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Sample records for ii pellet clad

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  2. 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)

  3. 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)

  4. 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)

  5. 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

  6. 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.)

  7. 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.)

  8. 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)

  9. 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)

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

    International Nuclear Information System (INIS)

    Esteves, A.M.

    1991-01-01

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

  11. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding

    International Nuclear Information System (INIS)

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed

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

    International Nuclear Information System (INIS)

    Engelman, G.P.

    1978-10-01

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

  13. A study of friction and axial effects in pellet-clad mechanical interaction

    International Nuclear Information System (INIS)

    Harriague, Santiago; Mayer, J.E.

    1982-01-01

    An analysis is made of the effect of friction and axial forces along the fuel rod in the pellet-cladding mechanical interaction in a commercial reactor under a power-up ramp. The effect of different pellet and rod shapes on their behaviour was also determined. A linear thermoelastic computer program was used in order to obtain the stiffness matrix of a compound structure from the stiffness of its components. Pellet-cladding displacements, localized deformations of the cladding in the interfaces between pellets, as well as pellet and cladding axial deformations were determined for different power axial profiles as well as for pellets with and without dishing and with height/diameter ratios of 1.7, 1 and 0.5. (M.E.L.) [es

  14. Simulation of a pellet-clad mechanical interaction with ABAQUS and its verification

    International Nuclear Information System (INIS)

    Cheon, J.-S.; Lee, B.-H.; Koo, Y.-H.; Sohn, D.-S.; Oh, J.-Y.

    2003-01-01

    Pellet-clad mechanical interaction (PCMI) during power transients for MOX fuel is modelled by a FE method. The PCMI model predicts well clad elongation during power ramp and relaxation during power hold except the fuel behaviour during a power decrease. Higher fiction factor results in the earlier occurrence of PCMI and more enhanced clad elongation. The relaxation is dependent on the irradiation creep rate of the pellet and axial compressive force. Verification of the PCMI model was done using recent MOX experimental data. Temperature and clad elongation for the fuel rod can be evaluated in a reasonable way

  15. 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

  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. 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

  18. 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.)

  19. 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)

  20. A study of friction and axial effects in pellet-clad mechanical interaction

    International Nuclear Information System (INIS)

    Harriague, S.; Meyer, J.E.

    1983-01-01

    An analysis is made of the effect of friction forces at the pellet-cladding contact points on the behaviour of a fuel rod under a power-up ramp. A thermoelastic description of the pellets is given; the stiffness matrix and initial displacements are obtained from a finite element calculation. The cladding is considered to behave as a thermoelastic thin shell. A method is developed to assemble the stiffness of each pellet and corresponding cladding section on a fuel rod, resulting in an explicit description of the whole stack. The assumption of thermoelasticity allows for a very fast calculation, even when including hundreds of pellets under an arbitrary axial distribution of power. Results showing the pattern of friction and axial forces, and relative and localized displacements along the rod, are presented. In most cases, pellets at the top of the stack slide with respect to the clad. As a result of the build-up of axial forces due to friction, pellets at lower positions in the fuel column may show, at the contact positions, no relative displacements with respect to the cladding. The effect of pellet dishing and L/D ratio on the axial strains and local deformations are shown. The predictions are consistent with the experimental observations on the effect of pellet shape. Finally, a discussion is made of the results of this study. The use of these results as a guideline for establishing proper boundary conditions in a non-linear PCMI model (i.e., including plasticity and pellet cracking) are also discussed. (author)

  1. 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.

  2. 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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  4. 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

  5. An example of coupling behaviour-damage-environment in polycrystals. Application to Pellet-Cladding Interaction

    International Nuclear Information System (INIS)

    Diard, Olivier

    2001-01-01

    Zircaloy-4 cladding is the first containment barrier for fission products, and its integrity must therefore be ensured in nominal and accidental situations. However, stress corrosion induced cracks may appear due to a strong pellet-cladding interaction. It is therefore important to model this interaction and crack growth and propagation to establish non-damage criteria. Thus, this research thesis aims at developing a modelling covering both issues (pellet-cladding interaction, and stress corrosion cracking) and allowing macroscopic and microscopic scales to be coupled. After a bibliographical synthesis on iodine-induced stress corrosion cracking and similar phenomena, the author presents the model proposed for the pellet-cladding interaction: phenomena to be taken into account, phenomenological and macroscopic behaviour laws used respectively for pellet and cladding. An extended version of an existing cladding viscoplastic model is proposed. Stress and strain fields in the cladding are obtained, notably in the contact zone. In the next part, the author presents various numerical tools developed or used to model multi-crystalline aggregates, and the model of crystalline plasticity used to simulate cladding behaviour at the microstructure scale. Effects of mesh density, element types and anisotropic elasticity are also discussed. The next chapter addresses the mechanical-chemical coupling. Some coupling formulas are presented for simple cases in order to define the effective diffusion coefficient. The last part reports the modelling of intergranular damage: definition of a damage criterion at the granular scale, assessment of stresses at grain boundaries, and effect of crystallographic neighbouring. A model of grain boundary damage is also proposed. This model is assessed on Failure Mechanics test samples and on simple microstructures. The application of the whole numerical model is reported [fr

  6. Effect of known clad and pellet reactions on the GEC ESL design of dry vault store

    International Nuclear Information System (INIS)

    Wheeler, D.

    1984-01-01

    The more important clad and pellet reactions and their temperature dependence are briefly reviewed, followed by an outline of the economical GEC ESL interim spent fuel storage concept that highlights: The ability of the concept to reduce the temperature of the fuel to values where the clad and pellet reactions are minimal. The containment philosophy that enables any consequences of the reactions to be safely retained to ALARA principles. The maintenance of an air storage environment that can never be lost. The utilisation of passive, naturally induced cooling regimes. The ability to continuously monitor for long-term degradation, together with ease of inspection at any time during storage

  7. 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

  8. 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.)

  9. 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

  10. 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)

  11. Flat cladding and pellets in the design of an irradiation target

    International Nuclear Information System (INIS)

    Yorio, Daniel; Denis, Alicia C.; Soba, Alejandro; Beuter, Oscar; Marajofsky, Adolfo

    2003-01-01

    The design of an enriched uranium irradiation target made of flat pellets and cladding is proposed in order to improve the fission Mo 99 production. The variation range of each one of the parameters is studied and the basic design of the target is given

  12. Finite element modeling of pellet-clad mechanical interaction with ABAQUS

    International Nuclear Information System (INIS)

    Cheon, C. S.; Lee, B. H.; Koo, Y. H.; Oh, J. Y.; Son, D. S.

    2002-01-01

    Pellet-clad mechanical interaction (PCMI) was modelled by an axisymmetric finite element method. Thermomechanical models of pellet and clad materials and a contact model for their interaction have been implemented in addition to the application of appropriate boundary conditions so that the FE model was configured. Temperature and displacement were evaluated through a coupled analysis using a general purposed FE code, ABAQUS. Also, a batch program has been developed to efficiently deal with a series of jobs such as making an interface with a fuel performance code, the generation of an input deck for ABAQUS code and its execution, and an interpretation of the output. Under various conditions, results from the present FE model were analyzed. Preliminary verification was conducted by comparing the clad elongation measured during an in-pile PCMI experiment with that calculated by means of the developed FE model

  13. 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)

  14. 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

  15. 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

  16. 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)

  17. Analysis of mechanical and chemical pellet-clad interaction during power ramps

    International Nuclear Information System (INIS)

    Vogl, W.; Hering, W.; Peehs; Lavake, J.

    1979-01-01

    A research and development program is being conducted by KWU and C-E to investigate Pellet/Clad Interaction (PCI) in LWR fuel rods during power ramping. Out-of-pile iodine stress corrosion cracking studies, in-pile ramp experiments and hot cell chemical and metallographical post-irradiation examinations are being performed to study and evaluate both the power limitations and the basic mechanisms of PCI as well as practical methods to improve ramping performance. (orig.)

  18. 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

  19. 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

  20. Pellet Cladding Mechanical Interaction Modeling Using the Extended Finite Element Method

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin W.; Jiang, Wen; Dolbow, John E.; Peco, Christian

    2016-09-01

    model stress concentrations induced by fuel fractures at the fuel/cladding interface during pellet cladding mechanical interaction (PCMI). This is accomplished by enhancing the thermal and mechanical contact enforcement algorithms employed by BISON to permit their use in conjunction with XFEM. The results from this methodology are demonstrated to be equivalent to those from using meshed discrete cracks. While the results of the two methods are equivalent for the case of a stationary crack, it is demonstrated that XFEM provides the additional flexibility of allowing arbitrary crack initiation and propagation during the analysis, and minimizes model setup effort for cases with stationary cracks.

  1. Streamlined analysis technique for the evaluation of pellet clad interaction in PWR reload cores

    International Nuclear Information System (INIS)

    Beard, Ch.; Morita, T.; Brown, J.

    2007-01-01

    For some applications, an analysis is required to explicitly demonstrate that fuel failure due to pellet-clad interaction (PCI) is prevented by the core limits and the protection system for both Condition I (normal operation) operation and for Condition II (events of moderate frequency) events. This analysis needs to address the entire range of normal operation allowed by the Technical Specifications and all Condition II transients. The obvious approach which has been utilized for many years is a simulation of normal operation power maneuvers followed by explicit Condition II transients as a function of key core parameters. This is a sampling approach and has concerns about the overall coverage of the potential space. An alternative approach is the 3D FAC power distribution analysis methodology that was based upon the Westinghouse Relaxed Axial Offset Control Strategy (RAOC) evaluation process. The 3D FAC methodology uses a parametric representation of variables affecting the power distributions, defining a grid mesh over a space of Condition I and Condition II parameters. The operation space is defined by a power range, temperature range, rod position range, axial offset range, core protection limits and representative xenon distributions. Then the 3D FAC evaluation consists of systematically calculating the 3D power distribution and margin to the core and fuel limits for each mesh point of this multi-dimensional space. The PCI margin is obtained by the comparison of the 3D power distributions over the Condition II space and the 3D maximum allowed power, which is dependent on the fuel rod history. The fuel history model utilizes the power history developed in the 3-dimensional nuclear analysis code to define local powers for the specified fuel rods to be analyzed. It tracks the rod history and provides the maximum allowed power for the point. This model is appropriate for base load operation, extended reduced power operation, return to power operation and

  2. Streamlined analysis technique for the evaluation of pellet clad interaction in PWR reload cores

    Energy Technology Data Exchange (ETDEWEB)

    Beard, Ch.; Morita, T.; Brown, J. [Westinghouse Electric Company, LLC, Nuclear Fuel Div., Pittsburgh, PA (United States)

    2007-07-01

    For some applications, an analysis is required to explicitly demonstrate that fuel failure due to pellet-clad interaction (PCI) is prevented by the core limits and the protection system for both Condition I (normal operation) operation and for Condition II (events of moderate frequency) events. This analysis needs to address the entire range of normal operation allowed by the Technical Specifications and all Condition II transients. The obvious approach which has been utilized for many years is a simulation of normal operation power maneuvers followed by explicit Condition II transients as a function of key core parameters. This is a sampling approach and has concerns about the overall coverage of the potential space. An alternative approach is the 3D FAC power distribution analysis methodology that was based upon the Westinghouse Relaxed Axial Offset Control Strategy (RAOC) evaluation process. The 3D FAC methodology uses a parametric representation of variables affecting the power distributions, defining a grid mesh over a space of Condition I and Condition II parameters. The operation space is defined by a power range, temperature range, rod position range, axial offset range, core protection limits and representative xenon distributions. Then the 3D FAC evaluation consists of systematically calculating the 3D power distribution and margin to the core and fuel limits for each mesh point of this multi-dimensional space. The PCI margin is obtained by the comparison of the 3D power distributions over the Condition II space and the 3D maximum allowed power, which is dependent on the fuel rod history. The fuel history model utilizes the power history developed in the 3-dimensional nuclear analysis code to define local powers for the specified fuel rods to be analyzed. It tracks the rod history and provides the maximum allowed power for the point. This model is appropriate for base load operation, extended reduced power operation, return to power operation and

  3. Effect of PWR Re-start ramp rate on pellet-cladding interactions

    International Nuclear Information System (INIS)

    Yagnik, S.K.; Chang, B.C.; Sunderland, D.J.

    2005-01-01

    To mitigate pellet-cladding interaction (PCI) leading to fuel rod failures, fuel suppliers specify reactor power ramp rate limitations during reactor start-up after an outage. Typical re-start ramp rates are restricted and range between 3-4% per hour of full reactor power above a threshold power level. Relaxation of threshold power and ramp rate restrictions has the potential to improve plant economics. The paper will compare known re-start power ascension procedures employed in the US, German, French and Korean PWRs after a refuelling outage. A technical basis for optimising power ascension procedures during reactor start-up can be developed using analytical modelling. The main objective of the modelling is to determine the potential for PCI failure for various combinations of threshold power levels and ramp rate levels. A key element of our analysis is to estimate the decrease in margin to cladding failure by ISCC based on a time-temperature-stress failure criterion fashioned Act a cumulative cladding damage index. The analysis approach and the cladding damage model will be described and the results from three case studies based on the FALCON fuel rod behaviour code will be reported. We conclude that the PCI behaviour is more affected by ramp rate and threshold power than by the fuel design and that the fuel power history is the most important parameter. (authors)

  4. 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

  5. PECITIS-II, a computer program to predict the performance of collapsible clad UO2 fuel elements

    International Nuclear Information System (INIS)

    Anand, A.K.; Anantharaman, K.; Sarda, V.

    1978-01-01

    The Indian power programme envisages the use of PHWRs, which use collapsible clad UO 2 fuel elements. A computer code, PECITIS-II, developed for the analysis of this type of fuel is described in detail. The sheath strain and fission gas pressure are evaluated by this method. The pellet clad gap conductance is calculated by Ross and Solute model. The pellet thermal expansion is calculated by assuming a two zone model, i.e. a plastic core surrounded by an elastic cracked annulus. (author)

  6. 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)

  7. Simulation of the thermomechanical interaction between pellet and cladding and fission gas release

    International Nuclear Information System (INIS)

    Denis, Alicia C.; Soba, Alejandro

    2000-01-01

    This paper summarizes the present status of a computer code that simulates some of the main phenomena occurring in a fuel element of a nuclear power reactor throughout its life. Temperature distribution, thermal expansion, elastic and plastic strains, creep, mechanical interaction between pellet and cladding, fission gas release, swelling and densification are modeled. Thermal expansion gives origin to elastic or plastic strains, which adequately describe the bamboo effect. The code assumes an axial symmetric rod and hence, cylindrical finite elements are employed for the discretization. The fission gas inventory is calculated by means of a diffusion model, which assumes spherical grains and uses also a finite element scheme. Once the temperature distribution in the pellet and the cladding is obtained and in order to reduce the calculation time, the rod is divided into five cylindrical rings where the temperature is averaged. In each ring the gas diffusion problem is solved in one representative grain and the results are then extended to the whole ring. The pressure, increased by the released gas, interacts with the stress field. Densification and swelling due to solid and gaseous fission products are also considered. Experiments, particularly those of the FUMEX series, are simulated with this code. A good agreement is obtained for the fuel center line temperature, the inside rod pressure and the fractional gas release. (author)

  8. 3D FE simulation of PCMI (Pellet-Cladding Mechanical Interaction) considering frictionless contact

    International Nuclear Information System (INIS)

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

    2014-01-01

    The goal of this code is coupling every aspect of physical phenomenon. Monodimensional FE model has been made for METEOR. It is good to evaluate the global behavior in high burn up levels. However, the multi-dimensional PCI analysis code is necessary to precisely analyze the stress distribution especially in case of the crack analysis. CAST3M 3D finite element code has been developed considering thermo-mechanical interaction in detail for TOUTATIS code. The advanced multidimensional code called ALCYONE has been developed considering chemical-physics and thermomechanical aspects. Although there are many codes that analyze pellet and cladding interaction, it is difficult to consider every physical aspect. In this paper, pellet to cladding mechanical interaction in 3D has been simulated with frictionless contact using the developed module, which is written in FORTRANN90. In this paper, 3D PCMI FE model is simulated with frictionless contact and elastic deformation. From the frictionless contact analysis, the interfacial pressure has been calculated and then this is used to obtain the solid heat coefficient which is a main factor to analyze the thermal distribution

  9. 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)

  10. 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%)

  11. 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

  12. 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

  13. Stress concentration during pellet cladding interaction: Comparison of closed-form solutions with 2D(r,θ) finite element simulations

    International Nuclear Information System (INIS)

    Sercombe, Jérôme; Masson, Renaud; Helfer, Thomas

    2013-01-01

    Highlights: • This paper presents closed-formed solutions concerning pellet cladding interaction. • First, the opening of a radial crack in a pellet fragment is estimated. • Second, the stresses in the cladding in front of the pellet crack are calculated. • The closed-formed solutions are found in good agreement with 2D FE simulations. • They are then used in the fuel code ALCYONE to model PCI during power ramps. -- Abstract: This paper presents two closed-form solutions that can be used to enrich the mechanical description of fuel pellets and cladding behavior in standard one-dimensional based fuel performance codes. The first one is concerned with the estimation of the opening of a radial crack in a pellet fragment induced by the radial thermal gradient in the pellet and limited by the pellet-clad contact pressure. The second one describes the stress distribution in a cladding bore in front of an opening pellet crack. A linear angular variation of the pellet-clad contact pressure and a constant prescribed radial displacement are considered. The closed-form solutions are checked by comparison to independent finite element models of the pellet fragment and of the cladding. Their ability to describe non-axisymmetric displacement and stress fields during loading histories representative of base irradiation and power ramps is then demonstrated by cross-comparison with the 2D pellet fragment-cladding model of the multi-dimensional fuel performance code ALCYONE. The calculated radial crack opening profiles at different times and the hoop stress concentration in the cladding at the top of the ramp are found in good agreement with ALCYONE

  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. 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

  16. 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

  17. Simulation of pellet-cladding thermomechanical interaction and fission gas release

    International Nuclear Information System (INIS)

    Denis, A.; Soba, A.

    2001-01-01

    This paper summarizes the present status of a computer code that describes some of the main phenomena occurring in a nuclear fuel element throughout its life. Temperature distribution, thermal expansion, elastic and plastic strains, creep, mechanical interaction between pellet and cladding, fission gas release, swelling and densification are modelized. The code assumes an axi-symmetric rod and hence, cylindrical finite elements are employed for the discretization. Due to the temperature dependence of the thermal conductivity, the heat conduction problem is non-linear. Thermal expansion gives origin to elastic or plastic strains, which adequately describe the bamboo effect. Plasticity renders the stress-strain problem non linear. The fission gas inventory is calculated by means of a diffusion model, which assumes spherical grains and uses a finite element scheme. In order to reduce the calculation time, the rod is divided into five cylindrical rings where the temperature is averaged. In each ring the gas diffusion problem is solved in one grain and the results are then extended to the whole ring. The pressure, increased by the released gas, interacts with the stress field. Densification and swelling due to solid and gaseous fission products are also considered. Experiments, particularly those of the FUMEX series, are simulated with this code. A good agreement is obtained for the fuel center line temperature, the inside rod pressure and the fractional gas release. (author)

  18. 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).

  19. Simulation of pellet-cladding thermomechanical interaction and fission gas release

    International Nuclear Information System (INIS)

    Denis, Alicia; Soba, Alejandro

    2003-01-01

    This paper summarizes the present status of a computer code that describes some of the main phenomena occurring in a nuclear fuel rod throughout its life. Temperature distribution, thermal expansion, elastic and plastic strains, creep, mechanical interaction between pellet and cladding, fission gas release, gas mixing, swelling, and densification are modeled. The modular structure of the code allows for the incorporation of models to simulate different phenomena and material properties. Collapsible rods can be also simulated. The code is bidimensional, assumes cylindrical symmetry for the rod and uses the finite element method to integrate the differential equations. The stress-strain and heat conduction problems are nonlinear due to plasticity and to the temperature dependence of the thermal conductivity. The fission gas inventory is calculated with a diffusion model, assuming spherical grains and using a one-dimensional finite element scheme. Pressure increase, swelling and densification are coupled with the stress field. Good results are obtained for the simulation of the irradiation tests of the first argentine prototypes of MOX fuels, where the bamboo effect is clearly observed, and of the FUMEX series for the fuel centerline temperature, the inside rod pressure and the fractional gas release.

  20. 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)

  1. Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-06-01

    Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO2) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

  2. 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)

  3. A Comparative Physics Study of Commercial PWR Cores using Metallic Micro-cell UO{sub 2}-Cr (or Mo) Pellets with Cr-based Cladding Coating

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hee; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of); In, Wang Kee [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this work, a comparative neutronic analysis of the cores using ATFs which include metallic micro-cell UO{sub 2}-Cr, UO{sub 2}-Mo pellets and Cr-based alloy coating on cladding was performed to show the effects of the ATF fuels on the core performance. In this study, the cores having different ATFs use the same initial uranium enrichments. The ATF concepts studied in this work are the metallic microcell UO{sub 2} pellets containing Cr or Mo with cladding outer coating composed of Cr-based alloy which have been suggested as the ATF concepts in KAERI (Korea Atomic Energy Research Institute). The metallic micro-cell pellets and Cr-based alloy coating can enhance thermal conductivity of fuel and reduce the production of hydrogen from the reaction of cladding with coolant, respectively. The objective of this work is to compare neutronic characteristics of commercial PWR equilibrium cores utilizing the different variations of metallic micro-cell UO{sub 2} pellets with cladding coating composed of Cr-based alloy. The results showed that the cores using UO{sub 2}-Cr and UO{sub 2}-Mo pellets with Cr-based alloy coating on cladding have reduced cycle lengths by 60 and 106 EFPDs, respectively, in comparison with the reference UO{sub 2} fueled core due to the reduced heavy metal inventories and large thermal absorption cross section but they do not have any significant differences in the core performances parameters. However, it is notable that the core fueled the micro-cell UO{sub 2}-Mo pellet and Cr-based alloy coating has considerably more negative MTC and slightly more negative FTC than the other cases. These characteristics of the core using micro-cell UO{sub 2}-Mo pellet and Cr-based alloy coating is due to the hard neutron spectrum and large capture resonance cross section of Mo isotopes.

  4. Design of absorber assemblies with intentional pellet-cladding mechanical interaction

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Birney, K.R.; Pitner, A.L.; Basmajian, J.A.

    1980-04-01

    A number of improvements in absorber assembly performance characteristics can be achieved through implementation of absorber cladding mechanical interaction (ACMI). Benefits include lower operating temperatures, less potential for material relocation, longer lifetime, and increased reactivity worth. Analyses indicate that substantial cladding strains may be attainable without significant risk of breach. However, actual in-reactor testing of ACMI in absorber elements will be required before design criteria can be revised to accept ACMI

  5. Cracking and healing behavior of UO2 as related to pellet-cladding mechanical interaction. Interim report, July 1976

    International Nuclear Information System (INIS)

    Kennedy, C.R.; Yaggee, F.L.; Voglewede, J.C.; Kupperman, D.S.; Wrona, B.J.; Ellingson, W.A.; Johanson, E.; Evans, A.G.

    1976-10-01

    A direct-electrical-heating apparatus has been designed and fabricated to investigate those nuclear-fuel-related phenomena involved in the gap closure-bridging annulus formation mechanism that can be reproduced in an out-of-reactor environment. Prototypic light-water-reactor UO 2 fuel-pellet temperature profiles have been generated utilizing high flow rates (approximately 700 liters/min) of helium coolant gas, and a recirculating system has been fabricated to permit tests of up to 1000 h. Simulated light-water-reactor single- and multiple-thermal-cycle experiments will be conducted on both unclad and ceramic (fused silica) clad UO 2 pellet stacks. A laser dilatometer with a resolution of 1.27 x 10 -2 mm (5 x 10 -4 in.) is used to measure pellet dimensional increase continuously during thermal cycling. Acoustic emissions from thermal-gradient cracking have been detected and correlated with crack length and crack area. The acoustic emissions are monitored continuously to provide instantaneous information about thermal-gradient cracking. Posttest metallography and fracture-mechanics measurements are utilized to characterize cracking and crack healing

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

    International Nuclear Information System (INIS)

    Tas, Fatma Burcu; Ergun, Sule

    2013-01-01

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

  7. 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

  8. 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

  9. 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

  10. Automatic mesh refinement and local multigrid methods for contact problems: application to the Pellet-Cladding mechanical Interaction

    International Nuclear Information System (INIS)

    Liu, Hao

    2016-01-01

    This Ph.D. work takes place within the framework of studies on Pellet-Cladding mechanical Interaction (PCI) which occurs in the fuel rods of pressurized water reactor. This manuscript focuses on automatic mesh refinement to simulate more accurately this phenomena while maintaining acceptable computational time and memory space for industrial calculations. An automatic mesh refinement strategy based on the combination of the Local Defect Correction multigrid method (LDC) with the Zienkiewicz and Zhu a posteriori error estimator is proposed. The estimated error is used to detect the zones to be refined, where the local sub-grids of the LDC method are generated. Several stopping criteria are studied to end the refinement process when the solution is accurate enough or when the refinement does not improve the global solution accuracy anymore. Numerical results for elastic 2D test cases with pressure discontinuity show the efficiency of the proposed strategy. The automatic mesh refinement in case of unilateral contact problems is then considered. The strategy previously introduced can be easily adapted to the multi-body refinement by estimating solution error on each body separately. Post-processing is often necessary to ensure the conformity of the refined areas regarding the contact boundaries. A variety of numerical experiments with elastic contact (with or without friction, with or without an initial gap) confirms the efficiency and adaptability of the proposed strategy. (author) [fr

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. In-reactor cladding breach of EBR-II driver-fuel elements

    International Nuclear Information System (INIS)

    Seidel, B.R.; Einziger, R.E.

    1977-01-01

    Knowledge of performance and minimum useful element lifetime of Mark-II driver-fuel elements is required to maintain a high plant operating capacity factor with maximum fuel utilization. To obtain such knowledge, intentional cladding breach has been obtained in four run-to-cladding-breach Mark-II experimental driver-fuel subassemblies operating under normal conditions in EBR-II. Breach and subsequent fission-product release proved benign to reactor operations. The breaches originated on the outer surface of the cladding in the root of the restrainer dimples and were intergranular. The Weibull distribution of lifetime accurately predicts the observed minimum useful element lifetime of 10 at.% burnup, with breach ensuing shortly thereafter

  18. 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

  19. Biosorption of copper (II) from aqueous solution by mycelial pellets ...

    African Journals Online (AJOL)

    The optimum diameter and pH for biosorption of copper (II) was found to be 1.0 to 1.2 mm and 4.0. Evaluation of the experimental data in terms of biosorption dynamics showed that the biosorption of copper (II) onto MPRO followed the pseudo-second-order kinetic model. The equilibrium data fitted very well to the Langmuir ...

  20. Review of session V of the ANS topical meeting, St. Charles, Il., USA, May 1977: ''Mechanisms for pellet cladding interactions''

    International Nuclear Information System (INIS)

    Wood, J.C.

    1977-07-01

    All seven authors were agreed that power ramping of UO 2 -Zircaloy fuel pins could cause clad defects that were not solely mechanical but of the stress corrosion cracking or liquid metal embrittlement type. Very strong circumstantial evidence for stress corrosion cracking was presented by relating the results of laboratory experiments and theoretical analyses with the behaviour of fuel in-reactor and its physical and chemical characteristics observed during post-irradiation examination. The most likely corrodant species to be responsible for defects are iodine, cadmium or cadmium dissolved in cesium. (author)

  1. 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.

  2. 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

  3. 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)

  4. Fabrication of lithium ceramic pellets, rings and single crystals for irradiation in BEATRIX-II

    International Nuclear Information System (INIS)

    Slagle, O.D.; Noda, K.; Takahashi, T.

    1989-04-01

    BEATRIX-II is an IEA sponsored experiment of lithium ceramic solid breeder materials in the FFTF/MOTA. Li 2 O solid pellets and annular ring specimens were fabricated for in-situ tritium release tests. In addition, a series of single crystal and polycrystalline lithium ceramic samples were fabricated to determine the irradiation behavior and beryllium compatibility. 6 refs., 10 figs., 4 tabs

  5. 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

  6. Oxidation behavior analysis of cladding during severe accidents with combined codes for Qinshan Phase II Nuclear Power Plant

    International Nuclear Information System (INIS)

    Shi, Xingwei; Cao, Xinrong; Liu, Zhengzhi

    2013-01-01

    Highlights: • A new verified oxidation model of cladding has been added in Severe Accident Program (SAP). • A coupled analysis method utilizing RELAP5 and SAP codes has been developed and applied to analyze a SA caused by LBLOCA. • Analysis of cladding oxidation under a SA for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) has been performed by SAP. • Estimation of the production of hydrogen has been achieved by coupled codes. - Abstract: Core behavior at a high temperature is extremely complicated during transition from Design Basic Accident (DBA) to the severe accident (SA) in Light Water Reactors (LWRs). The progression of core damage is strongly affected by the behavior of fuel cladding (oxidation, embrittlement and burst). A Severe Accident Program (SAP) is developed to simulate the process of fuel cladding oxidation, rupture and relocation of core debris based on the oxidation models of cladding, candling of melted material and mechanical slumping of core components. Relying on the thermal–hydraulic boundary parameters calculated by RELAP5 code, analysis of a SA caused by the large break loss-of-coolant accident (LBLOCA) without mitigating measures for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) was performed by SAP for finding the key sequences of accidents, estimating the amount of hydrogen generation and oxidation behavior of the cladding

  7. Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.; Mariani, R.D.

    1999-01-01

    Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U-Zr alloy fuel elements irradiated in the experimental breeder reactor II (EBR-II). We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments (hulls) that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr (and U) found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining. (orig.)

  8. Breached fuel pin contamination from Run Beyond Cladding Breach (RBCB) tests in EBR-II

    International Nuclear Information System (INIS)

    Colburn, R.P.; Strain, R.V.; Lambert, J.D.B.; Ukai, S.; Shibahara, I.

    1988-09-01

    Studies indicate there may be a large economic incentive to permit some continued reactor operation with breached fuel pin cladding. A major concern for this type of operation is the potential spread of contamination in the primary coolant system and its impact on plant maintenance. A study of the release and transport of contamination from naturally breached mixed oxide Liquid Metal Reactor (LMR) fuel pins was performed as part of the US Department of Energy/Power Reactor and Nuclear Fuel Development Corporation (DOE/PNC) Run Beyond Cladding Breach (RBCB) Program at EBR-II. The measurements were made using the Breached Fuel Test Facility (BFTF) at EBR-II with replaceable deposition samplers located approximately 1.5 meters from the breached fuel test assemblies. The effluent from the test assemblies containing the breached fuel pins was routed up through the samplers and past dedicated instrumentation in the BFTF before mixing with the main coolant flow stream. This paper discusses the first three contamination tests in this program. 2 refs., 5 figs., 2 tabs

  9. Cladding axial elongation models for FRAP-T6

    International Nuclear Information System (INIS)

    Shah, V.N.; Carlson, E.R.; Berna, G.A.

    1983-01-01

    This paper presents a description of the cladding axial elongation models developed at the Idaho National Engineering Laboratory (INEL) for use by the FRAP-T6 computer code in analyzing the response of fuel rods during reactor transients in light water reactors (LWR). The FRAP-T6 code contains models (FRACAS-II subcode) that analyze the structural response of a fuel rod including pellet-cladding-mechanical-interaction (PCMI). Recently, four models were incorporated into FRACAS-II to calculate cladding axial deformation: (a) axial PCMI, (b) trapped fuel stack, (c) fuel relocation, and (d) effective fuel thermal expansion. Comparisons of cladding axial elongation measurements from two experiments with the corresponding FRAP-T6 calculations are presented

  10. Target-plasma production by laser irradiation of a pellet in the Baseball II-T experiment

    International Nuclear Information System (INIS)

    Damm, C.C.; Foote, J.H.; Futch, A.H.; Goodman, R.K.; Hornady, R.S.; Osher, J.E.; Porter, G.D.

    1977-01-01

    One way to generate a plasma target that can be used in conjunction with an injected neutral beam to initiate a high-energy plasma in a steady-state, magnetic-mirror field is by the laser irradiation of a solid pellet located within the confinement region. In the Lawrence Livermore Laboratory Baseball II-T experiment, a CO 2 laser was used to provide a two-sided irradiation of an ammonia pellet; the maximum laser intensity on the pellet was approximately 4 x 10 12 W/cm 2 . The 150-μm-dia pellets were guided to the laser focal spot in the Baseball II-T magnetic field using steering voltages controlled by a microcomputer-based system. Diagnostics showed complete ionization of the pellet, average ion energies in the keV range, synchronized triggering of the laser and the neutral beam, and rapid expansion of the plasma to a diameter that was a good match to the diameter of the neutral beam. Predictions obtained from the LASNEX code compared well with measured results. Although the laser-pellet approach was proven usable as a target-plasma startup system, it would be much more complicated and expensive than the method in which streaming plasma is used to trap the neutal beams

  11. ALTERNATIVE BINDERS TO BENTONITE FOR IRON ORE PELLETIZING : PART II : EFFECTS ON METALLURGICAL AND CHEMICAL PROPERTIES

    Directory of Open Access Journals (Sweden)

    Osman Sivrikaya

    2014-07-01

    Full Text Available This study was started to find alternative binders to bentonite and to recover the low preheated and fired pellet mechanical strengths of organic binders-bonded pellets. Bentonite is considered as a chemical impurity for pellet chemistry due to acid constituents (SiO2 and Al2O3. Especially addition of silica-alumina bearing binders is detrimental for iron ore concentrate with high acidic content. Organic binders are the most studied binders since they are free in silica. Although they yield pellets with good wet strength; they have found limited application in industry since they fail to give sufficient physical and mechanical strength to preheated and fired pellets. It is investigated that how insufficient preheated and fired pellet strengths can be improved when organic binders are used as binder. The addition of a slag bonding/strength increasing constituent (free in acidic contents into pellet feed to provide pellet strength with the use of organic binders was proposed. Addition of boron compounds such as colemanite, tincal, borax pentahydrate, boric acid together with organic binders such as CMC, starch, dextrin and some organic based binders, into magnetite and hematite pellet mixture was tested. After determining the addition of boron compounds is beneficial to recover the low pellet physical and mechanical qualities in the first part of this study, in this second part, metallurgical and chemical properties (reducibility - swelling index – microstructure – mineralogy - chemical content of pellets produced with combined binders (an organic binder plus a boron compound were presented. The metallurgical and chemical tests results showed that good quality product pellets can be produced with combined binders when compared with the bentonite-bonded pellets. Hence, the suggested combined binders can be used as binder in place of bentonite in iron ore pelletizing without compromising the pellet chemistry.

