)

    Tyrpekl, Vaclav; Holzhäuser, Michael; Hein, Herwin; Vigier, Jean-Francois; Somers, Joseph; Svora, Petr

    2014-01-01

    Graphical abstract: Densification of HfO 2 –Y 2 O 3 micro-beads by Spark Plasma Sintering High density pellets with homogenous distribution of Hf and Y serve as neutron absorbers. - Abstract: Dense yttrium–stabilised hafnia pellets (91.35 wt.% HfO 2 and 8.65 wt.% Y 2 O 3 ) were prepared by spark plasma sintering consolidation of micro-beads synthesised by the “external gelation” sol–gel technique. This technique allows a preparation of HfO 2 –Y 2 O 3 beads with homogenous yttria–hafnia solid solution. A sintering time of 5 min at 1600 °C was sufficient to produce high density pellets (over 90% of the theoretical density) with significant reproducibility. The pellets have been machined in a lathe to the correct dimensions for use as neutron absorbers in an experimental test irradiation in the High Flux Reactor (HFR) in Petten, Holland, in order to investigate the safety of americium based nuclear fuels

  3. FUMAC-a new model for light water reactor fuel relocation and pellet-cladding interaction

    International Nuclear Information System (INIS)

    Walton, L.A.; Matheson, J.E.

    1984-01-01

    An improved approach to the mechanical modeling of fuel rod performance is presented. Previous computer modeling has centered around a unified finite element approach with both fuel pellets and cladding being represented by ring elements. The fuel mechanical analysis code (FUMAC) departs from these approaches in two areas. The pellet model is an empirically based deterministic algorithm, while the cladding model uses both plane stress and plane strain finite elements. The work describes a semiempirical fuel cracking and fragment relocation model, which is burnup and power-level dependent. The interaction of the pellet with the cladding is treated classically. The resulting thick cylinder stresses are used in conjunction with an orthotropic creep model to predict cladding ridging. The resulting ridging compares well with experimental data for both steady-state and transient operating conditions. Future work planned includes the integration of the finite element cladding model with the pellet model and refinement of the pellet relocation and thermal models. Transient performance predictions will be emphasized

  4. Pellet fueling of JET plasmas during ohmic, ICRF and NBI heating

    International Nuclear Information System (INIS)

    Gondhalekar, A.; Cheetham, A.; Bures, M.

    1986-01-01

    Pellet fueling experiments have been performed on JET using a single-shot pneumatic injector giving 4.6mm (4.5 x 10 21 D atoms) and 3.6mm (2.2 x 10 21 D atoms) diameter cylindrical deuterium pellets with velocity 0.8 ≤ V(km.s -1 ) ≤ 1.2. Z/sub eff/ 20 m -3 and T/sub e/(0) ≅ 1keV. Separately, high value of n/sub D/(0)tau/sub E/T/sub i/(0) = 1.3 x 10 20 m -3 .s.keV at T/sub i/90) = 6.5keV has been obtained with pellet fueling followed by NBI heating

  5. Construction and engineering report for advanced nuclear fuel development facility

    International Nuclear Information System (INIS)

    Cho, S. W.; Park, J. S.; Kwon, S.J.; Lee, K. W.; Kim, I. J.; Yu, C. H.

    2003-09-01

    The design and construction of the fuel technology development facility was aimed to accommodate general nuclear fuel research and development for the HANARO fuel fabrication and advanced fuel researches. 1. Building size and room function 1) Building total area : approx. 3,618m 2 , basement 1st floor, ground 3th floor 2) Room function : basement floor(machine room, electrical room, radioactive waste tank room), 1st floor(research reactor fuel fabrication facility, pyroprocess lab., metal fuel lab., nondestructive lab., pellet processing lab., access control room, sintering lab., etc), 2nd floor(thermal properties measurement lab., pellet characterization lab., powder analysis lab., microstructure analysis lab., etc), 3rd floor(AHU and ACU Room) 2. Special facility equipment 1) Environmental pollution protection equipment : ACU(2sets), 2) Emergency operating system : diesel generator(1set), 3) Nuclear material handle, storage and transport system : overhead crane(3sets), monorail hoist(1set), jib crane(2sets), tank(1set) 4) Air conditioning unit facility : AHU(3sets), packaged air conditioning unit(5sets), 5) Automatic control system and fire protection system : central control equipment(1set), lon device(1set), fire hose cabinet(3sets), fire pump(3sets) etc

  6. Study of uranium dioxide pellets by micro-acoustic techniques

    International Nuclear Information System (INIS)

    Roque, V.

    1999-01-01

    In order to reduce the volume of spent fuel to reprocess and to improve the productivity and the safety of the nuclear reactor, 'Electricite De France' aim to increase the average fuel discharge burn-up. To elaborate the safety reports, EDF develops codes to simulate the thermo-mechanical behaviour of the nuclear fuel element. These numeric simulations need to evaluate accurately and locally the evolution of the material and of its properties. One of the major concern today is the local characterisation of the intrinsic volume fraction porosity and the mechanical properties of the irradiated fuel. The fuel pellet fragmentation, the steep radial gradient in its physical properties evolution and the chemical evolution of the irradiated material make difficult nay the use of the conventional techniques. This leads EDF to pay interest for the use of two complementary techniques: micro-indentation on the one hand and acoustic methods on the other hand (acoustic microscopy and micro-echography), with an additional constrain to perform on active materials. The objective of this work has been to adapt the acoustic methods for an application on uranium dioxide pellets, used as nuclear fuel in Water Pressurised Reactor. Acquisitions protocols have been set to measure accurately the Rayleigh velocity and the longitudinal velocity of the UO 2 . Using these protocols, we have calibrated these acoustic methods by analysing non irradiated nuclear pellet which properties were well known. This process enable to quantify the effects of different physico-chemical parameters of the UO 2 on the ultrasonic velocities measured. Particularly, the large influence of the porosity has been demonstrated and empirical laws to express the evolution of the acoustic velocities as a function of the volume fraction porosity were established. Moreover, we have established a methodology to characterise the intrinsic elastic constants and the volume fraction porosity on irradiated UO 2 fuel pellets

  7. Development for analysis system of rods enrichment of nuclear fuels; Desarrollo de un sistema de analisis de enriquecimiento de barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E.L

    1998-11-01

    Nuclear industry is strongly regulated all over the world and quality assurance is important in every nuclear installation or process related with it. Nuclear fuel manufacture is not the exception. ININ was committed to manufacture four nuclear fuel bundles for the CFE nucleo electric station at Laguna Verde, Veracruz, under General Electric specifications and fulfilling all the requirements of this industry. One of the quality control requisites in nuclear fuel manufacture deals with the enrichment of the pellets inside the fuel bundle rods. To achieve the quality demanded in this aspect, the system described in this work was developed. With this system, developed at ININ it is possible to detect enrichment spikes since 0.4 % in a column of pellets with a 95 % confidence interval and to identify enrichment differences greater than 0.2 % e between homogeneous segments, also with a 95 % confidence interval. ININ delivered the four nuclear fuel bundles to CFE and these were introduced in the core of the nuclear reactor of Unit 1 in the fifth cycle. Nowadays they are producing energy and have shown a correct mechanical performance and neutronic behavior. (Author)

  8. Demonstration of fuel resistant to pellet-cladding interaction. Second semiannual report, January--June 1978

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.

    1978-09-01

    This program has as its ultimate objective the demonstration of an advanced fuel concept that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Since currently used fuel in the nuclear power industry is subject to the PCI failure mechanism, reactor operators limit the rates of power increases and thus reduce their capacity factors in order to protect the fuel. Two concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as ''barrier fuels'') have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress and reactive fission products during reactor service. The demonstration of one of these concepts in a commercial power reactor is planned for PHASE 2 of this program. The current plans for the demonstration will involve approximately 132 bundles of PCI-resistant fuel

  9. Development of nuclear fuel cycle technologies

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Matsumoto, Takashi; Suzuki, Kazumichi; Kawamura, Fumio

    1995-01-01

    In the long term plan for atomic energy that the Atomic Energy Commission decided the other day, the necessity of the technical development for establishing full scale fuel cycle for future was emphasized. Hitachi Ltd. has engaged in technical development and facility construction in the fields of uranium enrichment, MOX fuel fabrication, spent fuel reprocessing and so on. In uranium enrichment, it took part in the development of centrifuge process centering around Power Reactor and Nuclear Fuel Development Corporation (PNC), and took its share in the construction of the Rokkasho uranium enrichment plant of Japan Nuclear Fuel Service Co., Ltd. Also it cooperates with Laser Enrichment Technology Research Association. In Mox fuel fabrication, it took part in the construction of the facilities for Monju plutonium fuel production of PNC, for pellet production, fabrication and assembling processes. In spent fuel reprocessing, it cooperated with the technical development of maintenance and repair of Tokai reprocessing plant of PNC, and the construction of spent fuel stores in Rokkasho reprocessing plant is advanced. The centrifuge process and the atomic laser process of uranium enrichment are explained. The high reliability of spent fuel reprocessing plants and the advancement of spent fuel reprocessing process are reported. Hitachi Ltd. Intends to exert efforts for the technical development to establish nuclear fuel cycle which increases the importance hereafter. (K.I.)

  10. Nuclear fuel element

    International Nuclear Information System (INIS)

    Iwano, Yoshihiko.

    1993-01-01

    Microfine cracks having a depth of less than 10% of a pipe thickness are disposed radially from a central axis each at an interval of less than 100 micron over the entire inner circumferential surface of a zirconium alloy fuel cladding tube. For manufacturing such a nuclear fuel element, the inside of the cladding tube is at first filled with an electrolyte solution of potassium chloride. Then, electrolysis is conducted using the cladding tube as an anode and the electrolyte solution as a cathode, and the inner surface of the cladding tube with a zirconium dioxide layer having a predetermined thickness. Subsequently, the cladding tube is laid on a smooth steel plate and lightly compressed by other smooth steel plate to form microfine cracks in the zirconium dioxide layer on the inner surface of the cladding tube. Such a compressing operation is continuously applied to the cladding tube while rotating the cladding tube. This can inhibit progress of cracks on the inner surface of the cladding tube, thereby enabling to prevent failure of the cladding tube even if a pellet/cladding tube mechanical interaction is applied. Accordingly, reliability of the nuclear fuel elements is improved. (I.N.)

  11. Fuel pellet relocation behavior in fast reactor uranium-plutonium mixed oxide fuel pin at beginning-of-life

    International Nuclear Information System (INIS)

    Inoue, Masaki; Ukai, Shigeharu; Asaga, Takeo

    1999-08-01

    The effects of fabrication parameters, irradiation conditions and fuel microstructural feature on fuel pellet relocation behavior in fast reactor fuel pins were investigated. This work focused only on beginning-of-life conditions, when fuel centerline temperature depends largely on the behavior. Fuel pellet relocation behavior in Joyo Mk-II driver could not be characterized because of the lack of data. And the behavior in FFTF driver and its larger diameter type fuel pins could not be characterized because of the extensive lot-by-lot scatters. The behavior both in Monju type and in Joyo power-to-melt type fuel pins were similar to each other, and depends largely on the as-fabricated gap width while the effects of linear heat rate and the extent of microstructural evolution were negligible. And fuel pellet centerline melting seems to affect slightly the behavior. The correlation, which describes the extent of relocation both in Monju type and in Joyo power-to-melt type fuel pins, were newly formulated and extrapolated for Joyo Mk-II driver, FFTF driver and its larger diameter type fuel pins. And the behavior in Joyo Mk-II driver seemed to be similar. On the contrary, the similarity with JNC fuel pins was observed case-by-case in FFTF driver and its larger diameter type fuel pins. (author)

  12. Long island to Limerick, nuclear fuel transfer

    International Nuclear Information System (INIS)

    Jones, Bill

    1999-01-01

    The issue described is: how to move 33 shipments of radioactive nuclear fuel - 200 tons of enriched uranium pellets - on rail cars through the heart of Philadelphia, without upsetting politicians, the media and anti-nuclear activists, after a similar plan to move the fuel through New York City had been rejected in a political disaster. The answer to this is: Strategic Communications Planning. At PECO Energy's department of Corporate and Public Affairs, the research is quite clear that in risk management situations like this, the side that gets out front with the most credible information inevitably wins. That is exactly what was set out to do

  13. Quality assurance monitoring during nuclear fuel production in JSC 'TVEL'

    International Nuclear Information System (INIS)

    Filimonov, G.; Tchirkov, V.

    2000-01-01

    The paper describes Quality Assurance (QA) monitoring during fabrication of nuclear fuel in Russian Federation. Joint Stock Company 'TVEL', natural state monopoly of the type of holding that fabricates and supplies nuclear fuel for the NPPs of Russia, CIS and Europe, incorporates the major enterprises of the nuclear fuel cycle including JSC 'Mashinostroitelny zavod', Electrostal (fabrication of fuel pellets, rods and assemblies for different types of reactors), JSC 'Novosibirsky zavod khimconcentratov', Novosibirsk (fabrication of fuel rods and assemblies for WWER-440 and WWER-1000), JSC 'Tchepetsky mechanitchesky zavod', Tchepetsk (fabrication of Zr tubing). Monitoring of QA is an important element of Quality Management System (QMS) developed and implemented at the above-mentioned enterprises of the JSC 'TVEL' and it is performed on three levels including external and internal audits and author's supervision. Paper also describes short- and long-term policies of the JSC 'TVEL' in nuclear fuel quality field. (author)

  14. Nuclear fuel element and container

    International Nuclear Information System (INIS)

    Grubb, W.T.; King, L.H.

    1981-01-01

    The invention is based on the discovery that a substantial reduction in metal embrittlement or stress corrosion cracking from fuel pellet-cladding interaction can be achieved by the use of a copper layer or liner in proximity to the nuclear fuel, and an intermediate zirconium oxide barrier layer between the copper layer and the zirconium cladding substrate. The intermediate zirconia layer is a good copper diffusion barrier; also, if the zirconium cladding surface is modified prior to oxidation, copper can be deposited by electroless plating. A nuclear fuel element is described which comprises a central core of fuel material and an elongated container using the system outlined above. The method for making the container is again described. It comprises roughening or etching the surface of the zirconium or zirconium alloy container, oxidizing the resulting container, activating the oxidized surface to allow for the metallic coating of such surfaces by electroless deposition and further coating the activated-oxidized surface of the zirconium or zirconium alloy container with copper, iron or nickel or an alloy thereof. (U.K.)

  15. Combustion and emissions characterization of pelletized coal fuels. Technical report, December 1, 1992--February 28, 1993

    Energy Technology Data Exchange (ETDEWEB)

    Rajan, S. [Southern Illinois Univ., Carbondale, IL (United States). Dept. of Mechanical Engineering and Energy Processes

    1993-05-01

    The aim of this project is to demonstrate that sorbent-containing coal pellets made from low grade coal or coal wastes are viable clean burning fuels, and to compare their performance with that of standard run-of-mine coal. Fuels to be investigated are: (a) carbonated pellets containing calcium hydroxide sorbent, (b) coal fines-limestone pellets with cornstarch as binder, (c) pellets made from preparation plant recovered coal containing limestone sorbent and gasification tar as binder, and (d) a standard run-of-mine Illinois seam coal. The fuels will be tested in a laboratory scale 411 diameter circulating fluidized bed combustor. Progress this quarter has centered on the development of a hydraulic press based pellet mill capable of the high compaction pressures necessary to produce the gasification tar containing pellets outlined in (c) above. Limited quantities of the pellets have been made, and the process is being fine tuned before proceeding into the production mode. Tests show that the moisture content of the coal is an important parameter that needs to be fixed within narrow limits for a given coal and binder combination to produce acceptable pellets. Combustion tests with these pellet fuels and the standard coal are scheduled for the next quarter.

  16. Analysis of neutron flux depression across the pellet radius in CANDU fuel elements

    International Nuclear Information System (INIS)

    Sim, K.S.; Suk, H.C.

    1998-08-01

    The TUBRNP model, originally developed to perform the analysis of the flux depression across the pellet radius in LWR fuel elements, was improved for the application to CANDU fuel elements. The improved model was verified through comparison with existing CANDU model named FLUXDEP in prediction for various fuel conditions. A sensitivity study was also performed to investigate the effects on the flux depression of fuel initial enrichment and burnup, the contents of isotopes U-234 and U-236 and pellet diameter. (author). 9 refs., 8 figs

  17. Fracture toughness of WWER Uranium dioxide fuel pellets with various grain size

    International Nuclear Information System (INIS)

    Sivov, R.; Novikov, V.; Mikheev, E.; Fedotov, A.

    2015-01-01

    Uranium dioxide fuel pellets with grain sizes 13, 26, and 33 μm for WWER were investigated in the present work in order to determine crack formation and the fracture toughness.The investigation of crack formation in uranium oxide fuel pellets of the WWER-types showed that Young’s modulus and the microhardness of polycrystalline samples increase with increasing grain size, while the fracture toughness decreases. Characteristically, radial Palmqvist cracks form on the surface of uranium dioxide pellets for loads up to 1 kg. Transgranular propagation of cracks over distances several-fold larger than the length of the imprint diagonal is observed in pellets with large grains and small intragrain pores. Intergranular propagation of cracks along grain boundaries with branching occurs in pellets with small grains and low pore concentration on the grain boundaries. Blunting on large pores and at breaks in direction does not permit the cracks to reach a significant length

  18. Pellet injection and plasma behavior simulation code PEPSI

    International Nuclear Information System (INIS)

    Takase, Haruhiko; Tobita, Kenji; Nishio, Satoshi

    2003-08-01

    Fueling is one of the major issues on design of nuclear fusion reactor and the injection of solid hydrogen pellet to the core plasma is a useful method. On the design of a nuclear fusion reactor, it is necessary to determine requirements on the pellet size, the number of pellets, the injection speed and the injection cycle. PEllet injection and Plasma behavior SImulation code PEPSI has been developed to assess these parameters. PEPSI has two special features: 1) Adopting two numerical pellet models, Parks model and Strauss model, 2) Calculating fusion power and other plasma parameters in combination with a time-dependent one-dimensional transport model. This report describes the numerical models, numerical scheme, sequence of calculation, list of subroutines, list of variables and an example of calculation. (author)

  19. Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.

  20. Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Second semiannual report, July-December 1979

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.

    1980-03-01

    This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. In the current report period the nuclear design of the demonstration was begun. The design calls for 132 bundles of barrier fuel to be inserted into the core of Quad Cities Unit 2 at the beginning of Cycle 6. Laboratory and in-reactor tests were started to evaluate the stability of Zr-liner fuel which remains in service after a defect has occurred which allows water to enter the rod. Results to date on intentionally defected fuel indicate that the Zr-liner fuel is not rapidly degraded despite ingress of water

  1. quality assurance calculation in UO2 pellet manufacturing process

    International Nuclear Information System (INIS)

    Can, S.; Acarkan, S.; Guereli, L. and others

    1997-01-01

    A process qualification plan is prepared for preparation of quality assurance documentation in accordance with ISO-9000 series of standards, for sintered UO 2 pellets manufactured in the Nuclear Fuel Technology Department. The objectives of this plan are to determine quantitatively and statistically process capability of the pellet production, to check product properties (are) in conformance with specifications at the pre-( ) confidence levels, to prepare necessary documents and to assess the results. The product properties taking into account are chemical composition, cracks, density, microstructure and grain size. The statistical parameters used for qualification element of quality assurance are calculated.Statistical values for sintered pellets are: LENGTH/WEIGHT/DIAMETER/DENSITY/%TD: MEAN:13,395/16,808/12,293/10,679/97,400 STD:0,1651/ 0,252/0,0212/0,015/0,140. It was seen that sintered pellets manufactured in the Nuclear Fuel Technology Department meet the criteria within 95% confidence level. In this paper specifications, criteria and calculations will be explained in detail

  2. Modelling the bending/bowing of composite beams such as nuclear fuel

    International Nuclear Information System (INIS)

    Tayal, M.

    1989-01-01

    Arrays of tubes are used in many engineered structures, such as in nuclear fuel bundles and in steam generators. The tubes can bend (bow) due to in-service temperatures and loads. Assessments of bowing of nuclear fuel elements can help demonstrate the integrity of fuel and of surrounding components, as a function of operating conditions such as channel power. The BOW code calculates the bending of composite beams such as fuel elements, due to gradients of temperature and due to hydraulic forces. The deflections and rotations are calculated in both lateral directions, for given conditions of temperatures. Wet and dry operation of the sheath can be simulated. BOW accounts for the following physical phenomena: circumferential and axial variations in the temperatures of the sheath and of the pellet; cracking of pellets; grip and slip between the pellets and the sheath; hydraulic drag; restraint from endplates, from neighbouring elements, and from the pressure-tube; gravity; concentric or eccentric welds between endcaps and endplate; neutron flux gradients; and variations of material properties with temperature. The code is based on fundamental principles of mechanics. The governing equations are solved numerically using the finite element method. Several comparisons with closed-form equations shoe that the solutions of BOW are accurate. BOW's predictions for initial in-reactor bow are also consistent with two post-irradiation measurements

  3. Cracked pellet gap conductance model: comparison of FRAP-S calculations with measured fuel centerline temperatures

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Broughton, J.M.

    1975-03-01

    Fuel pellets crack extensively upon irradiation due both to thermal stresses induced by power changes and at high burnup, to accumulation of gaseous fission products at grain boundaries. Therefore, the distance between the fuel and cladding will be circumferentially nonuniform; varying between that calculated for intact operating fuel pellets and essentially zero (fuel segments in contact with the cladding wall). A model for calculation of temperatures in cracked pellets is proposed wherein the effective fuel to cladding gap conductance is calculated by taking a zero pressure contact conductance in series with an annular gap conductance. Comparisons of predicted and measured fuel centerline temperatures at beginning of life and at extended burnup are presented in support of the model. 13 references

  4. Nuclear fuel elements and assemblies

    International Nuclear Information System (INIS)

    Saito, Shozo; Maki, Hideo.

    1982-01-01

    Purpose: To facilitate the attainment of the uranium enrichment or gadolinia enrichment of a pellet filled in a fuel element. Constitution: The axial length of a pellet filled in a fuel element is set to predetermined sizes according to the uranium enrichment factor, gadolinia enrichment or their combination. Thus, the uranium enrichment factor or gadolinia enrichment can be identified by attaining the axial length of the pellet by using such a pellt. (Kamimura, M.)

  5. Nuclear fuel conversion and fabrication chemistry

    International Nuclear Information System (INIS)

    Lerch, R.E.; Norman, R.E.

    1984-01-01

    Following irradiation and reprocessing of nuclear fuel, two operations are performed to prepare the fuel for subsequent reuse as fuel: fuel conversion, and fuel fabrication. These operations complete the classical nuclear fuel cycle. Fuel conversion involves generating a solid form suitable for fabrication into nuclear fuel. For plutonium based fuels, either a pure PuO 2 material or a mixed PuO 2 -UO 2 fuel material is generated. Several methods are available for preparation of the pure PuO 2 including: oxalate or peroxide precipitation; or direct denitration. Once the pure PuO 2 is formed, it is fabricated into fuel by mechanically blending it with ceramic grade UO 2 . The UO 2 can be prepared by several methods which include direct denitration. ADU precipitation, AUC precipitation, and peroxide precipitation. Alternatively, UO 2 -PuO 2 can be generated directly using coprecipitation, direct co-denitration, or gel sphere processes. In coprecipitation, uranium and plutonium are either precipitated as ammonium diuranate and plutonium hydroxide or as a mixture of ammonium uranyl-plutonyl carbonate, filtered and dried. In direct thermal denitration, solutions of uranium and plutonium nitrates are heated causing concentration and, subsequently, direct denitration. In gel sphere conversion, solutions of uranium and plutonium nitrate containing additives are formed into spherical droplets, gelled, washed and dried. Refabrication of these UO 3 -PuO 2 starting materials is accomplished by calcination-reduction to UO 2 -PuO 2 followed by pellet fabrication. (orig.)

  6. A simulation of the temperature overshoot observed at high burnup in annular fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Baron, D [Electricite de France, Moret-sur-Loing (France); Couty, J C [Electricite de France (EDF), 69 - Villeurbanne (France)

    1997-08-01

    Instrumented experiments have been carried out in recent years to calibrate and improve temperature calculations at high burnup in PWR nuclear fuel rods. The introduction of a thermocouple in the fuel stack allows the experiment to record the centre-line temperature all along the irradiation or re-irradiation. The results obtained on fresh fuel have not revealed any abnormal behavior as have observations done on high burnup rods. In this case, a sudden overshoot has been recorded on the thermocouple temperature above an average power threshold. Several hypotheses have been suggested. Only two seem to be acceptable: one in relation to an effect of grain decohesion, another based on a modification of fuel chemistry. The apparent reversibility of the phenomena when power decreases led us to prefer the first explanation. Indeed, the introduction of a thermocouple means that annular fuel pellets must be used. These are either initially manufactured with a central hole or drilled after base irradiation, using the ``RISOE`` technique. One must bear in mind that the use of such annular pellets drastically changes the crack pattern as irradiation proceeds. This is due to a different stress field which, combined with a weakening of the grain binding energy, leads to a partial grain decohesion on the inner face of the annular pellet. Modification of the grain binding energy is related to the presence of an increasing local population of gas bubbles and metallic precipitates at grain boundaries, as swelling creates intergranular local stresses which also could probably enhance the grain decohesion process. This grain decohesion concerns a 250 to 350 {mu}m depth and shows a narrow cracks network through which released fission gas can flow, temporarily pushing the resident helium gas out. The low conductivity of these gaseous fission products and the numerous gas layers created this way could partly explain the unexpected temperatures measured in high burnup fuels. (Abstract

  7. A simulation of the temperature overshoot observed at high burnup in annular fuel pellets

    International Nuclear Information System (INIS)

    Baron, D.; Couty, J.C.

    1997-01-01

    Instrumented experiments have been carried out in recent years to calibrate and improve temperature calculations at high burnup in PWR nuclear fuel rods. The introduction of a thermocouple in the fuel stack allows the experiment to record the centre-line temperature all along the irradiation or re-irradiation. The results obtained on fresh fuel have not revealed any abnormal behavior as have observations done on high burnup rods. In this case, a sudden overshoot has been recorded on the thermocouple temperature above an average power threshold. Several hypotheses have been suggested. Only two seem to be acceptable: one in relation to an effect of grain decohesion, another based on a modification of fuel chemistry. The apparent reversibility of the phenomena when power decreases led us to prefer the first explanation. Indeed, the introduction of a thermocouple means that annular fuel pellets must be used. These are either initially manufactured with a central hole or drilled after base irradiation, using the ''RISOE'' technique. One must bear in mind that the use of such annular pellets drastically changes the crack pattern as irradiation proceeds. This is due to a different stress field which, combined with a weakening of the grain binding energy, leads to a partial grain decohesion on the inner face of the annular pellet. Modification of the grain binding energy is related to the presence of an increasing local population of gas bubbles and metallic precipitates at grain boundaries, as swelling creates intergranular local stresses which also could probably enhance the grain decohesion process. This grain decohesion concerns a 250 to 350 μm depth and shows a narrow cracks network through which released fission gas can flow, temporarily pushing the resident helium gas out. The low conductivity of these gaseous fission products and the numerous gas layers created this way could partly explain the unexpected temperatures measured in high burnup fuels. The purpose of

  8. Procedure for the fabrication of ceramic fuel pellets with an adjustable structure

    International Nuclear Information System (INIS)

    Henke, M.; Klemm, U.; Sobek, D.

    1986-01-01

    The invention concerns a procedure for the fabrication of ceramic fuel pellets of UO 2 , PuO 2 , ThO 2 and their mixtures with an adjustable structure. Before or during the milling the particle shaped fuel pellets have been added polyethylenglycol in a 20 - 60 % aqueous solution with an amount of 0.5 - 2.0 % in weight. This additive has an effect on a controlled pore formation and grain growth advancement

  9. Technical verification of advanced nuclear fuel for KSNPs

    International Nuclear Information System (INIS)

    Lee, C. B.; Bang, J. G.; Kim, D. H. and others

    2002-03-01

    KNFC has developed the advanced 16x16 fuel assembly for the Korean Standard Nuclear Plants through the three-year R and D project (from April 1999 to March 2002) under the Nuclear R and D program by MOST. The purpose of this project is to verify the advanced 16x16 fuel assembly for the Korean Standard Nuclear Plants being developed by KNFC during the same period. Verification tests for the advanced fuel assembly and its components such as characteristic test on the spacer grid spring and dimple, static buckling and dynamic impact test on the 5x5 partial spacer grid, the fuel rod vibration test supported by the PLUS7 mid-spacer grid, fretting wear test, turbulent flow structure test in wind tunnel and corrosion test were performed by using the KAERI facilities. Design reports and test results produced by KNFC were technically reviewed. For the domestic production of burnable poison rod, manufacturing technology of burnable poison pellets was developed

  10. Characterization methods of nuclear fuel materials

    International Nuclear Information System (INIS)

    Bustillos, O.W.R.; Teixeira, S.R.; Lordello, A.R.; Imakuma, K.; Rodrigues, C.

    1980-01-01

    In order to organize quality control and quality assurance activities for PWR fuel production, the laboratory of IPEN (Instituto de Pesquisas Energeticas e Nucleares) has developed various chemical and physical methods to be employed for fuel characterization of UO 2 pellets. The techniques developed are the determination of total residual gases by vacuum fusion method, determination of impurity elements by optical spectrography and characterization methods by X-ray diffraction. The development and the implementation of these techniques under the implementation of these techniques under the general scheme of characterization and quality control is the major theme discussed. (E.G.) [pt

  11. Improvement of fuel-element reliability by insertion of UO2 microspheres in the gap between pellet and clad

    International Nuclear Information System (INIS)

    Mehedinteanu, S.; Glodeanu, F.; Dobos, I.

