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Sample records for fuel pellet gf-47

  1. 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

  2. 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)

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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.)

  10. 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)

  11. 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

  12. 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.)

  13. 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

  14. 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

  15. 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.

  16. 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.)

  17. 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.)

  18. Proceedings: pellet fuels conference

    Energy Technology Data Exchange (ETDEWEB)

    1995-12-31

    The conference brought together professionals from the process- engineered-fuels (PEF), utility, paper, plastics, and boiler industries. Although the last two decades have produced technical breakthroughs, efforts to advance PEF must now focus on increasing commercial breakthroughs. Successful commercialization will depend on increasing supplier, consumer, and regulator confidence and support by demonstrating the performance and value of PEF products. Speakers provided updates on how PEF technology is evolving with respect to technical, economic, and regulatory challenges. Actions critical toward full commercialization of PEF were then considered. Discussion groups addressed materials sourcing, fuel processing and transportation, combustion, and ash handling.

  19. 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.)

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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)

  5. 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

  6. 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.)

  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. 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.)

  9. 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

  10. 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

  11. 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.)

  12. 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)

  13. 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.

  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. 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

  16. 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

  17. 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.)

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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.

  3. 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.

  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. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  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. 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.

  16. 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.)

  17. 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.)

  18. 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

  19. 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

  20. 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)

  1. 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)

  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. 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. 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)

  5. 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)

  6. 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).

  7. 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)

  8. Models of the ablation of fuel pellets

    International Nuclear Information System (INIS)

    Rozhanskij, V.A.; Senichenkov, I.Yu.

    2005-01-01

    One performed qualitative analysis of a model of neutral screening (NS) and of neutral-and-plasma screening (NPS). One listed basic physical processes governing formation of a screening cloud and evaporation rate. For the model one presents formulae linking evaporation rate and cloud parameters with parameters of background plasma and pellet. One carried out comparative evaluation of the efficiency and showed that the major share of energy flow of background electrons was trapped in a plasma cloud. One derived formulae for evaporation rate and for plasma parameters in terms of the model. One discusses how it happens that the model of neutral screening describes pellet evaporation rate adequately [ru

  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. 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 ...

  11. 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

  12. Summary of fueling by pellet injection

    International Nuclear Information System (INIS)

    Stewart, L.D.

    1978-01-01

    Model-based studies were presented which indicated in all cases that shielding will occur, but there was not total agreement in these studies on the mechanism of the shielding. The data from the pellet ablation experiment on ORMAK was explained by considering the plasma electron flux, incident on the pellet surface, to create an ablated neutral cloud which self-consistently attenuates the incident electron flux. The lack of total agreement in the studies comes about when extending this to tokamak reactor plasmas. Various groups contended either that this mechanism would continue to dominate in reactor plasmas, or that it would be modified by a comparable heat flux from alphas, or that it would be modified somewhat by electrostatic shielding because of electron flux induced charge buildup on the pellet, or that it would be modified by ionization of the neutral cloud yielding a plasma cloud shield, or that this same plasma cloud would exclude magnetic field causing deflection of the incident electron flux and therefore additional shielding

  13. 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.)

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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)

  19. 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

  20. 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.)

  1. 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

  2. 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.)

  3. 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)

  4. 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

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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.)

  10. 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

  11. 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)

  12. 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)

  13. 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

  14. 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)

  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. Microencapsulation and fabrication of fuel pellets for inertial confinement fusion

    International Nuclear Information System (INIS)

    Nolen, R.L. Jr.; Kool, L.B.

    1981-01-01

    Various microencapsulation techniques were evaluated for fabrication of thermonuclear fuel pellets for use in existing experimental facilities studying inertial confinement fusion and in future fusion-power reactors. Coacervation, spray drying, in situ polymerization, and physical microencapsulation methods were employed. Highly spherical, hollow polymeric shells were fabricated ranging in size from 20 to 7000 micron. In situ polymerization microencapsulation with poly(methyl methacrylate) provided large shells, but problems with local wall defects still must be solved. Extension to other polymeric systems met with limited success. Requirements for inertial confinement fusion targets are described, as are the methods that were used

  17. 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

  18. 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

  19. 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)

  20. 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.

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. Development of recycling processes for clean rejected MOX fuel pellets

    International Nuclear Information System (INIS)

    Khot, P.M.; Singh, G.; Shelke, B.K.; Surendra, B.; Yadav, M.K.; Mishra, A.K.; Afzal, Mohd.; Panakkal, J.P.

    2014-01-01

    Highlights: • Dry and wet (MWDD) methods were developed for 100% recycling of CRO (0.4–44% PuO 2 ). • Dry method showed higher productivity and comparable powder/product characteristics. • MWDD batches demonstrated improved powder/product characteristics to that of virgin. • Second/multiple recycling is possible with MWDD with better powder/product characteristics. • MWDD batches prepared by little milling showed better macroscopic homogeneity to that of virgin. - Abstract: The dry and wet recycling processes have been developed for 100% recycling of Clean Reject Oxide (CRO) generated during the fabrication of MOX fuel, as CRO contains significant amount of plutonium. Plutonium being strategic material need to be circumvented from its proliferation issues related to its storage for long period. It was difficult to recycle CRO containing higher Pu content even with multiple oxidation and reduction steps. The mechanical recycling comprising of jaw crushing and sieving has been coupled with thermal pulverization for recycling CRO with higher Pu content in dry recycling technique. In wet recycling, MicroWave Direct Denitration (MWDD) technique has been developed for 100% recycling of CRO. The powder prepared by dry and wet (MWDD) recycling techniques was characterized by XRD and BET techniques and their effects on the pellets were evaluated. (U,21%Pu)O 2 pellets fabricated from virgin powder and MWDD were characterized using optical microscopy and α-autoradiography and the results obtained were compared

  8. 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.

  9. 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.

  10. 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

  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. 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)

  13. 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.)

  14. 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.

  15. 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)

  16. 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)

  17. 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)

  18. 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.

  19. 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)

  20. 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...

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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.

  6. 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

  7. 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...

  8. 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.

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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)

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  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. 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.

  5. 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

  6. 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)

  7. 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)

  8. 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.)

  9. 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.

  10. 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)

  11. 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.)

  12. 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.)

  13. 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

  14. 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)

  15. 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)

  16. 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

  17. 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)

  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. 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. 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

  1. 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)

  2. 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)

  3. The quality analyses of olive cake fuel pellets - mathematical approach

    Directory of Open Access Journals (Sweden)

    Brlek Tea I.

    2016-01-01

    Full Text Available This article investigates the effect of processing parameters (conditioning temperature and binder content, on final quality of produced agro-pellets for heat energy generation, obtained from four different olive cultivars using different technological parameters. Technological, physical and chemical properties of pellets (carbon, hydrogen, nitrogen and sulphur content, particle density, abrasion length, moisture, ash content, higher and lower heating values, fixed carbon and volatile matter content have been determined to assess their quality. The performance of Artificial Neural Network (ANN was compared with the performance of second order polynomial (SOP model, as well as with the obtained experimental data in order to develop rapid and accurate mathematical model for prediction of final quality parameters of agro-pellets. SOP model showed high coefficients of determination (r2, between 0.692 and 0.955, while ANN model showed high prediction accuracy with r2 between 0.544 and 0.994. [Projekat Ministarstva nauke Republike Srbije, br. III 46005 i br. TR-31055

  4. 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

  5. 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.

  6. 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

  7. 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)

  8. 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

  9. 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

  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. 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)

  12. 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)

  13. 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

  14. 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)

  15. 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

  16. 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

  17. Procedure for the gasification of pelletized carbonaceous fuels

    Energy Technology Data Exchange (ETDEWEB)

    Ban, T.E.; Sheppard, J.C.; Marlowe, W.H.

    1980-08-07

    A continuous, travelling grate device is used for the production of low Btu gas with high hydrogen and carbon monoxide content from coke pellets. For the initiation of an oxidation zone the surface of one of the layers of a horizontally moved coal bed with a layer of a sorted recycled coal charge and a layer of fresh coal is ignited. The zone migrates in the form of a wave into the lower and upper layer reducing the coal which is to be found in zones in front of the forward migrating oxidation zones. The reactions are stopped before the reaction zones reaches the both extreme surfaces of the beds.

  18. 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

  19. A study of the effectiveness of hand protection when handling UO2 fuel pellets

    International Nuclear Information System (INIS)

    Washington, R.R.; Sullivan, D.F.

    1981-01-01

    Simple tests were performed to estimate the effectiveness of various forms of hand protection in reducing skin doses when handling UO 2 fuel pellets. Household rubber gloves (rubberized cotton) appeared to be the most effective of the varieties tested. Nylon gloves and latex finger cots were least effective. (author)

  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. 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

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

    DEFF Research Database (Denmark)

    Stelte, Wolfgang; Clemons, Craig; Holm, Jens K.

    2012-01-01

    and a high concentration of hydrophobic waxes on its outer surface that may limit the pellet strength. The present work studies the impact of the lignin glass transition on the pelletizing properties of wheat straw. Furthermore, the effect of surface waxes on the pelletizing process and pellet strength...... are investigated by comparing wheat straw before and after organic solvent extraction. The lignin glass transition temperature for wheat straw and extracted wheat straw is determined by dynamic mechanical thermal analysis. At a moisture content of 8%, transitions are identified at 53°C and 63°C, respectively....... Pellets are pressed from wheat straw and straw where the waxes have been extracted from. Two pelletizing temperatures were chosen—one below and one above the glass transition temperature of lignin. The pellets compression strength, density, and fracture surface were compared to each other. Pellets pressed...

  3. 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

  4. 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)

  5. 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)

  6. 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

  7. Fuel pellets from biomass: The importance of the pelletizing pressure and its dependency on the processing conditions

    DEFF Research Database (Denmark)

    Stelte, Wolfgang; Holm, Jens K.; Sanadi, Anand R.

    2011-01-01

    The aim of the present study was to identify the key factors affecting the pelletizing pressure in biomass pelletization processes. The impact of raw material type, pellet length, temperature, moisture content and particle size on the pressure build up in the press channel of a pellet mill...... act as lubricants, lowering the friction between the biomass and the press channel walls. The effect of moisture content on the pelletizing pressure was dependent on the raw material species. Different particle size fractions, from below 0.5 mm up to 2.8 mm diameter, were tested, and it was shown...

  8. 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

  9. 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)

  10. 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

  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. 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.

  13. 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)

  14. 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

  15. 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

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

    African Journals Online (AJOL)

    Lotter;Msola Hunter;Straub

    Biochar production averaged 59 and 29% of total fuel in the ND and Philips, respectively. Interviews of 30 ND TLUD stove users showed that 60% abandoned use within one month, 80% stating that they produce too much smoke and 40% stating that controlling the air vent is too much trouble. Seventy five percent said that ...

  17. 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)

  18. 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)

  19. Fuel pellets and optical systems for inertially confined fusion

    International Nuclear Information System (INIS)

    Hendricks, C.D.

    1979-01-01

    Current laser-driven ICF targets are complex sets of concentric spherical shells made from a variety of materials including the fuel (e.g., deuterium-tritium), glass, beryllium, gold, polymeric materials, organo-metallics, and several additional organic and inorganic materials depending on the particular experiments to be done. While it is not yet known what the reactor targets will be exactly, there is little reason to believe they will be just simple, low quality glass shells containing DT gas or simple spheres of deuterated polyethylene or other fuel. Consequently, many of the current targets, materials, and fabrication techniques are considered to be applicable to the long range problems of ICF reactor target fabrication. Many current material problems and fabrication techniques are discussed and various quality factors are presented in an attempt to bring an awareness of the possible fusion reactor target materials problems to the scientific and technical community

  20. 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.)

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

    Directory of Open Access Journals (Sweden)

    Petrović-Bećirović Sanja B.

    2015-01-01

    Full Text Available 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 obtained showed that in case of nominal load and combustion of pellets recommended by the stove manufacturer, stove efficiency of 80.03% was achieved. The use of lower quality pellet caused additional 1.13 kW reduction in heat output in case of nominal load and 0.63 kW in case of reduced load. This was attributed to less favourable properties and lower bulk and particle density of lower quality pellet. The use of different burner capacity has shown to have little effect on heat output and efficiency of the stove when pre-set values in the control system of the stove were not altered. It is concluded that replacement of the burner only is not sufficient to increase/decrease the declared capacity of the same stove configuration, meaning that additional measures are necessary. These measures include a new set up of the stove control system, which needs to be properly adjusted for each alteration in stove configuration. Without the adjustment mentioned, declared capacity of the stove cannot be altered, while its CO emission shall be considerably increased.

  2. 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.)

  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. 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.

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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.)

  10. 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)

  11. 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)

  12. 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).

  13. 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)

  14. 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

  15. 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.

  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. AGR fuel pin pellet-clad interaction failure limits and activity release fractions

    International Nuclear Information System (INIS)

    Hughes, H.; Hargreaves, R.

    1985-01-01

    The limiting conditions beyond which pellet-clad interaction can flail AGR fuel are described. They have been determined by many experiments involving post-irradiation examination and testing, loop experiments and cycling and up-rating of both individual fuel stringers and the whole WAGR core. The mechanisms causing this interaction are well understood and are quantitatively expressed in computer codes. Strain concentration effects over fuel cracks determine power cycling endurance while additional strain concentrations at clad ridges and from cross pin temperature gradients contribute to up-rating failures. An equation summarising tube burst test data so as to determine the ductility available at any transient is given. The hollow fuel and more ductile clad of the Civil AGR fuel pins leads to a much improved performance over the original fuel design. The Civil AGRs operate well within these limiting conditions and substantial increases beyond the design burn-up are confidently expected. The activity release on pin failure and its development during continued operation of failed fuel have also been investigated. A retention of radioiodine and caesium of 90-99% compared to the noble gases has been demonstrated. Measured fission gas releases into the free volume of Civil AGR fuel pins have been very low (< 0.1%)

  18. 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

  19. 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)

  20. Chemical aspects of pellet-cladding interaction in light water reactor fuel elements

    International Nuclear Information System (INIS)

    Olander, D.R.

    1982-01-01

    In contrast to the extensive literature on the mechanical aspects of pellet-cladding interaction (PCI) in light water reactor fuel elements, the chemical features of this phenomenon are so poorly understood that there is still disagreement concerning the chemical agent responsible. Since the earliest work by Rosenbaum, Davies and Pon, laboratory and in-reactor experiments designed to elucidate the mechanism of PCI fuel rod failures have concentrated almost exclusively on iodine. The assumption that this is the reponsible chemical agent is contained in models of PCI which have been constructed for incorporation into fuel performance codes. The evidence implicating iodine is circumstantial, being based primarily upon the volatility and significant fission yield of this element and on the microstructural similarity of the failed Zircaloy specimens exposed to iodine in laboratory stress corrosion cracking (SCC) tests to cladding failures by PCI

  1. 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)

  2. 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

  3. 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.

  4. Effect of fuel zinc content on toxicological responses of particulate matter from pellet combustion in vitro

    Energy Technology Data Exchange (ETDEWEB)

    Uski, O., E-mail: oskari.uski@uef.fi [University of Eastern Finland, Department of Environmental Science, P.O. Box 1627, FI-70211 Kuopio (Finland); National Institute for Health and Welfare, Department of Environmental Health, P.O. Box 95, FI-70701 Kuopio (Finland); Jalava, P.I., E-mail: pasi.jalava@uef.fi [University of Eastern Finland, Department of Environmental Science, P.O. Box 1627, FI-70211 Kuopio (Finland); Happo, M.S., E-mail: mikko.happo@uef.fi [University of Eastern Finland, Department of Environmental Science, P.O. Box 1627, FI-70211 Kuopio (Finland); Torvela, T., E-mail: tiina.torvela@uef.fi [University of Eastern Finland, Department of Environmental Science, P.O. Box 1627, FI-70211 Kuopio (Finland); Leskinen, J., E-mail: jani.leskinen@uef.fi [University of Eastern Finland, Department of Environmental Science, P.O. Box 1627, FI-70211 Kuopio (Finland); Mäki-Paakkanen, J., E-mail: jorma.maki-paakkanen@thl.fi [National Institute for Health and Welfare, Department of Environmental Health, P.O. Box 95, FI-70701 Kuopio (Finland); Tissari, J., E-mail: jarkko.tissari@uef.fi [University of Eastern Finland, Department of Environmental Science, P.O. Box 1627, FI-70211 Kuopio (Finland); Sippula, O., E-mail: olli.sippula@uef.fi [University of Eastern Finland, Department of Environmental Science, P.O. Box 1627, FI-70211 Kuopio (Finland); Lamberg, H., E-mail: heikki.lamberg@uef.fi [University of Eastern Finland, Department of Environmental Science, P.O. Box 1627, FI-70211 Kuopio (Finland); Jokiniemi, J., E-mail: jorma.jokiniemi@uef.fi [University of Eastern Finland, Department of Environmental Science, P.O. Box 1627, FI-70211 Kuopio (Finland); VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT, Espoo (Finland); and others

    2015-04-01

    Significant amounts of transition metals such as zinc, cadmium and copper can become enriched in the fine particle fraction during biomass combustion with Zn being one of the most abundant transition metals in wood combustion. These metals may have an important role in the toxicological properties of particulate matter (PM). Indeed, many epidemiological studies have found associations between mortality and PM Zn content. The role of Zn toxicity on combustion PM was investigated. Pellets enriched with 170, 480 and 2300 mg Zn/kg of fuel were manufactured. Emission samples were generated using a pellet boiler and the four types of PM samples; native, Zn-low, Zn-medium and Zn-high were collected with an impactor from diluted flue gas. The RAW 264.7 macrophage cell line was exposed for 24 h to different doses (15, 50,150 and 300 μg ml{sup −1}) of the emission samples to investigate their ability to cause cytotoxicity, to generate reactive oxygen species (ROS), to altering the cell cycle and to trigger genotoxicity as well as to promote inflammation. Zn enriched pellets combusted in a pellet boiler produced emission PM containing ZnO. Even the Zn-low sample caused extensive cell cycle arrest and there was massive cell death of RAW 264.7 macrophages at the two highest PM doses. Moreover, only the Zn-enriched emission samples induced a dose dependent ROS response in the exposed cells. Inflammatory responses were at a low level but macrophage inflammatory protein 2 reached a statistically significant level after exposure of RAW 264.7 macrophages to ZnO containing emission particles. ZnO content of the samples was associated with significant toxicity in almost all measured endpoints. Thus, ZnO may be a key component producing toxicological responses in the PM emissions from efficient wood combustion. Zn as well as the other transition metals, may contribute a significant amount to the ROS responses evoked by ambient PM. - Highlights: • Zinc powder was added into the

  5. Effect of fuel zinc content on toxicological responses of particulate matter from pellet combustion in vitro

    International Nuclear Information System (INIS)

    Uski, O.; Jalava, P.I.; Happo, M.S.; Torvela, T.; Leskinen, J.; Mäki-Paakkanen, J.; Tissari, J.; Sippula, O.; Lamberg, H.; Jokiniemi, J.

    2015-01-01

    Significant amounts of transition metals such as zinc, cadmium and copper can become enriched in the fine particle fraction during biomass combustion with Zn being one of the most abundant transition metals in wood combustion. These metals may have an important role in the toxicological properties of particulate matter (PM). Indeed, many epidemiological studies have found associations between mortality and PM Zn content. The role of Zn toxicity on combustion PM was investigated. Pellets enriched with 170, 480 and 2300 mg Zn/kg of fuel were manufactured. Emission samples were generated using a pellet boiler and the four types of PM samples; native, Zn-low, Zn-medium and Zn-high were collected with an impactor from diluted flue gas. The RAW 264.7 macrophage cell line was exposed for 24 h to different doses (15, 50,150 and 300 μg ml −1 ) of the emission samples to investigate their ability to cause cytotoxicity, to generate reactive oxygen species (ROS), to altering the cell cycle and to trigger genotoxicity as well as to promote inflammation. Zn enriched pellets combusted in a pellet boiler produced emission PM containing ZnO. Even the Zn-low sample caused extensive cell cycle arrest and there was massive cell death of RAW 264.7 macrophages at the two highest PM doses. Moreover, only the Zn-enriched emission samples induced a dose dependent ROS response in the exposed cells. Inflammatory responses were at a low level but macrophage inflammatory protein 2 reached a statistically significant level after exposure of RAW 264.7 macrophages to ZnO containing emission particles. ZnO content of the samples was associated with significant toxicity in almost all measured endpoints. Thus, ZnO may be a key component producing toxicological responses in the PM emissions from efficient wood combustion. Zn as well as the other transition metals, may contribute a significant amount to the ROS responses evoked by ambient PM. - Highlights: • Zinc powder was added into the pure

  6. 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

  7. 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.)

  8. QC methods and means during pellets and fuel rods manufacturing at JSC 'MSZ'

    International Nuclear Information System (INIS)

    Kouznetsov, A.I.

    2000-01-01

    The report contains the description of the main methods and devices used in fabrication of pellets and fuel rods to prove their conformity to the requirements of technical specifications. The basic principals, range and accuracy of methods and devices are considered in detail, as well as system of metrological support of measurements. The latter includes the metrological certification and periodical verification of the devices, metrological qualification of measurement procedures, standard samples provision and checking the correctness of the analyses performance. If one makes an overall review of testing methods used in different fuel production plants he will find that most part of methods and devices are very similar. There are still some variations in methods which could be a subject for interesting discussions among specialists. This report contains a brief review of testing methods and devices used at our plant. More detailed description is given to methods which differ from those commonly used. (author)

  9. 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)

  10. Pu-rich MOX agglomerate-by-agglomerate model for fuel pellet burnup analysis

    International Nuclear Information System (INIS)

    Chang, G.S.

    2004-01-01

    In support of potential licensing of the mixed oxide (MOX) fuel made from weapons-grade (WG) plutonium and depleted uranium for use in United States reactors, an experiment containing WG-MOX fuel is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The WG-MOX comprises five percent PuO 2 and 95% depleted UO 2 . Based on the Post Irradiation Examination (PIE) observation, the volume fraction (VF) of MOX agglomerates in the fuel pellet is about 16.67%, and PuO 2 concentration of 30.0 = (5 / 16.67 x 100) wt% in the agglomerate. A pressurized water reactor (PWR) unit WG-MOX lattice with Agglomerate-by-Agglomerate Fuel (AbAF) modeling has been developed. The effect of the irregular agglomerate distribution can be addressed through the use of the Monte Carlo AbAF model. The AbAF-calculated cumulative ratio of Agglomerate burnup to U-MAtrix burnup (AG/MA) is 9.17 at the beginning of life, and decreases to 2.88 at 50 GWd/t. The MCNP-AbAF-calculated results can be used to adjust the parameters in the MOX fuel fission gas release modeling. (author)

  11. 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)

  12. 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

  13. 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

  14. 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)

  15. 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)

  16. 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

  17. Correlation between UO2 powder and pellet quality in PHWR fuel manufacturing

    International Nuclear Information System (INIS)

    Glodeanu, F.; Spinzi, M.; Balan, V.

    1988-01-01

    Natural uranium dioxide fuel for heavy water reactors has a series of very tightly controlled quality factors: Chemical purity, density and microstructures. Although the fabrication history may consistently affect the fuel quality, the quality factor mentioned above are function mainly of the quality of the powder used as raw material. As regards the fulfilment of the requirements for very high density of the pellets, it was found that in a definite technology the raw material plays the decisive part. Except for the powder sinterability, one found other important subtile parameters, such as the degree of agglomeration and structural homogeneity. The fuel microstructure, very important for in-serive performances of the fuel, is related to a great extent to some powder characteristics (homogeneity, sinterability). This is why much stress was laid on UO 2 power quality evaluation both by standard methods and non-conventional ones (agglomeration, microscopy, X-rays). Some of the characteristics defined by product specification, such as powder sinterability, should be better defined to guarantee the final product quality. (orig.)

  18. Handbook for Small-Scale Densified Biomass Fuel (Pellets) Manufacturing for Local Markets.

    Energy Technology Data Exchange (ETDEWEB)

    Folk, Richard L.; Govett, Robert L.

    1992-07-01

    Wood pellet manufacturing in the Intermountain West is a recently founded and rapidly expanding energy industry for small-scale producers. Within a three-year period, the total number of manufacturers in the region has increased from seven to twelve (Folk et al., 1988). Small-scale industry development is evolving because a supply of raw materials from small and some medium-sized primary and secondary wood processors that has been largely unused. For the residue producer considering pellet fuel manufacturing, the wastewood generated from primary products often carries a cost associated with residue disposal when methods at-e stockpiling, landfilling or incinerating. Regional processors use these methods for a variety of reasons, including the relatively small amounts of residue produced, residue form, mixed residue types, high transportation costs and lack of a local market, convenience and absence of regulation. Direct costs associated with residue disposal include the expenses required to own and operate residue handling equipment, costs for operating and maintaining a combustor and tipping fees charged to accept wood waste at public landfills. Economic and social costs related to environmental concerns may also be incurred to include local air and water quality degradation from open-air combustion and leachate movement into streams and drinking water.