  12. On LMFBR corrosion. Part II: Consideration of the in-reactor fuel-cladding system

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Walker, C.T.; Whitlow, W.H.

    1976-05-01

    The scientific and technological aspects of LMFBR cladding corrosion are discussed in detail. Emphasis is placed on the influence of the irradiation environment and the effect of fuel and filler-gas impurities on the corrosion process. These studies are complemented by a concise review of out-of-pile simulation experiments that endeavour to clarify the role of the aggressive fission products cesium, tellurium and iodine. The principal models for cladding corrosion are presented and critically assessed. Areas of uncertainty are exposed and some pertinent experiments are suggested. Consideration is also given to some new observations regarding the role of stress in fuel-cladding reactions and the formation of ferrite in the corrosion zone of the cladding during irradiation. Finally, two technological solutions to the problem of cladding corrosion are proposed. These are based on the use of an oxygen buffer in the fuel and the application of a protective coating to the inner surface of the cladding

  13. Reflood behavior at low initial clad temperature in Slab Core Test Facility Core-II

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Sobajima, Makoto; Abe, Yutaka; Iwamura, Takamichi; Ohnuki, Akira; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Adachi, Hiromichi.

    1990-07-01

    In order to study the reflood behavior with low initial clad temperature, a reflood test was performed using the Slab Core Test Facility (SCTF) with initial clad temperature of 573 K. The test conditions of the test are identical with those of SCTF base case test S2-SH1 (initial clad temperature 1073 K) except the initial clad temperature. Through the comparison of results from these two tests, the following conclusions were obtained. (1) The low initial clad temperature resulted in the low differential pressures through the primary loops due to smaller steam generation in the core. (2) The low initial clad temperature caused the accumulated mass in the core to be increased and the accumulated mass in the downcomer to be decreased in the period of the lower plenum injection with accumulator (before 50s). In the later period of the cold leg injection with LPCI (after 100s), the water accumulation rates in the core and the downcomer were almost the same between both tests. (3) The low initial clad temperature resulted in the increase of the core inlet mass flow rate in the lower plenum injection period. However, the core inlet mass flow rate was almost the same regardless of the initial clad temperature in the later period of the cold leg injection period. (4) The low initial clad temperature resulted in the low turnaround temperature, high temperature rise and fast bottom quench front propagation. (5) In the region apart from the quench front, low initial clad temperature resulted in the lower heat transfer. In the region near the quench front, almost the same heat transfer coefficient was observed between both tests. (6) No flow oscillation with a long period was observed in the SCTF test with low initial clad temperature of 573 K, while it was remarkable in the Cylindrical Core Test Facility (CCTF) test which was performed with the same initial clad temperature. (J.P.N.)

  14. Cladding creepdown under compression

    International Nuclear Information System (INIS)

    Hobson, D.O.

    1977-01-01

    Light-water power reactors use Zircaloy tubing as cladding to contain the UO 2 fuel pellets. In-service operating conditions impose an external hydrostatic force on the cladding, causing it to creep down into eventual contact with the fuel. Knowledge of the rate of such creepdown is of great importance to modelers of fuel element performance. An experimental system was devised for studying creepdown that meets several severe requirements by providing (1) correct stress state, (2) multiple positions for measuring radial displacement of the cladding surface, (3) high-precision data, and (4) an experimental configuration compact enough to fit in-reactor. A microcomputer-controlled, eddy-current monitoring system was developed for this study and has proven highly successful in measuring cladding deformation with time at temperatures of 371 0 C (700 0 F) and higher, and at pressures as high as 21 MPa

  15. Cladding oxidation during air ingress. Part II: Synthesis of modelling results

    International Nuclear Information System (INIS)

    Beuzet, E.; Haurais, F.; Bals, C.; Coindreau, O.; Fernandez-Moguel, L.; Vasiliev, A.; Park, S.

    2016-01-01

    Highlights: • A state-of-the-art for air oxidation modelling in the frame of severe accident is done. • Air oxidation models from main severe accident codes are detailed. • Simulations from main severe accident codes are compared against experimental results. • Perspectives in terms of need for further model development and experiments are given. - Abstract: Air ingress is a potential risk in some low probable situations of severe accidents in a nuclear power plant. Air is a highly oxidizing atmosphere that can lead to an enhanced Zr-based cladding oxidation and core degradation affecting the release of fission products. This is particularly true speaking about ruthenium release, due to its high radiotoxicity and its ability to form highly volatile oxides in a significant manner in presence of air. The oxygen affinity is decreasing from the Zircaloy cladding, fuel and ruthenium inclusions. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues in such scenarios. In the past years, many works have been done on cladding oxidation by air under severe accident conditions. This paper with in addition the paper “Cladding oxidation during air ingress – Part I: Synthesis of experimental results” of this journal issue aim at assessing the state of the art on this phenomenon. In this paper, the modelling of air ingress phenomena in the main severe accident codes (ASTEC, ATHLET-CD, MAAP, MELCOR, RELAP/SCDAPSIM, SOCRAT) is described in details, as well as the validation against the integral experiments QUENCH-10, QUENCH-16 and PARAMETER-SF4. A full review of cladding oxidation by air is thus established.

  16. Fuel-cladding chemical interaction

    International Nuclear Information System (INIS)

    Gueneau, C.; Piron, J.P.; Dumas, J.C.; Bouineau, V.; Iglesias, F.C.; Lewis, B.J.

    2015-01-01

    The chemistry of the nuclear fuel is very complex. Its chemical composition changes with time due to the formation of fission products and depends on the temperature level history within the fuel pellet and the clad during operation. Firstly, in thermal reactors, zircaloy oxidation from reaction with UO 2 fuel under high-temperature conditions will be addressed. Then other fuel-cladding interaction phenomena occurring in fast reactors will be described. Large thermal gradients existing between the centre and the periphery of the pellet induce the radial redistribution of the fuel constituents. The fuel pellet can react with the clad by different corrosion processes which can involve actinide and/or fission product transport via gas, liquid or/and solid phases. All these phenomena are briefly described in the case of different kinds of fuels (oxide, carbide, nitride, metallic) to be used in fast reactors. The way these phenomena are taken into account in fuel performance codes is presented. (authors)

  17. 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.

  18. Structural cladding /clad structures

    DEFF Research Database (Denmark)

    Beim, Anne

    2012-01-01

    Structural Cladding /Clad Structures: Studies in Tectonic Building Practice A. Beim CINARK – Centre for Industrialized Architecture, Institute of Architectural Technology, The Royal Danish Academy of Fine Arts School of Architecture, Copenhagen, Denmark ABSTRACT: With point of departure in the pr......Structural Cladding /Clad Structures: Studies in Tectonic Building Practice A. Beim CINARK – Centre for Industrialized Architecture, Institute of Architectural Technology, The Royal Danish Academy of Fine Arts School of Architecture, Copenhagen, Denmark ABSTRACT: With point of departure...... to analyze, compare, and discuss how these various construction solutions point out strategies for development based on fundamentally different mindsets. The research questions address the following issues: How to learn from traditional construction principles: When do we see limitations of tectonic maneuver......, to ask for more restrictive building codes. As an example, in Denmark there are series of increasing demands in the current building legislations that are focused at enhancing the energy performance of buildings, which consequently foster rigid insulation standards and ask for improvement of air...

  19. Geometrical dimensioning of PWR UO2 pellets

    International Nuclear Information System (INIS)

    Silva, A.T.

    1988-08-01

    The finite element structural program SAP-IV is used to calculate UO 2 pellet strains developed under thermal gradients in pressurized water reactors. The applied procedure allows to analyse the influence of various aspects of pelet geometry on cladding strains and can be utilized for the dimensioning of UO 2 pellets. Pellets purchased with flat ends, with dishes pressed into both ends, shouders, and a 45-deg edge chamfer are analysed. The analyse results are compared with experiemtnal data. (author) [pt

  20. 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.

  1. Mechanical modelling of transient- to- failure SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L. E.

    2014-07-01

    The response of Sodium Fast Reactor (SFR) fuel rods to transient accident conditions is an important safety concern. During transients the cladding strain caused by the stress due to pellet cladding mechanical interaction (PCMI) can lead to failure. Due to the fact that SFR fuel rods are commonly clad with strengthened material made of stainless steel (SS), cladding is usually treated as an elastic-perfectly-plastic material. However, viscoplastic behaviour can contribute to mechanical strain at high temperature (> 1000 K). (Author)

  2. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    Shoup, R.L.

    1976-07-01

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

  3. Geometric dimensioning of UO2 pellets for PWR

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.

    1988-01-01

    The finite element structural program SAP-IV is used to calculate UO 2 pellet strains developed under thermal gradients in pressurized water reactors. The applied procedure allows to analyse the influence of various aspects of pellet geometry on cladding strains and can be utilized for the dimensioning of UO 2 pellets. Pellets purchased with flat ends, with dishes pressed into both ends, shouders, and a 45-deg edge chamfer are analysed. The analyse results are compared with experimental data.(autor) [pt

  4. 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.)

  5. Optimized Gen-II FeCrAl cladding production in large quantity for campaign testing

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sun, Zhiqian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-06-03

    There are two major objectives in this report; (1) to optimize microstructure control of ATF FeCrAl alloys during tube drawing processes, and (2) to provide an update on the progress of ATF FeCrAl tube production via commercial manufacturers. Experimental efforts have been made to optimize the process parameters balancing the tube fabricability, especially for tube drawing processes, and microstructure control of the final tube products. Lab-scale sheet materials of Gen II FeCrAl alloys (Mo-containing and Nb-containing FeCrAl alloys) were used in the study, combined with a stepwise warm-rolling process and intermediate annealing, aiming to simulate the tube drawing process in a commercial tube manufacturer. The intermediate annealing at 650ºC for 1h was suggested for the tube-drawing process of Mo-containing FeCrAl alloys because it successfully softened the material by recovering the work hardening introduced through the rolling step, without inducing grain coarsening due to recrystallization. The final tube product is expected to have stabilized deformed microstructure providing the improved tensile properties with sufficient ductility. Optimization efforts on Nb-containing FeCrAl alloys focused on the effect of alloying additions and annealing conditions on the stability of deformed microstructure. Relationships between the second-phase precipitates (Fe2Nb-Laves phase) and microstructure stability are discussed. FeCrAl tube production through commercial tube manufacturers is currently in progress. Three different manufacturers, Century Tubes, Inc. (CTI), Rhenium Alloys, Inc. (RAI), and Superior Tube Company, Inc. (STC), are providing capabilities for cold-drawing, warm-drawing, and HPTR cold-pilgering, respectively. The first two companies are currently working on large quantity tube production (expected 250 ft length) of Gen I model FeCrAl alloy (B136Y3, at CTI) and Gen II (C35M4, at RAI), with the process parameters obtained from the experimental

  6. Zirconium-barrier cladding attributes

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.; Rand, R.A.; Tucker, R.P.; Cheng, B.; Adamson, R.B.; Davies, J.H.; Armijo, J.S.; Wisner, S.B.

    1987-01-01

    This metallurgical study of Zr-barrier fuel cladding evaluates the importance of three salient attributes: (1) metallurgical bond between the zirconium liner and the Zircaloy substrate, (2) liner thickness (roughly 10% of the total cladding wall), and (3) softness (purity). The effect that each of these attributes has on the pellet-cladding interaction (PCI) resistance of the Zr-barrier fuel was studied by a combination of analytical model calculations and laboratory experiments using an expanding mandrel technique. Each of the attributes is shown to contribute to PCI resistance. The effect of the zirconium liner on fuel behavior during off-normal events in which steam comes in contact with the zirconium surface was studied experimentally. Simulations of loss-of-coolant accident (LOCA) showed that the behavior of Zr-barrier cladding is virtually indistinguishable from that of conventional Zircaloy cladding. If steam contacts the zirconium liner surface through a cladding perforation and the fuel rod is operated under normal power conditions, the zirconium liner is oxidized more rapidly than is Zircaloy, but the oxidation rate returns to the rate of Zircaloy oxidation when the oxide phase reaches the zirconium-Zircaloy metallurgical bond

  7. First results on the effect of fuel-cladding eccentricity

    International Nuclear Information System (INIS)

    Panka, I.; Kereszturi, A.

    2009-01-01

    In the traditional fuel-behaviour or hot channel calculations it is assumed that the fuel pellet is centered within the clad. However, in the real life the pellet could be positioned asymmetrically within the clad, which leads to asymmetric gap conductance and therefore it is worthwhile to investigate the magnitude of the effect on maximal fuel temperature and surface heat flux. In this paper our first experiences are presented on this topic. (Authors)

  8. 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.)

  9. Management of waste cladding hulls. Part II. An assessment of zirconium pyrophoricity and recommendations for handling waste hulls

    International Nuclear Information System (INIS)

    Kullen, B.J.; Levitz, N.M.; Steindler, M.J.

    1977-11-01

    This report reviews experience and research related to the pyrophoricity of zirconium and zirconium alloys. The results of recent investigations of the behavior of Zircaloy and some observations of industrial handling and treatment of Zircaloy tubing and scrap are also discussed. A model for the management of waste Zircaloy cladding hulls from light water reactor fuel reprocessing is offered, based on an evaluation of the reviewed information. It is concluded that waste Zircaloy cladding hulls do not constitute a pyrophoric hazard if, following the model flow sheet, finely divided metal is oxidized during the management procedure. Steps alternative to the model are described which yield zirconium in deactivated form and also accomplish varying degrees of transuranic decontamination. Information collected into appendixes is (1) a collation of zirconium pyrophoricity data from the literature, (2) calculated radioactivity contents in Zircaloy cladding hulls from spent LWR fuels, and (3) results of a laboratory study on volatilization of zirconium from Zircaloy using HCl or Cl 2

  10. 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

  11. Cladding creepdown model for FRAPCON-2

    International Nuclear Information System (INIS)

    Shah, V.N.; Tolli, J.E.

    1985-02-01

    This report presents a cladding deformation model developed to analyze cladding creepdown during steady state operation in both a pressurized water reactor (PWR) and a boiling water reactor (BWR). This model accounts for variations in zircaloy cladding heat treatment; cold worked and stress relieved material, typically used in a PWR, and fully recrystallized material, typically used in a BWR. The model calculates cladding creepdown as a function of hoop stress, fast neutron flux, exposure time, and temperature. This report also presents a comparison between cladding creep calculations by this model and corresponding measurements from the KWU/CE program, ORNL HOBBIE experiments, and EPRI/Westinghouse Engineering cooperative project. The comparisons show that the model calculates cladding creep strains well. The analyses of non-fueled rods by FRAPCON-2 show that the cladding creepdown model was correctly incorporated. Also, analysis of a PWR rod test case shows that the FRAPCON-2 code can analyze pellet-cladding mechanical interaction caused by cladding creepdown and fuel swelling

  12. 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

  13. A Novel Approach for Dry Powder Coating of Pellets with Ethylcellulose. Part II: Evaluation of Caffeine Release.

    Science.gov (United States)

    Albertini, Beatrice; Melegari, Cecilia; Bertoni, Serena; Dolci, Luisa Stella; Passerini, Nadia

    2018-04-01

    The objective of this study was to assess the efficacy and the capability of a novel ethylcellulose-based dry-coating system to obtain prolonged and stable release profiles of caffeine-loaded pellets. Lauric and oleic acids at a suitable proportion were used to plasticize ethylcellulose. The effect of coating level, percentage of drug loading, inert core particle size, and composition of the coating formulation including the anti-sticking agent on the drug release profile were fully investigated. A coating level of 15% w/w was the maximum layered amount which could modify the drug release. The best controlled drug release was obtained by atomizing talc (2.5% w/w) together with the solid plasticizer during the dry powder-coating process. SEM pictures revealed a substantial drug re-crystallization on the pellet surface, and the release studies evidenced that caffeine diffused through the plasticized polymer acting as pore former. Therefore, the phenomenon of caffeine migration across the coating layer had a strong influence on the permeability of the coating membrane. Comparing dry powder-coated pellets to aqueous film-coated ones, drug migration happened during storage, though more sustained release profiles were obtained. The developed dry powder-coating process enabled the production of stable caffeine sustained release pellets. Surprisingly, the release properties of the dry-coated pellets were mainly influenced by the way of addition of talc into the dry powder-coating blend and by the drug nature and affinity to the coating components. It would be interesting to study the efficacy of novel coating system using a different API.

  14. The fuel-cladding interfacial friction coefficient in water-cooled reactor fuel rods

    International Nuclear Information System (INIS)

    Smith, E.

    1979-01-01

    A central problem in the development of cladding failure criteria and of effective operational, design or material remedies is to know whether the cladding stress is enhanced significantly near cladding ridges, pellet chips or fuel pellet cracks; the latter may also be coincident with cladding ridges at pellet-pellet interfaces. As regards the fuel pellet crack source of cladding stress concentration, the magnitude of the uranium dioxide-Zircaloy interfacial friction coefficient μ governs the magnitude and distribution of the enhanced cladding stress. Considerable discussion, particularly at a Post-Conference Seminar associated with the SMIRT 4 Conference, has focussed on the value of μ, the author taking the view that it is unlikely to be large (< 0.5). The reasoning behind this view is as follows. A fuel pellet should fracture during a power ramp when the tensile hoop stress within the pellet exceeds the fuel's fracture stress. Since the preferred position for a fuel pellet crack to form is at the fuel-cladding interface midway between existing fuel cracks, where the interfacial shear stress changes sign, the pellet segment size after a power ramp provides a limit to the magnitude of the interfacial shear stresses and consequently to the value of μ. With this argument as a basis, the author's early work used the Gittus fuel rod model, in which there is a symmetric distribution of fuel pellet cracks and symmetric interfacial slippage, to show that μ < 0.5 if it is assumed that the average hoop stress within the cladding attains yield levels. It was therefore suggested that a high interfacial friction coefficient is unlikely to be operative during a power ramp; this result was used to support the view that interfacial friction effects do not play a dominant role in stress corrosion crack formation within the cladding. (orig.)

  15. Line-Shape Code Comparison through Modeling and Fitting of Experimental Spectra of the C ii 723-nm Line Emitted by the Ablation Cloud of a Carbon Pellet

    Directory of Open Access Journals (Sweden)

    Mohammed Koubiti

    2014-07-01

    Full Text Available Various codes of line-shape modeling are compared to each other through the profile of the C ii 723-nm line for typical plasma conditions encountered in the ablation clouds of carbon pellets, injected in magnetic fusion devices. Calculations were performed for a single electron density of 1017 cm−3 and two plasma temperatures (T = 2 and 4 eV. Ion and electron temperatures were assumed to be equal (Te = Ti = T. The magnetic field, B, was set equal to either to zero or 4 T. Comparisons between the line-shape modeling codes and two experimental spectra of the C ii 723-nm line, measured perpendicularly to the B-field in the Large Helical Device (LHD using linear polarizers, are also discussed.

  16. 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

  17. Nonequilibrium synthesis of NbAl3 and Nb-Al-V alloys by laser cladding. II - Oxidation behavior

    Science.gov (United States)

    Haasch, R. T.; Tewari, S. K.; Sircar, S.; Loxton, C. M.; Mazumder, J.

    1992-01-01

    Isothermal oxidation behaviors of NbAl3 alloy synthesized by laser cladding were investigated at temperatures between 800 and 1400 C, and the effect of vanadium microalloying on the oxidation of the laser-clad alloy was examined. The oxidation kinetics of the two alloys were monitored using thermal gravimetric weight gain data, and the bulk and surface chemistries were analyzed using XRD and XPS, respectively. It was found that NbAl3 did not form an exclusive layer of protective Al2O3. The oxidation products at 800 C were found to be a mixture of Nb2O5 and Al2O3. At 1200 C, a mixture of NbAlO4, Nb2O5, and Al2O3 formed; and at 1400 C, a mixture of NbAlO4, Al2O3, NbO2, NbO(2.432), and Nb2O5 formed. The addition of V led to a dramatic increase of the oxidation rate, which may be related to the formation of (Nb, V)2O5 and VO2, which grows in preference to protective Al2O3.

  18. Experimental assessment of fuel-cladding interactions

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-29

    A range of fuel concepts designed to better tolerate accident scenarios and reactor transients are currently undergoing fundamental development at national laboratories as well as university and industrial partners. Pellet-clad mechanical and chemical interaction can be expected to affect fuel failure rates experienced during steady state operation, as well as dramatically impact the response of the fuel form under loss of coolant and other accident scenarios. The importance of this aspect of fuel design prompted research initiated by AFC in FY14 to begin exploratory efforts to characterize this phenomenon for candidate fuelcladding systems of immediate interest. Continued efforts in FY15 and FY17 aimed to better understand and simulate initial pellet-clad interaction with little-to-no pressure on the pellet-clad interface. Reported here are the results from 1000 h heat treatments at 400, 500, and 600°C of diffusion couples pairing UN with a FeCrAl alloy, SiC, and Zr-based cladding candidate sealed in evacuated quartz ampoules. No gross reactions were observed, though trace elemental contaminants were identified.

  19. Critical cladding radius for hybrid cladding modes

    Science.gov (United States)

    Guyard, Romain; Leduc, Dominique; Lupi, Cyril; Lecieux, Yann

    2018-05-01

    In this article we explore some properties of the cladding modes guided by a step-index optical fiber. We show that the hybrid modes can be grouped by pairs and that it exists a critical cladding radius for which the modes of a pair share the same electromagnetic structure. We propose a robust method to determine the critical cladding radius and use it to perform a statistical study on the influence of the characteristics of the fiber on the critical cladding radius. Finally we show the importance of the critical cladding radius with respect to the coupling coefficient between the core mode and the cladding modes inside a long period grating.

  20. 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

  1. Fuel cladding behavior under rapid loading conditions

    Science.gov (United States)

    Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

    2016-02-01

    A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

  2. 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

  3. 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)

  4. 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.)

  5. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    Shoup, R.L.

    1976-01-01

    The composite of metals and alloys used in the fabrication of 238 Pu cardiac pacemaker fuel capsules resists the effects of high temperatures, high mechanical forces, and chemical corrosives and provides more than adequate protection to the fuel pellet even from deliberate attempts to dissolve the cladding in inorganic acids. This does not imply that opening a pacemaker fuel capsule by inorganic acids is impossible but that it would not be a wise choice

  6. 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.

  7. Development of optimized advanced austenitic steels (II). Evaluation of out-of-pile testing results of the test fuel claddings

    Energy Technology Data Exchange (ETDEWEB)

    Uwaba, Tomoyuki; Mizuta, Shunji; Ukai, Shigeharu [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2000-03-01

    14Cr-25Ni optimized advanced austenitic steels have been developed to improve the swelling resistance of 15Cr-20Ni austenitic stainless steels used for FBR fuel cladding. In this improvement, Ti, Nb, V and P were dissolved into 14Cr-25Ni matrix by means of the high-temperature solution treatment to make finely distributed and stabilized precipitates in the operation. Furthermore, at the final stage of cold-working, cold-working level increased and residual stress was reduced. In this study, as fabricated microstructure observation, solubility of alloying elements and grain size test in the manufacturing process were evaluated. Following results were obtained. (1) Spherical precipitates were observed in the grain. Most of them were identified as complexed carbide-nitride [Ti,Nb(C,N)] by EDX analysis. (2) The dissolved percentages of Ti and Ni in the matrix were about 70% and 30% respectively. Undissolved Ti and Nb may react with undissolved carbon and precipitate as MC carbides. (3) High-temperature solution treatment is effective for the sufficient solubility of alloying elements, but it is likely to induce very large grains, which is the cause of defective signal in the ultrasonic alloy testing. The results of the grain size test showed that the large grain size is reduced in low Nb (0.1wt%) alloy compared with the standard alloy (0.2wt% Nb), and the effectiveness for the grain size control by reducing the Nb content was confirmed. Also, it was suggested that the intermediate heat treatment and cold work conditions would possibly avoid the occurrence of the large grain at the final heat treatment. (author)

  8. Asymptotic Method for Cladding Stress Evaluation in PCMI

    International Nuclear Information System (INIS)

    Kim, Hyungkyu; Kim, Jaeyong; Yoon, Kyungho; Lee, Kanghee; Kang, Heungseok

    2014-01-01

    A PCMI (Pellet Cladding Mechanical Interaction) failure was first reported in the GETR (General Electric Test Reactor) at Vacellitos in 1963, and such failures are still occurring. Since the high stress values in the cladding tube has been of a crucial concern in PCMI studies, there have been many researches on the stress analysis of a cladding tube pressed by a pellet. Typical works can be found in some references. It has often been assumed, however, that the cracks in the pellet were equally spaced and the pellet was a rigid body. In addition, the friction coefficient was arbitrarily chosen so that a slipping between the pellets and cladding tube could not be logically defined. Moreover, the stress intensification due to the sharp edge of a pellet fragment has never been realistically considered. These problems above drove us to launch a framework of a PCMI study particularly on stress analysis technology to improve the present analysis method incorporating the actual PCMI conditions such as the stress intensification, arbitrary distribution of the pellet cracks, material properties (esp. pellet) and slipping behavior of the pellet/cladding interface. As a first step of this work, this paper introduces an asymptotic method that was originally developed for a stress analysis in the vicinity of a sharp notch of a homogeneous body. The intrinsic reason for applying this method is to simulate the stress singularity that is expected to take place at the sharp edge of a pellet fragment due to cracking during irradiation. As a first attempt of this work, an eigenvalue problem is formulated in the case of adhered contact, and the generalized stress intensity factors are defined and evaluated. Although some works obviously remain to be accomplished, for the present framework on the PCMI analysis (e. g., slipping behaviour, contact force etc.), it was addressed that the asymptotic method can produce the stress values that cause the cladding tube failure in PCMI more

  9. Potential for cladding thermal failure in LWRs during high temperature transients

    International Nuclear Information System (INIS)

    El Genk, M.S.

    1979-01-01

    The temperature increase in the fuel and the cladding during a PCM accident produces film boiling at the cladding surface which may induce zircaloy cladding failure, due to embrittlement, and fuel melting at the centerline of the fuel pellets. Molten fuel may extrude through radial cracks in the fuel and relocate in the fuel-cladding gap. Contact of extruded molten fuel with the cladding, which is at high temperature during film boiling, may induce cladding thermal failure due to melting. An assessment of central fuel melting and molten fuel extrusion into the fuel-cladding gap during a PCM accident is presented. The potential for thermal failure of the zircaloy cladding upon being contacted by molten fuel during such an accident is also analyzed and compared with the applicable experimental evidence

  10. Estimation of penetration depth of fission products in cladding Hull

    International Nuclear Information System (INIS)

    Kim, Hee Moon; Jung, Yang Hong; Yoo, Byong Ok; Choo, Yong Sun; Hong, Kwon Pyo

    2005-01-01

    A disposal and a reprocessing for spent fuel rod with high burnup need de-cladding procedure. Pellet in this rod has been separated from a cladding hull to reduce a radioactivity of hull by chemical and mechanical methods. But fission products and actinides(U,Pu) still remain inside of cladding hull by chemical bonding and fission spike, which is called as 'contamination'. More specific removal of this contamination would have been considered. In this study, the sorts of fission products and penetration depth in hull were observed by EPMA test. To analyze this behavior, SRIM 2000 code was also used as energies of fission products and an oxide thickness of hull

  11. 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.)

  12. Assessment of thin-walled cladding tube mechanical properties by segmented expanding Mandrel test

    International Nuclear Information System (INIS)

    Nilsson, Karl-Fredrik

    2015-01-01

    This paper presents the principles of the segmented expanding mandrel test for thin-walled cladding tubes, which can be used as a basic material characterisation test to determine stress-strain curves and ductility or as a test to simulate mechanical pellet-cladding interaction. The paper discusses the strengths and weaknesses of the test method and it illustrates how the test can be used to simulate hydride reorientations in zirconium claddings and quantify how hydride reorientation affects ductility. (authors)

  13. Test system to simulate transient overpower LMFBR cladding failure

    International Nuclear Information System (INIS)

    Barrus, H.G.; Feigenbutz, L.V.

    1981-01-01

    One of the HEDL programs has the objective to experimentally characterize fuel pin cladding failure due to cladding rupture or ripping. A new test system has been developed which simulates a transient mechanically-loaded fuel pin failure. In this new system the mechanical load is prototypic of a fuel pellet rapidly expanding against the cladding due to various causes such as fuel thermal expansion, fuel melting, and fuel swelling. This new test system is called the Fuel Cladding Mechanical Interaction Mandrel Loading Test (FCMI/MLT). The FCMI/MLT test system and the method used to rupture cladding specimens very rapidly to simulate a transient event are described. Also described is the automatic data acquisition and control system which is required to control the startup, operation and shutdown of the very fast tests, and needed to acquire and store large quantities of data in a short time

  14. 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

  15. 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...

  16. 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

  17. Fuel cladding mechanical interaction during power ramps

    International Nuclear Information System (INIS)

    Guerin, Y.

    1985-01-01

    Mechanical interaction between fuel and cladding may occur as a consequence of two types of phenomenon: i) fuel swelling especially at levels of caesium accumulation, and ii) thermal differential expansion during power changes. Slow overpower ramps which may occur during incidental events are of course one of the circumstances responsible for this second type of fuel cladding mechanical interaction (FCMI). Experiments and analysis of this problem that have been done at C.E.A. allow to determine the main parameters which will fix the level of stress and the risk of damage induced by the fuel in the cladding during overpower transients

  18. 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)

  19. 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

  20. 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.)

  1. 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

  2. An advanced cold moderator using solid methane pellets

    International Nuclear Information System (INIS)

    Foster, C.A.; Carpenter, J.M.

    2001-01-01

    This paper reports developments of the pellet formation and transport technologies required for producing a liquid helium or hydrogen cooled methane pellet moderator. The Phase I US DOE SBIR project, already completed, demonstrated the production of 3 mm transparent pellets of frozen methane and ammonia and transport of the pellets into a 40 cc observation cell cooled with liquid helium. The methane pellets, formed at 72 K, stuck together during the loading of the cell. Ammonia pellets did not stick and fell readily under vibration into a packed bed with a 60% fill fraction. A 60% fill fraction should produce a very significant increase in long-wavelength neutron production and advantages in shorter pulse widths as compared to a liquid hydrogen moderator. The work also demonstrated a method of rapidly changing the pellets in the moderator cell. The Phase II project, just now underway, will develop a full-scale pellet source and transport system with a 1.5 L 'moderator' cell. The Phase II effort will also produce an apparatus to sub-cool the methane pellets to below 20 K, which should prevent the methane pellets from sticking together. In addition to results of the phase I experiments, the presentation includes a short video of the pellets, and a description of plans for the Phase II project. (author)

  3. 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

  4. 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)

  5. 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

  6. 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.)

  7. 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

  8. 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.

  9. 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

  10. Gap conductance in Zircaloy-clad LWR fuel rods

    International Nuclear Information System (INIS)

    Ainscough, J.B.

    1982-04-01

    This report describes the procedures currently used to calculate fuel-cladding gap conductance in light water reactor fuel rods containing pelleted UO 2 in Zircaloy cladding, under both steady-state and transient conditions. The relevant theory is discussed together with some of the approximations usually made in performance modelling codes. The state of the physical property data which are needed for heat transfer calculations is examined and some of the relevant in- and out-of-reactor experimental work on fuel rod conductance is reviewed

  11. Laser cladding with powder

    NARCIS (Netherlands)

    Schneider, M.F.; Schneider, Marcel Fredrik

    1998-01-01

    This thesis is directed to laser cladding with powder and a CO2 laser as heat source. The laser beam intensity profile turned out to be an important pa6 Summary rameter in laser cladding. A numerical model was developed that allows the prediction of the surface temperature distribution that is

  12. 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

  13. Preliminary study of mechanical behavior for Cr coated Zr-4 Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do-Hyoung; Kim, Hak-Sung [Hanyang Univ., Seoul (Korea, Republic of); Kim, Hyo-Chan; Yang, Yong-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    To decrease the oxidation rate of Zr-based alloy components, many concepts of accident tolerant fuel (ATF) such as Mo-Zr cladding, SiC/SiCf cladding and iron-based alloy cladding are under development. One of the promised concept is the coated cladding which can remarkably increase the corrosion and wear resistance. Recently, KAERI is developing the Cr coated Zircaloy cladding as accident tolerance cladding. To coat the Cr powder on the Zircaloy, 3D laser coating technology has been employed because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. Therefore, for this work, the mechanical integrity of Cr coated Zircaloy should be evaluated to predict the safety of fuel cladding during the operating or accident of nuclear reactor. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr, which were referred from the literatures and experimental reports. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr. The pellet-clad mechanical interaction (PCMI) properties of Cr coated Zr-4 cladding were investigated by thermo-mechanical finite element analysis (FEA) simulation. The mechanical properties of Zr-4 and Cr was validated by simulation of ring compression test (RCT) of fuel cladding.

  14. 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

  15. 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)

  16. 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.

  17. 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)

  18. 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

  19. 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

  20. Comparison of cryogenic (hydrogen) and TESPEL (polystyrene) pellet particle deposition in a magnetically confined plasma

    Science.gov (United States)

    McCarthy, K. J.; Tamura, N.; Combs, S. K.; Panadero, N.; Ascabíbar, E.; Estrada, T.; García, R.; Hernández Sánchez, J.; López Fraguas, A.; Navarro, M.; Pastor, I.; Soleto, A.; TJ-II Team

    2017-10-01

    A cryogenic pellet injector (PI) and tracer encapsulated solid pellet (TESPEL) injector system has been operated in combination on the stellarator TJ-II. This unique arrangement has been created by piggy-backing a TESPEL injector onto the backend of a pipe-gun-type PI. The combined injector provides a powerful new tool for comparing ablation and penetration of polystyrene TESPEL pellets and solid hydrogen pellets, as well as for contrasting subsequent pellet particle deposition and plasma perturbation under analogous plasma conditions. For instance, a significantly larger increase in plasma line-averaged electron density, and electron content, is observed after a TESPEL pellet injection compared with an equivalent cryogenic pellet injection. Moreover, for these injections from the low-magnetic-field side of the plasma cross-section, TESPEL pellets deposit electrons deeper into the plasma core than cryogenic pellets. Finally, the physics behind these observations and possible implications for pellet injection studies are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  2. Development Status of Accident Tolerant Fuel Cladding for LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Yang, Jae-Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Hydrogen explosions and the release of radionuclides are caused by severe damage of current nuclear fuels, which are composed of fuel pellets and fuel cladding, during an accident. To reduce the damage to the public, the fuels have to enhance their integrity under an accident environment. Enhanced accident tolerance fuels (ATFs) can tolerate a loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations as well as operational transients, in comparison with the current UO{sub 2}-Zr alloy system used in the LWR. Surface modified Zr cladding as a new concept was suggested to apply an enhanced ATF cladding. The aim of the partial ODS treatment is to increase the high-temperature strength to suppress the ballooning/rupture behavior of fuel cladding during an accident event. The target of the surface coating is to increase the corrosion resistance during normal operation and increase the oxidation resistance during an accident event. The partial ODS treatment of Zircaloy-4 cladding can be produced using a laser beam scanning method with Y2O3 powder, and the surface Cr-alloy and Cr/FeCrAl coating on Zircaloy-4 cladding can be obtained after the development of 3D laser coating and arc ion plating technologies.

