WorldWideScience

Sample records for operation defective fuel

  1. Thermodynamic and kinetic modelling of fuel oxidation behaviour in operating defective fuel

    Lewis, B.J.; Thompson, W.T.; Akbari, F.; Thompson, D.M.; Thurgood, C.; Higgs, J.

    2004-01-01

    A theoretical treatment has been developed to predict the fuel oxidation behaviour in operating defective nuclear fuel elements. The equilibrium stoichiometry deviation in the hyper-stoichiometric fuel has been derived from thermodynamic considerations using a self-consistent set of thermodynamic properties for the U-O system, which emphasizes replication of solubilities and three-phase invariant conditions displayed in the U-O binary phase diagram. The kinetics model accounts for multi-phase transport including interstitial oxygen diffusion in the solid and gas-phase transport of hydrogen and steam in the fuel cracks. The fuel oxidation model is further coupled to a heat conduction model to account for the feedback effect of a reduced thermal conductivity in the hyper-stoichiometric fuel. A numerical solution has been developed using a finite-element technique with the FEMLAB software package. The model has been compared to available data from several in-reactor X-2 loop experiments with defective fuel conducted at the Chalk River Laboratories. The model has also been benchmarked against an O/U profile measurement for a spent defective fuel element discharged from a commercial reactor

  2. DIADEME: A computer code to assess in operation defective fuel characteristics and primary circuit contamination

    Genin, J.B. [DEN/DEC/S3C, CEA Cadarache, 13 - Saint-Paul-lez-Durance (France); Harrer, A. [EdF/SEPTEN, 69 - Villeurbanne (France); Musante, Y. [FRAMATOME-ANP, 69 - Lyon (France)

    2002-07-01

    DIADEME is a computer code developed within the framework of R and D cooperation between the French Atomic Energy Commission (CEA), Electricite de France (EdF) and FRAMATOME-ANP. Its aim is to assess in operation defective fuel characteristics and primary circuit contamination for actinides and long half-life fission products involved in health physics problems as well as in waste and decommissioning studies. DIADEME has been developed and qualified for the EDF nuclear power plants. For many years, both theoretical and experimental studies have been carried out at the CEA on the release of fission products and actinides out of defective fuel rods in operation, their migration and deposition in PWR primary circuits. These studies have allowed defect characteristic diagnosis methods to be developed, based on radiochemical measurements of the primary coolant. These methods are generally used along with gamma spectrometry measurements on primary water sampling. In order to be completely efficient, these methods can also be used in connection with an on-line primary water gamma spectrometry device. This permits to obtain the most comprehensive data on fission product activity evolutions at steady state and during operation transients, and allows the on-line characterization of the defective fuel assemblies. For long half-life fission products and for actinides, DIADEME is also able to assess the activities of soluble and insoluble forms in the primary water and in the chemical and voluminal control system (CVCS) filters and resins, as well as those activities deposited on primary circuit surfaces. (author)

  3. DIADEME: A computer code to assess in operation defective fuel characteristics and primary circuit contamination

    Genin, J.B.; Harrer, A.; Musante, Y.

    2002-01-01

    DIADEME is a computer code developed within the framework of R and D cooperation between the French Atomic Energy Commission (CEA), Electricite de France (EdF) and FRAMATOME-ANP. Its aim is to assess in operation defective fuel characteristics and primary circuit contamination for actinides and long half-life fission products involved in health physics problems as well as in waste and decommissioning studies. DIADEME has been developed and qualified for the EDF nuclear power plants. For many years, both theoretical and experimental studies have been carried out at the CEA on the release of fission products and actinides out of defective fuel rods in operation, their migration and deposition in PWR primary circuits. These studies have allowed defect characteristic diagnosis methods to be developed, based on radiochemical measurements of the primary coolant. These methods are generally used along with gamma spectrometry measurements on primary water sampling. In order to be completely efficient, these methods can also be used in connection with an on-line primary water gamma spectrometry device. This permits to obtain the most comprehensive data on fission product activity evolutions at steady state and during operation transients, and allows the on-line characterization of the defective fuel assemblies. For long half-life fission products and for actinides, DIADEME is also able to assess the activities of soluble and insoluble forms in the primary water and in the chemical and voluminal control system (CVCS) filters and resins, as well as those activities deposited on primary circuit surfaces. (author)

  4. Areva solutions for management of defective fuel

    Morlaes, I.; Vo Van, V.

    2014-01-01

    Defective fuel management is a major challenge for nuclear operators when all fuel must be long-term managed. This paper describes AREVA solutions for managing defective fuel. Transport AREVA performs shipments of defective fuel in Europe and proposes casks that are licensed for that purpose in Europe and in the USA. The paper presents the transport experience and the new European licensing approach of defective fuel transport. Dry Interim Storage AREVA is implementing the defective fuel storage in the USA, compliant with the Safety Authority's requirements. In Europe, AREVA is developing a new, more long-term oriented storage solution for defective fuel, the best available technology regarding safety requirements. The paper describes these storage solutions. Treatment Various types of defective fuel coming from around the world have been treated in the AREVA La Hague plant. Specific treatment procedures were developed when needed. The paper presents operational elements related to this experience. (authors)

  5. Impacts of reactor. Induced cladding defects on spent fuel storage

    Johnson, A.B.

    1978-01-01

    Defects arise in the fuel cladding on a small fraction of fuel rods during irradiation in water-cooled power reactors. Defects from mechanical damage in fuel handling and shipping have been almost negligible. No commercial water reactor fuel has yet been observed to develop defects while stored in spent fuel pools. In some pools, defective fuel is placed in closed canisters as it is removed from the reactor. However, hundreds of defective fuel bundles are stored in numerous pools on the same basis as intact fuel. Radioactive species carried into the pool from the reactor coolant must be dealt with by the pool purification system. However, additional radiation releases from the defective fuel during storage appear tu be minimal, with the possible exception of fuel discharged while the reactor is operating (CANDU fuel). Over approximately two decades, defective commercial fuel has been handled, stored, shipped and reprocessed. (author)

  6. A conceptual model for the fuel oxidation of defective fuel

    Higgs, J.D.; Lewis, B.J.; Thompson, W.T.; He, Z.

    2007-01-01

    A mechanistic conceptual model has been developed to predict the fuel oxidation behaviour in operating defective fuel elements for water-cooled nuclear reactors. This theoretical work accounts for gas-phase transport and sheath reactions in the fuel-to-sheath gap to determine the local oxygen potential. An improved thermodynamic analysis has also been incorporated into the model to describe the equilibrium state of the oxidized fuel. The fuel oxidation kinetics treatment accounts for multi-phase transport including normal diffusion and thermodiffusion for interstitial oxygen migration in the solid, as well as gas-phase transport in the fuel pellet cracks. The fuel oxidation treatment is further coupled to a heat conduction equation. A numerical solution of the coupled transport equations is obtained by a finite-element technique with the FEMLAB 3.1 software package. The model is able to provide radial-axial profiles of the oxygen-to-uranium ratio and the fuel temperatures as a function of time in the defective element for a wide range of element powers and defect sizes. The model results are assessed against coulometric titration measurements of the oxygen-to-metal profile for pellet samples taken from ten spent defective elements discharged from the National Research Universal Reactor at the Chalk River Laboratories and commercial reactors

  7. Behaviour of defective CANDU fuel: fuel oxidation kinetic and thermodynamic modelling

    Higgs, J.

    2005-01-01

    The thermal performance of operating CANDU fuel under defect conditions is affected by the ingress of heavy water into the fuel element. A mechanistic model has been developed to predict the extent of fuel oxidation in defective fuel and its affect on fuel thermal performance. A thermodynamic treatment of such oxidized fuel has been performed as a basis for the boundary conditions in the kinetic model. Both the kinetic and thermodynamic models have been benchmarked against recent experimental work. (author)

  8. Apparatus for locating defective nuclear fuel elements

    Lawrie, W.E.

    1979-01-01

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

  9. Physical models and codes for prediction of activity release from defective fuel rods under operation conditions and in leakage tests during refuelling

    Likhanskii, V.; Evdokimov, I.; Khoruzhii, O.; Sorokin, A.; Novikov, V.

    2003-01-01

    It is appropriate to use the dependences, based on physical models, in the design-analytical codes for improving of reliability of defective fuel rod detection and for determination of defect characteristics by activity measuring in the primary coolant. In the paper the results on development of some physical models and integral mechanistic codes, assigned for prediction of defective fuel rod behaviour are presented. The analysis of mass transfer and mass exchange between fuel rod and coolant showed that the rates of these processes depends on many factors, such as coolant turbulent flow, pressure, effective hydraulic diameter of defect, fuel rod geometric parameters. The models, which describe these dependences, have been created. The models of thermomechanical fuel behaviour, stable gaseous FP release were modified and new computer code RTOP-CA was created thereupon for description of defective fuel rod behaviour and activity release into the primary coolant. The model of fuel oxidation in in-pile conditions, which includes radiolysis and RTOP-LT after validation of physical models are planned to be used for prediction of defective fuel rods behaviour

  10. Detection of defective fuel rods in water reactors - a review

    Hartog, J.M.

    1980-01-01

    Consideration of the fundamental processes of fission product release within fuel pellets and at the pellet surface, and its transport in the fuel/cladding interspace and from fuel rod to coolant, indicates what radio-nuclides will be detectable in the coolant from small and large cladding failures. A better understanding of the aggregate fission product transport is required to allow reactor operators to interpret signals from detection systems in terms of quantitative cladding deterioration. This needs experimental investigation in a specially instrumented loop, as well as development of a technique to cause a rod to defect deliberately during steady power operation. (author)

  11. Fuel defect detection, localization and removal in Bruce Power units 3 through 8

    Stone, R.; Armstrong, J.; Iglesias, F.; Oduntan, R.; Lewis, B.

    2005-01-01

    Fuel element defects are occurring in Bruce 'A' and Bruce 'B' Units. A root-cause investigation is ongoing, however, a solution is not yet in-hand. Fuel defect management efforts have been undertaken, therefore, in the interim. Fuel defect management tools are in-place for all Bruce Units. These tools can be categorized as analysis-based or operations-based. Analysis-based tools include computer codes used primarily for fuel defect characterization, while operations-based tools include Unit-specific delayed-neutron ('DN') monitoring systems and gaseous fission product ('GFP') monitoring systems. Operations-based tools are used for fuel defect detection, localization and removal activities. Fuel and Physics staff use defect detection, localization and removal methodologies and guidelines to disposition fuel defects. Methodologies are 'standardized' or 'routine' procedures for implementing analysis-based and operations-based tools to disposition fuel defects during Unit start-up operation and during operation at high steady-state power levels. Guidelines at present serve to supplement fuel defect management methodologies during Unit power raise. (author)

  12. Fuel cracking in relation to fuel oxidation in support of an out-reactor instrumented defected fuel experiment

    Quastel, A.; Thiriet, C. [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Lewis, B., E-mail: brent.lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada); Corcoran, E., E-mail: emily.corcoran@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

    2014-07-01

    An experimental program funded by the CANDU Owners Group (COG) is studying an out-reactor instrumented defected fuel experiment in Stern Laboratories (Hamilton, Ontario) with guidance from Atomic Energy of Canada Limited (AECL). The objective of this test is to provide experimental data for validation of a mechanistic fuel oxidation model. In this experiment a defected fuel element with UO{sub 2} pellets will be internally heated with an electrical heater element, causing the fuel to crack. By defecting the sheath in-situ the fuel will be exposed to light water coolant near normal reactor operating conditions (pressure 10 MPa and temperature 265-310{sup o}C) causing fuel oxidation, especially near the hotter regions of the fuel in the cracks. The fuel thermal conductivity will change, resulting in a change in the temperature distribution of the fuel element. This paper provides 2D r-θ plane strain solid mechanics models to simulate fuel thermal expansion, where conditions for fuel crack propagation are investigated with the thermal J integral to predict fuel crack stress intensity factors. Finally since fuel crack geometry can affect fuel oxidation this paper shows that the solid mechanics model with pre-set radial cracks can be coupled to a 2D r-θ fuel oxidation model. (author)

  13. Canadian CANDU fuel development program and recent fuel operating experience

    Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

    1999-01-01

    This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the 2-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These 2 programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

  14. Defect networks and supersymmetric loop operators

    Bullimore, Mathew [Perimeter Institute for Theoretical Physics, 31 Caroline Street North, Waterloo, ON N2L 2Y5 (Canada)

    2015-02-10

    We consider topological defect networks with junctions in A{sub N−1} Toda CFT and the connection to supersymmetric loop operators in N=2 theories of class S on a four-sphere. Correlation functions in the presence of topological defect networks are computed by exploiting the monodromy of conformal blocks, generalising the notion of a Verlinde operator. Concentrating on a class of topological defects in A{sub 2} Toda theory, we find that the Verlinde operators generate an algebra whose structure is determined by a set of generalised skein relations that encode the representation theory of a quantum group. In the second half of the paper, we explore the dictionary between topological defect networks and supersymmetric loop operators in the N=2{sup ∗} theory by comparing to exact localisation computations. In this context, the the generalised skein relations are related to the operator product expansion of loop operators.

  15. Radioactivity of long-lived nuclides in the primary circuit of the reactor BOR-60 during operation with defective fuel elements

    Gryazev, V.M.; Kizin, V.D.; Lisitsyn, E.S.; Polyakov, V.I.; Chechetkin, Y.V.

    1978-06-01

    The summarized results of measurements of the enrichment and distribution of radioactive nuclides from corrosion and of fission products during the four years of operation of BOR-60, including a longer period of operation with detective fuel elements in the core, are presented. It is shown that for operation with approximately 1% leaking fuel rods radiation exposure becomes worse manily because of release and enrichment of cesium isotopes in the coolant. Of the other fission products, the largest contribution to the dose rate in pipework and components is given by 140 Ba / 140 La and 95 Nb. On operation with 0.1 to 0.2% of leaking fuel rods, this contribution is comparable to that of the corrosion products 60 Co and 54 Mn. The radioactivity of corrosion products in the circuit has not increased within the last three years and was about one order of magnitude lower than the theoretical values. The corrosion and fission products are nonuniformly distributed over the circuit. Concentration of 95 Nb and 60 Co in the pipe for 'cold' sodium is larger by a factor of 2 - 5 and of 140 Ba and 54 Mn by a factor of 10-20 than in the pipes for 'hot' sodium. Most of the cobalt was found to deposit in the heat exchanges. The effectiveness of emptying the pipes from coolant in order to reduce the dose sate is assessed. (orig.) [de

  16. Experimental study of defect power reactor fuel. Final report

    Forsyth, R.S.; Jonsson, T.