  19. Demonstration of fuel resistant to pellet-cladding interaction. Phase 2. First semiannual report, January-June 1979

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.

    1979-08-01

    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 first semiannual progress report for Phase 2 of this program (January-June 1979). Progress in the irradiation testing of barrier fuel and of unfueled barrier cladding specimens is reported

  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. 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)

  2. Device of measuring hydrogen in U-H-Zr fuel pellet

    International Nuclear Information System (INIS)

    Li Ming; Guo Qiaoru; Xu Xiuqing

    1992-01-01

    A device used for the determination of hydrogen content in U-H-Zr fuel pellet and its precision determination are introduced. The influence of different flow of carry gas, extracting temperature and collecting time on the determination of hydrogen content is researched separately. While the flow of carry gas is 85-100 mL/min and extracting temperature increasing to 1550 deg C, the best collecting time may be reduced to 18 min from 30 min, the determination of zirconium hydrogenate reference standard sample shows that the precision of six time determination is smaller than ±2%, determination error is 2%. The rate of recovery of hydrogen is 99.3%. The test shows that the device and its determination method are stable and operation is simple

  3. Cradle-to-Gate Life Cycle Assessment of Switchgrass Fuel Pellets Manufactured in the Southeastern United States

    Science.gov (United States)

    R. D. Bergman; D. L. Reed; A. M. Taylor; D. P. Harper; D. G. Hodges

    2015-01-01

    Developing renewable energy sources with low environmental impacts is becoming increasingly important as concerns about consuming fossil fuel sources grow. Cultivating, harvesting, drying, and densifying raw biomass feedstocks into pellets for easy handling and transport is one step forward in this endeavor. However, the corresponding environmental performances must be...

  4. 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).

  5. 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

  6. 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

  7. 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)

  8. Production of fuel pellets made of biomass. Saett vid framstaellning av pelleterat braensle av biomassa

    Energy Technology Data Exchange (ETDEWEB)

    Haeffner, E; Miller, A; Thrap-Olsen, O

    1986-11-17

    The pellets consists of dewatered peat mixed with milled biomass into a workable compound. The compound is led to a storage tank through which hot air from a heat exchanger is blown. The pre-dried compound then passes a pellet press. The pellets are moved to a heat insulated dryer with a perforated bottom through which hot gas is blown and when the pellets have a sufficient percentage of moisture they are cooled by blowing cool gas through the drier, thus producing durable pellets. (L.F.).

  9. 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.

  10. Effects of MnO-Al2O3 on the grain growth and high-temperature deformation strain of UO2 fuel pellets

    International Nuclear Information System (INIS)

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

    2010-01-01

    The fabrication and high-temperature deformation strain of MnO-Al 2 O 3 -doped UO 2 pellets were studied. The effects of additive composition and amount on the microstructure evolution of a UO 2 pellet were investigated. The compressive creep behaviors of MnO-Al 2 O 3 -doped UO 2 pellets were examined. The results indicated that a MnO-Al 2 O 3 binary additive can effectively promote the grain growth of UO 2 pellets. In addition, the high-temperature deformation strain of the UO 2 pellet can be improved significantly with 1,000 ppm 95MnO-5Al 2 O 3 (mol%). The developed MnO-Al 2 O 3 -additive-containing UO 2 pellets can be a potential candidate for a high-burn-up fuel and a pellet-cladding interaction (PCI) remedy. (author)

  11. Spent fuel UO2 matrix corrosion behaviour studies through alpha-doped UO2 pellets leaching

    International Nuclear Information System (INIS)

    Muzeau, B.; Jegou, C.; Broudic, V.

    2005-01-01

    Full text of publication follows: The option of direct disposal of spent nuclear fuel in a deep geological formation raises the need to investigate the long-term behaviour of the UO 2 matrix in aqueous media subjected to α-β-γ radiations. The β-γ emitters account for the most of the activity of spent fuel at the moment it is removed from the reactor, but diminish within a millennial time frame by over three orders of magnitude to less than the long-term activity. The latter persist over much longer time periods and must therefore be taken into account over geological disposal scale. In the present investigation the UO 2 matrix corrosion under alpha radiation is studied as a function of different parameters such as: the alpha activity, the carbonates and hydrogen concentrations,.. In order to study the effect of alpha radiolysis of water on the UO 2 matrix, 238/239 Pu doped UO 2 pellets (0.22 %wt. Pu total) were fabricated with different 238 Pu/ 239 Pu ratio to reproduce the alpha activity of a 47 GWd.t HMi -1 UOX spent fuel at different milestones in time (15, 50, 1500, 10000 and 40000 years). Undoped UO 2 pellets were also available as reference sample. Leaching experiments were conducted in deionized or carbonated water (NaHCO 3 1 mM), under Argon (O 2 2 30% gas mixture. Previous experiments conducted in deionized water under argon atmosphere, have shown a good correlation between alpha activity and uranium release for the 15-, 1500- and 40000-years alpha doped UO 2 batches. Besides, uranium release in the leachate is controlled either by the kinetics, or by the thermodynamics. Provided the solubility limit of uranium is not achieved, uranium concentration increases and is only limited by the kinetics, unless precipitation occurs and the uranium concentration remains constant over time. These controls are highly dependant on the solution chemistry (HCO 3 - , pH, Eh,..), the atmosphere (Ar, Ar/H 2 ,..), and the radiolysis strength. The experimental matrix

  12. The effect of the volumetric heat source distribution of the fuel pellet on the minimum DNBR ratio

    International Nuclear Information System (INIS)

    Hordosy, G.; Kereszturi, A.; Maroti, L.; Trosztel, I.

    1995-01-01

    The radial power distribution in a VVER-440 type fuel assembly is strongly non-uniform as a result of the water-gap between the shrouds and the moderator filled central tube. Consequently, it can be expected that the power density inside a single fuel rod is inhomogeneous, as well. In the paper the methodology and the results of coupled thermohydraulic and neutronic calculations are presented. The objective of the analysis was the investigation of the heat source distribution and the determination of the possible extent of the power non-uniformity in a corner rod which has always the highest peaking factor in a VVER-440 type assembly. The results of the analysis revealed that there can be a strong non-uniformity of power distribution inside a fuel pellet, and the effect depends first of all on the general assembly conditions, while the local subchannel parameters have only a slight influence on the pellet heat source distribution. (author)

  13. 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)

  14. Development of an integrated, unattended assay system for LWR-MOX fuel pellet trays

    International Nuclear Information System (INIS)

    Stewart, J.E.; Hatcher, C.R.; Pollat, L.L.

    1994-01-01

    Four identical unattended plutonium assay systems have been developed for use at the new light-water-reactor mixed oxide (LWR-MOX) fuel fabrication facility at Hanau, Germany. The systems provide quantitative plutonium verification for all MOX pellet trays entering or leaving a large, intermediate store. Pellet-tray transport and storage systems are highly automated. Data from the ''I-Point'' (information point) assay systems will be shared by the Euratom and International Atomic Energy Agency (IAEA) Inspectorates. The I-Point system integrates, for the first time, passive neutron coincidence counting (NCC) with electro-mechanical sensing (EMS) in unattended mode. Also, provisions have been made for adding high-resolution gamma spectroscopy. The system accumulates data for every tray entering or leaving the store between inspector visits. During an inspection, data are analyzed and compared with operator declarations for the previous inspection period, nominally one month. Specification of the I-point system resulted from a collaboration between the IAEA, Euratom, Siemens, and Los Alamos. Hardware was developed by Siemens and Los Alamos through a bilateral agreement between the German Federal Ministry of Research and Technology (BMFT) and the US DOE. Siemens also provided the EMS subsystem, including software. Through the USSupport Program to the IAEA, Los Alamos developed the NCC software (NCC COLLECT) and also the software for merging and reviewing the EMS and NCC data (MERGE/REVIEW). This paper describes the overall I-Point system, but emphasizes the NCC subsystem, along with the NCC COLLECT and MERGE/REVIEW codes. We also summarize comprehensive testing results that define the quality of assay performance

  15. 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

  16. 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

  17. 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

  18. 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.

  19. 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.)

  20. Handling system for nuclear fuel cans to a fuel pellet feeder

    International Nuclear Information System (INIS)

    Vere, B.; Mathevon, P.

    1985-01-01

    The handling system comprises a first array of conveyors which takes a batch of casings from a delivery rack, alters the spacing between the casings, and delivers them to a vibrating table feeder, a second array of conveyors which readjusts the spacing between casing to its initial value and transfers the casings to a removal rack, and automatic and synchronized control means for ensuring the displacements of casings always in the same direction. The increase of spacing between casings can be used, before feeding, to allow them to be weighed one after the other, and after feeding, for cleaning the end part of fuel cans [fr

  1. 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

  2. Fission distribution measurements of Atucha's fuel pellets with solid state track detectors

    International Nuclear Information System (INIS)

    Ricabarra, M.D. Bovisio de; Waisman, Dina.

    1979-08-01

    Distribution of fissions in a UO 2 rod has been measured by means of solid state detectors. Mica muscovite and Makrofol-N detectors were used in the experiment. The merits of mica muscovite relative to the Makrofol-N for the detection of fission fragments have been verified. However both fission track detectors closely agree (0,5%) in the final fission distribution of the UO 2 rod. Sensitivity of the detectors shows to be linear in the range between 50.000and 360.000 fission tracks per square centimeter. Due to the high spatial resolution this method is better than any other technique. Determination were made in UO 2 pellets similar to the fuel element of the Atucha reactor. The average fission rate in the rod has been measured within 0,8% error, and provides an accurate determination for the distribution of fissions in the rod wich is needed for the determination of energy liberated per fission in the natural uranium rod.(author) [es

  3. Application of railgun principle to high-velocity hydrogen pellet injection for magnetic fusion reactor fueling

    International Nuclear Information System (INIS)

    Kim, K.; Zhang, J.

    1992-01-01

    Three separate papers are included which report research progress during this period: (1) A new railgun configuration with perforated sidewalls, (2) development of a fuseless small-bore railgun for injection of high-speed hydrogen pellets into magnetically confined plasmas, and (3) controls and diagnostics on a fuseless railgun for solid hydrogen pellet injection

  4. Sintering control of fuel pellets from U O2 microspheres using Master Mix concept

    International Nuclear Information System (INIS)

    Assis, Gino de; Santos, Armindo

    1995-01-01

    Preliminary results of a pelletizing experiment with two different density Kernels are presented. The Kernels fabricated by the Sol-Gel process on several proportions were mixed. This procedure is called Master Mix. The blends were pelleted and some aspects are commented. (author). 4 refs., 4 figs., 2 tabs

  5. 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.

  6. 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.

  7. 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.

  8. Modelling of pellet cladding interaction during power ramps in PWR rods by means of Transuranus fuel rod analysis code

    International Nuclear Information System (INIS)

    Di Marcello, V.; Luzzi, L.

    2008-01-01

    Pellet-cladding interaction (PCI) in PWR type rods subjected to power ramps was analysed by means of TRANSURANUS (TU) fuel rod performance code. PCI phenomena depend on the fuel power history - i.e. by several irradiation and thermal induced phenomena occurring in the fuel rod and mutually interacting during its life in reactor - and may become critical for cladding integrity under accidental conditions. Ten test fuel rods, whose power histories and post irradiation experiment (PIE) data were available from the OECD/NEA-IAEA International Fuel Performance Experiment (UTE) database through the Studsvik SUPER-RAMP Project, were simulated by TRANSURANUS. During a power ramp pellet gaseous swelling can be inhibited by cladding pressure and can be over-predicted by a normal operation swelling model. This phenomenon was simulated by a new formulation of a fuel swelling model already available in the code, in order to consider hot pressing of inter-granular -fuel porosity due to the high hydrostatic stress resulting from PCI: it was found that TRANSURANUS, as a result of the proposed swelling formulation as well as of the accurate modelling of the other phenomena occurring during irradiation, gives correct predictions on PCI induced fuel rod failures. In addition, PCI failure threshold identified by TRANSURANUS was compared with the technological limits known in literature: the possibility of relaxing these limits for low burn-up values and the preponderance of the European fuel rod design in front of PCI emerged from TU analyses. Finally, a good agreement was found between TU evaluations and PIE data, with regard to fission gas release, fuel grain growth, and creep, corrosion and elongation of the cladding. (authors)

  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. Digital image processing: Cylindrical surface plane development of CAREM fuel pellets

    International Nuclear Information System (INIS)

    Caccavelli, J; Cativa Tolosa, S; Gommes, C

    2012-01-01

    As part of the development of fuel pellets (FPs) for nuclear reactor CAREM-25, is necessary to systematize the analysis of the mechanical integrity of the FPs that is now done manually by a human operator. Following specifications and standards of reference for this purpose, the FPs should be inspected visually for detecting material discontinuities in the FPs surfaces to minimize any deterioration, loss of material and excessive breakage during operation and load of fuel bars. The material discontinuities are classified into two defects: surface cracks and chips. For each of these surface defects exist acceptance criteria that determine if the fuel pellet (FP) as a whole is accepted or rejected. One criteria for surface cracks is that they do not exceed one third (1/3) of the circumferential surface of the FP. The FP has cylindrical shape, so some of these acceptance criteria make difficult to analyze the FP in a single photographic image. Depending on the axial rotation of the FP, the crack could not be entirely visualized on the picture frame. Even a single crack that appears in different parts of the FP rotated images may appear to be different cracks in the FP when it is actually one. For this reason it is necessary, for the automatic detection and measurement of surface defects, obtain the circumferential surface of the FP into a single image in order to decide the acceptance or reject of the FP. As the FP shape is cylindrical, it is possible to obtain the flat development of the cylindrical surface (surface unrolling) of the FPs into a single image combining the image set of the axial rotation of the FP. In this work, we expose the procedure to implement the flat development of the cylindrical surface (surface unrolling). Starting from a photographic image of the FP surface, which represents the projection of a cylinder in the plane, we obtain three-dimensional information of each point on the cylindrical surface of the FP (3D-mapping). Then, we can

  11. A fast-acting hydrogen gas source for staged pneumatic high-speed acceleration of fusion plasma fuel pellets

    International Nuclear Information System (INIS)

    Andersen, S.A.; Baekmark, L.

    1990-02-01

    This report describes a possible design of a fast, high-temperature, arc-driven hydrogen gas source module, to be used in a scheme for multistage high-speed pneumatic acceleration of fusion plasma fuel pellets. The potential of this scheme for operating with a moderate driving pressure at long acceleration path lengths is particular attractive for accelerating fragile hydrogen isotope ice pellets. From experiments with an ethanol-based arc unit, design parameters for a propeller module were assessed, and with a barrel-mounted ethanol module staged pneumatic acceleration of a plastic dummy pellet was demonstrated. In experiments with a hydrogenbased, cryogenic arc unit in which 200 joules of electrical energy were dissipated with a power level approaching 5 MW within 30 mus, the velocity of a 23-mg plastic pellet was increased from 1.7 to 2.4 km/s. Results in terms of barrel pressure transients and arc characteristics are described. (author) 20 ills., 8 refs

  12. Non-destructive studies of fuel pellets by neutron resonance absorption radiography and thermal neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Tremsin, A.S., E-mail: ast@ssl.berkeley.edu [University of California, Berkeley, CA 94720 (United States); Vogel, S.C.; Mocko, M.; Bourke, M.A.M.; Yuan, V.; Nelson, R.O.; Brown, D.W. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Feller, W.B. [NOVA Scientific, Inc., 10 Picker Rd., Sturbridge, MA 01566 (United States)

    2013-09-15

    fuel assemblies with intentionally introduced defects was investigated. The maps of elemental composition of pellets containing urania and tungsten were obtained simultaneously by resonance absorption imaging with spatial resolution better than ∼200 μm, while the voids and cracks were revealed by the transmission images obtained with thermal and cold neutrons. Our proof-of-principle experiments demonstrate that simultaneous acquisition of resonance and Bragg edge spectra enables concurrent mapping of isotope distributions, imaging of cracks and voids as well as measurements of some crystallographic parameters of fuel assemblies and their cladding. A detailed study of energy-dependent neutron statistics achievable at FP5 with our present detection system is also presented for a wide range of neutron energies.

  13. 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

  14. 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

  15. Development of a two-stage light gas gun to accelerate hydrogen pellets to high speeds for plasma fueling applications

    International Nuclear Information System (INIS)

    Combs, S.K.; Milora, S.L.; Foust, C.R.; Gouge, M.J.; Fehling, D.T.; Sparks, D.O.

    1988-01-01

    The development of a two-stage light gas gun to accelerate hydrogen isotope pellets to high speeds is under way at Oak Ridge National Laboratory. High velocities (>2 km/s) are desirable for plasma fueling applications, since the faster pellets can penetrate more deeply into large, hot plasmas and deposit atoms of fuel directly in a larger fraction of the plasma volume. In the initial configuration of the two-stage device, a 2.2-l volume (/ 3 for frozen hydrogen isotopes). However, the use of sabots to encase and protect the cryogenic pellets from the high peak pressures will probably be required to realize speeds of ∼3 km/s or greater. The experimental plan includes acceleration of hydrogen isotopes as soon as the gun geometry and operating parameters are optimized; theoretical models are being used to aid in this process. The hardware is being designed to accommodate repetitive operation, which is the objective of this research and is required for future applications. 25 refs., 6 figs., 1 tab

  16. Grain growth in thoria and thoria-base fuel pellets (LWBR development program)

    Energy Technology Data Exchange (ETDEWEB)

    Smid, R.J.

    1976-01-01

    The kinetics of grain growth in ThO/sub 2/-base sintered compacts were investigated to determine the cause of a nonuniform microstructural cross section. It was concluded that trace impurities which inhibit continuous grain growth at the pellet interior were removed by vaporization at the pellet exterior. This resulted in relatively normal grain growth at the pellet surface and discontinuous grain growth at the pellet interior. Calcining the starting ThO/sub 2/ powder to a slightly higher temperature removed inhibiting impurities but also decreased the driving force for grain growth by reducing the surface area of the powder. Mixing high and low temperature calcined ThO/sub 2/ resulted in improved grain growth. Increased oxygen partial pressure and temperature during sintering increased grain boundary mobility in spite of the inhibiting impurity. The specific inhibiting impurity was not isolated during this investigation.

  17. 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

  18. 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

  19. A new UO2 sintering technology for the recycling of defective fuel pellets

    International Nuclear Information System (INIS)

    Song, K. W.; Kim, K. S.; Jeong, Y. H.

    1998-01-01

    A new UO 2 sintering technology to recycle defective UO 2 pellets has been developed. The defective UO 2 pellets were oxidized in an air to produce U 3 O 8 powder, and the U 3 O 8 powder was mixed with fresh AUC-UO 2 powder in the range of 10 to 100 wt%. Nb 2 O 5 and TiO 2 are added to the mixed powder. The mixed powder was pressed and sintered at 1680 deg C for 4 hours in hydrogen. The density of UO 2 pellets without sintering agents decreased linearly with the U 3 O 8 content at the rate of 0.2 %TD per 1 wt% U 3 O 8 , and the density was below 93.5 %TD at the U 3 O 8 contents above 10 wt%. However, the mixed UO 2 and U 3 O 8 powder containing Nb 2 O 5 (≥0.3 wt%) and TiO 2 (≥0.1 wt%) yielded a sintered density above 94 %TD in all ranges of U 3 O 8 contents. It was found that higher mixing ratios of U 3 O 8 to UO 2 powder did not affect the grain size of UO 2 pellets under the addition of Nb 2 O 5 , but decreased the grain size of UO 2 pellets under the addition of TiO 2 . The doped UO 2 pellets have grain sizes larger than 20 μm, and have small density gain after re-sintering test, owing to large pores. Therefore, the sintering agents such as Nb 2 O 5 and TiO 2 can make highly densified UO 2 pellets from the powder comprising a large amount of U 3 O 8 powder

  20. Remote fabrication of (Th, {sup 233}U)O{sub 2} pellet-type fuels for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A

    1981-05-15

    Thorium fuels enriched with {sup 233}U must be fabricated in shielded cells because of high gamma and alpha activity. A conceptual design of a remotely operated plant to produce gamma-active pellet fuels has been made. The plant consists of eight fabrication canyons, two repair canyons, and several miscellaneous cells. Process equipment is modular, easily disconnected, and mounted on plates for easy removal. Equipment consists of a combination of robotics, hard automation, and conventional process equipment. The plant is operated from a central control room with the assistance of a sophisticated computer-based control and information system. Many of the automated process steps are preprogrammed on the control computer and executed on demand by the supervising operator. The technology to build such a plant exists today but needs to be adapted to the needs of the recycle fuel industry. (author)

  1. A model for predicting pellet-cladding interaction induced fuel rod failure, based on nonlinear fracture mechanics

    International Nuclear Information System (INIS)

    Jernkvist, L.O.

    1993-01-01

    A model for predicting pellet-cladding mechanical interaction induced fuel rod failure, suitable for implementation in finite element fuel-performance codes, is presented. Cladding failure is predicted by explicitly modelling the propagation of radial cracks under varying load conditions. Propagation is assumed to be due to either iodine induced stress corrosion cracking or ductile fracture. Nonlinear fracture mechanics concepts are utilized in modelling these two mechanisms of crack growth. The novelty of this approach is that the development of cracks, which may ultimately lead to fuel rod failure, can be treated as a dynamic and time-dependent process. The influence of cyclic loading, ramp rates and material creep on the failure mechanism can thereby be investigated. Results of numerical calculations, in which the failure model has been used to study the dependence of cladding creep rate on crack propagation velocity, are presented. (author)

  2. Fracture toughness and fracture surface energy of sintered uranium dioxide fuel pellets

    International Nuclear Information System (INIS)

    Kutty, T.R.G.; Chandrasekharan, K.N.; Panakkal, J.P.; Ghosh, J.K.

    1987-01-01

    The paper concerns the variation of fracture toughness Ksub(ic) and fracture surface energy γsub(s) in sintered uranium dioxide pellets in the density range 9.86 to 10.41 g cm -3 , using Vickers indentation technique. A minimum of four indentations were made on each pellet sample and the average crack length of each indentation and the hardness values were determined. The overall average crack-length datra and the data on volume fraction porosity in the pellets fitted a straight line, from which Ksub(ic) and γsub(s) were calculated. The fracture parameters of nonporous polycrystalline UO 2 , calculated from the experimental data, are presented in tabular form. (U.K.)

  3. 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).

  4. 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

  5. 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

  6. 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

  7. Dry recovery test of plutonium-uranium mixed oxide fuel pellets

    International Nuclear Information System (INIS)

    Kinugasa, Manabu; Kawamata, Kazuhiko; Kashima, Sadamitsu

    1981-01-01

    The oxidation conditions for pulverizing directly Pu-U mixed oxide pellets without mechanical crushing were examined to simplify the process and to reduce radiation exposure during the dry recovery of highly enriched Pu pellets. The specimens used were the Pusub(0.3) Usub(0.7) Osub(2-x) pellets with different density, which were sintered at 1650 deg C for 2 hours under an atmosphere of 5 % H 2 - N 2 . The oxidation experiment was carried out under several conditions. The oxidation products were examined by weight gain, X-ray diffraction, appearance pictures, SEM photographs and so on. From these studies, it can be concluded that the oxidation in NO 2 diluted with air was very powerful, but if only the coarse spalling of Pusub(0.3) Usub(0.7) O 2 sintered pellets is required, it is sufficient to oxidize them in air for 1 hr in a temperature range from 400 to 600 deg C. (Asami, T.)

  8. The dissolution of unirradiated UO2 fuel pellets under simulated disposal conditions

    International Nuclear Information System (INIS)

    Ollila, K.; Leino-Forsman, H.

    1993-03-01

    The dissolution behaviour of unirradiated UO 2 pellets was studied as a function of water composition under oxidizing and reducing conditions at 25 deg C. The waters included deionized water as the reference water, sodium bicarbonate solutions with varying bicarbonate content, and two different synthetic groundwaters. The release of uranium was measured during static batch dissolution experiments of long duration (3-4 years)

  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. 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

  11. Tritium pellet injector results

    International Nuclear Information System (INIS)

    Fisher, P.W.; Bauer, M.L.; Baylor, L.R.; Deleanu, L.E.; Fehling, D.T.; Milora, S.L.; Whitson, J.C.