  3. A model for hydrogen pickup for BWR cladding materials

    International Nuclear Information System (INIS)

    Hede, G.; Kaiser, U.

    2001-01-01

    It has been observed that rod elongation is driven by the hydrogen pickup but not by corrosion as such. Based on this a non-destructive method to determine clad hydrogen concentration has been developed. The method is based on the observation that there are three different mechanisms behind the rod growth: the effect of neutron irradiation on the Zircaloy microstructure, the volume increase of the cladding as an effect of hydride precipitation and axial pellet-cladding-mechanical-interaction (PCMI). The derived correlation is based on the experience of older cladding materials, inspected at hot-cell laboratories, that obtained high hydrogen levels (above 500 ppm) at lower burnup (assembly burnup below 50 MWd/kgU). Now this experience can be applied, by interpolation, on more modern cladding materials with a burnup beyond 50 MWd/kgU by analysis of the rod growth database of the respective cladding materials. Hence, the method enables an interpolation rather than an extrapolation of present day hydrogen pickup database, which improves the reliability and accuracy. Further, one can get a good estimate of the hydrogen pickup during an ongoing outage based on a non-destructive method. Finally, rod growth measurements are normally performed for a large population of rods, hence giving a good statistics compared to examination of a few rods at a hot cell. (author)

  4. Development of H2 pellet injectors for industrial marketing

    International Nuclear Information System (INIS)

    Visler, T.

    1988-09-01

    1. Discussion of the construction of injector installation at ETA-BETA II. 2. Production and experience with two different ''pipe-guns''. One for large pellets, diameter/length = 4.5-5 mm/8-20 mm and one for small pellets, diameter/length = 2 mm/3-4 mm. (author) 27 ills., 39 refs

  5. Cladding failure by local plastic instability

    International Nuclear Information System (INIS)

    Kramer, J.M.; Deitrich, L.W.

    1977-01-01

    Cladding failure is one of the major considerations in analysis of fast-reactor fuel pin behavior during hypothetical accident transients since time, location and nature of failure govern the early post-failure material motion and reactivity feedback. Out-of-Pile transient burst tests of both irradiated and unirradiated fast-reactor cladding show that local plastic instability, or bulging, often precedes rupture. To investigate the details of cladding bulging, a perturbation analysis of the equations governing the large deformation of a cylindrical shell has been developed. The overall deformation history is assumed to consist of a small perturbation epsilon of the radial displacement superimposed on large axisymmetric displacements. Computations have been carried out using high temperature properties of stainless steel in conjunction with various constitutive theories, including a generalization of the Endochronic Theory of Plasticity in which both time-independent and time-dependent plastic strains are modeled. Although the results of the calculations are all qualitatively similar, it appears that modeling of both time-independent and time-dependent plastic strains is necessary to interpret the transient burst test results. Sources for bulge formation that have been considered include initial geometric imperfections and thermal perturbations due to either eccentric fuel pellets or non-symmetric cooling. (Auth.)

  6. Stone cladding engineering

    National Research Council Canada - National Science Library

    Camposinhos, Rui de Sousa

    2014-01-01

    .... Straightforward formulae are provided for computing action on cladding, with special emphasis on the effect of seismic forces, including an extensive general methodology applied to non-structural elements...

  7. Eddy-current testing and analysis of a sample of Zircaloy fuel cladding for the OECD Halden 'Round-Robin' exercise (Phase II)

    International Nuclear Information System (INIS)

    Watson, P.C.; Cross, M.T.

    1987-02-01

    Two samples of Zircaloy fuel cladding were supplied, one containing pre-measured defects of known type and size, and the other containing unknown defects. Eddy-current testing techniques were used to ascertain the nature of the unknown defects. By using a high resolution encircling coil and a probe coil and then processing digitally the data with specially prepared software, nine internal defects, of volume 0.18 to 0.86 mm 3 were located positively and identified, despite interference from heavily fluctuating background signals. (author)

  8. 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...

  9. 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

  10. In-reactor measurement of clad strain: effect of power history

    International Nuclear Information System (INIS)

    Fehrenbach, P.J.; Morel, P.A.

    1980-01-01

    A series of experimental irradiations has been undertaken at CRNL to measure directly the in-reactor deformation of fuel elements while they are operating at power. Power histories have been chosen to allow investigation of power, time at power and burnup on pellet-clad interaction for element linear powers to 60kW/m. Results are presented which indicate that irradiation of a fresh fuel element at high power is effective in minimizing clad hoop stresses during subsequent ramps or cycles to that power. The effectiveness of this preconditioning appears to be due primarily to fuel densification rather than stress relaxation in the clad. (auth)

  11. Critical stability conditions of the fuel element cladding; Kriticni uslovi stabilnosti kosuljice G.E

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovic, M; Savic, D [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1968-12-15

    The role of the fuel element cladding being the first safety barrier, is to prevent contamination by the fission products. Construction of the fuel element cladding depends on the reactor type, coolant type, fuel type, technology of material fabrication, influence of the material on the neutron economy, thermal conditions, etc. That is why an optimum solution has to be found. This paper deals with mechanical properties of ceramic natural UO{sub 2} sintered fuel pellets in the zircaloy-2 cladding. This type of fuel is used in heavy water reactors.

  12. Method for the protection of the cladding tubes of fuel rods

    International Nuclear Information System (INIS)

    Steinberg, E.

    1978-01-01

    To present stress crack corrosion and to protect the cladding tubes of the fuel rods made of a circonium alloy from attack by iodine, the inward surfaces are provided with protective coatings. Therefore the casting tubes already filled with fuel element pellets are put under over-pressure at a temperature range between 300 and 500 0 C, until almost yield-point is reached. A small amount of H 2 O or H 2 O 2 , filled in, reacts with the cladding tube material to form the Zr-O 2 protective coating. Afterwards comes a pressure relief, and the cladding tube reaches its original dimensions. (DG) [de

  13. 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

  14. 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)

  15. Laser cladding of turbine blades

    International Nuclear Information System (INIS)

    Shepeleva, L.; Medres, B.; Kaplan, W.D.; Bamberger, M.

    2000-01-01

    A comparative study of two different techniques for the application of wear-resistant coatings for contact surfaces of shroud shelves of gas turbine engine blades (GTE) has been conducted. Wear-resistant coatings were applied on In713 by laser cladding with direct injection of the cladding powder into the melt pool. Laser cladding was conducted with a TRUMPF-2500, CW-CO 2 laser. The laser cladding was compared with commercially available plasma cladding with wire. Both plasma and laser cladded zones were characterized by optical and scanning electron microscopy. It was found that the laser cladded zone has a higher microhardness value (650-820 HV) compared with that of the plasma treated material (420-440 HV). This is a result of the significant reduction in grain size in the case of laser cladding. Unlike the plasma cladded zones, the laser treated material is free of micropores and microcracks. (orig.)

  16. Potential for fuel melting and cladding thermal failure during a PCM event in LWRs

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Croucher, D.W.

    1979-01-01

    The primary concern in nuclear reactor safety is to ensure that no conceivable accident, whether initiated by a failure of the reactor system or by incorrect operation, will lead to a dangerous release of radiation to the environment. A number of hypothesized off-normal power or cooling conditions, generally termed as power-cooling-mismatch (PCM) accidents, are considered in the safety analysis of light water reactors (LWRs). During a PCM accident, film boiling may occur at the cladding surface and cause a rapid temperature increase in the fuel and the cladding, perhaps producing embrittlement of the zircaloy cladding by oxidation. Molten fuel may be produced at the center of the pellets, extrude radially through open cracks in the outer, unmelted portion of the pellet and relocate in the fuel-cladding gap. If the amount of extruded molten fuel is sufficient to establish contact with the cladding, which is at a high temperature during film boiling, the zircaloy cladding may melt. The present work assesses the potential for central fuel melting and thermal failure of the zircaloy cladding due to melting upon being contacted by extruded molten UO 2 -fuel during a PCM event

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

    International Nuclear Information System (INIS)

    Koehler, L.

    1976-01-01

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

  18. 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.

  19. Initial Cladding Condition

    International Nuclear Information System (INIS)

    Siegmann, E.

    2000-01-01

    The purpose of this analysis is to describe the condition of commercial Zircaloy clad fuel as it is received at the Yucca Mountain Project (YMP) site. Most commercial nuclear fuel is encased in Zircaloy cladding. This analysis is developed to describe cladding degradation from the expected failure modes. This includes reactor operation impacts including incipient failures, potential degradation after reactor operation during spent fuel storage in pool and dry storage and impacts due to transportation. Degradation modes include cladding creep, and delayed hydride cracking during dry storage and transportation. Mechanical stresses from fuel handling and transportation vibrations are also included. This Analysis and Model Report (AMR) does not address any potential damage to assemblies that might occur at the YMP surface facilities. Ranges and uncertainties have been defined. This analysis will be the initial boundary condition for the analysis of cladding degradation inside the repository. In accordance with AP-2.13Q, ''Technical Product Development Planning'', a work plan (CRWMS M andO 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis will be used in

  20. Fabrication of ThO2, UO2, and PuO2-UO2 pellets

    International Nuclear Information System (INIS)

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

    1978-01-01

    Fabrication of ThO pellets for EBR-II irradiation testing and fabrication of UO 2 and PuO 2 -UO 2 pellets for United Kingdom Prototype Fast Reactor (PFR) irradiation testing is discussed. Effect of process parameters on density and microstructure of pellets fabricated by the cold press and sinter technique is reviewed

  1. Diffusion in cladding materials

    International Nuclear Information System (INIS)

    Anand, M.S.; Pande, B.M.; Agarwala, R.P.

    1992-01-01

    Aluminium has been used as a cladding material in most research reactors because its low neutron absorption cross section and ease of fabrication. However, it is not suitable for cladding in power reactors and as such zircaloy-2 is normally used as a clad because it can withstand high temperature. It has low neutron absorption cross section, good oxidation, corrosion, creep properties and possesses good mechanical strength. With the passage of time, further development in this branch of science took place and designers started looking for better neutron economy and less hydrogen pickup in PHW reactors. The motion of fission products in the cladding material could pose a problem after long operation. In order to understand their behaviour under reactor environment, it is essential to study first the diffusion under normal conditions. These studies will throw light on the interaction of defects with impurities which would in turn help in understanding the mechanism of diffusion. In this article, it is intended to discuss the diffusion behaviour of impurities in cladding materials.(i.e. aluminium, zircaloy-2, zirconium-niobium alloy etc.). (author). 94 refs., 4 figs., 3 tabs

  2. Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ito, Masahiro; Maeda, Koji [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan)

    2011-09-30

    Highlights: > We evaluated diametral strain of fast reactor MOX fuel pins irradiated to 130 GWd/t. > The strain was due to cladding void swelling and irradiation creep. > The irradiation creep was caused by internal gas pressure and PCMI. > The PCMI was associated with pellet swelling by rim structure or by cesium uranate. > The latter effect tended to increase the cumulative damage fraction of the cladding. - Abstract: The C3M irradiation test, which was conducted in the experimental fast reactor, 'Joyo', demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, 'Monju'. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and {sup 137}Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

  3. 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.

  4. Electra-Clad

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-05-04

    The study relates to the use of building-integrated photovoltaics. The Electra-Clad project sought to use steel-based cladding as a substrate for direct fabrication of a fully integrated solar panel of a design similar to the ICP standard glass-based panel. The five interrelated phases of the project are described. The study successfully demonstrated that the principles of the panel design are achievable and sound. But, despite intensive trials, a commercially realistic solar performance has not been achieved: the main failing was the poor solar conversion efficiency as the active area of the panel was increased in size. The problem lies with the coating used on the steel cladding substrates and it was concluded that a new type of coating will be required. ICP Solar Technologies UK carried out the work under contract to the DTI.

  5. Nuclear fuel cladding material

    International Nuclear Information System (INIS)

    Nakahigashi, Shigeo.

    1982-01-01

    Purpose: To largely improve the durability and the safety of fuel cladding material. Constitution: Diffusion preventive layers, e.g., aluminum or the like are covered on both sides of a zirconium alloy base layer of thin material, and corrosion resistant layers, e.g., copper or the like are covered thereon. This thin plate material is intimately wound in a circularly tubular shape in a plurality of layers to form a fuel cladding tube. With such construction, corrosion of the tube due to fuel and impurity can be prevented by the corrosion resistant layers, and the diffusion of the corrosion resistant material to the zirconium alloy can be prevented by the diffusion preventive layers. Since a plurality of layers are cladded, even if the corrosion resistant layers are damaged or cracked due to stress corrosion, only one layer is damaged or cracked, but the other layers are not affected. (Sekiya, K.)

  6. 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.

  7. 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.

  8. 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

  9. Mechanical resistance of UO{sub 2} pellet by means of free-fall-impact testing

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Tae-sik; Lee, Seung-jae; Kim, Jae-ik; Jo, Young-ho; Park, Bo-yong; Ko, Sang-ern [KEPCO NF, Daejeon (Korea, Republic of)

    2014-10-15

    A fuel rod failed during a power transient can be seen in Fig 1. and conjunction of a chipped pellet with a cladding crack has been observed in commercial reactors through the post-irradiation examinations. It revealed that missing-pellet-surface(MPS) was one of the reasons of the fuel failure. The mechanism of this failure mode that MPS induces the asymmetry of the pellet-cladding mechanical system mainly comprises a stress concentration at the inner surface resulting in non-classical PCI. The fracture toughness is largely close to material property. It is assumed that by optimizing surface design of UO{sub 2} pellet, the strength arises because theoretical strength is considerably affected by geometry as one of a parameter of factor 'f'. Pellet research for design optimization to achieve better resistance to external load should be accompanied with volumetric approach to the improvement of mechanical behavior of pellet being still ongoing. At this work, the resistance to external load is analyzed varying with the geometry of pellets and angles of impact on UO{sub 2} pellet surface by the free-fall-impact test method. The tested specimens were equivalently produced and sintered for having the same volumetric property such as sinter density and grain size expect the surface with different geometry design at the end face and shoulder which includes dish, chamfer and land in dimension and angle. Missing-pellet-surface(MPS) on UO{sub 2} pellet is inevitable behavior during manufacturing, handling and burning in reactor and brings about non-classical PCI behavior that could damage fuel rod integrity. For this reason, the free-fall-drop tester was developed by KEPCO NF Material Development laboratory in Daejeon for quantitatively investigating the mechanical behavior of UO{sub 2}. The free-fall-impact test is performed by dropping hammer on pellet shoulder with certain impact energy and at various angles. The result is quantitatively measured with weighing

  10. 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)

  11. 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

  12. 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...

  13. 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)

  14. 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

  15. 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.

  16. 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 (...

  17. 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

  18. 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.)

  19. 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

  20. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The reference fuel for Integral Fast Reactor (IFR) is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel

  1. 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

    under glove box conditions. Pellets of different geometry, from simple cylindrical to chamfered, dished and annular pellets have been fabricated and irradiated in research reactors although plain cylindrical pellets with L/D less than 1.2 have been used for MOX fuel loading in power reactors. Fully automated wet centreless grinding of MOX pellets using composite diamond wheel and subsequent ultrasonic cleaning has been used in the fabrication flowsheet. The MOX pellets undergo vacuum degassing at 400 deg. C to ensure low hydrogen content prior to loading of pellets into zircaloy clad fuel tubes. A novel sol-gel microsphere pelletisation route (SGMP) combined with LTS has also been developed and is briefly discussed. (author)

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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)

  7. The JET high frequency pellet injector project

    International Nuclear Information System (INIS)

    Geraud, Alain; Dentan, M.; Whitehead, A.; Butcher, P.; Communal, D.; Faisse, F.; Gedney, J.; Gros, G.; Guillaume, D.; Hackett, L.; Hennion, V.; Homfray, D.; Lucock, R.; McKivitt, J.; Sibbald, M.; Portafaix, C.; Perin, J.P.; Reade, M.; Sands, D.; Saille, A.

    2007-01-01

    A new deuterium ice pellet injector is in preparation for JET. It is designed to inject both small pellets (variable volume within 1-2 mm 3 ) at high frequency (up to 60 Hz) for ELM mitigation experiments and large pellets (volume within 35-70 mm 3 ) at moderate frequency (up to 15 Hz) for plasma fuelling. It is based on the screw extruder technology developed by PELIN and pneumatic acceleration. An injection line will connect the injector to the flight tubes already in place to convey the pellets toward the plasma either from the low field side or from the high field side of the torus. This injection line enables: (i) the pumping of the propellant gas, (ii) the provision of the vacuum interface with the torus and (iii) the selection of the flight tube to be used via a fast selector. All the interfaces have been designed and a prototype injector is being built, to demonstrate that the required performance is achievable

  8. 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)

  9. Ceramic Coatings for Clad (The C3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sickafus, Kurt E. [Univ. of Tennessee, Knoxville, TN (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Miller, Larry [Univ. of Tennessee, Knoxville, TN (United States); Weber, Bill [Univ. of Tennessee, Knoxville, TN (United States); Zhang, Yanwen [Univ. of Tennessee, Knoxville, TN (United States); Patel, Maulik [Univ. of Tennessee, Knoxville, TN (United States); Motta, Arthur [Pennsylvania State Univ., University Park, PA (United States); Wolfe, Doug [Pennsylvania State Univ., University Park, PA (United States); Fratoni, Max [Univ. of California, Berkeley, CA (United States); Raj, Rishi [Univ. of Colorado, Boulder, CO (United States); Plunkett, Kenneth [Univ. of Colorado, Boulder, CO (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Hollis, Kendall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Andy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Chris [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Comstock, Robert [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Partezana, Jonna [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Whittle, Karl [Univ. of Sheffield (United Kingdom); Preuss, Michael [Univ. of Manchester (United Kingdom); Withers, Philip [Univ. of Manchester (United Kingdom); Wilkinson, Angus [Univ. of Oxford (United Kingdom); Donnelly, Stephen [Univ. of Huddersfield (United Kingdom); Riley, Daniel [Australian Nuclear Science and Technology Organisation, Syndney (Australia)

    2017-02-14

    The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectives of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as

  10. Advances in appendage joining techniques for PHWR fuel cladding

    International Nuclear Information System (INIS)

    Desai, P.B.; Ray, T.K.; Date, V.G.; Purushotham, D.S.C.

    1995-01-01

    This paper describes work carried out at the BARC on the development of a technique to join tiny appendages (spacers and bearing pads) to thin cladding (before loading of UO 2 pellets) by resistance welding for PHWR fuel assemblies. The work includes qualifying the process for production environment, designing prototype equipment for regular production and quality monitoring. In the first phase of development, welding of appendages on UO 2 loaded elements was successfully developed, and is being used in production. Welding of appendages on to empty clad tubes is a superior technique for several reasons. Many problems associated with development of welding on empty tubes were resolved. work was initiated, in the second phase of the development task, to select a suitable technique to join appendages on empty clad tubes without any collapse of thin clad. Several alternatives were reviewed and assessed such as laser, full face welding, shim welding and shrink fitting ring spacers. Selection of a method using a mandrel and a modified electrode geometry was fully developed. Results were optimized and process development successfully completed. Appropriate weld monitoring techniques were also reviewed for their adaptation. This technique is useful for 19, 22 as well as 37 element assemblies. (author)

  11. 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

  12. 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

  13. Interactions of Zircaloy cladding with gallium: 1998 midyear status

    International Nuclear Information System (INIS)

    Wilson, D.F.; DiStefano, J.R.; Strizak, J.P.; King, J.F.; Manneschmidt, E.T.

    1998-06-01

    A program has been implemented to evaluate the effect of gallium in mixed-oxide (MOX) fuel derived from weapons-grade (WG) plutonium on Zircaloy cladding performance. The objective is to demonstrate that low levels of gallium will not compromise the performance of the MOX fuel system in a light-water reactor. The graded, four-phase experimental program was designed to evaluate the performance of prototypic Zircaloy cladding materials against (1) liquid gallium (Phase 1), (2) various concentrations of Ga 2 O 3 (Phase 2), (3) centrally heated surrogate fuel pellets with expected levels of gallium (Phase 3), and (4) centrally heated prototypic MOX fuel pellets (Phase 4). This status report describes the results of a series of tests for Phases 1 and 2. Three types of tests are being performed: (1) corrosion, (2) liquid metal embrittlement, and (3) corrosion-mechanical. These tests will determine corrosion mechanisms, thresholds for temperature and concentration of gallium that may delineate behavioral regimes, and changes in the mechanical properties of Zircaloy. Initial results have generally been favorable for the use of WG-MOX fuel. The MOX fuel cladding, Zircaloy, does react with gallium to form intermetallic compounds at ≥300 C; however, this reaction is limited by the mass of gallium and is therefore not expected to be significant with a low level (parts per million) of gallium in the MOX fuel. Although continued migration of gallium into the initially formed intermetallic compound can result in large stresses that may lead to distortion, this was shown to be extremely unlikely because of the low mass of gallium or gallium oxide present and expected clad temperatures below 400 C. Furthermore, no evidence for grain boundary penetration by gallium has been observed

  14. Interactions of zircaloy cladding with gallium -- 1997 status

    International Nuclear Information System (INIS)

    Wilson, D.F.; DiStefano, J.R.; King, J.F.; Manneschmidt, E.T.; Strizak, J.P.

    1997-11-01

    A four phase program has been implemented to evaluate the effect of gallium in mixed oxide (MOX) fuel derived from weapons grade (WG) plutonium on Zircaloy cladding performance. The objective is to demonstrate that low levels of gallium will not compromise the performance of the MOX fuel system in LWR. This graded, four phase experimental program will evaluate the performance of prototypic Zircaloy cladding materials against: (1) liquid gallium (Phase 1), (2) various concentrations of Ga 2 O 3 (Phase 2), (3) centrally heated surrogate fuel pellets with expected levels of gallium (Phase 3), and (4) centrally heated prototypic MOX fuel pellets (Phase 4). This status report describes the results of an initial series of tests for phases 1 and 2. Three types of tests are being performed: (1) corrosion, (2) liquid metal embrittlement (LME), and (3) corrosion mechanical. These tests are designed to determine the corrosion mechanisms, thresholds for temperature and concentration of gallium that may delineate behavioral regimes, and changes in mechanical properties of Zircaloy. Initial results have generally been favorable for the use of WG-MOX fuel. The MOX fuel cladding, Zircaloy, does react with gallium to form intermetallic compounds at ≥ 300 C; however, this reaction is limited by the mass of gallium and is therefore not expected to be significant with a low level (in parts per million) of gallium in the MOX fuel. While continued migration of gallium into the initially formed intermetallic compound results in large stresses that can lead to distortion, this is also highly unlikely because of the low mass of gallium or gallium oxide present and expected clad temperatures below 400 C. Furthermore, no evidence for grain boundary penetration by gallium has been observed

  15. Reuse potential of low-calcium bottom ash as aggregate through pelletization.

    Science.gov (United States)

    Geetha, S; Ramamurthy, K

    2010-01-01

    Coal combustion residues which include fly ash, bottom ash and boiler slag is one of the major pollutants as these residues require large land area for their disposal. Among these residues, utilization of bottom ash in the construction industry is very low. This paper explains the use of bottom ash through pelletization. Raw bottom ash could not be pelletized as such due to its coarseness. Though pulverized bottom ash could be pelletized, the pelletization efficiency was low, and the aggregates were too weak to withstand the handling stresses. To improve the pelletization efficiency, different clay and cementitious binders were used with bottom ash. The influence of different factors and their interaction effects were studied on the duration of pelletization process and the pelletization efficiency through fractional factorial design. Addition of binders facilitated conversion of low-calcium bottom ash into aggregates. To achieve maximum pelletization efficiency, the binder content and moisture requirements vary with type of binder. Addition of Ca(OH)(2) improved the (i) pelletization efficiency, (ii) reduced the duration of pelletization process from an average of 14-7 min, and (iii) reduced the binder dosage for a given pelletization efficiency. For aggregate with clay binders and cementitious binder, Ca(OH)(2) and binder dosage have significant effect in reducing the duration of pelletization process. 2010 Elsevier Ltd. All rights reserved.

  16. 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.

  17. An internal conical mandrel technique for fracture toughness measurements on nuclear fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sainte Catherine, C.; Le Boulch, D.; Carassou, S. [CEA Saclay, DEN/DMN, Bldg 625 P, Gif-Sur-Yvette, F-91191 (France); Lemaignan, C. [CEA Grenoble, 17 rue des Martyrs, Grenoble, F-38054 (France); Ramasubramanian, N. [ECCATEC Inc., 92 Deburn Drive, Toronto, Ontario (Canada)

    2006-07-01

    An understanding of the limiting stress level for crack initiation and propagation in a fuel cladding material is a fundamental requirement for the development of water reactor clad materials. Conventional tests, in use to evaluate fracture properties, are of limited help, because they are adapted from ASTM standards designed for thick materials, which differ significantly from fuel cladding geometry (small diameter thin-walled tubing). The Internal Conical Mandrel (ICM) test described here is designed to simulate the effect of fuel pellet diametrical increase on a cladding with an existing axial through-wall crack. It consists in forcing a cone, having a tapered increase in diameter, inside the Zircaloy cladding with an initial axial crack. The aim of this work is to quantify the crack initiation and propagation criteria for fuel cladding material. The crack propagation is monitored by a video system for obtaining crack extension {delta}a. A finite-element (FE) simulation of the ICM test is performed in order to derive J integrals. A node release technique is applied during the FE simulation for crack propagation and the J-resistance curves (J-{delta}a) are generated. This paper presents the test methodology, the J computation validation, and results for cold-worked stress relieved Zircaloy-4 cladding at 20 deg. and 300 deg. C and also for Al 7050-T7651 aluminum alloy tubing at 20 deg. C. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  19. 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)

  20. 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

  1. 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

  2. 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

  3. Analysis of fuel cladding chemical interaction in mixed oxide fuel pins

    International Nuclear Information System (INIS)

    Weber, J.W.; Dutt, D.S.

    1976-01-01

    An analysis is presented of the observed interaction between mixed oxide 75 wt percent UO 2 --25 wt percent PuO 2 fuel and 316--20 percent CW stainless steel cladding in LMFBR type fuel pins irradiated in EBR-II. A description is given of the test pins and their operating conditions together with, metallographic observations and measurements of the fuel/cladding reaction, and a correlation equation is developed relating depth of cladding attack to temperature and burnup. Some recent data on cladding reaction in fuel pins with low initial O/M in the fuel are given and compared with the correlation equation curves

  4. Results of the Gallium-Clad Phase 3 and Phase 4 tasks (canceled prior to completion)

    International Nuclear Information System (INIS)

    Morris, R.N.

    1998-08-01

    This report summarizes the results of the Gallium-Clad interactions Phase 3 and 4 tasks. Both tasks were to involve examining the out-of-pile stability of residual gallium in short fuel rods with an imposed thermal gradient. The thermal environment was to be created by an electrical heater in the center of the fuel rod and coolant flow on the rod outer cladding. Both tasks were canceled due to difficulties with fuel pellet fabrication, delays in the preparation of the test apparatus, and changes in the Fissile Materials Disposition program budget

  5. Observations of in-reactor endurance and rupture life for fueled and unfueled FTR cladding

    International Nuclear Information System (INIS)

    Lovell, A.J.; Christensen, B.Y.; Chin, B.A.

    1979-01-01

    Reactor component endurance limits are important to nuclear experimenters and operators. This paper investigates endurance limits of 316 CW fuel pin cladding. The objective of this paper is to compare and analyze two different sets of FTR fuel pin cladding data. The first data set is from unfueled pressurized cladding irradiated in the Experimental Breeder Reactor No. II (EBR-II). This data set was generated in an assembly in which the temperature was monitored and controlled. The second data set contains observations of breached and unbreached EBR-II test fuel pins covering a large range of temperature, power and burnup conditions

  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. Cladding tube manufacturing technology

    International Nuclear Information System (INIS)

    Hahn, R.; Jeong, Y. H.; Baek, B. J.; Kim, K. H.; Kim, S. J.; Choi, B. K.; Kim, J. M.

    1999-04-01

    This report gives an overview of the manufacturing routine of PWR cladding tubes. The routine essentially consists of a series of deformation and annealing processes which are necessary to transform the ingot geometry to tube dimensions. By changing shape, microstructure and structure-related properties are altered simultaneously. First, a short overview of the basics of that part of deformation geometry is given which is related to tube reducing operations. Then those processes of the manufacturing routine which change the microstructure are depicted, and the influence of certain process parameters on microstructure and material properties are shown. The influence of the resulting microstructure on material properties is not discussed in detail, since it is described in my previous report A lloy Development for High Burnup Cladding . Because of their paramount importance still up to now, and because manufacturing data and their influence on properties for other alloys are not so well established or published, the descriptions are mostly related to Zry4 tube manufacturing, and are only in short for other alloys. (author). 9 refs., 46 figs

  8. Cladding tube manufacturing technology

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, B.J.; Kim, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    This report gives an overview of the manufacturing routine of PWR cladding tubes. The routine essentially consists of a series of deformation and annealing processes which are necessary to transform the ingot geometry to tube dimensions. By changing shape, microstructure and structure-related properties are altered simultaneously. First, a short overview of the basics of that part of deformation geometry is given which is related to tube reducing operations. Then those processes of the manufacturing routine which change the microstructure are depicted, and the influence of certain process parameters on microstructure and material properties are shown. The influence of the resulting microstructure on material properties is not discussed in detail, since it is described in my previous report 'Alloy Development for High Burnup Cladding.' Because of their paramount importance still up to now, and because manufacturing data and their influence on properties for other alloys are not so well established or published, the descriptions are mostly related to Zry4 tube manufacturing, and are only in short for other alloys. (author). 9 refs., 46 figs.

  9. 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

  10. Performance of HT9 clad metallic fuel at high temperature

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Hayes, S.L.

    1992-01-01

    Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area

  11. CASTI handbook of cladding technology. 2. ed.

    International Nuclear Information System (INIS)

    Smith, L.; Celant, M.

    2000-01-01

    This updated (2000) CASTI handbook covers all aspects of clad products - the different means of manufacture, properties and applications in various industries. Topics include: an introduction to cladding technology, clad plate, clad pipes, bends, clad fittings, specification requirements of clad products, welding clad products, clad product application and case histories from around the world. Unique to this book is the documentation of case histories of major cladding projects from around the world and how the technology of that day has withstood the demands of time. Filled with over 100 photos and graphics illustrating the various cladding technology examples and products, this book truly documents the most recent technologies in the field of cladding technology used worldwide

  12. 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...

  13. 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

  14. 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)

  15. 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

  16. Stone cladding engineering

    CERN Document Server

    Sousa Camposinhos, Rui de

    2014-01-01

    This volume presents new methodologies for the design of dimension stone based on the concepts of structural design while preserving the excellence of stonemasonry practice in façade engineering. Straightforward formulae are provided for computing action on cladding, with special emphasis on the effect of seismic forces, including an extensive general methodology applied to non-structural elements. Based on the Load and Resistance Factor Design Format (LRDF), minimum slab thickness formulae are presented that take into consideration stress concentrations analysis based on the Finite Element Method (FEM) for the most commonly used modern anchorage systems. Calculation examples allow designers to solve several anchorage engineering problems in a detailed and objective manner, underlining the key parameters. The design of the anchorage metal parts, either in stainless steel or aluminum, is also presented.

  17. Modeling Dynamic Fracture of Cryogenic Pellets

    Energy Technology Data Exchange (ETDEWEB)

    Parks, Paul [General Atomics, San Diego, CA (United States)

    2016-06-30

    This work is part of an investigation with the long-range objective of predicting the size distribution function and velocity dispersion of shattered pellet fragments after a large cryogenic pellet impacts a solid surface at high velocity. The study is vitally important for the shattered pellet injection (SPI) technique, one of the leading technologies being implemented at ORNL for the mitigation of disruption damage on current tokamaks and ITER. The report contains three parts that are somewhat interwoven. In Part I we formulated a self-similar model for the expansion dynamics and velocity dispersion of the debris cloud following pellet impact against a thick (rigid) target plate. Also presented in Part I is an analytical fracture model that predicts the nominal or mean size of the fragments in the debris cloud and agrees well with known SPI data. The aim of Part II is to gain an understanding of the pellet fracturing process when a pellet is shattered inside a miter tube with a sharp bend. Because miter tubes have a thin stainless steel (SS) wall a permanent deformation (dishing) of the wall is produced at the site of the impact. A review of the literature indicates that most projectile impact on thin plates are those for which the target is deformed and the projectile is perfectly rigid. Such impacts result in “projectile embedding” where the projectile speed is reduced to zero during the interaction so that all the kinetic energy (KE) of the projectile goes into the energy stored in plastic deformation. Much of the literature deals with perforation of the target. The problem here is quite different; the softer pellet easily undergoes complete material failure causing only a small transfer of KE to stored energy of wall deformation. For the real miter tube, we derived a strain energy function for the wall deflection using a non-linear (plastic) stress-strain relation for 304 SS. Using a dishing profile identical to the linear Kirchkoff-Love profile (for lack

  18. 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

  19. 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

  20. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  1. Linear resonance acceleration of pellets

    International Nuclear Information System (INIS)

    Mills, R.G.