    1979-01-01

    With the accumulation of power reactor fuel operating experience, the study of the PCI phenomenon and the development of remedies have become important items in fuel research and development everywhere. The 'power-ramp' failure has drawn attention to the problem of obtaining high reliability from high burn-up fuel rods. Considerable attention has been paid to minimizing the cladding stresses imparted by fuel pellets during the power ramp. The paper describes a new concept of pellet-clad bonding by insertion of UO 2 microspheres in the gap. It is pointed out that the main advantages of this concept are: the low friction coefficient between pellet and clad; the accomodation of cracked pellet expansion by local microyielding of irradiation-embrittled clad; the reduced ridge height by use of undished pellets or other pellet shape; that the fine-sized UO 2 microspheres infiltrate around the pellets thus permitting the use of cracked or chipped pellets and also sintered pellets without the previously required grinding step needed for accurate sizing, etc. (author)

  12. Nuclear fuel cycle, nuclear fuel makes the rounds: choosing a closed fuel cycle, nuclear fuel cycle processes, front-end of the fuel cycle: from crude ore to enriched uranium, back-end of the fuel cycle: the second life of nuclear fuel, and tomorrow: multiple recycling while generating increasingly less waste

    International Nuclear Information System (INIS)

    Philippon, Patrick

    2016-01-01

    France has opted for a policy of processing and recycling spent fuel. This option has already been deployed commercially since the 1990's, but will reach its full potential with the fourth generation. The CEA developed the processes in use today, and is pursuing research to improve, extend, and adapt these technologies to tomorrow's challenges. France has opted for a 'closed cycle' to recycle the reusable materials in spent fuel (uranium and plutonium) and optimise ultimate waste management. France has opted for a 'closed' nuclear fuel cycle. Spent fuel is processed to recover the reusable materials: uranium and plutonium. The remaining components (fission products and minor actinides) are the ultimate waste. This info-graphic shows the main steps in the fuel cycle currently implemented commercially in France. From the mine to the reactor, a vast industrial system ensures the conversion of uranium contained in the ore to obtain uranium oxide (UOX) fuel pellets. Selective extraction, purification, enrichment - key scientific and technical challenges for the teams in the Nuclear Energy Division (DEN). The back-end stages of the fuel cycle for recycling the reusable materials in spent fuel and conditioning the final waste-forms have reached maturity. CEA teams are pursuing their research in support of industry to optimise these processes. Multi-recycle plutonium, make even better use of uranium resources and, over the longer term, explore the possibility of transmuting the most highly radioactive waste: these are the challenges facing future nuclear systems. (authors)

  13. Nuclear fuel element

    International Nuclear Information System (INIS)

    Watarumi, Kazutoshi.

    1992-01-01

    Hollow fuel pellets are piled at multi-stages in a cladding tube to form a pellet stack. A bundle of metal fine wires made of zirconium or an alloy thereof is inserted passing through the hollow portion of each of the hollow pellets over a length of the pellet stack. The metal fine wires are bundled by securing ring at a joining portions of the pellets. Then, the portion between both of adjacent rings is expanded radially and has a spring function biasing in the radial direction. With such a constitution, even if the pellet is expanded radially due to pallet gas swelling, the hollow portion is not closed, and the gas flow channel is ensured. In addition, even if the pellet is cracked due to thermal shocks, the pellet piece is prevented from dropping to the hollow portion. In this case, the thermal conduction between the pellets and the cladding tube is kept satisfactorily by the spring function of the metal wire bundle. (I.N.)

  14. Pelletizing of rice straws: A potential solid fuel from agricultural residues

    International Nuclear Information System (INIS)

    Puad, E.; Wan Asma, I; Shaharuddin, H.; Mahanim, S.; Rafidah, J.

    2010-01-01

    Full text: Rice straw is the dry stalks of rice plants, after the grain and chaff have been removed. More than 1 million tonnes of rice straw are produced in MADA in the northern region of Peninsular Malaysia annually. Burning in the open air is the common technique of disposal that contribute to air pollution. In this paper, a technique to convert these residues into solid fuel through pelletizing is presented. The pellets are manufactured from rice straw and sawdust in a disc pelletizer. The pellet properties are quite good with good resistance to mechanical disintegration. The pellets have densities between 1000 and 1200 kg/ m 3 . Overall, converting rice straw into pellets has increased its energy and reduced moisture content to a minimum of 8 % and 30 % respectively. The gross calorific value is about 15.6 MJ/ kg which is lower to sawdust pellet. The garnering of knowledge in the pelletization process provides a path to increase the use of this resource. Rice straw pellets can become an important renewable energy source in the future. (author)

  15. Argentina-LLNL-LANL Comparative Sample Analysis on UO2 fuel pellet CRM-125A for Nuclear Forensics

    Energy Technology Data Exchange (ETDEWEB)

    Kips, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-12-01

    The recent workshop on analytical plan development provided context and background for the next step in this engagement, i.e. a comparative sample analysis on CRM 125-A. This is a commercially available certified low-enriched uranium oxide fuel pellet material from New Brunswick National Laboratory (NBL) (see certificate in Annex 1).

  16. Fabrication of 0.5-inch diameter FBR mixed oxide fuel pellets

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Benecke, M.W.; McCord, R.B.

    1979-01-01

    Large diameter (0.535 inch) mixed oxide fuel pellets for Fast Breeder Reactor application were successfully fabricated by the cold-press-and-sinter technique. Enriched UO 2 , PuO 2 -UO 2 , and PuO 2 -ThO 2 compositions were fabricated into nominally 90% theoretical density pellets for the UO 2 and PuO 2 -UO 2 compositions, and 88% and 93% T.D. for the PuO 2 -ThO 2 compositions. Some processing adjustments were required to achieve satisfactory pellet quality and density. Furnace heating rate was reduced from 200 to 50 0 C/h for the organic binder burnout cycle for the large, 0.535-inch diameter pellets to eliminate pellet cracking during sintering. Additional preslugging steps and die wall lubrication during pressing were used to eliminate pressing cracks in the PuO 2 -ThO 2 pellets

  17. FRACAS: a subcode for the analysis of fuel pellet-cladding mechanical interaction

    International Nuclear Information System (INIS)

    Bohn, M.P.

    1977-04-01

    This report describes FRACAS (Fuel Rod and Cladding Analysis Subcode), a computer code which performs the mechanical analysis in the FRAP fuel rod codes. At each loadstep, FRACAS obtains a complete elastic-plastic-creep solution for the stresses, strains, and displacements in the fuel rod cladding. The cladding is modeled as a thin cylindrical shell with prescribed temperature, pressures, and radial displacement of the inside surface. The displacement of the fuel pellets is assumed to be due to thermal gradients only. Three different regimes of pellet-cladding mechanical interaction are considered: (a) open gap, (b) closed gap, and (c) trapped stack. Both transient and steady state creep calculations are performed. The capabilities of the code are illustrated by an example problem, and comparisons are made with data obtained from two experimental fuel rods

  18. Microgasification cookstoves and pellet fuels from waste biomass: A ...

    African Journals Online (AJOL)

    Microgasification cookstoves and pellet fuels from waste biomass: A cost and performance comparison with charcoal and natural gas in Tanzania. ... produce too much smoke and 40% stating that controlling the air vent is too much trouble.

  19. Cracking and relocation of UO2 fuel during nuclear operation

    International Nuclear Information System (INIS)

    Appelhans, A.D.; Dagbjartsson, S.J.

    1981-01-01

    Cracking and relocation of light water reactor (LWR) fuel pellets affect the axial gas flow path within nuclear reactor fuel rods and the thermal performance of the fuel. As part of the Nuclear Regulatory Commission's Water Reactor Safety Research Fuel Behavior Program, the Thermal Fuels Behavior Program of EG and G Idaho, Inc., is conducting fuel rod behavior studies in the Heavy Boiling Water Reactor in Halden, Norway. The Instrumental Fuel Assembly-430 (IFA-430) operated in that facility is a multipurpose assembly designed to provide information on fuel cracking and relocation, the long-term thermal response of LWR fuel rods subjected to various internal pressures and gas compositions, and the release of fission gases. This report presents the results of an analysis of fuel cracking and relocation phenomena as deduced from fuel rod axial gas flow and fuel temperature data from the first 6.5 GWd/tUO 2 burnup of the IFA-430

  20. Pellets used for nuclear reactor scram and a method for manufacturing them

    International Nuclear Information System (INIS)

    1974-01-01

    The invention deals with a pellet to be inserted in the core of a nuclear reactor for stopping the operation of the latter. The pellet is characterized in that it is in the form of a pellet capable of rolling easily, containing a neutron poison and a solid substance undergoing a change of state when it is raised to a predetermined temperature reached by the reactor-core, that change of state causing the pellet to desintegrate and inducing the deposition of the poison. This can be applied to the shut down of gas-cooled nuclear reactors [fr

  1. Ceramics as nuclear reactor fuels

    International Nuclear Information System (INIS)

    Reeve, K.D.

    1975-01-01

    Ceramics are widely accepted as nuclear reactor fuel materials, for both metal clad ceramic and all-ceramic fuel designs. Metal clad UO 2 is used commercially in large tonnages in five different power reactor designs. UO 2 pellets are made by familiar ceramic techniques but in a reactor they undergo complex thermal and chemical changes which must be thoroughly understood. Metal clad uranium-plutonium dioxide is used in present day fast breeder reactors, but may eventually be replaced by uranium-plutonium carbide or nitride. All-ceramic fuels, which are necessary for reactors operating above about 750 0 C, must incorporate one or more fission product retentive ceramic coatings. BeO-coated BeO matrix dispersion fuels and silicate glaze coated UO 2 -SiO 2 have been studied for specialised applications, but the only commercial high temperature fuel is based on graphite in which small fuel particles, each coated with vapour deposited carbon and silicon carbide, are dispersed. Ceramists have much to contribute to many aspects of fuel science and technology. (author)

  2. standards used for quality control of nuclear fuels

    International Nuclear Information System (INIS)

    Guereli, L; Can, S.

    1997-01-01

    Nuclear fuels and fuel materials are subject to stringent restrictions as to their quality. The standards and regulations that apply vary according to reactor type and country and the standards are stated in the quality assurance documents. The concept of quality assurance has altered the conventional quality control tests and procedures, defining which control tests are to be applied and how. Although most of the tests and measurements allow the determination of tolerances to be decided according to the agreement between the buyer and the seller, exacting procedures apply to which instruments and equipment are used for these tests and measurements, how these instruments are standardized.Detailed explanations of test methods and their documentation is a requirement in all standards. The purpose of this work is to study which standards, tests and measurements apply to the nuclear fuel production. Only the standards that apply to various stages of the nuclear fuel production (powder preparation, pellet production, fuel element and fuel assembly fabrication) are reviewed. Process and documentation control, design and licensing requirements and the frequency of inspections are quality assurance subjects. Some ASTM standards are given as examples

  3. Eight-shot pellet injector and fueling experiments at the HL-1M tokamak

    International Nuclear Information System (INIS)

    Xiao Zhenggui; Li Bo; Li Li

    2001-01-01

    An Eight-shot Pellet Injection (EPI) system has been proposed and developed in collaboration between STU (St. Petersburg State Technical University) of Russia and SWIP. In the EPI, the I n-situ c ondensation technique was used to produce the pellets in eight gun barrels respectively. The nominal pellet size (diameter of 1.0 mm and of 1.4 mm or 1.2 mm) is limited by the gun barrel inner diameter. The pellet length is adjusted by changing the g radient temperature o n the gun barrels and the amounts of filling fuel gas. Pellets are fired at speed range of 200 - 1200 m/s by He propellant with pressure of 2 - 6 MPa and then transferred to HL-1M vessel through an injection line that consists of two set of differential vacuum pumped chambers and guide tube combined with fast valves. In addition, this unit is equipped with diagnostics for pellet velocity and shape measure. The EPI has installed on HL-1M since 1996 for the multi-shot pellet fueling experiments. The typical characteristics including the peaked density profile and improved confinement, the deep penetration and suppression of soft X-ray sawteeth, the variance of rotation and flow of plasma in edge region as well as the photographing of pellet ablation clouds are presented

  4. Demonstration of fuel resistant to pellet-cladding interaction. First semiannual report, July-December 1977

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbaum, H.S. (comp.)

    1978-02-01

    Objective is the demonstration od advanced fuel concepts that are resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Since currently used fuel in the nuclear power industry is subject to the PCI failure mechanism, reactor operators limit the rates of power increases and thus reduce their capacity factors in order to protect the fuel. Two barrier concepts are being prepared for demonstration: (a) Cu-Barrier fuel and (b) Zr-Liner fuel. The large-scale demonstration of the PCI-resistant fuel is being designed generically to show feasibility of such a demonstration in a commercial power reactor of type BWR/3 having a steady-state core. Using the core of Quad Cities-1 reactor at the beginning of Cycle 6, the insertion of the demonstration PCI-resistant fuel and the reactor operational plan are being designed. Support laboratory tests to date for the Demonstration have shown that these barrier fuels (both the Cu-Barrier and the Zr-Liner types) are resistant to PCI. Four lead test assemblies (LTA) of the advanced PCI-resistant fuel are being fabricated for insertion into the Quad Cities-1 Boiling Water Reactor at the beginning of Cycle 5 (January 1979).

  5. Selective alpha autoradiography for monitoring thorium distribution in UO2-ThO2 fuel pellets

    International Nuclear Information System (INIS)

    Shriwastwa, B.B.; Raghunath, B.; Ghosh, J.K.

    1992-01-01

    Although natural uranium and thorium decay with similar alpha energies (4.20 and 3.98 MeV), their daughter products have different alpha characteristics. This has been exploited for selective alpha autoradiography for thoria in urania-thoria mixed nuclear fuel pellets. Difficulties in getting sufficient track density in alpha sensitive films due to the very low specific activity of natural uranium and thorium material were overcome by using a special film with annealing and pre-etching treatment. (orig./HP) [de

  6. Mixed U/Pu oxide fuel fabrication facility co-processed feed, pelletized fuel

    International Nuclear Information System (INIS)

    1978-09-01

    Two conceptual MOX fuel fabrication facilities are discussed in this study. The first facility in the main body of the report is for the fabrication of LWR uranium dioxide - plutonium dioxide (MOX) fuel using co-processed feed. The second facility in the addendum is for the fabrication of co-processed MOX fuel spiked with 60 Co. Both facilities produce pellet fuel. The spiked facility uses the same basic fabrication process as the conventional MOX plant but the fuel feed incorporates a high energy gamma emitter as a safeguard measure against diversion; additional shielding is added to protect personnel from radiation exposure, all operations are automated and remote, and normal maintenance is performed remotely. The report describes the fuel fabrication process and plant layout including scrap and waste processing; and maintenance, ventilation and safety measures

  7. Determination of Gd concentration profile in UO{sub 2}–Gd{sub 2}O{sub 3} fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Tobia, D., E-mail: dina.tobia@cab.cnea.gov.ar [Laboratorio de Resonancias Magnéticas, Centro Atómico Bariloche – CNEA and CONICET, 8400 S.C. de Bariloche (Argentina); Winkler, E.L.; Milano, J.; Butera, A. [Laboratorio de Resonancias Magnéticas, Centro Atómico Bariloche – CNEA and CONICET, 8400 S.C. de Bariloche (Argentina); Kempf, R. [División Caracterización de Combustibles Avanzados, Gerencia Ciclo Combustible Nuclear, Centro Atómico Constituyentes – CNEA, 1650 San Martín, Pcia. de Buenos Aires (Argentina); Bianchi, L.; Kaufmann, F. [Departamento de Combustibles Avanzados, Gerencia Ciclo Combustible Nuclear, Centro Atómico Constituyentes – CNEA, 1650 San Martín, Pcia. de Buenos Aires (Argentina)

    2014-08-01

    A transversal mapping of the Gd concentration was measured in UO{sub 2}–Gd{sub 2}O{sub 3} nuclear fuel pellets by electron paramagnetic resonance spectroscopy (EPR). The quantification was made from the comparison with a Gd{sub 2}O{sub 3} reference sample. The nominal concentration in the pellets is UO{sub 2}: 7.5% Gd{sub 2}O{sub 3}. A concentration gradient was found, which indicates that the Gd{sub 2}O{sub 3} amount diminishes towards the edges of the pellets. The concentration varies from (9.3 ± 0.5)% in the center to (5.8 ± 0.3)% in one of the edges. The method was found to be particularly suitable for the precise mapping of the distribution of Gd{sup 3+} ions in the UO{sub 2} matrix.

  8. Energy wood. Part 2b: Wood pellets and pellet space-heating systems

    International Nuclear Information System (INIS)

    Nussbaumer, T.

    2002-01-01

    The paper gives an overview on pellet utilization including all relevant process steps: Potential and properties of saw dust as raw material, pellet production with drying and pelletizing, standardization of wood pellets, storage and handling of pellets, combustion of wood pellets in stoves and boilers and applications for residential heating. In comparison to other wood fuels, wood pellets show several advantages: Low water content and high heating value, high energy density, and homogeneous properties thus enabling stationary combustion conditions. However, quality control is needed to ensure constant properties of the pellets and to avoid the utilization of contaminated raw materials for the pellet production. Typical data of efficiencies and emissions of pellet stoves and boilers are given and a life cycle analysis (LCA) of wood pellets in comparison to log wood and wood chips is described. The LCA shows that wood pellets are advantageous thanks to relatively low emissions. Hence, the utilization of wood pellet is proposed as a complementary technology to the combustion of wood chips and log wood. Finally, typical fuel cost of wood pellets in Switzerland are given and compared with light fuel oil. (author)

  9. Investigation into rationalization of low decontamination pellet fuel fabrication plant configuration

    International Nuclear Information System (INIS)

    Maekawa, Kazuhiko; Yoshimura, Tadahiro; Hoshino, Yasushi; Munekata, Hideki; Tamaki, Yoshihisa

    2005-02-01

    In feasibility studies on commercialized FBR cycle system, a comprehensive system investigation and properties evaluation for candidate FBR cycle systems has been implemented through view point of safety, economics, environmental burden reduction, non-proliferation resistivity, etc. As part of these studies, an investigation into rationalization of low decontamination pellet fuel fabrication plant configuration was carried out. Until last fiscal year, conceptual design studies of the fuel fabrication plant in 200t-HM/y scale were conducted, and system properties data concerning economics and environmental burden reduction of fuel fabrication plant was acquired. In addition to this, 50t-HM/y scale plant was also schematically studied. In this fiscal year, a rationalization study on conceptual design of 50t-HM/y scale plant was conducted with main aim of economic improvement, and the 200t-HM/y scale plant design was revised based on the recent R and D progress. The system properties data concerning economics and environmental burden reduction of fuel fabrication plant was also acquired. In both case of the 50t-HM/y and 200t-HM/y scale plant, it was suggested that the equipment costs were reduced in several percentages because of reduction of maintenance equipments and cut in line number at the pellet fabrication process although granulation process fro denitration converted powder and O/M control process for pellets were added. System properties data for comparative evaluation of candidate fuel fabrication systems was also prepared. (author)

  10. Production method of burnable poison incorporated fuel pellet by coating

    International Nuclear Information System (INIS)

    Naito, Naoyoshi.

    1993-01-01

    A cylindrical member is formed with an organic material which is melted, decomposed or evaporated by heating. Such organic materials include polyethylene and polyvinyl alcohol, for example. A predetermined amount of burnable poisons are homogeneously incorporated in the cylindrical member by a means, such as melting before fabricating it into a cylindrical shape. UO 2 fuel pellets are inserted to the cylindrical member and heated, to scatter only the organic materials, so that non-volatile burnable poisons are homogeneously left on the surface of the pellets. It is preferred that the cylindrical member having pellets inserted therein is inserted to a cladding tube and applied with a heat treatment. With such procedures, a UO 2 pellet is coated with burnable poisons by a convenient and compact device. In addition, grinding step after the coating is unnecessary. (I.N.)

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

    International Nuclear Information System (INIS)

    Denis, Alicia C.

    1999-01-01

    A fission gas release model is presented, which solves the atomic diffusion problem with xenon and krypton elements tramps produced by uranium fission during UO 2 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)

  12. Quantification of the effect of in-situ generated uranium metal on the experimentally determined O/U ratio of a sintered uranium dioxide fuel pellet

    International Nuclear Information System (INIS)

    Narasimha Murty, B.; Bharati Misra, U.; Yadav, R.B.; Srivastava, R.K.

    2005-01-01

    This paper describes quantitatively the effect of in-situ generated uranium metal (that could be formed due to the conducive manufacturing conditions) in a sintered uranium dioxide fuel pellet on the experimentally determined O/U ratio using analytical methods involving dissolution of the pellet material. To quantify the effect of in-situ generated uranium metal in the fuel pellet, a mathematical expression is derived for the actual O/U ratio in terms of the O/U ratio as determined by an experiment involving dissolution of the material and the quantity of uranium metal present in the uranium dioxide pellet. The utility of this derived mathematical expression is demonstrated by tabulating the calculated actual O/U ratios for varying amounts of uranium metal (from 5 to 95% in 5% intervals) and different O/U ratio values (from 2.001 to 2.015 in 0.001 intervals). This paper brings out the necessity of care to be exercised while interpreting the experimentally determined O/U ratio and emphasizes the fact that it is always safer to produce the nuclear fuel with oxygen to uranium ratios well below the specified maximum limit of 2.015. (author)

  13. A three-barrel repeating pneumatic pellet injector for plasma fueling of the Joint European Torus

    International Nuclear Information System (INIS)

    Combs, S.K.; Milora, S.L.; Baylor, L.R.; Foust, C.R.; Gethers, F.E.; Sparks, D.O.

    1987-01-01

    Pellet fueling, the injection of frozen hydrogen isotope pellets at high velocity, has been used to improve plasma performance in various tokamak experiments. In one recent experiment, the repeating pneumatic hydrogen pellet injector was used on the Tokamak Fusion Test Reactor (TFTR). This machine gun-like device, which was developed at the Oak Ridge National Laboratory (ORNL) with an objective of steady-state fueling applications, was characterized by a fixed pellet size and a maximum repetition rate of 4 to 6 Hz for several seconds. It was used to deliver deuterium pellets at speeds ranging from 1.0 to 1.5 km/s into TFTR plasma discharges. In the first experiments, injection of single, large (nominal 4-mm-diam) pellets provided high plasma densities in TFTR (1.8 x 10 14 cm -3 on axis). After a conversion to smaller (nominal 2.7-mm-diam) pellets, the pellet injector was operated in the repeating mode to gradually increase the plasma density, injecting up to five pellets on a single machine pulse. This resulted in central plasma densities approaching 4 x 10 14 cm -3 and n tau values of 1.4 x 10 14 cm -3 s. For plasma fueling applications on the Joint European Torus (JET), a pellet injector fashioned after the prototype repeating pneumatic design has been developed. The versatile injector features three repeating guns in a common vacuum enclosure; the guns provide pellets that are 2.7, 4.0, and 6.0 mm in diameter and can operate independently at repetition rates of 5, 2.5, and 1 Hz, respectively. The injector has been installed on JET. A description of the equipment is presented, emphasizing the differences from the original repeating device. Performance characteristics of the three pneumatic guns are also included

  14. Experimental determination of nuclear reaction rates in 238U and 235U along of the radius of fuel pellets of the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Mura, Luis Felipe Liambos

    2015-01-01

    This research presents and consolidates an alternative methodology for determining nuclear reaction rates along the radial direction of the fuel pellets which does not require high neutron flux. This technique is based on irradiating a thin UO 2 disk inserted into a removable fuel rod at the IPEN/MB-01 reactor core. Several gamma spectrometry are performed after irradiation using a HPGe detector. Six lead collimators with different diameters are sequentially alternated during this process, thus, the nuclear radioactive capture which occurs in 238 U and the fissions which occur in both 235 U and 238 U are measured according to six different radial regions of the fuel disk. Geometric efficiency corrections due to the introduction of collimators in HPGe detection system are determined by MCNP-5 code. The fission rate measurements are performed using the 99 Mo. This radionuclide was studied and proved ideal for these measurements because it is formed in linear behavior in the reactor core, have a high yield fission and emits low-energy photons. Measurements were performed irradiating UO 2 disks (with 4.3% enrichment) in the central position of the IPEN/MB-01 core at 100 watts power level during one hour. Some measurements were performed using a cadmium glove wrapped in the fuel rod to determine the nuclear reaction rates in the epithermal energy range. The experimental results obtained are compared with nuclear reaction rate calculations by means of MCNP-5 with ENDF/B-VII.0 data library showing discrepancies of up to 9% in 238 U capture rates and 14% for U fission rates for epithermal energies. Uncertainties regarding the nuclear capture rates have maximum values of 4.5% and the fission rates has maximum values of 11.3%. (author)

  15. Development of Nuclear Fuel Remote Fabrication Technology

    International Nuclear Information System (INIS)

    Lee, Jung Won; Yang, M. S.; Kim, S. S. and others

    2005-04-01

    The aim of this study is to develop the essential technology of dry refabrication using spent fuel materials in a laboratory scale on the basis of proliferation resistance policy. The emphasis is placed on the assessment and the development of the essential technology of dry refabrication using spent fuel materials. In this study, the remote fuel fabrication technology to make a dry refabricated fuel with an enhanced quality was established. And the instrumented fuel pellets and mini-elements were manufactured for the irradiation testing in HANARO. The design and development technology of the remote fabrication equipment and the remote operating and maintenance technology of the equipment in hot cell were also achieved. These achievements will be used in and applied to the future back-end fuel cycle and GEN-IV fuel cycle and be a milestone for Korea to be an advanced nuclear country in the world

  16. Grain Size and Phase Purity Characterization of U3Si2 Pellet Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hoggan, Rita E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tolman, Kevin R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cappia, Fabiola [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wagner, Adrian R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2018-05-01

    Characterization of U3Si2 fresh fuel pellets is important for quality assurance and validation of the finished product. Grain size measurement methods, phase identification methods using scanning electron microscopes equipped with energy dispersive spectroscopy and x-ray diffraction, and phase quantification methods via image analysis have been developed and implemented on U3Si2 pellet samples. A wide variety of samples have been characterized including representative pellets from an initial irradiation experiment, and samples produced using optimized methods to enhance phase purity from an extended fabrication effort. The average grain size for initial pellets was between 16 and 18 µm. The typical average grain size for pellets from the extended fabrication was between 20 and 30 µm with some samples exhibiting irregular grain growth. Pellets from the latter half of extended fabrication had a bimodal grain size distribution consisting of coarsened grains (>80 µm) surrounded by the typical (20-30 µm) grain structure around the surface. Phases identified in initial uranium silicide pellets included: U3Si2 as the main phase composing about 80 vol. %, Si rich phases (USi and U5Si4) composing about 13 vol. %, and UO2 composing about 5 vol. %. Initial batches from the extended U3Si2 pellet fabrication had similar phases and phase quantities. The latter half of the extended fabrication pellet batches did not contain Si rich phases, and had between 1-5% UO2: achieving U3Si2 phase purity between 95 vol. % and 98 vol. % U3Si2. The amount of UO2 in sintered U3Si2 pellets is correlated to the length of time between U3Si2 powder fabrication and pellet formation. These measurements provide information necessary to optimize fabrication efforts and a baseline for future work on this fuel compound.

  17. High-speed repetitive pellet injector for plasma fueling of magnetic confinement fusion devices

    International Nuclear Information System (INIS)

    Combs, S.K.; Baylor, L.R.; Foust, C.R.

    1993-01-01

    The projected fueling requirements of future magnetic confinement devices for controlled thermonuclear research [e.g., the International Thermonuclear Experimental Reactor (ITER)] indicate that a flexible plasma fueling capability is required. This includes a mix of traditional gas puffing and low- and high-velocity deuterium-tritium pellets. Conventional pellet injectors (based on light gas guns or centrifugal accelerators) can reliably provide frozen hydrogen pellets (1- to 6-mm-diam sizes tested) up to ∼1.3-km/s velocity at the appropriate pellet fueling rates (1 to 10 Hz or greater). For long-pulse operation in a higher velocity regime (>2 km/s), an experiment in collaboration between Oak Ridge National Laboratory (ORNL) and ENEA Frascati is under way. This activity will be carried out in the framework of a collaborative agreement between the US Department of Energy and European Atomic Energy Community -- ENEA Association. In this experiment, an existing ORNL hydrogen extruder (equipped with a pellet chambering mechanism/gun barrel assembly) and a Frascati two-stage light gas gun driver have been combined on a test facility at ORNL. Initial testing has been carried out with single deuterium pellets accelerated up to 2.05 km/s with the two-stage driver; in addition, some preliminary repetitive testing (to commission the diagnostics) was performed at reduced speeds, including sequences at 0.5 to 1 Hz and 10 to 30 pellets. The primary objective of this study is to demonstrate repetitive operation (up to ∼1 Hz) with speeds in the 2- to 3-km/s range. In addition, the strength of extruded hydrogen ice as opposed to that produced in situ by direct condensation in pipe guns can be investigated. The equipment and initial experimental results are described

  18. Three-dimensional FE analysis of the thermal-mechanical behaviors in the nuclear fuel rods

    International Nuclear Information System (INIS)

    Jiang Yijie; Cui Yi; Huo Yongzhong; Ding Shurong

    2011-01-01

    Highlights: → We establish three-dimensional finite element models for nuclear fuel rods. → The thermal-mechanical behaviors at the initial stage of burnup are obtained. → Several parameters on the in-pile performances are investigated. → The parameters have remarkable effects on the in-pile behaviors. → This study lays a foundation for optimal design and irradiation safety. - Abstract: In order to implement numerical simulation of the thermal-mechanical behaviors in the nuclear fuel rods, a three-dimensional finite element model is established. The thermal-mechanical behaviors at the initial stage of burnup in both the pellet and the cladding are obtained. Comparison of the obtained numerical results with those from experiments validates the developed finite element model. The effects of the constraint conditions, several operation and structural parameters on the thermal-mechanical performances of the fuel rod are investigated. The research results indicate that: (1) with increasing the heat generation rates from 0.15 to 0.6 W/mm 3 , the maximum temperature within the pellet increases by 99.3% and the maximum radial displacement at the outer surface of the pellet increases by 94.3%. And the maximum Mises stresses in the cladding all increase; while the maximum values of the first principal stresses within the pellet decrease as a whole; (2) with increasing the heat transfer coefficients between the cladding and the coolant, the internal temperatures reduce and the temperature gradient remains similar; when the heat transfer coefficient is lower than a critical value, the temperature change is sensitive to the heat transfer coefficient. The maximum temperature increases only 7.13% when h changes from 0.5 W/mm 2 K to 0.01 W/mm 2 K, while increases up to 54.7% when h decreases from 0.01 W/mm 2 K to 0.005 W/mm 2 K; (3) the initial gap sizes between the pellet and the cladding significantly affect the thermal-mechanical behaviors in the fuel rod; when the

  19. Properties of the high burnup structure in nuclear light water reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wiss, Thierry; Rondinella, Vincenzo V.; Konings, Rudy J.M. [European Commission, Joint Research Centre, Karlsruhe (Germany). Directorate Nuclear Safety and Security; and others

    2017-07-01

    The formation of the high burnup structure (HBS) is possibly the most significant example of the restructuring processes affecting commercial nuclear fuel in-pile. The HBS forms at the relatively cold outer rim of the fuel pellet, where the local burnup is 2-3 times higher than the average pellet burnup, under the combined effects of irradiation and thermo-mechanical conditions determined by the power regime and the fuel rod configuration. The main features of the transformation are the subdivision of the original fuel grains into new sub-micron grains, the relocation of the fission gas into newly formed intergranular pores, and the absence of large concentrations of extended defects in the fuel matrix inside the subdivided grains. The characterization of the newly formed structure and its impact on thermo-physical or mechanical properties is a key requirement to ensure that high burnup fuel operates within the safety margins. This paper presents a synthesis of the main findings from extensive studies performed at JRC-Karlsruhe during the last 25 years to determine properties and behaviour of the HBS. In particular, microstructural features, thermal transport, fission gas behaviour, and thermo-mechanical properties of the HBS will be discussed. The main conclusion of the experimental studies is that the HBS does not compromise the safety of nuclear fuel during normal operations.