    1988-01-01

    Injection of solid tritium pellets is considered to be the most promising way of fueling fusion reactors. The Tritium Proof-of- Principle (TPOP) experiment has demonstrated the feasibility of forming and accelerating tritium pellets. This injector is based on the pneumatic pipe-gun concept, in which pellets are formed in situ in the barrel and accelerated with high-pressure gas. This injector is ideal for tritium service because there are no moving parts inside the gun and because no excess tritium is required in the pellet production process. Removal of 3 He from tritium to prevent blocking of the cryopumping action by the noncondensible gas has been demonstrated with a cryogenic separator. Pellet velocities of 1280 m/s have been achieved for 4-mm-diam by 4-mm-long cylindrical tritium pellets with hydrogen propellant at 6.96 MPa (1000 psi). 10 refs., 10 figs

  12. Nuclear fuel technology - Determination of the O/M ratio in MOX pellets - Gravimetric method

    International Nuclear Information System (INIS)

    2008-01-01

    This International Standard describes a method for determining the oxygen-to-metal (O/M) ratio in mixed uranium-plutonium oxide pellets. The (U,Pu)O 2 x sample is submitted to controlled oxidation-reduction under thermodynamic conditions designed to change the O/M ratio to a value of 2,000. The initial stoichiometric deviation, X, is determined from the sample mass difference before and after heat treatment

  13. An evaluation of the influence of fuel design parameters and burnup on pellet/cladding interaction for boiling water reactor fuel rod through in-core diameter measurement

    International Nuclear Information System (INIS)

    Yanagisawa, K.

    1986-01-01

    The influence of design parameters and burning on pellet/cladding interaction (PCI) of current boiling water reactor fuel rods was studied through in-core diameter measurement. Thinner cladding and a smaller diametral gap enhanced the PCI during startup. At constant power, fuel with SiO 2 added greatly reduced PCI due to relaxation. The fuel with a small grain size greatly reduced PCI due to densification. Preirradiation of rods up to 23 MWd/kgU caused a large PCI not only in a small gap but also in a large gap rod. Relaxation and permanent deformation was small. In the power increase experiment, one rod experienced PCI failure. The spurt times of coolant radioactivity coincided well with the sudden drop of cladding axial strain and marked crack opening at the rod surface. The estimated hoop stress predicted by FEMAXI-III was 350 MPa at the failure

  14. 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.

  15. 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

  16. Method of improving the green strength of nuclear fuel pellets, and products thereof

    International Nuclear Information System (INIS)

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

    1984-01-01

    This invention provides a method of preparing an admixture comprising a particulate material and a fugitive binder for producing green pellets free of flaws and having improved strength, comprising the steps of: a) fluidizing and agitating a mass of particulate material with a fluidized bed system; b) adding a fugitive binder to the fluidizing and agitating mass of particulate material and blending the binder with the particulate material, said fugitive binder being comprised of ammonium bicarbonate, ammonium carbonate, ammonium bicarbonate, and mixtures thereof; c) aging the blended binder and particulate material for a period of greater than 48 hours; and d) forming the resulting aged blend by pressing into a green body

  17. ThO2-based pellet fuels - their properties, methods of fabrication, and irradiation performance: a critical assessment of the state of the technology and recommendations for further work

    Energy Technology Data Exchange (ETDEWEB)

    Hart, P.E.; Griffin, C.W.; Hsieh, K.A.; Matthews, R.B.; White, G.D.

    1979-09-01

    This critical assessment of the ThO/sub 2/-UO/sub 2/ pellet fuel technology was conducted in support of the Fuels Refabrication and Development Program (FRAD). Included in this critical review are the following areas: powder preparation; pellet fabrication; fuel chemical, physical, and mechanical properties; and fuel irradiation performance. The authors identify (1) areas where data are either deficient or lacking and (2) requirements for additional development and experimental work.

  18. ThO2-based pellet fuels - their properties, methods of fabrication, and irradiation performance: a critical assessment of the state of the technology and recommendations for further work

    International Nuclear Information System (INIS)

    Hart, P.E.; Griffin, C.W.; Hsieh, K.A.; Matthews, R.B.; White, G.D.

    1979-09-01

    This critical assessment of the ThO 2 -UO 2 pellet fuel technology was conducted in support of the Fuels Refabrication and Development Program (FRAD). Included in this critical review are the following areas: powder preparation; pellet fabrication; fuel chemical, physical, and mechanical properties; and fuel irradiation performance. The authors identify (1) areas where data are either deficient or lacking and (2) requirements for additional development and experimental work

  19. 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

  20. 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

  1. The influence of design and fuel parameters on the particle emissions from wood pellets combustion. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Wiinikka, Henrik; Gebart, Rikard [Energy Technology Centre, Piteaa (Sweden)

    2005-02-01

    Combustion of solid biomass under fixed bed conditions is a common technique to generate heat and power in both small and large scale grate furnaces (domestic boilers, stoves, district heating plants). Unfortunately, combustion of biomass will generate particle emissions containing both large fly ash particles and fine particles that consist of fly ash and soot. The large fly ash particles have been produced from fusion of non-volatile ash-forming species in burning char particle. The inorganic fine particles have been produced from nucleation of volatilised ash elements (K, Na, S, Cl and Zn). If the combustion is incomplete, soot particles are also produced from secondary reaction of tar. The particles in the fine fraction grows by coagulation and coalescence to a particle diameter around 0.1 pm. Since the smallest particles are very hard to collect in ordinary cleaning devices they contribute to the ambient air pollution. Furthermore, fine airborne particles have been correlated to adverse effects on the human health. It is therefore essential to minimize particle formation from the combustion process and thereby reduce the emissions of particulates to the ambient air. The aim with this project is to study particle emissions from small scale combustion of wood pellets and to investigate the impact of different operating, construction and fuel parameters on the amount and characteristic of the combustion generated particles. To address these issues, experiments were carried out in a 10 kW updraft fired wood pellets reactor that has been custom designed for systematic investigations of particle emissions. In the flue gas stack, particle emissions were sampled on a filter. The particle mass and number size distributions were analysed by a low pressure cascade impactor and a SMPS (Scanning Electron Mobility Particle Sizer). The results showed that the temperature and the flow pattern in the combustion zone affect the particle emissions. Increasing combustion

  2. 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.

  3. 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)

  4. 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.

  5. 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)

  6. 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.

  7. Fuel rod D07/B15 from Ringhals 2 PWR: Source material for corrosion/leach tests in groundwater. Fuel rod/pellet characterization program. Pt. 1

    International Nuclear Information System (INIS)

    Forsyth, R.

    1987-03-01

    A joint SKB/STUDSVIK experimental program to determine the corrosion rates and to establish the corrosion mechanisms of spent UO 2 fuel in groundwater under both oxidizing and reducing conditions is in progress in the Hot Cell Laboratory of Studsvik Energiteknik AB. High burnup fuel of both BWR and PWR type are studied. Characterization of the spent fuel at both rod and pellet level is an important part of the experimental program. Experiments on PWR fuel have been concentrated so far on specimens from one rod, manufacturer's number 03688, which had occupied position B15 in assembly D07. This assembly had been irradiated for 5 cycles in the Ringhals 2 reactor between 1977 and 1983. The calculated assembly burnup was 41.3 MWd/kg U. The present report is a collection of separate reports describing those items in the characterization program which have been performed so far. No overall summary of the experimental results is given here, and the report should be viewed as a collection of reference data. (orig.)

  8. Spent fuel UO{sub 2} matrix corrosion behaviour studies through alpha-doped UO{sub 2} pellets leaching

    Energy Technology Data Exchange (ETDEWEB)

    Muzeau, B.; Jegou, C.; Broudic, V. [CEA-Valrho DEN/DTCD/SECM Laboratoire des Materiaux et Procedes Actifs BP 17171 F-30207 Bagnols-sur-Ceze cedex (France)

    2005-07-01

    Full text of publication follows: The option of direct disposal of spent nuclear fuel in a deep geological formation raises the need to investigate the long-term behaviour of the UO{sub 2} matrix in aqueous media subjected to {alpha}-{beta}-{gamma} radiations. The {beta}-{gamma} emitters account for the most of the activity of spent fuel at the moment it is removed from the reactor, but diminish within a millennial time frame by over three orders of magnitude to less than the long-term activity. The latter persist over much longer time periods and must therefore be taken into account over geological disposal scale. In the present investigation the UO{sub 2} matrix corrosion under alpha radiation is studied as a function of different parameters such as: the alpha activity, the carbonates and hydrogen concentrations,.. In order to study the effect of alpha radiolysis of water on the UO{sub 2} matrix, {sup 238/239}Pu doped UO{sub 2} pellets (0.22 %wt. Pu total) were fabricated with different {sup 238}Pu/{sup 239}Pu ratio to reproduce the alpha activity of a 47 GWd.t{sub HMi}{sup -1} UOX spent fuel at different milestones in time (15, 50, 1500, 10000 and 40000 years). Undoped UO{sub 2} pellets were also available as reference sample. Leaching experiments were conducted in deionized or carbonated water (NaHCO{sub 3} 1 mM), under Argon (O{sub 2} < 0.1 ppm), or Ar/H{sub 2} 30% gas mixture. Previous experiments conducted in deionized water under argon atmosphere, have shown a good correlation between alpha activity and uranium release for the 15-, 1500- and 40000-years alpha doped UO{sub 2} batches. Besides, uranium release in the leachate is controlled either by the kinetics, or by the thermodynamics. Provided the solubility limit of uranium is not achieved, uranium concentration increases and is only limited by the kinetics, unless precipitation occurs and the uranium concentration remains constant over time. These controls are highly dependant on the solution chemistry

  9. 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)

  10. 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

  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. 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.

  14. Deposition of Cr, Nb, V, and Ti coatings on UO2-25w/oPuO2 fuel pellets by sputtering

    International Nuclear Information System (INIS)

    Gibby, R.L.; McClanahan, E.D.

    1976-01-01

    A sputtering deposition process was developed for application of metallic coatings on either the ends or circumferences of LMFBR mixed-oxide fuel pellets. Coatings of Cr, Nb, V and Ti were applied to over 860 pellets. Ceramography, emission spectrography, and spark source spectroscopy were used to characterize the coatings. Coating thicknesses were controlled to within +-0.0005 cm (0.0002 inch) for a coating thickness of 0.00127 cm (0.0005 inch) on the circumference and 0.00254 cm (0.001 inch) on the ends of pellets. Chemical impurities in the coatings were generally less than 0.5 wt percent. The coatings were adherent in all cases, although some interfacial separations were noted with Ti coatings. The results indicated that further optimization of coatings' parameters would result in improvement of the coatings

  15. A statistical analysis of pellet-clad interaction failures in water reactor fuel

    International Nuclear Information System (INIS)

    McDonald, S.G.; Fardo, R.D.; Sipush, P.J.; Kaiser, R.S.

    1981-01-01

    The primary objective of the statistical analysis was to develop a mathematical function that would predict PCI fuel rod failures as a function of the imposed operating conditions. Linear discriminant analysis of data from both test and commercial reactors was performed. The initial data base used encompassed 713 data points (117 failures and 596 non-failures) representing a wide variety of water cooled reactor fuel (PWR, BWR, CANDU, and SGHWR). When applied on a best-estimate basis, the resulting function simultaneously predicts approximately 80 percent of both the failure and non-failure data correctly. One of the most significant predictions of the analysis is that relatively large changes in power can be tolerated when the pre-ramp irradiation power is low, but that only small changes in power can be tolerated when the pre-ramp irradiation power is high. However, it is also predicted that fuel rods irradiated at low power will fail at lower final powers than those irradiated at high powers. Other results of the analysis are that fuel rods with high clad operating temperatures can withstand larger power increases that fuel rods with low clad operating temperatures, and that burnup has only a minimal effect on PCI performance after levels of approximately 10000 MWD/MTU have been exceeded. These trends in PCI performance and the operating parameters selected are believed to be consistent with mechanistic considerations. Published PCI data indicate that BWR fuel usually operates at higher local powers and changes in power, lower clad temperatures, and higher local ramp rates than PWR fuel

  16. Mox pellet reference material

    International Nuclear Information System (INIS)

    Perolat, J.P.

    1991-01-01

    A first batch of MOX pellets certified in plutonium and uranium has been prepared and characterised in France to meet the needs of laboratories which are engaged upon destructive analysis for safeguards purposes especially in fuel fabrication plants. The pellets sintering has been obtained in a special fabrication to achieve an homogeneity better than 0.1%. The plutonium and uranium characterisation by chemical analysis has been carried out by two laboratories using at least two different methods. 1 fig., 5 refs

  17. Fish pelleting

    African Journals Online (AJOL)

    PUBLICATIONS1

    fish meal pelletizing machine utilized 4kg of ingredients to produce 3.77kg pellets at an effi- ciency of .... Design and fabrication of fish meal pellet processing machine ... 53 ... horsepower for effective torque application on .... two edges were tacked with a spot weld to hold ... then welded on to the shaft making sure that the.

  18. An exercise to establish optimum procedures for the characterisation of porosity in UO2 fuel pellets

    International Nuclear Information System (INIS)

    Small, G.J.

    1980-05-01

    A standard metallographic preparation technique for UO 2 is proposed. The criteria for choosing the optimum route are that the specimen should be scratch-free and that the pores inherent to any sintered UO 2 pellet should be neither enlarged nor filled-in during preparation. Having met these criteria one has a specimen suitable for quantitative metallography which can be used to monitor porosity changes due to in-pile sintering. A procedure for analysing the porosity is suggested. This consists of imaging the specimen surface over a range of magnifications using both optical and Scanning Electron Microscopy in order to cover the range of pore sizes of interest (0.1 μm to 10 μm diameter). These images are then analysed to obtain figures for the distribution of pores as a function of diameter. Two methods of pore-size analysis are reviewed, the manual Zeiss Particle Size Analyser and a more sophisticated electronic instrument - the Quantimet. A comparison is made between these two instruments on the basis of accuracy, reproducibility and ease of operation. (author)

  19. Fission gas release from UO2 pellet fuel at high burn-up

    International Nuclear Information System (INIS)

    Vitanza, C.; Kolstad, E.; Graziani, U.

    1979-01-01

    Analysis of in-reactor measurements of fuel center temperature and rod internal pressure at the OECD Halden Reactor Project has led to the development of an empirical fission gas release model, which is described. The model originally derived from data obtained in the low and intermediate burn-up range, appears to give good predictions for rods irradiated to high exposures as well. PIE puncturing data from seven fuel rods, operated at relatively constant powers and peak center temperatures between 1900 and 2000 0 C up to approx. 40,000 MWd/t UO 2 , did not exhibit any burn-up enhancement on the fission gas release rate

  20. Pellet cladding interaction (PCI) fuel duty during normal operation of ASEA-ATOM BWRs

    International Nuclear Information System (INIS)

    Vaernild, O.; Olsson, S.

    1983-01-01

    Local power changes may under special conditions cause PCI fuel failures in a power reactor. By restricting the local power increase rate in certain situations it is possible to prevent PCI failures. Fine motion control rod drives, large operating range of the main recirculation pumps and an advanced burnable absorber design have minimized the impact of the PCI restrictions. With current ICFM schemes the power of an assembly is due to the burnup of the gadolinia gradually increasing during the first cycle of operation. After this the power is essentially decreasing monotonously during the remaining life of the assembly. Some assemblies are for short burnup intervals operated at very low power in control cells. The control rods in these cells may however be withdrawn without restrictions leading to energy production losses. Base load operation would in the normal case lead to very minor PCI loads on the fuel regardless of any PCI related operating restrictions. At the return to full power after a short shutdown or in connection with load follow operation, the xenon transient may cause PCI loads on the fuel. To avoid this a few hoursholdtime before going back to full power is recommended. (author)

  1. Pellet-cladding interaction (PCI) fuel duty during normal operation of ASEA-ATOM BWRs

    International Nuclear Information System (INIS)

    Vaernild, O.; Olsson, S.

    1985-01-01

    Local power changes may, under special conditions, cause PCI fuel failures in a power reactor. By restricting the local power increase rate in certain situations it is possible to prevent PCI failures. Fine motion control rod drives, large operating range of the main recirculation pumps and an advanced burnable absorber design have minimized the impact of the PCI restrictions. With current ICFM schemes the power of an assembly is due to the burnup of the gadolinia gradually increasing during the first cycle of operation. After this the power is essentially decreasing monotonously during the remaining life of the assembly. Some assemblies are for short burnup intervals operated at very low power in control cells. The control rods in these cells may, however, be withdrawn without restrictions leading to energy production losses. Base load operation would in the normal case lead to very minor PCI loads on the fuel regardless of any PCI-related operating restrictions. At the return to full power after a short shutdown or in connection with load follow operation, the xenon transient may cause PCI loads on the fuel. To avoid this a few hours hold-time before going back to full power is recommended. (author)

  2. Impact of wood pellets export on the development of their production in Serbia with the effects of substituting enegry from fossil fuels and reduction of carbon dioxide emission

    Directory of Open Access Journals (Sweden)

    Glavonjić Branko

    2016-01-01

    Full Text Available The paper presents the results of researching the impact of export on the production of wood pellets as well as the situation on the market for this wood fuel in Serbia. Objective of the research was to produce scientifically and professionally founded conclusions and the related adequate recommendations to the decision makers in order to improve the situation on wood pellets market in Serbia and eliminate the existing problems which significantly burden and slow down this development. Special objective of the research was to observe the contributions of wood pellets to the mitigation of climate changes using Serbia as the example. Results of the conducted research show that the expansion of the consumption (demand increase in the European Union countries in the last fifteen years and the related increase of export from Serbia are the most significant factors which have influenced the development of wood pellets production in Serbia. Parameters of econometric model of the impact of export on the increase of production show that production increase of 1.17% can be expected with the increase of export of 1%. Thus, the number of wood pellet producers has rapidly increased in the last ten years, from 2 producers in 2006 to 52 active producers in 2016. Increase of the number of producers was also accompanied by the increase of the installed capacities. At the end of 2015, total installed capacities for wood pellet production in Serbia reached 550 thousand tons, and the realized production was 229 thousand tons, or 41.6% of the installed capacity. Consumption of wood pellets in Serbia in the last four years achieved significant increase and reached the level of 89 thousand tons in 2015. However, concerning the segment of wood pellets consumption in Serbia, the situation is still unsatisfactory despite the fact that the consumption has been increasing year after year. Average price of 1 kWh of energy from wood pellets exported from Serbia was in the

  3. Fabrication of BN Nanosheet Reinforced ZrO{sub 2} Composite Pellets for Inert Matrix Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shukeir, Malik; Umer, Malik; Lee, Bin; Ryu, Ho Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    Plutonium also can be resulted from the dismantlement of nuclear weapons. This will result in the increase of the stockpile of plutonium. For that purpose many organizations are focusing their R-D work on the concept of Inert Matrix Fuel IMF, where a U-free matrix is used to eliminate the U-Pu conversion. R-D work was standardized around Zirconiabased IMF as a result of many screening and ranking studies performed on various candidates. Regardless of its outstanding radiation resistance, chemical stability and its high melting point, it has a very low thermal conductivity, which could be detrimental for the fuel matrix especially in case of accidents. A reinforcement phase could be used for the enhancement of the thermomechanical properties. Among many possible reinforcements, 2D structured nanosheets have emerged as an excellent candidate to enhance the thermal properties and mechanical properties simultaneously. In this approach Boron Nitride Nanosheets BNNS are used for that purpose. BNNS have a very low density, very high thermal conductivity, very high mechanical properties and high neutron absorption cross-section for Boron which is used frequently as a burnable poison. They have properties similar to graphene but they exhibit superior thermal stability in the oxide structure. Despite all the studies on other reinforcements, BNNS reinforced ZrO{sub 2} has not yet been reported. In this study, pure ZrO{sub 2} and partially stabilized Zirconia PSZ (using Yttria) ceramics are mixed with different volume fractions of BNNS.

  4. Modelling of thermal mechanical behaviour of high burn-Up VVER fuel at power transients with special emphasis on the impact of fission gas induced swelling of fuel pellets

    International Nuclear Information System (INIS)

    Novikov, V.; Medvedev, A.; Khvostov, G.; Bogatyr, S.; Kuzetsov, V.; Korystin, L.

    2005-01-01

    This paper is devoted to the modelling of unsteady state mechanical and thermo-physical behaviour of high burn-up VVER fuel at a power ramp. The contribution of the processes related to the kinetics of fission gas to the consequences of pellet-clad mechanical interaction is analysed by the example of integral VVER-440 rod 9 from the R7 experimental series, with a pellet burn-up in the active part at around 60 MWd/kgU. This fuel rod incurred ramp testing with a ramp value ΔW 1 ∼ 250 W/cm in the MIR research reactor. The experimentally revealed residual deformation of the clad by 30-40 microns in the 'hottest' portion of the rod, reaching a maximum linear power of up to 430 W/cm, is numerically justified on the basis of accounting for the unsteady state swelling and additional degradation of fuel thermal conductivity due to temperature-induced formation and development of gaseous porosity within the grains and on the grain boundaries. The good prediction capability of the START-3 code, coupled with the advanced model of fission gas related processes, with regard to the important mechanical (residual deformation of clad, pellet-clad gap size, central hole filling), thermal physical (fission gas release) and micro-structural (profiles of intra-granular concentration of the retained fission gas and fuel porosity across a pellet) consequences of the R7 test is shown. (authors)

  5. Effect of technological parameters and microstructure on mechanical strength of UO2 fuel pellets

    International Nuclear Information System (INIS)

    Radford, K.

    1980-01-01

    The effect of various peculiarities of tablet microstructure namely, sammury porosity (tablet density), grain size and pore distribution over sizes on technological parameters, is studied. It is shown that density decrease leads to a fast reduction of UO 2 tablet strength. The maximum effect on strength is produced by pore distribution over sizes, characterized by a median size, and not by the grain size, though a combined effect of those two factors is also observed. The important role of the technology of tablet production manifests itself in the fact that all operations bringing about the increase of pore or grain sizes leads to a reduction of strength. Such factors as powder origin, granule sizes, U 3 O 8 content and the amount of additions do not cause any considerable changes in the strength of tablets. Bend tests under conditions of biaxial loading should be considered as an ideal method of determining fuel tablets strength [ru

  6. Premium Fuel Production From Mining and Timber Waste Using Advanced Separation and Pelletizing Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Honaker, R. Q.; Taulbee, D.; Parekh, B. K.; Tao, D.

    2005-12-05

    The Commonwealth of Kentucky is one of the leading states in the production of both coal and timber. As a result of mining and processing coal, an estimated 3 million tons of fine coal are disposed annually to waste-slurry impoundments with an additional 500 million tons stored at a number of disposal sites around the state due to past practices. Likewise, the Kentucky timber industry discards nearly 35,000 tons of sawdust on the production site due to unfavorable economics of transporting the material to industrial boilers for use as a fuel. With an average heating value of 6,700 Btu/lb, the monetary value of the energy disposed in the form of sawdust is approximately $490,000 annually. Since the two industries are typically in close proximity, one promising avenue is to selectively recover and dewater the fine-coal particles and then briquette them with sawdust to produce a high-value fuel. The benefits are i) a premium fuel product that is low in moisture and can be handled, transported, and utilized in existing infrastructure, thereby avoiding significant additional capital investment and ii) a reduction in the amount of fine-waste material produced by the two industries that must now be disposed at a significant financial and environmental price. As such, the goal of this project was to evaluate the feasibility of producing a premium fuel with a heating value greater than 10,000 Btu/lb from waste materials generated by the coal and timber industries. Laboratory and pilot-scale testing of the briquetting process indicated that the goal was successfully achieved. Low-ash briquettes containing 5% to 10% sawdust were produced with energy values that were well in excess of 12,000 Btu/lb. A major economic hurdle associated with commercially briquetting coal is binder cost. Approximately fifty binder formulations, both with and without lime, were subjected to an extensive laboratory evaluation to assess their relative technical and economical effectiveness as binding

  7. Analysis of effects of pellet-cladding bonding on trapping of the released fission gases in high burnup KKL BWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Brankov, Vladimir [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Khvostov, Grigori; Mikityuk, Konstantin [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Pautz, Andreas [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Restani, Renato; Abolhassani, Sousan [Laboratory for Nuclear Materials at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Ledergerber, Guido [Kernkraftwerk Leibstadt, 5325 Leibstadt (Switzerland); Wiesenack, Wolfgang [Institutt for Energiteknikk - OECD Halden Reactor Project, Os Allé 5, 1777 Halden (Norway)

    2016-08-15

    Highlights: • Explanation for the scatter in measured fission gas release in high-BU BWR fuel rods. • Partial fuel-clad bond layer formation in high-BU BWR fuel. • Hypothesis for fission gas trapping facilitated by the pellet-cladding bond layer. • Correlation between burnup asymmetry and the quantity of trapped fission gas. • Implications of the trapped FG in LOCA transient. - Abstract: The first part of the paper presents results of a numerical analysis of the fuel behavior during base irradiation in the Kernkraftwerk Leibstadt Boiling Water Reactor (KKL BWR) using EPRI’s FALCON code coupled to GRSW-A – an advanced model for fuel swelling and fission gas release. Post-irradiation examinations conducted at the Paul Scherrer Institute’s (PSI) hot laboratory gave evidence of a distinct circumferential non-uniformity of local burnup at pellet surfaces. For several fuel samples, intact pellet-cladding bonding areas on the high burnup sides of the pellets at high burnup above ∼70 MWd/kgU were observed. It is hypothesized that a part of the fission gases, which are expected to be released by those areas, can be trapped and do not reach the rod plenum. In this paper, a simple approach to modeling of fission gas trapping is employed which reveals a potential correlation between the position of the rod within the fuel assembly (and therefore the degree of circumferential burnup non-uniformity) and the degree of fission gas trapping. A model is suggested to correlate the amount of locally trapped gas with the integral of the local contact pressure and the degree of circumferential burnup non-uniformity. The model is calibrated with available measurements of FGR from rod puncturing at the level of the plenums. In future work, the hypothesis about the axial distribution of trapped fission gas will be extrapolated to the Loss-Of-Coolant Accident (LOCA) analysis as an attempt to explain the fission gas release observed in some samples fabricated from

  8. Safety analyses for sodium-cooled fast reactors with pelletized and sphere-pac oxide fuels within the FP-7 European project PELGRIMM - 15386

    International Nuclear Information System (INIS)

    Maschek, W.; Andriolo, L.; Matzerath-Boccaccini, C.; Delage, F.; Parisi, C.; Del Nevo, A.; Abbate, G.; Schmitt, D.