    1978-01-01

    A possible requirement for the acceleration of macroscopic pellets to velocities exceeding 10 4 meters per second implies the development of new apparatus. A satisfactory approach might be the linear resonance accelerator. Such apparatus would require the charging of pellets to very high values not yet demonstrated. The incompatibility of phase stability with radial stability in these machines may require abandoning phase stability and adopting feedback control of the accelerating voltage to accommodate statistical fluctuations in the charge to mass ratio of successive pellets

  2. Pellet injection and toroidal confinement

    International Nuclear Information System (INIS)

    1989-12-01

    The proceedings of a technical committee meeting on pellet injection and toroidal confinement, held in Gut Ising, Federal Republic of Germany, 24-26 October, 1988, are given in this report. Most of the major fusion experiments are using pellet injectors; these were reported at this meeting. Studies of confinement, which is favorably affected, impurity transport, radiative energy losses, and affects on the ion temperature gradient instability were given. Studies of pellet ablation and effects on plasma profiles were presented. Finally, several papers described present and proposed injection guns. Refs, figs and tabs

  3. Reduction in degree of absorber-cladding mechanical interaction by shroud tube in control rods for the fast reactor

    International Nuclear Information System (INIS)

    Donomae, Takako; Katsuyama, Kozo; Tachi, Yoshiaki; Maeda, Koji; Yamamoto, Masaya; Soga, Tomonori

    2011-01-01

    Research and development of a long-life control rod for fast reactors is being conducted at Joyo. One of the challenges in developing a long-life control rod is the restraint of absorber-cladding mechanical interaction (ACMI). First, a helium-bonding rod was selected as a control rod for the experimental fast reactor Joyo, which is the first liquid metal fast reactor in Japan. Its lifetime was limited by ACMI, which is induced by the swelling and relocation of B 4 C pellets. To restrain ACMI, a shroud tube was inserted into the gap between the B 4 C pellets and the cladding tube. However, once B 4 C pellets cracked and broke into small fragments, relocation occurred. After this, the narrow gap closed immediately as the degree of B 4 C pellet swelling increased. To solve this problem, the gap was widened during design, and sodium was selected as the bonding material instead of helium to restrain the increase in pellet temperature. Irradiation testing of the modified sodium-bonding control rod confirmed that ACMI would be restrained by the shroud tube regardless of the occurrence of B 4 C pellet relocation. As a result of these improvements, the estimated lifetime of the control rod at Joyo was doubled. In this paper, the results of postirradiation examination are reported. (author)

  4. Deformation and collapse of zircaloy fuel rod cladding into plenum axial gaps

    International Nuclear Information System (INIS)

    Pfennigwerth, P.L.; Gorscak, D.A.; Selsley, I.A.

    1983-01-01

    To minimize support structure, blanket and reflector fuel rods of the thoria urania-fueled Light Water Breeder Reactor (LWBR) were designed with non-freestanding Zircaloy-4 cladding. An analytical model was developed to predict deformation of unirradiated cladding into axial gaps of fuel rod plenum regions where it is unsupported. This model uses the ACCEPT finite element computer program to calculate elastic-plastic deformation of cladding due to external pressure. The finite element is 20-node, triquadratic, isoparametric, and 3-dimensional. Its curved surface permits accurate modeling of the tube geometry, including geometric nonuniformities such as circumferential wall thickness variation and initial tube out-of-roundness. Progressive increases in axial gap length due to cladding elongation and fuel stack shrinkage are modeled, as are deformations of fuel pellets and stainless steel support sleeves which bound plenum axial gaps in LWBR type blanket fuel rods. Zircaloy-4 primary and secondary thermal creep representations were developed from uniaxial creep testing of fuel rod tubing. Creep response to multi-axial loading is modeled with a variation of Hill's formulation for anisotropic materials. Coefficients accounting for anisotropic thermal creep in Zircaloy-4 tubes were developed from creep testing of externally pressurized tubes having fixed axial gaps in the range 2.5 cm to 5.7 cm and radial clearances over simulated fuel pellets ranging from zero to 0.089 mm. (orig./RW)

  5. Production of floating pellets using appropriate methods | Suleiman ...

    African Journals Online (AJOL)

    The study investigated into the use of floating materials like candle wax, yeast and baking powder to achieve pellet buoyancy. Ten diets were formulated with incorporation of floating agents; Diet I-YBCT- (yeast-baking powder in cold water -toasted), Diet II-YBCU- (yeast-baking powder in cold water -untoasted) Diet III ...

  6. Handling of Deuterium Pellets for Plasma Refuelling

    DEFF Research Database (Denmark)

    Jensen, Peter Bjødstrup; Andersen, Verner

    1982-01-01

    The use of a guide tube technique to inject pellets in pellet-plasma experiments is described. The effect of the guide tube on the mass and speed of a slowly moving pellet ( nu approximately 150 m s-1) is negligible. To improve the divergence in trajectories of the pellets on leaving the guide tube...

  7. Pneumatic pellet injector for JT-60

    International Nuclear Information System (INIS)

    Onozuka, Masanori; Hiratsuka, Hajime; Kawasaki, Kouzo.

    1990-01-01

    The pneumatic 4-shot pellet injector has been installed and operated for JT-60 (JAERI Tokamak-60). The performance tests have proven that the device provides high speed pellets as planned. The maximum pellet velocity obtained in the hydrogen pellet tests is greater than 2.3km/s at 100 bar propellant gas. (author)

  8. Pneumatic pellet injector for JT-60

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Masanori (Mitsubishi Heavy Industries Ltd., Tokyo (Japan)); Hiratsuka, Hajime; Kawasaki, Kouzo

    1990-11-01

    The pneumatic 4-shot pellet injector has been installed and operated for JT-60 (JAERI Tokamak-60). The performance tests have proven that the device provides high speed pellets as planned. The maximum pellet velocity obtained in the hydrogen pellet tests is greater than 2.3km/s at 100 bar propellant gas. (author).

  9. Centrifuge pellet injector for JET

    International Nuclear Information System (INIS)

    Andelfinger, C.; Buchelt, E.; Jacobi, D.; Lackner, E.; Schilling, H.B.; Ulrich, M.; Weber, G.

    1983-08-01

    An engineering design of a centrifuge pellet injector for JET is reported as part of the Phase I contract number JE 2/9016. A rather detailed design is presented for the mechanical and electronic features. Stress calculations, dynamic behaviour and life estimates are considered. The interfaces to the JET vacuum system and CODAS are discussed. Proposals for the pellet diagnostics (velocity, mass and shape) are presented. (orig.)

  10. Physics of inertial confinement pellets

    International Nuclear Information System (INIS)

    Mead, W.C.

    1979-01-01

    An overview of inertial confinement fusion pellet physics is given. A discussion is presented of current estimated ICF driver requirements and a couple of pellet examples. The physics of driver/plasma coupling for two drivers which are being considered, namely a laser driver and a heavy ion accelerator driver, is described. Progress towards inertial confinement fusion that has been made using laser drivers in target experiments to date is discussed

  11. Post-irradiation examinations on the KNK II/1 fuel element NY-203 with 400 equivalent full-power days residence time and 10 % burnup

    International Nuclear Information System (INIS)

    Patzer, G.; Geier, F.

    1984-09-01

    The fuel assembly NY-203 has been irradiated in the first core of KNK II up to a burnup of about 10 % and a residence time of 400 equivalent full-power days. The assembly contained 211 fuel pins with 6.0 mm outer diameter and fuel pellets with the composition (U 0 .7Pu 0 .3)O 2 .00. The cladding material was the austenitic steel 1.4988 lg. Some selected pins were examined in the hot cells of the KfK Karlsruhe. The post-irradiation examinations did not reveal any critical design aspects [de

  12. 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)

  13. 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

  14. Fuel-cladding chemical interaction in mixed-oxide fuels

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Weber, J.W.; Devary, J.L.

    1978-10-01

    The character and extent of fuel-cladding chemical interaction (FCCI) was established for UO 2 -25 wt% PuO 2 clad with 20% cold worked Type 316 stainless steel irradiated at high cladding temperatures to peak burnups greater than 8 atom %. The data base consists of 153 data sets from fuel pins irradiated in EBR-II with peak burnups to 9.5 atom %, local cladding inner surface temperatures to 725 0 C, and exposure times to 415 equivalent full power days. As-fabricated oxygen-to-metal ratios (O/M) ranged from 1.938 to 1.984 with the bulk of the data in the range 1.96 to 1.98. HEDL P-15 pins provided data at low heat rates, approx. 200 W/cm, and P-23 series pins provided data at higher heat rates, approx. 400 W/cm. A design practice for breeder reactors is to consider an initial reduction of 50 microns in cladding thickness to compensate for possible FCCI. This approach was considered to be a conservative approximation in the absence of a comprehensive design correlation for extent of interaction. This work provides to the designer a statistically based correlation for depth of FCCI which reflects the influences of the major fuel and operating parameters on FCCI

  15. 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

  16. 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.)

  17. 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

  18. 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.

  19. Multi-shot type pellet injection device

    International Nuclear Information System (INIS)

    Onozuka, Masaki; Uchikawa, Takashi; Kuribayashi, Shitomi.

    1988-01-01

    Purpose: To inject pellets at high speed without melting or sublimating not-injected pellets even at a long pellet injection interval. Constitution: In the conventional multi-shot pellet injection device, the pellet injection interval is set depending on the plasma retention time. However, as the pellet injection interval is increased, not-injected pellets are melted or sublimated due to the introduced heat of acceleration gases supplied from an acceleration gas introduction pipe to give an effect on the dimensional shape of the pellets. In view of the above, a plurality of pellet forming and injection portions each comprising a carrier, an injection pipe and a holder are disposed independently of each other and pellets are formed and injected independently to thereby prevent the thermal effects of the acceleration gases. (Kamimura, M.)

  20. Multi-shot type pellet injection device

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Masaki; Uchikawa, Takashi; Kuribayashi, Shitomi.

    1988-07-27

    Purpose: To inject pellets at high speed without melting or sublimating not-injected pellets even at a long pellet injection interval. Constitution: In the conventional multi-shot pellet injection device, the pellet injection interval is set depending on the plasma retention time. However, as the pellet injection interval is increased, not-injected pellets are melted or sublimated due to the introduced heat of acceleration gases supplied from an acceleration gas introduction pipe to give an effect on the dimensional shape of the pellets. In view of the above, a plurality of pellet forming and injection portions each comprising a carrier, an injection pipe and a holder are disposed independently of each other and pellets are formed and injected independently to thereby prevent the thermal effects of the acceleration gases. (Kamimura, M.).

  1. 46 CFR 148.04-21 - Coconut meal pellets (also known as copra pellets).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Coconut meal pellets (also known as copra pellets). 148.04-21 Section 148.04-21 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) DANGEROUS... § 148.04-21 Coconut meal pellets (also known as copra pellets). (a) Coconut meal pellets; (1) Must...

  2. Gallium-cladding compatibility testing plan: Phase 3: Test plan for centrally heated surrogate rodlet test. Revision 2

    International Nuclear Information System (INIS)

    Morris, R.N.; Baldwin, C.A.; Wilson, D.F.

    1998-07-01

    The Fissile Materials Disposition Program (FMDP) is investigating the use of weapons grade plutonium in mixed oxide (MOX) fuel for light-water reactors (LWR). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons derived fuel may differ from the previous commercial fuels because of small amounts of gallium impurities. A concern presently exists that the gallium may migrate out of the fuel, react with and weaken the clad, and thereby promote loss of fuel pin integrity. Phases 1 and 2 of the gallium task are presently underway to investigate the types of reactions that occur between gallium and clad materials. This is a Level-2 document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-Oxide Fuel Irradiation Test Project Plan. This Plan summarizes the projected Phase 3 Gallium-Cladding compatibility heating test and the follow-on post test examination (PTE). This work will be performed using centrally-heated surrogate pellets, to avoid unnecessary complexities and costs associated with working with plutonium and an irradiation environment. Two sets of rodlets containing pellets prepared by two different methods will be heated. Both sets will have an initial bulk gallium content of approximately 10 ppm. The major emphasis of the PTE task will be to examine the material interactions, particularly indications of gallium transport from the pellets to the clad

  3. A New Material Constitutive Model for Predicting Cladding Failure

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Joe; Dunham, Robert [ANATECH Corp., San Diego, CA (United States); Rashid, Mark [University of California Davis, Davis, CA (United States); Machiels, Albert [EPRI, Palo Alto, CA (United States)

    2009-06-15

    An important issue in fuel performance and safety evaluations is the characterization of the effects of hydrides on cladding mechanical response and failure behavior. The hydride structure formed during power operation transforms the cladding into a complex multi-material composite, with through-thickness concentration profile that causes cladding ductility to vary by more than an order of magnitude between ID and OD. However, current practice of mechanical property testing treats the cladding as a homogeneous material characterized by a single stress-strain curve, regardless of its hydride morphology. Consequently, as irradiation conditions and hydrides evolution change, new material property testing is required, which results in a state of continuous need for valid material property data. A recently developed constitutive model, treats the cladding as a multi-material composite in which the metal and the hydride platelets are treated as separate material phases with their own elastic-plastic and fracture properties and interacting at their interfaces with appropriate constraint conditions between them to ensure strain and stress compatibility. An essential feature of the model is a multi-phase damage formulation that models the complex interaction between the hydride phases and the metal matrix and the coupled effect of radial and circumferential hydrides on cladding stress-strain response. This gives the model the capability of directly predicting cladding failure progression during the loading event and, as such, provides a unique tool for constructing failure criteria analytically where none could be developed by conventional material testing. Implementation of the model in a fuel behavior code provides the capability to predict in-reactor operational failures due to PCI or missing pellet surfaces (MPS) without having to rely on failure criteria. Even, a stronger motivation for use of the model is in the transportation accidents analysis of spent fuel

  4. 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

  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. Clad Treatment in KARMA Code and Library

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong-yeup; Lee, Hae-chan; Woo, Hae-seuk [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of)

    2016-05-15

    Zirconium is the main components in clad materials. The subgroup parameters of zirconium were generated with effective cross section which obtained by using flux distribution in clad region. It decreases absorption reaction rate differences with reference MCNP results. Use of composite nuclide is acceptable to increase efficiency but should be limited to specific target composition. Therefore, the use of the composite nuclide of Zircaloy-2 should be limited when HANA clad material is used for clad. Either using explicit components or generating composite nuclide for HANA is suggested. This paper investigates the clad analysis model for KARMA whether current method is applicable to HANA clad material.

  7. 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

  8. 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

  9. 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

  10. Clad Degradation - FEPs Screening Arguments

    International Nuclear Information System (INIS)

    E. Siegmann

    2004-01-01

    The purpose of this report is to document the screening of the cladding degradation features, events, and processes (FEPs) for commercial spent nuclear fuel (CSNF). This report also addresses the effect of some FEPs on both the cladding and the CSNF, DSNF, and HLW waste forms where it was considered appropriate to address the effects on both materials together. This report summarizes the work of others to screen clad degradation FEPs in a manner consistent with, and used in, the Total System Performance Assessment-License Application (TSPA-LA). This document was prepared according to ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA'' (BSC 2004a [DIRS 167796])

  11. Pin clad strains in Phenix

    International Nuclear Information System (INIS)

    Languille, A.

    1979-07-01

    The Phenix reactor has operated for 4 years in a satisfactory manner. The first 2 sub-assembly loadings contained pins clad in solution treated 316. The principal pin strains are: diametral strain (swelling and irradiation creep), ovality and spiral bending of the pin (interaction of wire and pin cluster and wrapper). A pin cluster irradiated to a dose of 80 dpa F reached a pin diameter strain of 5%. This strain is principally due to swelling (low fission gas pressure). The principal parameters governing the swelling are instantaneous dose, time and temperature for a given type of pin cladding. Other types of steel are or will be irradiated in Phenix. In particular, cold-worked titanium stabilised 316 steel should contribute towards a reduction in the pin clad strains and increase the target burn-up in this reactor. (author)

  12. 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

  13. Design Report for a 19-pin carbide test-bundle in a ring-subassembly of the test zone of KNK II/2

    International Nuclear Information System (INIS)

    Haefner, H.E.

    1982-03-01

    This report describes a 19-rod carbide test bundle in an annular oxide ring element placed at the position 201 of the test zone in the second core of KNK II as well as its behavior during the period of operation. The selected fuel rod concept includes low pellet density and a relatively large gap width as well as helium bonding between fuel and cladding. Characteristic design and operation data are: rod diameter 8.5 mm, pellet diameter 7.0 mm, maximum nominal linear rating 800 W/cm, maximum nominal burnup 70 MWd/kgHM. This report exclusively deals with the carbide test bundle and its individual components; it describes methods, criteria and results concerning the design. The annular carrier element with its head and foot is treated in a separate report. The loadability of the test bundle and its individual components is demonstrated by generally valid standards for strength criteria [de

  14. Interaction between thorium and potential clad materials

    International Nuclear Information System (INIS)

    Kale, G.B.; Gawde, P.S.; Sengupta, Pranesh

    2005-01-01

    Thorium based fuels are being used for nuclear reactors. The structural stability of fuel-clad assemblies in reactor systems depend upon the nature of interdiffusion reaction between fuel-cladding materials. Interdiffusion reaction thorium and various cladding materials is presented in this paper. (author)

  15. Study on characteristics of spent PWR cladding hull for categorizing into Non-TRU waste

    International Nuclear Information System (INIS)

    Jung, In Ha; Kim, Jong Ho; Park, Jang Jin; Shin, Jin Myeong; Lee, Ho Hee; Yang, Myung Seung

    2005-01-01

    AFCI and GEN-IV programs aim for decreasing the high level radioactive wastes to be disposed. They also try to get valuable materials to recycle as resources such as uranium and plutonium. On the other hand, cladding hull expected to be one-thirds in volume of spent fuel assembly has not studied so much in the point view of recycling to reuse. Since traditional process of reprocessing was wet process, cladding hull generating through the reprocessing process was unavoidably contaminated with TRU by acid solvent during the process. Therefore, cladding hull has been classified into TRU wastes or high level wastes. According to the strategy for TRU high level radioactive wastes of USA as well as Korea, it regulates in two respects. One is activity and the other is heat generation. In respect of activity, TRU waste contains more than 100 nCi/kg of alpha emits with longer half life than 20 years and higher than 92 in atomic number. Also, wastes are categorized into TRU waste when it generates higher than 2kW/m3, in the respect of heat generation. Our results as well as literatures, almost all of TRU nuclides in the cladding hull are responsible for remained uranium and plutonium owing to pellet-cladding interaction. In addition, recoiled fission products on the surface of the cladding hull serve as heat generator. Up to now, decontamination of the cladding hull generating from the reprocessing of wet process is regarded as valueless and un-economic works owing to the amount of second waste produced

  16. CO2 pellet blasting studies

    International Nuclear Information System (INIS)

    Archibald, K.E.

    1997-01-01

    Initial tests with CO 2 pellet blasting as a decontamination technique were completed in 1993 at the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory (INEL). During 1996, a number of additional CO 2 pellet blasting studies with Alpheus Cleaning Technologies, Oak Ridge National Laboratory, and Pennsylvania State University were conducted. After the testing with Alpheus was complete, an SDI-5 shaved CO 2 blasting unit was purchased by the ICPP to test and determine its capabilities before using in ICPP decontamination efforts. Results of the 1996 testing will be presented in this report

  17. Effects of feeding corn silage, pelleted, ensiled, or pelleted and ...

    African Journals Online (AJOL)

    2017-09-14

    Sep 14, 2017 ... However, lambs fed CS gained weight 21.3% faster than lambs fed AH (259 versus 213 ... ground, pelleted alfalfa more closely resembles a concentrate feed in terms of particle size. .... Vitamin E, 44 IU/g .... feed to reach their market weight versus 102.5 kg of feed for the lambs fed alfalfa haylage (P <0.01).

  18. Building giant plant for pellets; Bygger gigantanlegg for pellets

    Energy Technology Data Exchange (ETDEWEB)

    Nylund, Hilde Kari

    2008-07-01

    At Averoey outside of Kristiansund, Norway, one of the biggest pellet factories in the world is going to be built. The Norwegian companies Hafslund and Moere og Romsdal Biobrensel are together making an effort in bio energy. A brief presentation of the planned plant is provided

  19. Effects of feeding corn silage, pelleted, ensiled, or pelleted and ...

    African Journals Online (AJOL)

    Two experiments were conducted to evaluate the effects of corn silage (CS) and alfalfa (pelleted (AP), haylage (AH), or combination (APH)) on lamb growth and carcass characteristics. The objective of Experiment 1 (Exp. 1) was to compare AH to CS in lamb feedlot diets on lamb growth and carcass characteristics.

  20. Effect of continuous change of sintering atmosphere on the grain growth of Cr-doped UO2 pellets

    International Nuclear Information System (INIS)

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

    2010-01-01

    Cr-doped UO 2 pellet is one of the promising candidates for the high burn-up fuel in commercial LWRs. Major nuclear fuel vendors of such as AREVA or Westinghouse initiated the development of Cr-doped or Cr-containing additives doped UO 2 pellets since at the mid of 90's. Now, qualification programs are on-going to provide these pellets commercially. The main characteristics of the Cr-doped pellets are large-grain and visco-plasticity. Large grain pellet can reduce the corrosive fission gas release at high burn up. Viscoplastic soft pellets can lower the pressure to a cladding caused by a thermal expansion of a pellet at an elevated temperature during transient operations. Those advantages can provide room for additional power uprates and high burnup limits. Especially, PCI resistance improvement can be achieved by enlarging the pellet grain size and enhancing the fuel deformation at an elevated temperature. In this paper, to study the effect of oxygen partial pressure on grain growth in Cr-doped UO 2 pellets, Cr- doped UO 2 samples have been sintered with and without a step-wise change of sintering atmospheres. An introduction of a step-wise variation of oxygen partial pressure during the sintering enhances the grain growth of UO 2 pellets greatly. This step-wise sintering effect has been explained in terms of a continuous increase of Cr concentration along the grain boundary. The observed grain growth behavior under step-wisely changed sintering atmospheres demonstrates the possibility of reducing the amount of Cr 2 O 3 to minimum via control of oxygen partial pressure while keeping the large grain size

  1. ZIRCONIUM-CLADDING OF THORIUM

    Science.gov (United States)

    Beaver, R.J.

    1961-11-21

    A method of cladding thorium with zirconium is described. The quality of the bond achieved between thorium and zirconium by hot-rolling is improved by inserting and melting a thorium-zirconium alloy foil between the two materials prior to rolling. (AEC)

  2. Clad-coolant chemical interaction

    International Nuclear Information System (INIS)

    Iglesias, F.C.; Lewis, B.J.; Desgranges, C.; Toffolon, C.

    2015-01-01

    This paper provides an overview of the kinetics for zircaloy clad oxidation behaviour in steam and air during reactor accident conditions. The generation of chemical heat from metal/water reaction is considered. Low-temperature oxidation of zircaloy due to water-side corrosion is further described. (authors)

  3. Development of high performance cladding

    International Nuclear Information System (INIS)

    Kiuchi, Kiyoshi

    2003-01-01

    The developments of superior next-generation light water reactor are requested on the basis of general view points, such as improvement of safety, economics, reduction of radiation waste and effective utilization of plutonium, until 2030 year in which conventional reactor plants should be renovate. Improvements of stainless steel cladding for conventional high burn-up reactor to more than 100 GWd/t, developments of manufacturing technology for reduced moderation-light water reactor (RMWR) of breeding ratio beyond 1.0 and researches of water-materials interaction on super critical pressure-water cooled reactor are carried out in Japan Atomic Energy Research Institute. Stable austenite stainless steel has been selected for fuel element cladding of advanced boiling water reactor (ABWR). The austenite stain less has the superiority for anti-irradiation properties, corrosion resistance and mechanical strength. A hard spectrum of neutron energy up above 0.1 MeV takes place in core of the reduced moderation-light water reactor, as liquid metal-fast breeding reactor (LMFBR). High performance cladding for the RMWR fuel elements is required to get anti-irradiation properties, corrosion resistance and mechanical strength also. Slow strain rate test (SSRT) of SUS 304 and SUS 316 are carried out for studying stress corrosion cracking (SCC). Irradiation tests in LMFBR are intended to obtain irradiation data for damaged quantity of the cladding materials. (M. Suetake)

  4. The JET multi-pellet injector launcher

    International Nuclear Information System (INIS)

    Kupschus, P.; Bailey, W.; Gadeberg, M.; Hedley, L.; Twyman, P.; Szabo, T.; Evans, D.

    1987-01-01

    Under a collaborative agreement between the Joint European Torus JET and the United States Department of Energy US DOE, JET and Oak Ridge National Laboratory (ORNL) jointly built a multi-pellet injector for fuelling and re-fuelling of the JET plasma. A three-barrel repetitive pneumatic pellet Launcher - built by ORNL - is attached to a JET pellet launcher-machine interface (in short: Pellet Interface) which is the subject of this paper. The present Launcher-Interface combination provides deuterium or hydrogen injection at moderate pellet speeds for the next two operational periods on JET. The Pellet Interface, however, takes into account the future requirements of JET. It was designed to allow the attachment of the high speed pellet launchers now under development at JET and complies with the requirements of remote handling and tritium operation. In addition, the use of tritium pellets is being considered

  5. 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

  6. 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

  7. 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

  8. Development of high performance cladding materials

    International Nuclear Information System (INIS)

    Park, Jeong Yong; Jeong, Y. H.; Park, S. Y.

    2010-04-01

    The irradiation test for HANA claddings conducted and a series of evaluation for next-HANA claddings as well as their in-pile and out-of pile performances tests were also carried out at Halden research reactor. The 6th irradiation test have been completed successfully in Halden research reactor. As a result, HANA claddings showed high performance, such as corrosion resistance increased by 40% compared to Zircaloy-4. The high performance of HANA claddings in Halden test has enabled lead test rod program as the first step of the commercialization of HANA claddings. DB has been established for thermal and LOCA-related properties. It was confirmed from the thermal shock test that the integrity of HANA claddings was maintained in more expanded region than the criteria regulated by NRC. The manufacturing process of strips was established in order to apply HANA alloys, which were originally developed for the claddings, to the spacer grids. 250 kinds of model alloys for the next-generation claddings were designed and manufactured over 4 times and used to select the preliminary candidate alloys for the next-generation claddings. The selected candidate alloys showed 50% better corrosion resistance and 20% improved high temperature oxidation resistance compared to the foreign advanced claddings. We established the manufacturing condition controlling the performance of the dual-cooled claddings by changing the reduction rate in the cold working steps

  9. Study of laser cladding nuclear valve parts

    International Nuclear Information System (INIS)

    Shi Shihong; Wang Xinlin; Huang Guodong

    1998-12-01

    The mechanism of laser cladding is discussed by using heat transfer model of laser cladding, heat conduction model of laser cladding and convective transfer mass model of laser melt-pool. Subsequently the laser cladding speed limit and the influence of laser cladding parameters on cladding layer structure is analyzed. A 5 kW with CO 2 transverse flow is used in the research for cladding treatment of sealing surface of stop valve parts of nuclear power stations. The laser cladding layer is found to be 3.0 mm thick. The cladding surface is smooth and has no such defects as crack, gas pore, etc. A series of comparisons with plasma spurt welding and arc bead welding has been performed. The results show that there are higher grain grade and hardness, lower dilution and better performances of resistance to abrasion, wear and of anti-erosion in the laser cladding layer. The new technology of laser cladding can obviously improve the quality of nuclear valve parts. Consequently it is possible to lengthen the service life of nuclear valve and to raise the safety and reliability of the production system

  10. 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)

  11. Pellet injection into ASDEX upgrade plasmas

    International Nuclear Information System (INIS)

    Lang, P.T.; Zohm, H.; Buechl, K.; Fuchs, J.C.; Gehre, O.; Gruber, O.; Lang, R.S.; Mertens, V.; Neuhauser, J.; Salzmann, H.

    1996-04-01

    This work comprises results obtained using the new centrifuge injection system for the two first years of pellet injection experiments at Asdex Upgrade until the end of the 1995 experimental campaign. The main aim of the pellet injection investigation is to develop scenarios allowing for a more flexible plasma density control means of injection of cryogenic solid hydrogen pellets. Efforts have been made to develop scenarios allowing more flexible plasma density control by injecting cryogenic solid hydrogen pellets. While the injection of pellets during ohmic discharges was found to be most efficient and also improves the plasma performance, increasing the auxiliary heating power causes a detoriation of the pellet fuelling efficiency. A further strong reduction of the pellet fuelling efficiency by an additional process was observed for the more reactor-relevant conditions of shallow particle deposition during H-mode phases. With injection during type I ELMy H-mode phases, each pellet was found to trigger the release of an ELM and therefore cause particle losses mainly from the edge region. In the type I ELMy H-mode, only sufficient pellet penetration allowed noticeable, persistent particle deposition in the plasma by the pellets. Applying adequate pellet injection conditions and favourable scenarios using combined pellet/gas puff refuelling, significant density ramp-up to densities exceeding the empirical Greenwald limit by up to a factor of two was achieved even for strongly heated H-mode plasmas. (orig.)

  12. Pellet fired appliances. Market survey. 7. rev. ed.; Pelletheizungen. Marktuebersicht

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-01-15

    The market survey under consideration reports on pellet central heating systems and pellet fired appliances. The main chapters of this market survey are concerned to: (1) Information on wood pellets and pellet fired appliances; (2) Information about the interpretation of the market survey; (3) Survey of all compared pellet fired appliances with respect to the nominal power; (4) Price lists of pellet fired appliances and pellet central heating systems; (5) Type sheets of the compared pellet fired appliances and pellet central heating systems. Finally, this brochure contains the addresses of the produces and distribution partners of pellet fired appliances and pellet central heating systems.

  13. Flow-Induced Vibration Measurement of an Inner Cladding Tube in a Simulated Dual-Cooled Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang Hee; Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Young Ho; Kim, Jae Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    To create an internal coolant flow passage in a dual cooled fuel rod, an inner cladding tube cannot have intermediate supports enough to relieve its vibration. Thus it can be suffered from a flow-induced vibration (FIV) more severely than an outer cladding tube which will be supported by series of spacer grids. It may cause a fatigue failure at welding joints on the cladding's end plug or fluid elastic instability of long, slender inner cladding due to decrease of a critical flow velocity. This is one of the challenging technical issues when a dual cooled fuel assembly is to be realized into a conventional reactor core To study an actual vibration phenomenon of a dual cooled fuel rod, FIV tests using a small-scale test bundle are being carried out. Measurement results of inner cladding tube of two typically simulated rods are presented. Causes of the differences in the vibration amplitude and response spectrum of the inner cladding tube in terms of intermediate support condition and pellet stacking are discussed.

  14. 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)

  15. Performance of IN-706 and PE-16 cladding in mixed-oxide fuel pins

    International Nuclear Information System (INIS)

    Makenas, B.J.; Lawrence, L.A.; Jensen, B.W.

    1982-05-01

    Iron-nickel base, precipitation-strengthened alloys, IN-706 and PE-16, advanced alloy cladding considered for breeder reactor applications, were irradiated in mixed-oxide fuel pins in the HEDL-P-60 subassembly in EBR-II. Initial selection of candidate advanced alloys was done using only nonfueled materials test results. However, to establish the performance characteristics of the candidate cladding alloys, i.e., dimensional stability and structural integrity under conditions of high neutron flux, elevated temperature, and applied stress, it was necessary to irradiate fuel pins under typical operating conditions. Fuel pins were clad with solution treated IN-706 and PE-16 and irradiated to peak fluences of 6.1 x 10 22 n/cm 2 (E > .1 MeV) and 8.8 x 10 22 n/cm 2 (E > .1 MeV) respectively. Fabrication and operating parameters for the fuel pins with the advanced cladding alloy candidates are summarized. Irradiation of HEDL-P-60 was interrupted with the breach of a pin with IN-706 cladding at 5.1 at % and the test was terminated with cladding breach in a pin with PE-16 cladding at 7.6 at %

  16. Power from Pellets Technology and Applications

    CERN Document Server

    Döring, Stefan

    2013-01-01

    This book provides a practical description of the technology of pellet production on the basis of renewable sources as well as the utilization of pellets. The author explains what kinds of biomass are usable in addition to wood, how to produce pellets and how to use pellets to produce energy. Starting with the basics of combustion, gasification and the pelletizing process, several different technologies are described. The design, planning, construction and economic efficiency are discussed as well. The appendix gives useful advice about plant concepts, calculations, addresses, conversion tables and formulas.

  17. Laser cladding of quasicrystalline alloys

    International Nuclear Information System (INIS)

    Audebert, F.; Sirkin, H.; Colaco, R.; Vilar, R.

    1998-01-01

    Quasicrystals are a new class of ordinated structures with metastable characteristics room temperature. Quasicrystalline phases can be obtained by rapid quenching from the melt of some alloys. In general, quasicrystals present properties which make these alloys promising for wear and corrosion resistant coatings applications. During the last years, the development of quasicrystalline coatings by means of thermal spray techniques has been impulsed. However, no references have been found of their application by means of laser techniques. In this work four claddings of quasicrystalline compositions formed over aluminium substrate, produced by a continuous CO 2 laser using simultaneous powders mixture injection are presented. The claddings were characterized by X ray diffraction, scanning electron microscopy and Vickers microhardness. (Author) 18 refs

  18. Electrothermal plasma gun as a pellet injector

    International Nuclear Information System (INIS)

    Kincaid, R.W.; Bourham, M.A.

    1994-01-01

    The NCSU electrothermal plasma gun SIRENS has been used to accelerate plastic (Lexan polycarbonate) pellets, to determine the feasibility of the use of electrothermal guns as pellet injectors. The use of an electrothermal gun to inject frozen hydrogenic pellets requires a mechanism to provide protective shells (sabots) for shielding the pellet from ablation during acceleration into and through the barrel of the gun. The gun has been modified to accommodate acceleration of the plastic pellets using special acceleration barrels equipped with diagnostics for velocity and position of the pellet, and targets to absorb the pellet's energy on impact. The length of the acceleration path could be varied between 15 and 45 cm. The discharge energy of the electrothermal gun ranged from 2 to 6 kJ. The pellet velocities have been measured via a set of break wires. Pellet masses were varied between 0.5 and 1.0 grams. Preliminary results on 0.5 and 1.0 g pellets show that the exit velocity reaches 0.9 km/s at 6 kJ input energy to the source. Higher velocities of 1.5 and 2.7 km/s have been achieved using 0.5 and 1.0 gm pellets in 30 cm long barrel, without cleaning the barrel between the shots

  19. Initial deuterium pellet experiments on FTU

    International Nuclear Information System (INIS)

    Snipes, J.A.