  20. Fission gas release and pellet microstructure change of high burnup BWR fuel

    International Nuclear Information System (INIS)

    Itagaki, N.; Ohira, K.; Tsuda, K.; Fischer, G.; Ota, T.

    1998-01-01

    UO 2 fuel, with and without Gadolinium, irradiated for three, five, and six irradiation cycles up to about 60 GWd/t pellet burnup in a commercial BWR were studied. The fission gas release and the rim effect were investigated by the puncture test and gas analysis method, OM (optical microscope), SEM (scanning electron microscope), and EPMA (electron probe microanalyzer). The fission gas release rate of the fuel rods irradiated up to six cycles was below a few percent; there was no tendency for the fission gas release to increase abruptly with burnup. On the other hand, microstructure changes were revealed by OM and SEM examination at the rim position with burnup increase. Fission gas was found depleted at both the rim position and the pellet center region using EPMA. There was no correlation between the fission gas release measured by the puncture test and the fission gas depletion at the rim position using EPMA. However, the depletion of fission gas in the center region had good correlation with the fission gas release rate determined by the puncture test. In addition, because the burnup is very large at the rim position of high burnup fuel and also due to the fission rate of the produced Pu, the Xe/Kr ratio at the rim position of high burnup fuel is close to the value of the fission yield of Pu. The Xe/Kr ratio determined by the gas analysis after the puncture test was equivalent to the fuel average but not to the pellet rim position. From the results, it was concluded that fission gas at the rim position was released from the UO 2 matrix in high burnup, however, most of this released fission gas was held in the porous structure and not released from the pellet to the free volume. (author)

  1. Roll compaction and granulation system for nuclear fuel material

    International Nuclear Information System (INIS)

    Goldmann, L.H. Jr.; Holley, C.C.

    1981-01-01

    A roll compaction and roll granulation system has been designed and fabricated to replace conventional preslugging and crushing operations typically used in the fabrication of mixed oxide nuclear fuel pellets. This equipment will be of maintenance advantage with only the compaction and granulation rolls inside containment. The prototype is being tested and the results will be reported within a year

  2. A model for radial cesium transport in a fuel pellet

    International Nuclear Information System (INIS)

    Imoto, Shosuke

    1989-01-01

    In order to explain the radial redistribution of cesium in an irradiated pellet, a two-step release model is proposed. The first step involves the migration of cesium by atomic diffusion to some channels, such as grain boundaries and cracks, and the second step assumes a thermomigration down along the temperature gradient. Distribution profiles of cesium are obtained by numerical calculation with the present model assuming a constant and spatially uniform birth rate of cesium in the pellet. The result agrees well with the profile observed by micro-gamma scanning for the LWR fuel in the outer region of the pellet but diverges from it at the inner region. Discussion is made on the steady-state model hitherto generally utilized. (orig.)

  3. Study on the possibility of supercritical fluid extraction for reprocessing of spent nuclear fuel from high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Duan Wuhua; Zhu Liyang; Zhu Yongjun; Xu Jingming

    2011-01-01

    International interest in high temperature gas-cooled reactor (HTGR) has been increasing in recent years. It is important to study on reprocessing of spent nuclear fuel from HTGR for recovery of nuclear resource and reduction of nuclear waste. Treatment of UO 2 pellets for preparing fuel elements of the 10 MW high temperature gas-cooled reactor (HTR-10) using supercritical fluid extraction was investigated. UO 2 pellets are difficult to be directly dissolved and extracted with TBP-HNO 3 complex in supercritical CO 2 (SC-CO 2 ), and the extraction efficiency is only about 7% under experimental conditions. UO 2 pellets are also difficult to be converted completely into nitrate with N 2 O 4 . When UO 2 pellets break spontaneously into U 3 O 8 powders with particle size below 100 μm under O 2 flow and 600degc, the extraction efficiency of U 3 O 8 powders with TBP-HNO 3 complex in SC-CO 2 can reach more than 98%. U 3 O 8 powders are easy to be completely converted into nitrate with N 2 O 4 . The extraction efficiency of the nitrate product with TBP in SC-CO 2 can reach more than 99%. So it has a potential prospect that application of supercritical fluid extraction in reprocessing of spent nuclear fuel from HTGR. (author)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  5. Chemical characterization of nuclear fuel materials

    International Nuclear Information System (INIS)

    Ramakumar, K.L.

    2011-01-01

    control programme would result in reducing generation of analytical waste especially of plutonium. Selection of the evaluation techniques is made considering various aspects viz. accuracy, reliability, type of testing (non-destructive or otherwise), analysis time etc. In order to meet stringent chemical specifications, analyses of process intermediates and final pellets are essential. It should be realized that nuclear fission occurs only in the fissile atoms (e.g. 235 U or 239 Pu) and during nuclear fission a number of fission products starting from Br to Gd are produced making the nuclear fuel a highly heterogeneous and 'impure' system. But to start with the nuclear fuel should be as pure as possible. But invariably a number of trace impurities are always present in the fuel. There are a number of factors which need to be considered for realizing the designed performance of the nuclear fuel in side the reactor

  6. Simulation of the behaviour of nuclear fuel under high burnup conditions

    International Nuclear Information System (INIS)

    Soba, Alejandro; Lemes, Martin; González, Martin Emilio; Denis, Alicia; Romero, Luis

    2014-01-01

    Highlights: • Increasing the time of nuclear fuel into reactor generates high burnup structure. • We analyze model to simulate high burnup scenarios for UO 2 nuclear fuel. • We include these models in the DIONISIO 2.0 code. • Tests of our models are in very good agreement with experimental data. • We extend the range of predictability of our code up to 60 MWd/KgU average. - Abstract: In this paper we summarize all the models included in the latest version of the DIONISIO code related to the high burnup scenario. Due to the extension of nuclear fuels permanence under irradiation, physical and chemical modifications are developed in the fuel material, especially in the external corona of the pellet. The codes devoted to simulation of the rod behaviour under irradiation need to introduce modifications and new models in order to describe those phenomena and be capable to predict the behaviour in all the range of a general pressurized water reactor. A complex group of subroutines has been included in the code in order to predict the radial distribution of power density, burnup, concentration of diverse nuclides and porosity within the pellet. The behaviour of gadolinium as burnable poison also is modelled into the code. The results of some of the simulations performed with DIONISIO are presented to show the good agreement with the data selected for the FUMEX I/II/III exercises, compiled in the NEA data bank

  7. Selective alpha autoradiography for monitoring thorium distribution in UO[sub 2]-ThO[sub 2] fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Shriwastwa, B.B.; Raghunath, B.; Ghosh, J.K. (Bhabha Atomic Research Centre, Bombay (India))

    1992-10-01

    Although natural uranium and thorium decay with similar alpha energies (4.20 and 3.98 MeV), their daughter products have different alpha characteristics. This has been exploited for selective alpha autoradiography for thoria in urania-thoria mixed nuclear fuel pellets. Difficulties in getting sufficient track density in alpha sensitive films due to the very low specific activity of natural uranium and thorium material were overcome by using a special film with annealing and pre-etching treatment. (orig./HP).

  8. Incorporation of new tables into the preliminary structure of BDCN - nuclear fuel data base

    International Nuclear Information System (INIS)

    Pombo, J.B.S.; Andrade, M.C.

    1989-01-01

    The structure of the new tables related to fuel elements, fuel rods and fuel pellets have been incorporated into the developing structure of BDCN - nuclear fuel data base, are described. The list of all 44 tables structured up to now, with identification of its primary keys, as well as the reduced set of data has been entered to run the simulations are also presented. The simulation tests have been ran using a microcomputer.(M.I.)

  9. Dependence of sputtering erosion on fuel-pellet characteristics

    International Nuclear Information System (INIS)

    Bohachevsky, I.O.; Hafer, J.F.

    1977-11-01

    Conceptual designs of fusion reactors operating on the principle of inertial confinement require that the dependence of cavity-wall erosion on fuel-pellet energy yield, its mass, and representative atomic number be known. A simple approximate model of sputtering erosion is presented and explicit formulas are derived that express the total amount of eroded wall material in terms of the above three parameters

  10. Precedents for diversion-resistant nuclear fuel cycles

    International Nuclear Information System (INIS)

    Culler, F.L. Jr.

    1978-01-01

    The urgent need to limit the spread of nuclear weapons and to control the means of production of fissionable material has been the dominant force in the worldwide development of civilian nuclear power. The author follows the historical perspective for institutional control. To improve diversion resistance of the back end of the fuel recycle, the Civex process is proposed. The Civex process does not use new separation process principles or new methods for fuel fabrication. Rather, it is a combination of processes used and partially developed techniques for breeder fuel reprocessing and refabrication. Its characteristics are listed. The process steps and the design knowledge to meet these criteria, and to operate under conditions that provide maximum diversion resistance, can be adaptations of methods studied earlier and, in most cases, used for both military and civilian fuel recycle. The adaptations change the original techniques enough to make the technology different from that used for existing power reactors. The author discusses tried or partially demonstrated techniques from which Civex has been or could be fashioned. Separation processes discussed are bismuth phosphate; Purex; Thorex; fluoride volatility; pyrometallurgy. The Sol--Gel Uranium--Plutonium Spherepak and Pellet Fuels processes are discussed as candidates for Civex fuel-production methods. The author concludes that, in his opinion, the Civex process is as far as technology can go in the back end of the nuclear fuel cycle from illicit diversion of fissile materials

  11. Development of thermocouple re-instrumentation technique for irradiated fuel rod. Techniques for making center hole into UO2 pellets and thermocouple re-instrumentation to fuel rod

    International Nuclear Information System (INIS)

    Shimizu, Michio; Saito, Junichi; Oshima, Kunio

    1995-07-01

    The information on FP gas pressure and centerline temperature of fuel pellets during power transient is important to study the pellet clad interaction (PCI) mechanism of high burnup LWR fuel rods. At the Department of JMTR, a re-instrumentation technique of FP gas pressure gage for an irradiated fuel rod was developed in 1990. Furthermore, a thermocouple re-instrumentation technique was successfully developed in 1994. Two steps were taken to carry out the development program of the thermocouple re-instrumentation technique. In the first step, a drilling technique was developed for making a center hole of the irradiated fuel pellets. Various drilling tests were carried out using dummy of fuel rods consisted of Ba 2 FeO 3 pellets and Zry-2 cladding. On this work it is important to keep the pellets just the state cracked at a power reactor. In these tests, the technique to fix the pellets by frozen CO 2 was used during the drilling work. Also, diamond drills were used to make the center hole. These tests were completed successfully. A center hole, 54mm depth and 2.5mm diameter, was realized by these methods. The second step of this program is the in-pile demonstration test on an irradiated fuel rod instrumented dually a thermocouple and FP gas pressure gage. The demonstration test was carried out at the JMTR in 1995. (author)

  12. Role of analytical chemistry in the development of nuclear fuels

    International Nuclear Information System (INIS)

    Ramakumar, K.L.

    2012-01-01

    Analytical chemistry is indispensable and plays a pivotal role in the entire gamut of nuclear fuel cycle activities starting from ore refining, conversion, nuclear fuel fabrication, reactor operation, nuclear fuel reprocessing to waste management. As the fuel is the most critical component of the reactor where the fissions take place to produce power, extreme care should be taken to qualify the fuel. For example, in nuclear fuel fabrication, depending upon the reactor system, selection of nuclear fuel has to be made. The fuel for thermal reactors is normally uranium oxide either natural or slightly enriched. For research reactors it can be uranium metal or alloy. The fuel for FBR can be metal, alloy, oxide, carbide or nitride. India is planning an advanced heavy water reactor for utilization of vast resources of thorium in the country. Also research is going on to identify suitable metallic/alloy fuels for our future fast reactors and possible use in fast breeder test reactor. Other advanced fuel materials are also being investigated for thermal reactors for realizing increased performance levels. For example, advanced fuels made from UO 2 doped with Cr 2 O 3 and Al 2 O 3 are being suggested in LWR applications. These have shown to facilitate pellet densification during sintering and enlarge the pellet grain size. The chemistry of these materials has to be understood during the preparation to the stringent specification. A number of analytical parameters need to be determined as a part of chemical quality control of nuclear materials. Myriad of analytical techniques starting from the classical to sophisticated instrumentation techniques are available for this purpose. Insatiable urge of the analytical chemist enables to devise and adopt new superior methodologies in terms of reduction in the time of analysis, improvement in the measurement precision and accuracy, simplicity of the technique itself etc. Chemical quality control provides a means to ensure that the

  13. Furnace for the continuous sintering of pellets of ceramic nuclear fuel material

    International Nuclear Information System (INIS)

    Heyraud, J.

    1977-01-01

    The furnace comprises a hearth for the longitudinal displacement of pellet containers, means for injecting gas at both ends of the furnace, for sucking gas between preheating and sintering zones and for condensing the binder, means for displacing the containers from an introduction lock-chamber to an extraction lock-chamber, a conveyor belt which passes through a glove box and provides a leak-tight connection between the lock-chambers. A station for loading containers with pellet sub-containers prior to sintering and a station for unloading the pellet sub-containers after sintering are juxtaposed within the glove box. 3 claims, 1 drawing figure

  14. Calculations on the effect of pellet filling on the rewetting of overheated nuclear reactor fuel pins

    International Nuclear Information System (INIS)

    Pearson, K.G.; Loveless, J.

    1977-03-01

    Numerical solutions of the rewetting equations are presented which show the effect of filler material and gas gap on the rate of rewetting of an overheated fuel pin. It is shown that taking the presence of the fuel into account can lead to a large reduction in the calculated rewetting speed compared with a calculation which neglects the presence of fuel. The effect is most marked in conditions where rewetting speeds tend to be already low, such as at high pin temperatures and low ambient pressure. A comparison is made between the predictions of the present method and experimental data obtained on zircaloy and stainless steel pins filled with magnesia and with boron nitride. In all cases filling the pins produced a large reduction in rewetting speed and the agreement between the calculated and measured effect was encouraging. It is concluded that the presence of the UO 2 pellet filling should be taken into account when calculating rewetting speeds in safety assessments. (author)

  15. Fuel rods

    International Nuclear Information System (INIS)

    Adachi, Hajime; Ueda, Makoto

    1985-01-01

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

  16. Integrated analysis of oxide nuclear fuel sintering

    International Nuclear Information System (INIS)

    Baranov, V.; Kuzmin, R.; Tenishev, A.; Timoshin, I.; Khlunov, A.; Ivanov, A.; Petrov, I.

    2011-01-01

    Dilatometric and thermal-gravimetric investigations have been carried out for the sintering process of oxide nuclear fuel in gaseous Ar - 8% H 2 atmosphere at temperatures up to 1600 0 C. The pressed compacts were fabricated under real production conditions of the OAO MSZ with application of two different technologies, so called 'dry' and 'wet' technologies. Effects of the grain size growth after the heating to different temperatures were observed. In order to investigate the effects produced by rate of heating on properties of sintered fuel pellets, the heating rates were varied from 1 to 8 0 C per minute. Time of isothermal overexposure at maximal temperature (1600 0 C) was about 8 hours. Real production conditions were imitated. The results showed that the sintering process of the fuel pellets produced by two technologies differs. The samples sintered under different heating rates were studied with application of scanning electronic microscopy analysis for determination of mean grain size. A simulation of heating profile for industrial furnaces was performed to reduce the beam cycles and estimate the effects of variation of the isothermal overexposure temperatures. Based on this data, an optimization of the sintering conditions was performed in operations terms of OAO MSZ. (authors)

  17. Thoria-fuel irradiation. Program to irradiate 80% ThO2/20% UO2 ceramic pellets at the Savannah River Plant

    International Nuclear Information System (INIS)

    Pickett, J.B.

    1982-02-01

    This report describes the fabrication of proliferation-resistant thorium oxide/uranium oxide ceramic fuel pellets and preparations at the Savannah River Laboratory (SRL) to irradiate those materials. The materials were fabricated in order to study head end process steps (decladding, tritium removal, and dissolution) which would be required for an irradiated proliferation-resistant thorium based fuel. The thorium based materials were also to be studied to determine their ability to withstand average commercial light water reactor (LWR) irradiation conditions. This program was a portion of the Thorium Fuel Cycle Technology (TFCT) Program, and was coordinated by the Oak Ridge National Laboratory (ORNL) under the Consolidated Fuel Reprocessing Program (CFRP). The fuel materials were to be irradiated in a Savannah River Plant (SRP) reactor at conditions simulating the heat ratings and burnup of a commercial LWR. The program was terminated due to a de-emphasis of the TFCT Program, following completion of the fabrication of the fuel and the modified assemblies which were to be used in the SRP reactor. The reactor grade ceramic pellets were fabricated for SRL by Battelle, Pacific Northwest Laboratories. Five fuel types were prepared: 100% UO 2 pellets (control); 80% ThO 2 /20% UO 2 pellets; approximately 80% ThO 2 /20% UO 2 + 0.25 CaO (dissolution aid) pellets; 100% UO 2 hybrid pellets (prepared from sol-gel microspheres); and 100% ThO 2 pellets (control). All of the fuel materials were transferred to SRL from PNL and were stored pending a subsequent reactivation of the TFCT Programs

  18. A pellet model of DT ignitor and DD fuel for an ICF reactor without tritium breeding blanket

    International Nuclear Information System (INIS)

    Ido, Shunji; Tazima, Teruhiko.

    1983-01-01

    A pellet concept of a DT ignitor and DD fuel for an ICF reactor without a tritium breeding blanket is analytically examined under the condition that T is bred through the DD reactions. There is the additional restriction that the tritium breeding ratio in a pellet is unity, including the in situ DT burn in the DD region. Model calculations show that sufficiently high pellet gain can be obtained in a DT-DD pellet, when fuel rhoR increases to --40 g/cm 2 and the fraction of energy released in the DD region becomes dominant. One-dimensional neutronics calculations carried out for a reference pellet model with rhoR --40 g/cm 2 show that the neutron heating in the compressed pellet model is evident and the total energy of the neutrons escaping from the pellet is reduced from --2000 MJ to 330 MJ for a microexplosion of --3000 MJ. (author)

  19. Establishing QC/QA system in the fabrication of nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Suh, K.S.; Choi, S.K.; Park, H.G.; Park, T.G.; Chung, J.S.

    1980-01-01

    Quality control instruction manuals and inspection methods for UO 2 powder and zircaloy materials as the material control, and for UO 2 pellets and nuclear fuel rods as the process control were established. And for the establishment of Q.A programme, the technical specifications of the purchased materials, the control regulation of the measuring and testing equipments, and traceability chart as a part of document control have also been provided and practically applied to the fuel fabrication process

  20. Hot vacuum outgassing to ensure low hydrogen content in MOX fuel pellets for thermal reactors

    International Nuclear Information System (INIS)

    Majumdar, S.; Nair, M.R.; Kumar, Arun

    1983-01-01

    Hot vacuum outgassing treatment to ensure low hydrogen content in Mixed Oxide Fuel (MOX) pellets for thermal reactors has been described. Hypostoichiometric sintered MOX pellets retain more hydrogen than UO 2 pellets. The hydrogen content further increases with the addition of admixed lubricant and pore formers. However, low hydrogen content in the MOX pellets can be ensured by a hot vacuum outgassing treatment at a temperature between 773K to 823K for 2 hrs. (author)

  1. The pellet-cladding contact in a fuel rod and its simulation by finite elements

    International Nuclear Information System (INIS)

    Tanajura, C.A.S.

    1988-01-01

    A model to analyse the mechanical behavior of a fuel rod of a PWR is presented. We drew our attention to the phenomenon of pellet-pellet and pellet-cladding contact by taking advantage of a model which assumes the hypotheses of axisymmetry, elastic behavior with infinitesimal deformations and changes of the material properties with temperature. It also includes the effects of swelling and initial relocation. The problem of contact gives rise to a variational formulation which employs Lagrangian multipliers. With this approach an iterative scheme is constructed to obtain the solution. The finite element method is applied to space discretization. The model sensibility to some parameters and its performance concerning fuel rod behavior is discussed by means of numerical simulations. (author) [pt

  2. Burn characteristics of compressed fuel pellets for D-3He inertial fusion

    International Nuclear Information System (INIS)

    Nakao, Y.; Honda, T.; Honda, Y.; Kudo, K.; Nakashima, H.

    1992-01-01

    In this paper, the feasibility of using D- 3 He fuel in inertial confinement fusion is examined by using a hydrodynamics code that includes neutron and charged-particle transport routines. The use of a small amount of deuterium-tritium (D-T) ignitor is indispensable. Burn simulations are made for quasi-isobaric D-T/D- 3 He pellet models compressed to 5000 times the liquid density. Substantial fuel gains (∼500) are obtained from pellets having parameters ρR D-T = 3 g/cm 2 and ρR total = 14 g/cm 2 and a central spark temperature of 5 keV. The amount of driver energy needed to achieve these gains is estimated to be ∼ 30 MJ when the coupling efficiency is 10%. The driver energy requirement can be reduced by using spin-polarized D-T and D- 3 He fuels

  3. Modular nuclear fuel element, modular capsule for a such element and fabrication process for a modular capsule

    International Nuclear Information System (INIS)

    Chotard, A.

    1988-01-01

    The nuclear fuel rod is made by a tubular casing closed at both ends and containing a series of modular capsules with little play with the casing and made by a jacket closed by porous plugs at both ends and containing a stack of fuel pellets [fr

  4. PELLET: a computer routine for modeling pellet fueling in tokamak plasmas

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Iskra, M.A.; Howe, H.C.; Attenberger, S.E.

    1979-01-01

    Recent experimental results of frozen hydrogenic pellet injection into hot tokamak plasmas and substantial agreement with theoretical predictions have led to a much greater interest in pellets as a means of refueling plasmas. The computer routine PELLET has been developed and used as an aid in assessing pellet ablation models and the effects of pellets on plasma behavior. PELLET provides particle source profiles under various options for the ablation model and can be coupled either to a fluid transport code or to a brief routine which supplies the required input parameters

  5. Effect of the UO{sub 2} powder type and mixing method on microstructure of Mn-Al doped pellet

    Energy Technology Data Exchange (ETDEWEB)

    Na, Yeon Soo; Lim, Kwang Young; Choi, Min young; Jung, Tae Sik; Lee, Seung Jae; Yoo, Jong Sung [KEPCO, Daejeon (Korea, Republic of)

    2016-05-15

    Recently, the commercial LWRs are focused on the extending the burn-up and fuel cycle length in order to increase nuclear power plant economy as a maintenance and fuel cycle cost. Increasing the burn-up may lead to a faster and higher power variation such as a peak local linear power and normal operating transient (Load following operation). In such operating conditions, the risk of a fuel failure is considerably related to a pellet clad-interaction (PCI). So, recent development of advanced UO{sub 2} pellet for the LWRs is mainly focused on the large grain and soft pellet as they can reduce corrosive fission gas release and pellet-clad-interaction. In terms of the UO{sub 2} pellet, the prevention of PCI induced fuel failure can be achieved by enlarging the UO{sub 2} pellet grain size and enhancing the pellets deformation at an elevated temperature. In Korea, in order to increase the grain size and deformation of UO{sub 2} pellet on the high temperature, Mn-Al doped pellet with ADU (Ammonium Diuranate)-UO{sub 2} powder are developed in lab scale. But, the UO{sub 2} pellets for the commercial nuclear power plants in Korea are fabricated using the DC (Dry Conversion)-UO{sub 2} powder. So, it is necessary to understand the effect of microstructure on UO{sub 2} powder type for Mn-Al doped pellets. In this work, to investigate the effect of UO{sub 2} powder type and mixing method on the microstructure of the Mn-Al doped UO{sub 2} pellets, we fabricated the Mn-Al doped pellets using the DC-UO{sub 2} powder. The measurement of sintered density and mean grain size for fabricated pellets was performed, and then the results of test was evaluated in comparison with a Reference 2.

  6. Process for the fabrication of a nuclear fuel

    International Nuclear Information System (INIS)

    Hirose, Yasuo.

    1970-01-01

    Herein disclosed is a process for fabricating a nuclear fuel incorporating either uranium or plutonium. A pellet-like substrate consisting of a packed powder ceramic fuel such as uranium or plutonium is prepared with the horizontal surface of the body provided with a masking. Next, after impregnating the substrate voids with a solution consisting of a fissile material or mixture of fissile material and poison, the solvent is removed by a chemical deposition process which causes the impregnated material to migrate through capillary action toward the vicinity of the fuel body surface. Sintering and pyrolysis of the deposited material and masking are subsequently carried out to yield a fuel body having adjacent to its surface an intensely concentrated layer of either fissile material or a mixture of fissile material and poison. (Owens, K.J.)

  7. Advances in fuel pellet technology for improved performance at high burnup. Proceedings of a Technical Committee meeting

    International Nuclear Information System (INIS)

    1998-08-01

    The IAEA has recently completed two co-ordinated Research Programmes (CRPs) on The Development of Computer Models for Fuel Element Behaviour in Water Reactors, and on Fuel Modelling at Extended Burnup. Through these CRPs it became evident that there was a need to obtain data on fuel behaviour at high burnup. Data related o thermal behaviour, fission gas release and pellet to clad mechanical interaction were obtained and presented at the Technical Committee Meeting on Advances in Fuel Pellet Technology for Improved Performance at High Burnup which was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). The 34 papers from 10 countries are published in this proceedings and presented by a separate abstract. The papers were grouped in 6 sessions. First two sessions covered the fabrication of both UO 2 fuel and additives and MOX fuel. Sessions 3 and 4 covered the thermal behaviour of both types of fuel. The remaining two sessions dealt with fission gas release and the mechanical aspects of pellet to clad interaction

  8. Standard Specification for Nuclear Grade Zirconium Oxide Pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This specification applies to pellets of stabilized zirconium oxide used in nuclear reactors. 1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.

  9. 3D finite element analysis of a nuclear fuel rod with gap elements between the pellet and the cladding

    International Nuclear Information System (INIS)

    Kang, Chang-Hak; Lee, Sung-Uk; Yang, Dong-Yol; Kim, Hyo-Chan; Yang, Yong-Sik

    2016-01-01

    Nuclear fuel rods which comprises an important component of a nuclear power plant are composed of nuclear fuel and cladding. Simulating the nuclear fuel rod using a computer program is the universal method to verify its safety. The computer program used for this is called the fuel performance code. The main objective of this study is to simulate the nuclear fuel rod behavior considering the gap conductance using three-dimensional gap elements. Gap elements are used because, unlike other methods, this approach does not require special methods or other variables such as the Lagrange multiplier. In this work, a nuclear fuel rod has been simulated and the results are compared with the experimental results. (author)

  10. Burnable poison rod for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Funk, C.E.; Oneufer, A.S.

    1984-01-01

    A burnable poison rod for use in a nuclear reactor fuel assembly which includes concentrically disposed rods having an annular space therebetween which extends the full length of the rods. The inner rod is hollow to permit circulation of coolant therethrough. Annular burnable poison pellets are positioned in the annular space which is closed at both ends by plugs. A spring clip is located in the plenum space above the pellet stack in the rods. The spring clip is of cylindrical configuration having a gap in the material which provides two ends adapted to be squeezed toward each other. A cross section of the clip shows that its ends contain alternating flat and round edges, the round edges conforming to the outer rod inner surface to provide a retentive force which is releasably applied to the pellet stack as it grows during operation in a reactor

  11. Preliminary study of cost benefits associated with duplex fuel pellets of the LOWI type

    International Nuclear Information System (INIS)

    Ainscough, J.B.; Coucill, D.N.; Howl, D.A.; Jensen, A.; Misfeldt, I.