    2015-01-01

    The European FP-7 project PELGRIMM addresses the development of Minor-Actinide (MA) bearing oxide fuel for Sodium-cooled Fast Reactors. Optionally, both MA homogeneous recycling and heterogeneous recycling is investigated with pellet and sphere-pac fuel. A first safety assessment of sphere-pac fuelled cores should be given in the Work Package 4 of the project. This assessment is in continuity with the former FP-7 CP-ESFR project. Within the CP-ESFR project the CONF2 core design has been developed characterized by a core with a large upper sodium plenum to reduce the coolant void worth. This optimized core has been chosen for the safety analyses in PELGRIMM. The task within the PELGRIMM project is thus a safety assessment of the CONF2 core loaded either with pellets or with sphere-pac fuel. The investigations started with the design of the CONF2 core with sphere-pac fuel and the determination of core safety parameters and burn-up behavior. The neutronic analyses have been performed with the MCNPX code. Variants of the CONF2 core contain up to 4% Am in the fuel. The results revealed an extended void worth (core + upper plenum) for an Am free core of 1 up to 3 dollars for the 4% Am core. Thermal-hydraulic design analyses have been performed by RELAP5-3D. The accident simulations should be performed by different codes, some of which focus on the initiation phase of the accident, as SAS4A, BELLA and the MAT5DYN code, whereas the SIMMER-III code will also deal with the later accident phases and a potential whole core melting. The codes had to be adapted to the specifics of the sphere-pac fuel, in particular to the thermal conductivity and gap conditions. Analyses showed that the safety assessment has to take into account two main phases. Starting up the core, the green fuel shows a reduced fuel thermal conductivity. After restructuring within a couple of hours, the thermal conductivity recovers and the fuel temperature decreases. The main objective of the safety analyses

  9. Fabrication of ThO2 and ThO2-UO2 pellets for proliferation resistant fuels

    International Nuclear Information System (INIS)

    Matthews, R.B.; Davis, N.C.

    1979-10-01

    To meet this objective, batches of ThO 2 powders were compared and milling parameters, pressing and sintering conditions were established. A method for blending ThO 2 and UO 2 into homogeneous powders that press and sinter into 95% TD pellets was determined. The effect of UO 2 additions on ThO 2 -UO 2 pellet properties was determined and a process for fabricating irradiation test quality ThO 2 -20 wt% UO 2 pellets containing CaO as a dissolution aid was established

  10. How the user can influence particulate emissions from residential wood and pellet stoves: Emission factors for different fuels and burning conditions

    Science.gov (United States)

    Fachinger, Friederike; Drewnick, Frank; Gieré, Reto; Borrmann, Stephan

    2017-06-01

    For a common household wood stove and a pellet stove we investigated the dependence of emission factors for various gaseous and particulate pollutants on burning phase, burning condition, and fuel. Ideal and non-ideal burning conditions (dried wood, under- and overload, small logs, logs with bark, excess air) were used. We tested 11 hardwood species (apple, ash, bangkirai, birch, beech, cherry, hickory, oak, olive, plum, sugar maple), 4 softwood species (Douglas fir, pine, spruce, spruce/fir), treated softwood, beech and oak wood briquettes, paper briquettes, brown coal, wood chips, and herbaceous species (miscanthus, Chinese silver grass) as fuel. Particle composition (black carbon, non-refractory, and some semi-refractory species) was measured continuously. Repeatability was shown to be better for the pellet stove than for the wood stove. It was shown that the user has a strong influence on wood stove emission behavior both by selection of the fuel and of the burning conditions: Combustion efficiency was found to be low at both very low and very high burn rates, and influenced particle properties such as particle number, mass, and organic content in a complex way. No marked differences were found for the emissions from different wood species. For non-woody fuels, much higher emission factors could be observed (up to five-fold increase). Strongest enhancement of emission factors was found for burning of small or dried logs (up to six-fold), and usage of excess air (two- to three-fold). Real world pellet stove emissions can be expected to be much closer to laboratory-derived emission factors than wood stove emissions, due to lower dependence on user operation.

  11. Assessment of pelletized biofuels

    International Nuclear Information System (INIS)

    Samson, R.; Duxbury, P.; Drisdelle, M.; Lapointe, C.

    2000-04-01

    There has been an increased interest in the development of economical and convenient renewable energy fuels, resulting from concerns about climate change and rising oil prices. An opportunity to use agricultural land as a means of producing renewable fuels in large quantities, relying on wood and agricultural residues only has come up with recent advances in biomass feedstock development and conversion technologies. Increasing carbon storage in the landscape and displacing fossil fuels in combustion applications can be accomplished by using switchgrass and short rotation willow which abate greenhouse gas emissions. The potential of switchgrass and short rotation willow, as well as other biomass residues as new feedstocks for the pellet industry is studied in this document. Higher throughput rates are facilitated by using switchgrass, which shows potential as a pelleting feedstock. In addition, crop drying requires less energy than wood. By taking into consideration energy for switchgrass production, transportation to the conversion facility, preprocessing, pelleting, and marketing, the overall energy balance of switchgrass is 14.5:1. Research on alfalfa pelleting can be applied to switchgrass, as both exhibit a similar behaviour. The length of chop, the application of high temperature steam and the use of a die with a suitable length/diameter ratio are all factors that contribute to the successful pelleting of switchgrass. Switchgrass has a similar combustion efficiency (82 to 84 per cent) to wood (84 to 86 per cent), as determined by combustion trials conducted by the Canada Centre for Mineral and Energy Technology (CANMET) in the Dell-Point close coupled gasifier. The energy content is 96 per cent of the energy of wood pellets on a per tonne basis. Clinker formation was observed, which necessitated some adjustments of the cleaner grate settings. While stimulating rural development and export market opportunities, the high yielding closed loop biofuels show

  12. Lab and Bench-Scale Pelletization of Torrefied Wood Chips

    DEFF Research Database (Denmark)

    Shang, Lei; Nielsen, Niels Peter K.; Stelte, Wolfgang

    2013-01-01

    Combined torrefaction and pelletization is used to increase the fuel value of biomass by increasing its energy density and improving its handling and combustion properties. In the present study, a single-pellet press tool was used to screen for the effects of pellet die temperature, moisture cont...... of the torrefied pellets was higher and the particle size distribution after grinding the pellets was more uniform compared to conventional wood pellets....

  13. 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

  14. 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.

  15. 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)

  16. 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)

  17. Production of zinc pellets

    Science.gov (United States)

    Cooper, J.F.

    1996-11-26

    Uniform zinc pellets are formed for use in batteries having a stationary or moving slurry zinc particle electrode. The process involves the cathodic deposition of zinc in a finely divided morphology from battery reaction product onto a non-adhering electrode substrate. The mossy zinc is removed from the electrode substrate by the action of gravity, entrainment in a flowing electrolyte, or by mechanical action. The finely divided zinc particles are collected and pressed into pellets by a mechanical device such as an extruder, a roller and chopper, or a punch and die. The pure zinc pellets are returned to the zinc battery in a pumped slurry and have uniform size, density and reactivity. Applications include zinc-air fuel batteries, zinc-ferricyanide storage batteries, and zinc-nickel-oxide secondary batteries. 6 figs.

  18. Materials in Sweden for future production of fuel pellets. A review of possible materials in short- and medium long-term; Raavaror foer framtida tillverkning av braenslepellets i Sverige. En kartlaeggning av taenkbara alternativa raavaror paa kort och medellaang sikt

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

    The use of fuel pellets, mainly produced from sawdust and shavings from the Swedish sawmill industry, has increased during the 1990s among small-scale users such as private houses as well as large-scale users such as thermal power stations. During the last years this increase has continued for small-scale use. Due to a significant increase of the pellet prices the last couple of years the increase for the large-scale users seems to have stopped. It is reasonable to believe that these higher prices depend mostly on lack of raw materials for the fuel pellet production. The greater part of sawdust from Swedish saw mills is used in the pellet industry, the board industry or as an internal fuel. It is reasonable to assume a small increase of present raw material available for pellet production without a further decrease in the use for the board industry. Another sawmill by-product, dry chips, may increase in importance as a fuel pellet raw material and give a small contribution while the green chips should be for use in the pulp industry only. If the use of fuel pellets should increase there is a need for new raw materials. In the short-term, thinning material and cull tree could be alternatives that could give pellets with similar characteristics as present fuel pellets. For large-scale consumers with greater ability to handle problems concerning ash, such as sintering and fouling, as well as particle and gaseous emissions a further choice of raw materials could be possible, such as bark, peat and logging residues. In the longer term energy crops could be used as well as lignin, derived from energy effective pulp industry and from possible large-scale production of ethanol from woody biomass. Nearly all of the different raw materials studied in this review have higher amounts of substances not desirable in combustion such as potassium, chlorine and nitrogen. However, pelletizing gives an unique opportunity to mix different raw materials and possible additives in order

  19. Development of a pellet cutting and loading device for the JT-60 repetitive pellet injector

    International Nuclear Information System (INIS)

    Hiratsuka, Hajime; Ichige, Hisashi; Kizu, Kaname; Iwahashi, Takaaki; Honda, Masao

    2001-03-01

    In JT-60, a pellet injector that repetitively injects deuterium pellets is under development to supply fuel to high temperature plasmas and sustain high-density plasmas. The pellet injector generates cubic pellets and accelerates them with a straight-arm rotor by centrifugal force. In this acceleration method, it is important to supply pellets reliably and stably, to prevent pellet orbits from disordering and to stabilize the launching direction. To achieve higher performance of the injector, a pellet cutting and loading device that cuts a deuterium ice rod into cubic pellets and loads them to the pellet injector successively and stably has been developed. The pellet cutting and loading device can cut a deuterium ice rod produced at low temperature of -8 Pam 3 /s, cutting time of <3 ms, cutting frequency of 1-20 Hz and cutter stroke of 2.5 mm were confirmed in the device test. In the operation test after assembling this device to the centrifugal pellet injector, the operational performance of pellet injection frequency of ∼10 Hz, pellet speed of ∼690 m/s and pellet injection duration time of ∼3.5 s was achieved. Thus, the development of the pellet cutting and loading device contributed to the upgrade of the JT-60 pellet injector. (author)

  20. Producing bio-pellets from sunflower oil cake for use as an energy source

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Yuichi; Kato, Hitoshi; Kanai, Genta; Togashi, Tatsushi [National Agricultural Research Center (Japan)], E-mail: kobay@affrc.go.jp

    2008-07-01

    Pellet fuels were produced from ground sunflower oil cake using a pelletizer. The length, hardness, and powder characteristics of dried pellets depend on the initial water content of the oil cake. The appropriate values of water contents were 19.9 - 21.0% w.b. Oil cake pellets were found to contain 6.07% ash and 20.99 MJ/kg caloric value, which are within the standard range of wood pellets. Combustion experiments using a commercial pellet stove demonstrate that oil cake pellets burn as well as wood pellets. Oil cake pellets are useful as a fuel alternative to wood pellets. (author)

  1. Advances in estimation technology of thermal conductivity of irradiated fuels (1). Application of a thermal microscope to measure the thermal conductivity of the second phases in irradiated pellets

    International Nuclear Information System (INIS)

    Uno, Masayoshi; Murakami, Yukihiro

    2011-01-01

    CeO 2 sample as a surrogate for fuel and BaCeO 3 and BaMoO 4 samples as surrogates for the second phases, which have a lower thermal conductivity than the fuel matrix, were made. The thermal conductivity of these samples was measured by a thermal microscope. In this method, the thermal conductivity of a small region (e.g. 20 μm x 20 μm) of the sample can be measured. The valid thermal conductivity values for all the samples were obtained and the conditions of sample surface preparation and the thermal microscope measurement were found out. The thermal conductivity of a CeO 2 composite pellet which had the BaCeO 3 or BaMoO 4 second phase layer was also estimated. (author)

  2. Behaviour of fission gas in the rim region of high burn-up UO2 fuel pellets with particular reference to results from an XRF investigation

    International Nuclear Information System (INIS)

    Mogensen, M.; Walker, C.T.

    1999-01-01

    XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8-54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU. (orig.)

  3. Pellet injector development at ORNL

    International Nuclear Information System (INIS)

    Milora, S.L.; Argo, B.E.; Baylor, L.R.; Cole, M.J.; Combs, S.K.; Dyer, G.R.; Fehling, D.T.; Fisher, P.W.; Foster, C.A.; Foust, C.R.; Gouge, M.J.; Jernigan, T.C.; Langley, R.A.; Qualls, A.L.; Schechter, D.E.; Sparks, D.O.; Tsai, C.C.; Whealton, J.H.; Wilgen, J.B.; Schmidt, G.L.

    1992-01-01

    Plasma fueling systems for magnetic confinement experiments are under development at Oak Ridge National Laboratory (ORNL). ORNL has recently provided a four-shot tritium pellet injector with up to 4-mm-diam capability for the Tokamak Fusion Test Reactor (TFTR). This injector, which is based on the in situ condensation technique for pellet formation, features three single-stage gas guns that have been qualified in deuterium at up to 1.7 km/s and a two-stage light gas gun driver that has been operated at 2.8-km/s pellet speeds for deep penetration in the high-temperature TFTR supershot regime. Performance improvements to the centrifugal pellet injector for the Tore Supra tokamak are being made by modifying the storage-type pellet feed system, which has been redesigned to improve the reliability of delivery of pellets and to extend operation to longer pulse durations (up to 400 pellets). Two-stage light gas guns and electron-beam (e-beam) rocket accelerators for speeds in the range from 2 to 10 km/s are also under development. A repeating, two-stage light gas gun that has been developed can accelerate low-density plastic pellets at a 1-Hz repetition rate to speeds of 3 km/s. In a collaboration with ENEA-Frascati, a test facility has been prepared to study repetitive operation of a two-stage gas gun driver equipped with an extrusion-type deuterium pellet source. Extensive testing of the e-beam accelerator has demonstrated a parametric dependence of propellant burn velocity and pellet speed, in accordance with a model derived from the neutral gas shielding theory for pellet ablation in a magnetized plasma

  4. IAEA technical committee meeting on pellet injection

    International Nuclear Information System (INIS)

    1993-01-01

    The IAEA Technical Committee Meeting on Pellet Injection, May 10-12, 1993, at the Japan Atomic Energy Research Institute, Naka, Ibaraki-ken, Japan, was held to review the latest results on pellet injection and its effects on plasma confinement. In particular, topics included in the meeting include (i) pellet ablation and particle fueling results, (ii) pellet injection effects on confinement, including improved confinement modes, edge effects, magnetohydrodynamic activity and impurity transport, and (iii) injector technology and diagnostics using pellets. About 30 experts attended and 23 papers were presented. Refs, figs and tabs

  5. 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.)

  6. 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.

  7. Tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

    Gouge, M.J.; Baylor, L.R.; Cole, M.J.; Combs, S.K.; Dyer, G.R.; Fehling, D.T.; Fisher, P.W.; Foust, C.R.; Langley, R.A.; Milora, S.L.; Qualls, A.L.; Wilgen, J.B.; Schmidt, G.L.; Barnes, G.W.; Persing, R.G.

    1992-01-01

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) phase. The existing TFTR deuterium pellet injector (DPI) has been modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed to provide pellets ranging from 3.3 to 4.5 mm in diameter in arbitrarily programmable firing sequences at speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller. The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed, and the TPI was tested at ORNL with deuterium pellet. Results of the limited testing program at ORNL are described. The TPI is being installed on TFTR to support the D-D run period in 1992. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and secondary tritium containment systems and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  8. Surface analysis using X-ray photoelectron spectroscopy and X-ray diffraction of UO2 fuel pellets oxidised in air at 2300C and 2700C

    International Nuclear Information System (INIS)

    Tempest, P.A.; Tyler, J.W.

    1987-08-01

    Factors which affect the UO 2 → U 3 O 8 transformation have been investigated by sequentially oxidising UO 2 fuel pellets in air at 230 0 C and 270 0 C and monitoring the growth of U 3 O 7 and U 3 O 8 using X-ray photoelectron spectroscopy, X-ray diffraction and scanning electron microscopy. Initially oxidation proceeded at a linear rate by the inward diffusion of oxygen to form a complete layer of sub-stoichiometric U 3 O 7 . This phase was tetragonal with a c/a ratio of 1.015, significantly less than the value of 1.03 measured on UO 2 powder when oxidised under identical conditions. This difference and the preferred orientation exhibited by surface grains were caused by growth stresses induced in the pellet surface. Both intergranular and transgranular cracking occurred and became nucleation sites for the growth of U 3 O 8 . The linear oxidation period associated with U 3 O 7 growth was much shorter at 270 0 C than 230 0 C and U 3 O 8 nucleated earlier. Spallation and the production of particulate were only observed during the formation of U 3 O 8 when a 30% increase in volume arose from the U 3 O 7 → U 3 O 8 phase change. (author)

  9. 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

  10. Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Third semiannual report, January-June 1980

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbaum, H.S. (comp.)

    1980-09-01

    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 avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the work scope of this program one of these concepts is to be selected for demonstration in a commercial power reactor. It was decided to demonstrate Zr-liner in 132 bundles which have liners of either crystal-bar zirconium or of low-oxygen sponge zirconium in the reload for Quad Cities Unit 2, Cycle 6. Irradiation testing or barrier fuel was continued, and the superior PCI resistance of Zr-liner fuel was further substantiated in the current report period. Furthermore, an irradiation experiment in which Zr-liner fuel, having a deliberately fabricated cladding perforation, was operated at a linear heat generation rate of 35 kW/m to a burnup of approx. 3 MWd/kg U showed no unusual signs of degradation compared with a similarly defected reference fuel rod. Four lead test assemblies of barrier fuel (two of Zr-liner and two of Cu-barrier), presently under irradiation in Quad Cities Unit 1, have achieved a burnup of 11 MWd/kg U.

  11. Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Third semiannual report, January-June 1980

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.

    1980-09-01

    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 avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the work scope of this program one of these concepts is to be selected for demonstration in a commercial power reactor. It was decided to demonstrate Zr-liner in 132 bundles which have liners of either crystal-bar zirconium or of low-oxygen sponge zirconium in the reload for Quad Cities Unit 2, Cycle 6. Irradiation testing or barrier fuel was continued, and the superior PCI resistance of Zr-liner fuel was further substantiated in the current report period. Furthermore, an irradiation experiment in which Zr-liner fuel, having a deliberately fabricated cladding perforation, was operated at a linear heat generation rate of 35 kW/m to a burnup of approx. 3 MWd/kg U showed no unusual signs of degradation compared with a similarly defected reference fuel rod. Four lead test assemblies of barrier fuel (two of Zr-liner and two of Cu-barrier), presently under irradiation in Quad Cities Unit 1, have achieved a burnup of 11 MWd/kg U

  12. 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)

  13. Wood pellets : is it a reliable, sustainable, green energy option?

    International Nuclear Information System (INIS)

    Swaan, J.

    2006-01-01

    The Wood Pellet Association of Canada was formerly called the BC Pellet Fuel Manufacturers Association, and was renamed and re-organized in January 2006. The association serves as an advocate for the wood pellet industry in addition to conducting research projects. This power point presentation presented an overview of the wood pellet industry in North America and Europe. Canada's 23 pellet plants currently produce just over 1,000,000 tons of wood pellets annually. Pellet producers in the United States produce approximately 800,000 tons annually for the residential bagged market. There are currently 240 pellet plants in Europe, and district heating is the largest growth market for wood pellets in Europe. British Columbia (BC) pellet producers will ship 450,000 tons to European power plants in 2005. Wood pellet specifications were presented, with details of calorific values, moisture and ash contents. An outline of wood pellet production processes was provided. New pellet plants currently under construction were reviewed. Domestic, North American and overseas exports were discussed, along with production estimates for BC for the next 5 years. A chart of world production and consumption of wood pellets between 2000 to 2010 was presented. North American wood pellet technologies were described. The impact of the pine beetle infestation in BC on the wood pellet industry was evaluated, and a worldwide wood pellet production growth forecast was presented. Issues concerning off-gassing, emissions, and torrifracation were also discussed. tabs., figs

  14. Qualitative comparison of bremsstrahlung X-rays and 800 MeV protons for tomography of urania fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Morris, C. L.; Bourke, M.; Byler, D. D.; Chen, C. F.; Hogan, G.; Hunter, J. F.; Kwiatkowski, K.; Mariam, F. G.; McClellan, K. J.; Merrill, F.; Morley, D. J.; Saunders, A. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

    2013-02-15

    We present an assessment of x-rays and proton tomography as tools for studying the time dependence of the development of damage in fuel rods. We also show data taken with existing facilities at Los Alamos National Laboratory that support this assessment. Data on surrogate fuel rods have been taken using the 800 MeV proton radiography (pRad) facility at the Los Alamos Neutron Science Center (LANSCE), and with a 450 keV bremsstrahlung X-ray tomography facility. The proton radiography pRad facility at LANSCE can provide good position resolution (<70 {mu}m has been demonstrate, 20 {mu}m seems feasible with minor changes) for tomography on activated fuel rods. Bremsstrahlung x-rays may be able to provide better than 100 {mu}m resolution but further development of sources, collimation, and detectors is necessary for x-rays to deal with the background radiation for tomography of activated fuel rods.

  15. Qualitative comparison of bremsstrahlung X-rays and 800 MeV protons for tomography of urania fuel pellets

    International Nuclear Information System (INIS)

    Morris, C. L.; Bourke, M.; Byler, D. D.; Chen, C. F.; Hogan, G.; Hunter, J. F.; Kwiatkowski, K.; Mariam, F. G.; McClellan, K. J.; Merrill, F.; Morley, D. J.; Saunders, A.

    2013-01-01

    We present an assessment of x-rays and proton tomography as tools for studying the time dependence of the development of damage in fuel rods. We also show data taken with existing facilities at Los Alamos National Laboratory that support this assessment. Data on surrogate fuel rods have been taken using the 800 MeV proton radiography (pRad) facility at the Los Alamos Neutron Science Center (LANSCE), and with a 450 keV bremsstrahlung X-ray tomography facility. The proton radiography pRad facility at LANSCE can provide good position resolution (<70 μm has been demonstrate, 20 μm seems feasible with minor changes) for tomography on activated fuel rods. Bremsstrahlung x-rays may be able to provide better than 100 μm resolution but further development of sources, collimation, and detectors is necessary for x-rays to deal with the background radiation for tomography of activated fuel rods.

  16. 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)

  17. 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)

  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. Current generation by phased injection of pellets

    International Nuclear Information System (INIS)

    Fisch, N.J.

    1983-08-01

    By phasing the injection of frozen pellets into a tokamak plasma, it is possible to generate current. The current occurs when the electron flux to individual members of an array of pellets is asymmetric with respect to the magnetic field. The utility of this method for tokamak reactors, however, is unclear; the current, even though free in a pellet-fueled reactor, may not be large enough to be worth the trouble. Uncertainty as to the utility of this method is, in part, due to uncertainty as to proper modeling of the one-pellet problem

  20. Solid deuterium centrifuge pellet injector

    International Nuclear Information System (INIS)

    Foster, C.A.

    1982-01-01

    Pellet injectors are needed to fuel long pulse tokamak plasmas and other magnetic confinement devices. For this purpose, an apparatus has been developed that forms 1.3-mm-diam pellets of frozen deuterium at a rate of 40 pellets per second and accelerates them to a speed of 1 km/s. Pellets are formed by extruding a billet of solidified deuterium through a 1.3-mm-diam nozzle at a speed of 5 cm/s. The extruding deuterium is chopped with a razor knife, forming 1.3-mm right circular cylinders of solid deuterium. The pellets are accelerated by synchronously injecting them into a high speed rotating arbor containing a guide track, which carries them from a point near the center of rotation to the periphery. The pellets leave the wheel after 150 0 of rotation at double the tip speed. The centrifuge is formed in the shape of a centrifugal catenary and is constructed of high strength KEVLAR/epoxy composite. This arbon has been spin-tested to a tip speed of 1 km/s

  1. Solid deuterium centrifuge pellet injector

    International Nuclear Information System (INIS)

    Foster, C.A.