    1993-01-01

    Initial experiments have been performed with the Single Pellet INjector (SPIN) on FTU. SPIN is a two-stage cryogenic deuterium pellet injector capable of injection,a pellets with velocities up to 2.5 km/s. The nominal pellet mass for these experiments was approximately 1 x 10 20 atoms. These initial pellet experiments concentrated on studying pellet penetration under a variety of plasma conditions to compare with code predictions and to examine toroidal particle transport. The principal diagnostics used were two fast (∼1 μsec) photomultiplier tubes at nearly opposite toroidal locations with H α (D α ) interference filters (λ = 656 nm), a microwave cavity for pellet mass and velocity, a vertical array of soft x ray diodes without filters looking down onto the pellet, a DCN interferometer for electron density profiles, and a Michelson ECE system for electron temperature profiles. The time integral of the absolutely calibrated fast H α signal appears to give reasonable agreement with the expected pellet mass. Toroidal transport of deuterium ions from the pellet to nearly the opposite side of the tokamak agrees with calculated thermal deuterium velocities near the plasma edge. Comparison of the experimental results with code calculations using the Neutral Gas Shielding model show good agreement for the post-pellet electron temperature and density profiles and the H α profiles in some cases. Calculated penetration distances agree within 20%

  20. 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

  1. 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.)

  2. A centrifuge CO2 pellet cleaning system

    International Nuclear Information System (INIS)

    Foster, C.A.; Fisher, P.W.; Nelson, W.D.; Schechter, D.E.

    1993-01-01

    Centrifuge-based cryogenic pellet accelerator technology, originally developed at Oak Ridge National Laboratory (ORNL) for the purpose of refueling fusion reactors with high-speed pellets of frozen deuterium/tritium,is now being developed as a method of cleaning without the use of conventional solvents. In these applications large quantities of pellets made of frozen CO 2 or argon are accelerated in a high-speed rotor. The accelerated pellet stream is used to clean or etch surfaces. The advantage of this system is that the spent pellets and debris resulting from the cleaning process can be filtered leaving only the debris for disposal. This paper discusses the centrifuge CO 2 pellet cleaning system, the physics model of the pellet impacting the surface, the centrifuge apparatus, and some initial cleaning and etching tests

  3. CLAD DEGRADATION - FEPS SCREENING ARGUMENTS

    International Nuclear Information System (INIS)

    R. Schreiner

    2004-01-01

    The purpose of this report is to evaluate and document the screening of the clad degradation features, events, and processes (FEPs) with respect to modeling used to support the Total System Performance Assessment-License Application (TSPA-LA). This report also addresses the effect of certain FEPs on both the cladding and the commercial spent nuclear fuel (CSNF), DOE-owned spent nuclear fuel (DSNF), and defense high-level waste (DHLW) waste forms, as appropriate to address the effects on multiple materials and both components (FEPs 2.1.09.09.0A, 2.1.09.11.0A, 2.1.11.05.0A, 2.1.12.02.0A, and 2.1.12.03.0A). These FEPs are expected to affect the repository performance during the postclosure regulatory period of 10,000 years after permanent closure. Table 1-1 provides the list of cladding FEPs, including their screening decisions (include or exclude). The primary purpose of this report is to identify and document the analysis, screening decision, and TSPA-LA disposition (for included FEPs) or screening argument (for excluded FEPs) for these FEPs related to clad degradation. In some cases, where a FEP covers multiple technical areas and is shared with other FEP reports, this report may provide only a partial technical basis for the screening of the FEP. The full technical basis for shared FEPs is addressed collectively by the sharing FEP reports. The screening decisions and associated TSPA-LA dispositions or screening arguments from all of the FEP reports are cataloged in a project-specific FEPs database

  4. Pellet refueling program at Risoe

    International Nuclear Information System (INIS)

    Andersen, V.; Chang, C.T.; Joergensen, L.W.; Nielsen, P.; Sillesen, A.H.

    1978-01-01

    The pellet refueling work at Riso has up to now been concentrated at studying the ablation rate of hydrogen pellets in hydrogen and deuterium plasmas in the Puffatron device. The main results of these studies are well known and we shall only give a brief summary including some more recent results relating to the ablation process. The work on the Puffatron device has been completed and we are presently preparing to start ablation studies in a small Tokamak, Dante. This tokamak has only been constructed this summer and ablation studies are expected to begin in the beginning of 1978. We shall give the expected parameters of the tokamak plasma and indicate some of the planned work. In this presentation we shall also report on the theoretical work on refueling taking place at Riso. We have particularly been interested in the effect of α-particles which could significantly alter the conclusions made from present experiments

  5. Corrosion of research reactor aluminium clad spent fuel in water

    International Nuclear Information System (INIS)

    2009-12-01

    A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

  6. Model for pneumatic pellet injection

    International Nuclear Information System (INIS)

    Hogan, J.T.; Milora, S.L.; Schuresko, D.D.

    1983-07-01

    A hydrodynamic code has been developed to model the performance of pneumatic pellet injection systems. The code describes one dimensional, unsteady compressible gas dynamics, including gas friction and heat transfer to the walls in a system with variable area. The mass, momentum, and energy equations are solved with an iterated Lax-Wendroff scheme with additional numerical viscosity. The code is described and comparisons with experimental data are presented

  7. Interaction between aluminium oxide pellets and Zircaloy tubes in steam atmospheres at temperatures above 12000C

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Schanz, G.; Sepold, L.

    1988-09-01

    The burnable poison rods in light water reactors (LWR) consist of Al 2 O 3 /B 4 C pellets surrounded by Zircaloy-4 cladding tubes. In the Al 2 O 3 /B 4 C pellets of a LWR rod alumina is the main constituent (98.6 wt.-%) whereas boron carbide acts as neutron absorber. Failure of the Al 2 O 3 /Zircaloy test rods started at 1350 0 C when first droplets of molten material were observed running down the test bundle forming bundle blockages upon solidification. Post test examinations revealed that the process of liquefaction was initiated by a reduction of alumina by Zircaloy resulting in a (Zr, Al, O) melt which decomposed on cooldown into two metallic phases, a (Zr, Al) alloy and oxygen-stabilized a-Zr(O). The components of an extremely porous ceramic melt were also Zr, Al, and oxygen but with a higher oxygen content compared to the metallic melt. The ceramic melt decomposes on cooldown into an Al 2 O 3 /ZrO 2 eutectic with various amounts of primary constituents. Other types of relocated material were due to melting of essentially unreacted Zircaloy cladding and to debris formation by fracturing of oxidized cladding and Al 2 O 3 pellets stack residues. The interactions between Al 2 O 3 and Zircaloy occurring in a burnable poison rod are furthermore important for the behavior of the entire LWR core because the generated metals are able to attack the UO 2 chemically and dissolve or liquefy the fuel even below the melting point of Zircaloy (1760 0 C). As a result, fuel elements which contain burnable poison rods are expected to fail under severe accident conditions at about 1500 0 C. (orig./HP) [de

  8. Review and evaluation of cladding attack of LMFBR fuel

    International Nuclear Information System (INIS)

    Koizumi, M.; Nagai, S.; Furuya, H.; Muto, T.

    1977-01-01

    The behavior of cladding inner wall corrosion during irradiation was evaluated in terms of fuel density, fuel form, O/M ratio, plutonium concentration, cladding composition, cladding pretreatment, cladding inner diameter, burnup and cladding inner wall temperature. Factors which influence the corrosion are O/M ratio (oxygen to metal ratio), burn up, cladding inner diameter and cladding inner wall temperature. Maximum cladding inner wall corrosion depth was formulated as a function of O/M ratio, burn up and cladding inner wall temperature

  9. Pelletized ponderosa pine bark for adsorption of toxic heavy metals from water

    Directory of Open Access Journals (Sweden)

    Tshabalala, M. A.

    2007-02-01

    Full Text Available Bark flour from ponderosa pine (Pinus ponderosa was consolidated into pellets using citric acid as cross-linking agent. The pellets were evaluated for removal of toxic heavy metals from synthetic aqueous solutions. When soaked in water, pellets did not leach tannins, and they showed high adsorption capacity for Cu(II, Zn(II, Cd(II, and Ni(II under both equilibrium and dynamic adsorption conditions. The experimental data for Cd(II and Zn(II showed a better fit to the Langmuir than to the Freundlich isotherm. The Cu(II data best fit the Freundlich isotherm, and the Ni(II data fitted both Freundlich and Langmuir isotherms equally. According to the Freundlich constant KF, adsorption capacity of pelletized bark for the metal ions in aqueous solution, pH 5.1 ± 0.2, followed the order Cd(II > Cu(II > Zn(II >> Ni(II; according to the Langmuir constant b, adsorption affinity followed the order Cd(II >> Cu(II ≈ Zn(II >> Ni(II. Although data from dynamic column adsorption experiments did not show a good fit to the Thomas kinetic adsorption model, estimates of sorption affinity series of the metal ions on pelletized bark derived from this model were not consistent with the series derived from the Langmuir or Freundlich isotherms and followed the order Cu(II > Zn(II ≈ Cd(II > Ni(II. According to the Thomas kinetic model, the theoretical maximum amounts of metal that can be sorbed on the pelletized bark in a column at influent concentration of ≈10 mg/L and flow rate = 5 mL/min were estimated to be 57, 53, 50, and 27 mg/g for copper, zinc, cadmium, and nickel, respectively. This study demonstrated the potential for converting low-cost bark residues to value-added sorbents using starting materials and chemicals derived from renewable resources. These sorbents can be applied in the removal of toxic heavy metals from waste streams with heavy metal ion concentrations of up to 100 mg/L in the case of Cu(II.

  10. Mechanical behavior of fast reactor fuel pin cladding subjected to simulated overpower transients

    International Nuclear Information System (INIS)

    Johnson, G.D.; Hunter, C.W.

    1978-06-01

    Cladding mechanical property data for analysis and prediction of fuel pin transient behavior were obtained under experimental conditions in which the temperature ramps of reactor transients were simulated. All cladding specimens were 20% CW Type 316 stainless steel and were cut from EBR-II irradiated fuel pins. It was determined that irradiation degraded the cladding ductility and failure strength. Specimens that had been adjacent to the fuel exhibited the poorest properties. Correlations were developed to describe the effect of neutron fluence on the mechanical behavior of the cladding. Metallographic examinations were conducted to characterize the failure mode and to establish the nature of internal and external surface corrosion. Various mechanisms for the fuel adjacency effect were examined and results for helium concentration profiles were presented. Results from the simulated transient tests were compared with TREAT test results

  11. Improvement of the spectroscopic investigation of pellet ablation clouds

    International Nuclear Information System (INIS)

    Koubiti, M.; Ferri, S.; Godbert-Mouret, L.; Marandet, Y.; Rosato, J.; Stamm, R.; Goto, M.; Morita, S.

    2012-11-01

    The method allowing the characterization of the so-called ablation cloud of a pellet from its spectroscopic emission lines (intensities and shapes) is described. It is illustrated using measurements concerning carbon and aluminum pellets injected in the Large Helical Devices (LHD). The electron densities in pellet ablation clouds are sufficiently high that the energy levels of the main emitting species are at Local Thermodynamic Equilibrium (LTE). This justifies the electron temperature determination from the measured intensities using Boltzmann plots. In the case of carbon pellet, the C II 723 nm line was previously fitted with a convolution of a Lorentzian and a Gaussian profiles to determine the electron density. It is proposed here to use more elaborate theoretical profiles accounting for the Stark-Zeeman contributions in order to obtain more accurate plasma parameters especially for the high-resolution spectra in which both Zeeman and Stark features are visible. We present some preliminary comparisons with such spectra which were measured recently in LHD and discuss the possible improvement of the considered investigation technique once all the contributions to the line profile are effectively included. (author)

  12. Fundamentals and industrial applications of high power laser beam cladding

    International Nuclear Information System (INIS)

    Bruck, G.J.

    1988-01-01

    Laser beam cladding has been refined such that clad characteristics are precisely determined through routine process control. This paper reviews the state of the art of laser cladding optical equipment, as well as the fundamental process/clad relationships that have been developed for high power processing. Major categories of industrial laser cladding are described with examples chose to highlight particular process attributes

  13. 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)

  14. The role of a fuel element and its cladding in water cooled reactor dynamics

    International Nuclear Information System (INIS)

    Randles, J.

    1963-10-01

    To clarify the role of fuel element cladding in water reactor dynamics, the heat diffusion and transfer equations are solved in slab geometry for (a) an oscillatory fission power, (b) an oscillatory coolant temperature. From the resulting transfer functions a clear description of the effect of the cladding on the heat flow is obtained. A Mercury autocode programme for evaluating the transfer functions is described. In addition to the slab element, the heat diffusion equations are also solved for a cylindrical system exposed to an oscillatory fission power. The solutions are expressed as transfer functions and are obtainable numerically from another autocode programme. Both of these programmes are used to obtain the power out/ power in transfer function for a typical cylindrical and slab UO 2 fuel pellet clad in zircaloy. The results give a further indication of the effect of the cladding heat capacity over a wide frequency range. It is shown also that the effect of the geometrical difference between a slab and cylindrical fuel element is unimportant provided the surface to volume ratio of the fuel is the same in each case. (author)

  15. The role of a fuel element and its cladding in water cooled reactor dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Randles, J [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1963-10-15

    To clarify the role of fuel element cladding in water reactor dynamics, the heat diffusion and transfer equations are solved in slab geometry for (a) an oscillatory fission power, (b) an oscillatory coolant temperature. From the resulting transfer functions a clear description of the effect of the cladding on the heat flow is obtained. A Mercury autocode programme for evaluating the transfer functions is described. In addition to the slab element, the heat diffusion equations are also solved for a cylindrical system exposed to an oscillatory fission power. The solutions are expressed as transfer functions and are obtainable numerically from another autocode programme. Both of these programmes are used to obtain the power out/ power in transfer function for a typical cylindrical and slab UO{sub 2} fuel pellet clad in zircaloy. The results give a further indication of the effect of the cladding heat capacity over a wide frequency range. It is shown also that the effect of the geometrical difference between a slab and cylindrical fuel element is unimportant provided the surface to volume ratio of the fuel is the same in each case. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-15

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

  17. IFPE/IFA-508 and 515, PCMI Behaviour of Thin Cladding Rods, JAERI and HRP

    International Nuclear Information System (INIS)

    2007-01-01

    Description: To measure the integrated response of UO 2 and its cladding to conditions associated with PCI, the Japan Atomic Energy Research Institute carried out a series of experiments in the Halden BWR. The experiment involved two major objectives. The first was to study the influence of rod design parameters on PCI. Diametral gap, wall cladding thickness, SiO 2 additive, and pellet grain size were used as design parameters. The second objective was to study the influence of pre-irradiation (i.e. burnup) on PCI. The maximum burnup attained in the experiment was 23 MWd/kgU. These research results can be applied to current BWR-type fuel rods. The tests were performed between April 1977 and March 1981

  18. Unirradiated cladding rip-propagation tests

    International Nuclear Information System (INIS)

    Hu, W.L.; Hunter, C.W.

    1981-04-01

    The size of cladding rips which develop when a fuel pin fails can affect the subassembly cooling and determine how rapidly fuel escapes from the pin. The object of the Cladding Rip Propagation Test (CRPT) was to quantify the failure development of cladding so that a more realistic fuel pin failure modeling may be performed. The test results for unirradiated 20% CS 316 stainless steel cladding show significantly different rip propagation behavior at different temperatures. At room temperature, the rip growth is stable as the rip extension increases monotonically with the applied deformation. At 500 0 C, the rip propagation becomes unstable after a short period of stable rip propagation. The rapid propagation rate is approximately 200 m/s, and the critical rip length is 9 mm. At test temperatures above 850 0 C, the cladding exhibits very high failure resistances, and failure occurs by multiple cracking at high cladding deformation. 13 figures

  19. Effect of laser power on clad metal in laser-TIG combined metal cladding

    Science.gov (United States)

    Utsumi, Akihiro; Hino, Takanori; Matsuda, Jun; Tasoda, Takashi; Yoneda, Masafumi; Katsumura, Munehide; Yano, Tetsuo; Araki, Takao

    2003-03-01

    TIG arc welding has been used to date as a method for clad welding of white metal as bearing material. We propose a new clad welding process that combines a CO2 laser and a TIG arc, as a method for cladding at high speed. We hypothesized that this method would permit appropriate control of the melted quantity of base metal by varying the laser power. We carried out cladding while varying the laser power, and investigated the structure near the boundary between the clad layer and the base metal. Using the laser-TIG combined cladding, we found we were able to control appropriately the degree of dilution with the base metal. By applying this result to subsequent cladding, we were able to obtain a clad layer of high quality, which was slightly diluted with the base metal.

  20. Current status of the Run-Beyond-Cladding Breach (RBCB) tests for the Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Batte, G.L.; Pahl, R.G.; Hofman, G.L.

    1993-01-01

    This paper describes the results from the Integral Fast Reactor (IFR) metallic fuel Run-Beyond-Cladding-Breach (RBCB) experiments conducted in the Experimental Breeder Reactor II (EBR-II). Included in the report are scoping test results and the data collected from the prototypical tests as well as the exam results and discussion from a naturally occurring breach of one of the lead IFR fuel tests. All results showed a characteristic delayed neutron and fission gas release pattern that readily allows for identification and evaluation of cladding breach events. Also, cladding breaches are very small and do not propagate during extensive post breach operation. Loss of fuel from breached cladding was found to be insignificant. The paper will conclude with a brief description of future RBCB experiments planned for irradiation in EBR-II

  1. LASER SURFACE CLADDING FOR STRUCTURAL REPAIR

    OpenAIRE

    SANTANU PAUL

    2018-01-01

    Laser cladding is a powder deposition technique, which is used to deposit layers of clad material on a substrate to improve its surface properties. It has widespread application in the repair of dies and molds used in the automobile industry. These molds and dies are subjected to cyclic thermo-mechanical loading and therefore undergo localized damage and wear. The final clad quality and integrity is influenced by various physical phenomena, namely, melt pool morphology, microst...

  2. 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

  3. Laser surface cladding:a literature survey

    OpenAIRE

    Gedda, Hans

    2000-01-01

    This work consists of a literature survey of a laser surface cladding in order to investigate techniques to improve the cladding rate for the process. The high local heat input caused by the high power density of the laser generates stresses and the process is consider as slow when large areas are processed. To avoid these disadvantages the laser cladding process velocity can be increased three or four times by use of preheated wire instead of the powder delivery system. If laser cladding is ...

  4. Modelling cladding response to changing conditions

    Energy Technology Data Exchange (ETDEWEB)

    Tulkki, Ville; Ikonen, Timo [VTT Technical Research Centre of Finland ltd (Finland)

    2016-11-15

    The cladding of the nuclear fuel is subjected to varying conditions during fuel reactor life. Load drops and reversals can be modelled by taking cladding viscoelastic behaviour into account. Viscoelastic contribution to the deformation of metals is usually considered small enough to be ignored, and in many applications it merely contributes to the primary part of the creep curve. With nuclear fuel cladding the high temperature and irradiation as well as the need to analyse the variable load all emphasise the need to also inspect the viscoelasticity of the cladding.

  5. Pulsed Laser Cladding of Ni Based Powder

    Science.gov (United States)

    Pascu, A.; Stanciu, E. M.; Croitoru, C.; Roata, I. C.; Tierean, M. H.

    2017-06-01

    The aim of this paper is to optimize the operational parameters and quality of one step Metco Inconel 718 atomized powder laser cladded tracks, deposited on AISI 316 stainless steel substrate by means of a 1064 nm high power pulsed laser, together with a Precitec cladding head manipulated by a CLOOS 7 axes robot. The optimization of parameters and cladding quality has been assessed through Taguchi interaction matrix and graphical output. The study demonstrates that very good cladded layers with low dilution and increased mechanical proprieties could be fabricated using low laser energy density by involving a pulsed laser.

  6. Protective claddings for high strength chromium alloys

    Science.gov (United States)

    Collins, J. F.

    1971-01-01

    The application of a Cr-Y-Hf-Th alloy as a protective cladding for a high strength chromium alloy was investigated for its effectiveness in inhibiting nitrogen embrittlement of a core alloy. Cladding was accomplished by a combination of hot gas pressure bonding and roll cladding techniques. Based on bend DBTT, the cladding alloy was effective in inhibiting nitrogen embrittlement of the chromium core alloy for up to 720 ks (200hours) in air at 1422 K (2100 F). A significant increase in the bend DBTT occurred with longer time exposures at 1422 K or short time exposures at 1589 K (2400 F).

  7. RIA simulation tests using driver tube for ATF cladding

    Energy Technology Data Exchange (ETDEWEB)

    Cinbiz, Mahmut N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, N. R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lowden, R. R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, K. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

    Pellet-cladding mechanical interaction (PCMI) is a potential failure mechanism for accident-tolerant fuel (ATF) cladding candidates during a reactivity-initiated accident (RIA). This report summarizes Fiscal Year (FY) 2017 research activities that were undertaken to evaluate the PCMI-like hoop-strain-driven mechanical response of ATF cladding candidates. To achieve various RIA-like conditions, a modified-burst test (MBT) device was developed to produce different mechanical pulses. The calibration of the MBT instrument was accomplished by performing mechanical tests on unirradiated Generation-I iron-chromium-aluminum (FeCrAl) alloy samples. Shakedown tests were also conducted in both FY 2016 and FY 2017 using unirradiated hydrided ZIRLO™ tube samples. This milestone report focuses on testing of ATF materials, but the benchmark tests with hydrided ZIRLO™ tube samples are documented in a recent journal article.a For the calibration and benchmark tests, the hoop strain was monitored using strain gauges attached to the sample surface in the hoop direction. A novel digital image correlation (DIC) system composed of a single high-speed camera and an array of six mirrors was developed for the MBT instrument to better resolve the failure behavior of samples and to provide useful data for validation of high-fidelity modeling and simulation tools. The DIC system enable a 360° view of a sample’s outer surface. This feature was added to the instrument to determine the precise failure location on a sample’s surface for strain predictions. The DIC system was tested on several silicon carbide fiber/silicon carbide matrix (SiC/SiC) composite tube samples at various pressurization rates of the driver tube (which correspond to the strain rates for the samples). The hoop strains for various loading conditions were determined for the SiC/SiC composite tube samples. Future work is planned to enhance understanding of the failure behavior of the ATF cladding candidates of age

  8. Evaluation of remaining behavior of halogen on the fabrication of MOX pellet containing Am

    International Nuclear Information System (INIS)

    Ozaki, Yoko; Osaka, Masahiko; Obayashi, Hiroshi; Tanaka, Kenya

    2004-11-01

    It is important to limit the content of halogen elements, namely fluorine and chlorine that are sources of making cladding material corrode, in nuclear fuel from the viewpoint of quality assurance. The halogen content should be more carefully limited in the MOX fuel containing Americium (Am-MOX), which is fabricated in the Alpha-Gamma Facility (AGF) for irradiation testing to be conducted in the experimental fast reactor JOYO, because fluorine may remain in the sintered pellets owing to a formation of AmF 3 known to have a low vapor pressure and may exceeds the limit of 25 ppm. In this study, a series of experimental determination of halogen element in Am-MOX were performed by a combination method of pyrolysis and ion-chromatography for the purpose of an evaluation of behavior of remaining halogen through the sintering process. Oxygen potential, temperature and time were changed as experimental parameters and their effects on the remaining behavior of halogen were examined. It was confirmed that good pellets, which contained small amount of halogen, could be obtained by the sintering for 3 hour at 1700degC in the oxygen potential range from -520 to -390 kJ/mol. In order to analysis of fluorine chemical form in green pellet, thermal analysis was performed. AmF 3 and PuF 3 have been confirmed to remain in the green pellet. (author)

  9. Application of EMILAC to pellet injection

    International Nuclear Information System (INIS)

    Iwamura, Yasuhiro; Yamasaki, Takao; Nakamura, Hirone; Hashimoto, Mitsuo; Miya, Kenzo

    1987-01-01

    A new type of electromagnetic accelerator for pellet injection is proposed. Projectile of cylinder shape is accelerated with the repulsive force generated by a combination of two coils, which are different in purpose. And the accelerator is named EMILAC (Electro-Magnetic Inductive Linear Accelerator). In this paper, we investigate the method of applying EMILAC to pellet injection, and calculate the ablation rate of pellet. (author)

  10. Pellet injectors for steady state plasma fuelling

    International Nuclear Information System (INIS)

    Vinyar, I.; Geraud, A.; Yamada, H.; Lukin, A.; Sakamoto, R.; Skoblikov, S.; Umov, A.; Oda, Y.; Gros, G.; Krasilnikov, I.; Reznichenko, P.; Panchenko, V.

    2005-01-01

    Successful steady state operation of a fusion reactor should be supported by repetitive pellet injection of solidified hydrogen isotopes in order to produce high performance plasmas. This paper presents pneumatic pellet injectors and its implementation for long discharge on the LHD and TORE SUPRA, and a new centrifuge pellet injector test results. All injectors are fitted with screw extruders well suited for steady state operation

  11. Porosity influence on UO2 pellet fracture

    International Nuclear Information System (INIS)

    Quadros, N.F. de; Abreu Aires, M. de; Gentile, E.F.

    1976-01-01

    Compression tests were made with UO 2 pellets with grain size of 0,01 mm, approximately the same for all pellets, and with different porosities. The strain rate was 5,5 X 10 -5 sec -1 at room temperature. From fractographic studies and observations made during the compression tests, it was suggested that the pores and flaws resulting from sintering at 1650 0 C, play a fundamental role on the fracture mechanism of the UO 2 pellets [pt

  12. Modeling drying of iron ore pellets

    OpenAIRE

    Ljung, Anna-Lena

    2010-01-01

    Iron ore pellets are a highly refined product supplied to the steel making industry for use in blast furnaces or direct reduction processes. The use of pellets offers many advantages such as customer adopted products, transportability and mechanical strength yet the production is time and energy consuming. Being such, there is a natural driving force to enhance the pelletization in order to optimize production and improve quality. The aim with this thesis is to develop numerical models with w...

  13. 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)

  14. 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

  15. Capabilities of nitrogen admixed cryogenic deuterium pellets

    Energy Technology Data Exchange (ETDEWEB)

    Sharov, Igor; Sergeev, Vladimir [SPU, Saint-Petersburg (Russian Federation); Lang, Peter; Ploeckl, Bernhard; Cavedon, Marco [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Kocsis, Gabor; Szepesi, Tamas [Wigner RCP RMI, Budapest (Hungary); Collaboration: ASDEX Upgrade Team

    2015-05-01

    Operation at high core density with high energy confinement - as foreseen in a future fusion reactor like DEMO - is being investigated at ASDEX Upgrade tokamak. The efficiency of pellet fuelling from the high-field side usually increases with increasing injection speed. Due to the fragile nature of the deuterium ice, however, the increment of pellet mass losses and subsequent pellet fragmentations take place when the speed is increased. Studies show, that admixing of a small amount of nitrogen (N{sub 2}) into D{sub 2} gas can be favorable for the mechanical stability of pellets. This might be helpful for deeper pellet penetration. Besides, seeding by N{sub 2} can enhance plasma performance due to both increasing the energy confinement time and reducing the divertor heat load in the envisaged ELMy H-mode plasma scenario. Fuelling efficiency of N{sub 2}-admixed solid D{sub 2} pellets and their nitrogen seeding capabilities were investigated. It was found that both the overall plasma density increase and the measured averaged pellet penetration depth were smaller in case of the admixed (1% mol. in the gas resulting in about 0.8% in the ice) pellet fuelling. Possibility of the N{sub 2}-seeding by admixed pellets was confirmed by CXRS measurements of N{sup 7+} content in plasma.

  16. 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

  17. Impurity pellet injection experiments at TFTR

    International Nuclear Information System (INIS)

    Marmar, E.S.

    1992-01-01

    Impurity (Li and C) pellet injection experiments on TFTR have produced a number of new and significant results. (1) We observe reproducible improvements of TFTR supershots after wall-conditioning by Li pellet injection ('lithiumization'). (2) We have made accurate measurements of the pitch angle profiles of the internal magnetic field using two novel techniques. The first measures the internal field pitch from the polarization angles of Li + line emission from the pellet ablation cloud, while the second measures the pitch angle profiles by observing the tilt of the cigar-shaped Li + emission region of the ablation cloud. (3) Extensive measurements of impurity pellet penetration into plasmas with central temperatures ranging from ∼0.3 to ∼7 keV have been made and compared with available theoretical models. Other aspects of pellet cloud physics have been investigated. (4) Using pellets as a well defined perturbation has allowed study of transport phenomena. In the case of small pellet perturbations, the characteristics of the background plasmas are probed, while with large pellets, pellet induced effects are clearly observed. These main results are discussed in more detail in this paper

  18. 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)

  19. Ranking of lignocellulosic biomass pellets through multicriteria modeling

    Energy Technology Data Exchange (ETDEWEB)

    Sultana, A.; Kumar, A. [Alberta Univ., Edmonton, AB (Canada). Dept. of Mechanical Engineering

    2009-07-01

    A study was conducted in which pellets from different lignocellulosic biomass sources were ranked using a multicriteria assessment model. Five different pellet alternatives were compared based on 10 criteria. The pair-wise comparison was done in order to develop preference indices for various alternatives. The methodology used in this study was the Preference Ranking Organization Method for Enrichment and Evaluation (PROMETHEE). The biomass included wood pellets, straw pellets, switchgrass pellets, alfalfa pellets and poultry pellets. The study considered both quantitative and qualitative criteria such as energy consumption to produce the pellets, production cost, bulk density, NOx emissions, SOx emissions, deposit formation, net calorific value, moisture content, maturity of technology, and quality of material. A sensitivity analysis was performed by changing weights of criteria and threshold values of the criteria. Different scenarios were developed for ranking cost and environmental impacts. According to preliminary results, the wood pellet is the best energy source, followed by switchgrass pellets, straw pellets, alfalfa pellets and poultry pellets.

  20. Evaluation of the coat quality of sustained release pellets by individual pellet dissolution methodology.

    Science.gov (United States)

    Xu, Min; Liew, Celine Valeria; Heng, Paul Wan Sia

    2015-01-15

    This study explored the application of 400-DS dissolution apparatus 7 for individual pellet dissolution methodology by a design of experiment approach and compared its capability with that of the USP dissolution apparatus 1 and 2 for differentiating the coat quality of sustained release pellets. Drug loaded pellets were prepared by extrusion-spheronization from powder blends comprising 50%, w/w metformin, 25%, w/w microcrystalline cellulose and 25%, w/w lactose, and then coated with ethyl cellulose to produce sustained release pellets with 8% and 10%, w/w coat weight gains. Various pellet properties were investigated, including cumulative drug release behaviours of ensemble and individual pellets. When USP dissolution apparatus 1 and 2 were used for drug release study of the sustained release pellets prepared, floating and clumping of pellets were observed and confounded the release profiles of the ensemble pellets. Hence, the release profiles obtained did not characterize the actual drug release from individual pellet and the applicability of USP dissolution apparatus 1 and 2 to evaluate the coat quality of sustained release pellets was limited. The cumulative release profile of individual pellet using the 400-DS dissolution apparatus 7 was found to be more precise at distinguishing differences in the applied coat quality. The dip speed and dip interval of the reciprocating holder were critical operational parameters of 400-DS dissolution apparatus 7 that affected the drug release rate of a sustained release pellet during the individual dissolution study. The individual dissolution methodology using the 400-DS dissolution apparatus 7 is a promising technique to evaluate the individual pellet coat quality without the influence of confounding factors such as pellet floating and clumping observed during drug release test with dissolution apparatus 1 and 2, as well as to facilitate the elucidation of the actual drug release mechanism conferred by the applied sustained

  1. 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)

  2. 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)

  3. Procedure and apparatus for measuring the radial gap between fuel and surrounding cladding in a fuel rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Olshausen, K.D.

    1976-01-01

    A device is described for measuring non-destructively the annular fuel-cladding gap in an irradiated or fresh fuel rod. The principle applied is that a force is applied to an arm which presses the cladding diametrically, thus deforming it until it touches the fuel pellet. By presenting the values of the force applied and the deformation produced on an XY recorder, the width of the gap is obtained. Alternatively the gap width may be obtained digitally. Since the gap is so small that the deformation is within the elastic range, the fuel rod may be reloaded in the reactor for further irradiation. (JIW)

  4. BWR fuel clad behaviour following LOCA

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R.

    1996-01-01

    Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. (author). 7 refs, 6 figs, 3 tabs

  5. Cladding properties under simulated fuel pin transients

    International Nuclear Information System (INIS)

    Hunter, C.W.; Johnson, G.D.

    1975-01-01

    A description is given of the HEDL fuel pin testing program utilizing a recently developed Fuel Cladding Transient Tester (FCTT) to generate the requisite mechanical property information on irradiated and unirradiated fast reactor fuel cladding under temperature ramp conditions. The test procedure is described, and data are presented

  6. Analysis of corrosion behavior of KOFA cladding

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Kim, Ki Hang; Seo, Keum Seok; Chung, Jin Gon

    1994-01-01

    The corrosion behavior of KOFA cladding was analyzed using the oxide measurement data of KOFA fuel irradiated up to the fuel rod burnup of 35,000 MWD/MTU for two cycles in Kori-2. Even though KOFA cladding is a standard Zircaloy-4 manufactured by Westinghouse according to the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 specification, it was expected that in-pile corrosion behavior of KOFA cladding would not be equivalent to that of Siemens/KWU's cladding due to the differences in such manufacturing processes as cold work and heat treatment. The analysis of measured KOFA cladding oxidation showed that oxidation of KOFA cladding is at least 19 % lower than the design analysis based upon Siemens/KWU's HCW standard Zircaloy-4 cladding. Lower corrosion of KOFA cladding seems to result from the differences in the manufacturing processes and chemical composition although the burnup and oxide layer thickness of the measured fuel rods is relatively low and the amount of the oxidation data base is small

  7. Development of advanced zirconium fuel cladding

    International Nuclear Information System (INIS)

    Jeong, Young Hwan; Park, S. Y.; Lee, M. H.