    1982-01-01

    Two BWR fuel rods, one intact and one defect, with the same manufacturing and irradiation data have been examined in a comparative study. The defect rod has been irradiated in a defect condition during approximately one reactor cycle and has consequently some secondary defects. The defect rod has two penetrating defects at a distance of about 1.5 meters from each other. Comparison with the intact rod shows a large Cs loss from the defect rod, especially between the cladding defects, where the loss is measured to about 30 %. The leachibility in deionized water is higher for Cs, U and Cm for fuel from the defect rod. The leaching results are more complex for Sr-90, Pu and Am. The fuel in the defect rod has undergone a change of structure with gain growth and formation of oriented fuel structure. The cladding of the defect rod is hydrided locally in some parts of the lower part of the rod and furthermore over a more extended region near the end of the rod. (Authors)

  17. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    Clayton, J.C.

    1987-10-01

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated

  18. Fission product release from defected nuclear reactor fuel elements

    Lewis, B.J.

    1983-01-01

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

  19. Prediction of power-ramp defects in CANDU fuel

    Gillespie, P.; Wadsworth, S.; Daniels, T.

    2010-01-01

    Power ramps result in fuel pellet expansion and can lead to fuel sheath failures by fission product induced stress corrosion cracking (SCC). Historically, empirical models fit to experimental test data were used to predict the onset of power-ramp failures in CANDU fuel. In 1988, a power-ramped fuel defect event at PNGS-1 led to the refinement of these empirical models. This defect event has recently been re-analyzed and the empirical model updated. The empirical model is supported by a physically based model which can be used to extrapolate to fuel conditions (density, burnup) outside of the 1988 data set. (author)

  20. Method for the detection of defective nuclear fuel assemblies

    Lawrie, W.E.; Womack, R.E.; White, N.W. Jr.

    1978-01-01

    There is applied an ultrasonic transmitter on a tape carrier by means of which the ultrasonic transmitter can be guided underwater between the fuel assemblies. If a fuel assembly is defective, i.e. filled with water, the reflection coefficient at the front interface between cladding and inner space of the fuel assembly will decrease. Essential parts of the ultrasonic signal will move through the liquid and will not be reflected until the backward liquid/metal interface of the fuel assembly. This impulse echo is different from that of the gas-filled fuel assembly. (DG) [de

  1. Computed tomography on a defective CANDU fuel pencil end cap

    Lupton, L.R.

    1985-09-01

    Five tomographic slices through a defective end cap from a CANDU fuel pencil have been generated using a Co-60 source and a first generation translate-rotate tomography scanner. An anomaly in the density distribution that is believed to have resulted from the defect has been observed. However, with the 0.30 mm spatial resolution used, it has not been possible to state unequivocally whether the change in density is caused by a defect in the weld or a statistical anomaly in the data. It is concluded that a microtomography system, with a spatial resolution in the range of 0.1 mm, could detect the flaw

  2. Transport of volatile fission products in the fuel-to-sheath gap of defective fuel elements during normal and reactor accident conditions

    Lewis, B.J.; Bonin, H.W.

    1995-01-01

    An analytical treatment has been used to model the vapour transport of radioactive fission products released into the fuel-to-sheath gap of defective nuclear fuel elements. The model accounts for both diffusive and bulk-convective transport. Convective transport becomes important as a result of a significant release of gaseous fission products into the gap during a high-temperature reactor accident. However, during normal reactor operation, diffusion is shown to be the dominant process of transport. The model is based on an analysis of several in-reactor tests with operating defective fuel elements, and high-temperature annealing experiments with irradiated fuel specimens. ((orig.))

  3. A semi-mechanistic approach to calculate the probability of fuel defects

    Tayal, M.; Millen, E.; Sejnoha, R.

    1992-10-01

    In this paper the authors describe the status of a semi-mechanistic approach to the calculation of the probability of fuel defects. This approach expresses the defect probability in terms of fundamental parameters such as local stresses, local strains, and fission product concentration. The calculations of defect probability continue to reflect the influences of the conventional parameters like power ramp, burnup and CANLUB. In addition, the new approach provides a mechanism to account for the impacts of additional factors involving detailed fuel design and reactor operation, for example pellet density, pellet shape and size, sheath diameter and thickness, pellet/sheath clearance, and coolant temperature and pressure. The approach has been validated against a previous empirical correlation. AN illustrative example shows how the defect thresholds are influenced by changes in the internal design of the element and in the coolant pressure. (Author) (7 figs., tab., 12 refs.)

  4. Apparatus and method for grounding compressed fuel fueling operator

    Cohen, Joseph Perry; Farese, David John; Xu, Jianguo

    2002-06-11

    A safety system for grounding an operator at a fueling station prior to removing a fuel fill nozzle from a fuel tank upon completion of a fuel filling operation is provided which includes a fuel tank port in communication with the fuel tank for receiving and retaining the nozzle during the fuel filling operation and a grounding device adjacent to the fuel tank port which includes a grounding switch having a contact member that receives physical contact by the operator and where physical contact of the contact member activates the grounding switch. A releasable interlock is included that provides a lock position wherein the nozzle is locked into the port upon insertion of the nozzle into the port and a release position wherein the nozzle is releasable from the port upon completion of the fuel filling operation and after physical contact of the contact member is accomplished.

  5. Performance of artificially defected LWR fuel rods in an unlimited air dry storage atmosphere

    Einziger, R.E.; Knecht, R.L.; Cantley, D.A.; Cook, J.A.

    1983-09-01

    Thus far the tests are inconclusive as to whether breached LWR fuel can be stored at 230 0 C for long periods of time in air without fuel oxidation and dispersion. There is every indication, as expected, that there is no oxidation problem in an inert atmosphere. Only one of four defects exposed to unlimited air gave any indication of fuel oxidation. It has been suggested that this might be an incubation effect and continued operation would result in oxidation occurring at all four defects. As yet the destructive examination of the BWR rod has not been completed, so it is not possible to determine if cladding splitting was due to an anomoly in this test rod or something that can be expected in LWR rods in general. Thus far there is no indication of respirable particle dispersal even if fuel oxidation does occur

  6. Performance of water cooled nuclear power reactor fuels in India – Defects, failures and their mitigation

    Ganguly, Chaitanyamoy

    2015-01-01

    Water cooled and moderated nuclear power reactors account for more than 95% of the operating reactors in the world today. Light water reactors (LWRs) consisting of pressurized water reactor (PWR), their Russian counterpart namely VVER and boiling water reactor (BWR) will continue to dominate the nuclear power market. Pressurized heavy water reactor (PHWR), also known as CANDU, is the backbone of the nuclear power program in India. Updates on LWR and PHWR fuel performance are being periodically published by IAEA, OECD-NEA and the World Nuclear Association (WNA), highlighting fuel failure rate and the mitigation of fuel defects and failures. These reports clearly indicate that there has been significant improvement in in – pile fuel performance over the years and the present focus is to achieve zero fuel failure in high burn up and high performance fuels. The present paper summarizes the status of PHWR and LWR fuel performance in India, highlighting the manufacturing and the related quality control and inspection steps that are being followed at the PHWR fuel fabrication plant in order to achieve zero manufacturing defect which could contribute to achieving zero in – pile failure rate in operating and upcoming PHWR units in India. (author)

  7. Short, medium and long term consequences of inadequate defect fuel management

    Roberts, J.G.; McQueen, M.; Nashiem, R.; Ma, G.

    2011-01-01

    Defect fuel pencils result in short, medium and long term consequences to the environment within and external to the nuclear power station. The paper will describe these consequences and specify the Defect Fuel Management Practices required to avoid these consequences.

  8. Short, medium and long term consequences of inadequate defect fuel management

    Roberts, J.G. [CANTECH Associates Ltd., Burlington, ON (Canada); McQueen, M.; Nashiem, R.; Ma, G. [Bruce Power, Tiverton, ON (Canada)

    2011-07-01

    Defect fuel pencils result in short, medium and long term consequences to the environment within and external to the nuclear power station. The paper will describe these consequences and specify the Defect Fuel Management Practices required to avoid these consequences.

  9. The Canadian CANDU fuel development program and recent fuel operating experience

    Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

    1999-01-01

    This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the two-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These two programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

  10. Process and equipment for locating defective fuel rods of a reactor fuel element

    Jester, A.; Honig, H.

    1977-01-01

    By this equipment, well-known processes for determining defective fuel rods of a reactor fuel element are improved in such a fashion that defective fuel rods can be located individually, so that it is possible to replace them. The equipment consists of a cylindrical test vessel open above, which accommodates the element to be tested, so that an annular space is left between the latter's external circumference and the wall of the vessel, and so that the fuel rods project above the vessel. A bell in the shape of a frustrum of a cone is inverted over the test vessel, which has an infra-red measuring equipment at a certain distance above the tops of the fuel rods. The fuel element to be tested together with the test vessel and hood are immersed in a basin full of water, which displaces water by means of gas from the hood. The post-shutdown heat increases the temperature in the water space of the test vessel, which is stabilised at 100 0 C. In each defective fuel rod the water which has penetrated the defective fuel rod previously, or does so now, starts to boil. The steam rising in the fuel rod raises the temperature of the defective fuel rod compared to all the sound ones. The subsequent measurement easily determines this. Where one can expect interference with the measurement by appreciable amounts of gamma rays, the measuring equipment is removed from the path of radiation by mirror deflection in a suitably shaped measuring hood. (FW) [de

  11. On-line defected fuel monitoring using GFP data

    Livingstone, S.; Lewis, B.J.

    2008-01-01

    This paper describes the initial development of an on-line defected fuel diagnostic tool. The tool is based on coolant activity, and uses a quantitative and qualitative approach from existing mechanistic fission product release models, and also empirical rules based on commercial and experimental experience. The model departs from the usual methodology of analyzing steady-state fission product coolant activities, and instead uses steady-state fission product release rates calculated from the transient coolant activity data. An example of real-time defected fuel analysis work is presented using a prototype of this tool with station data. The model is in an early developmental stage, and this paper demonstrates the promising potential of this technique. (author)

  12. Detection and defect correction of operating process

    Vasendina, Elena; Plotnikova, Inna; Levitskaya, Anastasiya; Kvesko, Svetlana

    2016-01-01

    The article is devoted to the current problem of enterprise competitiveness rise in hard and competitive terms of business environment. The importance of modern equipment for detection of defects and their correction is explained. Production of chipboard is used as an object of research. Short description and main results of estimation efficiency of innovative solutions of enterprises are considered. (paper)

  13. Irradiation of defected SAP clad UO2 fuel in the X-7 organic loop

    Robertson, R.F.S.; Cracknell, A.G.; MacDonald, R.D.

    1961-10-01

    This report describes an experiment designed to test the behaviour under irradiation of a UO 2 fuel specimen clad in a defected SAP sheath and cooled by recirculating organic liquid. The specimen containing the defect was irradiated in the X-7 loop in the NRX reactor from the 25th of November until the 13th of December 1960. Up to the 13th of December the behaviour was analogous to that seen with defected UO 2 specimens clad in zircaloy which were irradiated in water loops. Reactor power transients resulted in peaking of gamma ray activities in the loop, but on steady operation these activities tended to fall to a steady state level, Over this period the pressure drop across the fuel increased by a factor of two, the increases occurring after reactor shut downs and start ups. On 13th December the pressure drop increased rapidly, after a reactor shut down and start up, to over five times its original value and the activities in the loop rose to a high level. The specimen was removed and examination showed that the sheath was very badly split and that the volume between the fuel and the sheath was filled with a hard black organic substance. This report gives full details of the irradiation and of the post -irradiation examination. Correlation of the observed phenomenon is attempted and a preliminary assessment of the problems which would be associated with defect fuel in an organic reactor is given. (author)

  14. A device for tracking-down the defective fuel rods in a reactor

    Preda, Marin; Ciocanescu, Marin; Barbos, Dumitru; Rogociu, Ioan

    2008-01-01

    The paper gives first the fuel element description and its operation. If a cladding defect arises, some of the fission isotopes pass into the primary cooling system and, as these isotopes are extremely radio-active, the danger of primary cooling system contamination occurs what entails expensive decontamination operations. For identification of the bundle containing the defective pins a simple, modular device was designed and made. It works by pointing-out the bundle(s) which has at least one defective fuel pin. After tracking, the fuel bundle is picked-up from the core and searching is continued to point-out the defective pin inside post-irradiation-hot cells. For dosimetric survey in the reactor hall, an aerosol detector was used. When an accident arises the released noble gases will be detected by this detector. The detector can give no information where the damage is located for one of the fuel pins inside the irradiation devices (loop or capsule) can also get defective and consequently it can release radioactive noble gases in the reactor hall. For avoiding this a radioactive survey device for core cooling agent was mounted by the primary cooling system. The device for defective fuel rod identification in the nuclear reactor is composed of the following components: - a device for water sampling from the fuel bundle; - a suction valve; - a handling tool; - an electric pump; - ionic filters; - a flexible hose. When fission isotopes arise in primary cooling system, the device is brought to the edge of the reactor pool in a sharp positioning. By means of the handling tool the sampling device is inserted at the top of the fuel bundle. The suction inlet circuit and the electric pump are filled with pool water, and after that the ionic filter and outlet circuit are filled also. The electric pump is actuated and the following circuit is operated: fuel bundle, electric pump, ionic filter, pool. For avoiding the overheating of the pump, part of the flow is by

  15. A study on the behavior of defected LWR spent fuel

    You, Gil Sung; Kim, Eun Ka; Kim, Keon Sik; Suh, Hang Suck; Kim, Seung Jung; Ro, Seung Gy; Park, Chong Mook; Ji, Pyung Gook

    1992-03-01

    To investigate the storage behavior of the defective LWR spent fuel rods, the characteristic changes of fuel and cladding are to be measured and analyzed. In addition, the oxidation study in air on non-irradiated and irradiated U0 2 was performed. No changes were observed in the tested fuel rods after 30 month storage. The Cs-134, 137 released rapidly during the initial 3 months of storage, but remained in constant value after 3 month storage and the release was almost ceased after 30 month storage. The weight gain of non-irradiated U0 2 samples showed a trend of S type curves and the activation energies were 11OKJ/mol above 350 deg C. and 143KJ/mol below 350 deg C. But irradiated U0 2 showed a rapid increase at initial stage of oxidation and a decrease at later stage when compared with the results of non-irradiated U0 2 . (Author)

  16. Canadian fuel development program and recent operational experience

    Cox, D.S.; Kohn, E.; Lau, J.H.K.; Dicke, G.J.; Macici, N.N.; Sancton, R.W.