    1983-01-01

    Duplex UO 2 pellets, which consist of an outer enriched annulus and a depleted or natural core, can provide a solution to the problem of stress corrosion cracking failures, which have led to constraints being placed on ramp rates in power reactors. An analysis of the reactor physics and the performance of duplex pellets is presented in the context of a 17 X 17 pressurized water reactor fuel rod design. The study has been based on the particular type of duplex pellet in which the core and the annulus are physically separate; this is called ''LOWI'' after the Danish design. At low burnup, this fuel shows a significant improvement in power ramp performance compared with standard fuel. At higher burnup, the benefits are less certain but as the severity of the ramp will usually be less in high burnup fuel simply because of the reduced rating, the reduction in benefit may not be significant. If the gap between the core and annulus persists to high burnup, there will be no loss of benefit. Economic calculations and a cost-benefit analysis are presented to show the number of extra full-power hours of reactor operation that must be obtained in order to outweigh the additional fabrication costs associated with this fuel

  12. Proposal for Ultrasonic Technique for evaluation elastic constants in UO2 pellets

    International Nuclear Information System (INIS)

    Lopes, Alessandra Susanne Viana Ragone; Baroni, Douglas Brandao; Bittencourt, Marcelo de Siqueira Queiroz; Souza, Mauro Carlos Lopes

    2015-01-01

    Pellets of uranium dioxide are used as fuel in nuclear power reactors, in which are exposed to high thermal gradients. This high energy will initiate fusion in the central part of the pellet. The expansion of the uranium dioxide pellets, resulting from fission products, can cause fissures or cracks, therefore, the study of their behavior is important. This work aims to develop and propose an ultrasonic technique to evaluate the elastic constants of UO 2 pellets. However, because of the difficulties in handling nuclear material, we proposed an initial study of alumina specimens. Alumina pellets are also ceramic material and their porosity and dimensions are in the similar range of dioxide uranium pellets. They also are used as thermal insulation in the fuel rods, operating under the same conditions. They were fabricated and used in two different sets of 10 alumina pellets with densities of 92% and 96%. The developed ultrasonic technique evaluates the traveling time of ultrasonic waves, longitudinal and transverse, and correlates the observed time and the elastic constants of the materials. Equations relating the speed of the ultrasonic wave to the elastic modulus, shear modulus and Poisson's ratio have led to these elastic constants, with graphics of correlation that showed excellent agreement with the literature available for Alumina. In view of the results and the ease of implementation of this technique, we believe that it may easily be used for dioxide uranium pellets, justifying further studies for that application. (author)

  13. Improved fueling and transport barrier formation with pellet injection from different locations on DIII-D

    International Nuclear Information System (INIS)

    Baylor, L.R.; Jernigan, T.C.; Gohil, P.

    2001-01-01

    Pellet injection has been employed on DIII-D from different injection locations to optimize the mass deposition for density profile control and internal transport barrier formation. Transport barriers have been formed deep in the plasma core with central mass deposition from high field side (HFS) injected pellets and in the edge with pellets that trigger L-mode to H-mode transitions. Pellets injected from all locations can trigger the H-mode transition, which depends on the edge density gradient created and not on the radial extent of the pellet deposition. Pellets injected from inside the magnetic axis from the inner wall or vertical port lead to stronger central mass deposition than pellets injected from the low field side (LFS) and thus yield deeper more efficient fueling. (author)

  14. Impact of fuel quality and burner capacity on the performance of wood pellet stove

    OpenAIRE

    Petrović-Bećirović Sanja B.; Manić Nebojša G.; Stojiljković Dragoslava D.

    2015-01-01

    Pellet stoves may play an important role in Serbia in the future when fossil fuel fired conventional heating appliances are replaced by more efficient and environmentally friendly devices. Experimental investigation was conducted in order to examine the influence of wood pellet quality, as well as burner capacity (6, 8 and 10 kW), used in the same stove configuration, on the performance of pellet stove with declared nameplate capacity of 8 kW. The results o...

  15. Review of hydrogen pellet injection technology for plasma fueling applications

    International Nuclear Information System (INIS)

    Milora, S.L.

    1989-01-01

    In the past several years, steady progress has been made worldwide in the development of high-speed hydrogen pellet injectors for fueling magnetically confined plasmas. Several fueling systems based on the conventional pneumatic and centrifuge acceleration concepts have been put into practice on a wide variety of toroidal plasma confinement devices. Long-pulse fueling has been demonstrated in the parameter range 0.8--1.3 km/s, for pellets up to 6 mm in diameter, and at delivery rates up to 40 Hz. Conventional systems have demonstrated the technology to speeds approaching 2 km/s, and several more exotic accelerator concepts are under development to meet the more demanding requirements of the next generation of reactor-grade plasmas. These include a gas gun that can operate in tritium, the two-stage light gas gun, electrothermal guns, electromagnetic rail guns, and an electron-beam-driven thruster. Although these devices are in various stages of development, velocities of 3.8 km/s have already been achieved with two-stage light gas guns, and the prospects for attaining 5 km/s in the near future appear good

  16. Manufacture, delivery and marketing of wood pellets

    International Nuclear Information System (INIS)

    Huhtanen, T.

    2001-01-01

    Wood pellet is a cheap fuel, the use of which can easily bee automated. Pellet heating can be carried out with a stoker or a pellet burner, which can be mounted to oil and solid fuels boiler or to solid fuel boilers. Vapo Oy delivers wood pellet to farms and detached houses via Agrimarket stores. Vapo Oy delivers pellets to large real estates, municipalities, industry, greenhouses and power plants directly as bulk. The pellets are delivered either by trailers or lorries equipped with fan-operated unloaders. The use of wood pellets is a suitable fuel especially for real estates, the boiler output of which is 20 - 1000 kW. Vapo Oy manufactures wood pellets of cutter chips, grinding dust and sawdust. The raw material for Ilomantsi pellet plant is purchased from the province of North Karelia. The capacity of pelletizing plant is 45 000 t of pellets per year, half of which is exported mainly to Sweden and Denmark

  17. Development of the pellet grinding and inspection system

    International Nuclear Information System (INIS)

    Yamaguchi, Toshihiro; Kawasaki, Takao; Inui, Toshihiko; Yamada, Hiroyuki.

    1996-01-01

    The pellet grinding and inspection system is the facility of producing the MOX fuel for FBRs, and it was delivered in March, 1996 to the No. 3 Development Office of the plutonium fuel shop in Tokai of Power Reactor and Nuclear Fuel Development Corporation. This facility realized the high speed processing using the compact facility by combining the grinding of pellets, the inspection of outer density and appearance, the rationalization of the facility and the high performance automation equipment. The production processes of MOX comprise the processes of powder, sinter, finish and inspection, fabrication and assembling. The requirement for developing these facilities is shown. The features of this pellet grinding and inspection system are the compact facility by the combination of functions, high speed processing, the reduction of holdup in processes and the improvement of maintenance. These features are explained. The main components are simple sorting equipment, grinder, grinding dust recovering equipment, outside diameter and density sorting equipment and appearance inspection equipment. The constitution and the main functions of the control facility are described. The operational process is explained. (K.I.)

  18. Pelletizing of NaF granules as adsorbent for fluorides of nuclear fuel materials, (2)

    International Nuclear Information System (INIS)

    Kimura, Syojiro; Tsutsui, Tenson; Kanagawa, Akira.

    1976-01-01

    Experimental studies on compressibility of NaF granules are investigated. Brittleness and porosity of a tableted pellet depend on compressive power applied to the granules. Relation of the compressive power and distortion of NaF granules agree with Kawakita's formula. About the brittleness of NaF pellet tableted with optional pressure, the follow experimental formula is formed between axial destructive power and compressive power. Dsub(p)=f.psup(n) Then, relation of the axial destructive power and porosity of NaF pellet is given from the above mentioned experimental formula and Kawakita's formula. At the lower value of the porosity than 0.4, the axial destructive power varies steeply as compared with variation of the porosity. At optional porosity, the brittleness of the tableted NaF pellet are lower than Harshow Chem. Co. NaF.HF pellet. So it requires some process raising the brittleness of the pellet after the compression. (auth.)

  19. Effect of additives in sintering UO2-7wt%Gd2O3 fuel pellets

    International Nuclear Information System (INIS)

    Santos, L.R.; Riella, H.G.

    2009-01-01

    Gadolinium has been used as burnable poison for reactivity control in modern PWRs. The incorporation of Gd 2 O 3 powder directly into the UO 2 powder enables longer fuel cycles and optimized fuel utilization. Nevertheless, processing by this method leads to difficulties while obtaining sintered pellets with the minimum required density. The process for manufacturing UO 2 - Gd 2 O 3 generates scraps that should be reused. The main scraps are green and sintered pellets, which must be calcined to U 3 O 8 to return to the fabrication process. Also, the incorporation of Gd 2 O 3 in UO 2 requires the use of an additive to improve the sintering process, in order to achieve the physical properties specified for the mixed fuel, mainly density and microstructure. This paper describes the effect of the addition of fabrication scraps on the properties of the UO 2 -Gd 2 O 3 fuel. Aluminum hydroxide Al(OH) 3 was also incorporated to the fuel as a sintering aid. The results shown that the use of 2000 ppm of Al(OH) 3 as additive allow to fabricate good pellets with up to 10 wt% of recycled scraps. (author)

  20. Method of manufacturing UO2 pellet

    International Nuclear Information System (INIS)

    Harada, Yuhei; Asami, Yasuji.

    1989-01-01

    The present invention concerns a method of manufacturing UO 2 pellets with less FP gas release and having fine structure for moderating PCMI. At first, oxide nuclear fuel pellets are placed in a sintering furnance and preliminarily sintered in a H 2 gas atmosphere at 1400 - 1600 degC. In this step, sintering is progressed to about 90 % TD, by which closed cells are formed substantially completely. Then, when sintering is further advanced at an identical temperature in a CO 2 gas atmosphere, growth of the crystal grains is advanced at the central portion of the pellets. Then, reductive heat treatment is applied at the identical temperature in a H 2 gas atmosphere. As a result, pellets having a fine double structure with the larger grain size region being in the central portion and smaller grain size region in the outer periphery can be obtained. (I.J.)

  1. Oxygen-to-metal ratio control during fabrication of mixed oxide fast breeder reactor fuel pellets

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Benecke, M.W.; Jentzen, W.R.; McCord, R.B.

    1979-05-01

    Oxygen-to-metal ratio (O/M) of mixed oxide fuel pellets can be controlled during fabrication by proper selection of binder (type and content) and sintering conditions. Sintering condition adjustments involved the passing of Ar--8% H 2 sintering gas across a cryostat ice bath controlled to temperatures ranging from -5 to -60 0 C to control as-sintered pellet O/M ratio. As-sintered fuel pellet O/M decreased with increasing Sterotex binder and PuO 2 concentrations, increasing sintering temperature, and decreasing sintering gas dew point. Approximate relationships between Sterotex binder level and O/M were established for PuO 2 --UO 2 and PuO 2 --ThO 2 fuels. O/M was relatively insensitive to Carbowax binder concentration. Several methods of increasing O/M using post-sintering pellet heat treatments were demonstrated, with the most reliable being a two-step process of first raising the O/M to 2.00 (stoichiometric) at 650 0 C in Ar--8% H 2 bubbled through H 2 O, followed by hydrogen reduction to specification O/M in oxygen-gettered Ar-8% H 2 at temperatures ranging from 1200 to 1690 0 C

  2. FBR pellet fabrication - density and dimensional control

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Schaus, P.S.

    1982-01-01

    The fuel pellet fabricating experience described in this paper involved pellet processing tests using mixed oxide (PuO 2 -UO 2 ) powders to produce fast breeder reactor (FBR) fuel pellets. Objectives of the pellet processing tests were to establish processing parameters for sintered-to-size fuel pellets to be used in an irradiation test in the Fast Flux Test Facility and to establish baseline fabrication control information. 26 figures, 7 tables

  3. Calculation of pellet radial power distributions with a Monte Carlo burnup code

    International Nuclear Information System (INIS)

    Suzuki, Motomu; Yamamoto, Toru; Nakata, Tetsuo

    2010-01-01

    The Japan Nuclear Energy Safety Organization (JNES) has been working on an irradiation test program of high-burnup MOX fuel at Halden Boiling Water Reactor (HBWR). MOX and UO 2 fuel rods had been irradiated up to about 64 GWd/t (rod avg.) as a Japanese utilities research program (1st phase), and using those fuel rods, in-situ measurement of fuel pellet centerline temperature was done during the 2nd phase of irradiation as the JNES test program. As part of analysis of the temperature data, power distributions in a pellet radial direction were analyzed by using a Monte Carlo burnup code MVP-BURN. In addition, the calculated results of deterministic burnup codes SRAC and PLUTON for the same problem were compared with those of MVP-BURN to evaluate their accuracy. Burnup calculations with an assembly model were performed by using MVP-BURN and those with a pin cell model by using SRAC and PLUTON. The cell pitch and, therefore, fuel to moderator ratio in the pin cell calculation was determined from the comparison of neutron energy spectra with those of MVP-BURN. The fuel pellet radial distributions of burnup and fission reaction rates at the end of the 1st phase irradiation were compared between the three codes. The MVP-BURN calculation results show a large peaking in the burnup and fission rates in the pellet outer region for the UO 2 and MOX pellets. The SRAC calculations give very close results to those of the MVP-BURN. On the other hand, the PLUTON calculations show larger burnup for the UO 2 and lower burnup for the MOX pellets in the pellet outer region than those of MVP-BURN, which lead to larger fission rates for the UO 2 and lower fission rates for the MOX pellets, respectively. (author)

  4. Comparative study of the different industrial manufacturing routes for UO2 pellet specifications through the wet process

    International Nuclear Information System (INIS)

    Palheiros, Franklin; Gonzaga, Reinaldo; Soares, Alexandre

    2009-01-01

    In the fuel cycle, converting UF 6 to UO 2 powder is an intermediate step for fabrication of pellets for fuel assemblies to be used in nuclear power plants. The basic proposal common to the different powder fabrication processes is to provide raw material capable of being processed into the form of pellets. The wet processes is the most often used industrially and are divided in two categories: the ADU (Ammonium Diuranate) and AUC (Ammonium Uranyl Carbonate) processes, whose names originate in the precipitate obtained in aqueous solution during the intermediate steps of UO 2 powder fabrication. It has known that the powder characteristics have a considerable influence in the UO 2 pellet manufacturing and quality characteristics. INB has used the AUC process to produce UO 2 pellets and supply fuel to Angra 1 and 2 Nuclear Power Plants. Despite of this process is characterized by the precipitation of a different intermediate precipitate compared to the ADU route (i.e., (NH 4 ) 4 UO 2 (CO 3 ) 3 , in the AUC process, and (NH 4 ) 2 U 2 O 7 in ADU process) leading to some slight differences in the final pellet microstructure, it has been considered that the models that predict the pellet behavior during irradiation in a nuclear reactor are basically the same compared to those used to predict the pellets form the ADU process. In order to evaluate how different the pellets originated from these two industrial routes are, this paper aims to compare the INB production historical data (Angra 1, Cycles 14 and 15) with the key parameters of a common product specification from the ADU process. (author)

  5. A MULTIDIMENSIONAL AND MULTIPHYSICS APPROACH TO NUCLEAR FUEL BEHAVIOR SIMULATION

    Energy Technology Data Exchange (ETDEWEB)

    R. L. Williamson; J. D. Hales; S. R. Novascone; M. R. Tonks; D. R. Gaston; C. J. Permann; D. Andrs; R. C. Martineau

    2012-04-01

    Important aspects of fuel rod behavior, for example pellet-clad mechanical interaction (PCMI), fuel fracture, oxide formation, non-axisymmetric cooling, and response to fuel manufacturing defects, are inherently multidimensional in addition to being complicated multiphysics problems. Many current modeling tools are strictly 2D axisymmetric or even 1.5D. This paper outlines the capabilities of a new fuel modeling tool able to analyze either 2D axisymmetric or fully 3D models. These capabilities include temperature-dependent thermal conductivity of fuel; swelling and densification; fuel creep; pellet fracture; fission gas release; cladding creep; irradiation growth; and gap mechanics (contact and gap heat transfer). The need for multiphysics, multidimensional modeling is then demonstrated through a discussion of results for a set of example problems. The first, a 10-pellet rodlet, demonstrates the viability of the solution method employed. This example highlights the effect of our smeared cracking model and also shows the multidimensional nature of discrete fuel pellet modeling. The second example relies on our the multidimensional, multiphysics approach to analyze a missing pellet surface problem. As a final example, we show a lower-length-scale simulation coupled to a continuum-scale simulation.

  6. Simulation of the irradiation-induced micro-thermo-mechanical behaviors evolution in ADS nuclear fuel pellets

    Science.gov (United States)

    Ding, Shurong; Zhao, Yunmei; Wan, Jibo; Gong, Xin; Wang, Canglong; Yang, Lei; Huo, Yongzhong

    2013-11-01

    An Accelerator Driven System (ADS) is dedicated to Minor Actinides (MA) transmutation. The fuels for ADS are highly innovative, which are composite fuel pellets with the fuel particles containing MA phases dispersed in a MgO or Mo matrix. Assuming that the fuel particles are distributed periodically in the MgO matrix, a three-dimensional finite element model is developed. The three-dimensional incremental large-deformation constitutive relations for the fuel particles and matrix are separately built, and a method is accordingly constructed to implement simulation of the micro-thermo-mechanical behaviors evolution. Evolutions of the temperature and mechanical fields are given and discussed. With irradiation creep included in the MgO matrix constitutive relation, the conclusions can be drawn as that (1) irradiation creep has a remarkable effect on the mechanical behaviors evolution in the matrix; (2) irradiation creep plays an important role in the damage mechanism interpretation of ceramic matrix fuel pellets. Thermal conductivity The thermal conductivity model is adopted as KUO2 = K0·FD·FP·FM·FR, which was proposed by Lucuta et al. [10] to adapt to the high burnup conditions with consideration of the effects of temperature, burnup, porosity and fission products. K0 is the thermal conductivity of fully dense un-irradiated UO2, as Eq. (1) in W/m K; FD, FP are the adjust factors reflecting the effects of dissolved and precipitated fission products; FM and FR are factors due to porosity and irradiation effects. The adopted thermal conductivity varies with temperature and burnup, which expresses its degradation with burnup, with the terms as k0={1}/{0.0375+2.165×10-4T}+{4.715×109}/{T2}exp-{16361}/{T} FD={1.09}/{B3.265}+{0.0643}/{√{B}}√{T}artan{1}/{1.09/B3.265}+{0.0643}/{√{B}}√{T} FP=1+0.019B/3-0.019B{1}/{1+exp(1200-T100)} FM={1-P}/{1+(s-1)P} FR=1-{0.2}/{1+expT-90080} Thermal expansion The engineering strain of thermal expansion [11] is given as {ΔL}/{L0

  7. Modelling the cracking of pressurised water reactor fuel pellets and its consequences on the mechanical behaviour of the fuel rod; Etude de l'impact de la fissuration des combustibles nucleaires oxyde sur le comportement normal et incidentel des crayons combustible

    Energy Technology Data Exchange (ETDEWEB)

    Helfer, Th

    2006-03-15

    This thesis aims to model the cracking of pressurised water reactor fuel pellets and its consequences on the mechanical behaviour of the fuel rod. Fuel cracking has two main consequences. It relieves the stress in the pellet, upon which the majority of the mechanical and physico-chemical phenomena are dependent. It also leads to pellet fragmentation. Taking fuel cracking into account is therefore necessary to adequately predict the mechanical loading of the cladding during the course of an irradiation. The local approach to fracture was chosen to describe fuel pellet cracking. Practical considerations brought us to favour a quasi-static description of fuel cracking by means of a local damage models. These models describe the appearance of cracks by a local loss of rigidity of the material. Such a description leads to numerical difficulties, such as mesh dependency of the results and abrupt changes in the equilibrium state of the mechanical structure during unstable crack propagations. A particular attention was paid to these difficulties because they condition the use of such models in engineering studies. This work was performed within the framework of the ALCYONE fuel performance package developed at CEA/DEC/SESC which relies on the PLEIADES software platform. ALCYONE provides users with various approaches for modelling nuclear fuel behaviour, which differ in terms of the type geometry considered for the fuel rod. A specific model was developed and implemented to describe fuel cracking for each of these approaches. The 2D axisymmetric fuel rod model is the most innovative and was particularly studied. We show that it is able to assess, thanks to an appropriate description of fuel cracking, the main geometrical changes of the fuel rod occurring under normal and off-normal operating conditions. (author)

  8. The MARINE experiment: Irradiation of sphere-pac fuel and pellets of UO{sub 2−x} for americium breeding blanket concept

    Energy Technology Data Exchange (ETDEWEB)

    D' Agata, E., E-mail: elio.dagata@ec.europa.eu [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Hania, P.R. [Nuclear Research and Consultancy Group, P.O. Box 25, NL-1755 ZG Petten (Netherlands); Freis, D.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Bejaoui, S. [Commissariat à l’Energie Atomique et aux Energies Alternatives, DEN/DEC, F-13108 St. Paul lez Durance Cedex (France); Charpin, F.F.; Baas, P.J.; Okel, R.A.F.; Til, S. van [Nuclear Research and Consultancy Group, P.O. Box 25, NL-1755 ZG Petten (Netherlands); Lapetite, J.-M. [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Delage, F. [Commissariat à l’Energie Atomique et aux Energies Alternatives, DEN/DEC, F-13108 St. Paul lez Durance Cedex (France)

    2017-01-15

    Highlights: • MARINE is designed to check the behaviour of MABB sphere-pac concept. • MABB sphere-pac are compared with MABB pellet. • Swelling and helium release behaviour will be the main output of the experiment. • An experiment to check sphere-pac MADF fuel behaviour has been already performed. - Abstract: Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like {sup 241}Am is therefore an option for the reduction of radiotoxicity and heat production of waste packages to be stored in a repository. The MARINE irradiation experiment is the latest of a series of European experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS, MARIOS, SPHERE) performed in the High Flux Reactor (HFR). The MARINE experiment is developed and carried out in the framework of the collaborative research project PELGRIMM of the EURATOM 7th Framework Programme (FP7). During the past years of experimental works in the field of transmutation and tests of innovative nuclear fuels, the release or trapping of helium as well as swelling have been shown to be the key issues for the design of such kind of fuel both as drivers and even more for Am-bearing blanket targets (due to the higher Am contents). The main objective of the MARINE experiment is to study the in-pile behaviour of uranium oxide fuel containing 13% of americium and to compare the behaviour of sphere-pac versus pellet fuel, in particular the role of microstructure and temperature on fission gas release and He on fuel swelling. The MARINE experiment will be irradiated in 2016 in the HFR in Petten (The Netherlands) and is expected to be completed in spring 2017. This paper discusses the rationale and objective of the MARINE experiment and provides a general description of its design for which some innovative features have been adopted.

  9. The manufacture process and properties of (U, Gd)O2 burnable poisonous fuel pellets

    International Nuclear Information System (INIS)

    Yi Wei; Tang Yueming; Dai Shengping; Yang Youqing; Zuo Guoping; Wu Shihong; Gu Xiaofei; Gu Mingfei

    2006-03-01

    The main properties of important raw powder materials used in the (U, Gd)O 2 burnable poisonous fuel pellets production line of NPIC are presented. The powders included UO 2 , Gd 2 O 3 , (U, Gd) 3 O 8 and necessary additives, such as ammonium oxalate and zinc stearate. And the main properties of (U, Gd)O 2 burnable poisonous fuel pellets and the manufacture processes, such as ball-milling blending, granulation, pressing, sintering and grinding are also described. Moreover, the main effect of the process parameters controlled in the manufacture process have been discussed. (authors)

  10. Dry refabrication technology development of spent nuclear fuel

    International Nuclear Information System (INIS)

    Park, Geun Il; Lee, J. W.; Song, K. C.

    2012-04-01

    Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed

  11. Dry refabrication technology development of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Geun Il; Lee, J. W.; Song, K. C.; and others

    2012-04-15

    Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed.

  12. Study of UO2-10WT%Gd2O3 fuel pellets obtained by seeding method using AUC co-precipitation and mechanical mixing processes

    International Nuclear Information System (INIS)

    Lima, M.M.F.; Ferraz, W.B.A.; Santos, M.M. dos; Pinto, L.C.M.; Santos, A.

    2008-01-01

    The use of gadolinium and uranium mixed oxide as a nuclear fuel aims to obtain a fuel with a performance better than that of UO 2 fuel. In this work, seeding method was used to improve ionic diffusivity during sintering to produce high density pellets containing coarse grains by co-precipitation and mechanical mixing processes. Sintered UO 2 -10 wt% Gd 2 O 3 pellets were obtained using the reference processes with 2 wt% and 5 wt% UO 2 seeds with two granulometries, less than 20 μm and between 20 and 38 μm. Characterisation was carried out by chemical analysis, surface area, X-ray diffraction, SEM, WDS, image analysis, and densitometry. The seeding method using mechanical mixing process was more effective than the co-precipitation method. Furthermore, mechanical mixing process resulted in an increase in density of UO 2 -10wt% Gd 2 O 3 with seeds in relation to that of UO 2 -10wt% Gd 2 O 3 without seeds. (author)

  13. Manufacture of wood-pellets doubles. Biowatti Oy started a wood pellet plant in Turenki

    International Nuclear Information System (INIS)

    Rantanen, M.

    1999-01-01

    Wood pellets have many advantages compared to other fuels. It is longest processed biofuel with favorable energy content. It is simple to use, transport and store. Heating with wood pellets is cheaper than with light fuel oil, and approximately as cheap as utilization of heavy fuel oil, about 110 FIM/MWh. The taxable price of wood pellets is about 550 FIM/t. Stokers and American iron stoves are equally suitable for combustion of wood pellets. Chip fueled stokers are preferred in Finland, but they are also suitable for the combustion of wood pellets. Wood pellets is an environmentally friendly product, because it does not increase the CO 2 load in the atmosphere, and its sulfur and soot emissions are relatively small. The wood pelletizing plant of Biowatti Oy in Turenki was started in an old sugar mill. The Turenki sugar mill was chosen because the technology of the closed sugar factory was suitable for production of wood pellets nearly as such, and required only by slight modifications. A press, designed for briquetting of sugar beat clippings makes the pellets. The Turenki mill will double the volume of wood pellet manufacture in Finland during the next few years. At the start the annual wood pellet production will be 20 000 tons, but the environmental permit allows the production to be increased to 70 000 tons. At first the mill uses planing machine chips as a raw material in the production. It is the most suitable raw material, because it is already dry (moisture content 8-10%), and all it needs is milling and pelletizing. Another possible raw material is sawdust, which moisture content is higher than with planing machine chips. Most of the wood pellets produced are exported e.g. to Sweden, Denmark and Middle Europe. In Sweden there are over 10 000 single-family houses using wood pellets. Biowatti's largest customer is a power plant located in Stockholm, which combusts annually about 200 000 tons of wood pellets

  14. Nuclear fuels

    International Nuclear Information System (INIS)

    Gangwani, Saloni; Chakrabortty, Sumita

    2011-01-01

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

  15. Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Fourth semiannual report, July-December 1980

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.

    1981-03-01

    This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts have been developed for possible demonstration: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the scope of this program one of these concepts had to be selected for a large-scale demonstration in a commercial power reactor. The selection was made to demonstrate Zr-liner fuel and to include bundles which have liners prepared from either low oxygen sponge zirconium or of crystal bar zirconium. The demonstration is intended to include a total of 132 barrier bundles in the reload for Quad Cities Unit 2, Cycle 6. In the current report period changes in the nuclear design were made to respond to changes in the Energy Utilization Plan for Quad Cities Unit 2. Bundle designs were completed, and were licensed for use in a BWR/3. The core specific licensing will be done as part of the reload license for Quad Cities Unit 2, Cycle 6

  16. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye

    2013-01-01

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses

  17. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

  18. Techniques for chamfer and taper grinding of oxide fuel pellets (LWBR Development Program)

    International Nuclear Information System (INIS)

    Johnson, R.G.R.; Allison, J.W.

    1981-10-01

    Floor mounted centerless grinding machines were adapted for shaping the edges of cylindrical oxide fuel pellets for the Light Water Breeder Reactor (LWBR) by plunge grinding. Edge configurations consisted of chamfers, either 0.015 inch x 45 0 or 0.006 inch x 45 0 , or tapers 0.150 inch long x .0025 inch deep. Grinding was done by plunging the pellet against a shaped grinding wheel which ground both the diameter to the required size and shaped the edges of the pellet. Two plunges per pellet were required to complete the operation. Separate wheels were needed for grinding either a chamfer or a taper, the set up was adjustable to vary the size of the chamfer or taper as needed. The set up also had the flexibility to accommodate the multiple pellet lengths and diameters required by the LWBR design. Tight manufacturing tolerances in the chamfer and taper dimensions required the use of dimensional control charts and statistical sampling plans as process controls

  19. Fuel chemistry and pellet-clad interaction related to high burnup fuel. Proceedings of the technical committee

    International Nuclear Information System (INIS)

    2000-10-01

    The purpose of the meeting was to review new developments in clad failures. Major findings regarding the causes of clad failures are presented in this publication, with the main topics being fuel chemistry and fission product behaviour, swelling and pellet-cladding mechanical interaction, cladding failure mechanism at high burnup, thermal properties and fuel behaviour in off-normal conditions. This publication contains 17 individual presentations delivered at the meeting; each of them was indexed separately

  20. Disposition of excess plutonium using ''off-spec'' MOX pellets as a sintered ceramic waste form

    International Nuclear Information System (INIS)

    Armantrout, G.A.; Jardine, L.J.