    1983-01-01

    Pellet injectors are needed to fuel long pulse tokamak plasmas and other magnetic confinement devices. For this purpose, an apparatus has been developed that forms 1.3-mm-diam pellets of frozen deuterium at a rate of 40 pellets per second and accelerates them to a speed of 1 km/s. Pellets are formed by extruding a billet of solidified deuterium through a 1.3-mm-diam nozzle at a speed of 5 cm/s. The extruding deuterium is chopped with a razor knife, forming 1.3-mm right circular cylinders of solid deuterium. The pellets are accelerated by synchronously injecting them into a high speed rotating arbor containing a guide track, which carries them from a point near the center of rotation to the periphery. The pellets leave the wheel after 150 0 of rotation at double the tip speed. The centrifuge is formed in the shape of a centrifugal catenary and is constructed of high strength Kevlar/epoxy composite. This arbor has been spin-tested to a tip speed of 1 km/s

  2. Study of radiation effects on zircaloy 4 microstructure (Impact on susceptibility to fuel pellet-cladding interaction in PWR)

    International Nuclear Information System (INIS)

    Lefebvre, F.

    1989-01-01

    In PWR the fast neutron flux is an important parameter for fuel can aging by modification of zircaloy-4 microstructure: amorphisation and dissolution of intermetallic precipitates. These phenomena are both analysed and their influence on fuel-cladding interaction is discussed. Irradiations by 1 MeV electrons, Ar ions, Kr ions and fast neutrons are realized for comparison of damages with different defect creation kinetics. Amorphisation is explained as the crystal amorphous state transformation allowing precipitate dissolution by creation of a chemical potential gradient between matrix and amorphous phase. Progressive dissolution of precipitates produced by irradiation decrease the number of potential sites for stress corrosion cracking, improving rupture resistance of the alloy by fuel-cladding interaction [fr

  3. 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)

  4. 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)

  5. Quality control and testing UO2 powder and sintering pellets for nuclear fuel for LWR in out of pile condition

    International Nuclear Information System (INIS)

    Djuricic, Lj.; Katanic, J.; Stefanovic, M.

    1976-01-01

    The analysis of chemical and physical characteristics of fuels based on UO2 from the point of view of requested properties in the nuclear application, of the foreign technical methods of characterisation and domestic experience is given as one of the first steps toward standardization in the field in the state

  6. 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)

  7. Why pellet fuelling of large machines?

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Arguments for pellet fueling as a way to optimize the density profile in large machines with respect to the density limit, the beta limit, energy confinement and requirements for hydrogen and helium pumping are reviewed. It is concluded that pellets can be used as a way to overcome the density limit and enhance energy confinement but there is currently no clear argument for density profile shaping. Pumping requirements are lowered for deep fueling

  8. Raw materials for pellets; Rohstoffe fuer Pellets

    Energy Technology Data Exchange (ETDEWEB)

    Neumann, H.

    2008-01-15

    In order to keep the pellet prices stable, producers look for new raw materials. Sawdust as a former basis also competes with the manufacturers of chip boards and paper. Three classes of quality are discussed by the pellet manufacturers: (a) the DINplus pellet as a premium segment for which high-quality sawdust are used; (b) a wood pellet from natural wood with varying quality for the utilization in larger plants with filters; (c) the inexpensive industrial wood pellet which deviates from the DINplus commodity regarding to the ingredients and form and could be fired in larger power stations.

  9. Development of repeating pneumatic pellet injector

    Energy Technology Data Exchange (ETDEWEB)

    Oda, Y.; Onozuka, M.; Shimomura, T. (Mitsubishi Heavy Industries Ltd., Kobe (Japan)) (and others)

    1990-01-01

    A repeating pneumatic pellet injector has been constructed to experiment with the technique of continuous injection for fueling fusion reactors. This device is composed of a cryogenic extruder and a gun assembly in (among others) a high-vacuum vessel, diagnostic vessels, LHe, fuel-gas and propellant-gas supply systems, control and data acquisition systems, etc. The performance tests, using hydrogen, have proved that the device provides the function of extruding frozen hydrogen ribbons at the speed of 6 mm s{sup -1}, chambering pellet at the rate of 5 Hz, and injecting pellet at the speed of 900 m s{sup -1}, as planned. (author).

  10. Development of repeating pneumatic pellet injector

    International Nuclear Information System (INIS)

    Oda, Y.; Onozuka, M.; Shimomura, T.

    1990-01-01

    A repeating pneumatic pellet injector has been constructed to experiment with the technique of continuous injection for fueling fusion reactors. This device is composed of a cryogenic extruder and a gun assembly in (among others) a high-vacuum vessel, diagnostic vessels, LHe, fuel-gas and propellant-gas supply systems, control and data acquisition systems, etc. The performance tests, using hydrogen, have proved that the device provides the function of extruding frozen hydrogen ribbons at the speed of 6 mm s -1 , chambering pellet at the rate of 5 Hz, and injecting pellet at the speed of 900 m s -1 , as planned. (author)

  11. Durability of pellets made from different wood fuel assortments - The effect of moisture content, particle size and temperature development during storage. A laboratory-scale study; Haallfasthet hos braenslepellets tillverkade av olika traedbraenslesortiment - betydelsen av fukthalt, fraktionsstorlek och temperaturutveckling under lagringen. En laboratoriestudie

    Energy Technology Data Exchange (ETDEWEB)

    Lehtikangas, Paeivi [Swedish Univ. of Agricultural Sciences, Uppsala (Sweden). Dept. of Forest Management and Products

    2001-02-01

    Durability of fuel pellets is one of the most important quality variables. The aim of this pilot study was to investigate the effect of moisture content and particle size distribution on the durability of pellets made out of sawdust, logging residues and bark from Norway spruce and Scots pine. Moreover, the durability of pellets made out of logging residues which were exposed to high temperature development was investigated. Pellets were manufactured using a laboratory press CPM Europe. The determined quality parameters were moisture content, percentage of accept-pellets (percentage of pellets of the total amount of raw material), temperature and density of individual pellets. The pelleting process was controlled by monitoring the raw material input to the press, vibration and power consumption. From the results of this study the following preliminary conclusions can be drawn: * Increase moisture content and temperature during pelleting had a positive correlation with the percentage of the accept-pellets, especially concerning the bark assortments, and * Pellets made out of particles smaller than 1 mm resulted in a significantly (Student's t-test) higher percentage of accept-pellets than pellets made out of particles of 1-2 mm. Moreover, the process temperature was probably too low (max. 86 deg C) to activate softening of the lignin. A lower press temperature was sufficient to produce the same percentage of accept-pellets of fresh bark than with other raw materials. A possible reason for that is the presence of high amounts of extractives in bark, which could develop an adhesive nature during thermal treatment.

  12. Mitigation of divertor heat flux by high-frequency ELM pacing with non-fuel pellet injection in DIII-D

    Directory of Open Access Journals (Sweden)

    A. Bortolon

    2017-08-01

    Full Text Available Experiments have been conducted on DIII-D investigating high repetition rate injection of non-fuel pellets as a tool for pacing Edge Localized Modes (ELMs and mitigating their transient divertor heat loads. Effective ELM pacing was obtained with injection of Li granules in different H-mode scenarios, at frequencies 3–5 times larger than the natural ELM frequency, with subsequent reduction of strike-point heat flux (Bortolon et al., Nucl. Fus., 56, 056008, 2016. However, in scenarios with high pedestal density (∼6 ×1019m−3, the magnitude of granule triggered ELMs shows a broad distribution, in terms of stored energy loss and peak heat flux, challenging the effectiveness of ELM mitigation. Furthermore, transient heat-flux deposition correlated with granule injections was observed far from the strike-points. Field line tracing suggest this phenomenon to be consistent with particle loss into the mid-plane far scrape-off layer, at toroidal location of the granule injection.

  13. Large area quantitative X-ray mapping of (U,Pu)O2 nuclear fuel pellets using wavelength dispersive electron probe microanalysis

    International Nuclear Information System (INIS)

    Bremier, S.; Haas, D.; Somers, J.; Walker, C.T.

    2003-01-01

    The work presented is an example of how large area compositional mapping (≥1 mm 2 ) can be used to provide quantitative information on element distribution and specimen homogeneity. High-resolution was accomplished by producing a collage of X-ray maps acquired using classical conditions; magnification x400, spatial resolution 256x256 pixels. The individual images, each measuring roughly 250x250 μm, were converted to quantitative maps using the HIMAX reg software package and the XMAS reg matrix correction from SAMx. The quantitative gray-level large area X-ray picture was pieced together using the 'Multiple Image Alignment' function of the ANALYSIS reg image processing software. This software was also used to convert the gray-level pictures to false color images. The specimens investigated were transverse sections of MOX fuel pellets. Results are presented for the distribution of Pu by area fraction and cumulative area fraction, the size distribution of regions of high Pu concentration and average separation of these regions

  14. Temperature distribution on fuel rods: a study on the effect of eccentricity in the position of UO2 pellets

    International Nuclear Information System (INIS)

    Gaspar Junior, Joao Carlos Aguiar

    2010-01-01

    This work proposes the development of a method of solving equations of heat transfer applied in fuel rods using the finite element method, in order to evaluate the performance and safety of the nuclear system. Was prepared in a Fortran program to evaluate the equations governing the problem, the boundary conditions and apply the properties of materials on a steady state. This program uses the mesh generation input and graphical output generated by the program GID. The method was validated against the analytical solution found in the book Todreas and Kazimi with error less than 0.2% and with respect to the improved analytical solution of Nijsing for axisymmetry rod and eccentricity rod with error less than a 3.6%. Applications have been developed with the use of correlations for properties with the temperature dependence of resolution axisymmetry rod and the resolution of a rod with eccentricity. The method developed, should it be implemented, would allow the assessment of fuel rods in the given situations and other scenarios, as well as adding a tool of substantial value in the analysis of rods. (author)

  15. Pellet injectors for JET

    International Nuclear Information System (INIS)

    Andelfinger, C.; Buechl, K.; Lang, R.S.; Schilling, H.B.; Ulrich, M.

    1981-09-01

    Pellet injection for the purpose of refuelling and diagnostic of fusion experiments is considered for the parameters of JET. The feasibility of injectors for single pellets and for quasistationary refuelling is discussed. Model calculations on pellet ablation with JET parameters show the required pellet velocity ( 3 ). For single pellet injection a light gas gun, for refuelling a centrifuge accelerator is proposed. For the latter the mechanical stress problems are discussed. Control and data acquisition systems are outlined. (orig.)

  16. Optimization of bentonite pellet properties

    International Nuclear Information System (INIS)

    Sanden, Torbjoern; Andersson, Linus; Jonsson, Esther; Fritzell, Anni

    2012-01-01

    Document available in extended abstract form only. SKB in Sweden is developing and implementing concepts for the final disposal of spent nuclear fuel. A KBS-3V repository consists of a deposition tunnel with copper canisters containing spent fuel placed in vertical deposition holes. The canisters are embedded in highly compacted bentonite. After emplacement of canisters and bentonite blocks, the tunnels will be backfilled and sealed with an in-situ cast plug at the entrance. The main concept for backfilling the deposition tunnels imply pre compacted blocks of bentonite stacked on a bed of bentonite pellet. The remaining slot between blocks and rock will be filled with bentonite pellets. The work described in this abstract is a part of the ASKAR-project which main goal is to make a system design based on the selected concept for backfilling. Immediately after starting the backfill installation, inflowing water from the rock will come in contact with the pellet filling and thereby influence the characteristics of the pellet filling. The pellet filling helps to increase the average density of the backfill, but one of the most important properties beside this is the water storing capacity which will prevent water from reaching the backfill front where it would disturb and influence the quality of the installation. If water flows through the pellet filling out to the backfilling front, there will be erosion of material which also will affect the quality of the installed backfill. In order to optimize the properties regarding water storing capacity and sensitivity for erosion a number of tests have been made with different pellet types. The tests were made in different scales and with equipment specially designed for the purpose. The performed tests can be divided in four parts: 1. Standard tests (determining water content and density of pellet fillings and individual pellets, compressibility of the pellet fillings and strength of the individual pellets); 2. Erosion

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  18. The spectrographic analysis of plutonium oxide or mixed plutonium oxide/uranium oxide fuel pellets by the dried residue technique

    International Nuclear Information System (INIS)

    Jarbo, G.J.; Faught, P.; Hildebrandt, B.

    1980-05-01

    An emission spectrographic method for the quantitative determination of metallic impurities in plutonium oxide and mixed plutonium oxide/uranium oxide is described. The fuel is dissolved in nitric acid and the plutonium and/or uranium extracted with tributyl phosphate. A small aliquot of the aqueous residue is dried on a 'mini' pyrolitic graphite plate and excited by high voltage AC spark in an oxygen atmosphere. Spectra are recorded in a region which has been specially selected to record simultaneously lines of boron and cadmium in the 2nd order and all the other elements of interest in the 1st order. Indium is used as an internal standard. The excitation of very small quantities of the uraniumm/plutonium free residue by high voltage spark, together with three separate levels of containment reduce the hazards to personnel and the environment to a minimum with limited effect on sensitivity and accuracy of the results. (auth)

  19. 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)

  20. 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 pellets. The magnitude of these bulk properly reductions was found to be primarily dependent on the amount of crack area in the transverse plane of the fuel. The model described herein uses a simple description of the crack geometry to couple the fuel rod thermal and mechanical behaviors by relating in-reactor data to Hooke's Law and a crack compliance model. Data from the NRC/PNL Halden experiment IFA-432 show that for a typical helium-filled BWR-design rod at 30 kW/m, the effective thermal conductivity and elastic moduli of the cracked fuel are 4/5 and 1/40 of that for solid pellets, respectively

  1. The pellet handbook: the production and thermal utilisation of pellets

    National Research Council Canada - National Science Library

    Obernberger, Ingwald; Thek, Gerold

    2010-01-01

    ...: - International overview of standards for pellets - Evaluation of raw materials and raw material potentials - Quality and properties of pellets - Technical evaluation of the pellet production process...

  2. Pellet injector research at ORNL

    International Nuclear Information System (INIS)

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

    1988-01-01

    Advanced plasma fueling systems for magnetic confinement devices are under development a the Oak Ridge National Laboratory (ORNL). The general approach is that of producing and accelerating frozen hydrogen isotope pellets at speeds in the range 1-2 km/s and higher. Recently, ORNL provided pneumataic-based pellet fueling systems for two of the world's largest tokamak experiments, the Tokamak Fusion Test Reactor (TFTR) and the Joint European Torus (JET). A new versatile centrifuge type injector is being readied at ORNL for use on the Tore Supra tokamak. Also, a new simplified eight-shot injector design has been developed for use on the Princeton Beta Experiment (PBX) and the Advanced Toroidal Facility (ATF). In addition to these confinement physics related activities, ORNL is pursuing advanced technologies to achieve pellet velocities significantly in excess of 2 km/s and is carrying out a Tritium Proof-of-Principle (TPOP) experiment in which the fabrication and acceleration of tritium pellets have already been demonstrated. This paper describes these ongoing activities. 25 refs., 9 figs

  3. Completion of UO{sub 2} pellets production and fuel rods load for the RA-8 critical facility; Finalizacion de la produccion de pastillas y carga de barras combustibles de UO{sub 2} para el conjunto critico RA-8

    Energy Technology Data Exchange (ETDEWEB)

    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 [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Combustibles Nucleares

    1999-07-01

    The Advanced Fuels Division produced fuel pellets of {sup 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{sub 2} with 3.4% enrichment in {sup 235}U, therefore the {sup 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)

  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. Wood pellets for stoker burner

    International Nuclear Information System (INIS)

    Nykaenen, S.

    2000-01-01

    The author of this article has had a stoker for several years. Wood chips and sod peat has been used as fuels in the stoker, either separately or mixed. Last winter there occurred problems with the sod peat due to poor quality. Wood pellets, delivered by Vapo Oy were tested in the stoker. The price of the pellets seemed to be a little high 400 FIM/500 kg large sack. If the sack is returned in good condition 50 FIM deposit will be repaid to the customer. However, Vapo Oy informed that the calorific value of wood pellets is three times higher than that of sod peat so it should not be more expensive than sod peat. When testing the wood pellets in the stoker, the silo of the stoker was filled with wood pellets. The adjustments were first left to position used for sod peat. However, after the fire had ignited well, the adjustments had to be decreased. The content of the silo was combusted totally. The combustion of the content of the 400 litter silo took 4 days and 22 hours. Respectively combustion of 400 l silo of good quality sod peat took 2 days. The water temperature with wood pellets remained at 80 deg C, while with sod peat it dropped to 70 deg C. The main disadvantage of peat with small loads is the unhomogenous composition of the peat. The results of this test showed that wood pellets will give better efficiency than peat, especially when using small burner heads. The utilization of them is easier, and the amount of ash formed in combustion is significantly smaller than with peat. Wood pellets are always homogenous and dry if you do not spoil it with unproper storage. Pellets do not require large storages, the storage volume needed being less than a half of the volume needed for sod peat. When using large sacks the amount needed can even be transported at the trunk of a passenger car. Depending on the area to be heated, a large sack is sufficient for heating for 2-3 weeks. Filling of stoker every 2-5 day is not an enormous task

  6. Proceedings of the 2. world conference on Pellets

    International Nuclear Information System (INIS)

    2006-07-01

    The conference and exhibition had over 1000 participants from 60 different countries. Subject areas covered by the conference were: Raw Materials For Densification; Pellet Production Technologies; Pellet Burning Technologies; Supply Chain Logistics; Environmental Issues; Marketing of Densified Fuels; Co-Generation of Heat and Electricity from Densified Fuels; 57 contributions have been separately indexed for the database

  7. Mechanistic considerations used in the development of the probability of failure in transient increases in power (PROFIT) pellet-zircaloy cladding (thermo-mechanical-chemical) interactions (pci) fuel failure model

    International Nuclear Information System (INIS)

    Pankaskie, P.J.

    1980-05-01

    A fuel Pellet-Zircaloy Cladding (thermo-mechanical-chemical) interactions (PCI) failure model for estimating the Probability of Failure in Transient Increases in Power (PROFIT) was developed. PROFIT is based on (1) standard statistical methods applied to available PCI fuel failure data and (2) a mechanistic analysis of the environmental and strain-rate-dependent stress versus strain characteristics of Zircaloy cladding. The statistical analysis of fuel failures attributable to PCI suggested that parameters in addition to power, transient increase in power, and burnup are needed to define PCI fuel failures in terms of probability estimates with known confidence limits. The PROFIT model, therefore, introduces an environmental and strain-rate dependent Strain Energy Absorption to Failure (SEAF) concept to account for the stress versus strain anomalies attributable to interstitial-dislocation interaction effects in the Zircaloy cladding

  8. Factors Affecting the Sintering of UO2 Pellets

    International Nuclear Information System (INIS)

    El-Hakim, E.; Afifi, Y.K.

    1999-01-01

    Sintering of UO 2 pellets is affected by many parameters such as; UO 2 powder parameters, the conditions followed for preparing the green UO 2 pellets and the sintering scheme(heating and cooling rate, soaking time and temperature). The aim of this work is to study the effect of some these parameters on the characteristics of the sintered UO 2 pellets were qualified according to the technical specifications of Candu fuel. Pressed green pellets at different pressing force (15 to 50 k N) were sintered at 1650 ±20 degree for two hours to study the effect of pressing force on the sintered pellets characteristics; visual inspection, pellet dimensions, density and shrinkage ratio. Compacted green pellets at a pressing force of 48 k N were sintered at different sintering temperature (1600± 20 degree, 1650 ±20 degree, 1700± 20 degree) for two hours to study the effect of sintering temperature on the sintered pellets characteristics. The effect of the heating rate (200,300 and 400 degree per hour) on the sintered pellets characteristics was also investigated. It was found that the pressing force used to compact the green pellets had an effect on the density of the sintered pellets. Pellets pressed at 15 k N have a density of 10.3 g/cm 3 while, those pressed at 50 k N have a density of 10.6 g/cm 3. It was observed that increasing the heating rate to 400 degree /h lead to cracked pellets

  9. Repeating pneumatic pellet injector in JAERI

    International Nuclear Information System (INIS)

    Kasai, Satoshi; Hasegawa, Kouichi; Suzuki, Sadaaki; Miura, Yukitoshi; Oda, Yasushi; Onozuka, Masanori; Tsujimura, Seiichi.

    1992-09-01

    A repeating pneumatic pellet injector has been developed and constructed at Japan Atomic Energy Research Institute. This injector can provide repetitive pellet injection to fuel tokamak plasmas for an extended period of time, aiming at the improvement of plasma performance. The pellets with nearly identical speed and mass can be repeatedly injected with a repetition rate of 2-3.3 Hz and a speed of up to 1.7 km/s by controlling the temperature of the cryogenic system, the piston speed and the pressure of the propellant gas. (author)

  10. Repeating pneumatic pellet injector in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kasai, Satoshi; Hasegawa, Kouichi; Suzuki, Sadaaki; Miura, Yukitoshi (Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment); Oda, Yasushi; Onozuka, Masanori; Tsujimura, Seiichi.

    1992-09-01

    A repeating pneumatic pellet injector has been developed and constructed at Japan Atomic Energy Research Institute. This injector can provide repetitive pellet injection to fuel tokamak plasmas for an extended period of time, aiming at the improvement of plasma performance. The pellets with nearly identical speed and mass can be repeatedly injected with a repetition rate of 2-3.3 Hz and a speed of up to 1.7 km/s by controlling the temperature of the cryogenic system, the piston speed and the pressure of the propellant gas. (author).

  11. Pneumatic pellet injectors for TFTR and JET

    International Nuclear Information System (INIS)

    Combs, S.K.; Milora, S.L.

    1986-01-01

    This paper describes the development of pneumatic hydrogen pellet injectors for plasma fueling applications on the Tokamak Fusion Test Reactor (TFTR) and the Joint European Torus (JET). The performance parameters of these injectors represent an extension of previous experience and include pellet sizes in the range 2-6 mm in diameter and speeds approaching 2 km/s. Design features and operating characteristics of these pneumatic injectors are presented

  12. Methane pellet moderator development

    International Nuclear Information System (INIS)

    Foster, C.A.; Schechter, D.E.; Carpenter, J.M.

    2004-01-01

    A methane pellet moderator assembly consisting of a pelletizer, a helium cooled sub-cooling tunnel, a liquid helium cooled cryogenic pellet storage hopper and a 1.5L moderator cell has been constructed for the purpose demonstrating a system for use in high-power spallation sources. (orig.)

  13. Pellet injection in WVIIA

    International Nuclear Information System (INIS)

    Renner, H.; Wuersohing, E.; Weller, A.; Jaeckel, H.; Hartfuss, H.; Hacker, H.; Ringler, H.; Buechl, K.

    1986-01-01

    The results of pellet injection experiments in the Wendelstein VII A stellarator are presented. The injector was a single shot pneumatic gun using deuterium pellets. Experiments were carried out in both ECRH and NI plasmas. Data is shown for plasma density, energy confinement, penetration depth and pellet ablation. Results are compared to a neutral gas shielding model

  14. 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

  15. Neutron absorber pellets

    International Nuclear Information System (INIS)

    Radford, K.C.

    1983-01-01

    An annular burnable poison pellet of aluminium oxide - boron carbide (Al 2 O 3 - B 4 C) adapted for positioning in the annular space of concentrically disposed zircaloy tubes. Each tubular pellet is fabricated from Al 2 O 3 powders of moderate sintering activity which serves as a matrix for B 4 C medium size particle distribution. Special pellet moisture controls are incorporated in the pellet for moisture stability and the pellet is sintered in the temperature range of 1630 deg to 1650 deg C. This method of fabrication produces a pellet about 2 inch long with a wall thickness of from 0.020 inch to 0.040 inch. Fabricating each pellet to about 70% theoretical density gives an optimum compromise between fabricability, microstructure, strength and moisture absorption. (author)

  16. 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

  17. Development of pellet injection systems for ITER

    International Nuclear Information System (INIS)

    Combs, S.K.; Gouge, M.J.; Baylor, L.R.

    1995-01-01

    Oak Ridge National Laboratory (ORNL) has been developing innovative pellet injection systems for plasma fueling experiments on magnetic fusion confinement devices for about 20 years. Recently, the ORNL development has focused on meeting the complex fueling needs of the International Thermonuclear Experimental Reactor (ITER). In this paper, we describe the ongoing research and development activities that will lead to a ITER prototype pellet injector test stand. The present effort addresses three main areas: (1) an improved pellet feed and delivery system for centrifuge injectors, (2) a long-pulse (up to steady-state) hydrogen extruder system, and (3) tritium extruder technology. The final prototype system must be fully tritium compatible and will be used to demonstrate the operating parameters and the reliability required for the ITER fueling application

  18. Performance characterization of pneumatic single pellet injection system

    International Nuclear Information System (INIS)

    Schuresko, D.D.; Milora, S.L.; Hogan, J.T.; Foster, C.A.; Combs, S.K.