    1995-01-01

    This paper provides an overview of the current Canadian CANDU fuel R and D programs and operational experience. The details of operational experience for fuel in Canadian reactors are summarized for the period 1991-1994; excellent fuel performance has been sustained, with steady-state bundle defect rates currently as low as 0.02%. The status of introducing long 37-element bundles, and bundles with rounded bearing pads is reviewed. These minor changes in fuel design have been selectively introduced in response to operational constraints (end-plate cracking and pressure-tube fretting) at Ontario Hydro's Bruce-B and Darlington stations. The R and D programs are generating a more complete understanding of CANDU fuel behaviour, while the CANDU Owners Group (COG) Fuel Technology Program is being re-aligned to a more exclusive focus on the needs of operating stations. Technical highlights and realized benefits from the COG program are summarized. Re-organization of AECL to provide a one-company focus, with an outward looking view to new CANDU markets, has strengthened R and D in advanced fuel cycles. Progress in AECL's key fuel cycle programs is also summarized. (author)

  17. Secondary hydriding of defected zircaloy-clad fuel rods

    Olander, D.R.; Vaknin, S.

    1993-01-01

    The phenomenon of secondary hydriding in LWR fuel rods is critically reviewed. The current understanding of the process is summarized with emphasis on the sources of hydrogen in the rod provided by chemical reaction of water (steam) introduced via a primary defect in the cladding. As often noted in the literature, the role of hydrogen peroxide produced by steam radiolysis is to provide sources of hydrogen by cladding and fuel oxidation that are absent without fission-fragment irradiation of the gas. Quantitative description of the evolution of the chemical state inside the fuel rod is achieved by combining the chemical kinetics of the reactions between the gas and the fuel and cladding with the transport by diffusion of components of the gas in the gap. The chemistry-gas transport model provides the framework into which therate constants of the reactions between the gases in the gap and the fuel and cladding are incorporated. The output of the model calculation is the H 2 0/H 2 ratio in the gas and the degree of claddingand fuel oxidation as functions of distance from the primary defect. This output, when combined with a criterion for the onset of massive hydriding of the cladding, can provide a prediction of the time and location of a potential secondary hydriding failure. The chemistry-gas transport model is the starting point for mechanical and H-in-Zr migration analyses intended to determine the nature of the cladding failure caused by the development of the massive hydride on the inner wall

  18. Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

    Ham, Y.S.; Kerr, P.; Sitaraman, S.; Swan, R. [Global Security Directorate, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Rossa, R. [SCK-CEN, Mol (Belgium); Liljenfeldt, H. [SKB in Oskarshamn (Sweden)

    2015-07-01

    The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)

  19. Fuel performance and operation experience of WWER-440 fuel in improved fuel cycle

    Gagarinski, A.; Proselkov, V.; Semchenkov, Yu.

    2007-01-01

    The paper summarizes WWER-440 second-generation fuel operation experience in improved fuel cycles using the example of Kola NPP units 3 and 4. Basic parameters of fuel assemblies, fuel rods and uranium-gadolinium fuel rods, as well as the principal neutronic parameters and burn-up achieved in fuel assemblies are presented. The paper also contains some data concerning the activity of coolant during operation (Authors)

  20. Monitoring for fuel sheath defects in three shipments of irradiated CANDU nuclear fuel

    Johnson, H.M.

    1978-01-01

    Analyses of radioactive gases within the Pegase shipping flask were performed at the outset and at the completion of three shipments of irradiated nuclear fuel from the Douglas Point Generating Station to Whiteshell Nuclear Research Establishment. No increases in the concentration of active gases, volatiles or particulates were observed. The activity of the WR-1 bay water rose only marginally due to the storage of the fuel. Other tests indicated that minimal surface contamination was present. These data established that defects in fuel element sheaths did not arise during the transport or the handling of this irradiated fuel. The observation has significance for the prospect of irradiated nuclear fuel transfer and handling in preparation for storage or disposal. (author)

  1. The cost of operating with failed fuel at Virginia power

    Ford, C.A.

    1988-01-01

    Virginia Power has completed a study of the costs incurred due to fuel failures in its pressurized water reactors. This study was prompted by histories of high primary coolant activity and subsequent fuel inspections at the North Anna and Surry power stations. The study included an evaluation of the total costs of fuel failures as well as an evaluation of the economics of postirradiation fuel inspections. The major costs of fuel failures included personnel radiation exposure, permanently discharged failed fuel, radwaste generation, increased labor requirements, containment entry delays due to airborne radioactivity, and ramp rate restrictions. Although fuel failures affect a utility in several other areas, the items evaluated in the study were thought to be the most significant of the costs. The study indicated that performing a postirradiation failed fuel examination can be economically justified at tramp-corrected 131 I levels of > 0.015 μCi/g. The savings to the utility can be on the order of several million dollars. Additionally, the cost penalty of performing a fuel inspection at lower iodine levels is generally in the range of $200,000. This economic penalty is expected to be outweighed by the intangible benefits of operating with a defect-free core

  2. LOFT instrumented fuel design and operating experience

    Russell, M.L.

    1979-01-01

    A summary description of the Loss-of-Fluid Test (LOFT) system instrumented core construction details and operating experience through reactor startup and loss-of-coolant experiment (LOCE) operations performed to date are discussed. The discussion includes details of the test instrumentation attachment to the fuel assembly, the structural response of the fuel modules to the forces generated by a double-ended break of a pressurized water reactor (PWR) coolant pipe at the inlet to the reactor vessel, the durability of the LOFT fuel and test instrumentation, and the plans for incorporation of improved fuel assembly test instrumentation features in the LOFT core

  3. Operating a fuel cell using landfill gas

    Trippel, C.E.; Preston, J.L. Jr.; Trocciola, J.; Spiegel, R.

    1996-12-31

    An ONSI PC25{trademark}, 200 kW (nominal capacity) phosphoric acid fuel cell operating on landfill gas is installed at the Town of Groton Flanders Road landfill in Groton, Connecticut. This joint project by the Connecticut Light & Power Company (CL&P) which is an operating company of Northeast Utilities, the Town of Groton, International Fuel Cells (IFC), and the US EPA is intended to demonstrate the viability of installing, operating and maintaining a fuel cell operating on landfill gas at a landfill site. The goals of the project are to evaluate the fuel cell and gas pretreatment unit operation, test modifications to simplify the GPU design and demonstrate reliability of the entire system.

  4. Reliabilityy and operating margins of LWR fuels

    Strasser, A.A.; Lindquist, K.O.

    1977-01-01

    The margins to fuel thermal operating limits under normal and accident conditions are key to plant operating flexibility and impact on availability and capacity factor. Fuel performance problems that do not result in clad breach, can reduce these margins. However, most have or can be solved with design changes. Regulatory changes have been major factors in eroding these margins. Various methods for regaining the margins are discussed

  5. Short, medium and long term consequences of inadequate defect fuel management

    Roberts, J.G. [CANTECH Associates Limited, Burlington, Ontario (Canada); Nashiem, R.; McQueen, M.; Ma, G. [Bruce Power, Tiverton, Ontario (Canada)

    2011-07-01

    Defect fuel pencils result in short, medium and long term consequences to the environment within and external to the nuclear power station. The paper will describe these consequences and specify the Defect Fuel Management Practices required to avoid these consequences. (author)

  6. Short, medium and long term consequences of inadequate defect fuel management

    Roberts, J.G.; Nashiem, R.; McQueen, M.; Ma, G.

    2011-01-01

    Defect fuel pencils result in short, medium and long term consequences to the environment within and external to the nuclear power station. The paper will describe these consequences and specify the Defect Fuel Management Practices required to avoid these consequences. (author)

  7. Short, medium and long term consequences of inadequate defect fuel management

    Roberts, J.G., E-mail: alchemy@tnt21.com [CANTECH Associates Ltd., Burlington, Ontario (Canada); Nashiem, R.; McQueen, M.; Ma, G., E-mail: Rod.nashiem@brucepower.com, E-mail: Maureen.mcqueen@brucepower.com, E-mail: guoping.ma@brucepower.com [Bruce Power, Tiverton (Canada)

    2010-07-01

    Defect fuel pencils result in short, medium and long term consequences to the environment within and external to the nuclear power station. The paper will describe these consequences and specify the Defect Fuel Management Practices required to avoid these consequences. (author)

  8. Short, medium and long term consequences of inadequate defect fuel management

    Roberts, J.G.; Nashiem, R.; McQueen, M.; Ma, G.

    2010-01-01

    Defect fuel pencils result in short, medium and long term consequences to the environment within and external to the nuclear power station. The paper will describe these consequences and specify the Defect Fuel Management Practices required to avoid these consequences. (author)

  9. Factors affecting defective fraction of biso-coated HTGR fuel particles during in-block carbonization

    Caputo, A.J.; Johnson, D.R.; Bayne, C.K.

    1977-01-01

    The performance of Biso-coated thoria fuel particles during the in-block processing step of HTGR fuel element refabrication was evaluated. The effect of various process variables (heating rate, particle crushing strength, horizontal and/or vertical position in the fuel element blocks, and fuel hole permeability) on pitch coke yield, defective fraction of fuel particles, matrix structure, and matrix porosity was evaluated. Of the variables tested, only heating rate had a significant effect on pitch coke yield while both heating rate and particle crushing strength had a significant effect on defective fraction of fuel particles

  10. Operations monitoring concept. Consolidated Fuel Reprocessing Program

    Kerr, H.T.

    1985-01-01

    Operations monitoring is a safeguards concept which could be applied in future fuel cycle facilities to significantly enhance the effectiveness of an integrated safeguards system. In general, a variety of operations monitoring techniques could be developed for both international and domestic safeguards application. The goal of this presentation is to describe specific examples of operations monitoring techniques as may be applied in a fuel reprocessing facility. The operations monitoring concept involves monitoring certain in-plant equipment, personnel, and materials to detect conditions indicative of the diversion of nuclear material. An operations monitoring subsystem should be designed to monitor operations only to the extent necessary to achieve specified safeguards objectives; there is no intent to monitor all operations in the facility. The objectives of the operations monitoring subsystem include: verification of reported data; detection of undeclared uses of equipment; and alerting the inspector to potential diversion activities. 1 fig

  11. Spent Nuclear Fuel Project operational staffing plan

    Debban, B.L.

    1996-03-01

    Using the Spent Nuclear Fuel (SNF) Project's current process flow concepts and knowledge from cognizant engineering and operational personnel, an initial assessment of the SNF Project radiological exposure and resource requirements was completed. A small project team completed a step by step analysis of fuel movement in the K Basins to the new interim storage location, the Canister Storage Building (CSB). This analysis looked at fuel retrieval, conditioning of the fuel, and transportation of the fuel. This plan describes the staffing structure for fuel processing, fuel movement, and the maintenance and operation (M ampersand O) staffing requirements of the facilities. This initial draft does not identify the support function resources required for M ampersand O, i.e., administrative and engineering (technical support). These will be included in future revisions to the plan. This plan looks at the resource requirements for the SNF subprojects, specifically, the operations of the facilities, balances resources where applicable, rotates crews where applicable, and attempts to use individuals in multi-task assignments. This plan does not apply to the construction phase of planned projects that affect staffing levels of K Basins

  12. Operating a locomotive on liquid methane fuel

    Stolz, J.L.

    1992-01-01

    This paper reports that several years ago, Burlington Northern Railroad looked into the feasibility of operating a diesel railroad locomotive to also run on compressed natural gas in a dual-fuel mode. Recognizing the large volume of on-board storage required and other limitations of CNG in the application, a program was begun to fuel a locomotive with liquefied natural gas. Because natural gas composition can vary with source and processing, it was considered desirable to use essentially pure liquid methane as the engine fuel. Initial testing results show the locomotive system achieved full diesel-rated power when operating on liquid methane and with equivalent fuel efficiency. Extended testing, including an American Association of Railroad 500-hour durability test, was undertaken to obtain information on engine life, wear rate and lubrication oil life

  13. Nuclear fuel operation at Balakovo NPP

    Morozov, A.

    2015-01-01

    The presentation addressed the positive experience of the TVS-2M assemblies implementation at Balakovo NPP in 18 month fuel cycles, at uprated power (104%) and the usage of the axial profiled Gd-rods in order to minimize the power peaking factors and linear heat rate in the upper part in some of the fuel rods. The results of the test operation of fuel rods with different claddings, made by E110M, E125 and E635M alloys at Balakovo NPP were also provided. The recently observed problem with the “white crust” on the cladding surfaces was also discussed

  14. Trends in Operational Fuel Use

    2016-08-01

    prices , to around $140 per bbl, that had the Chief of Naval Operations worried about whether the Navy could afford sufficient steaming days and flying...kt full power rate), and with rates less than 7 bbl/hour (7 is the burn rate of the Allison501-K34 generator at 2000 kw – a basic hotel load) [5...modeling, and simulation—to help develop, evaluate, and implement policies, practices, and programs that make people, budgets , and assets more

  15. Operation of spent fuel storage facilities

    1994-01-01

    This Safety Guide was prepared as part of the IAEA's programme on safety of spent fuel storage. This is for interim spent fuel storage facilities that are not integral part of an operating nuclear power plant. Following the introduction, Section 2 describes key activities in the operation of spent fuel storage facilities. Section 3 lists the basic safety considerations for storage facility operation, the fundamental safety objectives being subcriticality, heat removal and radiation protection. Recommendations for organizing the management of a facility are contained in Section 4. Section 5 deals with aspects of training and qualification; Section 6 describes the phases of the commissioning of a spent fuel storage facility. Section 7 describes operational limits and conditions, while Section 8 deals with operating procedures and instructions. Section 9 deals with maintenance, testing, examination and inspection. Section 10 presents recommendations for radiation and environmental protection. Recommendations for the quality assurance (QA) system are presented in Section 11. Section 12 describes the aspects of safeguards and physical protection to be taken into account during operations; Section 13 gives guidance for decommissioning. 15 refs, 5 tabs

  16. Novel materials for fuel cells operating on liquid fuels

    César A. C. Sequeira

    2017-05-01

    Full Text Available Towards commercialization of fuel cell products in the coming years, the fuel cell systems are being redefined by means of lowering costs of basic elements, such as electrolytes and membranes, electrode and catalyst materials, as well as of increasing power density and long-term stability. Among different kinds of fuel cells, low-temperature polymer electrolyte membrane fuel cells (PEMFCs are of major importance, but their problems related to hydrogen storage and distribution are forcing the development of liquid fuels such as methanol, ethanol, sodium borohydride and ammonia. In respect to hydrogen, methanol is cheaper, easier to handle, transport and store, and has a high theoretical energy density. The second most studied liquid fuel is ethanol, but it is necessary to note that the highest theoretically energy conversion efficiency should be reached in a cell operating on sodium borohydride alkaline solution. It is clear that proper solutions need to be developed, by using novel catalysts, namely nanostructured single phase and composite materials, oxidant enrichment technologies and catalytic activity increasing. In this paper these main directions will be considered.