    2007-04-01

    This report includes the manufacturing technology developed for HANA TM claddings, a series of their characterization results as well as the results of their in-pile and out-of pile performances tests which were carried out to develop some fuel claddings for a high burn-up (70,000MWd/mtU) which are competitive in the world market. Some of the HANA TM claddings, which had been manufactured based on the results from the 1st and 2nd phases of the project, have been tested in a research reactor in Halden of Norway for an in-pile performance qualification. The results of the in-pile test showed that the performance of the HANA TM claddings for corrosion and creep was better than 50% compared to that of Zircaloy-4 or A cladding. It was also found that the out-of pile performance of the HANA TM claddings for such as LOCA and RIA in some accident conditions corrosion creep, tensile, burst and fatigue was superior or equivalent to that of the Zircaloy-4 or A cladding. The project also produced the other many data which were required to get a license for an in-pile test of HANA TM claddings in a commercial reactor. The data for the qualification or characterization were provided for KNFC to assist their activities to get the license for the in-pile test of HANA TM Lead Test Rods(LTR) in a commercial reactor

  8. Corrosion characteristics of K-claddings

    International Nuclear Information System (INIS)

    Park, J. Y.; Choi, B. K.; Jung, Y. H.; Jung, Y. H.

    2004-01-01

    The Improvement of the corrosion resistance of nuclear fuel claddings is the critical issue for the successful development of the high burn-up fuel. KAERI have developed the K-claddings having a superior corrosion resistance by controlling the alloying element addition and optimizing the manufacturing process. The comparative evaluation of the corrosion resistance for K-claddings and the foreign claddings was performed and the effect of the heat treatment on the corrosion behavior of K-claddings was also examined. Corrosion tests were carried out in the conditions of 360 .deg. C pure water, PWR-simulating loop and 400 .deg. C steam, From the results of the corrosion tests, it was found that the corrosion resistance of K-claddings is superior to those of Zry4 and A claddings and K6 showed a better corrosion resistance than K3. The corrosion behavior of K-cladding was strongly influenced by the final annealing rather than the intermediate annealing, and the corrosion resistance increased with decreasing the final annealing temperature

  9. Cladding Alloys for Fluoride Salt Compatibility

    Energy Technology Data Exchange (ETDEWEB)

    Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Walker, Larry R [ORNL; Santella, Michael L [ORNL; Holcomb, David Eugene [ORNL

    2011-06-01

    This report provides an overview of several candidate technologies for cladding nickel-based corrosion protection layers onto high-temperature structural alloys. The report also provides a brief overview of the welding and weld performance issues associated with joining nickel-clad nickel-based alloys. From the available techniques, two cladding technologies were selected for initial evaluation. The first technique is a line-of-sight method that would be useful for cladding large structures such as vessel interiors or large piping. The line-of-sight method is a laser-based surface cladding technique in which a high-purity nickel powder mixed into a polymer binder is first sprayed onto the surface, baked, and then rapidly melted using a high-power laser. The second technique is a vapor phase technique based on the nickel-carbonyl process that is suitable for cladding inaccessible surfaces such as the interior surfaces of heat exchangers. An initial evaluation for performed on the quality of nickel claddings processed using the two selected cladding techniques.

  10. Micromorphology of pelletized soil conditioners

    Science.gov (United States)

    Hirsch, Florian; Dietrich, Nils; Knoop, Christine; Raab, Thomas

    2017-04-01

    Soil conditioners produced by anaerobic digestion and subsequent composting of organic household waste, bear the potential to improve unproductive farmland together with a reduced input risk of unwanted pollutants into the soils. Within the VeNGA project (http://www.biogas-network.de/venga), soil conditioners from anaerobically digested organic household waste are tested for their potential to increase plant growth in glasshouse and field experiments. Because the production techniques of these soil conditioners may influence their physical and chemical behaviour in the soil, two different techniques for pelletizing the soil conditioners where applied. We present findings from a pot experiment with cereal that has been sampled after two months for micromorphological analyses. We visualize the decomposition and the physical behaviour of the soil conditioners. Pellets produced in an agglomeration mixer result in dense balls, that are only slightly decomposed after the trial. But the soil conditioners created under pressure in a screw extruder are rich in voids and have the potential of retaining more soil water.

  11. Flaw behavior in mechanically loaded clad plates

    International Nuclear Information System (INIS)

    Iskander, S.K.; Robinson, G.C.; Oland, C.B.

    1989-01-01

    A small crack near the inner surface of clad nuclear reactor pressure vessels is an important consideration in the safety assessment of the structural integrity of the vessel. Four-point bend tests on large plate specimens, conforming to ASTM specification for pressure vessel plates, alloy steels, quenched and tempered, Mn-Mo and Mn-Mo-Ni (A533) grade B six clad and two unclad with stainless steels 308, 309 and 312 weld wires, were performed to determine the effect of cladding upon the propagation of small surface cracks subjected to stress states. Results indicated that the tough surface layer composed of cladding and/or heat-affected zone has enhanced the load-bearing capacity of plates under conditions where unclad plates have ruptured. The results are interpreted in terms of fracture mechanics. The behavior of flaws in clad reactor pressure vessels is examined in the light of the test results. 11 refs., 8 figs., 2 tabs

  12. Some proposed mechanisms for internal cladding corrosion

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Whitlow, W.H.

    1977-01-01

    In spite of extensive research during recent years, a comprehensive model for internal cladding corrosion in fast reactor oxide fuel pins has not yet been established. In this paper, a model is proposed which accounts for many of the features normally associated with this type of corrosion. The model is composed of a number of parts which describe the chronological sequence of events at the fuel/cladding interface. The corrosion reaction is visualised as being primarily chemical in character, involving the cladding steel, the fuel and the more aggressive fission products, notably caesium in the presence of oxygen. The model attempts to explain how corrosion starts, how it depends on the oxygen potential, why it occurs non-uniformly; also covered are phase changes within the cladding steel and morphological features such as the intergranular form of attack and the distribution of corrosion products in the fuel/cladding gap. (author)

  13. Some proposed mechanisms for internal cladding corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Bradbury, M H; Pickering, S; Whitlow, W H [EURATOM (United Kingdom)

    1977-04-01

    In spite of extensive research during recent years, a comprehensive model for internal cladding corrosion in fast reactor oxide fuel pins has not yet been established. In this paper, a model is proposed which accounts for many of the features normally associated with this type of corrosion. The model is composed of a number of parts which describe the chronological sequence of events at the fuel/cladding interface. The corrosion reaction is visualised as being primarily chemical in character, involving the cladding steel, the fuel and the more aggressive fission products, notably caesium in the presence of oxygen. The model attempts to explain how corrosion starts, how it depends on the oxygen potential, why it occurs non-uniformly; also covered are phase changes within the cladding steel and morphological features such as the intergranular form of attack and the distribution of corrosion products in the fuel/cladding gap. (author)

  14. Injection of pellets into the TCA tokamak

    International Nuclear Information System (INIS)

    Martin, Y.

    1993-05-01

    This thesis presents experimental results from the analysis of the ablation process of pellets injected into the TCA tokamak. The determination of scaling laws relating the pellet penetration to the pellet and plasma parameters preceding injection, were used to improve the understanding of the interaction of the pellet with the plasma since a) the pellet and plasma conditions preceding injection were varied over a large range, and b) the estimation of the penetration depth takes into account the influence of striations in the deposition profile. Over 400 pellets with a range of sizes and speeds were injected into a range of plasma parameters in order to create a database from which the scaling laws could be deduced. The ablation characteristics were principally measured with two CCD video cameras, which provided good spatial resolution, and two filtered photomultiplier tubes, which provided good temporal resolution of the light emitted from the pellet ablation cloud. In the text, the traditional methods of analysing these diagnostics are examined with special reference to the presumptions that a) the pellet velocity is constant in the plasma, and b) the light intensity determined from the ablation cloud is proportional to the ablation rate. After successive data reduction from the database, in order to separate the effects of varying different parameters, the main observations were that, a) the pellet penetration varies as the square root of the pellet velocity, b) the scaling laws for the other parameters strongly depend on whether the pellet has sufficient velocity to reach the q=1 rational magnetic surface in the tokamak. (author) 45 refs

  15. 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

  16. 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

  17. 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

  18. 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.)

  19. 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

  20. 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.)

  1. Hydrogen Pellet-Rotating Plasma Interaction

    DEFF Research Database (Denmark)

    Jørgensen, L. W.; Sillesen, Alfred Hegaard; Øster, Flemming

    1977-01-01

    Spectroscopic measurements on the interaction between solid hydrogen pellets and rotating plasmas are reported. It was found that the light emitted is specific to the pellet material, and that the velocity of the ablated H-atoms is of the order of l0^4 m/s. The investigation was carried out...

  2. Emissions from burning of softwood pellets

    International Nuclear Information System (INIS)

    Olsson, Maria; Kjaellstrand, Jennica

    2004-01-01

    Softwood pellets from three different Swedish manufacturers were burnt in laboratory scale to determine compounds emitted. The emissions were sampled on Tenax cartridges and assessed by gas chromatography and mass spectrometry. No large differences in the emissions from pellets from different manufacturers were observed. The major primary semi-volatile compounds released during flaming burning were 2-methoxyphenols from lignin. The methoxyphenols are of interest due to their antioxidant effect, which may counteract health hazards of aromatic hydrocarbons. Glowing combustion released the carcinogenic benzene as the predominant aromatic compound. However, the benzene emissions were lower than from flaming burning. To relate the results from the laboratory burnings to emissions from pellet burners and pellet stoves, chimney emissions were determined for different burning equipments. The pellet burner emitted benzene as the major aromatic compound, whereas the stove and boiler emitted phenolic antioxidants together with benzene. As the demand for pellets increases, different biomass wastes will be considered as raw materials. Ecological aspects and pollution hazards indicate that wood pellets should be used primarily for residential heating, whereas controlled large-scale combustion should be preferred for pellets made of most other types of biomass waste. (Author)

  3. 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.)

  4. Automatic pellet density checking machine using vision technique

    International Nuclear Information System (INIS)

    Kumar, Suman; Raju, Y.S.; Raj Kumar, J.V.; Sairam, S.; Sheela; Hemantha Rao, G.V.S.

    2012-01-01

    Uranium di-oxide powder prepared through chemical process is converted to green pellets through the powder metallurgy route of precompaction and final compaction operations. These green pellets are kept in a molybdenum boat, which consists of a molybdenum base and a shroud. The boats are passed through the high temperature sintering furnaces to achieve required density of pellets. At present MIL standard 105 E is followed for measuring density of sintered pellets in the boat. As per AQL 2.5 of MIL standard, five pellets are collected from each boat, which contains approximately 800 nos of pellets. The densities of these collected pellets are measured. If anyone pellet density is less than the required value, the entire boat of pellets are rejected and sent back for dissolution for further processing. An Automatic Pellet Density Checking Machine (APDCM) was developed to salvage the acceptable density pellets from the rejected boat of pellets

  5. Fabrication and characterization of absorber pellets for FFTF irradiation testing

    International Nuclear Information System (INIS)

    Wilson, C.N.; Hollenberg, G.W.

    1981-01-01

    Methods used for characterization of B 4 C powder and fabricated pellets are summarized. Fabrication techniques used at HEDL for absorber test pellets are reviewed and selected powder and pellet characterization data are presented

  6. Pellet-plasma interactions in tokamaks

    DEFF Research Database (Denmark)

    Chang, C.T.

    1991-01-01

    confinement time, offset by the accumulation of impurities at the plasma core is brought into focus. A possible remedy is suggested to diminish the effect of the impurities. Plausible arguments are presented to explain the apparent controversial observations on the propagation of a fast cooling front ahead......The ablation of a refuelling pellet of solid hydrogen isotopes is governed by the plasma state, especially the density and energy distribution of the electrons. On the other hand, the cryogenic pellet gives rise to perturbations of the plasma temperature and density. Based on extensive experimental...... data, the interaction between the pellet and the plasma is reviewed. Among the subjects discussed are the MHD activity, evolution of temperature and density profiles, and the behaviour of impurities following the injection of a pellet (or pellets). The beneficial effect of density peaking on the energy...

  7. Vertical pellet injection in FTU discharges

    International Nuclear Information System (INIS)

    Giovannozzi, E.; Annibaldi, S.V.; Buratti, P.

    2005-01-01

    Central fuelling and pellet enhanced performance modes have been obtained with pellets injected vertically from the high field side on the FTU tokamak. Four phases have been recognized: ablation of the pellets, drifting plasmoids, MHD modes which take the density to the centre of the discharge and finally an anomalous drift which further increases the density peaking. Pellet ablation data have been compared with values from a pellet ablation and deposition code. Comparison between 0.8 and 1.1 MA discharges at a high magnetic field (B T = 7 T) has been carried out: a higher performance has been obtained with the latter due to the higher target density and the larger inversion radius which would increase the effects of m = 1 modes to take the density to the plasma centre

  8. 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....

  9. Metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, Nina

    This work concerns planar optical waveguide sensors for biosensing applications, with the focus on deep-probe sensing for micron-scale biological objects like bacteria and whole cells. In the last two decades planar metal-clad waveguides have been brieflyintroduced in the literature applied...... for various biosensing applications, however a thorough study of the sensor configurations has not been presented, but is the main subject of this thesis. Optical sensors are generally well suited for bio-sensing asthey show high sensitivity and give an immediate response for minute changes in the refractive...... index of a sample, due to the high sensitivity of optical bio-sensors detection of non-labeled biological objects can be performed. The majority of opticalsensors presented in the literature and commercially available optical sensors are based on evanescent wave sensing, however most of these sensors...

  10. 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)

  11. 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

  12. 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

  13. Chemical compatibility between cladding alloys and advanced fuels

    International Nuclear Information System (INIS)

    Fee, D.C.; Johnson, C.E.

    1975-05-01

    The National Advanced Fuels Program requires chemical, mechanical, and thermophysical properties data for cladding alloys. The compatibility behavior of cladding alloys with advanced fuels is critically reviewed. in carbide fuel pins, the principal compatibility problem is cladding carburization, diffusion of carbon into the cladding matrix accompanied by carbide precipitation. Carburization changes the mechanical properties of the cladding alloy. The extent of carburization increases in sodium (versus gas) bonded fuels. The depth of carburization increases with increasing sesquicarbide (M 2 C 3 ) content of the fuel. In nitride fuel pins, the principal compatibility problem is cladding nitriding, diffusion of nitrogen into the cladding matrix accompanied by nitride precipitation. Nitriding changes the mechanical properties of the cladding alloy. In both carbide and nitride fuel pins, fission products do not migrate appreciably to the cladding and do not appear to contribute to cladding attack. 77 references. (U.S.)

  14. A centrifuge CO2 pellet cleaning system

    Science.gov (United States)

    Foster, C. A.; Fisher, P. W.; Nelson, W. D.; Schechter, D. E.

    1995-01-01

    An advanced turbine/CO2 pellet accelerator is being evaluated as a depaint technology at Oak Ridge National Laboratory (ORNL). The program, sponsored by Warner Robins Air Logistics Center (ALC), Robins Air Force Base, Georgia, has developed a robot-compatible apparatus that efficiently accelerates pellets of dry ice with a high-speed rotating wheel. In comparison to the more conventional compressed air 'sandblast' pellet accelerators, the turbine system can achieve higher pellet speeds, has precise speed control, and is more than ten times as efficient. A preliminary study of the apparatus as a depaint technology has been undertaken. Depaint rates of military epoxy/urethane paint systems on 2024 and 7075 aluminum panels as a function of pellet speed and throughput have been measured. In addition, methods of enhancing the strip rate by combining infra-red heat lamps with pellet blasting and by combining the use of environmentally benign solvents with the pellet blasting have also been studied. The design and operation of the apparatus will be discussed along with data obtained from the depaint studies.

  15. 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)

  16. 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

  17. 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

  18. Friction Surface Cladding of AA1050 on AA2024-T351; influence of clad layer thickness and tool rotation rate

    NARCIS (Netherlands)

    Liu, Shaojie; Bor, Teunis Cornelis; Geijselaers, Hubertus J.M.; Akkerman, Remko

    2015-01-01

    Friction Surfacing Cladding (FSC) is a recently developed solid state process to deposit thin metallic clad layers on a substrate. The process employs a rotating tool with a central opening to supply clad material and support the distribution and bonding of the clad material to the substrate. The

  19. A reliable protocol for the isolation of viable, chondrogenically differentiated human mesenchymal stem cells from high-density pellet cultures.

    Science.gov (United States)

    Ullah, Mujib; Hamouda, Houda; Stich, Stefan; Sittinger, Michael; Ringe, Jochen

    2012-12-01

    Administration of chondrogenically differentiated mesenchymal stem cells (MSC) is discussed as a promising approach for the regenerative treatment of injured or diseased cartilage. The high-density pellet culture is the standard culture for chondrogenic differentiation, but cells in pellets secrete extracellular matrix (ECM) that they become entrapped in. Protocols for cell isolation from pellets often result in cell damage and dedifferentiation towards less differentiated MSC. Therefore, our aim was to develop a reliable protocol for the isolation of viable, chondrogenically differentiated MSC from high-density pellet cultures. Human bone marrow MSC were chondrogenically stimulated with transforming growth factor-β3, and the cartilaginous structure of the pellets was verified by alcian blue staining of cartilage proteoglycans, antibody staining of cartilage collagen type II, and quantitative real-time reverse-transcription polymerase chain reaction of the marker genes COL2A1 and SOX9. Trypsin and collagenases II and P were tested alone or in combination, and for different concentrations and times, to find a protocol for optimized pellet digestion. Whereas trypsin was not able to release viable cells, 90-min digestion with 300 U of collagenase II, 20 U of collagenase P, and 2 mM CaCl2 worked quite well and resulted in about 2.5×10(5) cells/pellet. The protocol was further optimized for the separation of released cells and ECM from each other. Cells were alcian blue and collagen type II positive and expressed COL2A1 and SOX9, verifying a chondrogenic character. However, they had different morphological shapes. The ECM was also uniformly alcian blue and collagen type II positive but showed different organizational and structural forms. To conclude, our protocol allows the reliable isolation of a defined number of viable, chondrogenically differentiated MSC from high-density pellet cultures. Such cells, as well as the ECM components, are of interest as

  20. Fuel-clad heat transfer coefficient of a defected fuel rod

    International Nuclear Information System (INIS)

    Bruet, M.; Stora, J.P.

    1976-01-01

    A special rod has been built with a stack of UO 2 pellets inside a thick zircaloy clad. The atmosphere inside the fuel rod can be changed and particularly the introduction of water is possible. The capsule was inserted in the Siloe pool reactor in a special device equipped with a neutron flux monitor. The fuel centerline temperature and the temperature at a certain radius of the clad were recorded by two thermocouples. The temperature profiles in the fuel and in the cladding have been calculated and then the heat transfer coefficient. In order to check the proper functioning of the device, two runs were successively achieved with a helium atmosphere. Then the helium atmosphere inside the fuel rod was removed and replaced by water. The heat transfer coefficients derived from the measurements at low power level are in agreement with the values given by the model based on thermal conductivity. However, for higher power levels, the heat transfer coefficients become higher than those based on the calculated gap

  1. Exploring the potential of polacrilin potassium as a novel superdisintegrant in microcrystalline cellulose based pellets prepared by extrusion-spheronization

    Directory of Open Access Journals (Sweden)

    Amita K Joshi

    2011-01-01

    Full Text Available Polacrilin potassium (PP, an ion exchange resin, was used as a superdisintegrant to improve the dissolution of rifampicin, from microcrystalline cellulose (MCC based pellets prepared by extrusion-spheronization. Production of fast release pellets by extrusion-spheronization using MCC is a complicated process. In the present study, pellets were prepared containing 50% w/w rifampicin (BCS class II drug and 40% w/w MCC as extrusion-spheronization aid. Different levels of PP and lactose ratio investigated were 0:10, 2:8, 4:6, 6:4, 8:2, and 10:0. Pellets were evaluated for yield, size, size distribution, shape, porosity, friability, residual moisture, and dissolution efficiency (DE at 30 minutes. Incorporation of this novel superdisintegrant had no adverse effect on the mechanical and micromeritic characteristics of pellets. All the batches of pellets showed high yields′, ~90%; narrow particle size distribution; aspect ratio, 1.0-1.1; friability, <1%; and porosity, 45.51-49.84%. Dissolution profiles were compared using model-independent approaches; DE and similarity factor, f 2 . Addition of Polacrilin results in significant improvement in the DE of rifampicin. The dissolution profiles were significantly different from the dissolution profile of pellets formulated without PP. This preliminary study indicates that PP can serve as an effective superdisintegrant in MCC pellets prepared by extrusion-spheronization.

  2. Engineered zircaloy cladding modifications for improved accident tolerance of LWR fuel: US DOE NEUP Integrated Research Project

    International Nuclear Information System (INIS)

    Heuser, Brent

    2013-01-01

    cladding composition to promote precipitation of minor phase(s) during fabrication. These precipitates will be stable under normal operation, but dissolve during the temperature excursions; the migration of solute elements to the free surface will then shift the reaction away from oxide formation. This pathway is referred to as the 'bulk self-healing' solution. A synergistic response of the fuel rod is anticipated in which the combined mitigation of brittle exothermic oxide formation and associated reduction in cladding temperature lead to accident tolerance with respect to cladding failure. The proposed cladding modifications potentially may influence neutronics and thermal hydraulics, both under normal operation and off-normal scenarios; a favourable reactor system response must therefore be demonstrated for both solution pathways. The objectives of the proposed IRP is four-fold: 1) demonstration of the performance of modified cladding material under normal BWR and PWR operation with respect to corrosion, in particular, stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC); 2) the mitigation of accelerated cladding oxidation during off-normal scenarios that fall below unchecked LOCA events, as well as uncovering scenarios that involve used fuel in on-site storage pools; 3) the benchmarking of the fuel performance code against the databases developed in 1 and 2; 4) demonstration of overall reactor system performance with the proposed modifications to the pellet and cladding

  3. MODELLING OF NUCLEAR FUEL CLADDING TUBES CORROSION

    Directory of Open Access Journals (Sweden)

    Miroslav Cech

    2016-12-01

    Full Text Available This paper describes materials made of zirconium-based alloys used for nuclear fuel cladding fabrication. It is focused on corrosion problems their theoretical description and modeling in nuclear engineering.

  4. Method of processing spent fuel cladding tubes

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi; Ouchi, Atsuhiro; Imahashi, Hiromichi.

    1986-01-01

    Purpose: To decrease the residual activity of spent fuel cladding tubes in a short period of time and enable safety storage with simple storage equipments. Constitution: Spent fuel cladding tubes made of zirconium alloys discharged from a nuclear fuel reprocessing step are exposed to a grain boundary embrittling atmosphere to cause grain boundary destruction. This causes grain boundary fractures to the zirconium crystal grains as the matrix of nuclear fuels and then precipitation products precipitated to the grain boundary fractures are removed. The zirconium constituting the nuclear fuel cladding tube and other ingredient elements contained in the precipitation products are separated in this removing step and they are separately stored respectively. As a result, zirconium constituting most part of the composition of the spent nuclear fuel cladding tubes can be stored safely at a low activity level. (Takahashi, M.)

  5. GSGG edge cladding development: Final technical report

    International Nuclear Information System (INIS)

    Izumitani, T.; Meissner, H.E.; Toratani, H.

    1986-01-01

    The objectives of this project have been: (1) Investigate the possibility of chemical etching of GSGG crystal slabs to obtain increased strength. (2) Design and construct a simplified mold assembly for casting cladding glass to the edges of crystal slabs of different dimensions. (3) Conduct casting experiments to evaluate the redesigned mold assembly and to determine stresses as function of thermal expansion coefficient of cladding glass. (4) Clad larger sizes of GGG slabs as they become available. These tasks have been achieved. Chemical etching of GSGG slabs does not appear possible with any other acid than H 3 PO 4 at temperatures above 300 0 C. A mold assembly has been constructed which allowed casting cladding glass around the edges of the largest GGG slabs available (10 x 20 x 160 mm) without causing breakage through the annealing step

  6. Duplex stainless steel surface bay laser cladding

    International Nuclear Information System (INIS)

    Amigo, V.; Pineda, Y.; Segovia, F.; Vicente, A.

    2004-01-01

    Laser cladding is one of the most promising techniques to restore damaged surfaces and achieve properties similar to those of the base metal. In this work, duplex stainless steels have been cladded by a nickel alloy under different processing conditions. The influence of the beam speed and defocusing variables ha been evaluated in the microstructure both of the cladding and heat affected zone, HAZ. These results have been correlated to mechanical properties by means of microhardness measurements from cladding area to base metal through the interface. This technique has shown to be very appropriate to obtain controlled mechanical properties as they are determined by the solidification microstructure, originated by the transfer of mass and heat in the system. (Author) 21 refs

  7. Corrosion behaviour of cladded nickel base alloys

    International Nuclear Information System (INIS)

    Brandl, W.; Ruczinski, D.; Nolde, M.; Blum, J.

    1995-01-01

    As a consequence of the high cost of nickel base alloys their use as surface layers is convenient. In this paper the properties of SA-as well as RES-cladded NiMo 16Cr16Ti and NiCr21Mo14W being produced in single and multi-layer technique are compared and discussed with respect to their corrosion behaviour. Decisive criteria describing the qualities of the claddings are the mass loss, the susceptibility against intergranular corrosion and the pitting corrosion resistance. The results prove that RES cladding is the most suitable technique to produce corrosion resistant nickel base coatings. The corrosion behaviour of a two-layer RES deposition shows a better resistance against pitting than a three layer SAW cladding. 7 refs

  8. CREEP STRAIN CORRELATION FOR IRRADIATED CLADDING

    International Nuclear Information System (INIS)

    P. Macheret

    2001-01-01

    In an attempt to predict the creep deformation of spent nuclear fuel cladding under the repository conditions, different correlations have been developed. One of them, which will be referred to as Murty's correlation in the following, and whose expression is given in Henningson (1998), was developed on the basis of experimental points related to unirradiated Zircaloy cladding (Henningson 1998, p. 56). The objective of this calculation is to adapt Murty's correlation to experimental points pertaining to irradiated Zircaloy cladding. The scope of the calculation is provided by the range of experimental parameters characterized by Zircaloy cladding temperature between 292 C and 420 C, hoop stress between 50 and 630 MPa, and test time extending to 8000 h. As for the burnup of the experimental samples, it ranges between 0.478 and 64 MWd/kgU (i.e., megawatt day per kilogram of uranium), but this is not a parameter of the adapted correlation

  9. 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

  10. 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)

  11. 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)

  12. 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).

  13. 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)

  14. 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).

  15. Impurity pellet injection experiments at TFTR

    International Nuclear Information System (INIS)

    Marmar, E.S.

    1991-01-01

    Impurity (Li and C) pellet experiments, which began at TFTR in 1989, and are expected to continue at least through 1991, have continued to produce new and significant results. The most significant of these are: (1) improvements in TFTR supershots after wall-conditioning by Li pellet injection; (2) accurate measurements of the pitch angle profiles of the internal magnetic field using the polarization angles of line emission from Li + in the pellet ablation cloud; and (3) initial measurements of pitch angle profiles using the tilt of the LI + emission region of the ablation cloud which is stretched out along the field lines

  16. 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)

  17. 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)

  18. Optimization of metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, N.; Horvath, R.; Pedersen, H.C.

    2005-01-01

    The present paper deals with the optimization of metal-clad waveguides for sensor applications to achieve high sensitivity for adlayer and refractive index measurements. By using the Fresnel reflection coefficients both the angular shift and the width of the resonances in the sensorgrams are taken...... into account. Our optimization shows that it is possible for metal-clad waveguides to achieve a sensitivity improvement of 600% compared to surface-plasmon-resonance sensors....

  19. A proposal for pellet production from residual woody biomass in the island of Majorca (Spain

    Directory of Open Access Journals (Sweden)

    Javier Sánchez

    2015-09-01

    Full Text Available The use of residual biomass for energy purposes is of great interest in isolated areas like Majorca for waste reduction, energy sufficiency and renewable energies development. In addition, densification processes lead to easy-to-automate solid biofuels which additionally have higher energy density. The present study aims at (i the estimation of the potential of residual biomass from woody crops as well as from agri-food and wood industries in Majorca, and (ii the analysis of the optimal location of potential pellet plants by means of a GIS approach (location-allocation analysis and a cost evaluation of the pellets production chain. The residual biomass potential from woody crops in Majorca Island was estimated at 35,874 metric tons dry matter (t DM per year, while the wood and agri-food industries produced annually 21,494 t DM and 2717 t DM, respectively. Thus, there would be enough resource available for the installation of 10 pellet plants of 6400 t·year−1 capacity. These plants were optimally located throughout the island of Mallorca with a maximum threshold distance of 28 km for biomass transport from the production points. Values found for the biomass cost at the pellet plant ranged between 57.1 €·t−1 and 63.4 €·t−1 for biomass transport distance of 10 and 28 km. The cost of pelleting amounted to 56.7 €·t−1; adding the concepts of business fee, pellet transport and profit margin (15%, the total cost of pelleting was estimated at 116.6 €·t−1. The present study provides a proposal for pellet production from residual woody biomass that would supply up to 2.8% of the primary energy consumed by the domestic and services sector in the Balearic Islands.

  20. Laser cladding technology to small diameter pipes

    International Nuclear Information System (INIS)

    Fujimagari, H.; Hagiwara, M.; Kojima, T.

    2000-01-01

    A laser cladding method which produces a highly corrosion-resistant material coating layers (cladding) on the austenitic stainless steel (type 304 SS) pipe inner surface was developed to prevent SCC (stress corrosion cracking) occurrence. This technology is applicable to a narrow and long distance area from operators, because of the good accessibility of the YAG (yttrium-aluminum-garnet) laser beam that can be transmitted through an optical fiber. In this method a mixed paste metallic powder and heating-resistive organic solvent are firstly placed on the inner surface of a small pipe, and then a YAG laser beam transmitted through an optical fiber irradiates to the pasted area. A mixed paste will be melted and form a cladding layer subsequently. A cladding layer shows as excellent corrosion resistance property. This laser cladding (LC) method had already applied to several domestic nuclear power plants and had obtained a good reputation. This report introduces the outline of laser cladding technology, the developed equipment for practical application in the field, and the circumstance in actual plant application. (orig.)

  1. Production and ejection of solid hydrogen-isotope pellet (single pellet)

    International Nuclear Information System (INIS)

    Kasai, Satoshi; Hasegawa, Koichi; Miura, Yukitoshi; Ishibori, Ikuo

    1986-03-01

    The pneumatic gun type pellet injector (single pellet) has been constructed, which is basic type used at ORNL. The pellet in the carrier is 1.65 mm in diameter and 1.65 mm in length, and another is 1 mmD x 1 mmL. Hydrogen pellet velocity of about 900 m/s was observed at propellant gas (He) pressure of 14 kg/cm 2 . In the injection experiment into a plasma, typical velocity is 714 ∼ 833 m/s. These values are 80 ∼ 95 % of velocity calculated from the ideal gun model. The ejected pellet size is 71 ∼ 90 % of the hole size in the carrier disk (1.65 mmD x 1.65 mmL) and 46 ∼ 56 % (1 mmD x 1 mmL). The spread in the pellet trajectories is about 26 mm in diameter at a plasma center. (author)

  2. Studies on the use of nuclear fuel kernels in cladding tubes

    International Nuclear Information System (INIS)

    Thomas, G.

    1981-12-01

    Two approaches for using UO 2 -kernels in cladding tubes have been investigated, viz. the preparation of dense sphere-pacs and direct pelletizing (spherical). A theoretical study on the packing of spheres of different sizes showed that practical experiments were required. Model tests were, therefore, carried out, mostly with glass spheres. The most important results obtained are: A packing density of 80% can be exceeded if spheres of two sizes are used; quick and simple packing can be achieved with the mixing chute presented here; spheres pacs with a density of 90% for LWR cannot be prepared with kernels of practicable sizes; packing results can be translated to other tube diameters and to spheres and tubes made of other materials. The only suitable way to prepare dense pellets from kernels is pressing with a floating matrix at about 10 kbar, followed by removal under residual load. The kernels used should be produced without PVA and be reduced between 500 0 C and 800 0 C. Sintering is best accomplished in a limited oxidizing atmosphere at 1100 0 C with subsequent reduction. Stable pellets with up to 96% of their theoretical density could be produced this way. (orig.) [de

  3. Measurement of the friction coefficient between UO2 and cladding tube

    International Nuclear Information System (INIS)

    Tachibana, Toshimichi; Narita, Daisuke; Kaneko, Hiromitsu; Honda, Yutaka

    1978-01-01

    Most of fuel rods used for light water reactors or fast reactors consist of the cladding tubes filled with UO 2 -PuO 2 pellets. The measurement was made on the coefficient of static friction and the coefficient of dynamic friction in helium under high contact load on UO 2 /Zry-2 and UO 2 /SUS 316 combined samples at the temperature ranging from room temperature to 400 deg. C and from room temperature to 600 deg. C, respectively. The coefficient of static friction for Zry-2 tube and UO 2 pellets was 0.32 +- 0.08 at room temperature and 0.47 +- 0.07 at 400 deg. C, and increased with temperature rise in this temperature range. The coefficient of static friction between 316 stainless steel tube and UO 2 pellets was 0.29 +- 0.04 at room temperature and 1.2 +- 0.2 at 600 deg. C, and increased with temperature rise in this temperature range. The coefficient of dynamic friction for both UO 2 /Zry-2 and UO 2 /SUS 316 combinations seems to be equal to or about 10% excess of the coefficient of static friction. The coefficient of static friction for UO 2 /SUS 316 combination decreased with the increasing number of repetition, when repeating slip several times on the same contact surfaces. (Kobatake, H.)

  4. Performance testing of refractory alloy-clad fuel elements for space reactors

    International Nuclear Information System (INIS)

    Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K.

    1985-01-01

    Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO 2 ) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

  5. SIFAIL: a subprogram to calculate cladding deformation and damage for fast reactor fuel pins

    International Nuclear Information System (INIS)

    Wilson, D.R.; Dutt, D.S.