    1996-02-01

    The authors describe a potential strategy for the disposition of excess weapons plutonium in a way that minimizes (1) technological risks, (2) implementation costs and completion schedules, and (3) requirements for constructing and operating new or duplicative Pu disposition facilities. This is accomplished by an optimized combination of (1) using existing nuclear power reactors to ''burn'' relatively pure excess Pu inventories as mixed oxide (MOX) fuel and (2) using the same MOX fuel fabrication facilities to fabricate contaminated or impure excess Pu inventories into an ''off-spec'' MOX solid ceramic waste form for geologic disposition. Diversion protection for the SCWF to meet the ''spent fuel standard'' introduced by the National Academy of Sciences can be achieved in at least three ways. (1) One can utilize the radiation field from defense high-level nuclear waste by first packaging the SCWF pellets in 2- to 4-L cans that are subsequently encapsulated in radioactive glass in the Defense Waste Processing Facility (DWPF) glass canisters (a ''can-in-canister'' approach). (2) One can add 137 Cs (recovered from defense wastes at Hanford and currently stored as CsCl in capsules) to an encapsulating matrix such as cement for the SCWF pellets in a small hot-cell facility and thus fabricate large monolithic forms. (3) The SCWF can be fabricated into reactor fuel-like pellets and placed in tubes similar to fuel assemblies, which can then be mixed in sealed repository containers with irradiated spent nuclear fuel for geologic disposition

  1. Pellet clad interaction analysis of AFA 3G fuel rod

    International Nuclear Information System (INIS)

    Liu Tong; Shen Caifen; Jiao Yongjun; Lu Huaquan; Zhou Zhou

    2002-01-01

    The author described Pellet Clad Interaction (PCI) analysis of AFA 3G fuel rod during condition II transients for GNPS 18-months alternating equilibrium cycles. It provided PCI technical limit, analytical methods and computer code used in the analyses of condition II transients and thermal-mechanical. Finally, given main calculation results and the conclusion for GNPS 18-months cycles

  2. Nuclear fuel, with emphasis on its utilization in pressurized water reactor

    International Nuclear Information System (INIS)

    Khazaneh, R.; Roshanzamir, M.

    1997-01-01

    Production processes of nuclear fuel on one hand and using nuclear fuels in reactors, particularly PWR Type reactors on the other hand is investigated. The first chapter reviews the relationship between fuel and reactors; The principals of reactor physics in relation with fuel are described shortly. The second chapter reviews uranium exploration and extraction as well as production of uranium concentrate and uranium dioxides. The third chapter is specified to the different procedures of uranium enrichment. In the fourth chapter, processing of uranium dioxide powder and fuel pellet is described. In the fifth chapter fabrication of fuel rod and fuel assemblies is explained thoroughly. The sixth chapter devoted to the different phenomena which occur ed in fuel structure and can during operational time of reactor; damage to fuel rods and developing theoretical models to describe these phenomena and analysis of fuel structure. The seventh chapter discusses how fuel rods are to be experimented during fabrication, operation and development of technology. The eighth chapter explains different fuels such as uranium compounds and mixed oxide fuel of uranium Gadolinium and uranium plutonium and the process of fabrication of zircaloy. In the tenth chapter, fuel reprocessing is investigated and the difficulties of developing this technology is referred

  3. Developing and testing a vertical sintering furnace for remote nuclear applications

    International Nuclear Information System (INIS)

    Nesbitt, J.F.; Ryer, C.M.

    1980-01-01

    Horizontal-type furnaces used to sinter fuel pellets on a production basis are large and thus impractical for remote applications. However, research has shown that vertical-type furnaces are adaptable for use and are cheaper to operate and maintain. In 1979, Pacific Northwest Laboratory, working under the auspices of the Department of Energy's Fuel Refabrication and Development (FRAD) Program, began developing an advanced concept for a remotely operated furnace designed specifically to sinter nuclear fuel pellets. The FRAD Program at PNL ended before the sintering of nuclear fuels could be completely verified. However during 1979, PNL performed a sufficient number and variety of tests to establish that nuclear fuel pellets can be sintered in a vertical furnace

  4. Axisym finite element code: modifications for pellet-cladding mechanical interaction analysis

    International Nuclear Information System (INIS)

    Engelman, G.P.

    1978-10-01

    Local strain concentrations in nuclear fuel rods are known to be potential sites for failure initiation. Assessment of such strain concentrations requires a two-dimensional analysis of stress and strain in both the fuel and the cladding during pellet-cladding mechanical interaction. To provide such a capability in the FRAP (Fuel Rod Analysis Program) codes, the AXISYM code (a small finite element program developed at the Idaho National Engineering Laboratory) was modified to perform a detailed fuel rod deformation analysis. This report describes the modifications which were made to the AXISYM code to adapt it for fuel rod analysis and presents comparisons made between the two-dimensional AXISYM code and the FRACAS-II code. FRACAS-II is the one-dimensional (generalized plane strain) fuel rod mechanical deformation subcode used in the FRAP codes. Predictions of these two codes should be comparable away from the fuel pellet free ends if the state of deformation at the pellet midplane is near that of generalized plane strain. The excellent agreement obtained in these comparisons checks both the correctness of the AXISYM code modifications as well as the validity of the assumption of generalized plane strain upon which the FRACAS-II subcode is based

  5. Technology and distribution of pellets. Experience about the European network on wood pellets

    International Nuclear Information System (INIS)

    Rapp, S.W.

    1999-01-01

    Wood pellets might become the most important alternative to fossil fuels in the near future. As a bio-fuel it has the following characteristics: heat value, min 4.7 kWh/kg; ash fraction less than 1.0 vol. %; humidity less than 10 vol. %; diameter (rod shaped) min 6 mm and volumetric weight about 650 kg/m 3 . About 2.1 t pellets substitute 1000 l fuel oil. Sweden and Austria have more than 15 year experience in using wood pellets, followed by Germany. They are an environmentally friendly alternative for private houses, for district heating plants and especially suitable for densely built-up and inhabited areas. Having high energy density they can be transported to the areas with high energy requirements. Among their advantages are: low humidity, easy transport and storage, can be produced by renewable raw materials and provide new local jobs, fit for renewable energy systems with closed cycle. Disadvantages include: relatively more expensive for private houses compared to oil and gas and necessity of two times larger storage space than oil. Wood pellets are produced by all kind of paper waste and wood wastes from industry. They are especially suitable for small boiler plants and the oil burner can be replaced by a pellet burner in the same boiler. The leading producer of wood pellets is Sweden, of pellet stoves - USA. Pellet stoves, pellet burners and pellet boilers both for private houses and for heating plants are manufactured also in Sweden, Denmark,Finland, Germany, Austria and Ireland

  6. Fuel assemblies

    International Nuclear Information System (INIS)

    Echigoya, Hironori; Nomata, Terumitsu.

    1983-01-01

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

  7. Development of TIG Welding System for a Nuclear Fuel Test Rig

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Changyoung; Ahn, Sungho; Hong, Jintae; Kim, Kahye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rig and rods. To manufacture the nuclear fuel test rig, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rig and rods jointing the various sensors and end caps on a fuel cladding tube, which is charged with fuel pellets and component parts. Thus, we designed and fabricated the precision welding system consisting of an orbital TIG welder, a low-pressure chamber, and a high-pressure chamber. Using this system, the performance tests were performed with the round and seal spot welds for each welding condition. This paper describes not only the contents for the fabrication of precision TIG welding system but also some results from weld tests using the low-pressure and high-pressure chambers to verify the performance of this system. The TIG welding system was developed to manufacture the nuclear fuel test rig and rods. It has been configured to be able to weld the nuclear fuel test rigs and rods by applying the TIG welder using a low-pressure chamber and a high-pressure chamber. The performance tests using this system were performed with the round and seal spot welds for the welding conditions. The soundness of the orbital TIG welding system was confirmed through performance tests in the low-pressure and high-pressure chambers.

  8. Development of TIG Welding System for a Nuclear Fuel Test Rig

    International Nuclear Information System (INIS)

    Joung, Changyoung; Ahn, Sungho; Hong, Jintae; Kim, Kahye

    2013-01-01

    The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rig and rods. To manufacture the nuclear fuel test rig, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rig and rods jointing the various sensors and end caps on a fuel cladding tube, which is charged with fuel pellets and component parts. Thus, we designed and fabricated the precision welding system consisting of an orbital TIG welder, a low-pressure chamber, and a high-pressure chamber. Using this system, the performance tests were performed with the round and seal spot welds for each welding condition. This paper describes not only the contents for the fabrication of precision TIG welding system but also some results from weld tests using the low-pressure and high-pressure chambers to verify the performance of this system. The TIG welding system was developed to manufacture the nuclear fuel test rig and rods. It has been configured to be able to weld the nuclear fuel test rigs and rods by applying the TIG welder using a low-pressure chamber and a high-pressure chamber. The performance tests using this system were performed with the round and seal spot welds for the welding conditions. The soundness of the orbital TIG welding system was confirmed through performance tests in the low-pressure and high-pressure chambers

  9. Numerical solution of the elastic non-axial contact between pellet and cladding of fuel rod in PWR

    International Nuclear Information System (INIS)

    Zymak, J.

    1987-08-01

    Elastic non-axial contacts between the pellet and the cladding of a fuel rod in a pressurized water reactor were calculated. The existence and the uniqueness of the solution were proved. The problem was approximated by the finite element method and quadratic programming was used for the solution. The results will be used in the solution of the probabilistic model of a fuel rod with non-axial pellets in a PWR. (author). 10 figs., 4 tabs., 10 refs

  10. Production of nuclear ceramic fuel for nuclear power plants at 'Ulba metallurgical plant' OSC

    International Nuclear Information System (INIS)

    Khadeev, V.G.

    2000-01-01

    The paper describes the flow-sheet of production of uranium dioxide powders and nuclear ceramic fuel pellets of them existing at the facility. 'UMP' OSC applies ADU extraction process of UO2 powders production. An indisputable success of the process is the possibility of use of the wide range of raw materials. Uranium hexafluoride, uranium oxides, uranium metal, uranium tetrafluoride, uranyl salts, uranium ore concentrates, all possible types of uranium-containing materials the processing of which by routine methods is difficult (ashes, scraps, etc.) are used as the raw materials. In addition, a reprocessed nuclear fuel can be used for fuel production. The quality of uranium dioxide powder produced does not depend on the type of uranium raw material used. High selectivity of extraction refining makes possible to obtain material with rather low impurities content that meets practically all specifications for uranium dioxide known to us. Ceramic and process features of uranium dioxide powders, namely, specific surface, bulk density, grain size and sinterability make possible to produce nuclear ceramic fuel with specified features. Quality of uranium dioxide powders produced by 'UMP' OSC was highly rated by General Electric company that is one of the leading companies from fuel manufactures in the USA market . It has certified 'UMP' OSC as its supplier. Currently, our company makes great efforts on establishing production of uranium dioxide powders with natural isotopes content for production of fuel for CANDU reactors. Trial lots of such powders are under tests at some companies manufacturing fuel for this type reactors in Canada, USA and Corea

  11. Induction plasma deposition technology for nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Jung, I. H.; Bae, K. K.; Lee, J. W.; Kim, T. K.; Yang, M. S.

    1998-01-01

    A study on induction plasma deposition with ceramic materials, yttria-stabilized-zirconia ZrO 2 -Y 2 O 3 (m.p. 2640 degree C), was conducted with a view of developing a new method for nuclear fuel fabrication. Before making dense pellets of more than 96%T.D., the spraying condition was optimized through the process parameters, such as chamber pressure, plasma plate power, powder spraying distance, sheath gas composition, probe position, particle size and powders of different morphology. The results with a 5mm thick deposit on rectangular planar graphite substrates showed a 97.11% theoretical density when the sheath gas flow rate was Ar/H 2 120/20 l/min, probe position 8cm, particle size -75 μm and spraying distance 22cm by AMDRY146 powder. The degree of influence of the main effects on density were powder morphology, particle size, sheath gas composition, plate power and spraying distance, in that order. Among the two parameter interactions, the sheath gas composition and chamber pressure affects density greatly. By using the multi-pellets mold of wheel type, the pellet density did not exceed 94%T.D., owing to the spraying angle

  12. ORNL pellet acceleration program

    International Nuclear Information System (INIS)

    Foster, C.A.; Milora, S.L.

    1978-01-01

    The Oak Ridge National Laboratory (ORNL) pellet fueling program is centered around developing equipment to accelerate large pellets of solidified hydrogen to high speeds. This equipment will be used to experimentally determine pellet-plasma interaction physics on contemporary tokamaks. The pellet experiments performed on the Oak Ridge Tokamak (ORMAK) indicated that much larger, faster pellets would be advantageous. In order to produce and accelerate pellets of the order of 1 to 6 mm in diameter, two apparatuses have been designed and are being constructed. The first will make H 2 pellets by extruding a filament of hydrogen and mechanically chopping it into pellets. The pellets formed will be mechanically accelerated with a high speed arbor to a speed of 950 m/sec. This technique may be extended to speeds up to 5000 m/sec, which makes it a prime candidate for a reactor fueling device. In the second technique, a hydrogen pellet will be formed, loaded into a miniature rifle, and accelerated by means of high pressure hydrogen gas. This technique should be capable of speeds of the order of 1000 m/sec. While this technique does not offer the high performance of the mechanical accelerator, its relative simplicity makes it attractive for near-term experiments

  13. Changing the energy climate: clean and green heat from grass biofuel pellets

    International Nuclear Information System (INIS)

    Jannasch, R.; Samson, R.; DeMaio, A.; Adams, T.; Ho Lem, C.

    2001-01-01

    Uncertain energy supplies and international agreements to reduce greenhouse gas (GHG) emissions have created unique opportunities for biofuel development. Pelleted fuels from warm season grasses such as switchgrass (Panicum virgatum) can be grown for $3-4/GigaJoule (GJ) with only minor emissions of CO 2 . Using close-coupled gasifer combustion technology, switchgrass fuel pellets emit 86%, 91%, 71% and 89% less CO 2 than electricity, heating oil, natural gas and propane, respectively. Every 100 ha of switchgrass converted into pellet form and used to displace fossil fuel for space-heating prevents the emission of 1000 tonnes of CO 2 . Heating an average Ontario house with a 90GJ heat demand costs $1213 with switchgrass pellets compared to $2234, $1664, $882 and $3251 with electricity, heating oil, natural gas and propane, respectively. An estimated 23.4 million acres of agricultural land in Canada could potentially be converted to perennial grass biofuel production. The depressed farm sector would benefit economically from energy farming. Low-grade heat energy derived from grass pellets could displace some of the 30,000 GigaWatt Hours of electricity currently used for home heating in Quebec, Ontario and Manitoba. Surplus electricity could be exported or used to replace nuclear or coal burning plants. Contrary to prevailing beliefs that reducing GHG emissions will raise societal energy costs, pelletized grass biofuels could provide consumers with less expensive and more GHG-friendly heating options than most fossil energy sources. If the political support and direction exist to implement the Kyoto Protocol as intended, grass pellets could well become a heating fuel of choice in North America. (author)

  14. Fabrication of the instrumented fuel rods for the 3-Pin Fuel Test Loop at HANARO

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Park, Sung Jae; Shin, Yoon Tag; Lee, Jong Min; Ahn, Sung Ho; Kim, Soo Sung; Kim, Bong Goo; Kim, Young Ki; Lee, Ki Hong; Kim, Kwan Hyun

    2008-09-01

    The 3-Pin Fuel Test Loop(hereinafter referred to as the '3-Pin FTL') facility has been installed at HANARO(High-flux Advanced Neutron Application Reactor) and the 3-Pin FTL is under a test operation. The purpose of this report is to fabricate the instrumented fuel rods for the 3-Pin FTL. The fabrication of these fuel rods was based on experiences and technologies of the instrumented fuel rods for an irradiation fuel capsule. The three instrumented fuel rods of the 3-Pin FTL have been designed. The one fuel rod(180 .deg. ) was designed to measure the centerline temperature of the nuclear fuels and the internal pressure of the fuel rod, and others(60 .deg. and 300 .deg. ) were designed to measure the centerline temperature of the fuel pellets. The claddings were made of the reference material 1 and 2 and new material 1 and 2. And nuclear fuel was used UO 2 (2.0w/o) pellet type with large grain and standard grain. The major procedures of fabrication are followings: (1) the assembling and weld of fuel rods with the pellet mockups and the sensor mockups for the qualification tests, (2) the qualification tests(dimension measurements, tensile tests, metallography examinations and helium leak tests) of weld, (3) the assembling and weld of instrumented fuel rods with the nuclear pellets and the sensors for the irradiation test, and (4) the qualification tests(the helium leak test, the dimensional measurement, electric resistance measurements of sensors) of test fuel rods. Satisfactory results were obtained for all the qualification tests of the instrumented fuel rods for the 3-Pin FTL. Therefore the three instrumented fuel rods for the 3-Pin FTL have been fabricated successfully. These will be installed in the In-Pile Section of 3-Pin FTL. And the irradiation test of these fuel rods is planned from the early next year for about 3 years at HANARO

  15. Determination of fission gas release of spent nuclear fuel in puncturing test and in leaching experiments under anoxic conditions

    Energy Technology Data Exchange (ETDEWEB)

    González-Robles, E., E-mail: ernesto.gonzalez-robles@kit.edu [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Metz, V. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Wegen, D.H. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Herm, M. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Papaioannou, D. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Bohnert, E. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Gretter, R. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Müller, N. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Nasyrow, R.; Weerd, W. de; Wiss, T. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Kienzler, B. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany)

    2016-10-15

    During reactor operation the fission gases Kr and Xe are formed within the UO{sub 2} matrix of nuclear fuel. Their quantification is important to evaluate their impact on critical parameters regarding the fuel behaviour during irradiation and (long-term) interim storage, such as internal pressure of the fuel rod and fuel swelling. Moreover the content of Kr and Xe in the plenum of a fuel rod and their content in the UO{sub 2} fuel itself are widely used as indicators for the release properties of {sup 129}I, {sup 137}Cs, and other safety relevant radionuclides with respect to final disposal of spent nuclear fuel. The present study deals with the fission gas release from spent nuclear fuel exposed to simulated groundwater in comparison with the fission gas previously released to the fuel rod plenum during irradiation in reactor. In a unique approach we determined both the Kr and Xe inventories in the plenum by means of a puncturing test and in leaching experiments with a cladded fuel pellet and fuel fragments in bicarbonate water under 3.2 bar H{sub 2} overpressure. The fractional inventory of the fission gases released during irradiation into the plenum was (8.3 ± 0.9) %. The fraction of inventory of fission gases released during the leaching experiments was (17 ± 2) % after 333 days of leaching of the cladded pellet and (25 ± 2) % after 447 days of leaching of the fuel fragments, respectively. The relatively high release of fission gases in the experiment with fuel fragments was caused by the increased accessibility of water to the Kr and Xe occluded in the fuel.

  16. Results of REIMEP '89 UO2 pellet

    International Nuclear Information System (INIS)

    Mayer, K.; Alonso, A.; Bievre, P. de; Lycke, W.; Bolle, W. de; Gallet, M.; Hendrickx, F.

    1991-01-01

    The interest in the safeguards of fissile material focuses on a limited number of compounds which play key roles in the nuclear fuel cycle. Amongst these materials Uranium Dioxide pellets are of considerable importance as they enter the reactors in order to generate energy. In LWR's pellets with an initial 235 U content of about 3 mass % are used, whereas natural or depleted material is applied for the breeding zone in FBR's. The 89/90 round o REIMEP covered Uranium materials with 235 U abundances in the range of natural or depleted material. UO 2 pellets were distributed to 21 laboratories for analysis. The participating laboratories were asked to determine the Uranium content and the isotopic composition of the material. The results reported by the participants are presented as graphs thus giving a picture of the state-of-the-practice

  17. Emission of Metals from Pelletized and Uncompressed Biomass Fuels Combustion in Rural Household Stoves in China

    Science.gov (United States)

    Zhang, Wei; Tong, Yindong; Wang, Huanhuan; Chen, Long; Ou, Langbo; Wang, Xuejun; Liu, Guohua; Zhu, Yan

    2014-07-01

    Effort of reducing CO2 emissions in developing countries may require an increasing utilization of biomass fuels. Biomass pellets seem well-suited for residential biomass markets. However, there is limited quantitative information on pollutant emissions from biomass pellets burning, especially those measured in real applications. In this study, biomass pellets and raw biomass fuels were burned in a pellet burner and a conventional stove respectively, in rural households, and metal emissions were determined. Results showed that the emission factors (EFs) ranged 3.20-5.57 (Pb), 5.20-7.58 (Cu), 0.11-0.23 (Cd), 12.67-39.00 (As), 0.59-1.31 mg/kg (Ni) for pellets, and 0.73-1.34 (Pb), 0.92-4.48 (Cu), 0.08-0.14 (Cd), 7.29-13.22 (As), 0.28-0.62 (Ni) mg/kg for raw biomass. For unit energy delivered to cooking vessels, the EFs ranged 0.42-0.77 (Pb), 0.79-1.16 (Cu), 0.01-0.03 (Cd), 1.93-5.09 (As), 0.08-0.19 mg/MJ (Ni) for pellets, and 0.30-0.56 (Pb), 0.41-1.86 (Cu), 0.04-0.06 (Cd), 3.25-5.49 (As), 0.12-0.26 (Ni) mg/MJ for raw biomass. This study found that moisture, volatile matter and modified combustion efficiency were the important factors affecting metal emissions. Comparisons of the mass-based and task-based EFs found that biomass pellets produced higher metal emissions than the same amount of raw biomass. However, metal emissions from pellets were not higher in terms of unit energy delivered.

  18. Method of manufacturing sintered nuclear fuel

    International Nuclear Information System (INIS)

    Watarumi, Kazutoshi.

    1984-01-01

    Purpose: To obtain composite pellets with an improved strength. Method: A core mainly composed of fuel materials is previously prepared, embedded into the central portion of a pellet, silted therearound with cladding material, and then pressmolded and sintered. For instance, a rugby-ball like core body with the maximum outer diameter of 6 mm and the height of 6 mm is made by compressive molding with uranium dioxide powder, then coating material comprising the same powder incorporated with 0.1 % by weight of SiC fibers is filled around the core body, which is molded into a composite pellet by means of pressing and then sintered at 1600 0 C, to obtain a sintered pellet of 93.5 % theoretical density. As the result of the compression test for the pellet, it showed a strength greater by 15 % than that of the similar mono-layer pellet. (Kamimura, M.)

  19. Fuel performance-REP, Seminars on nuclear fuel performance based on basic underlining phenomena, proceedings

    International Nuclear Information System (INIS)

    2008-01-01

    Description: The need for further improving the understanding of basic phenomena underlying nuclear fuel behaviour has been recognised both by fuel vendors, experts in fuel research in the different laboratories and committees and working groups coordinating international activities. The OECD/NEA Nuclear Science Committee has established an Experts Group addressing this issue. This has led to establishing an International Fuel Performance Experiments Database (IFPE) that should help model evaluation and validation. Many years ago the IAEA established an International Working Group on Fuel Performance and Technology (IWGFPT) that led to the FUMEX-I and FUMEX-II (Fuel Modelling Exercise) which has had an important impact on code improvements. Both international organisations, with the support of national organisations, co-operate in establishing and maintaining the Database and to build confidence in the predictive power of the models through international comparison exercises. But above all the different parties have agreed that seminars focussed on specific phenomena would be beneficial to exchange current knowledge, identify outstanding problems and agree on common action that would lead to improved understanding of the phenomena. A series of three seminars has been initiated by the Commissariat a l'Energie Atomique (CEA), Electricite de France (EdF), Framatome and Cogema under the aegis of the OECD/NEA and the IAEA. 1. Thermal Performance of High Burn-Up LWR Fuel at Cadarache, France, from 3 to 6 of March 1998. Thermal performance occupies the most important aspect of the fuel performance modelling. Not only is it extremely important from a safety point of view, but also many of the material properties of interest and behaviour, such as transport properties like fuel creep and fission gas release are thermally activated processes. Thus, in order to model these processes correctly, it is critical to calculate temperatures and their distribution as accurately as

  20. The US pellet market

    International Nuclear Information System (INIS)

    Elliot, S.

    2007-01-01

    Bear Mountain is the largest producer of pellets, firelogs, animal beddings, and barbecue pellets in Western United States. The company's branded products are sold directly to more than 400 retail dealers. This presentation included a series of graphs depicting Bear Mountain's USA pellet sales in tons from 2002 to 2007; truckloads to various distribution areas; pellet stoves and insert units shipped from 1998 to 2006; and hearth appliance shipments from 1998 to 2006. It was noted that in the United States, 98 per cent of the pellets sold come in 40 pound bags and are delivered to retailers by truck. Space is needed for inventory purposes, as each customer may use 2 to 4 tons. The pellets are used in small ash capacity room heaters. The pellet producers buy sawdust from area mills. It was noted that the soft housing market combined with competition for pulp and paper has pinched the supply of pellets. Pellets were in short supply in the west coast during the winter of 2006-2007 and in eastern United States during the winters of 2004-2005 and 2005-2006, indicating that summer production of pellets is required in order to meet winter demand. The key demand factors for pellets include stove sales; pellet pricing; pricing of other fuels; and, weather. The key supply factors for pellets include availability of sawdust; logistics; competition; and cost. The greatest challenge facing pellet producers is the high cost of freight. It was concluded that 2008 will be another year of uncertainty for pellet producers, due to the abundant supply of pellets in the east and midwest, and stabilized alternative fuel pricing. tabs., figs

  1. Fuel-cladding chemical interaction

    International Nuclear Information System (INIS)

    Gueneau, C.; Piron, J.P.; Dumas, J.C.; Bouineau, V.; Iglesias, F.C.; Lewis, B.J.

    2015-01-01

    The chemistry of the nuclear fuel is very complex. Its chemical composition changes with time due to the formation of fission products and depends on the temperature level history within the fuel pellet and the clad during operation. Firstly, in thermal reactors, zircaloy oxidation from reaction with UO 2 fuel under high-temperature conditions will be addressed. Then other fuel-cladding interaction phenomena occurring in fast reactors will be described. Large thermal gradients existing between the centre and the periphery of the pellet induce the radial redistribution of the fuel constituents. The fuel pellet can react with the clad by different corrosion processes which can involve actinide and/or fission product transport via gas, liquid or/and solid phases. All these phenomena are briefly described in the case of different kinds of fuels (oxide, carbide, nitride, metallic) to be used in fast reactors. The way these phenomena are taken into account in fuel performance codes is presented. (authors)

  2. Pellet bed reactor for nuclear propelled vehicles: Part 2: Missions and vehicle integration trades

    International Nuclear Information System (INIS)

    Haloulakos, V.E.

    1991-01-01

    Mission and vehicle integration tradeoffs involving the use of the pellet bed reactor (PBR) for nuclear powered vehicles is discussed, with much of the information being given in viewgraph form. Information is given on propellant tank geometries, shield weight requirements for conventional tank configurations, effective specific impulse, radiation mapping, radiation dose rate after shutdown, space transfer vehicle design data, a Mars mission summary, sample pellet bed nuclear orbit transfer vehicle mass breakdown, and payload fraction vs. velocity increment

  3. Pellet bed reactor for nuclear propelled vehicles: Part 2: Missions and vehicle integration trades

    Science.gov (United States)

    Haloulakos, V. E.

    1991-01-01

    Mission and vehicle integration tradeoffs involving the use of the pellet bed reactor (PBR) for nuclear powered vehicles is discussed, with much of the information being given in viewgraph form. Information is given on propellant tank geometries, shield weight requirements for conventional tank configurations, effective specific impulse, radiation mapping, radiation dose rate after shutdown, space transfer vehicle design data, a Mars mission summary, sample pellet bed nuclear orbit transfer vehicle mass breakdown, and payload fraction vs. velocity increment.

  4. A statistical approach to nuclear fuel design and performance

    Science.gov (United States)

    Cunning, Travis Andrew

    As CANDU fuel failures can have significant economic and operational consequences on the Canadian nuclear power industry, it is essential that factors impacting fuel performance are adequately understood. Current industrial practice relies on deterministic safety analysis and the highly conservative "limit of operating envelope" approach, where all parameters are assumed to be at their limits simultaneously. This results in a conservative prediction of event consequences with little consideration given to the high quality and precision of current manufacturing processes. This study employs a novel approach to the prediction of CANDU fuel reliability. Probability distributions are fitted to actual fuel manufacturing datasets provided by Cameco Fuel Manufacturing, Inc. They are used to form input for two industry-standard fuel performance codes: ELESTRES for the steady-state case and ELOCA for the transient case---a hypothesized 80% reactor outlet header break loss of coolant accident. Using a Monte Carlo technique for input generation, 105 independent trials are conducted and probability distributions are fitted to key model output quantities. Comparing model output against recognized industrial acceptance criteria, no fuel failures are predicted for either case. Output distributions are well removed from failure limit values, implying that margin exists in current fuel manufacturing and design. To validate the results and attempt to reduce the simulation burden of the methodology, two dimensional reduction methods are assessed. Using just 36 trials, both methods are able to produce output distributions that agree strongly with those obtained via the brute-force Monte Carlo method, often to a relative discrepancy of less than 0.3% when predicting the first statistical moment, and a relative discrepancy of less than 5% when predicting the second statistical moment. In terms of global sensitivity, pellet density proves to have the greatest impact on fuel performance

  5. A method for determining an effective porosity correction factor for thermal conductivity in fast reactor uranium-plutonium oxide fuel pellets

    International Nuclear Information System (INIS)

    Inoue, Masaki; Abe, Kazuyuki; Sato, Isamu

    2000-01-01

    A reliable method has been developed for determining an effective porosity correction factor for calculating a realistic thermal conductivity for fast reactor uranium-plutonium (mixed) oxide fuel pellets. By using image analysis of the ceramographs of transverse sections of mixed-oxide fuel pellets, the fuel morphology could be classified into two basic types. One is a 'two-phase' type that consists of small pores dispersed in the fuel matrix. The other is a 'three-phase' type that has large pores in addition to the small pores dispersed in the fuel matrix. The pore sizes are divided into two categories, large and small, at the 30 μm area equivalent diameter. These classifications lead to an equation for calculating an effective porosity correction factor by accounting for the small and large pore volume fractions and coefficients. This new analytical method for determining the effective porosity correction factor for calculating the realistic thermal conductivity of mixed-oxide fuel was also experimentally confirmed for high-, medium- and low-density fuel pellets

  6. Trend of fuel for light water reactors and development hereafter

    International Nuclear Information System (INIS)

    Ichikawa, Michio; Maru, Akira; Shimoshige, Takanori

    1993-01-01

    Recently, the heightening of fuel burnup has been actively advanced internationally. Its degree is different according to the policy and the economical factors in respective countries. The extension of the period of operation cycle urges high burnup in view of economy. The circumstances in USA, Europe and Japan are explained. The corrosion of zircaloy cladding is the factor of limiting fuel life. The state of corrosion in reactors is different in BWRs and PWRs, and both cases are explained. The emission of FP gas from pellets to fuel rods raises the internal pressure of the fuel rods, and affects the gap conductance between pellets and cladding tubes. In the fuel for LWRs, plutonium is formed locally and burns in pellet rim part. This rim effect is discussed. The irradiation growth of fuel rods, creep down and pellet-cladding interaction are explained. The MOX fuel for LWRs and the trend of development of new type fuel are reported. The fuel for BWRs of Hitachi Ltd. and Toshiba Corp. and Nuclear Fuel Industries Ltd., the fuel for PWRs of Mitsubishi Heavy Industries Ltd. and Nuclear fuel Industries Ltd., and the recent development of the fuel cladding tubes for LWRs are described. (K.I.)