  17. Fuel failure detection in operating reactors

    Seigel, B.; Hagen, H.H.

    1977-12-01

    Activity detectors in commercial BWRs and PWRs are examined to determine their capability to detect a small number of fuel rod failures during reactor operation. The off-gas system radiation monitor in a BWR and the letdown line radiation monitor in a PWR are calculated to have this capability, and events are cited that support this analysis. Other common detectors are found to be insensitive to small numbers of fuel failures. While adequate detectors exist for normal and transient operation, those detectors would not perform rapidly enough to be useful during accidents; in most accidents, however, primary system sensors (pressure, temperature, level) would provide adequate warning. Advanced methods of fuel failure detection are mentioned

  18. 14 CFR 27.961 - Fuel system hot weather operation.

    2010-01-01

    ... AIRCRAFT AIRWORTHINESS STANDARDS: NORMAL CATEGORY ROTORCRAFT Powerplant Fuel System § 27.961 Fuel system hot weather operation. Each suction lift fuel system and other fuel systems with features conducive to... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel system hot weather operation. 27.961...

  19. 14 CFR 29.961 - Fuel system hot weather operation.

    2010-01-01

    ... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.961 Fuel system hot weather operation. Each suction lift fuel system and other fuel systems conducive to vapor... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel system hot weather operation. 29.961...

  20. Spent Fuel Storage Operation - Lessons Learned

    2013-12-01

    Experience gained in planning, constructing, licensing, operating, managing and modifying spent fuel storage facilities in some Member States now exceeds 50 years. Continual improvement is only achieved through post-project review and ongoing evaluation of operations and processes. This publication is aimed at collating and sharing lessons learned. Hopefully, the information provided will assist Member States that already have a developed storage capability and also those considering development of a spent nuclear fuel storage capability in making informed decisions when managing their spent nuclear fuel. This publication is expected to complement the ongoing Coordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR-III); the scope of which prioritizes facility operational practices in lieu of fuel and structural components behaviour over extended durations. The origins of the current publication stem from a consultants meeting held on 10-12 December 2007 in Vienna, with three participants from the IAEA, Slovenia and USA, where an initial questionnaire on spent fuel storage was formulated (Annex I). The resultant questionnaire was circulated to participants of a technical meeting, Spent Fuel Storage Operations - Lessons Learned. The technical meeting was held in Vienna on 13-16 October 2008, and sixteen participants from ten countries attended. A consultants meeting took place on 18-20 May 2009 in Vienna, with five participants from the IAEA, Slovenia, UK and USA. The participants reviewed the completed questionnaires and produced an initial draft of this publication. A third consultants meeting took place on 9-11 March 2010, which six participants from Canada, Hungary, IAEA, Slovenia and the USA attended. The meeting formulated a second questionnaire (Annex II) as a mechanism for gaining further input for this publication. A final consultants meeting was arranged on 20-22 June 2011 in Vienna. Six participants from Hungary, IAEA, Japan

  1. Rapid detection of defects in fuel-cell electrodes using infrared reactive-flow-through technique

    Das, Prodip K.; Weber, Adam Z.; Bender, Guido; Manak, Austin; Bittinat, Daniel; Herring, Andrew M.; Ulsh, Michael

    2014-09-01

    As fuel cells become more prominent, new manufacturing and production methods will need to be developed to deal efficiently and effectively with increased demand. One necessary component of this industrial growth is the accurate measurement of the variability in the manufacturing process. In this study, we present a diagnostic system that combines infrared thermography with a reactive-flow-through technique to detect catalyst-loading defects in fuel-cell gas-diffusion electrodes accurately with high spatial and temporal resolutions. Experimental results are compared with model predictions of thermal response with good agreement. Data analysis, operating-condition impacts, and detection limits are explored using both experiments and simulation. Overall, the results demonstrate the potential of this technique to measure defects on the millimeter length scale with temporal resolutions appropriate for use on a web-line. Thus we present the first development stage of a next-generation non-destructive diagnostic tool, which may be amenable to eventual use on roll-to-roll manufacturing lines.

  2. 14 CFR 25.961 - Fuel system hot weather operation.

    2010-01-01

    ... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.961 Fuel system hot weather operation. (a) The fuel system must perform satisfactorily in hot weather operation. This... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel system hot weather operation. 25.961...

  3. Environmental impact of nuclear fuel cycle operations

    Wilkinson, W.L.

    1989-09-01

    This paper considers the environmental impact of nuclear fuel cycle operations, particularly those operated by British Nuclear Fuels plc, which include uranium conversion, fuel fabrication, uranium enrichment, irradiated fuel transport and storage, reprocessing, uranium recycle and waste treatment and disposal. Quantitative assessments have been made of the impact of the liquid and gaseous discharges to the environment from all stages in the fuel cycle. An upper limit to the possible health effects is readily obtained using the codified recommendations of the International Commission on Radiological Protection. This contrasts with the lack of knowledge concerning the health effects of many other pollutants, including those resulting from the burning of fossil fuels. Most of the liquid and gaseous discharges result at the reprocessing stage and although their impact on the environment and on human health is small, they have given rise to much public concern. Reductions in discharges at Sellafield over the last few years have been quite dramatic, which shows what can be done provided the necessary very large investment is undertaken. The cost-effectiveness of this investment must be considered. Some of it has gone beyond the point of justification in terms of health benefit, having been undertaken in response to public and political pressure, some of it on an international scale. The potential for significant off-site impact from accidents in the fuel cycle has been quantitatively assessed and shown to be very limited. Waste disposal will also have an insignificant impact in terms of risk. It is also shown that it is insignificant in relation to terrestrial radioactivity and therefore in relation to the human environment. 14 refs, 5 figs, 2 tabs

  4. Behavior of defective LWR-type fuel rods irradiated under postulated accident conditions

    Hobbins, R.R.; Croucher, D.W.; Seiffert, S.L.; Cook, B.A.; Kerwin, D.K.; Mehner, A.S.; Ploger, S.A.

    1979-05-01

    The irradiation experiments reported here have been conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc., for the United States Nuclear Regulatory Commission in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Five of the rods were irradiated in PCM tests and one in a LOC test. During these tests, the six rods lost cladding integrity prior to or during the transient phase of the test due to either manufacturing defects or intentional rod design and operation. Of the five defective rods tested under PCM conditions, one (Rod IE-008, Test IE-1) had a hydride rupture below the region of the rod, which was in film boiling during the transient; two (Rod A-0021, Test PCM-3 and Rod IE-019, Test IE-5) contained defects (a pin hole and a small axial crack, respectively) within the film boiling zone; and two (Rod 201-1, Test PCM-1 and Rod 205-8, Test PCM-5) failed by cladding embrittlement within the film boiling zone. Rod 312-3 was waterlogged before being subjected to LOC conditions in Test LLR-3

  5. Fuel improvement and WWER-1000 FA main operational results

    Rozhkov, V.; Enin, A.; Bezborodov, Y.; Petrov, V.

    2003-01-01

    The JSC NCCP experience of WWER-1000 Fuel Assemblies (FAs) fabrication and operation confirms the adequate feasibility and efficiency of fuel operation in 3-4-x fuel cycles, high operating reliability and competitive capacity as compared with foreign analogues. The work on fuel improvement is aimed at an improvement of the operating reliability and an enhancement of the fuel use efficiency in WWER-1000 advanced FAs

  6. Defect trap model of gas behaviour in UO2 fuel during irradiation

    Szuta, A.

    2003-01-01

    Fission gas behaviour is one of the central concern in the fuel design, performance and hypothetical accident analysis. The report 'Defect trap model of gas behaviour in UO 2 fuel during irradiation' is the worldwide literature review of problems studied, experimental results and solutions proposed in related topics. Some of them were described in details in the report chapters. They are: anomalies in the experimental results; fission gas retention in the UO 2 fuel; microstructure of the UO 2 fuel after irradiation; fission gas release models; defect trap model of fission gas behaviour; fission gas release from UO 2 single crystal during low temperature irradiation in terms of a defect trap model; analysis of dynamic release of fission gases from single crystal UO 2 during low temperature irradiation in terms of defect trap model; behaviour of fission gas products in single crystal UO 2 during intermediate temperature irradiation in terms of a defect trap model; modification of re-crystallization temperature of UO 2 in function of burnup and its impact on fission gas release; apparent diffusion coefficient; formation of nanostructures in UO 2 fuel at high burnup; applications of the defect trap model to the gas leaking fuel elements number assessment in the nuclear power station (VVER-PWR)

  7. Positron annihilation spectroscopy study of lattice defects in non-irradiated doped and un-doped fuels

    Chollet Mélanie

    2017-01-01

    Full Text Available Fission gas behavior within the fuel structure plays a major role for the safety of nuclear fuels during operation in the nuclear power plant. Fission gas distribution and retention is determined by both, micro- and lattice-structure of the fuel matrix. The ADOPT (Advanced Doped Pellet Technology fuel, containing chromium and aluminum additives, shows larger grain sizes than standard (undoped UO2 fuel, enhancing the fission gas retention properties of the matrix. However, the additions of such trivalent cations shall also induce defects in the lattice. In this study, we investigated the microstructure of such doped fuels as well as a reference standard UO2 by positron annihilation spectroscopy (PAS. Although this technique is particularly sensitive to lattice point defects in materials, a wider application in the UO2 research is still missing. The PAS-lifetime components were measured in the hotlab facility of PSI using a 22Na source sandwiched between two 500-μm-thin sample discs. The values of lifetime at the center and the rim of both samples, examined to check at the radial homogeneity of the pellets, are not significantly different. The mean lifetimes were found to be longer in the ADOPT material, 220 ps, than in standard UO2, 190 ps, which indicates a larger presence of additional defects, presumably generated by the dopants. While two-component decomposition (bulk + one defect component could be performed for the standard material, only one lifetime component was found in the doped material. The absence of the bulk component in the ADOPT sample refers to a saturated positron trapping (i.e., all positrons are trapped at defects. In order to associate a type of lattice defect to each PAS component, interpretation of the PAS experimental observations was conducted with respect to existing experimental and modeling studies. This work has shown the efficiency of PAS to detect lattice point defects in UO2 produced by Cr and Al oxides

  8. Operational support of a safe operating envelope for fuel

    Chapman, T.J.; Gibb, R.A.

    1998-01-01

    The mandate of a station safety analysis group is to ensure that the station is operated and maintained in a manner consistent with the basis for our understanding of the safety consequences of process or human failures. As operating experience has developed an awareness of the significance of fuel manufacture and operating conditions on safety consequences has also grown. This awareness has led to a program that is designed to ensure that these influences are appropriately considered. This paper describes the projects that make up this program. (author)

  9. Operational limitations of light water reactors relating to fuel performance

    Cheng, H.S.

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed

  10. Determination and microscopic study of incipient defects in irradiated power reactor fuel rods. Final report

    Pasupathi, V.; Perrin, J.S.; Roberts, E.

    1978-05-01

    This report presents the results of nondestructive and destructive examinations carried out on the Point Beach-1 (PWR) and Dresden-3 (BWR) candidate fuel rods selected for the study of pellet-clad interaction (PCI) induced incipient defects. In addition, the report includes results of examination of sections from Oskarshamn-1 (BWR) fuel rods. Eddy current examination of Point Beach-1 rods showed indications of possible incipient defects in the fuel rods. The profilometry and the gamma scan data also indicated that the source of the eddy current indications may be incipient defects. No failed rods or rods with incipient failure were found in the sample from Point Beach-1. Despite the lack of success in finding incipient defects and filed rods, the mechanism for fuel rod failures in Point Beach-1 is postulated to be PCI-related, with high startup rates and fuel handling being the key elements. Nine out of the 10 candidate fuel rods from Dresden-3 (BWR) were failed, and all the failed rods had leaked water so that the initial mechanism was observed. Examination of clad inner surfaces of the specimens from failed and unfailed rods showed fuel deposits of widely varying appearance. The deposits were found to contain uranium, cesium, and tellurium. Transmission electron microscopy of clad specimens showed evidence of microscopic strain. Metallographic examination of fuel pellets from the peak transient power location showed extensive grain boundary separation and axial movement of the fuel indicative of rapid release of fission products. Examination of Oskarshamn clad specimens did not show any stress corrosion crack (SCC) type defects. The defects found in the examinations appear to be related to secondary hydriding. The clad inner surface of the Oskarshamn specimens also showed uranium-rich deposits of varying features

  11. Atomic-scale effects of chromium-doping on defect behaviour in uranium dioxide fuel

    Guo, Zhexi; Ngayam-Happy, Raoul; Krack, Matthias; Pautz, Andreas

    2017-01-01

    The effects of doping conventional UO 2 fuel with chromium are studied through atomistic simulations using empirical force field methods. We first analyse the stable structures of unirradiated doped fuel by determining the preferred lattice configuration of chromium ions and oxygen vacancies within the matrix. In order to understand the physical effects of the dopants, we investigate the energy change upon inserting isolated defects and Frenkel pairs in the vicinity of chromium. The behaviour of point defects is then studied with collision cascade simulations and relaxation of doped simulation cells containing Frenkel pairs. The defective structures are analysed using an in-house tool named ASTRAM. Results indicate definite effects of chromium-doping on the ease with which defects are formed. Moreover, the extent of Cr effects on the residual damage following a displacement cascade is dependent on the dopant distribution and concentration in the fuel matrix.

  12. Atomic-scale effects of chromium-doping on defect behaviour in uranium dioxide fuel

    Guo, Zhexi; Ngayam-Happy, Raoul, E-mail: raoul.ngayam-happy@psi.ch; Krack, Matthias; Pautz, Andreas

    2017-05-15

    The effects of doping conventional UO{sub 2} fuel with chromium are studied through atomistic simulations using empirical force field methods. We first analyse the stable structures of unirradiated doped fuel by determining the preferred lattice configuration of chromium ions and oxygen vacancies within the matrix. In order to understand the physical effects of the dopants, we investigate the energy change upon inserting isolated defects and Frenkel pairs in the vicinity of chromium. The behaviour of point defects is then studied with collision cascade simulations and relaxation of doped simulation cells containing Frenkel pairs. The defective structures are analysed using an in-house tool named ASTRAM. Results indicate definite effects of chromium-doping on the ease with which defects are formed. Moreover, the extent of Cr effects on the residual damage following a displacement cascade is dependent on the dopant distribution and concentration in the fuel matrix.