    1983-01-01

    The Oak Ridge National Laboratory single-shot pellet injector, which has been used in plasma fueling experiments on ISX and PDX, has been upgraded and extensively instrumented in order to study the gas dyamics of pneumatic pellet injection. An improved pellet transport line was developed which utilizes a 0.3-cm-diam by 100-cm-long guide tube. Pellet gun performance was characterized by measurements of breech and muzzle dynamic pressures and by pellet velocity and mass determinations. Velocities of up to 1.4 km/s were achieved for intact hydrogen pellets using hydrogen propellant at 5-MPa breech pressure. These data have been compared with new pellet acceleration calculations which include the effects of propellant friction, heat transfer, time-dependent boundary conditions, and finite gun geometry. These results provide a basis for the extrapolation of present-day pneumatic injection system performance to velocities in excess of 2 km/s

  19. Performance characterization of pneumatic single pellet injection system

    International Nuclear Information System (INIS)

    Schuresko, D.D.; Milora, S.L.; Hogan, J.T.; Foster, C.A.; Combs, S.K.

    1982-01-01

    The Oak Ridge National Laboratory single-shot pellet injector, which has been used in plasma fueling experiments on ISX and PDX, has been upgraded and extensively instrumented in order to study the gas dynamics of pneumatic pellet injection. An improved pellet transport line was developed which utilizes a 0.3-cm-diam by 100-cm-long guide tube. Pellet gun performance was characterized by measurements of breech and muzzle dynamic pressures and by pellet velocity and mass determinations. Velocities up to 1.4 km/s were achieved for intact hydrogen pellets using hydrogen propellant at 5-MPa breech pressure. These data have been compared with new pellet acceleration calculations which include the effects of propellant friction, heat transfer, time-dependent boundary conditions, and finite gun geometry. These results provide a basis for the extrapolation of present-day pneumatic injection system performance to velocities in excess of 2 km/s

  20. 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.)

  1. Particle density determination of pellets and briquettes

    Energy Technology Data Exchange (ETDEWEB)

    Rabier, Fabienne; Temmerman, Michaeel [Centre wallon de Recherches agronomiques, Departement de Genie rural, CRA-W, Chaussee de Namur, 146, B 5030 Gembloux (Belgium); Boehm, Thorsten; Hartmann, Hans [Technologie und Foerderzentrum fuer Nachwachsende Rohstoffe, TFZ, Schulgasse 18, D 94315 Straubing (Germany); Daugbjerg Jensen, Peter [Forest and Landscape, The Royal Veterinary and Agricultural University, Rolighedsvej 23, DK 1958 Frederiksberg C (Denmark); Rathbauer, Josef [Bundesanstalt fuer Landtechnik, BLT, Rottenhauer Strasse,1 A 3250 Wieselburg (Austria); Carrasco, Juan; Fernandez, Miguel [Centro de investigaciones Energeticas, Medioambientales y Tecnologicas, CIEMAT, Avenida Complutense, 22 E 28040 Madrid (Spain)

    2006-11-15

    Several methods and procedures for the determination of particle density of pellets and briquettes were tested and evaluated. Round robin trials were organized involving five European laboratories, which measured the particle densities of 15 pellet and five briquette types. The test included stereometric methods, methods based on liquid displacement (hydrostatic and buoyancy) applying different procedures and one method based on solid displacement. From the results for both pellets and briquettes, it became clear that the application of a method based on either liquid or solid displacement (only tested on pellet samples) leads to an improved reproducibility compared to a stereometric method. For both, pellets and briquettes, the variability of measurements strongly depends on the fuel type itself. For briquettes, the three methods tested based on liquid displacement lead to similar results. A coating of the samples with paraffin did not improve the repeatability and the reproducibility. Determinations with pellets proved to be most reliable when the buoyancy method was applied using a wetting agent to reduce surface tensions without sample coating. This method gave the best values for repeatability and reproducibility, thus less replications are required to reach a given accuracy level. For wood pellets, the method based on solid displacement gave better values of repeatability, however, this instrument was tested at only one laboratory. (author)

  2. Pellet injector research and development at ORNL

    International Nuclear Information System (INIS)

    Combs, S.K.; Argo, B.E.; Baylor, L.R.; 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, E.; Sparks, D.O.; Tsai, C.C.; Wilgen, J.B.; Whealton, J.W.

    1993-01-01

    A variety of pellet injector designs have been developed at ORNL including single-shot guns that inject one pellet, multiple-shot guns that inject four and eight pellets, machine gun-types (single- and multiple-barrel) that can inject up to >100 pellets, and centrifugal accelerators (mechanical devices that are inherently capable of high repetition rates and long-pulse operation). With these devices, macroscopic pellets (1--6 mm in diameter) composed of hydrogen isotopes are typically accelerated to speeds of ∼1.0 to 2.0 km/s for injection into plasmas of experimental fusion devices. In the past few years, steady progress has been made at ORNL in the development and application of pellet injectors for fueling present-day and future fusion devices. In this paper, we briefly describe some research and development activities at ORNL, including: (1) two recent applications and a new one on large experimental fusion devices, (2) high-velocity pellet injector development, and (3) tritium injector research

  3. Pelletization of fine coals. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Sastry, K.V.S.

    1995-12-31

    Coal is one of the most abundant energy resources in the US with nearly 800 million tons of it being mined annually. Process and environmental demands for low-ash, low-sulfur coals and economic constraints for high productivity are leading the coal industry to use such modern mining methods as longwall mining and such newer coal processing techniques as froth flotation, oil agglomeration, chemical cleaning and synthetic fuel production. All these processes are faced with one common problem area--fine coals. Dealing effectively with these fine coals during handling, storage, transportation, and/or processing continues to be a challenge facing the industry. Agglomeration by the unit operation of pelletization consists of tumbling moist fines in drums or discs. Past experimental work and limited commercial practice have shown that pelletization can alleviate the problems associated with fine coals. However, it was recognized that there exists a serious need for delineating the fundamental principles of fine coal pelletization. Accordingly, a research program has been carried involving four specific topics: (i) experimental investigation of coal pelletization kinetics, (ii) understanding the surface principles of coal pelletization, (iii) modeling of coal pelletization processes, and (iv) simulation of fine coal pelletization circuits. This report summarizes the major findings and provides relevant details of the research effort.

  4. Pellets - the advance of refined bioenergy

    International Nuclear Information System (INIS)

    Dahlstroem, J.E.

    1997-01-01

    This conference paper discusses the role of pellets in the use of bioenergy in Sweden. Pellets (P) have many advantages: (1) P are dry and can be stored, (2) P create local jobs, (3) P burn without seriously polluting the environment, (4) P are made from domestic and renewable resources, (5) P have high energy density, (6) P fit well in an energy system adapted to nature, (6) P are an economical alternative, both on a small scale and on a large scale. Pellets are more laborious to use than oil or electricity and require about three times as much storage space as oil. The Swedish pellets manufacturers per 1997 are listed. Locally pellets are most conveniently transported as bulk cargo and delivered to a silo by means of pressurized air. Long-distance transport use train or ship. At present, pellets are most often used in large or medium-sized heat plants, but equipment exists for use from private houses and up to the size of MW. Pellets may become the most important alternative to the fossil fuels which along with electricity today are dominating the small scale market. 1 fig., 1 table

  5. Device for introducing radiative pellets in a tube

    International Nuclear Information System (INIS)

    Michel, A.; Milesi, A.

    1983-01-01

    Fuel sheaths are filled through a device comprising a funnel-guide with a bore having a diameter and slightly higher than pellet diameter and slightly lower than fuel can inside diameter. The flaring part of the funnel is toward facing a pellet distributor placed in a containment cell. The fuel can is tightened and aligned for a close contact with the funnel-guide [fr

  6. New pellet production and acceleration technologies for high speed pellet injection system 'HIPEL' in large helical device

    International Nuclear Information System (INIS)

    Viniar, I.; Sudo, S.

    1994-12-01

    New technologies of pellet production and acceleration for fueling and diagnostics purposes in large thermonuclear reactors are proposed. The technologies are intended to apply to the multiple-pellet injection system 'HIPEL' for Large Helical Device of NIFS in Japan. The pellet production technology has already been tested in a pipe-gun type pellet injector. It will realize the repeating pellet injection by means of decreasing of the pellet formation time into the pipe-gun barrel. The acceleration technology is based upon a new pump tube operation in two-stage gas gun and also upon a new conception of the allowable pressure acting on a pellet into a barrel. Some preliminary estimations have been made, and principles of a pump tube construction providing for a reliable long term operation in the repeating mode without any troubles from a piston are proposed. (author)

  7. 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.)

  8. 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

  9. A four-pellet pneumatic injection system in the JT-60

    International Nuclear Information System (INIS)

    Hiratsuka, Hajime; Kawasaki, Kouzo; Miyo, Yasuhiko; Yoshioka, Yuji; Ohta, Kazuya; Shimizu, Masatsugu; Kondo, Ikuo; Onozuka, Masanori; Shimomura, Tomoyoshi; Iwamoto, Syuichi; Hashiri, Noboru

    1991-01-01

    A four-pellet pneumatic injection system has been developed for plasma fueling of the JT-60. The JT-60 pellet injector is capable of accelerating separately four cylindrical pellets 3.0 mm in diameter x 3.0 mm long for two pellets and 4.0 mm in diameter x 4.0 mm long for the remaining two. The JT-60 pellet injector was installed on the JT-60 tokamak machine at the end of 1988. Obtained pellet velocity was higher than 2300 m/s by propellant gases of up to 100 bar and the pellet fueling efficiency achieved was around 70% for both dimensions of pellets. This paper describes the design, injection operation and performance test results of the JT-60 pellet injector. (orig.)

  10. A four-pellet pneumatic injection system in the JT-60

    Energy Technology Data Exchange (ETDEWEB)

    Hiratsuka, Hajime; Kawasaki, Kouzo; Miyo, Yasuhiko; Yoshioka, Yuji; Ohta, Kazuya; Shimizu, Masatsugu; Kondo, Ikuo (Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan)); Onozuka, Masanori; Shimomura, Tomoyoshi; Iwamoto, Syuichi; Hashiri, Noboru (Mitsubishi Heavy Industries Ltd., Kobe (Japan))

    1991-05-01

    A four-pellet pneumatic injection system has been developed for plasma fueling of the JT-60. The JT-60 pellet injector is capable of accelerating separately four cylindrical pellets 3.0 mm in diameter x 3.0 mm long for two pellets and 4.0 mm in diameter x 4.0 mm long for the remaining two. The JT-60 pellet injector was installed on the JT-60 tokamak machine at the end of 1988. Obtained pellet velocity was higher than 2300 m/s by propellant gases of up to 100 bar and the pellet fueling efficiency achieved was around 70% for both dimensions of pellets. This paper describes the design, injection operation and performance test results of the JT-60 pellet injector. (orig.).

  11. Fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Hajime.

    1995-01-01

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

  12. 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)

  13. Deuterium pellet injector gun design

    International Nuclear Information System (INIS)

    Lunsford, R.V.; Wysor, R.B.; Bryan, W.E.; Shipley, W.D.; Combs, S.K.; Foust, C.R.; Milora, S.L.; Fisher, P.W.

    1985-01-01

    The Deuterium Pellet Injector (DPI), an eight-pellet pneumatic injector, is being designed and fabricated for the Tokamak Fusion Test Reactor (TFTR). It will accelerate eight pellets, 4 by 4 mm maximum, to greater than 1500 m/s. It utilizes a unique pellet-forming mechanism, a cooled pellet storage wheel, and improved propellant gas scavenging

  14. 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

  15. Pellet injection experiments on tokamaks in ASIPP, China

    International Nuclear Information System (INIS)

    Yang, Y.; Bao, Y.; Li, J.; Gu, X.; He, Y.

    2001-01-01

    Pellet injection has been proved to be an effective method for deep fueling of fusion devices. Improvements of both the particle confinement and the energy confinement were observed in many experiments. In HT-6M and HT-7 tokamaks, single and multi-pellet experiments are tried, and attractive results are obtained. (author)

  16. Optimization of backfill pellet properties AASKAR DP2 - Laboratory tests

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Linus; Sanden, Torbjoern [Clay Technology AB, Lund (Sweden)

    2012-12-15

    Bentonite pellets are planned to be used as a part of the backfill in the Swedish spent nuclear fuel deep repository concept KBS-3. This report describes testing and evaluation of different backfill pellet candidates. The work completed included testing of both pellet material and pellet type. The materials tested were sourced from India (ASHA), Greece (IBECO, 2 products) and Wyoming USA (MX-80 clay). The majority of the tests were completed on the ASHA clay as well as the IBECO-RWC-BF products, with only limited testing of the others. The pellets tested were manufactured using both extrusion and roller compaction techniques and had different sizes and geometries. The following tests have been performed and are presented in this report: 1. General tests. Water content, bulk density and dry density have been determined for both the pellet filling and the individual pellets. The compressibility of the pellet filling was tested with CRS-tests and the strength of the individual pellets was tested with a special compression test. The water content varied from 11.3% to 18.7% and was highest for the extruded pellets. The dry density was somewhat higher for the roller-compacted pellets and their compressibility was lower. The strength of the individual pellets was generally higher for the extruded pellets. 2. Erosion. The pellet filling will be exposed to groundwater inflow when installed in the tunnel. This flow could possibly cause significant erosion on the pellet filling. Erosion tests have been performed with comparisons in erosion resistance made on the various material- and pellet-types. The influence of variations in water salinity and flow rates was also tested. The IBECO extruded 6- and 10- mm diameter rods and the compacted Posiva spec.-A pellet filling seem to have the lowest tendency to erode. It is also the IBECO extruded pellet filling that withstands variations in water salinity and flow rates best. 3. Water storing capacity. The pellet filling

  17. Reciprocating pellet press

    Science.gov (United States)

    Jones, Charles W.

    1981-04-07

    A machine for pressing loose powder into pellets using a series of reciprocating motions has an interchangeable punch and die as its only accurately machines parts. The machine reciprocates horizontally between powder receiving and pressing positions. It reciprocates vertically to press, strip and release a pellet.

  18. Uranium dioxide pellets

    International Nuclear Information System (INIS)

    Zawidzki, T.W.

    1979-01-01

    Sintered uranium dioxide pellets composed of particles of size > 50 microns suitable for power reactor use are made by incorporating a small amount of sulphur into the uranium dioxide before sintering. The increase in grain size achieved results in an improvement in overall efficiency when such pellets are used in a power reactor. (author)

  19. SAF line pellet gaging

    International Nuclear Information System (INIS)

    Jedlovec, D.R.; Bowen, W.W.; Brown, R.L.

    1983-10-01

    Automated and remotely controlled pellet inspection operations will be utilized in the Secure Automated Fabrication (SAF) line. A prototypic pellet gage was designed and tested to verify conformance to the functions and requirements for measurement of diameter, surface flaws and weight-per-unit length

  20. 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

  1. 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

  2. Characteristics of an electron-beam rocket pellet accelerator

    International Nuclear Information System (INIS)

    Tsai, C.C.; Foster, C.A.; Schechter, D.E.

    1989-01-01

    An electron-beam rocket pellet accelerator has been designed, built, assembled, and tested as a proof-of-principle (POP) apparatus. The main goal of accelerators based on this concept is to use intense electron-beam heating and ablation of a hydrogen propellant stick to accelerate deuterium and/or tritium pellets to ultrahigh speeds (10 to 20 km/s) for plasma fueling of next-generation fusion devices such as the International Thermonuclear Engineering Reactor (ITER). The POP apparatus is described and initial results of pellet acceleration experiments are presented. Conceptual ultrahigh-speed pellet accelerators are discussed. 14 refs., 8 figs

  3. 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.

  4. Tritium proof-of-principle pellet injector

    International Nuclear Information System (INIS)

    Fisher, P.W.

    1991-07-01

    The tritium proof-of-principle (TPOP) experiment was designed and built by Oak Ridge National Laboratory (ORNL) to demonstrate the formation and acceleration of the world's first tritium pellets for fueling of future fusion reactors. The experiment was first used to produce hydrogen and deuterium pellets at ORNL. It was then moved to the Tritium Systems Test Assembly at Los Alamos National Laboratory for the production of tritium pellets. The injector used in situ condensation to produce cylindrical pellets in a 1-m-long, 4-mm-ID barrel. A cryogenic 3 He separator, which was an integral part of the gun assembly, was capable of lowering 3 He levels in the feed gas to <0.005%. The experiment was housed to a glovebox for tritium containment. Nearly 1500 pellets were produced during the course of the experiment, and about a third of these were pure tritium or mixtures of deuterium and tritium. Over 100 kCi of tritium was processed through the experiment without incident. Tritium pellet velocities of 1400 m/s were achieved with high-pressure hydrogen propellant. The design, operation, and results of this experiment are summarized. 34 refs., 44 figs., 3 tabs

  5. Present and future trends in pellet markets, raw materials, and supply logistics in Sweden and Finland

    International Nuclear Information System (INIS)

    Selkimaeki, Mari; Mola-Yudego, Blas; Roeser, Dominik; Prinz, Robert; Sikanen, Lauri

    2010-01-01

    Wood pellets have become an important fuel in heat and power production, and pellet markets are currently undergoing rapid development. In this paper, the pellet markets, raw materials and supply structures are analyzed for Sweden and Finland, based on a database of the current location and production capacity of the pellet producers, complemented with existing reports and literature. In Sweden, a total of 94 pellet plants/producers were identified, producing 1.4 million tonnes of pellets, while the domestic consumption was 1.7 million tonnes, and about 400,000 t of pellets were imported to fulfil the demand in 2007. In Finland, 24 pellet plants/producers were identified and the production was around 330,000 t while the domestic consumption was 117,000 t in 2007. In Finland, pellet market has been long time export oriented, whereas domestic consumption has been growing mainly in the small scale consumer sector, estimated 15,000 households had pellet heating systems in 2008. In the future, the increasing number of pellet users will require a reliable delivery network and good equipment for bulk pellet deliveries. Provision of new raw materials and ensuring the good quality of pellets through the whole production, delivery and handling chain will be fundamental in order to increase the use of pellets and sustain the ability to compete with other fuels. (author)

  6. Pellets for fusion reactor refueling. Annual progress report, January 1, 1976--December 31, 1976

    International Nuclear Information System (INIS)

    Turnbull, R.J.; Kim, K.

    1977-01-01

    The purpose of this research is to test the feasibility of refueling fusion reactors using solid pellets composed of fuel elements. A solid hydrogen pellet generator has been constructed and experiments have been done to inject the pellets into the ORMAK Tokamak. A theory has been developed to describe the pellet ablation in the plasma, and an excellent agreement has been found between the theory and the experiment. Techniques for charging the pellets have been developed in order to accelerate and control them. Other works currently under way include the development of techniques for accelerating the pellets for refueling purpose. Evaluation of electrostatic acceleration has also been performed

  7. Small scale pelletizing equipment for agriculture; Smaaskalig pelleteringsanlaeggning foer lantbruket

    Energy Technology Data Exchange (ETDEWEB)

    Paulrud, Susanne (The Swedish Environmental Research Inst. Ltd., Stockholm (Sweden)); Wallin, Mikael (Sweden Powers Chippers AB, Boraas (Sweden))

    2009-06-15

    Refining agricultural raw materials is one way for farmers to increase the value of their products. For example, briquettes or pellets made from straw, reed canary grass or hemp can sell for a higher price than in loose or baled forms. The aim of this project was to develop and build a plant for small-scale production of pellets on the farm. Working together with the farmer, the plant would then be tested and adapted for small-scale production of pellets from straw, reed canary grass and hemp. The project also aimed to investigate and summarise suitable systems and solutions for breaking up bales in preparation for use in the pellets module. A pelleting system has been developed and built as a container module (measuring 6 x 2.80 x 2.50 m) by Sweden Power Chippers (SPC). The container system includes a fuel store with push floor, a grinder, an SPC pellet press (pp150, capacity 150 kg/hour for woodbased material), a conveyor belt and a cooling system. The practical operation of the plant was tested on two Swedish farms: Laattra gaard in Vingaaker and Ek gaard in Vara. The bales were broken up in preparation for pelleting using a straw mill of model Tomahawk 505M. The project has demonstrated that the SPC plant has the capacity to be used for agricultural pelleting of fuels from straw, canary reed grass and hemp. Some modification and continued adjustment of the feed system for the fuel remains to be done in order to optimise and ensure the reliability of the pelleting process. A certain amount of modifications to the plant is required to enable cost-effective transportation between different locations. Tests showed that each batch of fuel was unique, even from the same raw material, and that optimisation of the dies is necessary for each specific case. Training is required to run the plant. The farmers have been able to run the plant themselves, for example, starting up the plant, changing the sieve on the grinder, changing dies etc. In order for such small

  8. Validation of a multiparameter model to investigate torrefied biomass pelletization behavior

    DEFF Research Database (Denmark)

    Puig Arnavat, Maria; Ahrenfeldt, Jesper; Henriksen, Ulrik Birk

    2017-01-01

    The present study aims to apply and validate a simple model describing the forces that are built up along the dies of a pellet press matrix to the pelletization of torrefied biomass. The model combines a theoretical background with the use of a single pellet press to describe the pelletizing...... behavior of torrefied material in an industrial scale pellet mill. Wet torrefaction and dry torrefaction pretreatments are considered in the study. Both torrefaction concepts produce a fuel with enhanced properties including a lower moisture content, higher calorific value, and better friability. The fuel...... and to avoid time consuming as well as expensive trial and error experiments....

  9. Hydrogen pellet injection into Alcator C

    International Nuclear Information System (INIS)

    Greenwald, M.

    1983-09-01

    A four-shot pneumatic pellet injector, based on an ORNL design, has been built and operated on the Alcator C tokamak at MIT. The injector fires four independently-timed frozen hydrogen pellets with velocities in the range 8 x 10 4 - 1 x 10 5 cm/sec. Each contains 6 x 10 19 particles which corresponds to = 2 x 10 14 /cm 3 . The objectives of this experiment are to study pellet fueling and penetration, particle confinement, dependence of energy confinement on density profile and fueling mode, and edge physics and recycling as a function of fueling mode. Typical pre-injection plasmas have had anti n/sub e/ = 2 - 3 x 10 14 , Bt = 80 - 100 kG, Ip = 400 - 500 kA, T/sub e/(0) = 1200 - 1500 ev. A single pellet injected into this plasma will roughly double the electron density. Record plasma densities have been obtained by multiple injections. Line average densities in excess of 8 x 10 14 have been achieved, with highly peaked profiles. Central densities of 1.5 - 2 x 10 15 have been measured

  10. Optimization of a multi-parameter model for biomass pelletization to investigate temperature dependence and to facilitate fast testing of pelletization behavior

    DEFF Research Database (Denmark)

    Holm, Jens Kai; Stelte, Wolfgang; Posselt, Dorthe

    2011-01-01

    Pelletization of biomass residues increases the energy density, reduces storage and transportation costs and results in a homogeneous product with well-defined physical properties. However, raw materials for fuel pellet production consist of ligno-cellulosic biomass from various resources...... and error” experiments and personal experience. However in recent years the utilization of single pellet press units for testing the biomass pelletizing properties has attracted more attention. The present study outlines an approach where single pellet press testing is combined with modeling to mimic...... the pelletizing behavior of new types of biomass in a large scale pellet mill. This enables a fast estimation of key process parameters such as optimal press channel length and moisture content. Secondly, the study addresses the question of the origin of the observed relationship between pelletizing pressure...

  11. 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.)

  12. Fundamentals of Biomass pellet production

    DEFF Research Database (Denmark)

    Holm, Jens Kai; Henriksen, Ulrik Birk; Hustad, Johan Einar

    2005-01-01

    Pelletizing experiments along with modelling of the pelletizing process have been carried out with the aim of understanding the fundamental physico-chemical mechanisms that control the quality and durability of biomass pellets. A small-scale California pellet mill (25 kg/h) located with the Biomass...

  13. Fire and fire extinguishment in silos. An experimental study[Storage of wood fuel pellets]; Brand och brandslaeckning i siloanlaeggningar. En experimentell studie

    Energy Technology Data Exchange (ETDEWEB)

    Persson, Henry; Blomqvist, Per; Zhenghua Yan

    2007-01-15

    A series of four tests have been conducted with wood pellets stored in a reduced size silo. The tests were conducted in order to increase the knowledge on fire development, detection and extinction technique in silo fires. The project originated from a pre-study on the extinction of silo fires made for the Swedish Rescue Service Agency (SRV). The test silo was built of concrete rings, had a diameter of 1 m, and a height of almost 6 m, which gave a volume of 4.4 m{sup 3}. The silo was filled with wood pellets up to a height of 5 m. A local auto ignition was imitated by a coiled heating wire placed in the pellet bulk centrally in the silo and a self sustaining pyrolysis zone was established within one hour. The silo was instrumented with almost 100 thermocouples as a mean to follow the development of the pyrolysis zone and later the efficiency of the extinguishment. Gas analyses were further made, both in the top of the silo, and at four different levels in the pellet bulk. After 30 hours the extinguishment was initiated using nitrogen (N{sub 2}) and carbon dioxide (CO{sub 2}), respectively. The gas was injected into the bottom of the silo. Two tests were also conducted where gas injection was combined foam application in the top of the silo. The tests showed that the pyrolysis zone preferably spreads downwards in the silo, while moisture and pyrolysis gases form a wave that slowly spreads upwards. It was difficult to detect the fire before the main 'gas wave' reached the pellet surface in the silo top, and detection time was about 20 hours in these tests. The spread of the pyrolysis zone, downward, was even slower. The slow development is probably an explanation of why real silo fires often are rather extensive once discovered. Inerting the silo with either nitrogen or carbon dioxide worked out well in the tests. The gas must be in gas phase and slowly introduced into the silo, as close to the bottom as possible. An efficient extinction implies an air

  14. Pellet transfer apparatus and method

    International Nuclear Information System (INIS)

    DiGrande, J.T.; Huggins, T.B. Sr.; Lambert, D.V.; Roberts, E.