  7. Nuclear fuel

    International Nuclear Information System (INIS)

    D Hondt, P.

    1998-01-01

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

  8. Microwave based oxidation process for recycling the off-specification (U,Pu)O{sub 2} fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Singh, G., E-mail: gitendars@barctara.gov.in [Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, 401 502 (India); Khot, P.M. [Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, 401 502 (India); Kumar, Pradeep [Integrated Fuel Fabrication Facility (IFFF), Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Bhatt, R.B.; Behere, P.G.; Afzal, Mohd [Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, 401 502 (India)

    2017-02-15

    This paper reports development of a process named MicroWave Direct Oxidation (MWDO) for recycling the off-specification (U,Pu)O{sub 2} mixed oxide (MOX) fuel pellets generated during fabrication of typical fast reactor fuels. MWDO is a two-stage, single-cycle process based on oxidative pulverisation of pellets using 2450 MHz microwave. The powder sinterability was evaluated by bulk density and BET specific surface area. The oxidised powders were analyzed for phases using XRD and stoichiometry by thermogravimetry. The sinterability was significantly enhanced by carrying out oxidation in higher oxygen partial pressure and by subjecting MOX to multiple micronisation-oxidation cycles. After three cycles, the recycled powder from (U,28%Pu)O{sub 2} resulted surface area >3 m{sup 2}/g and 100% re-used for MOX fabrication. The flow sheet was developed for maximum utilization of recycled powder describable by a parameter called Scrap Recycling Ratio (SRR). The process demonstrates smaller processing cycle, better powder properties and higher oxidative pulverisation over conventional method. - Highlights: • A process for recycling the off-specification (U,Pu)O{sub 2} sintered fuel pellets of fast reactors was demonstrated. • The method is a two-stage, single cycle process based on oxidative pulverization of MOX pellets using 2450 MHz microwave. • The process demonstrated utilization of recycled powder with SRR of 1.

  9. Reductions in emissions of carbonaceous particulate matter and polycyclic aromatic hydrocarbons from combustion of biomass pellets in comparison with raw fuel burning.

    Science.gov (United States)

    Shen, Guofeng; Tao, Shu; Wei, Siye; Zhang, Yanyan; Wang, Rong; Wang, Bin; Li, Wei; Shen, Huizhong; Huang, Ye; Chen, Yuanchen; Chen, Han; Yang, Yifeng; Wang, Wei; Wei, Wen; Wang, Xilong; Liu, Wenxing; Wang, Xuejun; Masse Simonich, Staci L y

    2012-06-05

    Biomass pellets are emerging as a cleaner alternative to traditional biomass fuels. The potential benefits of using biomass pellets include improving energy utilization efficiency and reducing emissions of air pollutants. To assess the environmental, climate, and health significance of replacing traditional fuels with biomass pellets, it is critical to measure the emission factors (EFs) of various pollutants from pellet burning. However, only a few field measurements have been conducted on the emissions of carbon monoxide (CO), particulate matter (PM), and polycyclic aromatic hydrocarbons (PAHs) from the combustion of pellets. In this study, pine wood and corn straw pellets were burned in a pellet burner (2.6 kW), and the EFs of CO, organic carbon, elemental carbon, PM, and PAHs (EF(CO), EF(OC), EF(EC), EF(PM), and EF(PAH)) were determined. The average EF(CO), EF(OC), EF(EC), and EF(PM) were 1520 ± 1170, 8.68 ± 11.4, 11.2 ± 8.7, and 188 ± 87 mg/MJ for corn straw pellets and 266 ± 137, 5.74 ± 7.17, 2.02 ± 1.57, and 71.0 ± 54.0 mg/MJ for pine wood pellets, respectively. Total carbonaceous carbon constituted 8 to 14% of the PM mass emitted. The measured values of EF(PAH) for the two pellets were 1.02 ± 0.64 and 0.506 ± 0.360 mg/MJ, respectively. The secondary side air supply in the pellet burner did not change the EFs of most pollutants significantly (p > 0.05). The only exceptions were EF(OC) and EF(PM) for pine wood pellets because of reduced combustion temperatures with the increased air supply. In comparison with EFs for the raw pine wood and corn straw, EF(CO), EF(OC), EF(EC), and EF(PM) for pellets were significantly lower than those for raw fuels (p 0.05). Based on the measured EFs and thermal efficiencies, it was estimated that 95, 98, 98, 88, and 71% reductions in the total emissions of CO, OC, EC, PM, and PAHs could be achieved by replacing the raw biomass fuels combusted in traditional cooking stoves with pellets burned in modern pellet burners.

  10. Domestic nuclear fuels supply: possibility of an independent technology

    International Nuclear Information System (INIS)

    Cirimello, R.O.

    1982-01-01

    After considering the different energy sources, their consumption and their respective periods of exploitation, technological considerations in the nuclear fuel field are made. The main subject is the Domestic Supply Project of Embalse Fuel (CANDU type). The different aspects which had to be developed during the realization of this project still under progress, and which are fundamental for the command of the technology, are described: 1) Qualification of the produced fuel elements: fuel elements' characteristics; the reactors' operating parameters, and the prototype fuel elements' characteristics; 2) Development of materials and/or suppliers: the obtainment of UO 2 and its physical properties are considered, as well as those of Zircaloy-4, the development of suppliers and the respective developments for the obtainment of materials such as beryllium, helium and colloidal graphite; 3) Processes development; the following processes are studied and defined: UO 2 pellets fabrication with UO 2 granulated powder; beryllium coating under vaccum; and induction brazing of bearing pads and spacers, end cap and end plate resistance welding and stamping of Zircaloy components, graphite-coating of cladding's internal face; 4) Development of special production equipments; automatic equipment for end cap-to-cladding resistance welding among others. The need for a specific program of quality assurance for nuclear fuels supply is emphasized and the basic criteria are established. The IAEA's quality asssurance requirements are also analyzed. (M.E.L.) [es

  11. Completion of UO2 pellets production and fuel rods load for the RA-8 critical facility

    International Nuclear Information System (INIS)

    Marajofsky, Adolfo; Perez, Lidia E.; Thern, Gerardo G.; Altamirano, Jorge S.; Benitez, Ana M.; Cardenas, Hugo R.; Becerra, Fabian A.; Perez, Aldo E.; Fuente, Mariano de la

    1999-01-01

    The Advanced Fuels Division produced fuel pellets of 235 U with 1.8% and 3.6% enrichment and Zry-4 cladding loads for the RA-8 reactor at Pilcaniyeu Technological Unit. For economical and availability reasons, the powder acquired was initially UO 2 with 3.4% enrichment in 235 U, therefore the 235 U powder with 1.8% enrichment was produced by mechanical mixture. The production of fuel pellets for both enrichments was carried out by cold pressing and sintering processes in reducing atmosphere. The load of Zry-4 claddings was performed manually. The production stages can be divided into setup, qualification and production. This production allows not only to fulfill satisfactorily the new fuel rods supply for the RA-8 reactor but also to count with a new equipment and skilled personnel as well as to meet quality and assurance control methods for future pilot-scale production and even new fuel elements production. (author)

  12. Demonstration of fuel resistant to pellet-cladding interaction. Phase I. Final report

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.

    1979-03-01

    This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel, and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress, and reactive fission products during reactor service. This is the final report for PHASE 1 of this program. Support tests have shown that the barrier fuel resists PCI far better than does the conventional Zircaloy-clad fuel. Power ramp tests thus far have shown good PCI resistance for Cu-barrier fuel at burnup > 12 MWd/kg-U and for Zr-liner fuel > 16 MWd/kg-U. The program calls for continued testing to still higher burnup levels in PHASE 2

  13. Characterizing high-temperature deformation of internally heated nuclear fuel element simulators

    Energy Technology Data Exchange (ETDEWEB)

    Belov, A.I.; Fong, R.W.L.; Leitch, B.W.; Nitheanandan, T.; Williams, A., E-mail: alexander.belov@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The sag behaviour of a simulated nuclear fuel element during high-temperature transients has been investigated in an experiment utilizing an internal indirect heating method. The major motivation of the experiment was to improve understanding of the dominant mechanisms underlying the element thermo-mechanical response under loss-of-coolant accident conditions and to obtain accurate experimental data to support development of 3-D computational fuel element models. The experiment was conducted using an electrically heated CANDU fuel element simulator. Three consecutive thermal cycles with peak temperatures up to ≈1000 {sup o}C were applied to the element. The element sag deflections and sheath temperatures were measured. On heating up to 600 {sup o}C, only minor lateral deflections of the element were observed. Further heating to above 700 {sup o}C resulted in an element multi-rate creep and significant permanent bow. Post-test visual and X-ray examinations revealed a pronounced necking of the sheath at the pellet-to-pellet interface locations. A wall thickness reduction was detected in the necked region that is interpreted as a sheath longitudinal strain localization effect. The sheath cross-sectioning showed signs of a 'hard' pellet-cladding interaction due to the applied cycles. A 3-D model of the experiment was generated using the ANSYS finite element code. As a fully coupled thermal mechanical simulation is computationally expensive, it was deemed sufficient to use the measured sheath temperatures as a boundary condition, and thus an uncoupled mechanical simulation only was conducted. The ANSYS simulation results match the experiment sag observations well up to the point at which the fuel element started cooling down. (author)

  14. Current status of the tasks performed by the Nuclear Fuel Department for the CAREM project

    International Nuclear Information System (INIS)

    Kaufmann, Federico; Perez, Lidia E.; Perez, Aaldo; Marchi, Daniel E.

    2009-01-01

    CAREM Project required to the Nuclear Fuels Department to perform the necessary tasks to define the powder characteristics and natural UO 2 -Gd 2 O 3 , UO 2 -1.8 and 3.1 % enrichment fuel pellets manufacturing and control parameters. To start with these tasks it was necessary to hire and train staff, begin the licensing process of the facilities, update infrastructure works and equipment of existing facilities. A brief description of the tasks performed is shown. (author)

  15. Quality control in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Abdelhalim, A.S.; Elsayed, A.A.; Shaaban, H.I.

    1988-01-01

    The department of metallurgy, NRC Inchass is embarking on a programme of on a laboratory scale, fuel pins containing uranium dioxide pellets are going to be produced. The department is making use of the expertise and equipment at present available and is going to utilize the new fuel pin fabrication unit which would be shortly in operation. The fabrication and testing of uranium dioxide pellets then gradually adapt them and develop, a national know how in this field. This would also involve building up of indigenous experience through proper training of qualified personnel. That are applied to ensure quality of U o 2 pellets, the techniques implemented, the equipment used and the specifications of the equipment presently available. The following parameters are subject to quality control tests: density. O/U ration, hydrogen content, microstructure, each property will be discussed, measurements related to U o 2 powders, including flow ability, bulk density, O/U ratio, bet surface area and water content will be critically discussed. Relevant tests to ensure Q C of pellets are reviewed. These include surface integrity, density, dimensions, microstructure.4 fig., 1 tab

  16. Simulation of pellet-cladding interaction with the Pleiades fuel performance software environment

    International Nuclear Information System (INIS)

    Michel, B.; Nonon, C.; Sercombe, J.; Michel, F.; Marelle, V.

    2013-01-01

    This paper focuses on the PLEIADES fuel performance software environment and its application to the modeling of pellet-cladding interaction (PCI). The PLEIADES platform has been under development for 10 yr; a unified software environment, including the multidimensional finite element solver CAST3M, has been used to develop eight computation schemes now under operation. Among the latter, the ALCYONE application is devoted to pressurized water reactor fuel rod behavior. This application provides a three-dimensional (3-D) model for a detailed analysis of fuel element behavior and enables validation through comparing simulation and post-irradiation examination results (cladding residual diameter and ridges, dishing filling, pellet cracking, etc.). These last years the 3-D computation scheme of the ALCYONE application has been enriched with a complete set of physical models to take into account thermomechanical and chemical-physical behavior of the fuel element under irradiation. These models have been validated through the ALCYONE application on a large experimental database composed of approximately 400 study cases. The strong point of the ALCYONE application concerns the local approach of stress-corrosion-cracking rupture under PCI, which can be computed with the 3-D finite element solver. Further developments for PCI modeling in the PLEIADES platform are devoted to a new mesh refinement method for assessing stress-and-strain concentration (multigrid technique) and a new component for assessing fission product chemical recombination. (authors)

  17. Wood pellet seminar

    International Nuclear Information System (INIS)

    Aarniala, M.; Puhakka, A.

    2001-01-01

    The objective of the wood pellet seminar, arranged by OPET Finland and North Karelia Polytechnic, was to deliver information on wood pellets, pellet burners and boilers, heating systems and building, as well as on the activities of wood energy advisors. The first day of the seminar consisted of presentations of equipment and products, and of advisory desks for builders. The second day of the seminar consisted of presentations held by wood pellet experts. Pellet markets, the economy and production, the development of the pellet markets and their problems (in Austria), the economy of heating of real estates by different fuel alternatives, the production, delivery and marketing of wood pellets, the utilization of wood pellet in different utilization sites, the use of wood pellets in detached houses, pellet burners and fireplaces, and conversion of communal real estate houses to use wood pellets were discussed in the presentations. The presentations held in the third day discussed the utilization of wood pellets in power plants, the regional promotion of the production and the use of pellets. The seminar consisted also of visits to pellet manufacturing plant and two pellet burning heating plants

  18. Electrochemical reduction of CerMet fuels for transmutation using surrogate CeO2-Mo pellets

    Science.gov (United States)

    Claux, B.; Souček, P.; Malmbeck, R.; Rodrigues, A.; Glatz, J.-P.

    2017-08-01

    One of the concepts chosen for the transmutation of minor actinides in Accelerator Driven Systems or fast reactors proposes the use of fuels and targets containing minor actinides oxides embedded in an inert matrix either composed of molybdenum metal (CerMet fuel) or of ceramic magnesium oxide (CerCer fuel). Since the sufficient transmutation cannot be achieved in a single step, it requires multi-recycling of the fuel including recovery of the not transmuted minor actinides. In the present work, a pyrochemical process for treatment of Mo metal inert matrix based CerMet fuels is studied, particularly the electroreduction in molten chloride salt as a head-end step required prior the main separation process. At the initial stage, different inactive pellets simulating the fuel containing CeO2 as minor actinide surrogates were examined. The main studied parameters of the process efficiency were the porosity and composition of the pellets and the process parameters as current density and passed charge. The results indicated the feasibility of the process, gave insight into its limiting parameters and defined the parameters for the future experiment on minor actinide containing material.

  19. Simulating Dynamic Fracture in Oxide Fuel Pellets Using Cohesive Zone Models

    Energy Technology Data Exchange (ETDEWEB)

    R. L. Williamson

    2009-08-01

    It is well known that oxide fuels crack during the first rise to power, with continued fracture occurring during steady operation and especially during power ramps or accidental transients. Fractures have a very strong influence on the stress state in the fuel which, in turn, drives critical phenomena such as fission gas release, fuel creep, and eventual fuel/clad mechanical interaction. Recently, interest has been expressed in discrete fracture methods, such as the cohesive zone approach. Such models are attractive from a mechanistic and physical standpoint, since they reflect the localized nature of cracking. The precise locations where fractures initiate, as well as the crack evolution characteristics, are determined as part of the solution. This paper explores the use of finite element cohesive zone concepts to predict dynamic crack behavior in oxide fuel pellets during power-up, steady operation, and power ramping. The aim of this work is first to provide an assessment of cohesive zone models for application to fuel cracking and explore important numerical issues associated with this fracture approach. A further objective is to provide basic insight into where and when cracks form, how they interact, and how cracking effects the stress field in a fuel pellet. The ABAQUS commercial finite element code, which includes powerful cohesive zone capabilities, was used for this study. Fully-coupled thermo-mechanical behavior is employed, including the effects of thermal expansion, swelling due to solid and gaseous fission products, and thermal creep. Crack initiation is determined by a temperature-dependent maximum stress criterion, based on measured fracture strengths for UO2. Damage evolution is governed by a traction-separation relation, calibrated to data from temperature and burn-up dependent fracture toughness measurements. Numerical models are first developed in 2D based on both axisymmetric (to explore axial cracking) and plane strain (to explore radial

  20. Fuel Cell Electrodes Based on Carbon Nanotube/Metallic Nanoparticles Hybrids Formed on Porous Stainless Steel Pellets

    Directory of Open Access Journals (Sweden)

    S. M. Khantimerov

    2013-01-01

    Full Text Available The preparation of carbon nanotube/metallic particle hybrids using pressed porous stainless steel pellets as a substrate is described. The catalytic growth of carbon nanotubes was carried out by CVD on a nickel catalyst obtained by impregnation of pellets with a highly dispersive colloidal solution of nickel acetate tetrahydrate in ethanol. Granular polyethylene was used as the carbon source. Metallic particles were deposited by thermal evaporation of Pt and Ag using pellets with grown carbon nanotubes as a base. The use of such composites as fuel cell electrodes is discussed.

  1. Development of Innovative Accident Tolerant High Thermal Conductivity UO2-Diamond Composite Fuel Pellets

    Energy Technology Data Exchange (ETDEWEB)

    Tulenko, James [Univ. of Florida, Gainesville, FL (United States); Subhash, Ghatu [Univ. of Florida, Gainesville, FL (United States)

    2016-01-01

    The University of Florida (UF) evaluated a composite fuel consisting of UO2 powder mixed with diamond micro particles as a candidate as an accident-tolerant fuel (ATF). The research group had previous extensive experience researching with diamond micro particles as an addition to reactor coolant for improved plant thermal performance. The purpose of this research work was to utilize diamond micro particles to develop UO2-Diamond composite fuel pellets with significantly enhanced thermal properties, beyond that already being measured in the previous UF research projects of UO2 – SiC and UO2 – Carbon Nanotube fuel pins. UF is proving with the current research results that the addition of diamond micro particles to UO2 may greatly enhanced the thermal conductivity of the UO2 pellets producing an accident-tolerant fuel. The Beginning of life benefits have been proven and fuel samples are being irradiated in the ATR reactor to confirm that the thermal conductivity improvements are still present under irradiation.

  2. FUJI - a comparative irradiation test with pellet, sphere-pac, and vipac fuel

    International Nuclear Information System (INIS)

    Hellwig, C.; Bakker, K.; Ozawa, T.; Nakamura, M.; Kihara, Y.

    2004-01-01

    Particle fuels such as sphere-pac and vipac fuels have been considered as promising fuel systems for fast reactors, due to their inherent potential in remote operation, cost reduction and incineration of minor actinides or low-decontaminated plutonium. The FUJI test addresses the questions of fabrication of MOX particle fuels with high Pu content (20%) and its irradiation behaviour during the start-up phase. Four kinds of fuel, i.e. MOX sphere-pac, MOX vipac, MOX pellet and Np-MOX sphere-pac fuel, have been and will be simultaneously irradiated under identical conditions in the High Flux Reactor in Petten. First results show that the particle fuel undergoes a dramatic structure change already at the very beginning of the irradiation when the maximum power is reached. The structural changes, i.e. the formation of a central void and the densification of fuel, decrease the fuel central temperature. Thus the fast and strong restructuring helps to prevent central fuel melting at high power levels. (authors)

  3. Development of nuclear fuel rod inspection technique using ultrasonic resonance phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Myung Sun; Lee, Jong Po; Ju, Young Sang [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-11-01

    Acoustic resonance scattering from a nuclear fuel rod in water is analyzed. A new model for the background which is attributed to the interference of reflected wave and diffracted wave is found and here named {sup t}he inherent background{sup .} The resonance spectrum of a fuel rod is obtained by subtracting the inherent background from the scattered pressure. And also analyzed are the effect of material damping of cladding tube and pellet on the resonance spectrum of a fuel rod. The propagation characteristics of circumferential waves which cause the resonances of cladding tube is produced and the appropriate resonance modes for the application to the inspection of assembled fuel rods are selected. The resonance modes are experimentally measured for pre- and post-irradiated fuel rods and the validation of the fuel rod inspection using ultrasonic resonance phenomenon is examined. And thin ultrasonic sensors accessible into the narrow interval (about 2-3mm) between assembled fuel rods are designed and manufactured. 14 refs. (Author).

  4. Discrete Modeling of Early-Life Thermal Fracture in Ceramic Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Huang, Hai [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dolbow, John E. [Duke Univ., Durham, NC (United States); Hales, Jason D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Fracturing of ceramic fuel pellets heavily influences performance of light water reactor (LWR) fuel. Early in the life of fuel, starting with the initial power ramp, large thermal gradients cause high tensile hoop and axial stresses in the outer region of the fuel pellets, resulting in the formation of radial and axial cracks. Circumferential cracks form due to thermal gradients that occur when the power is ramped down. These thermal cracks cause the fuel to expand radially, closing the pellet/cladding gap and enhancing the thermal conductance across that gap, while decreasing the effective conductivity of the fuel in directions normal to the cracking. At lower length scales, formation of microcracks is an important contributor to the decrease in bulk thermal conductivity that occurs over the life of the fuel as the burnup increases. Because of the important effects that fracture has on fuel performance, a realistic, physically based fracture modeling capability is essential to predict fuel behavior in a wide variety of normal and abnormal conditions. Modeling fracture within the context of the finite element method, which is based on continuous interpolations of solution variables, has always been challenging because fracture is an inherently discontinuous phenomenon. Work is underway at Idaho National Laboratory to apply two modeling techniques model fracture as a discrete displacement discontinuity to nuclear fuel: The extended finite element method (XFEM), and discrete element method (DEM). XFEM is based on the standard finite element method, but with enhancements to represent discontinuous behavior. DEM represents a solid as a network of particles connected by bonds, which can arbitrarily fail if a fracture criterion is reached. This paper presents initial results applying the aforementioned techniques to model fuel fracturing. This work has initially focused on early life behavior of ceramic LWR fuel. A coupled thermal-mechanical XFEM method that includes

  5. Industrias Nucleares do Brasil in the context of the Brazilian nuclear program

    International Nuclear Information System (INIS)

    1996-10-01

    The activities carried out by Industrias Nucleares Brasileiras (INB) related to the nuclear fuel cycle are described. These activities comprise presently uranium prospecting and processing and fuel elements assembly.Starting in 1997,INB will also perform the reconversion of enriched uranium hexafluoride and the fabrication of fuel pellets.Furthermore, INB produces as well rare earth oxides

  6. UN TRISO Compaction in SiC for FCM Fuel Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Trammell, Michael P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kiggans, James O. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jolly, Brian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-11-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Advanced Fuels Campaign (AFC) is conducting research and development to elevate the technology readiness level of Fully Ceramic Microencapsulated (FCM) fuels, a candidate nuclear fuel with potentially enhanced accident tolerance due to very high fission product retention. One of the early activities in FY17 was to demonstrate production of FCM pellets with uranium nitride TRISO particles. This was carried out in preparation of the larger pellet production campaign in support of the upcoming irradiation testing of this fuel form at INL’s Advanced Test Reactor.

  7. Fuel performance analysis code 'FAIR'

    International Nuclear Information System (INIS)

    Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1994-01-01

    For modelling nuclear reactor fuel rod behaviour of water cooled reactors under severe power maneuvering and high burnups, a mechanistic fuel performance analysis code FAIR has been developed. The code incorporates finite element based thermomechanical module, physically based fission gas release module and relevant models for modelling fuel related phenomena, such as, pellet cracking, densification and swelling, radial flux redistribution across the pellet due to the build up of plutonium near the pellet surface, pellet clad mechanical interaction/stress corrosion cracking (PCMI/SSC) failure of sheath etc. The code follows the established principles of fuel rod analysis programmes, such as coupling of thermal and mechanical solutions along with the fission gas release calculations, analysing different axial segments of fuel rod simultaneously, providing means for performing local analysis such as clad ridging analysis etc. The modular nature of the code offers flexibility in affecting modifications easily to the code for modelling MOX fuels and thorium based fuels. For performing analysis of fuel rods subjected to very long power histories within a reasonable amount of time, the code has been parallelised and is commissioned on the ANUPAM parallel processing system developed at Bhabha Atomic Research Centre (BARC). (author). 37 refs

  8. Out-pile test of non-instrumented capsule for the advanced PWR fuel pellets in HANARO irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Oh, D. S.; Bang, J. K.; Kim, Y. M.; Yang, Y. S.; Jeong, Y. H.; Jeon, H. K.; Ryu, J. S. [KAERI, Taejon (Korea, Republic of)

    2002-05-01

    Non-instrumental capsule were designed and fabricated to irradiate the advanced pellet developed for the high burn-up LWR fuel in the HANARO in-pile capsule. This capsule was out-pie tested at Cold Test Loop-I in KAERI. From the pressure drop test results, it is noted that the flow velocity across the non-instrumented capsule of advanced PWR fuel pellet corresponding to the pressure drop of 200 kPa is measured to be about 7.45 kg/sec. Vibration frequency for the capsule ranges from 13.0 to 32.3 Hz. RMS displacement for non-instrumented capsule of advanced PWR fuel pellet is less than 11.6 {mu}m, and the maximum displacement is less that 30.5 {mu}m. The flow rate for endurance test were 8.19 kg/s, which was 110% of 7.45 kg/s. And the endurance test was carried out for 100 days and 17 hours. The test results found not to the wear satisfied to the limits of pressure drop, flow rate, vibration and wear in the non-instrumented capsule.

  9. Current developments of fuel fabrication technologies at the plutonium fuel production facility, PFPF

    International Nuclear Information System (INIS)

    Asakura, K.; Aono, S.; Yamaguchi, T.; Deguchi, M.

    2000-01-01

    The Japan Nuclear Cycle Development Institute, JNC, designed, constructed and has operated the Plutonium Fuel Production Facility, PFPF, at the JNC Tokai Works to supply MOX fuels to the proto-type Fast Breeder Reactor, FBR, 'MONJU' and the experimental FBR 'JOYO' with 5 tonMOX/year of fabrication capability. Reduction of personal radiation exposure to a large amount of plutonium is one of the most important subjects in the development of MOX fabrication facility on a large scale. As the solution of this issue, the PFPF has introduced automated and/or remote controlled equipment in conjunction with computer controlled operation scheme. The PFPF started its operation in 1988 with JOYO reload fuel fabrication and has demonstrated MOX fuel fabrication on a large scale through JOYO and MONJU fuel fabrication for this decade. Through these operations, it has become obvious that several numbers of equipment initially installed in the PFPF need improvements in their performance and maintenance for commercial utilization of plutonium in the future. Furthermore, fuel fabrication of low density MOX pellets adopted in the MONJU fuel required a complete inspection because of difficulties in pellet fabrication compared with high density pellet for JOYO. This paper describes new pressing equipment with a powder recovery system, and pellet finishing and inspection equipment which has multiple functions, such as grinding measurements of outer diameter and density, and inspection of appearance to improve efficiency in the pellet finishing and inspection steps. Another development of technology concerning an annular pellet and an innovative process for MOX fuel fabrication are also described in this paper. (author)

  10. Laser-prearc railgun: Development for the application to a fuel pellet injector of a nuclear fusion reactor

    Science.gov (United States)

    Tamura, H.; Sawaoka, A. B.; Oda, Y.; Onozuka, M.; Kuribayashi, S.; Shimizu, K.

    1992-05-01

    The laser-prearc railgun, that utilizes the phenomenon of laser-induced arc formation, was constructed and tested with plastic pellet projectiles. We envision our railgun as especially well suited as a solid hydrogen pellet injector for magnetic confinement fusion. The system consisted of a gas gun for preacceleration of a pellet and a railgun for its primary acceleration. A Q-switched ruby laser was used to induce electrical breakdown of propellant helium gas behind a dielectric pellet in the railgun. The present railgun was shown to accelerate a plastic pellet up to a velocity of 2.4 km/s.

  11. Thermomechanical analysis of nuclear fuel elements

    International Nuclear Information System (INIS)

    Hernandez L, H.

    1997-01-01

    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)

  12. Some aspects of a technology of processing weapons grade plutonium to nuclear fuel

    International Nuclear Information System (INIS)

    Bibilashvili, Y.; Glagovsky, E.M.; Zakharkin, B.S.; Orlov, V.K.; Reshetnikov, F.G.; Rogozkin, B.G.; Soloni-N, M.I.