  13. Fluid flows due to leakproofness defects of nuclear fuel cans: application to sipping in PWR

    Parrat, D.

    1983-10-01

    This thesis establishes the theoritical laws describing fluid and gas transfers through channels representative of nuclear fuel can defects. These laws, applied to craks on fuel cans (samples) and on new or spent experimental rods, allow to know their size. A computerized modelling has been developed to analyse more particularly the examination by ''quantitative permeation testing'', giving the equivalent size of a channel from the observed signal. The utilization of this process to non-tight fuel assemblies of pressurized water reactors leads to the establishment of a classification of defects according to their importance [fr

  14. 14 CFR 23.961 - Fuel system hot weather operation.

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel system hot weather operation. 23.961... AIRCRAFT AIRWORTHINESS STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System § 23.961 Fuel system hot weather operation. Each fuel system must be free from vapor lock...

  15. 3D modeling of missing pellet surface defects in BWR fuel

    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

  16. Operator aid system for Dhruva fueling machine

    Misra, S.M.; Ramaswamy, L.R.; Gohel, N.; Bharadwaj, G.; Ranade, M.R.; Khadilkar, M.G.

    1997-01-01

    Systems with significant software contents are replacing the old hardware logic systems. These systems not only are versatile but are easy to make changes in the program. Extensive use of such systems in critical real-time operation environment warrants not only excessive training on simulators, documentation but also fault tolerant system to bring the operation to a safe state in case of error. With new graphic user software interface and advancement in personal computer hardware design, the dynamic status of the physical environment can be shown on the visual display at near real time. These visual aids along with the software covering all the interlocks aids an operator in his professional work. This paper highlights the operator aid system for Dhruva fueling machine. (author). 6 refs., 1 fig

  17. Device for detecting defective nuclear reactor fuel rods

    Steven, J.

    1976-01-01

    A moisture sensor is provided for a nuclear fuel rod for water-cooled nuclear reactors wherein moisture can be present. The fuel rod has an end cap and a charge of nuclear fuel. The moisture sensor is disposed between the end cap and the charge and serves to detect a leak in the fuel rod. The moisture sensor includes a capsule-like housing having an inner space and having openings through which moisture can pass into the inner space in the event of a leak in the fuel rod. Ferromagnetic material is disposed in the inner space of the housing together with a moisture detector responsive to moisture for altering the diposition of the ferromagnetic material in the inner space. 5 claims, 6 drawing figures

  18. Fuel performance in NPD while operating with two-phase coolant

    Bain, A.S.

    1978-03-01

    The NPD reactor operated as a boiling heavy water reactor from October 27, 1968 to April 18, 1971. At 25 MWe the steam quality at the steam generator inlet was 13 wt%, and fuel channel outlet steam qualities ranged from 2 to 22 wt%. During this period ammonia was used for oxygen suppression and pH control. At equilibrium the coolant had 7 mg NH 3 /kg D 2 O, 60 ml D 2 /kg D 2 O and 20 ml N 2 /kg D 2 O. The performance of the fuel was excellent during the time that NPD operated in the boiling mode. No indications were observed of dimensional changes, inter-element fretting, fuel/sheath interaction, excessive oxidation, excessive deuterium concentrations, or unusual migration of hydrogen and deuterium to the cooler end plugs. One element defected; although the defect mechanism could not be identified at the time, we now believe the defect was associated with faulty bar stock for end plugs. The behaviour of the defective element was similar to that for other defective elements in CANDU reactors. No problems were encountered in removing the defected bundle from the reactor. (author)

  19. 46 CFR 108.487 - Helicopter deck fueling operations.

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Helicopter deck fueling operations. 108.487 Section 108... DESIGN AND EQUIPMENT Fire Extinguishing Systems Fire Protection for Helicopter Facilities § 108.487 Helicopter deck fueling operations. (a) Each helicopter landing deck on which fueling operations are...

  20. The Defect Inspection on the Irradiated Fuel Rod by Eddy Current Test

    Koo, D. S.; Park, Y. K.; Kim, E. K.

    1996-01-01

    The eddy current test(ECT) probe of differential encircling coil type was designed and fabricated, and the optimum condition of ECT was derived for the examination of the irradiated fuel rod. The correlation between ECT test frequency and phase and amplitude was derived by performing the test of the standard rig that includes inner notches, outer notches and through-holes. The defect of through-hole was predicted by ECT at the G33-N2 fuel rod irradiated in the Kori-1 nuclear power reactor. The metallographic examination on the G33-N2 fuel rod was Performed at the defect location predicted by ECT. The result of metallographic examination for the G33-N2 fuel rod was in good agreement with that of ECT. This proves that the evaluation for integrity of irradiated fuel rod by ECT is reliable

  1. Method to mount defect fuel elements i transport casks

    Borgers, H.; Deleryd, R.

    1996-01-01

    Leaching or otherwise failed fuel elements are mounted in special containers that fit into specially designed chambers in a transportation cask for transport to reprocessing or long-time storage. The fuel elements are entered into the container under water in a pool. The interior of the container is dried before transfer to the cask. Before closing the cask, its interior, and the exterior of the container are dried. 2 figs

  2. CANDU fuel performance

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

    1982-01-01

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

  3. Natural uranium metallic fuel elements: fabrication and operating experience

    Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

    1980-01-01

    The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

  4. Quarterly Progress Report Fuels Development Operation: October - December 1959

    Cadwell, J. J. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation; Tobin, J. C. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Physical Metallurgy; Minor, J. E. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuel Element Design; Evans, E. A. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Ceramic Fuels Development; Bush, S. H. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuels Fabrication Development

    1960-01-15

    The present Quarterly Report is the continuation of a series issued by the new Fuels Development operation. Reports in this series combine portions of the quarterly reports by the former Metallurgy Research and Fuel Technology Sub-Sections. Work reported includes research conducted by the Physical Metallurgy Operation, and research and development conducted by Fuel Design, Fuels Fabrication Development and Ceramic Fuels Development Operations. Studies formerly reported by the Radiometallurgy, Metallography, and Welding and Corrosion Units, in addition to portions of the Fuels Technology work, are reported elsewhere.

  5. Quarterly Progress Report Fuels Development Operation: January - March 1958

    Cadwell, J. J. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation; Tobin, J. C. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Physical Metallurgy; Minor, J. E. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuel Element Design; Evans, E. A. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Ceramic Fuels Development; Bush, S. H. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuels Fabrication Development

    1958-04-15

    The present Quarterly Report is the continuation of a series issued by the new Fuels Development operation. Reports in this series combine portions of the quarterly reports by the former Metallurgy Research and Fuel Technology Sub-Sections. Work reported includes research conducted by the Physical Metallurgy Operation, and research and development conducted by Fuel Design, Fuels Fabrication Development and Ceramic Fuels Development Operations. Studies formerly reported by the Radiometallurgy, Metallography, and Welding and Corrosion Units, in addition to portions of the Fuels Technology work, are reported elsewhere.

  6. Quarterly Progress Report Fuels Development Operation: July - September 1957

    Bush, S. H. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Physical Metallurgy; Minor, J. E. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuel Element Design; Evans, E. A. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Ceramic Fuels Development; Wallace, W. P. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuels Fabrication Development

    1957-10-15

    The present Quarterly Report is the continuation of a series issued by the new Fuels Development operation. Reports in this series combine portions of the quarterly reports by the former Metallurgy Research and Fuel Technology Sub-Sections. Work reported includes research conducted by the Physical Metallurgy Operation, and research and development conducted by Fuel Design, Fuels Fabrication Development and Ceramic Fuels Development Operations. Studies formerly reported by the Radiometallurgy, Metallography, and Welding and Corrosion Units, in addition to portions of the Fuels Technology work, are reported elsewhere.

  7. Fuel management for TRIGA reactor operators

    Totenbier, R.E.; Levine, S.H.

    1980-01-01

    One responsibility of the Supervisor of Reactor Operations is to follow the TRIGA core depletion and recommend core loading changes for refueling and special experiments. Calculations required to analyze such changes normally use digital computers and are extremely difficult to perform for one who is not familiar with computer language and nuclear reactor diffusion theory codes. The TRICOM/SCRAM program developed to perform such calculations for the Penn State TRIGA Breazeale Reactor (PSBR), has a very simple input format and is one which can be used by persons having no knowledge of computer codes. The person running the program need not understand computer language such as Fortran, but should be familiar with reactor core geometry and effects of loading changes. To further simplify the input requirements but still allow for all of the studies normally needed by the reactor operations supervisor, the options required for input have been isolated to two. Given a master deck of computer cards one needs to change only three cards; a title card, core energy history information card and one with core changes. With this input, the program can provide individual fuel element burn-up for a given period of operation and the k eff of the core. If a new loading is desired, a new master deck containing the changes is also automatically provided. The life of a new core loading can be estimated by feeding in projected core burn-up factors and observing the resulting loss in individual fuel elements. The code input and output formats have now been made sufficiently convenient and informative as to be incorporated into a standard activity for the Reactor Operations Supervisor. (author)

  8. Cracking and relocation of UO2 fuel during nuclear operation

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

    1981-01-01

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

  9. Evaluation of Biodiesel Fuels to Reduce Fossil Fuel Use in Corps of Engineers Floating Plant Operations

    2016-07-01

    ER D C/ CH L TR -1 6- 11 Dredging Operations and Environmental Research Program Evaluation of Biodiesel Fuels to Reduce Fossil Fuel Use... Fuels to Reduce Fossil Fuel Use in Corps of Engineers Floating Plant Operations Michael Tubman and Timothy Welp Coastal and Hydraulics Laboratory...sensitive emissions, increase use of renewable energy, and reduce the use of fossil fuels was conducted with funding from the U.S. Army Corps of

  10. A new criterion for defective used nuclear fuel in dry storage condition

    Desgranges, Lionel [CEA Cadarache Bat 316 13108Saint-Paul lez Durance (France); Poulesquen, Arnaud; Ferry, Cecile [CEA Saclay Bat. 450 point courrier no 40 - 91191 Gif-sur-Yvette cedex (France)

    2008-07-01

    In the frame of the PRECCI program, the mechanisms associated to the oxidation of used nuclear fuel were studied. Experiments on un-irradiated UO{sub 2}, evidenced that the oxidation proceeded in three stages (namely UO{sub 2} {yields} U{sub 4}O{sub 9}, U{sub 4}O{sub 9} {yields} U{sub 3}O{sub 7} and U{sub 3}O{sub 7} {yields} U{sub 3}O{sub 8}) instead of 2 as previously proposed in literature. Experiments on irradiated fuel fragments evidenced the existence of an oxidation front inside the fuel fragment associated to some fission gas release. The degradation of a fuel rod slice was simulated in situ with CROCODILE experiment, and shown to be due to the fuel swelling. Finally a new criterion was proposed defining a safe duration for defective used fuel in a dry air facility in the case of an accident scenario where a breach in a container would put a defective fuel rod in contact with atmosphere. The criterion is related to the time needed to form a given thickness of U{sub 4}O{sub 9{gamma}}{sub +{psi}} with a higher stoichiometry than a given value, in irradiated grains. This U{sub 4}O{sub 9{gamma}}{sub +{psi}} layer thickness is assumed to simulate the onset of the ceramic fragmentation. This safety duration was calculated thanks to a new modeling of fuel fragment oxidation.

  11. Optimization of Fuel Cell System Operating Conditions for Fuel Cell Vehicles

    Zhao, Hengbing; Burke, Andy

    2008-01-01

    Proton Exchange Membrane fuel cell (PEMFC) technology for use in fuel cell vehicles and other applications has been intensively developed in recent decades. Besides the fuel cell stack, air and fuel control and thermal and water management are major challenges in the development of the fuel cell for vehicle applications. The air supply system can have a major impact on overall system efficiency. In this paper a fuel cell system model for optimizing system operating conditions was developed wh...

  12. Alternate aircraft fuels prospects and operational implications

    Witcofski, R. D.

    1977-01-01

    The paper discusses NASA studies of the potentials of coal-derived aviation fuels, specifically synthetic aviation kerosene, liquid methane, and liquid hydrogen. Topics include areas of fuel production, air terminal requirements for aircraft fueling (for liquid hydrogen only), and the performance characteristics of aircraft designed to utilize alternate fuels. Energy requirements associated with the production of each of the three selected fuels are determined, and fuel prices are estimated. Subsonic commercial air transports using liquid hydrogen fuel have been analyzed, and their performance and the performance of aircraft which use commercial aviation kerosene are compared. Environmental and safety issues are considered.

  13. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented

  14. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented. 6 references, 4 figures

  15. Defect sizing of post-irradiated nuclear fuels using grayscale thresholding in their radiographic images

    Chaudhary, Usman Khurshid; Iqbal, Masood; Ahmad, Munir

    2010-01-01

    Quantification of different types of material defects in a number of reference standard post-irradiated nuclear fuel image samples have been carried out by virtue of developing a computer program that takes radiographic images of the fuel as input. The program is based on user adjustable grayscale thresholding in the regime of image segmentation whereby it selects and counts the pixels having graylevel values less than or equal to the computed threshold. It can size the defects due to chipping in nuclear fuel, cracks, voids, melting, deformation, inclusion of foreign materials, heavy isotope accumulation, non-uniformity, etc. The classes of fuel range from those of research and power reactors to fast breeders and from pellets to annular and vibro-compacted fuel. The program has been validated against ground truth realities of some locally fabricated metallic plates having drilled holes of known sizes simulated as defects in them in which the results indicate that it either correctly selects and quantifies at least 94% of the actual required regions of interest in a given image or it gives less than 8.1% false alarm rate. Also, the developed program is independent of image size.

  16. Defect sizing of post-irradiated nuclear fuels using grayscale thresholding in their radiographic images

    Chaudhary, Usman Khurshid, E-mail: ukhurshid@hotmail.co [Department of Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan); Iqbal, Masood, E-mail: masiqbal@hotmail.co [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad 45650 (Pakistan); Ahmad, Munir [Nondestructive Testing Group, Directorate of Technology, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad 45650 (Pakistan)

    2010-10-15

    Quantification of different types of material defects in a number of reference standard post-irradiated nuclear fuel image samples have been carried out by virtue of developing a computer program that takes radiographic images of the fuel as input. The program is based on user adjustable grayscale thresholding in the regime of image segmentation whereby it selects and counts the pixels having graylevel values less than or equal to the computed threshold. It can size the defects due to chipping in nuclear fuel, cracks, voids, melting, deformation, inclusion of foreign materials, heavy isotope accumulation, non-uniformity, etc. The classes of fuel range from those of research and power reactors to fast breeders and from pellets to annular and vibro-compacted fuel. The program has been validated against ground truth realities of some locally fabricated metallic plates having drilled holes of known sizes simulated as defects in them in which the results indicate that it either correctly selects and quantifies at least 94% of the actual required regions of interest in a given image or it gives less than 8.1% false alarm rate. Also, the developed program is independent of image size.