    1991-01-01

    This patent describes a pellet inspection system having a station for inspecting a predetermined parameter of a pellet. It comprises means for aligning and guiding pellets in a first row to be advanced along a linear path past the pellet inspecting station and in a second row previously advanced along the linear path past the pellet inspecting station; and a transfer mechanism operable for engaging at least one of the pellets in each of the first and second rows and moving from an initial position through a forward stroke to advance the first and second rows of pellets along the liner path such that the inspecting station can inspect the preselected parameter of the pellets in the first row as they are advanced successively , the transfer mechanism being operable for disengaging the pellets and moving through a return stroke relative to the stationary advanced first and second rows of pellets back to the initial position

  15. 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

  16. ELM mitigation with pellet ELM triggering and implications for PFCs and plasma performance in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Baylor, Larry R. [ORNL; Lang, P. [EURATOM / UKAEA, Abingdon, UK; Allen, S. L. [Lawrence Livermore National Laboratory (LLNL); Lasnier, C. J. [Lawrence Livermore National Laboratory (LLNL); Meitner, Steven J. [ORNL; Combs, Stephen Kirk [ORNL; Commaux, Nicolas JC [ORNL; Loarte, A. [ITER Organization, Cadarache, France; Jernigan, Thomas C. [ORNL

    2015-08-01

    The triggering of rapid small edge localized modes (ELMs) by high frequency pellet injection has been proposed as a method to prevent large naturally occurring ELMs that can erode the ITER plasma facing components (PFCs). Deuterium pellet injection has been used to successfully demonstrate the on-demand triggering of edge localized modes (ELMs) at much higher rates and with much smaller intensity than natural ELMs. The proposed hypothesis for the triggering mechanism of ELMs by pellets is the local pressure perturbation resulting from reheating of the pellet cloud that can exceed the local high-n ballooning mode threshold where the pellet is injected. Nonlinear MHD simulations of the pellet ELM triggering show destabilization of high-n ballooning modes by such a local pressure perturbation.A review of the recent pellet ELM triggering results from ASDEX Upgrade (AUG), DIII-D, and JET reveals that a number of uncertainties about this ELM mitigation technique still remain. These include the heat flux impact pattern on the divertor and wall from pellet triggered and natural ELMs, the necessary pellet size and injection location to reliably trigger ELMs, and the level of fueling to be expected from ELM triggering pellets and synergy with larger fueling pellets. The implications of these issues for pellet ELM mitigation in ITER and its impact on the PFCs are presented along with the design features of the pellet injection system for ITER.

  17. 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

  18. Design of pellet surface grooves for fission gas plenum

    International Nuclear Information System (INIS)

    Carter, T.J.; Jones, L.R.; Macici, N.; Miller, G.C.

    1986-01-01

    In the Canada deuterium uranium pressurized heavy water reactor, short (50-cm) Zircaloy-4 clad bundles are fueled on-power. Although internal void volume within the fuel rods is adequate for the present once-through natural uranium cycle, the authors have investigated methods for increasing the internal gas storage volume needed in high-power, high-burnup, experimental ceramic fuels. This present work sought to prove the methodology for design of gas storage volume within the fuel pellets - specifically the use of grooves pressed or machined into the relatively cool pellet/cladding interface. Preanalysis and design of pellet groove shape and volume was accomplished using the TRUMP heat transfer code. Postirradiation examination (PIE) was used to check the initial design and heat transfer assumptions. Fission gas release was found to be higher for the grooved pellet rods than for the comparison rods with hollow or unmodified pellets. This had been expected from the initial TRUMP thermal analyses. The ELESIM fuel modeling code was used to check in-reactor performance, but some modifications were necessary to accommodate the loss of heat transfer surface to the grooves. It was concluded that for plenum design purposes, circumferential pellet grooves could be adequately modeled by the codes TRUMP and ELESIM

  19. Pellet injector research and development at ORNL

    International Nuclear Information System (INIS)

    Combs, S.K.; Barber, G.C.; Baylor, L.R.

    1994-01-01

    Oak Ridge National Laboratory has been developing pellet injectors for plasma fueling experiments on magnetic confinement devices for more than 15 years. Recent major applications of the ORNL development program include (1) a tritium-compatible four-shot pneumatic injector for the Tokamak Fusion Test Reactor, (2) a centrifuge pellet injector for the Tore Supra tokamak, and most recently (3) a three-barrel repeating pneumatic injector for the DIII-D tokamak. In addition to applications, ORNL is developing advanced technologies, including high-speed pellet injectors, tritium injectors, and long-pulse pellet feed systems. The high-speed research involves a collaboration between ORNL and ENEA-Frascati in the development of a repeating two-stage light gas gun based on an extrusion-type pellet feed system. Construction of a new tritium-compatible, extruder-based repeating pneumatic injector (8-mm-diam) is complete and will replace the pipe gun in the original tritium proof-of-principle experiment. The development of a steady-state feed system in which three standard extruders operate in tandem is under way. These research and development activities are relevant to the International Thermonuclear Experimental Reactor and are briefly described in this paper

  20. Monitoring and data acquisition of the high speed hydrogen pellet in SPINS

    Energy Technology Data Exchange (ETDEWEB)

    Mukherjee, Samiran Shanti, E-mail: samiran@ipr.res.in; Mishra, Jyotishankar; Gangradey, Ranjana; Dutta, Pramit; Rastogi, Naveen; Panchal, Paresh; Nayak, Pratik; Agarwal, Jyoti; Bairagi, Pawan; Patel, Haresh; Sharma, Hardik

    2016-11-15

    Highlights: • Pellet INjector System with monitoring and data acquisition is described. • A high speed camera was used to view pellet size, and its flight trajectory. • PXI based high speed control system is used data acquisition. • Pellets of length 2–4.8 mm and speed 250–750 m/s were obtained. - Abstract: Injection of solid hydrogen pellets is an efficient way of replenishing the spent fuel in high temperature plasmas. Aiming that, a Single Pellet INjector System (SPINS) is developed at Institute for Plasma Research (IPR), India, to initiate pellet injection related research in SST-1. The pellet injector is controlled by a PXI system based data acquisition and control (DAC) system for pellet formation, precise firing control, data collection and diagnostics. The velocity of high speed moving pellets is estimated by using two sets of light gate diagnostic. Apart from light gate, a fast framing camera is used to measure the pellet size and its speed. The pellet images are captured at a frame rate of ∼200,000 frames per second at (128 × 64) pixel resolution with an exposure time of 1 μs. Using these diagnostic, various cylindrical pellets of length ranging from 2 to 4.8 mm and speed 250–750 m/s were successfully obtained. This paper describes the control and data acquisition system of SPINS, the techniques for measurement of pellet velocity and capturing images of high speed moving pellet.

  1. 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)

  2. Portuguese pellets market: Analysis of the production and utilization constrains

    International Nuclear Information System (INIS)

    Monteiro, Eliseu; Mantha, Vishveshwar; Rouboa, Abel

    2012-01-01

    As opposite in Portugal, the wood pellets market is booming in Europe. In this work, possible reasons for this market behavior are foreseen according to the key indicators of biomass availability, costs and legal framework. Two major constrains are found in the Portuguese pellets market: the first one is the lack of an internal consumption, being the market based on exportations. The second one is the shortage of raw material mainly due to the competition with the biomass power plants. Therefore, the combination of the biomass power plants with pellet production plants seems to be the best option for the pellets production in the actual Portuguese scenario. The main constrains for pellets market has been to convince small-scale customers that pellets are a good alternative fuel, mainly due to the investment needed and the strong competition with natural gas. Besides some benefits in the acquisition of new equipment for renewable energy, they are insufficient to cover the huge discrepancy of the investment in pellets heating. However, pellets are already economic interesting for large utilizations. In order cover a large amount of households, additional public support is needed to cover the supplementary costs of the pellets heating systems. - Highlights: ► There is a lack of internal consumption being the pellets market based on exportation. ► The shortage of raw material is mainly due to the biomass power plants. ► Combining pellet plants with biomass power plants seems to be a wise solution. ► The tax benefits of renewable energy equipments are not enough to cover the higher investment. ► Pellets are already economic interesting for large utilizations in the Portuguese scenario.

  3. Energy wood. Part 2b: Wood pellets and pellet space-heating systems; Holzenergie Teil 2b: Holzpellets und Pelletheizungen / Energie du bois Partie 2b: Granules de bois et installations de chauffage a granules de bois

    Energy Technology Data Exchange (ETDEWEB)

    Nussbaumer, T. [Verenum, Zuerich (Switzerland)

    2002-07-01

    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)

  4. Description of pelletizing facility

    Energy Technology Data Exchange (ETDEWEB)

    Vojin Cokorilo; Dinko Knezevic; Vladimir Milisavljevic [University of Belgrade, Belgrade (Serbia). Faculty of Mining and Geology

    2006-07-01

    A lot of electrical energy in Serbia was used for heating, mainly for domestics. As it is the most expensive source for heating the government announced a National Program of Energy Efficiency with only one aim, to reduce the consumption of electric energy for the heating. One of the contributions to mentioned reduction is production of coal pellets from the fine coal and its use for domestic heating but also for heating of schools, hospitals, military barracks etc. Annual production of fine coal in Serbia is 300,000 tons. The stacks of fine coal present difficulties at each deep mine because of environmental pollution, spontaneous combustion, low price, smaller market etc. To overcome the difficulties and to give the contribution to National Program of Energy Efficiency researchers from the Department of Mining Engineering, the University of Belgrade designed and realized the project of fine coal pelletizing. This paper describes technical aspect of this project. Using a CPM machine Model 7900, a laboratory facility, then a semi-industrial pelletizing facility followed by an industrial facility was set up and produced good quality pellets. The plant comprised a coal fines hopper, conveyor belt, hopper for screw conveyor, screw conveyor, continuous mixer conditioner, binder reservoir, pump and pipelines, pellet mill, product conveyor belt and product hopper. 4 refs., 3 figs., 1 tab.

  5. Torrefaction of wood pellets: New solutions

    Science.gov (United States)

    Zaichenko, V. M.; Shterenberg, V. Ya.

    2017-10-01

    The current state of the market of conventional and torrefied wood pellets and the trends of its development have been analyzed. The advantages and disadvantages of pellets of both types have been compared with other alternative fuels. The consumer segment in which wood pellets are the most competitive has been determined. The original torrefaction technology using exhaust gas heat from a standard gas engine that was developed at the Joint Institute for High Technologies and the scheme of an experimental unit for the elaboration of the technology have been presented. The scheme of the combined operation of a torrefaction unit and a standard hot water boiler, which makes it possible to utilize the heat of exhaust steam-and-gas products of torrefaction with the simultaneous prevention of emissions of harmful substances into the environment, has been proposed. The required correlation between the capacity of the torrefaction unit and the heating boiler house has been estimated for optimal operation under the conditions of the isolated urban village in a region that is distant from the areas of extraction of traditional fuels and, at the same time, has quite sufficient resources of raw materials for the production of wood pellets.

  6. Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

    International Nuclear Information System (INIS)

    Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

    1982-01-01

    The document reports in detail the fuel pin fabrication data and describes the irradiation conditions and history. All the relevant results of the non-destructive and destructive post-irradiation examinations are reported. They include: visual inspection and chemical analysis of crud; length and diameter measurements; neutron radiography and gamma scanning; juncture tests and fission gas analysis (including residual gas in fuel samples); microscopy and alpha + beta/gamma autoradiography; microprobe investigations; burn-up and isotopic analysis; and hydrogen analysis in clad. The data and observations obtained are discussed in detail and conclusions are given. The irradiation and post-irradiation examinations of the R-109 pins have shown the safe, pre-calculable performance of LWR fuel pins containing mixed-oxide sphere-pac fuel with the fissile material mainly present in the large spheres

  7. Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

    International Nuclear Information System (INIS)

    Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

    1981-04-01

    Three fuel pin bundles, R-109/1, 2 and 3, were irradiated in a PWR loop in the HFR at Petten during respectively 131, 57 and 57 effective full power days at average powers of approximately 39 kW.m -1 and at peak powers of approximately 60 kW.m -1 . The results of the post-irradiation examinations of these fuel bundles are presented. (Auth.)

  8. Uranium dioxide pellets

    International Nuclear Information System (INIS)

    Zawidzki, T.W.

    1982-01-01

    A process for the preparation of a sintered, high density, large crystal grain size uranium dioxide pellet is described which involves: (i) reacting a uranyl nitrate of formula UO 2 (NO 3 ) 2 .6H 2 O with a sulphur source, at a temperature of from about 300 deg. C to provide a sulphur-containing uranium trioxide; (ii) reacting the thus-obtained modified uranium trioxide with ammonium nitrate to form an insoluble sulphur-containing ammonium uranate; (iii) neutralizing the thus-formed slurry with ammonium hydroxide to precipitate out as an insoluble ammonium uranate the remaining dissolved uranium; (iv) recovering the thus-formed precipitates in a dry state; (v) reducing the dry precipitate to UO 2 , and forming it into 'green' pellets; and (vi) sintering the pellets in a hydrogen atmosphere at an elevated temperature

  9. 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)

  10. Pneumatic hydrogen pellet injection system for the ISX tokamak

    International Nuclear Information System (INIS)

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

    1979-01-01

    We describe the design and operation of the solid hydrogen pellet injection system used in plasma refueling experiments on the ISX tokamak. The gun-type injector operates on the principle of gas dynamic acceleration of cold pellets confined laterally in a tube. The device is cooled by flowing liquid helium refrigerant, and pellets are formed in situ. Room temperature helium gas at moderate pressure is used as the propellant. The prototype device injected single hydrogen pellets into the tokamak discharge at a nominal 330 m/s. The tokamak plasma fuel content was observed to increase by (0.5--1.2) x 10 19 particles subsequent to pellet injection. A simple modification to the existing design has extended the performance to 1000 m/s. At higher propellant operating pressures (28 bars), the muzzle velocity is 20% less than predicted by an idealized constant area expansion process

  11. Modelling of pellet-clad interaction during power ramps

    International Nuclear Information System (INIS)

    Zhou, G.; Lindback, J.E.; Schutte, H.C.; Jernkvist, L.O.; Massih, A.R.; Massih, A.R.

    2005-01-01

    A computational method to describe the pellet-clad interaction phenomenon is presented. The method accounts for the mechanical contact between fragmented pellets and the zircaloy clad, as well as for chemical reaction of fission products with zircaloy during power ramps. Possible pellet-clad contact states, soft, hard and friction, are taken into account in the computational algorithm. The clad is treated as an elastic-plastic-viscoplastic material with irradiation hardening. Iodine-induced stress corrosion cracking is described by using a fracture mechanics-based model for crack propagation. This integrated approach is used to evaluate two power ramp experiments made on boiling water reactor fuel rods in test reactors. The influence of the pellet-clad coefficient of friction on clad deformation is evaluated and discussed. Also, clad deformations, pellet-clad gap size and fission product gas release for one of the ramped rods are calculated and compared with measured data. (authors)

  12. Modeling pellet impact drilling process

    Science.gov (United States)

    Kovalyov, A. V.; Ryabchikov, S. Ya; Isaev, Ye D.; Ulyanova, O. S.

    2016-03-01

    The paper describes pellet impact drilling which could be used to increase the drilling speed and the rate of penetration when drilling hard rocks. Pellet impact drilling implies rock destruction by metal pellets with high kinetic energy in the immediate vicinity of the earth formation encountered. The pellets are circulated in the bottom hole by a high velocity fluid jet, which is the principle component of the ejector pellet impact drill bit. The experiments conducted has allowed modeling the process of pellet impact drilling, which creates the scientific and methodological basis for engineering design of drilling operations under different geo-technical conditions.

  13. Thermal-mechanical properties of cracked UO2 pellets

    International Nuclear Information System (INIS)

    Williford, R.E.; Mohr, C.L.; Lanning, D.D.

    1980-11-01

    A series of experiments (IFA-431, 432, 513, and 527) sponsored by the Fuel Behavior Research Branch of the USNRC are being irradiated in the Halden Boiling Water Reactor to better define LWR fuel behavior over the normal operating range of power reactor fuel rods. One fuel behavior variable of interest is the thermally induced cracking of UO 2 fuel pellets. The effects of pellet cracking on the effective thermal conductivity and elastic moduli for the fragmented fuel were found to be primarily dependent on the free area in the r, theta plane of the fuel rod. The free area is defined as the area within the cladding inner surface that is not occupied by the fuel fragments themselves

  14. 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

  15. Modification of ELESTRES code with new database of flux depression across the pellet radius

    International Nuclear Information System (INIS)

    Sim, Ki Sub; Park, Kwang Suk; Byun, Taek Sang; Suk, Ho Chun

    1995-01-01

    Modification of ELESTRES CANDU fuel performance code with new database of flux depression across the pellet radius is described, and application results of the improved ELESTRES to the fuel performance data are described. (Author) 4 refs., 4 figs

  16. Tritium pellet injector for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Gouge, M.J.; Baylor, L.R.; Combs, S.K.; Fisher, P.W.; Foust, C.R.; Milora, S.L.

    1992-01-01

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the CY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  17. Considerations on the DEMO pellet fuelling system

    Energy Technology Data Exchange (ETDEWEB)

    Lang, P.T., E-mail: peter.lang@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Day, Ch. [Karlsruhe Institute of Technology, 76021 Karlsruhe (Germany); Fable, E. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Igitkhanov, Y. [Karlsruhe Institute of Technology, 76021 Karlsruhe (Germany); Köchl, F. [Association EURATOM-Ö AW/ATI, Atominstitut, TU Wien, 1020 Vienna (Austria); Mooney, R. [Culham Centre for Fusion Energy, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Pegourie, B. [CEA, IRFM, 13108 Saint-Paul-lez-Durance (France); Ploeckl, B. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Wenninger, R. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); EFDA, Garching (Germany); Zohm, H. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-10-15

    Graphical abstract: - Highlights: • Considerations are made for a core particle fuelling system covering all DEMO requirements. • Particle deposition beyond the pedestal top is needed to achieve efficient fuelling. • Conventional pellet technology enabling launching from the torus inboard side can be used. • Efforts have been taken for integrating a suitable pellet guiding system into the EU DEMO model. • In addition, further techniques bearing potential for advanced fuelling performance are considered. - Abstract: The Demonstration Fusion Power Reactor DEMO is the step foreseen to bridge the gap between ITER and the first commercial fusion power plant. One key element in the European work plan for DEMO is the elaboration of a conceptual design for a suitable core particle fuelling system. First considerations for such a system are presented in this contribution. Following the well-considered ITER solution, most analysis performed in this study assumes conventional pellet technology will be used for the fuelling system. However, taking advantage of the less compressed time frame for the DEMO project, several other techniques thought to bear potential for advanced fuelling performance are considered as well. In a first, basic analysis all actuation parameters at hand and their implications on the fuelling performance were considered. Tentative transport modeling of a reference scenario strongly indicates only particles deposited inside the plasma pedestal allow for efficient fuelling. Shallow edge fuelling results in an unbearable burden on the fuel cycle. Sufficiently deep particle deposition seems technically achievable, provided pellets are launched from the torus inboard at sufficient speed. All components required for a DEMO pellet system capable for high speed inboard pellet launch are already available or can be developed in due time with reasonable efforts. Furthermore, steps to integrate this solution into the EU DEMO model are taken.

  18. Considerations on the DEMO pellet fuelling system

    International Nuclear Information System (INIS)

    Lang, P.T.; Day, Ch.; Fable, E.; Igitkhanov, Y.; Köchl, F.; Mooney, R.; Pegourie, B.; Ploeckl, B.; Wenninger, R.; Zohm, H.

    2015-01-01

    Graphical abstract: - Highlights: • Considerations are made for a core particle fuelling system covering all DEMO requirements. • Particle deposition beyond the pedestal top is needed to achieve efficient fuelling. • Conventional pellet technology enabling launching from the torus inboard side can be used. • Efforts have been taken for integrating a suitable pellet guiding system into the EU DEMO model. • In addition, further techniques bearing potential for advanced fuelling performance are considered. - Abstract: The Demonstration Fusion Power Reactor DEMO is the step foreseen to bridge the gap between ITER and the first commercial fusion power plant. One key element in the European work plan for DEMO is the elaboration of a conceptual design for a suitable core particle fuelling system. First considerations for such a system are presented in this contribution. Following the well-considered ITER solution, most analysis performed in this study assumes conventional pellet technology will be used for the fuelling system. However, taking advantage of the less compressed time frame for the DEMO project, several other techniques thought to bear potential for advanced fuelling performance are considered as well. In a first, basic analysis all actuation parameters at hand and their implications on the fuelling performance were considered. Tentative transport modeling of a reference scenario strongly indicates only particles deposited inside the plasma pedestal allow for efficient fuelling. Shallow edge fuelling results in an unbearable burden on the fuel cycle. Sufficiently deep particle deposition seems technically achievable, provided pellets are launched from the torus inboard at sufficient speed. All components required for a DEMO pellet system capable for high speed inboard pellet launch are already available or can be developed in due time with reasonable efforts. Furthermore, steps to integrate this solution into the EU DEMO model are taken.

  19. User's guide for FREG-3: a computer program to analyze pellet-cladding gap conductance in accordance with fuel-rod irradiation history

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Izumi, Fumio; Fujita, Misao; Ishibashi, Akihiro; Otsubo, Naoaki.

    1976-10-01

    The present report describes user's manual for program FREG-3, and provides a general description of the program and instructions of input/output. FREG-3 estimates the temperature distribution in a fuel rod and the stored energy based on the distribution. The temperature distribution is calculated in accordance with fuel-rod irradiation history. Mechanical properties and models in handling specific problems, such as densification and relocation, are optional in the program. The options are to be given by key word. If appropriate options are selected, the program is used not only as a safety evaluation code, but also as a best evaluation code. (auth.)

  20. 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

  1. Production and Innovative Applications of Cryogenic Solid Pellets

    International Nuclear Information System (INIS)

    Baylor, L.R.; Combs, S.K.; Fisher, P.W.; Foster, C.A.; Foust, C.R.; Gouge, M.J.; Milora, S.L.

    1999-01-01

    For over two decades Oak Ridge National Laboratory has been developing cryogenic pellet injectors for fueling hot, magnetic fusion plasmas. Cryogenic solid pellets of all three hydrogen isotopes have been produced in a size range of 1- to 10-mm diameter and accelerated to speeds from <100 to ∼3000 m/s. The pellets have been formed discretely by cryocondensation in gun barrels and also by extrusion of cryogenic solids at mass flow rates up to ∼0.26 g/s and production rates up to ten pellets per second. The pellets traverse the hot plasma in a fraction of a millisecond and continuously ablate, providing fresh hydrogenic fuel to the interior of the plasma. From this initial application, uses of this technology have expanded to include (1) cryogenic xenon drops or solids for use as a debris-less target in a laser plasma source of X-rays for advanced lithography systems, (2) solid argon and carbon dioxide pellets for surface cleaning or decontamination, and (3) methane pellets in a liquid hydrogen bath for use as an innovative moderator of cold neutrons. Methods of production and acceleration/transport of these cryogenic solids will be described, and examples will be given of their use in prototype systems

  2. Experimental study of curved guide tubes for pellet injection

    International Nuclear Information System (INIS)

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

    1997-01-01

    The use of curved guide tubes for transporting frozen hydrogen pellets offers great flexibility for pellet injection into plasma devices. While this technique has been previously employed, an increased interest in its applicability has been generated with the recent ASDEX Upgrade experimental data for magnetic high-field side (HFS) pellet injection. In these innovative experiments, the pellet penetration appeared to be significantly deeper than for the standard magnetic low-field side injection scheme, along with corresponding greater fueling efficiencies. Thus, some of the major experimental fusion devices are planning experiments with HFS pellet injection. Because of the complex geometries of experimental fusion devices, installations with multiple curved guide tube sections will be required for HFS pellet injection. To more thoroughly understand and document the capability of curved guide tubes, an experimental study is under way at the Oak Ridge National Laboratory (ORNL). In particular, configurations and pellet parameters applicable for the DIII-D tokamak and the International Thermonuclear Experimental Reactor (ITER) were simulated in laboratory experiments. Initial test results with nominal 2.7- and 10-mm-diam deuterium pellets are presented and discussed

  3. Degradation of copepod fecal pellets

    DEFF Research Database (Denmark)

    Poulsen, Louise K.; Iversen, Morten

    2008-01-01

    amount of fecal pellets. The total degradation rate of pellets by the natural plankton community of Oresund followed the phytoplankton biomass, with maximum degradation rate during the spring bloom (2.5 +/- 0.49 d(-1)) and minimum (0.52 +/- 0.14 d(-1)) during late winter. Total pellet removal rate ranged...