    1979-05-01

    SIFAIL is a series of subroutines used in conjunction with the thermal performance models of SIEX to assist in the evaluation of mechanical performance of mixed uranium plutonium oxide fuel pins. Cladding deformations due to swelling and creep are calculated. These have been compared to post-irradiation data from fuel pin tests in EBR-II. Several fuel pin cladding failure criteria (cumulative damage, total strain, and thermal creep strain) are evaluated to provide the fuel pin designer with a basis to select design parameters. SIFAIL allows the user many property options for cladding material. Code input is limited to geometric and environmental parameters, with a consistent set of material properties provided by the code. The simplified, yet adequate, thin wall stress--strain calculations provide a reliable estimate of fuel pin mechanical performance, while requiring a small amount of core storage and computer running time

  6. Tritium recovery from lithium oxide pellets

    International Nuclear Information System (INIS)

    Bertone, P.C.; Jassby, D.L.

    1984-01-01

    The TFTR Lithium Blanket Module is an assembly containing 650 kg of lithium oxide that will be used to test the ability of neutronics codes to model the tritium breeding characteristics of limited-coverage breeding zones in a tokamak. It is required that tritium concentrations as low as 0.1 nCi/g bred in both metallic lithium samples and lithium oxide pellets be measured with an uncertainty not exceeding +- 6%. A tritium assay technique for the metallic samples which meets this criterion has been developed. Two assay techniques for the lithium oxide pellets are being investigated. In one, the pellets are heated in a flowing stream of hydrogen, while in the other, the pellets are dissolved in 12 M hydrochloric acid

  7. Operation of the lithium pellet injector

    International Nuclear Information System (INIS)

    Khlopenkov, K.V.; Sudo, S.; Sergeev, V.Yu.

    1996-05-01

    A lithium pellet injection requires an accurate handling with lithium and special technique of loading the pellets. Thus, the technology for this has been developed based on the following conditions: 1) Because of chemical activity of lithium it is necessary to operate in a glove-box with the noble gas atmosphere (He, Ar, etc.). 2) A special procedure of replacing the glove-box atmosphere allows to achieve high purity of the noble gas. 3) When making the pellets it is better to keep the clean lithium in the liquid hexane so as to maintain lithium purity. 4) The pressure of the accelerating gas for Li pellets should be not less than 30 atm. (author)

  8. 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.)

  9. Pellet injection requirements for TORE SUPRA

    International Nuclear Information System (INIS)

    Lafferranderie, J.

    1986-01-01

    The main parameters of TORE SUPRA are outlined and pellet injection requirements to meet plasma density goals are discussed. Topics considered include plasma buildup, plasma refueling and penetration depth

  10. The unaccountability case of plastic pellet pollution.

    Science.gov (United States)

    Karlsson, Therese M; Arneborg, Lars; Broström, Göran; Almroth, Bethanie Carney; Gipperth, Lena; Hassellöv, Martin

    2018-04-01

    Plastic preproduction pellets are found in environmental samples all over the world and their presence is often linked to spills during production and transportation. To better understand how these pellets end up in the environment we assessed the release of plastic pellets from a polyethylene production site in a case study area on the Swedish west coast. The case study encompasses; field measurements to evaluate the level of pollution and pathways, models and drifters to investigate the potential spread and a revision of the legal framework and the company permits. This case study show that millions of pellets are released from the production site annually but also that there are national and international legal frameworks that if implemented could help prevent these spills. Bearing in mind the negative effects observed by plastic pollution there is an urgent need to increase the responsibility and accountability of these spills. Copyright © 2018 The Author(s). Published by Elsevier Ltd.. All rights reserved.

  11. Kinetic depletion model for pellet ablation

    International Nuclear Information System (INIS)

    Kuteev, Boris V.

    2001-11-01

    A kinetic model for depletion effect, which determines pellet ablation when the pellet passes a rational magnetic surface, is formulated. The model predicts a moderate decrease of the ablation rate compared with the earlier considered monoenergy versions [1, 2]. For typical T-10 conditions the ablation rate reduces by a reactor of 2.5 when the 1-mm pellet penetrates through the plasma center. A substantial deceleration of pellets -about 15% per centimeter of low shire rational q region; is predicted. Penetration for Low Field Side and High Field Side injections is considered taking into account modification of the electron distribution function by toroidal magnetic field. It is shown that Shafranov shift and toroidal effects yield the penetration length for HFS injection higher by a factor of 1.5. This fact should be taken into account when plasma-shielding effects on penetration are considered. (author)

  12. UO2 pellet and manufacturing method

    International Nuclear Information System (INIS)

    Komada, Kiichi; Nishinaka, Keiji; Adachi, Kazunori; Fujiwara, Shuji.

    1995-01-01

    The present invention concerns an uranium dioxide pellet having a large crystal grain size. The grain size of the pellet is enlarged to increase the distance of an FP gas generated in the crystal grain to reach the grain boundary and, as a result, decrease the releasing speed of the FP gas. A UO 2 powder having a specific surface area of from 5 to 50m 2 /g is used as a starting powder in a step of forming a molding product, and chlorine or a chlorine compound is added in such an amount that the chlorine content in the UO 2 pellet is from 3 to 25ppm, in one of a production step, a molding step or a sintering step for UO 2 powder. With such procedures, a UO 2 pellet having a large crystal grain size can be prepared with good reproducibility. (T.M.)

  13. Influence of specimen design on the ductility of zircaloy cladding: Experiment and analysis

    International Nuclear Information System (INIS)

    Bates, D. W.; Majumdar, S.; Koss, D. A.; Motta, A. T.

    1999-01-01

    In a reactivity-initiated accident (RIA), a control rod ejection or drop causes a sudden increase in reactor power, which in turn deposits a large amount of energy into the fuel. The resulting thermal expansion and fission gas release loads the cladding into the plastic regime and may cause it to fail. In order to predict cladding survivability, there has been considerable interest and effort in supplementing integral WA tests with separate-effects ring tests of cladding tubes. Such tests can give one insight into failure mechanisms and measure relevant mechanical properties (such as yield strength, uniform elongation, uniaxial stress-strain curve, etc.), for use in computer codes that attempt to predict cladding response during an RIA. The accuracy of such model predictions obviously depends on appropriate and accurate failure data. This study concerns itself with the proper development of ring tensile tests that (i) are similar to the loading conditions present in an RIA, (ii) measure the relevant mechanical properties and (iii) provide insight regarding the influence of the strain paths on the failure mechanisms present if Zircaloy cladding. Based on both experiments and computational modeling, the authors investigate the failure of Zircaloy tubing as a function of specimen geometry, and discuss the limitations of certain ring-test geometries in yielding failure ductility data that are applicable to RIA situations

  14. Short Communication: Emission of Oxygenated Polycyclic Aromatic Hydrocarbons from Biomass Pellet Burning in a Modern Burner for Cooking in China.

    Science.gov (United States)

    Shen, Guofeng; Wei, Siye; Zhang, Yanyan; Wang, Rong; Wang, Bin; Li, Wei; Shen, Huizhong; Huang, Ye; Chen, Yuanchen; Chen, Han; Wei, Wen; Tao, Shu

    2012-12-01

    Biomass pellets are undergoing fast deployment widely in the world, including China. To this stage, there were limited studies on the emissions of various organic pollutants from the burning of those pellets. In addition to parent polycyclic aromatic hydrocarbons, oxygenated PAHs (oPAHs) have been received increased concerns. In this study, emission factors of oPAHs (EF oPAHs ) were measured for two types of pellets made from corn straw and pine wood, respectively. Two combustion modes with (mode II) and without (mode I) secondary side air supply in a modern pellet burner were investigated. For the purpose of comparison, EF oPAHs for raw fuels combusted in a traditional cooking stove were also measured. EF oPAHs were 348±305 and 396±387 µg/kg in the combustion mode II for pine wood and corn straw pellets, respectively. In mode I, measured EF oPAHs were 77.7±49.4 and 189±118 µg/kg, respectively. EFs in mode II were higher (2-5 times) than those in mode I mainly due to the decreased combustion temperature under more excess air. Compared to EF oPAHs for raw corn straw and pine wood burned in a traditional cooking stove, total EF oPAHs for the pellets in mode I were significantly lower ( p < 0.05 ), likely due to increased combustion efficiencies and change in fuel properties. However, the difference between raw biomass fuels and the pellets burned in mode II was not statistically significant. Taking both the increased thermal efficiencies and decreased EFs into consideration, substantial reduction in oPAH emission can be expected if the biomass pellets can be extensively used by rural residents.

  15. Emission of oxygenated polycyclic aromatic hydrocarbons from biomass pellet burning in a modern burner for cooking in China

    Science.gov (United States)

    Shen, Guofeng; Wei, Siye; Zhang, Yanyan; Wang, Rong; Wang, Bin; Li, Wei; Shen, Huizhong; Huang, Ye; Chen, Yuanchen; Chen, Han; Wei, Wen; Tao, Shu

    2012-12-01

    Biomass pellets are undergoing fast deployment widely in the world, including China. To this stage, there were limited studies on the emissions of various organic pollutants from the burning of those pellets. In addition to parent polycyclic aromatic hydrocarbons, oxygenated PAHs (oPAHs) have been received increased concerns. In this study, emission factors of oPAHs (EFoPAHs) were measured for two types of pellets made from corn straw and pine wood, respectively. Two combustion modes with (mode II) and without (mode I) secondary side air supply in a modern pellet burner were investigated. For the purpose of comparison, EFoPAHs for raw fuels combusted in a traditional cooking stove were also measured. EFoPAHs were 348 ± 305 and 396 ± 387 μg kg-1 in the combustion mode II for pine wood and corn straw pellets, respectively. In mode I, measured EFoPAHs were 77.7 ± 49.4 and 189 ± 118 μg kg-1, respectively. EFs in mode II were higher (2-5 times) than those in mode I mainly due to the decreased combustion temperature under more excess air. Compared to EFoPAHs for raw corn straw and pine wood burned in a traditional cooking stove, total EFoPAHs for the pellets in mode I were significantly lower (p pellets burned in mode II was not statistically significant. Taking both the increased thermal efficiencies and decreased EFs into consideration, substantial reduction in oPAH emission can be expected if the biomass pellets can be extensively used by rural residents.

  16. Production, use and reuse of Dutch calcite in drinking water pellet softening

    NARCIS (Netherlands)

    Palmen, LJ; Schetters, M.J.A.; van der Hoek, J.P.; Kramer, O.J.I.; Kors, L.J.; Hofs, B; Koppers, H

    2014-01-01

    In The Netherlands, 50% of the drinking water is treated with pellet softening for various reasons: i) public health (heavy metal solubility), ii) costs (warm water device maintenance, energy and soap requirement), iii) environmental benefits (energy and soap requirement) and iv) customer comfort

  17. 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

  18. 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)

  19. 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

  20. 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)

  1. 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

  2. New automated pellet/powder assay system

    International Nuclear Information System (INIS)

    Olsen, R.N.

    1975-01-01

    This paper discusses an automated, high precision, pellet/ powder assay system. The system is an active assay system using a small isotopic neutron source and a coincidence detection system. The handling of the pellet powder samples has been automated and a programmable calculator has been integrated into the system to provide control and data analysis. The versatile system can assay uranium or plutonium in either active or passive modes

  3. Clad Degradation- Summary and Abstraction for LA

    International Nuclear Information System (INIS)

    D. Stahl

    2004-01-01

    The purpose of this model report is to develop the summary cladding degradation abstraction that will be used in the Total System Performance Assessment for the License Application (TSPA-LA). Most civilian commercial nuclear fuel is encased in Zircaloy cladding. The model addressed in this report is intended to describe the postulated condition of commercial Zircaloy-clad fuel as a function of postclosure time after it is placed in the repository. Earlier total system performance assessments analyzed the waste form as exposed UO 2 , which was available for degradation at the intrinsic dissolution rate. Water in the waste package quickly became saturated with many of the radionuclides, limiting their release rate. In the total system performance assessments for the Viability Assessment and the Site Recommendation, cladding was analyzed as part of the waste form, limiting the amount of fuel available at any time for degradation. The current model is divided into two stages. The first considers predisposal rod failures (most of which occur during reactor operation and associated activities) and postdisposal mechanical failure (from static loading of rocks) as mechanisms for perforating the cladding. Other fuel failure mechanisms including those caused by handling or transportation have been screened out (excluded) or are treated elsewhere. All stainless-steel-clad fuel, which makes up a small percentage of the overall amount of fuel to be stored, is modeled as failed upon placement in the waste packages. The second stage of the degradation model is the splitting of the cladding from the reaction of water or moist air and UO 2 . The splitting has been observed to be rapid in comparison to the total system performance assessment time steps and is modeled to be instantaneous. After the cladding splits, the rind buildup inside the cladding widens the split, increasing the diffusion area from the fuel rind to the waste package interior. This model report summarizes the

  4. Modeling of the cold work stress relieved Zircaloy-4 cladding tubes mechanical behavior under PWR operating conditions

    International Nuclear Information System (INIS)

    Richard, F.; Delobelle, P.; Leclercq, S.; Bouffioux, P.; Rousselier, G.

    2003-01-01

    This paper proposes a damaged viscoplastic model to simulate, for different isotherms (320, 350, 380, 400 and 420 degC), the out-of-flux anisotropic mechanical behavior of cold work stress relieved Zircaloy-4 cladding tubes over the fluence range 0-85.1024 nm -2 (E > 1 MeV). The model, identified from uni and biaxial tests conducted at 350 and 400 degC, is validated from tests performed at 320, 380 and 420 degC. This model is able to simulate strain hardening under internal pressure followed by a stress relaxation period (thermal creep), which is representative of a pellet cladding mechanical interaction occurring during a power transient (class 2 incidental condition). Both the integration of a scalar state variable, characterizing the damage caused by a bombardment with neutrons, and the modification of the static recovery law allowed us to simulate the fast neutron flux effect (irradiation creep). (author)

  5. 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

  6. Advanced turbine/CO2 pellet accelerator

    International Nuclear Information System (INIS)

    Foster, C.A.; Fisher, P.W.

    1994-01-01

    An advanced turbine/CO 2 pellet accelerator is being evaluated as a depaint technology at Oak Ridge National Laboratory. The program, sponsored by Warner Robins Air Logistics Center, Robins Air Force Base, Georgia, has developed a robot-compatible apparatus that efficiently accelerates pellets of dry ice with a high-speed rotating wheel. In comparison to the more conventional compressed air sandblast pellet accelerators, the turbine system can achieve higher pellet speeds, has precise speed control, and is more than ten times as efficient. A preliminary study of the apparatus as a depaint technology has been undertaken. Depaint rates of military epoxy/urethane paint systems on 2024 and 7075 aluminum panels as a function of pellet speed and throughput have been measured. In addition, methods of enhancing the strip rate by combining infra-red heat lamps with pellet blasting have also been studied. The design and operation of the apparatus will be discussed along with data obtained from the depaint studies. Applications include removal of epoxy-based points from aircraft and the cleaning of surfaces contaminated with toxic, hazardous, or radioactive substances. The lack of a secondary contaminated waste stream is of great benefit

  7. Characteristics of pellet injuries to the orbit.

    Science.gov (United States)

    Kükner, A Sahap; Yilmaz, Turgut; Celebi, Serdal; Karslioğlu, Safak; Alagöz, Gürsoy; Serin, Didem; Acar, M Akif; Ozveren, M Faik

    2009-01-01

    To investigate the features of orbital injuries by pellets fired from the front. Retrospective, 4 cases of pellet injuries. Five orbits of 4 patients who sustained pellet injuries received from the front were reviewed retrospectively. The course of injury and results were assessed. Radiological examinations were reviewed. The patients were evaluated between December 1996 and June 2004. Five orbits of 4 patients sustained injuries caused by pellets fired from an anterior direction. The globe in the injured orbit was intact in 2 cases. Severe loss of vision was also present in these 2 globes due to optic nerve involvement. Final visual acuity was down to no light perception in 4 eyes and limited to light perception in 1 eye. The prognosis of orbital pellet injuries is, unfortunately, poor. A pellet passing through the floor of the orbit often causes double perforation of the globe and, once in the orbital aperture, it travels towards the apex as a result of the conical shape of the orbit and lodges in the optic canal or its entrance, severely damaging the optic nerve. Surgery or other treatments are usually unsuccessful. Even if the globe is intact, vision is usually severely impaired. Copyright 2009 S. Karger AG, Basel.

  8. 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

  9. 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

  10. Wood pellets offer a competitive energy option in Sweden

    International Nuclear Information System (INIS)

    2001-01-01

    The market for wood pellets in Sweden grew rapidly during the 1990s and production now stands at around 550,000 tonnes/year. More efficient combustion technology, pellet transportation, pellet storage and pellet delivery have also been developed. The pellets, which are produced by some 25 plants, are used in family houses, large-scale district heating plants, and combined heat and power (CHP) plants. Most of the pellets are made from biomass resources such as forest residues and sawdust and shavings from wood mills. Pellet production, the energy content of saw mill by-products, the current market and its potential for future expansion, the way in which the pellets are used in different combustion systems, the theoretical market potential for wood pellet heating installations in small houses and the Swedish P-certificate system for the certification of pellet stoves and burners are described

  11. Zircaloy PWR fuel cladding deformation tests under mainly convective cooling conditions

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.

    1980-01-01

    In a loss-of-coolant accident the temperature of the cladding of the fuel rods may rise to levels (650-810 0 C) where the ductility of Zircaloy is high (approximately 80%). The net outward pressure which will obtain if the coolant pressure falls to a small fraction of its normal working value produces stresses in the cladding which can result in large strain through secondary creep. An earlier study of the deformation of specimens of PWR Zircaloy cladding tubing 450 mm long under internal pressure had shown that strains of over 50% could be produced over considerable lengths (greater than twenty tube diameters). Extended deformation of this sort might be unacceptable if it occurred in a fuel element. The previous tests had been carried out under conditions of uniform radiative heat loss, and the work reported here extends the study to conditions of mainly convective heat loss believed to be more representative of a fuel element following a loss of coolant. Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 845 0 C in flowing steam at atmospheric pressure. Internal test pressures were in the range 2.9-11.0 MPa (400-1600 1b/in 2 ). Maximum strains were observed of the same magnitude as those seen in the previous tests, but the shape of the deformation differed; in these tests the deformation progressively increased in the direction of the steam flow. These results are compared with those from multi-rod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behaviour of fuel elements in a loss-of-coolant accident are outlined. (author)

  12. 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

  13. Trapping of pellet cloud radiation in thermonuclear plasmas

    International Nuclear Information System (INIS)

    Sergeev, V.Yu.; Miroshinikov, I.V.; Sudo, Shigeru; Namba, C.; Lisitsa, V.S.

    2001-01-01

    The experimental and theoretical data on radiation trapping in clouds of pellets injected into thermonuclear plasmas are presented. The theoretical modeling is performed in terms of equivalent Stark spectral line widths under condition of LTE (Sakha-Boltzman) in pellet cloud plasmas. It is shown that a domain of blackbody radiation could exist in hydrogen pellet clouds resulting in ''pellet disappearance'' effect which is absent in a case of impurity pellet clouds. Reasons for this difference are discussed. (author)

  14. Fabrication of chamfered uranium-plutonium mixed carbide pellets

    International Nuclear Information System (INIS)

    Arai, Yasuo; Iwai, Takashi; Shiozawa, Kenichi; Handa, Muneo

    1985-10-01

    Chamfered uranium-plutonium mixed carbide pellets for high burnup irradiation test in JMTR were fabricated in glove boxes with purified argon gas. The size of die and punch in a press was decided from pellet densities and dimensions including the angle of chamfered parts. No chip or crack caused by adopting chamfered pellets was found in both pressing and sintering stages. In addition to mixed carbide pellets, uranium carbide pellets used as insulators were also successfully fabricated. (author)

  15. Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-08-31

    The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229Th, 230Th, 232U, 233U, and 234U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

  16. Accident tolerant fuel cladding development: Promise, status, and challenges

    Science.gov (United States)

    Terrani, Kurt A.

    2018-04-01

    The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

  17. Residual stress and crack initiation in laser clad composite layer with Co-based alloy and WC + NiCr

    International Nuclear Information System (INIS)

    Lee, Changmin; Park, Hyungkwon; Yoo, Jaehong; Lee, Changhee; Woo, WanChuck; Park, Sunhong

    2015-01-01

    Highlights: • Major problem, clad cracking in laser cladding process, was researched. • Residual stress measurements were performed quantitatively by neutron diffraction method along the surface of specimens. • Relationship between the residual stress and crack initiation was showed clearly. • Ceramic particle effect in the metal matrix was showed from the results of residual stress measurements. • Initiation sites of generating clad cracks were specifically studied in MMC coatings. - Abstract: Although laser cladding process has been widely used to improve the wear and corrosion resistance, there are unwanted cracking issues during and/or after laser cladding. This study investigates the tendency of Co-based WC + NiCr composite layers to cracking during the laser cladding process. Residual stress distributions of the specimen are measured using neutron diffraction and elucidate the correlation between the residual stress and the cracking in three types of cylindrical specimens; (i) no cladding substrate only, (ii) cladding with 100% stellite#6, and (iii) cladding with 55% stellite#6 and 45% technolase40s. The microstructure of the clad layer was composed of Co-based dendrite and brittle eutectic phases at the dendritic boundaries. And WC particles were distributed on the matrix forming intermediate composition region by partial melting of the surface of particles. The overlaid specimen exhibited tensile residual stress, which was accumulated through the beads due to contraction of the coating layer generated by rapid solidification, while the non-clad specimen showed compressive. Also, the specimen overlaid with 55 wt% stellite#6 and 45 wt% technolase40s showed a tensile stress higher than the specimen overlaid with 100% stellite#6 possibly, due to the difference between thermal expansion coefficients of the matrix and WC particles. Such tensile stresses can be potential driving force to provide an easy crack path ways for large brittle fractures

  18. Residual stress and crack initiation in laser clad composite layer with Co-based alloy and WC + NiCr

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Changmin; Park, Hyungkwon; Yoo, Jaehong [Division of Materials Science and Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); Lee, Changhee, E-mail: chlee@hanyang.ac.kr [Division of Materials Science and Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); Woo, WanChuck [Neutron Science Division, Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Park, Sunhong [Research Institute of Industrial Science & Technology, Hyo-ja-dong, Po-Hang, Kyoung-buk, San 32 (Korea, Republic of)

    2015-08-01

    Highlights: • Major problem, clad cracking in laser cladding process, was researched. • Residual stress measurements were performed quantitatively by neutron diffraction method along the surface of specimens. • Relationship between the residual stress and crack initiation was showed clearly. • Ceramic particle effect in the metal matrix was showed from the results of residual stress measurements. • Initiation sites of generating clad cracks were specifically studied in MMC coatings. - Abstract: Although laser cladding process has been widely used to improve the wear and corrosion resistance, there are unwanted cracking issues during and/or after laser cladding. This study investigates the tendency of Co-based WC + NiCr composite layers to cracking during the laser cladding process. Residual stress distributions of the specimen are measured using neutron diffraction and elucidate the correlation between the residual stress and the cracking in three types of cylindrical specimens; (i) no cladding substrate only, (ii) cladding with 100% stellite#6, and (iii) cladding with 55% stellite#6 and 45% technolase40s. The microstructure of the clad layer was composed of Co-based dendrite and brittle eutectic phases at the dendritic boundaries. And WC particles were distributed on the matrix forming intermediate composition region by partial melting of the surface of particles. The overlaid specimen exhibited tensile residual stress, which was accumulated through the beads due to contraction of the coating layer generated by rapid solidification, while the non-clad specimen showed compressive. Also, the specimen overlaid with 55 wt% stellite#6 and 45 wt% technolase40s showed a tensile stress higher than the specimen overlaid with 100% stellite#6 possibly, due to the difference between thermal expansion coefficients of the matrix and WC particles. Such tensile stresses can be potential driving force to provide an easy crack path ways for large brittle fractures

  19. Fuel assembly and fuel cladding tube

    International Nuclear Information System (INIS)

    Tsutsumi, Shinro; Ito, Ken-ichi; Inagaki, Masatoshi; Nakajima, Junjiro.

    1996-01-01

    A fuel cladding tube is a zirconium liner tube formed by lining a pure zirconium layer on the inner side of a zirconium alloy tube. The fuel cladding tube is formed by extrusion molding of a composite billet formed by inserting a pure zirconium billet into a zirconium alloy billet. Accordingly, the pure zirconium layer and the zirconium alloy tube are strongly joined by metal bond. The fuel cladding tube has an external oxide film on the outer surface of the zirconium alloy tube and an internal oxide film on the inner side of the pure zirconium layer. The external oxide film has a thickness preferably of about 1μm. The internal oxide film has a thickness of not more than 10μm, preferably, from 1 to 5μm. With such a constitution, flaws to be formed on both inner and outer surfaces of the cladding tube upon assembling a fuel assembly can be reduced thereby enabling to reduce the amount of hydrogen absorbed to the cladding tube. (I.N.)

  20. Clad buffer rod sensors for liquid metals

    International Nuclear Information System (INIS)

    Jen, C.-K.; Ihara, I.

    1999-01-01

    Clad buffer rods, consisting of a core and a cladding, have been developed for ultrasonic monitoring of liquid metal processing. The cores of these rods are made of low ultrasonic-loss materials and the claddings are fabricated by thermal spray techniques. The clad geometry ensures proper ultrasonic guidance. The lengths of these rods ranges from tens of centimeters to 1m. On-line ultrasonic level measurements in liquid metals such as magnesium at 700 deg C and aluminum at 960 deg C are presented to demonstrate their operation at high temperature and their high ultrasonic performance. A spherical concave lens is machined at the rod end for improving the spatial resolution. High quality ultrasonic images have been obtained in the liquid zinc at 600 deg C. High spatial resolution is needed for the detection of inclusions in liquid metals during processing. We also show that the elastic properties such as density, longitudinal and shear wave velocities of liquid metals can be measured using a transducer which generates and receives both longitudinal and shear waves and is mounted at the end of a clad buffer rod. (author)

  1. 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

  2. Development of repetitive railgun pellet accelerator and steady-state pellet supply system

    International Nuclear Information System (INIS)

    Oda, Y.; Onozuka, M.; Azuma, K.; Kasai, S.; Hasegawa, K.

    1995-01-01

    A railgun system for repetitive high-speed pellet acceleration and steady-state pellet supply system has been developed and investigated. Using a 2m-long railgun system, the hydrogen pellet was accelerated to 2.6km/sec by the supplied energy of 1.7kJ. It is expected that the hydrogen pellet can be accelerated to 3km/sec using the present pneumatic pellet accelerator and a 2m-long augment railgun. Screw-driven hydrogen-isotope filament extruding system has been fabricated and will be tested to examine its applicability to the steady-state extrusion of the solid hydrogen-isotope filament

  3. Development of repetitive railgun pellet accelerator and steady-state pellet supply system

    Energy Technology Data Exchange (ETDEWEB)

    Oda, Y.; Onozuka, M.; Azuma, K. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan); Kasai, S.; Hasegawa, K. [Japan Atomic Energy Research Inst., Naka (Japan)

    1995-12-31

    A railgun system for repetitive high-speed pellet acceleration and steady-state pellet supply system has been developed and investigated. Using a 2m-long railgun system, the hydrogen pellet was accelerated to 2.6km/sec by the supplied energy of 1.7kJ. It is expected that the hydrogen pellet can be accelerated to 3km/sec using the present pneumatic pellet accelerator and a 2m-long augment railgun. Screw-driven hydrogen-isotope filament extruding system has been fabricated and will be tested to examine its applicability to the steady-state extrusion of the solid hydrogen-isotope filament.

  4. Capturing reflected cladding modes from a fiber Bragg grating with a double-clad fiber coupler.

    Science.gov (United States)

    Baiad, Mohamad Diaa; Gagné, Mathieu; Lemire-Renaud, Simon; De Montigny, Etienne; Madore, Wendy-Julie; Godbout, Nicolas; Boudoux, Caroline; Kashyap, Raman

    2013-03-25

    We present a novel measurement scheme using a double-clad fiber coupler (DCFC) and a fiber Bragg grating (FBG) to resolve cladding modes. Direct measurement of the optical spectra and power in the cladding modes is obtained through the use of a specially designed DCFC spliced to a highly reflective FBG written into slightly etched standard photosensitive single mode fiber to match the inner cladding diameter of the DCFC. The DCFC is made by tapering and fusing two double-clad fibers (DCF) together. The device is capable of capturing backward propagating low and high order cladding modes simply and efficiently. Also, we demonstrate the capability of such a device to measure the surrounding refractive index (SRI) with an extremely high sensitivity of 69.769 ± 0.035 μW/RIU and a resolution of 1.433 × 10(-5) ± 8 × 10(-9) RIU between 1.37 and 1.45 RIU. The device provides a large SRI operating range from 1.30 to 1.45 RIU with sufficient discrimination for all individual captured cladding modes. The proposed scheme can be adapted to many different types of bend, temperature, refractive index and other evanescent wave based sensors.

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

    International Nuclear Information System (INIS)

    Rapp, S.W.

    1999-01-01

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

  6. Inspection system for Zircaloy clad fuel rods

    International Nuclear Information System (INIS)

    Yancey, M.E.; Porter, E.H.; Hansen, H.R.

    1975-10-01

    A description is presented of the design, development, and performance of a remote scanning system for nondestructive examination of fuel rods. Characteristics that are examined include microcracking of fuel rod cladding, fuel-cladding interaction, cladding thickness, fuel rod diameter variation, and fuel rod bowing. Microcracking of both the inner and outer fuel rod surfaces and variations in wall thickness are detected by using a pulsed eddy current technique developed by Argonne National Laboratory (ANL). Fuel rod diameter variation and fuel rod bowing are detected by using two linear variable differential transformers (LVDTs) and a signal conditioning system. The system's mechanical features include variable scanning speeds, a precision indexing system, and a servomechanism to maintain proper probe alignment. Initial results indicate that the system is a very useful mechanism for characterizing irradiated fuel rods

  7. Fuel cladding mechanical properties for transient analysis

    International Nuclear Information System (INIS)

    Johnson, G.D.; Hunter, C.W.; Hanson, J.E.

    1976-01-01

    Out-of-pile simulated transient tests have been conducted on irradiated fast-reactor fuel pin cladding specimens at heating rates of 10 0 F/s (5.6 0 K/s) and 200 0 F/s (111 0 K/s) to generate mechanical property information for use in describing cladding behavior during off-normal events. Mechanical property data were then analyzed, applying the Larson-Miller Parameter to the effects of heating rate and neutron fluence. Data from simulated transient tests on TREAT-tested fuel pins demonstrate that Plant Protective System termination of 3$/s transients prevents significant damage to cladding. The breach opening produced during simulated transient testing is shown to decrease in size with increasing neutron fluence

  8. Laser cladding to select new glassy alloys

    International Nuclear Information System (INIS)

    Medrano, L.L.O.; Afonso, C.R.M.; Kiminami, C.S.; Gargarella, P.; Ramasco, B.

    2016-01-01

    A new experimental technique used to analyze the effect of compositional variation and cooling rate in the phase formation in a multicomponent system is the laser cladding. This work have evaluated the use of laser cladding to discover a new bulk metallic glass (BMG) in the Al-Co-Zr system. Coatings with composition variation have made by laser cladding using Al-Co-Zr alloys powders and the samples produced have been characterized by X ray diffraction, microscopy and energy-dispersive X-ray spectroscopy. The results did not show the composition variation as expected, because of incomplete melting during laser process. It was measured a composition variation tendency that allowed the glass forming investigation by the glass formation criterion λ+Δh 1/2 . The results have showed no glass formation in the coating samples, which prove a limited capacity of Zr-Co-Al system to form glass (author)

  9. Potential effects of gallium on cladding materials

    International Nuclear Information System (INIS)

    Wilson, D.F.; Beahm, E.C.; Besmann, T.M.; DeVan, J.H.; DiStefano, J.R.; Gat, U.; Greene, S.R.; Rittenhouse, P.L.; Worley, B.A.

    1997-10-01

    This paper identifies and examines issues concerning the incorporation of gallium in weapons derived plutonium in light water reactor (LWR) MOX fuels. Particular attention is given to the more likely effects of the gallium on the behavior of the cladding material. The chemistry of weapons grade (WG) MOX, including possible consequences of gallium within plutonium agglomerates, was assessed. Based on the calculated oxidation potentials of MOX fuel, the effect that gallium may have on reactions involving fission products and possible impact on cladding performance were postulated. Gallium transport mechanisms are discussed. With an understanding of oxidation potentials and assumptions of mechanisms for gallium transport, possible effects of gallium on corrosion of cladding were evaluated. Potential and unresolved issues and suggested research and development (R and D) required to provide missing information are presented

  10. Stress corrosion testing of irradiated cladding tubes

    International Nuclear Information System (INIS)

    Lunde, L.; Olshausen, K.D.

    1980-01-01

    Samples from two fuel rods with different cladding have been stress corrosion tested by closed-end argon-iodine pressurization at 320 0 C. The fuel rods with stress relieved and recrystallized Zircaloy-2 had received burnups of 10.000 and 20.000 MWd/ton UO 2 , respectively. It was found that the SCC failure stress was unchanged or slightly higher for the irradiated than for the unirradiated control tubes. The tubes failed consistently in the end with the lowest irradiation dose. The diameter increase of the irradiated cladding during the test was 1.1% for the stress-relieved samples and 0.24% for the recrystallized samples. SEM examination revealed no major differences between irradiated and unirradiated cladding. A ''semi-ductile'' fracture zone in recrystallized material is described in some detail. (author)

  11. The control system for the multiple-pellet injector on the Joint European Torus

    International Nuclear Information System (INIS)

    Baylor, L.R.; Jernigan, T.C.; Stewart, K.A.

    1989-01-01

    A stand-alone control and data acquisition system for the Oak Ridge National Laboratory (ORNL) multiple-pellet injector installed on the Joint European Torus (JET) has been designed and installed with the injector. This system, which is based on a MicroVAX II computer and a programmable logic controller (PLC), is an upgrade of previous systems designed for ORNL pellet injectors installed on other fusion experiments. The primary control system upgrades are in the user interface, in the automation of sequential injector operation, and in the analysis of the transient data acquired for each pellet fired. The system is integrated into the JET CODAS environment through CAMAC communications modules with customized communications software. Routine operation of the injector is automated and requires no operator intervention. Details of the hardware and software design and the operation of the system are presented in this paper. 4 refs., 3 figs

  12. The enhanced ASDEX Upgrade pellet centrifuge launcher

    Energy Technology Data Exchange (ETDEWEB)

    Plöckl, B.; Lang, P. T. [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany)

    2013-10-15

    Pellets played an important role in the program of ASDEX Upgrade serving both for investigations on efficient particle fuelling and high density scenarios but also for pioneering work on Edge Localised Mode (ELM) pacing and mitigation. Initially designed for launching fuelling pellets from the magnetic low field side, the system was converted already some time ago to inject pellets from the magnetic high field side as much higher fuelling efficiency was found using this configuration. In operation for more than 20 years, the pellet launching system had to undergo a major revision and upgrading, in particular of its control system. Furthermore, the control system installed adjacent to the launcher had to be transferred to a more distant location enforcing a complete galvanic separation from torus potential and a fully remote control solution. Changing from a hybrid system consisting of PLC S5/S7 and some hard wired relay control to a state of the art PLC system allowed the introduction of several new operational options enabling more flexibility in the pellet experiments. This article describes the new system architecture of control hardware and software, the operating procedure, and the extended operational window. First successful applications for ELM pacing and triggering studies are presented as well as utilization for the development of high density scenarios.