  17. Fuel supply shutdown facility interim operational safety requirements

    Besser, R.L.; Brehm, J.R.; Benecke, M.W.; Remaize, J.A.

    1995-01-01

    These Interim Operational Safety Requirements (IOSR) for the Fuel Supply Shutdown (FSS) facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls to ensure safe operation. The IOSRs apply to the fuel material storage buildings in various modes (operation, storage, surveillance)

  18. Cause of defect in the end plug welding of the JOYO fuel pin

    Ouchi, Masaru; Otani, Seiji; Onisi, Koichi; Tateisi, Yoshinori; Ikawa, Yukio.

    1976-01-01

    About twelve thousand fuel pins for the JOYO core fuel were fabricated, and their end plug welding was inspected by X-ray radiography. The defect fractions were 0.2 percent for the lower end plugs and 1.8 percent for the upper, respectively. It had been known that the defect was due to ''line porosity''. In this study, the cause of the ''line porosity defect'' was investigated by the welding experiment performed on some dummy specimens of three different types; open end; closed end; and closed end with dummy pellets and a spring. The position of electrodes was varied for changing the arc gap from 0.3 mm to 1.2 mm. The experimental results are summarized in tables. The results showed that no defect was found in the open end type specimens even with the arc gap of 1.2 mm. Whereas in the other two types of specimens, the defect fraction of 60 to 75 percent was observed with the same arc gap. As for the effect of the arc gap, it was shown that 0.3 mm is the best among 0.3 mm, 0.5 mm and 1.2 mm. No defect was observed in the third type of specimens with the arc gap of 0.3 mm. In summary, it was found that the line porosity defect did not depend on the shape of the end plugs. It is considered to be dependent on both the structure of dummy fuel pins and the position of electrodes. (Aoki, K.)

  19. Solid Oxide Fuel Cells Operating on Alternative and Renewable Fuels

    Wang, Xiaoxing; Quan, Wenying; Xiao, Jing; Peduzzi, Emanuela; Fujii, Mamoru; Sun, Funxia; Shalaby, Cigdem; Li, Yan; Xie, Chao; Ma, Xiaoliang; Johnson, David; Lee, Jeong; Fedkin, Mark; LaBarbera, Mark; Das, Debanjan; Thompson, David; Lvov, Serguei; Song, Chunshan

    2014-09-30

    This DOE project at the Pennsylvania State University (Penn State) initially involved Siemens Energy, Inc. to (1) develop new fuel processing approaches for using selected alternative and renewable fuels – anaerobic digester gas (ADG) and commercial diesel fuel (with 15 ppm sulfur) – in solid oxide fuel cell (SOFC) power generation systems; and (2) conduct integrated fuel processor – SOFC system tests to evaluate the performance of the fuel processors and overall systems. Siemens Energy Inc. was to provide SOFC system to Penn State for testing. The Siemens work was carried out at Siemens Energy Inc. in Pittsburgh, PA. The unexpected restructuring in Siemens organization, however, led to the elimination of the Siemens Stationary Fuel Cell Division within the company. Unfortunately, this led to the Siemens subcontract with Penn State ending on September 23rd, 2010. SOFC system was never delivered to Penn State. With the assistance of NETL project manager, the Penn State team has since developed a collaborative research with Delphi as the new subcontractor and this work involved the testing of a stack of planar solid oxide fuel cells from Delphi.

  20. Analysis of fuel operational reliability and fuel failures

    Smiesko, I.

    1999-01-01

    In this lecture the fuel failure (loss of fuel rod (cladding) integrity, corruption of second barrier for fission product release from duel and their consequences (increase of primary coolant activity; increase of fission product releases to environment; increase of rad-waste activities and potential increase of personnel exposure) are discussed

  1. Post operation: The changing characteristics of nuclear fuel cycle costs

    Frank, F.J.

    1986-01-01

    Fundamental changes have occurred in the nuclear fuel cycle. These changes forged by market forces, legislative action, and regulatory climate appear to be a long term characteristic of the nuclear fuel cycle. The nature of these changes and the resulting emerging importance of post-operation and its impact on fuel cycle costs are examined

  2. Tailoring Vantage 5 (fuel) to suit each operator's need

    Chapin, D L; Secker, J R [Westinghouse Electric Corp., Philadelphia, PA (USA)

    1990-03-01

    By the end of 1989, Westinghouse Vantage 5 fuel had been reloaded into 36 nuclear power plants. The fuel offers a number of features operators can choose from to suit their own particular needs. Experience so far has shown the fuel to have performed well, with coolant activity levels remaining low. (author).

  3. Romanian nuclear fuel fabrication and in-reactor fuel operational experience

    Budan, O.

    2003-01-01

    A review of the Romanian nuclear program since mid 60's is made. After 1990, the new Romanian nuclear power authority, RENEL-GEN, elaborated a realistic Nuclear Fuel Program. This program went through the Romanian nuclear fuel plant qualification with the Canadian (AECL and ZPI) support, restarting in January 1995 of the industrial nuclear fuel production, quality evaluation of the fuel produced before 1990 and the recovery of this fuel. This new policy produced good results. FCN is since 1995 the only CANDU fuel supplier from outside Canada recognised by AECL as an authorised CANDU fuel manufacturer. The in-reactor performances and behaviour of the fuel manufactured by FCN after its qualification have been excellent. Very low - more then five times lesser than the design value - fuel defect rate has been recorded up to now and the average discharge of this fuel was with about 9% greater than the design value. Since mid 1998 when SNN took charge of the production of nuclear generated electricity, FCN made significant progresses in development and procurement of new and more efficient equipment and is now very close to double its fuel production capacity. After the completion of the recovery of the fuel produced before June 1990, FCN is already prepared to shift its fuel production to the so-called 'heavy' bundle containing about 19.3 kg of Uranium per bundle

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

    Murayama, Y.; Issiki, M.

    1995-01-01

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

  5. Safety analysis to support a safe operating envelope for fuel

    Gibb, R.A.; Reid, P.J.

    1998-01-01

    This paper presents an approach for defining a safe operating envelope for fuel. 'Safe operating envelope' is defined as an envelope of fuel parameters defined for application in safety analysis that can be related to, or used to define, the acceptable range of fuel conditions due to operational transients or deviations in fuel manufacturing processes. The paper describes the motivation for developing such a methodology. The methodology involved four steps: the update of fission product inventories, the review of sheath failure criteria, a review of input parameters to be used in fuel modelling codes, and the development of an improved fission product release code. This paper discusses the aspects of fuel sheath failure criteria that pertain to operating or manufacturing conditions and to the evaluation and selection of modelling input data. The other steps are not addressed in this paper since they have been presented elsewhere. (author)

  6. Fuel Supply Shutdown Facility Interim Operational Safety Requirements

    BENECKE, M.W.

    2000-01-01

    The Interim Operational Safety Requirements for the Fuel Supply Shutdown (FSS) Facility define acceptable conditions, safe boundaries, bases thereof, and management of administrative controls to ensure safe operation of the facility

  7. Spent fuel receipt and storage at the Morris Operation

    Astrom, K.A.; Eger, K.J.

    1978-06-01

    Operating and maintenance activities in an independent spent fuel storage facility are described, and current regulations governing such activities are summarized. This report is based on activities at General Electric's licensed storage facility located near Morris, Illinois, and includes photographs of cask and fuel handling equipment used during routine operations

  8. Method of operating a direct dme fuel cell system

    2011-01-01

    The present invention relates to a method of operating a fuel cell system comprising one or more fuel cells with a proton exchange membrane, wherein the membrane is composed of a polymeric material comprising acid-doped polybenzimidazole (PBI). The method comprises adjusting the operating...

  9. Operational results of WWER fuel fabricated by MSZ (Elektrostal, Russia)

    Asatiani, I.; Balabanov, S.; Beglov, A.; Khryashchev, D.

    2009-01-01

    The presentation brings forth a statistical analysis of the WWER fuel manufactured by OAO MSZ, operational experience. A necessity of such an analysis is determined by the fact that objective operational results prove the appropriateness of the solutions and decisions made by vendor, designer, manufacturer and utility, as well as motivates further fuel improvements. (authors)

  10. Modeling defect and fission gas properties in U-Si fuels

    Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Noordhoek, Mark [Univ. of South Carolina, Columbia, SC (United States); Besmann, Theodore [Univ. of South Carolina, Columbia, SC (United States); Middleburgh, Simon C. [Westinghouse Electric Sweden, Vasteras (Sweden); Lahoda, E. J. [Westinghouse Electric Company LLC, Cranberry Woods, PA (United States); Chernatynskiy, Aleksandr [Missouri University of Science and Technology; Grimes, Robin W. [Imperial College, London (United Kingdom)

    2017-04-27

    Uranium silicides, in particular U3Si2, are being explored as an advanced nuclear fuel with increased accident tolerance as well as competitive economics compared to the baseline UO2 fuel. They benefit from high thermal conductivity (metallic) compared to UO2 fuel (insulator or semi-conductor) used in current Light Water Reactors (LWRs). The U-Si fuels also have higher fissile density. In order to perform meaningful engineering scale nuclear fuel performance simulations, the material properties of the fuel, including the response to irradiation environments, must be known. Unfortunately, the data available for USi fuels are rather limited, in particular for the temperature range where LWRs would operate. The ATF HIP is using multi-scale modeling and simulations to address this knowledge gap.

  11. Modeling defect and fission gas properties in U-Si fuels

    Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Noordhoek, Mark J. [Univ. of South Carolina, Columbia, SC (United States); Besmann, Theodore M. [Univ. of South Carolina, Columbia, SC (United States); Middleburgh, Simon C. [Westinghouse Electric Sweden, Vasteras (Sweden); Lahoda, E. J. [Westinghouse Electric Company LLC, Cranberry Woods, PA (United States); Chernatynskiy, Aleksandr [Missouri Univ. of Science and Technology, Rolla, MO (United States); Grimes, Robin W. [Imperial College, London (United Kingdom)

    2017-04-14

    Uranium silicides, in particular U3Si2, are being explored as an advanced nuclear fuel with increased accident tolerance as well as competitive economics compared to the baseline UO2 fuel. They benefit from high thermal conductivity (metallic) compared to UO2 fuel (insulator or semi-conductor) used in current Light Water Reactors (LWRs). The U-Si fuels also have higher fissile density. In order to perform meaningful engineering scale nuclear fuel performance simulations, the material properties of the fuel, including the response to irradiation environments, must be known. Unfortunately, the data available for USi fuels are rather limited, in particular for the temperature range where LWRs would operate. The ATF HIP is using multi-scale modeling and simulations to address this knowledge gap.

  12. Electrochemical Methods for Reprocessing Defective Fuel Elements and for Decontaminating Equipment

    Mikheykin, S. V.; Rybakov, K. A.; Simonov, V. P.

    2002-01-01

    Reprocessing of fuel elements receives much consideration in nuclear engineering. Chemical and electrochemical methods are used for the purpose. For difficultly soluble materials based on zirconium alloys chemical methods are not suitable. Chemical reprocessing of defective or irradiated fuel elements requires special methods for their decladding because the dissolution of the clad material in nitric acid is either impossible (stainless steel, Zr alloys) or quite slow (aluminium). Fuel elements are cut in air-tight glove-boxes equipped with a dust collector and a feeder for crushed material. Chemical treatment is not free from limitations. For this reason we started a study of the feasibility of electrochemical methods for reprocessing defective and irradiated fuel elements. A simplified electrochemical technology developed makes it possible to recover expensive materials which were earlier wasted or required multi-step treatment. The method and an electrochemical cell are suitable for essentially complete dissolution of any fuel elements, specifically those made of materials which are difficultly soluble by chemical methods

  13. Effect of cladding defect size on the oxidation of irradiated spent LWR [light-water reactor] fuel below 3690C

    Einziger, R.E.; Strain, R.V.

    1984-01-01

    Tests on spent fuel fragments and rod segments were conducted between 250 and 360 0 C to relate temperature, defect size, and fuel oxidation rate with time-to-cladding-splitting. Defect sizes from 760 μm diameter down to 8 μm, the size of an SCC type breach, were used. Above 283 0 C, the time-to-cladding-splitting was longer for the smaller defects. The enhancement of the incubation time by smaller defects steadily decreased with temperature and was not detected at 250 0 C. 18 refs., 10 figs., 4 tabs

  14. A Method of Operating a Fuel Cell

    2013-01-01

    The present invention relates to a method of determining the net water drag coefficient (rd) in a fuel cell. By measuring the velocity of the fluid stream at the outlet of the anode, rd can be determined. Real time monitoring and adjustments of the water balance of a fuel cell may be therefore...

  15. Tests of the SNR fuel pin behaviour in case of operational transients in the HFR Petten

    Plitz, H.

    1989-05-01

    The loadings on fast reactor fuel pins under operational transients (power and temperature increases in the design area) have been studied in the High-Flux-Reactor HFR in Petten with sodium cooled irradiation capsules. The results of the first campaign of transient experiments are described in the report. No cladding defects have been observed, and the fuel pins of the Mark-I and Mark-II type resisted to linear power levels of more than 800 W/cm, thus demonstrating the required design margins. The plans for further experiments are outlined

  16. Halden fuel and material experiments beyond operational and safety limits

    Volkov, Boris; Wiesenack, Wolfgang; McGrath, M.; Tverberg, T.

    2014-01-01

    One of the main tasks of any research reactor is to investigate the behavior of nuclear fuel and materials prior to their introduction into the market. For commercial NPPs, it is important both to test nuclear fuels at a fuel burn-up exceeding current limits and to investigate reactor materials for higher irradiation dose. For fuel vendors such tests enable verification of fuel reliability or for the safety limits to be found under different operational conditions and accident situations. For the latter, in-pile experiments have to be performed beyond some normal limits. The program of fuel tests performed in the Halden reactor is aimed mainly at determining: The thermal FGR threshold, which may limit fuel operational power with burn-up increase, the “lift-off effect” when rod internal pressure exceeds coolant pressure, the effects of high burn-up on fuel behavior under power ramps, fuel relocation under LOCA simulation at higher burn-up, the effect of dry-out on high burn-up fuel rod integrity. This paper reviews some of the experiments performed in the Halden reactor for understanding some of the limits for standard fuel utilization with the aim of contributing to the development of innovative fuels and cladding materials that could be used beyond these limits. (author)

  17. Safeguards operations in the integral fast reactor fuel cycle

    Goff, K.M.; Benedict, R.W.; Brumbach, S.B.; Dickerman, C.E.; Tompot, R.W.