  4. Review: study of single-pellet injection experiments and development of pellet injector in JFT-2M

    International Nuclear Information System (INIS)

    Kasai, Satoshi; Miura, Yukitoshi; Hasegawa, Kouichi; Sengoku, Seio

    1987-10-01

    The single pellet injector developed for JFT-2M and the improvement of plasma characteristics in the auxiliary-heated discharges by single-pellet injection are reviewed for the period 1982 - 1986. The pellet injector is a pneumatic type and the designed pellet size is 1.65 mmD x 1.65 mmL and 1 mmD x 1 mmL. The hydrogen, deuterium and mixed (H 2 + D 2 ) pellets can be produced with good reproducibility. Maximum pellet velocity is about 970 m/s (pellet is deuterium and propellant gas is hydrogen). In the pellet injection experiments into auxiliary-heated (NB, ICRF) divertor or limiter discharges, the plasma confinement time is improved by a factor of 1.4 - 1.7 compared with the confinement time in the Ohmic discharges. The achieved confinement time is longer than that on the high confinement mode (H-mode) in gas fueled discharges, although the phenomena are transient. (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. Report on the survey of the commercialization of wood biomass energy. Project on the production of wood pellet fuel; 2001 nendo mokushitsu baiomasu energy jigyoka chosa hokokusho. Mokushitsu peretto nenryo seizo jigyo

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-02-01

    For the purpose of regenerating forestry and contributing to the policy for prevention of global warming, a potential study was made of the commercialization of the wood pellet using low-quality wood materials such as thinnings and wood pieces from lumber mill. In the study, based on the survey of raw materials of wood pellet and the demand amount, the scale of pellet production was assumed, and subjects were arranged toward the basic design of plant, evaluation of economical efficiency and commercialization. As a result of the study, the following subjects were extracted. In the study, the supply of lumbers of 2,800 t/y and securing of demand of about 1,600 t/y were set forth as a premise, but the subject was to secure the initial demand. The pellet combustor was higher in price than the kerosene combustor, and for the imported combustion equipment, the combustion of white pellet was supposed. It is necessary to develop combustor of pellets including the bark. In the trial calculation of the unit price of heat utilization (yen/Mcal), the pellet stove was about 3.3 times as high in price as the kerosene stove. It is necessary to reduce the pellet price down to 30 yen/kg or so by decreasing the cost of pellet production. (NEDO)

  7. 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)

  8. Table-top pellet injector (TATOP) for impurity pellet injection

    Energy Technology Data Exchange (ETDEWEB)

    Szepesi, Tamás, E-mail: szepesi.tamas@wigner.mta.hu [Wigner RCP, RMI, Konkoly Thege 29-33, H-1121 Budapest (Hungary); Herrmann, Albrecht [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Kocsis, Gábor; Kovács, Ádám; Németh, József [Wigner RCP, RMI, Konkoly Thege 29-33, H-1121 Budapest (Hungary); Ploeckl, Bernhard [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-10-15

    Highlights: • A portable pellet injector for solid state pellets was designed. • Aims to study ELM triggering potential of impurity pellets. • Aims for multi-machine comparison of pellet–plasma interaction. • Max. pellet speed: 450 m/s, max. rate: 25 Hz. • Pellet size: 0.5–1.5 mm (diameter). - Abstract: A table-top pellet injector (TATOP) has been designed to fulfill the following scientific aims: to study the ELM triggering potential of impurity pellets, and to make pellet injection experiments comparable over several fusion machines. The TATOP is based on a centrifugal accelerator therefore the complete system is run in vacuum, ensuring the compatibility with fusion devices. The injector is able to launch any solid material (stable at room temperature) in form of balls with a diameter in the 0.5–1.5 mm range. The device hosts three individual pellet tanks that can contain e.g. pellets of different materials, and the user can select from those without opening the vacuum chamber. A key element of the accelerator is a two-stage stop cylinder that reduces the spatial scatter of pellets exiting the acceleration arm below 6°, enabling the efficient collection of all fired pellets. The injector has a maximum launch speed of 450 m/s. The launching of pellets can be done individually by providing TTL triggers for the injector, giving a high level of freedom for the experimenter when designing pellet trains. However, the (temporary) firing rate cannot be larger than 25 Hz. TATOP characterization was done in a test bed; however, the project is still in progress and before application at a fusion oriented experiment.

  9. From a single pellet press to a bench scale pellet mill - Pelletizing six different biomass feedstocks

    DEFF Research Database (Denmark)

    Puig Arnavat, Maria; Shang, Lei; Sárossy, Zsuzsa

    2016-01-01

    The increasing demand for biomass pellets requires the investigation of alternative raw materials for pelletizetion. In the present paper, the pelletization process of fescue, alfalfa, sorghum, triticale, miscanthus and willow is studied to determine if results obtained in a single pellet press (...

  10. Pellet imaging techniques on ASDEX

    International Nuclear Information System (INIS)

    Wurden, G.A.; Buechl, K.; Hofmann, J.; Lang, R.; Loch, R.; Rudyj, A.; Sandmann, W.

    1990-01-01

    As part of a USDOE/ASDEX collaboration, a detailed examination of pellet ablation in ASDEX with a variety of diagnostics has allowed a better understanding of a number of features of hydrogen ice pellet ablation in a plasma. In particular, fast gated photos with an intensified Xybion CCD video camera allow in-situ velocity measurements of the pellet as it penetrates the plasma. With time resolution of typically 100 nanoseconds and exposures every 50 microseconds, the evolution of each pellet in a multi-pellet ASDEX tokamak plasma discharge can be followed. When the pellet cloud track has striations, the light intensity profile through the cloud is hollow (dark near the pellet), whereas at the beginning or near the end of the pellet trajectory the track is typically smooth (without striations) and has a gaussian-peaked light emission profile. New, single pellet Stark broadened D α D β , and D γ spectra, obtained with a tangentially viewing scanning mirror/spectrometer with Reticon array readout, are consistent with cloud densities of 2 x 10 17 cm -3 or higher in the regions of strongest light emission. A spatially resolved array of D α detectors shows that the light variations during the pellet ablation are not caused solely by a modulation of the incoming energy flux as the pellet crosses rational q-surfaces, but instead are a result of a dynamic, non-stationary, ablation process. 20 refs., 4 figs

  11. Thermodynamics of pellet-cladding interaction

    International Nuclear Information System (INIS)

    Kyoh, Bunkei; Fuji, Kensho

    1987-01-01

    Equilibrium thermodynamic calculations are performed on the U-Zr-Cs-I-O system that is assumed to exist in the fuel-cladding gap of light water reactor (LWR) fuel under pellet-cladding interaction (PCI) failure condition. For this purpose a computer program called SOLGASMIX-PV for the calculation of complex multi-component equilibria is used, and the results of postirradiation examination are interpreted. The analysis of the thermodynamics of the system U-Zr-Cs-I-O indicates that cesium and iodine are assumed to be released from fuel pellet into the fuel-cladding gap as CsI, therefore, the Cs/I ratio in fuel-cladding bonding zone is one. The important condensed phases in this region are UO 2 , U 3 O 8 , Cs 2 U 2 O 7 , Cs 2 U 15 O 46 , ZrO 2 and CsI, and the major gaseous species are CsI, I 2 and I. Under this situation where Cs/I ratio is one, cesium-zirconate is not present. If, however, cesium rich phase is partially present then cesium will be associated with zirconium, possibly as Cs 2 ZrO 3 . (author)

  12. Description of pelletizing facility

    Directory of Open Access Journals (Sweden)

    Čokorilo Vojin

    2006-01-01

    Full Text Available A lot of electrical energy in Serbia was used for heating, mainly for domes- tics. As it is the most expensive source for heating the government announced a National Program of Energy Efficiency with only one aim, to reduce the consumption of electric energy for the heating. One of the contributions to mentioned reduction is production of coal pellets from the fine coal and its use for domestic heating but also for heating of schools, hospitals, military barracks, etc. Annual production of fine coal in Serbia is 300,000 tons. The stacks of fine coal make a lot difficulties to the each mine because of environmental pollution, spontaneous combustion, low price, smaller market, etc. To prevent the difficulties and to give the contribution to National Program of Energy Efficiency researchers from the Department of Mining Engineering, University of Belgrade, designed and realized the project of fine coal pelletizing. This paper describes technical aspect of this project.

  13. 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)

  14. Opportunities for Pellet Trade - Towards a Single European Pellet Market

    International Nuclear Information System (INIS)

    Pigaht, Maurice; Janssen, Rainer; Rutz, Dominik; Boehm, Thorsten; Vasen, Norbert; Vegas, Laura; Karapanagiotis, Nicolas

    2006-01-01

    The potential for Pellets trade in Europe was researched and assessed. Such trade is of key importance for the development of a European pellet market of sufficient supply, demand, price and quality standards. Three target markets were taken as case studies for the trade assessment: Greece, Spain and Italy. All three markets stand to profit greatly from international trade. For these markets, pellet imports could supply the basis for the development of a domestic boiler market. At the same time, pellet exports would allow the planning of larger pellet production plants. Whilst these additional costs amount to some 10-20% of the Pellets price, they are financially acceptable, especially for new markets and 'peaks' in the demand/supply of established markets

  15. Deuterium pellet injection in the TFR Tokamak

    International Nuclear Information System (INIS)

    Lazare, O.

    1985-07-01

    Injecting fresh fuel deep inside the plasma of a thermonuclear reactor appears to be necessary; the only way to do that is to inject fast solid deuterium pellets. The existing theoretical, technical and experimental aspects of this method are presented. The experiments on TFR have confirmed that injecting pellets is technically feasible; a new kind of injector is presented. The injection does not degrade stability nor confinement of the plasma. The study of the transient phenomena occuring during the injection has proved to be an efficient way to investigate particles and energy transport in the discharge; in particular, a fast transport phenomenon, similar to those occuring during disruptions, has been studied in details. Conclusions about disruptions are drawn. (Ref 101) [fr

  16. Modeling of the PWR fuel mechanical behaviour and particularly study of the pellet-cladding interaction in a fuel rod; Contribution a la modelisation du comportement mecanique des combustibles REP sous irradiation, avec en particulier le traitement de l`interaction pastille-gaine dans un crayon combustible

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. PELLETS AND PELLETIZATION: EMERGING TRENDS IN THE PHARMA INDUSTRY.

    Science.gov (United States)

    Zaman, Muhammad; Saeed-Ul-Hassan, Syed; Sarfraz, Rai Muhammad; Batool, Nighat; Qureshi, Muhammad Junaid; Akram, Muhammad Abdullah; Munir, Saiqa; Danish, Zeeshan

    2016-11-01

    The present time is considered as an era of advancements in drug delivery systems. Different novel approaches are under investigation that range from uniparticulate to multi particulate system, macro to micro and nano particulate systems. Pelletization is one of the novel drug delivery technique that provides an effective way to deliver the drug in modified pattern. It is advantageous in providing site specific delivery of the drug. Drugs with unpleasant taste, poor bioavailability and short biological half-life can be delivered efficiently through pellets. Their reduced size makes them more valuable as compared to the conventional drug deliv- ery system. Different techniques are used to fabricate the pellets such as extrusion and spheronization, hot melt extrusion, powder layering, suspension or solution layering, freeze pelletization and pelletization by direct compression method. Various natural polymers including xanthan gum, guar gum, tragacanth and gum acacia, semisynthetic polymers like cellulose derivatives, synthetic polymers like derivatives of acrylamides, can be used in pellets formulation. Information provided in this review is collected from various national and intemational research articles, review articles and literature available in the books. The purpose of the current review is to discuss pellets, their characterizations, different techniques of pelletization and the polymers with potential of being suitable for pellets formulation.

  18. Recycling process of Mn-Al doped large grain UO2 pellets

    International Nuclear Information System (INIS)

    Nam, Ik Hui; Yang, Jae Ho; Rhee, Young Woo; Kim, Dong Joo; Kim, Jong Hun; Kim, Keon Sik; Song, Kun Woo

    2010-01-01

    To reduce the fuel cycle costs and the total mass of spent light water reactor (LWR) fuels, it is necessary to extend the fuel discharged burn-up. Research on fuel pellets focuses on increasing the pellet density and grain size to increase the uranium contents and the high burnup safety margins for LWRs. KAERI are developing the large grain UO 2 pellet for the same purpose. Small amount of additives doping technology are used to increase the grain size and the high temperature deformation of UO 2 pellets. Various promising additive candidates had been developed during the last 3 years and the MnO-Al 2 O 3 doped UO 2 fuel pellet is one of the most promising candidates. In a commercial UO 2 fuel pellet manufacturing process, defective UO 2 pellets or scraps are produced and those should be reused. A common recycling method for defective UO 2 pellets or scraps is that they are oxidized in air at about 450 .deg. C to make U 3 O 8 powder and then added to UO 2 powder. In the oxidation of a UO 2 pellet, the oxygen propagates along the grain boundary. The U 3 O 8 formation on the grain boundary causes a spallation of the grains. So, size and shape of U 3 O 8 powder deeply depend on the initial grain size of UO 2 pellets. In the case of Mn-Al doped large grain pellets, the average grain size is about 45μm and about 5 times larger than a typical un-doped UO 2 pellet which has grain size of about 8∼10μm. That big difference in grain size is expected to cause a big difference in recycled U 3 O 8 powder morphology. Addition of U 3 O 8 to UO 2 leads to a drop in the pellet density, impeding a grain growth and the formation of graph- like pore segregates. Such degradation of the UO 2 pellet properties by adding the recycled U 3 O 8 powder depend on the U 3 O 8 powder properties. So, it is necessary to understand the property and its effect on the pellet of the recycled U 3 O 8 . This paper shows a preliminary result about the recycled U 3 O 8 powder which was obtained by

  19. 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

  20. 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

  1. Carbon pellet cloud striations

    International Nuclear Information System (INIS)

    Parks, P.B.

    1989-01-01

    Fine scale striations, with alternating rows of bright and dark zones, have been observed in the ablation clouds of carbon pellets injected into the TEXT tokamak. The striations extend along the magnetic field for about 1 cm with quite regular cross-field variations characterized by a wavelength of a few mm. Their potential as a diagnostic tool for measuring q-profiles in tokamaks provides motivation for investigating the origin of the striations. The authors propose that the striations are not due to a sequence of high and low ablation rates because of the finite thermal magnetic islands localized at rational surfaces, q = m/n, would be responsible for reducing the electron flux to the pellet region; the length of the closed field line which forms the local magnetic axis of the island is too long to prevent a depletion of plasma electrons in a flux tube intercepting the pellet for the duration 2 rp / vp . Instead, they propose that striations are the manifestation of the saturated state of growing fluctuations inside the cloud. The instability is generated by E x B rotation of the ablation cloud. The outward centrifugal force points down the ablation density gradient inducing the Rayleigh-Taylor instability. The instability is not present for wave numbers along the field lines, which may explain why the striations are long and uniform in that direction. The E field develops inside the ablation cloud as a result of cold electron return currents which are induced to cancel the incoming hot plasma electron current streaming along the field lines

  2. Pellets direct from the forest

    International Nuclear Information System (INIS)

    Keel, A.

    2006-01-01

    This article takes a look at developments in the market for wood pellets and their production from forest wood. The general situation in the booming pellets market is reviewed and the potential of this climate-neutral form of heating is discussed. Figures and prognoses on the use of wood pellets are presented. In particular, the potential for the use of forestry wood supplies to augment the use of wood wastes and sawdust from sawmills is looked at

  3. Power matching for pellet fusion

    International Nuclear Information System (INIS)

    Martin, R.L.; Arnold, R.C.

    1976-01-01

    The number of beams required for optimum power transfer from a given power source to the surface of a pellet is derived. The result is valid for linear optical systems, hence, for pellet fusion by laser or high energy ion beams. The optimum number of beams turns out to be inconceivably large for any practical system. Practical pellet fusion by lasers or high energy heavy ion beams must thus compromise physical principles in favor of reduced cost and optical complexity

  4. Direct dissolution and supercritical fluid extraction of uranium from UO2 powder, granule, green pellet and sintered pellet

    International Nuclear Information System (INIS)

    Rao, Ankita; Kumar, Pradeep; Ramakumar, K.L.

    2009-01-01

    In the present work, direct dissolution and extraction of UO 2 from the solid rejects various stages of fuel fabrication viz. powder granules green pellet and, sintered pellet has been studied. Powder and granules could be easily dissolved in TBP-HNO 3 complex at 50 deg C., whereas in case of green and sintered pellets at elevated temperature at raised to 80 deg C in TBP-HNO 3 complex. With supercritical (SC) CO 2 alone the efficiency was ∼70%. But with SC CO 2 +2.5% TBP, the efficiency was ∼95% for powder and granules, and ∼60% for green and sintered pellets. Nearly complete extraction (∼99%) was achievable for SC CO 2 + 2.5 % TTA in all cases. The method has distinct advantage of elimination of acid usage and minimization of liquid waste generation. (author)

  5. Conceptual design of ICF reactor SENRI, Part II. Advances in design and pellet gain scaling

    International Nuclear Information System (INIS)

    Ido, S.; Mima, K.; Nakai, S.; Tsuji, R.; Yamanaka, C.

    1984-01-01

    This chapter reviews the recent design studies on reactor concepts with magnetically guided lithium flow, SENRI-I, SENRI-IA and SENRI-II. The routes from the present status to power reactors and an advanced fuel pellet concept is also discussed. Topics covered include pellet design, magnetohydrodynamic design of liquid lithium flow; reactor cavity concepts with magnetically guided lithium flow, a thermo-hydraulic analysis, a tritium recovery system; and an advanced fuel pellet concept for an inertial confinement fusion (ICF) reactor without a tritium breeding blanket. An advanced fuel pellet for an ICF reactor without a T breeder was studied in the model calculations, which showed sufficiently high values of pellet gain. Includes a table and 8 diagrams

  6. Hydrogen pellet injection device

    International Nuclear Information System (INIS)

    Kanno, Masahiro.

    1992-01-01

    In a hydrogen pellet injection device, a nozzle block having a hydrogen gas supply channel is disposed at the inner side of a main cryogenic housing, and an electric resistor is attached to the block. Further, a nozzle block and a hydrogen gas introduction pipe are attached by way of a thermal insulating spacer. Electric current is supplied to the resistor to positively heat the nozzle block and melt remaining solid hydrogen in the hydrogen gas supply channel. Further, the effect of temperature elevation due to the resistor is prevented from reaching the side of the hydrogen gas introduction pipe by the thermal insulation spacer. That is, the temperature of the nozzle block is directly and positively elevated, to melt the solid hydrogen rapidly. Preparation operation from the injection of the hydrogen pellet to the next injection can be completed in a shorter period of time compared with a conventional case thereby enabling to make the test more efficient. Further, only the temperature of the nozzle block is elevated with no effect of temperature elevation due to the resistor to other components by the thermal insulation flange. (N.H.)

  7. Investigating the properties of residues. Characterization of pellets from fermentation residues; Den Eigenschaften der Reststoffe auf der Spur. Untersuchung widmet sich der Charakterisierung von Pellets aus Gaerresten

    Energy Technology Data Exchange (ETDEWEB)

    Kratzeisen, Martin; Mueller, Joachim [Hohenheim Univ., Stuttgart (Germany). Inst. fuer Agrartechnik; Starcevic, Nikica [Hohenheim Univ., Stuttgart (Germany). Inst. fuer Agrartechnik; Strabag Umweltanlagen GmbH, Muenchen (Germany). Projekt Produktentwicklung/Schlammbehandlung

    2009-09-15

    Fermentation residues are by-products of the biogas process. Farmers use them as fertilizers, but as the size of biogas plants grows, so does the residues volume. It is now too much for local use, and transport to other sites is expensive. Fuel pellets production may be an alternative. Pellets from fermentation residues are not accepted as yet because too little is known about their characteristics. The contribution describes an investigation that intends to identify the fuel characteristics of pellets from fermentation residues. (orig.)

  8. Wood pellet production costs under Austrian and in comparison to Swedish framework conditions

    NARCIS (Netherlands)

    Obernberger, I.; Thek, G.

    2004-01-01

    Owing to the rapidly increasing importance of pellets as high-quality biomass fuel in Austria and Europe within the last years, many companies, mainly from the wood industry, are thinking of entering this market. The calculation of the production costs before starting a pellet plant is essential for

  9. A pellet-clad interaction failure criterion

    International Nuclear Information System (INIS)

    Howl, D.A.; Coucill, D.N.; Marechal, A.J.C.

    1983-01-01

    A Pellet-Clad Interaction (PCI) failure criterion, enabling the number of fuel rod failures in a reactor core to be determined for a variety of normal and fault conditions, is required for safety analysis. The criterion currently being used for the safety analysis of the Pressurized Water Reactor planned for Sizewell in the UK is defined and justified in this paper. The criterion is based upon a threshold clad stress which diminishes with increasing fast neutron dose. This concept is consistent with the mechanism of clad failure being stress corrosion cracking (SCC); providing excess corrodant is always present, the dominant parameter determining the propagation of SCC defects is stress. In applying the criterion, the SLEUTH-SEER 77 fuel performance computer code is used to calculate the peak clad stress, allowing for concentrations due to pellet hourglassing and the effect of radial cracks in the fuel. The method has been validated by analysis of PCI failures in various in-reactor experiments, particularly in the well-characterised power ramp tests in the Steam Generating Heavy Water Reactor (SGHWR) at Winfrith. It is also in accord with out-of-reactor tests with iodine and irradiated Zircaloy clad, such as those carried out at Kjeller in Norway. (author)

  10. Pelletizing properties of torrefied spruce

    DEFF Research Database (Denmark)

    Stelte, Wolfgang; Clemons, Craig; Holm, Jens K.

    2011-01-01

    analysis revealed a cohesive failure mechanism due to strong inter-particle bonding in spruce pellets as a resulting from a plastic flow of the amorphous wood polymers, forming solid polymer bridges between adjacent particles. Fracture surfaces of pellets made from torrefied spruce possessed gaps and voids...

  11. 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

  12. 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.

  13. Tritium proof-of-principle pellet injector results

    International Nuclear Information System (INIS)

    Fisher, P.W.; Fehling, D.T.; Gouge, M.J.; Milora, S.L.

    1989-01-01

    The tritium proof-of-principle (TPOP) experiment was built by Oak Ridge National Laboratory (ORNL) to demonstrate the feasibility of forming solid tritium pellets and accelerating them to high velocities for fueling future fusion reactors. TPOP used a pneumatic pipe-gun with a 4-mm-i.d. by 1-m-long barrel. Nearly 1500 pellets were fired by the gun during the course of the experiment; about a third of these were tritium or mixtures of deuterium and tritium. The system also contained a cryogenic 3 He separator that reduced the 3 He level to <0.005%. Pure tritium pellets were accelerated to 1400 m/s. Experiments evaluated the effect of cryostat temperature and fill pressure on pellet size, the production of pellets from mixtures of tritium and deuterium, and the effect of aging on pellet integrity. The tritium phase of these experiments was performed at the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. About 100 kCi of tritium was processed through the apparatus without incident. 8 refs., 7 figs

  14. Potential greenhouse gas benefits of transatlantic wood pellet trade

    International Nuclear Information System (INIS)

    Dwivedi, Puneet; Khanna, Madhu; Bailis, Robert; Ghilardi, Adrian

    2014-01-01

    Power utility companies in the United Kingdom are using imported wood pellets from the southern region of the United States for electricity generation to meet the legally binding mandate of sourcing 15% of the nation’s total energy consumption from renewable sources by 2020. This study ascertains relative savings in greenhouse gas (GHG) emissions for a unit of electricity generated using imported wood pellet in the United Kingdom under 930 different scenarios: three woody feedstocks (logging residues, pulpwood, and logging residues and pulpwood combined), two forest management choices (intensive and non-intensive), 31 plantation rotation ages (year 10 to year 40 in steps of 1 year), and five power plant capacities (20–100 MW in steps of 20 MW). Relative savings in GHG emissions with respect to a unit of electricity derived from fossil fuels in the United Kingdom range between 50% and 68% depending upon the capacity of power plant and rotation age. Relative savings in GHG emissions increase with higher power plant capacity. GHG emissions related to wood pellet production and transatlantic shipment of wood pellets typically contribute about 48% and 31% of total GHG emissions, respectively. Overall, use of imported wood pellets for electricity generation could help in reducing the United Kingdom’s GHG emissions. We suggest that future research be directed to evaluation of the impacts of additional forest management practices, changing climate, and soil carbon on the overall savings in GHG emissions related to transatlantic wood pellet trade. (paper)

  15. Design of a tritium pellet injector for TFTR