    2000-01-01

    The concept by Russia to use fissile weapons-grade materials, which are being recovered from nuclear pits in the process of disarmament, is based on an assessment of weapons-grade plutonium as an important energy source intended for use in nuclear power plants. However, in the path of involving plutonium excessive from the purposes of national safety into industrial power engineering there are a lot of problems, from which effectiveness and terms of its disposition are being dependent upon. Those problems have political, economical, financial and environmental character. This report outlines several technology problems of processing weapons-grade metallic plutonium into MOX-fuel for reactors based on thermal and fast neutrons, in particular, the issue of conversion of the metal into dioxide from the viewpoint of fabrication of pelletized MOX-fuel. The processing of metallic weapons-grade plutonium into nuclear fuel is a rather complicated and multi-stage process, every stage of which is its own production. Some of the stages are absent in production of MOX-fuel, for instance the stage of the conversion, i.e. transferring of metallic plutonium into dioxide of the ceramic quality. At this stage of plutonium utilization some tasks must be resolved as follows: I. As a result of the conversion, a material purified from ballast and radiogenic admixtures has to be obtained. This one will be applied to fabricate pelletized MOX-fuel going from morphological, physico-mechanical and technological properties. II. It is well known that metallic gallium, which is used as an alloying addition in weapons-grade plutonium, actively reacts with multiple metals. Therefore, an important issue is to study the effect of gallium on the technology of MOX-fuel production, quality of the pellets, as well as the interaction of gallium oxide with zirconium and steel shells of fuel elements depending upon the content of gallium in the fuel. The rate of the interaction of gallium oxide

  13. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Berrill, Mark A [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Pugmire, Dave [ORNL; Dilts, Gary [Los Alamos National Laboratory (LANL); Banfield, James E [ORNL

    2012-02-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

  14. Study of the feasibility of mixing Refuse Derived Fuels with wood pellets through the grey and Fuzzy theory

    Energy Technology Data Exchange (ETDEWEB)

    Moran, J.C.; Miguez, J.L.; Porteiro, J.; Patino, D.; Granada, E.; Collazo, J. [Universidad de Vigo E.T.S. Ingenieros Industriales, Lagoas-Marcosende, s/n. Dpto. Ing. Mecanica Maquinas y Motores Termicos, 36200 Vigo (Pontevedra) (Spain)

    2009-12-15

    This paper presents a combined grey relational and fuzzy analysis for the evaluation of the environmental feasibility of burning mixtures of pellet and RDF (Refuse Derived Fuel) in a small pellet boiler-stove. RDF is obtained from Municipal Solid Waste (MSW) and contains a biomass fraction and a non-organic fraction (plastic). As a first step, both fuels are characterized to define their properties. A special feeding system is also used to improve the stove plant and to facilitate pellet distribution, which maintains a constant rate between the two fuels. Small scale energy converters, such as chimneys, boilers, stoves, etc., which produce heat and/or hot water by burning biomass (wood, pellets, briquettes, etc.), are especially suited to domestic purposes. However, in common commercial combustion conditions, this kind of use still has some disadvantages: some emissions (volatile organic carbons, carbon monoxide or NO{sub x}) may still be high, and it is difficult to compare the quality and performance of equipment working in very different combustion conditions. The grey relational analysis of different energy and emission variables leads to the definition of a new single variable called the grey relational grade (GRG). Thus, evaluation and optimisation of complicated multiple responses can be converted into the optimisation of a standardised single variable. The aim of the work is to research the most feasible mixture of pellets according to a grey relational analysis, taking into consideration energy-related, financial and environmental aspects. (author)

  15. Development of nuclear fuel for integrated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO{sub 2}-based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO{sub 2}-based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method.

  16. Development of nuclear fuel for integrated reactor

    International Nuclear Information System (INIS)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M.

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO 2 -based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO 2 -based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method

  17. Niobia-doped UO2 fuel manufacturing experience at British nuclear fuels Ltd

    International Nuclear Information System (INIS)

    Marsh, G.; Wood, G.A.; Perkins, C.P.

    1998-01-01

    BNFL Fuel Division has made niobia doped fuel for over twenty years in its Springfields Research and Development facilities. This paper reviews this experience together with feedback from successful in-reactor and laboratory tests. Recent experience in qualifying and manufacturing niobia doped fuel pellets for a European PWR will be described. (author)

  18. Design study of advanced nuclear fuel recycle system. Conceptual study of recycle system using molten salt

    International Nuclear Information System (INIS)

    Kakehi, I.; Shirai, N.; Hatano, M.; Kajitani, M.; Yonezawa, S.; Kawai, T.; Kawamura, F.; Tobe, K.; Takahashi, K.

    1996-12-01

    For the purpose of developing the future nuclear fuel recycle system, the design study of the advanced nuclear fuel recycle system is being conducted. This report describes intermediate accomplishments in the conceptual system study of the advanced nuclear fuel recycle system. Fundamental concepts of this system is the recycle system using molten salt which intend to break through the conventional concepts of purex and pellet fuel system. Contents of studies in this period are as follows, 1)feasibility study of the process by Cd-cathode for nitride fuel, 2)application study for the molten salt of low melting point (AlCl3+organic salt), 3)research for decladding (advantage of decladding by heat treatment), 4)behavior of FPs in electrorefining (behavior of iodine and volatile FP chlorides, FPs behavior in chlorination), 5)criticality analysis in electrorefiner, 6)drawing of off-gas flow diagram, 7)drawing of process machinery concept (cathode processor, vibration packing), 8)evaluation for the amounts of the high level radioactive wastes, 9)quality of the recycle fuels (FPs contamination of recycle fuel), 10)conceptual study of in-cell handling system, 11)meaning of the advanced nuclear fuel recycle system. The conceptual system study will be completed in describing concepts of the system and discussing issues for the developments. (author)

  19. Industrias Nucleares do Brasil in the context of the Brazilian nuclear program; A INB no contexto do programa nuclear brasileiro

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-10-01

    The activities carried out by Industrias Nucleares Brasileiras (INB) related to the nuclear fuel cycle are described. These activities comprise presently uranium prospecting and processing and fuel elements assembly.Starting in 1997,INB will also perform the reconversion of enriched uranium hexafluoride and the fabrication of fuel pellets.Furthermore, INB produces as well rare earth oxides 2 figs., 1 tab.

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

    International Nuclear Information System (INIS)

    Domoratskii, E.P.

    1994-01-01

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

  1. Testing of reactor fuel materials using nuclear techniques

    International Nuclear Information System (INIS)

    Khouri, M.T.F.C.

    1978-01-01

    The tests presented here apply to: the quantitative determination of uranium in the core of fuel element plates by the detection of the number of neutrons produced in photo induced reactions in uranium; the determination of 235 U proportion in uranium dioxide samples, in the form of uranyl nitrate, by the technique of the detection of tracks produced by fission fragments and in pellet samples by passive gamma spectrometry and the checking of uranium homogenization distribution in fuel plates and uranium dioxide pellets. (Author) [pt

  2. Second jet workshop on pellet injection: pellet fueling program in the United States. Summary

    International Nuclear Information System (INIS)

    Milora, S.L.

    1983-01-01

    S. Milora described the US programme on pellet injection. It has four parts: (1) a confinement experimental program; (2) pellet injector development; (3) theoretical support; and (4) tritium pellet study for TFTR

  3. WWER-1000 nuclear fuel manufacturing process at PJSC MSZ

    International Nuclear Information System (INIS)

    Morylev, A.; Bagdatyeva, E.; Aksenov, P.

    2015-01-01

    In this report a brief description of WWER-1000 fuel manufacturing process steps at PJSC MSZ as: uranium dioxide powder fabrication; fuel pellet manufacture fuel rod manufacture working assembly and fuel assembly manufacture is given. The implemented innovations are also presented

  4. Preliminary pellet injection experiment in the Gamma 10 tandem mirror

    Energy Technology Data Exchange (ETDEWEB)

    Kawamori, Eiichirou; Tamano, Teruo; Nakashima, Yousuke; Yoshikawa, Masayuki; Kobayashi, Shinji; Cho, Teruji; Ishii, Kameo; Yatsu, Kiyoshi [Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki (Japan); Mase, Atsushi [Advanced Sceince and Technology Center for Cooperative Research, Kyushu University, Kasuga, Fukuoka (Japan)

    2000-07-01

    In the GAMMA 10 tandem mirror, pellet injection experiments have been started as a solution for the density limit problem. This is the first pellet injection experiment in open systems. We describe the possibilities of confinement of pellet fueled particles. For that, we measure the number of end loss particles and compare them with pellet fueled ones in various conditions of confining potentials. The deterioration of confining potential with the pellet injection is a fundamental issue. The results show that the ion confining potential recover faster than central electron temperature due to thermal barrier. We also consider the operating space for fueling method. It is demonstrated that the operating space for pellet injection exceeds gas fueled one on hot ion mode plasmas. (author)

  5. Nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-10-01

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

  6. Canada's nuclear fuel industry: An overview. Background paper

    International Nuclear Information System (INIS)

    Nixon, A.

    1993-11-01

    Canada was among the first countries to mine and process uranium-bearing ores. Such ores contain trace amounts of radium, which was in great demand for medical treatment and for use by research laboratories in the early part of the century. For the last half century, the same basic processes have been used to extract uranium from its ores and convert it to a form suitable for use in nuclear reactors. The process described here is that currently in use in Canada. Mining can take a variety of forms, from open-pit to deep, hard-rock. Mining is typically the most costly step in the process, particularly for lower-grade ores. The ore is crushed and ground in the mill to the consistency of fine sand from which the uranium is extracted chemically to produce the impure concentrate known as yellowcake. In the next step, the impure uranium concentrate is chemically refined into highly purified, nuclear-grade, uranium trioxide (UO 3 ). Uranium trioxide is then converted, in two separate chemical processes, into uranium dioxide (UO 2 ) which is destined for domestic consumption and uranium hexafluoride (UF 6 ) which is exported. In Canada, fabrication is the final step of the fuel production process. Uranium dioxide powder is compressed and sintered into very dense ceramic pellets which are then sealed in zirconium tubes and assembled into fuel bundles for Candu reactors. This background paper will review the Canadian nuclear fuels industry. 1 fig

  7. Quality properties of fuel pellets from forest biomass

    Energy Technology Data Exchange (ETDEWEB)

    Lehtikangas, P.

    1999-07-01

    Nine pellet assortments, made of fresh and stored sawdust, bark and logging residues (a mixture of Norway spruce and Scots pine) were tested directly after production and after 5 months of storage in large bags (volume about 1 m{sup 3} loose pellets) for moisture content, heating value and ash content. Dimensions, bulk density, density of individual pellets and durability were also determined. Moreover, sintering risk and contents of sulphur, chlorine, and lignin of fresh pellets were determined. It is concluded that bark and logging residues are suitable raw materials for pellets production, especially regarding durability and if the ash content is controlled. Pellets density had no effect on its durability, unlike lignin content which was positively correlated. The pellets had higher ash content and lower calorific heating value than the raw materials, probably due to loss of volatiles during drying. In general, the quality changes during storage were not large, but notable. The results showed that storage led to negative effects on durability, especially on pellets made of fresh materials. The average length of pellets was decreased due to breakage during storage. Microbial growth was noticed in some of the pellet assortments. Pellets made out of fresh logging residues were found to be weakest after storage. The tendency to reach the equilibrium with the ambient moisture content should be taken into consideration during production due to the risk of decreasing durability.

  8. Fuel pellets from biomass - Processing, bonding, raw materials

    Energy Technology Data Exchange (ETDEWEB)

    Stelte, W.

    2011-12-15

    The present study investigates several important aspects of biomass pelletization. Seven individual studies have been conducted and linked together, in order to push forward the research frontier of biomass pelletization processes. The first study was to investigate influence of the different processing parameters on the pressure built up in the press channel of a pellet mill. It showed that the major factor was the press channel length as well as temperature, moisture content, particle size and extractive content. Furthermore, extractive migration to the pellet surface at an elevated temperature played an important role. The second study presented a method of how key processing parameters can be estimated, based on a pellet model and a small number of fast and simple laboratory trials using a single pellet press. The third study investigated the bonding mechanisms within a biomass pellet, which indicate that different mechanisms are involved depending on biomass type and pelletizing conditions. Interpenetration of polymer chains and close intermolecular distance resulting in better secondary bonding were assumed to be the key factors for high mechanical properties of the formed pellets. The outcome of this study resulted in study four and five investigating the role of lignin glass transition for biomass pelletization. It was demonstrated that the softening temperature of lignin was dependent on species and moisture content. In typical processing conditions and at 8% (wt) moisture content, transitions were identified to be at approximately 53-63 deg. C for wheat straw and about 91 deg. C for spruce lignin. Furthermore, the effects of wheat straw extractives on the pelletizing properties and pellet stability were investigated. The sixth and seventh study applied the developed methodology to test the pelletizing properties of thermally pre-treated (torrefied) biomass from spruce and wheat straw. The results indicated that high torrefaction temperatures above 275 deg

  9. The Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    2011-08-01

    This brochure describes the nuclear fuel cycle, which is an industrial process involving various activities to produce electricity from uranium in nuclear power reactors. The cycle starts with the mining of uranium and ends with the disposal of nuclear waste. The raw material for today's nuclear fuel is uranium. It must be processed through a series of steps to produce an efficient fuel for generating electricity. Used fuel also needs to be taken care of for reuse and disposal. The nuclear fuel cycle includes the 'front end', i.e. preparation of the fuel, the 'service period' in which fuel is used during reactor operation to generate electricity, and the 'back end', i.e. the safe management of spent nuclear fuel including reprocessing and reuse and disposal. If spent fuel is not reprocessed, the fuel cycle is referred to as an 'open' or 'once-through' fuel cycle; if spent fuel is reprocessed, and partly reused, it is referred to as a 'closed' nuclear fuel cycle.

  10. Some regularity of the grain size distribution in nuclear fuel with controllable structure

    International Nuclear Information System (INIS)

    Loktev, Igor

    2008-01-01

    It is known, the fission gas release from ceramic nuclear fuel depends from average size of grains. To increase grain size they use additives which activate sintering of pellets. However, grain size distribution influences on fission gas release also. Fuel with different structures, but with the same average size of grains has different fission gas release. Other structure elements, which influence operational behavior of fuel, are pores and inclusions. Earlier, in Kyoto, questions of distribution of grain size for fuel with 'natural' structure were discussed. Some regularity of grain size distribution of fuel with controllable structure and high average size of grains are considered in the report. Influence of inclusions and pores on an error of the automated definition of parameters of structure is shown. The criterion, which describe of behavior of fuel with specific grain size distribution, is offered

  11. Development of advanced LWR fuel pellet technology

    International Nuclear Information System (INIS)

    Song, Kun Woo; Kang, K.W.; Kim, K. S.; Yang, J. H.; Kim, Y. M.; Kim, J. H.; Bang, J. B.; Kim, D. H.; Bae, S. O.; Jung, Y. H.; Lee, Y. S.; Kim, B. G.; Kim, S. H.

    2000-03-01

    A UO 2 pellet was designed to have a grain size of larger than 12 μm, and a new duplex design that UO 2 -Gd 2 O 3 is in the core and UO 2 -Er 2 O 3 in the periphery was proposed. A master mixing method was developed to make a uniform mixture of UO 2 and additives. The open porosity of UO 2 pellet was reduced by only mixing AUC-UO 2 powder with ADU-UO 2 or milled powder. Duplex compaction tools (die and punch) were designed and fabricated, and duplex compacting procedures were developed to fabricate the duplex BA pellet. In UO 2 sintering, the relations between sintering variables (additive, sintering gas, sintering temperature) and pellet properties (density, grain size, pore size) were experimentally found. The UO 2 -U 3 O 8 powder which is inherently not sinterable to high density could be sintered well with the aid of additives. U 3 O 8 single crystals were added to UO 2 powder, and homogeneous powder mixture was pressed and sintered in a reducing atmosphere. This technology leads to a large-grained pellet of 12-20 μm. In UO 2 -Gd 2 O 3 sintering, the relations between sintering variables (additives, sintering gas) and pellet properties (density, grain size) were experimentally found. The developed technology of fabricating a large-grained UO 2 pellet has been optimized in a lab scale. Pellet properties were investigated in the fields of (1) creep properties, (2) thermal properties, (3) O/M ratios and (4) unit cell lattice. (author)

  12. Uncertainty and sensitivity analysis of the nuclear fuel thermal behavior

    Energy Technology Data Exchange (ETDEWEB)

    Boulore, A., E-mail: antoine.boulore@cea.fr [Commissariat a l' Energie Atomique (CEA), DEN, Fuel Research Department, 13108 Saint-Paul-lez-Durance (France); Struzik, C. [Commissariat a l' Energie Atomique (CEA), DEN, Fuel Research Department, 13108 Saint-Paul-lez-Durance (France); Gaudier, F. [Commissariat a l' Energie Atomique (CEA), DEN, Systems and Structure Modeling Department, 91191 Gif-sur-Yvette (France)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer A complete quantitative method for uncertainty propagation and sensitivity analysis is applied. Black-Right-Pointing-Pointer The thermal conductivity of UO{sub 2} is modeled as a random variable. Black-Right-Pointing-Pointer The first source of uncertainty is the linear heat rate. Black-Right-Pointing-Pointer The second source of uncertainty is the thermal conductivity of the fuel. - Abstract: In the global framework of nuclear fuel behavior simulation, the response of the models describing the physical phenomena occurring during the irradiation in reactor is mainly conditioned by the confidence in the calculated temperature of the fuel. Amongst all parameters influencing the temperature calculation in our fuel rod simulation code (METEOR V2), several sources of uncertainty have been identified as being the most sensitive: thermal conductivity of UO{sub 2}, radial distribution of power in the fuel pellet, local linear heat rate in the fuel rod, geometry of the pellet and thermal transfer in the gap. Expert judgment and inverse methods have been used to model the uncertainty of these parameters using theoretical distributions and correlation matrices. Propagation of these uncertainties in the METEOR V2 code using the URANIE framework and a Monte-Carlo technique has been performed in different experimental irradiations of UO{sub 2} fuel. At every time step of the simulated experiments, we get a temperature statistical distribution which results from the initial distributions of the uncertain parameters. We then can estimate confidence intervals of the calculated temperature. In order to quantify the sensitivity of the calculated temperature to each of the uncertain input parameters and data, we have also performed a sensitivity analysis using the Sobol' indices at first order.

  13. Coordinated irradiation plan for the Fuel Refabrication and Development Program

    International Nuclear Information System (INIS)

    Barner, J.O.

    1979-04-01

    The Department of Energy's Fuel Refabrication and Development (FRAD) Program is developing a number of proliferation-resistant fuel systems and forms for alternative use in nuclear reactors. A major portion of the program is the development of irradiation behavioral information for the fuel system/forms with the ultimate objective of qualifying the design for licensing and commercial utilization. The nuclear fuel systems under development include denatured thoria--urania fuels and spiked urania--plutonia or thoria--plutonia fuels. The fuel forms being considered include pellet fuel produced from mechanically mixed or coprecipitated feed materials, pellet fuel fabricated from partially calcined gel-derived or freeze-dried spheres (hybrid fuel) and packed-particle fuel produced from sintered gel-derived spheres (sphere-pac). This document describes the coordinated development program that will be used to test and demonstrate the irradiation performance of alternative fuels

  14. Fuel and nuclear fuel cycle

    International Nuclear Information System (INIS)

    Prunier, C.

    1998-01-01

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

  15. Recent advances in the theory and simulation of pellet ablation and fast fuel relocation in tokamaks

    International Nuclear Information System (INIS)

    Parks, P.B.; Baylor, L.R.; Ishizaki, R.; Jardin, S.C.; Samtaney, R.

    2005-01-01

    This paper presents new theory and simulation of pellet ablation, and the rapid cross-field redistribution of the ionized pellet mass following pellet injection in tokamaks. The first 2-D time-dependent simulations describing the expansion of pellet ablation flow against the magnetic field is presented here using the Eulerian code CAP. The early-time expansion is characterized by the formation of an ellipsoidal diamagnetic cavity surrounding the pellet, which diverts heat flux around the pellet, thereby reducing the ablation rate. Near-pellet cloud properties from CAP provide initial conditions for the subsequent ExB advection of the ionized clouds caused by polarization in the inhomogeneous toroidal magnetic field. The first complete set of time-dependent equations describing mass redistribution has been developed and solved for numerically using the PRL code. New effects identified, including curvature drive by near sonic field-aligned flows, rotational transform of the magnetic field lines and magnetic shear are considered from the viewpoint of the parallel vorticity equation. Close agreement between theory and experimental fuel deposition profiles are obtained for both inner and outer wall pellet injection on the DIII-D tokamak, providing improved predictive capability for ITER. A new 3-D MHD simulation code AMR was started, which provides the required fine-scale mesh size needed for accurate modeling of pellet clouds having sharp perpendicular-to-B gradients. (author)

  16. Processing of excess fermentation remainders to compact fuel pellets; Verarbeitung ueberschuessiger Gaerreste zu kompakten Brennstoffpellets

    Energy Technology Data Exchange (ETDEWEB)

    Kirsten, Claudia; Lenz, Volker; Schroeder, Hans-Werner; Repke, Jens-Uwe [Deutsches Biomasseforschungszentrum (DBFZ) gemeinnuetzige GmbH, Leipzig (Germany). AG Brennstoffe im Bereich Thermo-Chemische Konversion

    2013-10-01

    With a growing number of biogas plants in Germany the amounts of digestate grows too. In regions with high livestock farming and low agriculture the disposal of digestate as fertilizer can be problematic for energetic related use of digestate, both combustion technologies and fuel-qualities have to be optimized. Depending on the fermentation process, the variability of the ingredients and the structure of the digestate, special challenges for the processing of the raw material have to be overcome. Thus, analysis of the grinding and pelletizing behavior of digestate is presented. Fuel pellets with a durability of up to 98 wt.-% and a bulk density up to 700 kg/m{sup 3} could be produced, which is in accordance with the requirements on the physical-mechanical properties of EN 14961-6. (orig.)

  17. Romanian nuclear fuel program

    International Nuclear Information System (INIS)

    Budan, O.

    1999-01-01

    The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified and authorised CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle

  18. Nuclear fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding, and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

  19. Nuclear fuel element

    International Nuclear Information System (INIS)

    Thompson, J.R.; Rowland, T.C.

    1976-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting, fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

  20. Modeling of the PWR fuel mechanical behaviour and particularly study of the pellet-cladding interaction in a fuel rod

    International Nuclear Information System (INIS)

    Hourdequin, N.

    1995-05-01

    In Pressurized Water Reactor (PWR) power plants, fuel cladding constitutes the first containment barrier against radioactive contamination. Computer codes, developed with the help of a large experimental knowledge, try to predict cladding failures which must be limited in order to maintain a maximal safety level. Until now, fuel rod design calculus with unidimensional codes were adequate to prevent cladding failures in standard PWR's operating conditions. But now, the need of nuclear power plant availability increases. That leads to more constraining operating condition in which cladding failures are strongly influenced by the fuel rod mechanical behaviour, mainly at high power level. Then, the pellet-cladding interaction (PCI) becomes important, and is characterized by local effects which description expects a multidimensional modelization. This is the aim of the TOUTATIS 2D-3D code, that this thesis contributes to develop. This code allows to predict non-axisymmetric behaviour too, as rod buckling which has been observed in some irradiation experiments and identified with the help of TOUTATIS. By another way, PCI is influenced by under irradiation experiments and identified with the help of TOUTATIS which includes a densification model and a swelling model. The latter can only be used in standard operating conditions. However, the processing structure of this modulus provides the possibility to include any type of model corresponding with other operating conditions. In last, we show the result of these fuel volume variations on the cladding mechanical conditions. (author). 25 refs., 89 figs., 2 tabs., 12 photos., 5 appends

  1. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    Shoup, R.L.

    1976-01-01

    The composite of metals and alloys used in the fabrication of 238 Pu cardiac pacemaker fuel capsules resists the effects of high temperatures, high mechanical forces, and chemical corrosives and provides more than adequate protection to the fuel pellet even from deliberate attempts to dissolve the cladding in inorganic acids. This does not imply that opening a pacemaker fuel capsule by inorganic acids is impossible but that it would not be a wise choice

  2. Modelling of pellet-cladding interaction in PWR's

    International Nuclear Information System (INIS)

    Esteves, A.M.; Silva, A.T. e.

    1992-01-01

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyses the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. (author)

  3. Apparatus and method for loading pellets into fuel rods

    International Nuclear Information System (INIS)

    Widener, W.H.

    1991-01-01

    An apparatus for feeding a column of aligned cylindrical pellets along a longitudinal path of travel and while identifying a pellet of improper size. It comprises guide surface means adapted for supporting a plurality of serially arranged and longitudinally oriented cylindrical pellets, and such that the pellets are adapted to be slidably and longitudinally advanced along the guide surface means to define an advancing column of pellets, and pellet segregation means positioned adjacent one end of the guide surface means for permitting each advancing pellet having a cross-sectional diameter equal to a predetermined minimum diameter to advance thereacross while permitting each advancing pellet having a cross-sectional diameter less than the predetermined minimum diameter to drop to a level below the level of the remaining pellets in the advancing column

  4. Nuclear-fuel-cycle education: Module 1. Nuclear fuel cycle overview

    International Nuclear Information System (INIS)

    Eckhoff, N.D.

    1981-07-01

    This educational module is an overview of the nuclear-fule-cycle. The overview covers nuclear energy resources, the present and future US nuclear industry, the industry view of nuclear power, the International Nuclear Fuel Cycle Evaluation program, the Union of Concerned Scientists view of the nuclear-fuel-cycle, an analysis of this viewpoint, resource requirements for a model light water reactor, and world nuclear power considerations

  5. Development of advanced LWR fuel pellet technology

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kun Woo; Kang, K.W.; Kim, K. S.; Yang, J. H.; Kim, Y. M.; Kim, J. H.; Bang, J. B.; Kim, D. H.; Bae, S. O.; Jung, Y. H.; Lee, Y. S.; Kim, B. G.; Kim, S. H

    2000-03-01

    A UO{sub 2} pellet was designed to have a grain size of larger than 12 {mu}m, and a new duplex design that UO{sub 2}-Gd{sub 2}O{sub 3} is in the core and UO{sub 2}-Er{sub 2}O{sub 3} in the periphery was proposed. A master mixing method was developed to make a uniform mixture of UO{sub 2} and additives. The open porosity of UO{sub 2} pellet was reduced by only mixing AUC-UO{sub 2} powder with ADU-UO{sub 2} or milled powder. Duplex compaction tools (die and punch) were designed and fabricated, and duplex compacting procedures were developed to fabricate the duplex BA pellet. In UO{sub 2} sintering, the relations between sintering variables (additive, sintering gas, sintering temperature) and pellet properties (density, grain size, pore size) were experimentally found. The UO{sub 2}-U{sub 3}O{sub 8} powder which is inherently not sinterable to high density could be sintered well with the aid of additives. U{sub 3}O{sub 8} single crystals were added to UO{sub 2} powder, and homogeneous powder mixture was pressed and sintered in a reducing atmosphere. This technology leads to a large-grained pellet of 12-20 {mu}m. In UO{sub 2}-Gd{sub 2}O{sub 3} sintering, the relations between sintering variables (additives, sintering gas) and pellet properties (density, grain size) were experimentally found. The developed technology of fabricating a large-grained UO{sub 2} pellet has been optimized in a lab scale. Pellet properties were investigated in the fields of (1) creep properties, (2) thermal properties, (3) O/M ratios and (4) unit cell lattice. (author)

  6. Development of U-frame bending system for studying the vibration integrity of spent nuclear fuel

    Science.gov (United States)

    Wang, Hong; Wang, Jy-An John; Tan, Ting; Jiang, Hao; Cox, Thomas S.; Howard, Rob L.; Bevard, Bruce B.; Flanagan, Michelle

    2013-09-01

    A bending fatigue system developed to evaluate the response of spent nuclear fuel rods to vibration loads is presented. A U-frame testing setup is used for imposing bending loads on the fuel rod specimen. The U-frame setup consists of two rigid arms, side connecting plates to the rigid arms, and linkages to a universal testing machine. The test specimen's curvature is obtained through a three-point deflection measurement method. The tests using surrogate specimens with stainless steel cladding revealed increased flexural rigidity under unidirectional cyclic bending, significant effect of cladding-pellets bonding on the response of surrogate rods, and substantial cyclic softening in reverse bending mode. These phenomena may cast light on the expected response of a spent nuclear fuel rod. The developed U-frame system is thus verified and demonstrated to be ready for further pursuit in hot-cell tests.

  7. Specific features of the determination of the pellet-cladding gap of the fuel rods by non-destructive method

    International Nuclear Information System (INIS)

    Amosov, S.V.; Pavlov, S.V.

    2002-01-01

    This report describes the specific features of determining the pellet-cladding gap of the irradiated WWER-1000 fuel rods by nondestructive method. The method is based on the elastic radial deformation of the cladding up to its contact with the fuel. The value of deformation of cladding till its contacting fuel when radial force changes from F max to 0 is proposed as a measuring parameter for determination of the diametrical gap. Because of the features of compression method, the obtained gap value is not analog of the gap measured on micrograph of the fuel rod cross-section. Results of metallography can provide only qualitative evaluation of its method efficiency. Comparison of the values determined by non-destructive method and metallography for WWER-1000 fuel rods with burnup from 25 to 55 MWd/kg U testified that the results of compression method can be used as a low estimate of the pellet-cladding gap value. (author)

  8. Combustion tests with different pellet qualities

    International Nuclear Information System (INIS)

    Bachs, A.; Dahlstroem, J.E.; Persson, Henrik; Tullin, C.

    1999-05-01

    Eight different pellet qualities with the diameters 6, 8 and 10 mm, from eight different producers has been tested in three pellet burners and two pellet stoves. The objective was to investigate how different diameter affect the emissions of CO, OGC and NO x . Previous experience has indicated that the pellet diameter could have significant importance for the combustion. This was not verified in the study. It showed contradictory that the diameter has a minor effect on the combustion result. The study shows that different combustion equipment give different emission. For e g hydrocarbon emissions the difference is a factor 2.2 between the 'best' and the 'worst' equipment fired on full load. The difference increases to 2.7 with lower load. The choice of fuel has a big importance for the quality of the combustion. For hydrocarbons the emissions could in an extreme situation differ with a factor 25 between 'best' and 'worst' fuel. More normally the difference is about a factor of five. Nitrogen oxide emissions are to a major part related to the nitrogen contents in the fuel. The difference between the 'best' and 'worst' fuel is in the range of a factor two. Tests with the same fuel in different equipment gives a variation of 20-30%. The combustion result depends on both the pellet quality and the equipment and there is no fuel that is good in all equipment. The big variation in combustion results shows that there is a big indifference between fuels used for small scale heating Project report from the program: Small scale combustion of biofuels. 2 refs, 15 figs, 5 tabs

  9. Spectroscopic methods for characterization of nuclear fuels

    International Nuclear Information System (INIS)

    Sastry, M.D.