    1994-01-01

    Argonne National Laboratory is currently demonstrating the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The safeguards aspects of the fuel cycle demonstration must be approved by the United States Department of Energy, but a further goal of the program is to develop a safeguards system that could gain acceptance from the Nuclear Regulatory Commission and International Atomic Energy Agency. This fuel cycle is described with emphasis on aspects that differ from aqueous reprocessing and on its improved safeguardability due to decreased attractiveness and diversion potential of all process streams, including the fuel product

  18. Defects and Interfaces on PtPb Nanoplates Boost Fuel Cell Electrocatalysis.

    Sun, Yingjun; Liang, Yanxia; Luo, Mingchuan; Lv, Fan; Qin, Yingnan; Wang, Lei; Xu, Chuan; Fu, Engang; Guo, Shaojun

    2018-01-01

    Nanostructured Pt is the most efficient single-metal catalyst for fuel cell technology. Great efforts have been devoted to optimizing the Pt-based alloy nanocrystals with desired structure, composition, and shape for boosting the electrocatalytic activity. However, these well-known controls still show the limited ability in maximizing the Pt utilization efficiency for achieving more efficient fuel cell catalysis. Herein, a new strategy for maximizing the fuel cell catalysis by controlling/tuning the defects and interfaces of PtPb nanoplates using ion irradiation technique is reported. The defects and interfaces on PtPb nanoplates, controlled by the fluence of incident C + ions, make them exhibit the volcano-like electrocatalytic activity for methanol oxidation reaction (MOR), ethanol oxidation reaction (EOR), and oxygen reduction reaction (ORR) as a function of ion irradiation fluence. The optimized PtPb nanoplates with the mixed structure of dislocations, subgrain boundaries, and small amorphous domains are the most active for MOR, EOR, and ORR. They can also maintain high catalytic stability in acid solution. This work highlights the impact and significance of inducing/controlling the defects and interfaces on Pt-based nanocrystals toward maximizing the catalytic performance by advanced ion irradiation strategy. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  19. Behaviour of fission products in PWR primary coolant and defected fuel rods evaluation

    Bourgeois, P.; Stora, J.P.

    1979-01-01

    The activity surveillance of the PWR primary coolant by γ spectometry gives some informations on fuel failures. The activity of different nuclides e.g. Xenons, Kryptons, Iodines, can be correlated with the number of the defected fuel rods. Therefore the precharacterization with eventually a prelocalization of the related fuel assemblies direct the sipping-test and allows a saving of time during refueling. A model is proposed to calculate the number of the defected rods from the activity measurements of the primary coolant. A semi-empirical model of the release of the fission products has been built from the activity measurements of the primary coolant in a 900 MWe PWR. This model allows to calculate the number of the defected rods and also a typical parameter of the mean damage. Fission product release is described by three stages: release from uranium dioxide, transport across the gas gap and behaviour in the primary coolant. The model of release from the oxide considers a diffusion process in the grains with trapping. The release then occurs either directly to free surfaces or with a delay due to a transit into closed porosity of the oxide. The amount released is the same for iodine and rare gas. With the gas gap transit is associated a transport time and a probability of trapping for the iodines. In the primary coolant the purification and the radioactive decay are considered. (orig.)

  20. Fuel deposits, chemistry and CANDU® reactor operation

    Roberts, J.G.

    2014-01-01

    'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU® reactor, the first being the Nuclear Power Demonstration - 2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channelled to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5. The difference being that during 'hot conditioning' of CANDU® heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

  1. Operation control device for a nuclear reactor fuel exchanger

    Aida, Takashi.

    1984-01-01

    Purpose: To provide a operation control device for a nuclear reactor fuel exchanger with reduced size and weight capable of optionally meeting the complicated and versatile mode of the operation scope. Constitution: The operation range of a fuel exchanger is finely divided so as to attain the state capable of discriminating between operation-allowable range and operation-inhibitive range, which are stored in a memory circuit. Upon operating the fuel exchanger, the position is detected and a divided range data corresponding to the present position is taken out from the memory circuit so as to determine whether the fuel exchanger is to be run or stopped. Use of reduced size and compact IC circuits (calculation circuit, memory circuit, data latch circuit) and input/output interface circuits or the likes contributes to the size reduction of the exchanger control system to enlarge the floor maintenance space. (Moriyama, K.)

  2. Techniques and results of examination of fission product release from VVER fuel rods with artificial defects and a burnup of ∼60 MWd/kgU at the MIR loop facility

    Burukin, A.; Goryachev, A.; Ilyenko, S.; Izhutov, A.; Konyashov, V.; Shishin, V.; Shulimov, V.; Luzanova, L.; Miglo, V.

    2009-01-01

    Complex of equipment and several techniques for examination of radioactive fission product release from defective fuel rods were developed, prepared and tested at the PV-1 loop facility of the MIR reactor. During the first test, which was conducted at the PV-1 loop facility and aimed at testing of developed equipment and techniques, measurement of radioactive fission product release from an experimental re-fabricated fuel rod with a burnup of ∼60 MWd/kgU and an artificial defect was performed under design-basis steady-state operating conditions of the VVER-1000 reactor. PIE of all main parameters of the experimental defective fuel rod did not reveal any state peculiarities which could be caused by the artificial defect, i.e. fuel and cladding characteristics in the defect area did not differ from the initial ones (before testing) as well as their characteristics in areas distant from the defect; they are typical for fuel rods with a similar irradiation history in the VVER NPP. The gap in the experimental fuel rod was bridged due to close contact between fuel and cladding at increased fuel burnup; it can appreciable reduce release of radioactive fission products into the PV-1 primary coolant. This suggestion and quantitative characteristics of effect of gap bridging in a high-burnup fuel rod on radioactive fission product release should be investigated during the next tests performed at the PV-1 loop facility. Values of radioactive fission product release measured during the first test at the PV-1 loop facility in the MIR reactor will be used for development of an empirical engineering model in order to take into account high burnup effects and their impact on fission product release from fuel and defective fuel rods

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

    Pasupathi, V.; Perrin, J.S.

    1979-01-01

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

  4. Alternate aircraft fuels: Prospects and operational implications

    Witcofski, R. D.

    1977-01-01

    The potential use of coal-derived aviation fuels was assessed. The studies addressed the prices and thermal efficiencies associated with the production of coal-derived aviation kerosene, liquid methane and liquid hydrogen and the air terminal requirements and subsonic transport performance when utilizing liquid hydrogen. The fuel production studies indicated that liquid methane can be produced at a lower price and with a higher thermal efficiency than aviation kerosene or liquid hydrogen. Ground facilities of liquefaction, storage, distribution and refueling of liquid hydrogen fueled aircraft at airports appear technically feasibile. The aircraft studies indicate modest onboard energy savings for hydrogen compared to conventional fuels. Liquid hydrogen was found to be superior to both aviation kerosene and liquid methane from the standpoint of aircraft engine emissions.

  5. Nuclear fuel in water reactors: Manufacturing technology, operational experience and development objectives in the Federal Republic of Germany

    Holzer, R.; Knoedler, D.

    1977-01-01

    The nuclear fuel industry in the Federal Republic of Germany comprises the full range of manufacturing capabilities for pressurized-, boiling- and heavy-water reactor technology. The existing manufacturing companies are Reaktor-Brennelement Union (RBU) and Alkem. RBU makes natural and enriched UO 2 -fuel assemblies, starting with powder preparation. Facilities to produce UO 2 -gadolinia and UO 2 -ThO 2 fuel are also available. Alkem manufactures mixed-oxide UO 2 /PuO 2 fuel and fuel rods. Zircaloy cladding tubes are produced by Nuklearrohr-Gesellschaft (NRG) and Mannesmannroehren-Werke (MRW). Construction of a new fuel manufacturing plant has been announced by Exxon. Supplementary to quality control, an integrated quality assurance system has been established between the reactor vendor's fuel design and engineering division and the existing manufacturing companies for fuel and tubing. Operating experience with LWR and HWR fuel dates back to 1964/65 and has shown good performance. Possible reasons for a small fraction of defective rods could be identified quickly by a fast feedback system incorporating close co-operation between Kraftwerk Union (KWU) and the utilities. KWU combines fuel development, hot-cell and pool-side service facilities as well as fuel technology linked to manufacturing. The responsibility of KWU for core and fuel design, which enabled an integral optimization, was also an important reason for the successful operation and design flexibility. (author)

  6. Reprocessing on the whole fuel cycle operations

    Megy, J.

    1983-11-01

    Spent fuel reprocessing, in France, is become an industrial reality which takes an importance place in several fields: place surely essential in the fuel cycle from the energetic material economy and waste management point of view; place priority in the CEA (Commissariat a l'Energie Atomique) research and development programs; place in the industry where it is an important activity sector with the realizations in progress [fr

  7. Results of trial operation of the WWER advanced fuel assemblies

    Vasilchenko, I.; Dragunov, Y.; Mikhalchuk, A.

    2001-01-01

    The paper describes results from experimental operation of advanced WWER-1000 fuel assemblies (AFA) at five units in Balakovo NPP. Advanced fuel is developed according to the concept of standard WWER-1000 fuel assembly (jacket-free). The new features includes: 1) zirconium guiding channels (alloy E-635 and E-110) and spacer grids (alloy E-110); 2) integrated burnable absorber gadolinium; 3) extended service life of fuel assemblies (FA) and absorber rods (possibility of repair of FA); 4) improved adoption to reactor conditions. Some results of AFA pilot operation of a three year operation are presented and analyses of effectiveness of improvements are made concerning application of zirconium channels and grids; application of integrated burnable absorbers; extension of FA and absorbing rods service life and FA repairability. These new features of WWER-1000 fuel design allow: 1) to reduce the average fuel enrichment to the 3.77% instead of 4.31% in U-235; 2) to reduce the FA axial load in reactor hot state by 40%,; 3) increasing of fuel operation in reactor to the 30000 effective days with possibility to have a 5-year residence time in the reactor. The design of new generation FA for WWER-440 reactors involves few key changes. Fuel inventory in new fuel design is increased due to elongation of fuel stack and reducing the diameter of the central hole. Vibration stability is enhanced as a result of: no-play junction of the fuel rod with the lower grid; change of SG arrangements; strengthening of the lower grid unit; secure of the central tube in the gap. Water-uranium ration is increased. Introduction of all these kinds of modernization in a 5-year fuel cycle reduces fuel component in the energy cost to the 7%

  8. Development of wireless vehicle remote control for fuel lid operation

    Sulaiman, N.; Jadin, M. S.; Najib, M. S.; Mustafa, M.; Azmi, S. N. F.

    2018-04-01

    Nowadays, the evolution of the vehicle technology had made the vehicle especially car to be equipped with a remote control to control the operation of the locking and unlocking system of the car’s door and rear’s bonnet. However, for the fuel or petrol lid, it merely can be opened from inside the car’s cabin by handling the fuel level inside the car’s cabin to open the fuel lid. The petrol lid can be closed by pushing the lid by hand. Due to the high usage of using fuel lever to open the fuel lid when refilling the fuel, the car driver might encounter the malfunction of fuel lid (fail to open) when pushing or pulling the fuel lever. Thus, the main aim of the research is to enhance the operation of an existing car remote control where the car fuel lid can be controlled using two techniques; remote control-based and smartphone-based. The remote control is constructed using Arduino microcontroller, wireless sensors and XCTU software to set the transmitting and receiving parameters. Meanwhile, the smartphone can control the operation of the fuel lid by communicating with Arduino microcontroller which is attached to the fuel lid using Bluetooth sensor to open the petrol lid. In order to avoid the conflict of instruction between wireless systems with the existing mechanical-based system, the servo motor will be employed to release the fuel lid merely after receiving the instruction from Arduino microcontroller and smartphone. As a conclusion, the prototype of the multipurpose vehicle remote control is successfully invented, constructed and tested. The car fuel lid can be opened either using remote control or smartphone in a sequential manner. Therefore, the outcome of the project can be used to serve as an alternative solution to solve the car fuel lid problem even though the problem rarely occurred.

  9. Development of evaluation method of fuel failure fraction during the High Temperature Engineering Test Reactor operation

    Sawa, Kazuhiro; Yoshimuta, Shigeharu; Tobita, Tsutomu; Sato, Masashi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1997-05-01

    The High Temperature Engineering Test Reactor (HTTR) uses coated particles as fuel. During normal operation, short-lived noble gases are mainly released by diffusion from fuel particles with defects in their coating layers (i.e., failed particle). Since noble gases do not plate out on the inner surfaces of primary cooling system, their activities in primary coolant reflect fuel failure fraction in the core. An evaluation method was developed to predict failure fraction of coated fuel particles during normal operation of the HTTR. The method predicts core-average and hot plenum regionwise failure fractions based on the fractional releases, (R/B)s, of noble gases. The (R/B)s are calculated by fission gas concentration measurements in the primary cooling system of the HTTR. Recent fabrication data show that through-coatings failure fraction is extremely low. Then, fractional release from matrix contamination uranium, which is background for accurate evaluation of the fuel failure fraction, should be precisely predicted. This report describes an evaluation method of fuel failure fraction from measurements in the HTTR together with a fission gas release model from fuel compact containing failed particles and matrix contamination uranium. (author)

  10. Software in support of fuel operation in WWERS

    Evdokimov, I.A; Novikov, V.V; Ugrumov, A.V; Shishkin, A.A

    2013-01-01

    A software package comprising computer codes and fuel monitoring tools is under development in Russia in support of WWER fuel operation. The software package includes an expert computer system designed for failure diagnosis in course of reactor operation, prediction of activity evolution in primary coolant and express analysis of pellet-to-cladding mechanical interaction (PCMI) on rod-by-rod basis under normal and transient modes of operation. Coupled with the expert system, the first version of a graphical interface computer program is developed for NPP operating bodies. One of the features of this program is to launch automatically a fuel performance code for a series of detailed calculations for fuel rods with severe PCMI. The particular rods for calculations are determined by the expert system during the express core analysis. A greater attention is paid to recent results in prediction of fuel behavior after a primary failure has occurred. One of the major risks to further operation of leaking fuel comes from secondary fuel degradation due to massive cladding hydriding. Threshold conditions for initiation of secondary hydriding have been found on the basis of physical modeling. Final criteria of secondary failure occurrence were deduced by applying the model to analysis of post-irradiation examinations of leaking WWER fuel. (authors)

  11. Operational reliability testing of FBR fuel in EBR-II

    Asaga, Takeo; Ukai, Shigeharu; Nomura, Shigeo; Shikakura, Sakae

    1991-01-01

    The operational reliability testing of FBR fuel has been conducting in EBR-II as a DOE/PNC collaboration program. This paper reviews the achieved summary of Phase-I test as well as outline of progressing Phase-II test. In Phase-I test, the reliability of FBR fuel pins including 'MONJU' fuel was demonstrated at the event of operational transient. Continued operation of the failed pins was also shown to be feasible without affecting the plant operation. The objectives of the Phase-II test is to extend the data base relating with the operational reliability for long life fuel, and to supply the highly quantitative evaluation. The valuable insight obtained in Phase-II test are considerably expected to be useful toward the achievement of commercial FBR. (author)

  12. Wide Operating Voltage Range Fuel Cell Battery Charger

    Hernandez Botella, Juan Carlos; Mira Albert, Maria del Carmen; Sen, Gokhan

    2014-01-01

    DC-DC converters for fuel cell applications require wide voltage range operation due to the unique fuel cell characteristic curve. Primary parallel isolated boost converter (PPIBC) is a boost derived topology for low voltage high current applications reaching an efficiency figure up to 98...... by two the converter input-to-output voltage gain. This allows covering the conditions when the fuel cell stack operates in the activation region (maximum output voltage) and increases the degrees of freedom for converter optimization. The transition between operating modes is studied because represents...