  13. The enhanced ASDEX Upgrade pellet centrifuge launcher

    International Nuclear Information System (INIS)

    Plöckl, B.; Lang, P. T.

    2013-01-01

    Pellets played an important role in the program of ASDEX Upgrade serving both for investigations on efficient particle fuelling and high density scenarios but also for pioneering work on Edge Localised Mode (ELM) pacing and mitigation. Initially designed for launching fuelling pellets from the magnetic low field side, the system was converted already some time ago to inject pellets from the magnetic high field side as much higher fuelling efficiency was found using this configuration. In operation for more than 20 years, the pellet launching system had to undergo a major revision and upgrading, in particular of its control system. Furthermore, the control system installed adjacent to the launcher had to be transferred to a more distant location enforcing a complete galvanic separation from torus potential and a fully remote control solution. Changing from a hybrid system consisting of PLC S5/S7 and some hard wired relay control to a state of the art PLC system allowed the introduction of several new operational options enabling more flexibility in the pellet experiments. This article describes the new system architecture of control hardware and software, the operating procedure, and the extended operational window. First successful applications for ELM pacing and triggering studies are presented as well as utilization for the development of high density scenarios

  14. Lithium Pellet Injector Development for NSTX

    International Nuclear Information System (INIS)

    Gettelfinger, G.; Dong, J.; Gernhardt, R.; Kugel, H.; Sichta, P.; Timberlake, J.

    2003-01-01

    A pellet injector suitable for the injection of lithium and other low-Z pellets of varying mass into plasmas at precise velocities from 5 to 500 m/s is being developed for use on NSTX (National Spherical Torus Experiment). The ability to inject low-Z impurities will significantly expand NSTX experimental capability for a broad range of diagnostic and operational applications. The architecture employs a pellet-carrying cartridge propelled through a guide tube by deuterium gas. Abrupt deceleration of the cartridge at the end of the guide tube results in the pellet continuing along its intended path, thereby giving controlled reproducible velocities for a variety of pellets materials and a reduced gas load to the torus. The planned injector assembly has four hundred guide tubes contained in a rotating magazine with eight tubes provided for injection into plasmas. A PC-based control system is being developed as well and will be described elsewhere in these Proceedings. The development path and mechanical performance of the injector will be described

  15. 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

  16. 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.)

  17. 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

  18. Railgun pellet injection system for fusion experimental devices

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Oda, Y. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Azuma, K. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Satake, K. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Kasai, S. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun 319-11 (Japan); Hasegawa, K. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun 319-11 (Japan)

    1995-11-01

    A railgun pellet injection system has been developed for fusion experimental devices. Using a low electric energy railgun system, hydrogen pellet acceleration tests have been conducted to investigate the application of the electromagnetic railgun system for high speed pellet injection into fusion plasmas. In the system, the pellet is pre-accelerated before railgun acceleration. A laser beam is used to induce plasma armature. The ignited plasma armature is accelerated by an electromagnetic force that accelerates the pellet. Under the same operational conditions, the energy conversion coefficient for the dummy pellets was around 0.4%, while that for the hydrogen pellets was around 0.12%. The highest hydrogen pellet velocity was 1.4 km s{sup -1} using a 1 m long railgun. Based on the findings, it is estimated that the hydrogen pellet has the potential to be accelerated to 5 km s{sup -1} using a 3 m long railgun. (orig.).

  19. Railgun pellet injection system for fusion experimental devices

    International Nuclear Information System (INIS)

    Onozuka, M.; Hasegawa, K.

    1995-01-01

    A railgun pellet injection system has been developed for fusion experimental devices. Using a low electric energy railgun system, hydrogen pellet acceleration tests have been conducted to investigate the application of the electromagnetic railgun system for high speed pellet injection into fusion plasmas. In the system, the pellet is pre-accelerated before railgun acceleration. A laser beam is used to induce plasma armature. The ignited plasma armature is accelerated by an electromagnetic force that accelerates the pellet. Under the same operational conditions, the energy conversion coefficient for the dummy pellets was around 0.4%, while that for the hydrogen pellets was around 0.12%. The highest hydrogen pellet velocity was 1.4 km s -1 using a 1 m long railgun. Based on the findings, it is estimated that the hydrogen pellet has the potential to be accelerated to 5 km s -1 using a 3 m long railgun. (orig.)

  20. Pelletized waste form demonstration program, October 1980-March 1981

    International Nuclear Information System (INIS)

    Lewis, E.L.; Herbert, R.F. Jr.

    1981-01-01

    During the last six months, performance testing of waste/cement pellets was continued. These evaluations included leachability tests and compressive strength tests of cold soil/cement pellets of various compositions. Fractional leach rates (g/cm 2 /day) after 21 months of testing were, in all cases -5 g/cm 2 /day (Mound Acceptance Value). Based upon these recent data, the pressed waste/cement pellets appeared to be a suitable matrix for the immobilization of low-level transuranic wastes. The installation of the Carver custom pellet press was completed. Plutonium-238 contaminated (< 100 nCi/g) ash/cement pellets were produced at a rate of 360 pellets/hr. Pellets of two different compositions were produced, 50% ash/50% cement and 65% ash/35% cement. The compressive strength of sample pellets was slightly lower than expected. Static MCC-1 leachability testing as well as long-term radiolysis testing of sample pellets are scheduled

  1. Stress corrosion crack initiation of Zircaloy-4 cladding tubes in an iodine vapor environment during creep, relaxation, and constant strain rate tests

    Science.gov (United States)

    Jezequel, T.; Auzoux, Q.; Le Boulch, D.; Bono, M.; Andrieu, E.; Blanc, C.; Chabretou, V.; Mozzani, N.; Rautenberg, M.

    2018-02-01

    During accidental power transient conditions with Pellet Cladding Interaction (PCI), the synergistic effect of the stress and strain imposed on the cladding by thermal expansion of the fuel, and corrosion by iodine released as a fission product, may lead to cladding failure by Stress Corrosion Cracking (SCC). In this study, internal pressure tests were conducted on unirradiated cold-worked stress-relieved Zircaloy-4 cladding tubes in an iodine vapor environment. The goal was to investigate the influence of loading type (constant pressure tests, constant circumferential strain rate tests, or constant circumferential strain tests) and test temperature (320, 350, or 380 °C) on iodine-induced stress corrosion cracking (I-SCC). The experimental results obtained with different loading types were consistent with each other. The apparent threshold hoop stress for I-SCC was found to be independent of the test temperature. SEM micrographs of the tested samples showed many pits distributed over the inner surface, which tended to coalesce into large pits in which a microcrack could initiate. A model for the time-to-failure of a cladding tube was developed using finite element simulations of the viscoplastic mechanical behavior of the material and a modified Kachanov's damage growth model. The times-to-failure predicted by this model are consistent with the experimental data.

  2. Comparison of corrosion behavior between fusion cladded and explosive cladded Inconel 625/plain carbon steel bimetal plates

    International Nuclear Information System (INIS)

    Zareie Rajani, H.R.; Akbari Mousavi, S.A.A.; Madani Sani, F.

    2013-01-01

    Highlights: ► Both explosive and fusion cladding aggravate the corrosion resistance of Inconel 625. ► Fusion cladding is more detrimental to nonuniform corrosion resistance. ► Single-layered fusion coat does not show any repassivation ability. ► Adding more layers enhance the corrosion resistance of fusion cladding Inconel 625. ► High impact energy spoils the corrosion resistance of explosive cladding Inconel 625. -- Abstract: One of the main concerns in cladding Inconel 625 superalloy on desired substrates is deterioration of corrosion resistance due to cladding process. The present study aims to compare the effect of fusion cladding and explosive cladding procedures on corrosion behavior of Inconel 625 cladding on plain carbon steel as substrate. Also, an attempt has been made to investigate the role of load ratio and numbers of fusion layers in corrosion behavior of explosive and fusion cladding Inconel 625 respectively. In all cases, the cyclic polarization as an electrochemical method has been applied to assess the corrosion behavior. According to the obtained results, both cladding methods aggravate the corrosion resistance of Inconel 625. However, the fusion cladding process is more detrimental to nonuniform corrosion resistance, where the chemical nonuniformity of fusion cladding superalloy issuing from microsegregation, development of secondary phases and contamination of clad through dilution hinders formation of a stable passive layer. Moreover, it is observed that adding more fusion layers can enhance the nonuniform corrosion resistance of fusion cladding Inconel 625, though this resistance still remains weaker than explosive cladding superalloy. Also, the results indicate that raising the impact energy in explosive cladding procedure drops the corrosion resistance of Inconel 625.

  3. Multilayer cladding with hyperbolic dispersion for plasmonic waveguides

    DEFF Research Database (Denmark)

    Babicheva, Viktoriia; Shalaginov, Mikhail Y.; Ishii, Satoshi

    2015-01-01

    We study the properties of plasmonic waveguides with a dielectric core and multilayer metal-dielectric claddings that possess hyperbolic dispersion. The waveguides hyperbolic multilayer claddings show better performance in comparison to conventional plasmonic waveguides. © OSA 2015....

  4. Cladding using a 15 kW CO2 laser

    International Nuclear Information System (INIS)

    Vesely, E.J.; Verma, S.K.

    1989-01-01

    Laser alloying or cladding differs little in principle from the traditional forms of weld overlays, but lasers as a heat source offer some distinct advantages. With the selective heating attainable using high power lasers, good metallurgical bond of the clad layer, minimal dilution and typically, a very fine homogeneous microstructure can be obtained in the clad layer. This is a review of work in laser cladding using the 15 kW CO 2 laser. The authors discuss the ability of the laser clad surface to increase the high temperature oxidation resistance of a low-alloy carbon steel (4140). Examples of clads subjected to high- temperature thermal cycling of nickel-20% aluminum and TaC + 4140 clad low-alloy steel and straight high-temperature oxidation of Stellite 6-304L cladding on a 4140 substrate are given

  5. Evaluation of fast experimental reactor claddings, (2)

    International Nuclear Information System (INIS)

    Miura, Makoto; Nagaki, Hiroshi; Koyama, Masahiro; Tanaka, Yasumasa

    1974-01-01

    Thin-walled fine tubes of Type 316 austenitic stainless steel are used for fuel cladding in Joyo (experimental FBR). The material exhibits the change of the mechanical properties in long-time annealing at high temperature, resulting from the precipitation of carbide in structure. In this connection, the experiment and the results on the changes of the microstructure and mechanical properties (proof stress and hardness) are described. The test specimens are the fuel cladding tubes produced for trial for Joyo core and those for FFTF core made in the U.S.A. They were heated between 400 0 and 850 0 C for 1000 hr in vacuum. (Mori, K.)

  6. Management of cladding hulls and fuel hardware

    International Nuclear Information System (INIS)

    1985-01-01

    The reprocessing of spent fuel from power reactors based on chop-leach technology produces a solid waste product of cladding hulls and other metallic residues. This report describes the current situation in the management of fuel cladding hulls and hardware. Information is presented on the material composition of such waste together with the heating effects due to neutron-induced activation products and fuel contamination. As no country has established a final disposal route and the corresponding repository, this report also discusses possible disposal routes and various disposal options under consideration at present

  7. Inpile (in PWR) testing of cladding materials

    International Nuclear Information System (INIS)

    Hahn, R.; Jeong, Y. H.; Baek, B. J.; Kim, K. H.; Kim, S. J.; Choi, B. K.; Kim, J. M.

    1999-04-01

    As an introduction, the reasons to perform inpile tests are depicted. An overview over general inpile test procedure is given, and test details which are necessary for the development of new alloys for high burnup claddings, like sample geometries and measuring techniques for inpile corrosion testing, are described in detail. Tests for the creep and length change behavior of cladding tubes are described briefly. Finally, conclusions are drawn and literature citations for further test details are given. (author). 9 refs., 2 tabs., 17 figs

  8. Microstructure of laser cladded martensitic stainless steel

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2006-08-01

    Full Text Available and martensite with 10% ferrite for Material B. Table 7 - Proposed martensitic stainless steel alloys for laser cladding Material C* Cr Ni Mn Si Mo Co Ms (ºC)* Cr eq Ni eq Material A 0.4 13 - 1 0.5 2.5 5.5 120 16.5 12.5 Material B 0.2 15 2 1 0.7 2.5 5.5 117... dilution, low heat input, less distortion, increased mechanical and corrosion properties excellent repeatability and control of process parameters. Solidification of laser cladded martensitic stainless steel is primarily austenitic. Microstructures...

  9. Phosphate-core silica-clad Er/Yb-doped optical fiber and cladding pumped laser.

    Science.gov (United States)

    Egorova, O N; Semjonov, S L; Velmiskin, V V; Yatsenko, Yu P; Sverchkov, S E; Galagan, B I; Denker, B I; Dianov, E M

    2014-04-07

    We present a composite optical fiber with a Er/Yb co-doped phosphate-glass core in a silica glass cladding as well as cladding pumped laser. The fabrication process, optical properties, and lasing parameters are described. The slope efficiency under 980 nm cladding pumping reached 39% with respect to the absorbed pump power and 28% with respect to the coupled pump power. Due to high doping level of the phosphate core optimal length was several times shorter than that of silica core fibers.

  10. Film Coating of Nifedipine Extended Release Pellets in a Fluid Bed Coater with a Wurster Insert

    Directory of Open Access Journals (Sweden)

    Luciane Franquelin Gomes de Souza

    2014-01-01

    Full Text Available The objective of this work was to study the coating process of nifedipine extended release pellets using Opadry and Opadry II, in a fluid bed coater with a Wurster insert. The coating process was studied using a complete experimental design of two factors at two levels for each polymer. The variables studied were the inlet air temperature and the coating suspension flow rate. The agglomerate fraction and coating efficiency were the analyzed response variables. The air temperature was the variable that most influenced the coating efficiency for both polymers. In addition, a study of the dissolution profiles of coated and uncoated pellets using 0.5% sodium lauryl sulfate in simulated gastric fluid without enzymes (pH 1.2 was conducted. The results showed a prolonged release profile for the coated and uncoated pellets that was very similar to the standards established by the U.S. Pharmacopoeia. The drug content and the release profiles were not significantly affected by storage at 40°C and 75% relative humidity. However, when exposed to direct sunlight and fluorescent light (light from fluorescent bulbs, the coated pellets lost only 5% of the drug content, while the uncoated ones lost more than 35%; furthermore, the dissolution profile of the uncoated pellets was faster.

  11. Film Coating of Nifedipine Extended Release Pellets in a Fluid Bed Coater with a Wurster Insert

    Science.gov (United States)

    de Souza, Luciane Franquelin Gomes; Nitz, Marcello; Taranto, Osvaldir Pereira

    2014-01-01

    The objective of this work was to study the coating process of nifedipine extended release pellets using Opadry and Opadry II, in a fluid bed coater with a Wurster insert. The coating process was studied using a complete experimental design of two factors at two levels for each polymer. The variables studied were the inlet air temperature and the coating suspension flow rate. The agglomerate fraction and coating efficiency were the analyzed response variables. The air temperature was the variable that most influenced the coating efficiency for both polymers. In addition, a study of the dissolution profiles of coated and uncoated pellets using 0.5% sodium lauryl sulfate in simulated gastric fluid without enzymes (pH 1.2) was conducted. The results showed a prolonged release profile for the coated and uncoated pellets that was very similar to the standards established by the U.S. Pharmacopoeia. The drug content and the release profiles were not significantly affected by storage at 40°C and 75% relative humidity. However, when exposed to direct sunlight and fluorescent light (light from fluorescent bulbs), the coated pellets lost only 5% of the drug content, while the uncoated ones lost more than 35%; furthermore, the dissolution profile of the uncoated pellets was faster. PMID:24772426

  12. Influences of sodium carbonate on physicochemical properties of lansoprazole in designed multiple coating pellets.

    Science.gov (United States)

    He, Wei; Yang, Min; Fan, Jun Hong; Feng, Cai Xia; Zhang, Su Juan; Wang, Jin Xu; Guan, Pei Pei; Wu, Wei

    2010-09-01

    Lansoprazole (LSP), a proton-pump inhibitor, belongs to class II drug. It is especially instable to heat, light, and acidic media, indicating that fabrication of a formulation stabilizing the drug is difficult. The addition of alkaline stabilizer is the most powerful method to protect the drug in solid formulations under detrimental environment. The purpose of the study was to characterize the designed multiple coating pellets of LSP containing an alkaline stabilizer (sodium carbonate) and assess the effect of the stabilizer on the physicochemical properties of the drug. The coated pellets were prepared by layer-layer film coating with a fluid-bed coater. In vitro release and acid-resistance studies were carried out in simulated gastric fluid and simulated intestinal fluid, respectively. Furthermore, the moisture-uptake test was performed to evaluate the influence of sodium carbonate on the drug stability. The results indicate that the drug exists in the amorphous state or small (nanometer size) particles without crystallization even after storage at 40°C/75% for 5 months. The addition of sodium carbonate to the pellet protects the drug from degradation in simulated gastric fluid in a dose-dependent manner. The moisture absorbed into the pellets has a detrimental effect on the drug stability. The extent of drug degradation is directly correlated with the content of moisture absorption. In conclusion, these results suggest that the presence of sodium carbonate influence the physicochemical properties of LSP, and the designed multiple coating pellets enhance the drug stability.

  13. Construction of in-situ creep strain test facility for the SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Heo, Hyeong Min; Kim, Jun Hwan; Kim, Sung Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, in-situ laser inspection creep test machine was developed for the measuring the creep strain of SFR fuel cladding materials. Ferritic-martensitic steels are being considered as an attractive candidate material for a fuel cladding of a SFR due to their low expansion coefficients, high thermal conductivities and excellent irradiation resistances to a void swelling. HT9 steel (12CrMoVW) is initially developed as a material for power plants in Europe in the 1960. This steel has experienced to expose up to 200dpa in FFTE and EBR-II. Ferritic-Martensitic steel's maximum creep strength in existence is 180Mpa for 106 hour 600 .deg., but HT9 steel is 60Mpa. Because SFR is difficult to secure in developing and applying materials, HT9 steel has accumulated validated data and is suitable for SFR component. And also, because of its superior dimensional stability against fast neutron irradiation, Ferritic-martensitic steel of 9Cr and 12Cr steels, such as HT9 and FC92(12Cr-2W) are preferable to utilize in the fuel cladding of an SFR in KAERI. The pressurized thermal creep test of HT9 and FC92 claddings are being conducted in KAERI, but the change of creep strain in cladding is not easy to measure during the creep test due to its pressurized and closed conditions. In this paper, in-situ laser inspection pressurized creep test machine developed for SFR fuel cladding specimens is described. Moreover, the creep strain rate of HT9 at 650 .deg. C was examined from the in-situ laser inspection pressurized creep test machine.

  14. Analysis of coaxial laser micro cladding processing conditions

    OpenAIRE

    Tarasova, Tatiana Vasilievna; Gvozdeva, Galina Olegovna; Nowotny, Steffen; Ableyeva, Riana R.; Dolzhikova, Evgenia Yu

    2018-01-01

    The laser build-up cladding is a well-known technique for repair, coatings and additive manufacturing tasks. Modern equipment for the laser cladding enables material to be deposited with the lateral resolution of about 100 μm and to manufacture miniature precise parts. However, the micro cladding regimes are unknown. Determination of these regimes is an expensive task as a well-known relation between laser cladding parameters and melt pool dimensions are changing by technology micro-miniaturi...

  15. Electron beam cladding of titanium on stainless steel plate

    International Nuclear Information System (INIS)

    Tomie, Michio; Abe, Nobuyuki; Yamada, Masanori; Noguchi, Shuichi.

    1990-01-01

    Fundamental characteristics of electron beam cladding was investigated. Titanium foil of 0.2mm thickness was cladded on stainless steel plate of 3mm thickness by scanning electron beam. Surface roughness and cladded layer were analyzed by surface roughness tester, microscope, scanning electron microscope and electron probe micro analyzer. Electron beam conditions were discussed for these fundamental characteristics. It is found that the energy density of the electron beam is one of the most important factor for cladding. (author)

  16. 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.)

  17. Microbiological survey of birds of prey pellets.

    Science.gov (United States)

    Dipineto, Ludovico; Bossa, Luigi Maria De Luca; Pace, Antonino; Russo, Tamara Pasqualina; Gargiulo, Antonio; Ciccarelli, Francesca; Raia, Pasquale; Caputo, Vincenzo; Fioretti, Alessandro

    2015-08-01

    A microbiological survey of 73 pellets collected from different birds of prey species housed at the Wildlife Rescue and Rehabilitation Center of Napoli (southern Italy) was performed. Pellets were analyzed by culture and biochemical methods as well as by serotyping and polymerase chain reaction. We isolated a wide range of bacteria some of them also pathogens for humans (i.e. Salmonella enterica serotype Typhimurium, Campylobacter coli, Escherichia coli O serogroups). This study highlights the potential role of birds of prey as asymptomatic carriers of pathogenic bacteria which could be disseminated in the environment not only through the birds of prey feces but also through their pellets. Copyright © 2015 Elsevier Ltd. All rights reserved.

  18. 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

  19. Shock implosion of a small homogeneous pellet

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Y.; Mishkin, E.A.; Alejaldre, C.

    1985-10-01

    A small spherical, or cylindrical, pellet is imploded by an intensive, evenly distributed, short energy pulse. At the surface of the pellet the matter ionizes, its temperature and pressure rapidly rise, and the ablated plasma, by reaction, implodes the inner nucleus of the pellet. The involved structure of the energy absorbing zone is idealized and a sharp deflagration front is considered. With an almost square energy pulse, slightly dropping with time, the solution of the mass, momentum, and energy conservation equations of the compressed matter, is self-similar. The differential equation of the nondimensional position of the deflagration front, its integral, and the magnitude and shape of the outside energy pulse are derived. The process of ablation is shown to depend solely on the nondimensional velocity of the gas just ahead of the deflagration front, minus the speed of sound, or the ratio of the gas densities across the deflagration front.

  20. Shock implosion of a small homogeneous pellet

    International Nuclear Information System (INIS)

    Fujimoto, Y.; Mishkin, E.A.; Alejaldre, C.

    1985-01-01

    A small spherical, or cylindrical, pellet is imploded by an intensive, evenly distributed, short energy pulse. At the surface of the pellet the matter ionizes, its temperature and pressure rapidly rise, and the ablated plasma, by reaction, implodes the inner nucleus of the pellet. The involved structure of the energy absorbing zone is idealized and a sharp deflagration front is considered. With an almost square energy pulse, slightly dropping with time, the solution of the mass, momentum, and energy conservation equations of the compressed matter, is self-similar. The differential equation of the nondimensional position of the deflagration front, its integral, and the magnitude and shape of the outside energy pulse are derived. The process of ablation is shown to depend solely on the nondimensional velocity of the gas just ahead of the deflagration front, minus the speed of sound, or the ratio of the gas densities across the deflagration front

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Polarization effects in silicon-clad optical waveguides

    Science.gov (United States)

    Carson, R. F.; Batchman, T. E.

    1984-01-01

    By changing the thickness of a semiconductor cladding layer deposited on a planar dielectric waveguide, the TE or TM propagating modes may be selectively attenuated. This polarization effect is due to the periodic coupling between the lossless propagating modes of the dielectric slab waveguide and the lossy modes of the cladding layer. Experimental tests involving silicon claddings show high selectivity for either polarization.

  8. Influence of fuel-cladding system deviations from the model of continuous cylinders on the parameters of WWER fuel element working ability

    International Nuclear Information System (INIS)

    Scheglov, A.

    1994-01-01

    In the programs of fuel rod computation, fuel and cladding are usually presented in the form of coaxial cylinders, which can change their sizes, mechanical and thermal-physical properties. The real fuel element has some typical deviations from this continuous coaxial cylinders (CCC) model as: axial asymmetry of fuel-cladding system (due to the oval form of the cladding, cracking and other type of fuel pallet damage, axial asymmetry of the volumetric heat release), gaps between the pallets (and heat release peaking in fuel near the gap), chambers in the pallets. As a result of these deviations actual fuel rod parameters of working ability - temperature, stresses, thermal fluxes relieved from the cladding, geometry changes - in some locations can greatly vary from the ones calculated according to CCC model. The influence of these deviations is extremely important while calculating the fuel rod, because they are a part of the mechanical excess coefficient. The author reviews the influence of these factors using specific examples. He applies his own two-dimensional codes based on the Finite Elements Method for calculations of temperature fields, stresses and deformation in the fuel rod elements. It is shown that consideration of these deviations, as a rule, leads to the increase of the maximum fuel temperature in the WWER pellets (characterized by a large central hole), temperature of the cladding, thermal flux, relieved by the coolant from the cladding, and stresses in the cladding. It is necessary to consider these factors for both validation of the fuel element working ability and interpretation of the experimental results. 4 tabs., 3 figs., 5 refs

  9. Influence of fuel-cladding system deviations from the model of continuous cylinders on the parameters of WWER fuel element working ability

    Energy Technology Data Exchange (ETDEWEB)

    Scheglov, A [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    1994-12-31

    In the programs of fuel rod computation, fuel and cladding are usually presented in the form of coaxial cylinders, which can change their sizes, mechanical and thermal-physical properties. The real fuel element has some typical deviations from this continuous coaxial cylinders (CCC) model as: axial asymmetry of fuel-cladding system (due to the oval form of the cladding, cracking and other type of fuel pallet damage, axial asymmetry of the volumetric heat release), gaps between the pallets (and heat release peaking in fuel near the gap), chambers in the pallets. As a result of these deviations actual fuel rod parameters of working ability - temperature, stresses, thermal fluxes relieved from the cladding, geometry changes - in some locations can greatly vary from the ones calculated according to CCC model. The influence of these deviations is extremely important while calculating the fuel rod, because they are a part of the mechanical excess coefficient. The author reviews the influence of these factors using specific examples. He applies his own two-dimensional codes based on the Finite Elements Method for calculations of temperature fields, stresses and deformation in the fuel rod elements. It is shown that consideration of these deviations, as a rule, leads to the increase of the maximum fuel temperature in the WWER pellets (characterized by a large central hole), temperature of the cladding, thermal flux, relieved by the coolant from the cladding, and stresses in the cladding. It is necessary to consider these factors for both validation of the fuel element working ability and interpretation of the experimental results. 4 tabs., 3 figs., 5 refs.

  10. The investigation of fast reactor fuel pin start up behaviour in the irradiation experiment DUELL II

    International Nuclear Information System (INIS)

    Freund, D.; Geithoff, D.

    1988-04-01

    The irradiation experiments DUELL-II within the SNR-300 operational Transient Experimental Program deal with the investigation of fresh mixed oxide fuel behaviour at start-up. The irradiation has been carried out in the HFR Petten in four so-called DUELL capsules with two fuel pin samples each. The fuel pins with a total length of 453 mm contained a fuel column of 150 mm length, consisting of high dense (U,Pu)O 2-x fuel with an initial porosity of 4%, a Pu-content of 20.9%, and an O/Me ratio of 1.96. The fuel pellet diameter was 6.37 mm, the outer diameter of the SS cladding, material No. 1.4970, was 7.6 mm. The irradiation included four phases, consisting of preconditioning at 85% nominal power (corresponds to 550 W/cm), a following increase to full power, and two following full power periods of 1 and 10 days, respectively. Post irradiation examination showed incomplete fuel restructuring in the first capsules with central void diameters of 800 μm in the hot plane, complete restructuring in the last capsule, leading to central voids of approximately 1 mm diameter. The residual gaps between fuel and clad varied between 25 and 44 μm. The clad inner surface did not show any corrosion attack. The analysis of fuel restructuring has been carried out with the computer code SATURN-S showing good agreement with the PIE results. The analysis led to a series of model improvements, especially for crack volume and relocation modelling. (orig./GL) [de

  11. International Trade of Wood Pellets (Brochure)

    Energy Technology Data Exchange (ETDEWEB)

    2013-05-01

    The production of wood pellets has increased dramatically in recent years due in large part to aggressive emissions policy in the European Union; the main markets that currently supply the European market are North America and Russia. However, current market circumstances and trade dynamics could change depending on the development of emerging markets, foreign exchange rates, and the evolution of carbon policies. This fact sheet outlines the existing and potential participants in the wood pellets market, along with historical data on production, trade, and prices.

  12. Pellet ablation and ablation model development

    International Nuclear Information System (INIS)

    Houlberg, W.A.

    1989-01-01

    A broad survey of pellet ablation is given, based primarily on information presented at this meeting. The implications of various experimental observations for ablation theory are derived from qualitative arguments of the physics involved. The major elements of a more complete ablation theory are then outlined in terms of these observations. This is followed by a few suggestions on improving the connections between theory and experimental results through examination of ablation data. Although this is a rather aggressive undertaking for such a brief (and undoubtedly incomplete) assessment, some of the discussion may help us advance the understanding of pellet ablation. 17 refs

  13. 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

  14. 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

  15. Investigations of the interaction between ballooning Zircaloy cladding and emergency core cooling

    International Nuclear Information System (INIS)

    Wiehr, K.; Barth, S.; Erbacher, F.; Hame, W.; Harten, U.; Just, W.; Megerle, A.; Mueller, S.; Neitzel, H.J.; Reimann; Schaeffner, P.; Schmidt, H.

    1975-01-01

    The development of fabrication methods for the production of fuel rod simulators has been largely terminated. For welding of Zircaloy-4 and inconel 600 explosive welding has proved to be promissory in preliminary tests. A prototype fuel rod simulator was tested at full power. Its performance was faultless and the fuel rod and ring pellets could be easily dismantled and reused after the experiment. Planning of the test rig and electricity supply were terminated. Most of the assembly work has been finished. For electric heating of the fuel rod simulators a special device was built and tested which allows to program the power control. The radiographic system recording ballooning of the Zircaloy clad was erected outside the test space and put into operation. First trial pictures yielded good results. (orig.) [de

  16. Introduction program of M5TM cladding in Japan

    International Nuclear Information System (INIS)

    Mardon, Jean Paul; Kaneko, Nori

    2008-01-01

    Experience from irradiation in PWR has confirmed that M5 TM possesses all the properties required for upgraded operation including new fuel management approaches and high duty reactor operation. Specifically, the alloy M5 TM has demonstrated impressive improvements over Zircaloy-4 for fuel rod cladding and fuel assembly structural components. Moreover, several irradiation campaigns have been worldwide performed in order to confirm the excellent M5 TM in-pile behavior in very demanding PWR irradiation conditions (high void fraction, heat flux, temperature, lithium content and Zinc injection). Regarding licensing, the authorization for loading M5 TM alloy has been granted by US, UK, South Korean, German, Chinese, South-African, Swedish and Belgian Safety Authorities. Also the French Nuclear Safety Authority has given individually its authorization to load all-M5 TM fuel assembly batches in 1300MWe plants and a generic license to load all-M5 TM fuel in EDF N4 reactors and M5 TM fuel clad in 900MWe reactors for MOX parity fuel management. Licensing is also now underway in Switzerland, Finland, Brazil and Spain. The M5 TM alloy has demonstrated its superiority at burn-ups beyond current licensing limits, through operations in PWR at fuel rod burn-ups exceeding 71GWd/tU in the United States and 78GWd/tU in Europe. The Japanese nuclear industry has planned a stepwise approach to increase the burn-up of the fuel. Step-I fuel (48GWd/tU Fuel Assembly maximum burn-up) which was introduced in the late 80s. In the 90s started the licensing of the Step-II fuel (55GWd/tU Fuel Assembly maximum burn-up). Because the extension of the burn-up is important to reduce discharge fuel and cycle cost, the Japanese industry has plans to further extend the burn-up. In such burn-up region, fuel cladding with even better corrosion properties and very low hydrogen pick-up shall be necessary. M5 TM alloy, with high anticorrosion/hydriding properties, is suitable for not only the Step-II fuel

  17. 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

  18. 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)

  19. 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)

  20. Decay rate of reindeer pellet-groups

    Directory of Open Access Journals (Sweden)

    Anna Skarin

    2008-06-01

    Full Text Available Counting of animal faecal pellet groups to estimate habitat use and population densities is a well known method in wildlife research. Using pellet-group counts often require knowledge about the decay rate of the faeces. The decay rate of a faecal pellet group may be different depending on e.g. substrate, size of the pellet group and species. Pellet-group decay rates has been estimated for a number of wildlife species but never before for reindeer (Rangifer tarandus. During 2001 to 2005 a field experiment estimating the decay rate of reindeer pellet groups was performed in the Swedish mountains close to Ammarnäs. In total the decay rate of 382 pellet groups in three different habitat types (alpine heath, birch forest and spruce forest was estimated. The slowest decay rate was found in alpine heath and there the pellet groups persisted for at least four years. If decay was assumed to take place only during the bare ground season, the estimated exponential decay rate was -0.027 pellet groups/week in the same habitat. In the forest, the decay was faster and the pellet groups did not persist more than two years. Performing pellet group counts to estimate habitat use in dry habitats, such as alpine heath, I will recommend using the faecal standing crop method. Using this method makes it possible to catch the animals’ general habitat use over several years. Abstract in Swedish / Sammanfattning:Nedbrytningshastighet av renspillningInom viltforskningen har spillningsinventeringar använts under flera årtionden för att uppskatta habitatval och populationstäthet hos olika djurslag. För att kunna använda data från spillningsinventeringar krävs ofta att man vet hur lång tid det tar för spillningen att brytas ner. Nedbrytningshastigheten är olika beroende på marktyp och djurslag. Nedbrytningshastighet på spillning har studerats för bland annat olika typer av hjortdjur, men det har inte studerats på ren (Rangifer tarandus tidigare. I omr