    1999-01-01

    Spectroscopic techniques have contributed immensely in the characterisation and speciation of materials relevant to a variety of applications. These techniques have time tested credentials and continue to expand into newer areas. In the field of nuclear fuel fabrication, atomic spectroscopic methods are used for monitoring the trace metallic constituents in the starting materials and end product, and for monitoring process pick up. The current status of atomic spectroscopic methods for the determination of trace metallic constituents in nuclear fuel materials will be briefly reviewed and new approaches will be described with a special emphasis on inductively coupled plasma techniques and ETV-ICP-AES hyphenated techniques. Special emphasis will also be given in highlighting the importance of chemical separation procedures for the optimum utilization of potential of ICP. The presentation will also include newer techniques like Photo Acoustic Spectroscopy, and Electron Paramagnetic Resonance (EPR) Imaging. PAS results on uranium and plutonium oxides will be described with a reference to the determination of U 4+ /U 6+ concentration in U 3 O 8 . EPR imaging techniques for speciation and their spatial distribution in solids will be described and its potential use for Gd 3+ containing UO 2 pellets (used for flux flattening) will be highlighted. (author)

  10. Pelletizing and combustion of wood from thinning; Pelletering och foerbraenning av gallringsvirke

    Energy Technology Data Exchange (ETDEWEB)

    Oerberg, Haakan; Thyrel, Mikael; Kalen, Gunnar; Larsson, Sylvia

    2007-12-14

    This work has been done in order to find new raw material sources for an expanding pellet industry, combined with finding a use for a forest product that has no market today. The raw material has been forest from early thinning in two typical stands in Vaesterbotten. The purpose has been to evaluate this material as a raw material for producing pellets. Two typical stands have been chosen. One stand with only pine trees and one mixed stand dominated by birch. The soil of these stands was poor. Half of the trees were delimbed by harvest and half of the trees were not delimbed. This formed four different assortments that were handled in the study. After harvesting the assortments were transported to an asphalt area to be stored. Half of the material was stored during one summer and half of the material was stored during one year and one summer. The different assortments were upgraded to pellets and test combusted in the research plant BTC at the Swedish University of Agricultural Sciences, in Umeaa. The upgrading process contains of the following steps: 1.Chipping by a mobile chipper. 2.Low temperature drying (85 deg C). 3. Coarse shredding ({phi}15 mm). 4. Fine shredding ({phi}4-6 mm) and 5. Pelletizing (Die: {phi}8). Samples for fuel analysis were taken during the chipping. Analyses shows that the net calorific value for delimbed assortments are about 0,3 MJ/kg DM higher than for limbed assortments. Pellets made of the assortments Mixed limbed and Pine limbed has shown a net calorific value comparable to stem wood pellets. Pellets made of Birch delimbed show a net calorific value 0,4 MJ/kg DM lower than stem wood pellets. Analyses show that ash contents of the assortment Mixed delimbed was 1 %-unit higher compared to stem wood pellets. The assortment Pine delimbed and Birch delimbed has showed an ash contents comparable with stem wood pellets. The ash melting characteristics can reduce the value of a raw material. Low ash melting temperature for a fuel might cause

  11. Pellets standard on the way

    International Nuclear Information System (INIS)

    Laeng, H.-P.

    2001-01-01

    This short article introduces the Swiss standard that has been adapted from the German standard for heating pellets made of untreated wood. The various requirements placed on the materials used in the manufacture of the pellets and their influence on the pollution emissions produced by boilers and ovens using the pellets as a heating fuel are listed. Further points in the standard referring to declarations to be made by the manufacturer, size and specific weight of the pellets and instructions for the storage and burning of the pellets are discussed

  12. Pneumatic pellet injector for JET

    International Nuclear Information System (INIS)

    Andelfinger, C.; Buechl, K.; Jacobi, D.; Sandmann, W.; Schiedeck, J.; Schilling, H.B.; Weber, G.

    1983-07-01

    Pellet injection is a useful tool for plasma diagnostics of tokamaks. Pellets can be applied for investigation of particle, energy and impurity transport, fueling efficiency and magnetic surfaces. Design, operation and control of a single shot pneumatic pellet gun is described in detail including all supplies, the vacuum system and the diagnostics of the pellet. The arrangement of this injector in the torus hall and the interfaces to the JET system and CODAS are considered. A guide tube system for pellet injection is discussed but it will not be recommended for JET. (orig.)

  13. Development of the centrifugal pellet injector for JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Kizu, Kaname; Hiratsuka, Hajime; Ichige, Hisashi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    2001-03-01

    For core fueling of JT-60U plasmas, a repetitive pellet injector which centrifugally accelerates D{sub 2} cubic pellets using a straight rod has been developed. This centrifugal pellet injector can eject trains of up to 40 cubic pellets at frequencies of 1-10 Hz and velocities of 0.3-1.0 km/s. The average pellet mass is 3.6x10{sup 20} atoms/pellet below 0.7 m/s. Key techniques for the development were a mesh structured acceleration component for removing gas sublimated from the pellet and a funnel with an appropriate angle connected just behind the acceleration chamber for introducing the pellet to plasma without destruction. Using the mesh structured components, the horizontal angular distribution of pellets ejected became narrow, because irregular pellet motion caused by sublimated gas was reduced. To investigate the performance of the injector, pellet injection experiments from the low magnetic field side (LFS) were conducted using ohmic heating plasmas. Central fueling and enhanced fueling rate have been observed. D{alpha} intensity around the divertor region was reduced in a pellet injection plasma compared to gas puffing, indicating low recycling rate was maintained with the pellet injection. (author)

  14. Development of the centrifugal pellet injector for JT-60U

    International Nuclear Information System (INIS)

    Kizu, Kaname; Hiratsuka, Hajime; Ichige, Hisashi

    2001-03-01

    For core fueling of JT-60U plasmas, a repetitive pellet injector which centrifugally accelerates D 2 cubic pellets using a straight rod has been developed. This centrifugal pellet injector can eject trains of up to 40 cubic pellets at frequencies of 1-10 Hz and velocities of 0.3-1.0 km/s. The average pellet mass is 3.6x10 20 atoms/pellet below 0.7 m/s. Key techniques for the development were a mesh structured acceleration component for removing gas sublimated from the pellet and a funnel with an appropriate angle connected just behind the acceleration chamber for introducing the pellet to plasma without destruction. Using the mesh structured components, the horizontal angular distribution of pellets ejected became narrow, because irregular pellet motion caused by sublimated gas was reduced. To investigate the performance of the injector, pellet injection experiments from the low magnetic field side (LFS) were conducted using ohmic heating plasmas. Central fueling and enhanced fueling rate have been observed. Dα intensity around the divertor region was reduced in a pellet injection plasma compared to gas puffing, indicating low recycling rate was maintained with the pellet injection. (author)

  15. Impact of pellet-cladding interaction on fuel integrity: a status report

    International Nuclear Information System (INIS)

    Pankaskie, P.J.

    1978-02-01

    There appears to be a general consensus that pellet/cladding interaction (PCI) is one of the principal limitations on reactor core power cycling. The economic importance of PCI, as fuel service limiting, is evidenced by the fact that all USLWR fuel suppliers impose some operating restrictions and/or recommendations on rates and magnitudes of power increases for both startup and demand load response modes of operation. In contrast to the economic aspects of PCI, there does not appear to be a similar attitude with regard to the safety significance of PCI in operating USLWRs. The apparent incidence of PCI failures accompanying a transient increase in core/rod power, however, provides a basis for some system safety conern. The predominant role of the economics of PCI failures has led to the individual development, by USLWR fuel suppliers, of specific operating recommendations for minimization of PCI fuel failures under more or less normal operation

  16. Pellet injector development and experiments at ORNL

    International Nuclear Information System (INIS)

    Baylor, L.R.; Argo, B.E.; Barber, G.C.; Combs, S.K.; Cole, M.J.; Dyer, G.R.; Fehling, D.T.; Fisher, P.W.; Foster, C.A.; Foust, C.R.; Gouge, M.J.; Jernigan, T.C.; Langley, R.A.; Milora, S.L.; Qualls, A.L.; Schechter, D.E.; Sparks, D.O.; Tsai, C.C.; Wilgen, J.B.; Whealton, J.H.

    1993-01-01

    The development of pellet injectors for plasma fueling of magnetic confinement fusion experiments has been under way at Oak Ridge National Laboratory (ORNL) for the past 15 years. Recently, ORNL provided a tritium-compatible four-shot pneumatic injector for the Tokamak Fusion Test Reactor (TFTR) based on the in situ condensation technique that features three single-stage gas guns and an advanced two-stage light gas gun driver. In another application, ORNL supplied the Tore Supra tokamak with a centrifuge pellet injector in 1989 for pellet fueling experiments that has achieved record numbers of injected pellets into a discharge. Work is progressing on an upgrade to that injector to extend the number of pellets to 400 and improve pellet repeatability. In a new application, the ORNL three barrel repeating pneumatic injector has been returned from JET and is being readied for installation on the DIII-D device for fueling and enhanced plasma performance experiments. In addition to these experimental applications, ORNL is developing advanced injector technologies, including high-velocity pellet injectors, tritium pellet injectors, and long-pulse feed systems. The two-stage light gas gun and electron-beam-driven rocket are the acceleration techniques under investigation for achieving high velocity. A tritium proof-of-principle (TPOP) experiment has demonstrated the feasibility of tritium pellet production and acceleration. A new tritium-compatible, extruder-based, repeating pneumatic injector is being fabricated to replace the pipe gun in the TPOP experiment and will explore issues related to the extrudability of tritium and acceleration of large tritium pellets. The tritium pellet formation experiments and development of long-pulse pellet feed systems are especially relevant to the International Tokamak Engineering Reactor (ITER)

  17. Specification of PWR UO2 pellet design parameters with the fuel performance code FRAPCON-1

    International Nuclear Information System (INIS)

    Silva, A.T.; Marra Neto, A.

    1988-08-01

    UO 2 pellet design parameters are analysed to verify their influence in the fuel basic properties and in its performance under irradiation in pressurized water reactors. Three groups of parameters are discussed: 1) content of fissionable and impurity materials; 2) stoichiometry; 3) density pore morpholoy, and microstructure. A methodology is applied with the fuel performance program FRAPCON-1 to specify these parameters. (author [pt

  18. Effects of fuel particle size and fission-fragment-enhanced irradiation creep on the in-pile behavior in CERCER composite pellets

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yunmei [Institute of Mechanics and Computational Engineering, Department of Aeronautics and Astronautics, Fudan University, Shanghai 200433 (China); Ding, Shurong, E-mail: dsr1971@163.com [Institute of Mechanics and Computational Engineering, Department of Aeronautics and Astronautics, Fudan University, Shanghai 200433 (China); Zhang, Xunchao; Wang, Canglong; Yang, Lei [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China)

    2016-12-15

    The micro-scale finite element models for CERCER pellets with different-sized fuel particles are developed. With consideration of a grain-scale mechanistic irradiation swelling model in the fuel particles and the irradiation creep in the matrix, numerical simulations are performed to explore the effects of the particle size and the fission-fragment-enhanced irradiation creep on the thermo-mechanical behavior of CERCER pellets. The enhanced irradiation creep effect is applied in the 10 μm-thick fission fragment damage matrix layer surrounding the fuel particles. The obtained results indicate that (1) lower maximum temperature occurs in the cases with smaller-sized particles, and the effects of particle size on the mechanical behavior in pellets are intricate; (2) the first principal stress and radial axial stress remain compressive in the fission fragment damage layer at higher burnup, thus the mechanism of radial cracking found in the experiment can be better explained. - Highlights: • A grain-scale gas swelling model considering the development of recrystallization and resolution is adopted for particles. • The influence of fission-gas-induced porosity is considered in the constitutive relations for particles. • A simulation method is developed for the multi-scale thermo-mechanical behavior. • The effects of fuel particle size and fission-fragment-enhanced irradiation creep are investigated in pellets.

  19. PBX/TFTR pellet program PPPL

    International Nuclear Information System (INIS)

    Schmidt, G.

    1986-01-01

    Goals, current results and plans for pellet injection work for the PBX and TFTR programs are outlined. The present PBX injector is a prototype for ORNL 4 pellet condensing injectors. It has demonstrated that pellet injection on PBX can be used to increase overall density and alter the density profile. Future PBX operation requires reliable operation in deuterium and tritium, multiple pellet capability and ability to vary the size of pellets. These goals will require the construction of a new injector similar to the TFTR DPI system. It has also been demonstrated that pellets can efficiently fuel TFTR, producing a clean, high density plasma. Issues which are still outstanding include isotope exchange effects, use of different pellet sizes, optimization of pellet density perturbations and pellet penetration at high beam power

  20. High burn-up structure in nuclear fuel: impact on fuel behavior - 4005

    International Nuclear Information System (INIS)

    Noirot, J.; Pontillon, Y.; Zacharie-Aubrun, I.; Hanifi, K.; Bienvenu, P.; Lamontagne, J.; Desgranges, L.

    2016-01-01

    When UO 2 and (U,Pu)O 2 fuels locally reach high burn-up, a major change in the microstructure takes place. The initial grains are replaced by thousands of much smaller grains, fission gases form micrometric bubbles and metallic fission products form precipitates. This occurs typically at the rim of the pellets and in heterogeneous MOX fuel Pu rich agglomerates. The high burn-up at the rim of the pellets is due to a high capture of epithermal neutrons by 238 U leading locally to a higher concentration of fissile Pu than in the rest of the pellet. In the heterogeneous MOX fuels, this rim effect is also active, but most of the high burn-up structure (HBS) formation is linked to the high local concentration of fissile Pu in the Pu agglomerates. This Pu distribution leads to sharp borders between HBS and non-HBS areas. It has been shown that the size of the new grains, of the bubbles and of the precipitates increase with the irradiation local temperatures. Other parameters have been shown to have an influence on the HBS initiation threshold, such as the irradiation density rate, the fuel composition with an effect of the Pu presence, but also of the Gd concentration in poisoned fuels, some of the studied additives, like Cr, and, maybe some of the impurities. It has been shown by indirect and direct approaches that HBS formation is not the main contributor to the increase of fission gas release at high burn-up and that the HBS areas are not the main source of the released gases. The impact of HBS on the fuel behavior during ramp on high burn-up fuels is still unclear. This short paper is followed by the slides of the presentation

  1. Fracture of Zircaloy cladding by interactions with uranium dioxide pellets in LWR fuel rods. Technical report 10

    International Nuclear Information System (INIS)

    Smith, E.; Ranjan, G.V.; Cipolla, R.C.

    1976-11-01

    Power reactor fuel rod failures can be caused by uranium dioxide fuel pellet-Zircaloy cladding interactions. The report summarizes the current position attained in a detailed theoretical study of Zircaloy cladding fracture caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn-up. It is shown that stress corrosion crack growth in irradiated Zircaloy must be able to proceed at very low stress intensifications if uniform friction effects are operative at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs (i.e., ''uniform'' conditions). Otherwise, the observed fuel rod failures must be due to departures from ''uniform'' conditions, and a very high interfacial friction coefficient and particularly fuel-cladding bonding, are means of providing sufficient stess intensification at a cladding crack tip to explain the occurrence of cladding fractures. The results of the investigation focus attention on the necessity for reliable experimental data on the stress corrosion crack growth behavior of irradiated Zircaloy, and for further investigations on the correlation between local fuel-cladding bonding and stress corrosion cracking

  2. Development of Chemical Technology in Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Jee, Kwang Yong; Kim, W. H.; Kim, J. S.

    2007-06-01

    This project mainly concentrates on the development of technologies related to elemental analysis for the mass balance of pyro-chemical process, on the development of in-line measurement system for high temperature molten salt, and on the development of radiation shielded LA-ICP-MS and micro-XRD system to evaluate the integrity of nuclear fuel. Chemical analysis methods for the quantitative determination of fissile elements, minor actinide elements, fission products, chemical additive and corrosion products in Uranium Metal Ingots are established. It will be applied to the evaluation of mass balance in electrolytic reduction process for the optimization of the process. Optical fiber based UV-VIS spectrophotometer combined with reaction cell was developed for the measurement of reactions in high temperature molten salt. This system is applicable to in-line monitoring of electro-refining process and contribute to clarify the chemical reactions. Radiation shielded LA-ICP-MS and micro-XRD systems are planned to be used for the analysis of isotopic distribution and structural changes from core to rim of spent nuclear fuel pellet, respectively. The developed techniques can contribute to produce database needed for authorization and practical use of ultra high burn-up fuel. In addition, it can be applicable to the other industries such as microelectronics, nano material science and semiconductor to analyze micro region

  3. Fuel Pellets from Biomass. Processing, Bonding, Raw Materials

    DEFF Research Database (Denmark)

    Stelte, Wolfgang

    in an increasing interest in biomass densification technologies, such as pelletization and briquetting. The global pellet market has developed quickly, and strong growth is to be expected for the coming years. Due to an increasing demand for biomass, the traditionally used wood residues from sawmills and pulp...... influence of the different processing parameters on the pressure built up in the press channel of a pellet mill. It showed that the major factor was the press channel length as well as temperature, moisture content, particle size and extractive content. Furthermore, extractive migration to the pellet...... surface at an elevated temperature played an important role. The second study presented a method of how key processing parameters can be estimated, based on a pellet model and a small number of fast and simple laboratory trials using a single pellet press. The third study investigated the bonding...

  4. A review on the development of the MOX fuel fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, See Hyung; Lee, Yung Woo; Sohn, Dong Sung; Yang, Myung Seung; Bae, Kee Kwang; Nah, Sang Hoh; Kim, Han Soo; Lee, Jung Won; Kim, Bong Koo; Song, Keun Woo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    Development of the Mixed Oxide(MOX) fuel fabrication technology was reviewed in this study. Firstly, the feasibility of Pu utilization for nuclear fuel was analyzed by comparison of nuclear characteristics between U and Pu. Secondly, the feature and problem of processes developed so far was revealed and analyzed by reviewing each process in terms of technical difficulties and in connection with the pellet characteristics. Also, fabrication facilities currently existing were analyzed to understand particularities and circumstances in view of Pu handling, and finally, in-reactor behaviors of MOX fuel was compared with those of U fuel to understand how the Pu has an effect on fuel was compared with those of U fuel to understand how the Pu has an effect on fuel pellet structure and fuel rod. 73 figs., 15 tabs., 58 refs. (Author).

  5. Wood pellet use in Sweden. A systems approach to the residential sector

    International Nuclear Information System (INIS)

    Vinterbaeck, Johan

    2000-01-01

    This empirically based thesis deals with a biofuel market in a systems context with focus on Sweden. Fuel pellets is a new consumer market for wood products. Initially used mainly by large-scale heating plants, wood pellets expanded into the Swedish residential heating market in the mid 1990s. The overall aim of this work is to provide a deeper understanding of the system for small-scale use of densified wood fuels. The objective was to provide a mapping and logistic analysis of fuel and delivery chains primarily for wood pellets. The description includes both technical as well as economic and organisational aspects. The thesis in particular investigates (i) experience from practical densification operations in the past, (ii) wood pellet retailers in Sweden, (iii) wood pellet consumers in Austria, Sweden and the United States, (iv) imports of wood pellets, and (v) forecasting of pellet consumption and inventory management for wood pellet distributors. Previous international studies revealed that the availability of cheap raw materials for fuel production and the price and availability of the most important competing fuels: coal, oil and natural gas were important factors that have guided production and use of densified wood and bark fuels. A major network of wood pellet distributors was mapped. It was concluded from a survey to these retailers that the Swedish residential market was now firmly in place and that the price of wood pellets was competitive with prices of traditional national fuels. A majority of pellet users in Austria, Sweden and the United States were pleased with pellet heating. One way to improve pellet distribution systems would be to optimise inventory management. An internal model for optimising inventory management, Pell-Sim, was constructed. For Sweden, wood pellets in 1997 represented the second most traded biofuel assortment, with 4.35 PJ or 18% of the total biofuel imports. Contrary to trade with other biofuel assortments, wood pellet trade

  6. Analysis of pellet cladding mechanical interaction using computational simulation

    Energy Technology Data Exchange (ETDEWEB)

    Berretta, José R.; Suman, Ricardo B.; Faria, Danilo P.; Rodi, Paulo A., E-mail: jose.berretta@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LabRisco/USP), São Paulo, SP (Brazil). Laboratório de Análise, Avaliação e Gerenciamento de Riscos

    2017-07-01

    During the operation of Pressurized Water Reactors (PWR), specifically under power transients, the fuel pellet experiences many phenomena, such as swelling and thermal expansion. These dimensional changes in the fuel pellet can enable occurrence of contact it and the cladding along the fuel rod. Thus, pellet cladding mechanical interaction (PCMI), due this contact, induces stress increase at the contact points during a period, until the accommodation of the cladding to the stress increases. This accommodation occurs by means of the cladding strain, which can produce failure, if the fuel rod deformation is permanent or the burst limit of the cladding is reached. Therefore, the mechanical behavior of the cladding during the occurrence of PCMI under power transients shall be investigated during the fuel rod design. Considering the Accident Tolerant Fuel program which aims to develop new materials to be used as cladding in PWR, one important design condition to be evaluated is the cladding behavior under PCMI. The purpose of this paper is to analyze the effects of the PCMI on a typical PWR fuel rod geometry with stainless steel cladding under normal power transients using computational simulation (ANSYS code). The PCMI was analyzed considering four geometric situations at the region of interaction between pellet and cladding. The first case, called “perfect fuel model” was used as reference for comparison. In the second case, it was considered the occurrence of a pellet crack with the loss of a chip. The goal for the next two cases was that a pellet chip was positioned into the gap of pellet-cladding, in the situations described in the first two cases. (author)

  7. Operation results of 3-rd generation nuclear fuel WWER-440 in initial period

    International Nuclear Information System (INIS)

    Adeev, V.; Panov, A.

    2011-01-01

    On unit 4 of Kola NPP trial operation of 3-rd generation's fuel began in 2010. Fuel assemblies of 3-rd generation (FA-3) have a number of design features that provide better operational characteristics. Concise description of a design and the basic advantages of fuel of 3-rd generation are described in articles. Increasing of efficiency of nuclear fuel usage will be achieved by reduction of the parasitic capture of thermal neutrons in constructional materials (weight of zirconium is reduced), optimization of uranium-water relation (increase in fuel elements step), increasing of uranium loading (usage of fuel pellets with increased diameter and without central hole in them). By results of trial operation mass transition to use of given type of assemblies in WWER-440 is possible. This report presents the basic outcomes of the trial operation, a brief survey of the obtained data. The basic characteristics of the reactor core with fuel of 3-rd generation are resulted in work. (authors)

  8. Nuclear fuel cycle facilities and RP: the case of Brazil

    International Nuclear Information System (INIS)

    Tranjan Filho, Alfredo; Costa, Cesar Gustavo S.

    2008-01-01

    Full text: The renewed nuclear energy scenario, national and worldwide, calls for the strengthening of all activities involving the nuclear fuel production, from uranium extraction at the mines to fuel assemblies delivery at the nuclear power plants, which in Brazil is the mission of the Industrias Nucleares do Brasil (INB). With only a third of its territory prospected, Brazil currently has the sixth largest uranium reserve in the world. Brazil's three main deposits are: the Caldas mine (in the state of Minas Gerais) the first mineral-industrial complex that processed uranium, developed in 1982, and presently being decommissioned; Caetite mine and processing facility (located in the state of Bahia), nowadays operational and with a current production capacity of 400 tonnes per year of uranium concentrates, being in trend of doubling its annual capacity; and the Itataia/Santa Quiteria deposit (in Ceara State), the largest geological uranium reserve in Brazil, although its feasible future production depends on the exploration of the phosphate associated to it. Concerning the nuclear fuel fabrication, INB plant at Resende (in the state of Rio de Janeiro) is responsible for the conversion of Uf 6 to UO 2 the production of fuel pellets and the assembly of the fuel elements, in order to supply the demands of Brazil's two operating PWR (Angra 1 and Angra 2). In addition, in May 2006, INB-Resende inaugurated the uranium enrichment facility, employing the ultra-centrifugation technology. Today still in its first phase of operation, when completed the enrichment facility is intended to provide 100 percent of the domestic requirements, eventually by the year 2015. Detailing present status and future perspectives of INB, in face of the global and national renaissance of nuclear energy, this paper addresses the Radiation Protection (RP) aspects related to INB's achievements and performance, as well as the pressing future challenges to be dealt with, in order to guarantee

  9. Effect of continuous change of sintering atmosphere on the grain growth of Cr-doped UO2 pellets

    International Nuclear Information System (INIS)

    Yang, Jae Ho; Nam, Ik Hui; Kim, Jong Hun; Rhee, Young Woo; Kim, Dong Joo; Kim, Keon Sik; Song, Kun Woo

    2010-01-01

    Cr-doped UO 2 pellet is one of the promising candidates for the high burn-up fuel in commercial LWRs. Major nuclear fuel vendors of such as AREVA or Westinghouse initiated the development of Cr-doped or Cr-containing additives doped UO 2 pellets since at the mid of 90's. Now, qualification programs are on-going to provide these pellets commercially. The main characteristics of the Cr-doped pellets are large-grain and visco-plasticity. Large grain pellet can reduce the corrosive fission gas release at high burn up. Viscoplastic soft pellets can lower the pressure to a cladding caused by a thermal expansion of a pellet at an elevated temperature during transient operations. Those advantages can provide room for additional power uprates and high burnup limits. Especially, PCI resistance improvement can be achieved by enlarging the pellet grain size and enhancing the fuel deformation at an elevated temperature. In this paper, to study the effect of oxygen partial pressure on grain growth in Cr-doped UO 2 pellets, Cr- doped UO 2 samples have been sintered with and without a step-wise change of sintering atmospheres. An introduction of a step-wise variation of oxygen partial pressure during the sintering enhances the grain growth of UO 2 pellets greatly. This step-wise sintering effect has been explained in terms of a continuous increase of Cr concentration along the grain boundary. The observed grain growth behavior under step-wisely changed sintering atmospheres demonstrates the possibility of reducing the amount of Cr 2 O 3 to minimum via control of oxygen partial pressure while keeping the large grain size

  10. Wood pellets in a power plant - mixed combustion of coal and wood pellets

    International Nuclear Information System (INIS)

    Nupponen, M.

    2001-01-01

    The author reviews in his presentation the development of Turku Energia, the organization of the company, the key figures of the company in 2000, as well as the purchase of energy in 2000. He also presents the purchase of basic heat load, the energy production plants of the company, the sales of heat in 2000, the emissions of the plants, and the fuel consumption of the plants in 2000. The operating experiences of the plants are also presented. The experiences gained in Turku Energia on mixed combustion of coal and wood pellets show that the mixing ratios, used at the plants, have no effect on the burning properties of the boiler, and the use of wood pellets with coal reduce the SO 2 and NO x emissions slightly. Simultaneously the CO 2 share of the wood pellets is removed from the emissions calculations. Several positive effects were observed, including the disappearance of the coal smell of the bunker, positive publicity of the utilization of wood pellets, and the subsidies for utilization of indigenous fuels in power generation. The problems seen include the tendency of wood pellets to arc the silos, especially when the pellets include high quantities of dust, and the loading of the trucks and the pneumatic unloading of the trucks break the pellets. Additionally the wood pellets bounce on the conveyor so they drop easily from the conveyor, the screw conveyors designed for conveying grain are too weak and they get stuck easily, and static electricity is easily generated in the plastic pipe used as the discharge pipe for wood pellet (sparkling tendency). This disadvantage has been overcome by using metal net and grounding

  11. Nuclear Fuel Reprocessing

    International Nuclear Information System (INIS)

    Simpson, Michael F.; Law, Jack D.

    2010-01-01

    This is a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. Nuclear reprocessing is the chemical treatment of spent fuel involving separation of its various constituents. Principally, it is used to recover useful actinides from the spent fuel. Radioactive waste that cannot be re-used is separated into streams for consolidation into waste forms. The first known application of nuclear reprocessing was within the Manhattan Project to recover material for nuclear weapons. Currently, reprocessing has a peaceful application in the nuclear fuel cycle. A variety of chemical methods have been proposed and demonstrated for reprocessing of nuclear fuel. The two most widely investigated and implemented methods are generally referred to as aqueous reprocessing and pyroprocessing. Each of these technologies is described in detail in Section 3 with numerous references to published articles. Reprocessing of nuclear fuel as part of a fuel cycle can be used both to recover fissionable actinides and to stabilize radioactive fission products into durable waste forms. It can also be used as part of a breeder reactor fuel cycle that could result in a 14-fold or higher increase in energy utilization per unit of natural uranium. Reprocessing can also impact the need for geologic repositories for spent fuel. The volume of waste that needs to be sent to such a repository can be reduced by first subjecting the spent fuel to reprocessing. The extent to which volume reduction can occur is currently under study by the United States Department of Energy via research at various national laboratories and universities. Reprocessing can also separate fissile and non-fissile radioactive elements for transmutation.

  12. Compliance characteristics of cracked UO2 pellets

    International Nuclear Information System (INIS)

    Williford, R.E.; Mohr, C.L.; Lanning, D.D.

    1981-01-01

    The thermally induced cracking of UO 2 fuel pellets causes simultaneous reductions of the bulk (extrinsic) fuel thermal conductivity and elastic moduli to values significantly less than those for solid <