  13. Fuel deposits, chemistry and CANDU reactor operation

    Roberts, J.G.

    2013-01-01

    'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU reactor, the first being the Nuclear Power Demonstration-2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channel led to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5, and subsequently utilized for each CANDU unit since. The difference being that during 'hot conditioning' of CANDU heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

  14. Operational indices of WWER-1000 fuel assemblies and their improvements

    Vasilchenko, I.; Demin, E.

    1994-01-01

    The most general design features of WWER-1000 fuel assembly are discussed. The following advantages of design are stated as well as their operational confirmation and occurrences: 1) 'packing' density (tight-lattice) of fuel rods within the fuel assemblies; 2) simple handling of fuel assemblies and its small vulnerability; 3) good conditions for coolant mixing; 4) protection of the absorber rods against coolant effect; 5) adaptability to manufacture that provides stable quality. The main operational indices gathered during a ten-year period (1982-1992) at 17 WWER-1000 units in Russia and Ukraine are outlined. Provisions for emergency protection reliability are described. Future directions to improve fuel economy and control rod operability are discussed. 1 fig

  15. Operational indices of WWER-1000 fuel assemblies and their improvements

    Vasilchenko, I; Demin, E [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation)

    1994-12-31

    The most general design features of WWER-1000 fuel assembly are discussed. The following advantages of design are stated as well as their operational confirmation and occurrences: (1) `packing` density (tight-lattice) of fuel rods within the fuel assemblies; (2) simple handling of fuel assemblies and its small vulnerability; (3) good conditions for coolant mixing; (4) protection of the absorber rods against coolant effect; (5) adaptability to manufacture that provides stable quality. The main operational indices gathered during a ten-year period (1982-1992) at 17 WWER-1000 units in Russia and Ukraine are outlined. Provisions for emergency protection reliability are described. Future directions to improve fuel economy and control rod operability are discussed. 1 fig.

  16. Algebra of 2D periodic operators with local and perpendicular defects

    Kutsenko, Anton

    2016-01-01

    We show that 2D periodic operators with local and perpendicular defects form an algebra. We provide an algorithm for finding spectrum for such operators. While the continuous spectral components can be computed by simple algebraic operations on some matrix-valued functions and a few number...

  17. Operational requirements of spherical HTR fuel elements and their performance

    Roellig, K.; Theymann, W.

    1985-01-01

    The German development of spherical fuel elements with coated fuel particles led to a product design which fulfils the operational requirements for all HTR applications with mean gas exit temperatures from 700 deg C (electricity and steam generation) up to 950 deg C (supply of nuclear process heat). In spite of this relatively wide span for a parameter with strong impact on fuel element behaviour, almost identical fuel specifications can be used for the different reactor purposes. For pebble bed reactors with relatively low gas exit temperatures of 700 deg C, the ample design margins of the fuel elements offer the possibility to enlarge the scope of their in-service duties and, simultaneously, to improve fuel cycle economics. This is demonstrated for the HTR-500, an electricity and steam generating 500 MWel eq plant presently proposed as follow-up project to the THTR-300. Due to the low operating temperatures of the HTR-500 core, the fuel can be concentrated in about 70% of the pebbles of the core thus saving fuel cycle costs. Under all design accident conditions fuel temperatures are maintained below 1250 deg C. This allows a significant reduction in the engineered activity barriers outside the primary circuit, in particular for the loss of coolant accident. Furthermore, access to major primary circuit components and the reuse of the fuel elements after any design accident are possible. (author)

  18. Application of robotics in remote fuel fabrication operations

    Nyman, D.H.; Nagamoto, T.T.

    1984-01-01

    The Secure Automated Fabrication (SAF) line, an automated and remotely controlled manufacturing process, is scheduled for startup in 1987 and will produce mixed uranium/plutonium oxide fuel pins for the Fast Flux Test Facility (FFTF). The application of robotics in the fuel fabrication and supporting operations is described

  19. Operational Experience of Nuclear Fuel in Finnish Nuclear Power Plants (with Emphasis on WWER Fuel)

    Teraesvirta, R.

    2009-01-01

    The four operating nuclear reactors in Finland, Loviisa-1 and -2 and Olkiluoto-1 and -2 have now operated approximately 30 years. The overall operational experience has been excellent. Load factors of all units have been for years among the highest in the world. The development of the fuel designs during the years has enabled remarkable improvement in the fuel performance in terms of burnup. Average discharge burnup has increased more than 30 percent in all Finnish reactor units. A systematic inspection of spent fuel assemblies, and especially all failed fuel assemblies, is a good and useful practise employed in Finland. A possibility to inspect the fuel on site using a pool side inspection facility is a relatively economic way to find out root causes of fuel failures and thereby facilitate developing remedies to prevent similar failures in the future

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

    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. Evolution of PHWR fuel transfer system based on operating experience

    Parvatikar, R.S.; Singh, Jaipal; Chaturvedi, P.C.; Bhambra, H.S.

    2006-01-01

    Fuel Transfer System facilitates loading of new fuel into Fuelling Machine, receipt of spent fuel from Fuelling Machine and its further transportation to Storage Bay. To overcome the limitations of transferring a pair of bundles in the single tube Airlock and Transfer Arm in RAPS-1 and 2/MAPS, a new concept of six tube Transfer Magazine was introduced in NAPS. This resulted in simultaneous loading of new fuel from Transfer Magazine into the Fuelling Machine and unloading of spent fuel from the Fuelling Machine through the exchange mode. It further facilitated the parallel/simultaneous operation of refuelling by Fuelling Machines on the reactor and transferring of spent fuel bundles from the Transfer Magazine to the bay. This new design of Fuel Transfer System was adopted for all standardised 220 MWe PHWRs. Based on the experience gained in 220 MWe PHWRs in the area of operation and maintenance, a number of improvements have been carried out over the years. These aspects have been further strengthened and refined in the Fuel Transfer System of 540 MWe units. The operating experience of the system indicates that the presence of heavy water in the Transfer Magazine poses limitations in its maintenance in the Fuel Transfer room. Further, Surveillance and maintenance of large number of under water equipment and associated valves, rams and underwater sensors is putting extra burden on the O and M efforts. A new concept of mobile light water filled Transfer Machine has been evolved for proposed 700 MWe PHWR units to simplify Fuel Transfer System. This has been made possible by adopting snout level control in the Fuelling Machine, elimination of Shuttle Transport System and locating the Storage Bay adjacent to the Reactor Building. This paper describes the evolution of Fuel Transfer System concepts and various improvements based on the experience gained in the operation and maintenance of the system. (author)

  2. Fuel design and operational experience in Loviisa NPP, future trends in fuel issues

    Terasvirta, R.

    2001-01-01

    This paper summarizes the past operational experience of nuclear fuel with reference to most significant design changes during the years. In general, the fuel behaviour in Loviisa NPP in terms of leaking fuel assemblies has been good. The major improvements by fuel design changes in Lovissa NPP, including rod elongation margin, change in the pellet design and manufacturing process, upper grid modifications, change of material in the spacer grids and reduction of the shroud tube thickness are discussed and related to the number of failed fuel assemblies. The detailed investigation of fuel failure rates as function of different fuel and operation characteristics allows to classify the leaking fuel assemblies according to the cause of failure. In a brief discussion concerning new changes in the safety guide for nuclear design limits, re-issued by the Finnish Safety Authority (STUK), the frequencies for class 1 and class 2 accidents are determined. Another change in this guide is the introduction of design limits for the number of fuel rods experiencing DNB in class 1 accidents and number of failed rods in class 2 accidents. It is concluded that as far as normal operation is concerned, there seems to be sufficiently large margin between present operational limits in Loviisa and the design limits. The real limits do not come from fuel behaviour in the normal operation or operational occurrences but from the accident behaviour. At the moment, fuel assembly burnup extension beyond 45 MWd/kgU is clearly out of the question before further information and positive results are obtained on high burnup fuel behaviour in accident conditions

  3. LOFT fuel design and operating experience

    Russell, M.L.

    1978-01-01

    The purpose of the LOFT fuel is to provide a pressurized water reactor core that has (1) test instrumentation for measurement of core conditions and (2) materials and geometric features to ensure heat transfer, hydraulic, mechanical, chemical, metallurgical and nuclear behaviors are typical of large pressurized water reactors (LPWRS) during the loss-of-coolant accident (LOCA) sequence. The LOFT core is unique because it is designed for exposure to several LOCAs without loss of function

  4. Typical IAEA operations at a fuel fabrication plant

    Morsy, S.

    1984-01-01

    The IAEA operations performed at a typical Fuel Fabrication Plant are explained. To make the analysis less general the case of Low Enriched Uranium (LEU) Fuel Fabrication Plants is considered. Many of the conclusions drawn from this analysis could be extended to other types of fabrication plants. The safeguards objectives and goals at LEU Fuel Fabrication Plants are defined followed by a brief description of the fabrication process. The basic philosophy behind nuclear material stratification and the concept of Material Balance Areas (MBA's) and Key Measurement Points (KMP's) is explained. The Agency operations and verification methods used during physical inventory verifications are illustrated

  5. Fuel element structure - design, production and operational behaviour

    Pott, G.; Dietz, W.

    1985-01-01

    The lectures held at the meeting of the fuel element section of the Kerntechnische Gesellschaft gives a survey of developments in fuel element structure design for PWR-type, BWR-type and fast breeder reactors. For better utilization of the fuel, concepts have been developed for re-usable, removable and thus repairable fuel elements. Furthermore, the manufacturing methods for fuel element structures were refined to achieve better quality and more efficient manufacturing methods. Statements on the dimensional behaviour and on the mechanical stability of fuel element structures in normal and accident operation could be made on the basis of post-irradiation inspections. Finally, the design, manufacture and irradiation behaviour of graphite reflectors in HTGR-type reactors are described. The 12 lectures have been recorded in the data base separately. (RF) [de

  6. Improvement of the vibration of the test fuel(Type-B) with a guide tube under operational condition

    Sohn, Dong Seung; Yim, Jeong Sik; Lim, I. C. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-04-01

    The Type-B test fuel for the Hanaro has a flexible guide tube on top of the fuel to lead and guide the instrumentation wires. Depending on the flow condition in the reactor, the fuel is susceptible to vibration. During the test operation of the fuel, a fairly large amplitude vibration was observed and the possibility of flow tube contact with adjacent flow tubes, due to the excessive vibration of the fuel, and consequent wear or defect of the flow tubes were raised. Thus, to know the vibration characteristics as well as whether the flow tube contact each other, analyses of the Type-B fuel the dummy fuel were performed by BEVIRA and ANSYS. Besides the analyses, vibration tests using the dummy fuel in air and with Type-B fuel in the core at zero power under operational flow condition were executed. The results from the analyses were compared with those from tests to validate the analyses. From the deflection test of the dummy fuel in air to get the maximum displacement of the flow tube at the top, the flow tube were found to contact each other. For the prevention of the contact of the flow tubes caused by the excessive vibration of the guide tube, an additional support to the guide tube was proposed. With the additional support, analysis and in core vibration test under operational flow condition were conducted and there found to be no excessive vibration any more. 6 refs., 16 figs., 6 tabs. (Author)

  7. LOFT fuel design and operating experience

    Russell, M.L.

    1979-01-01

    The objective of the LOFT fuel design and fabrication effort was to provide a pressurized water reactor core that has (1) materials and geometric features to ensure that heat transfer, hydraulic, mechanical, chemical, metallurgical and nuclear behaviors are typical of large pressurized water reactors (PWR) during the loss-of-coolant accident (LOCA) sequence and (2) test instrumentation for measurement of core conditions. The LOFT core is unique because it is designed for exposure to several LOCAs without loss of function. This paper summarizes the design effort and extent to which the design objectives have been achieved

  8. Energy Fuels Nuclear, Inc. Arizona Strip Operations

    Pool, T.C.

    1993-01-01

    Founded in 1975 by uranium pioneer, Robert W. Adams, Energy Fuels Nuclear, Inc. (EFNI) emerged as the largest US uranium mining company by the mid-1980s. Confronting the challenges of declining uranium market prices and the development of high-grade ore bodies in Australia and Canada, EFNI aggressively pursued exploration and development of breccia-pipe ore bodies in Northwestern Arizona. As a result, EFNI's production for the Arizona Strip of 18.9 million pounds U 3 O 8 over the period 1980 through 1991, maintained the company's status as a leading US uranium producer

  9. How to operate safely steam generators with multiple tube through-wall defects

    Hernalsteen, P.

    1993-01-01

    For a Nuclear Power Plant (NPP) of the Pressurized Water Reactor (PWR) type, the Steam Generator (SG) tube bundle represents the major but also the thinnest part of the primary pressure boundary. To the extent that no tube material has yet been identified to be immune to corrosion, defects may initiate in service and easily propagate through wall. While not a desirable feature, a Through Wall Deep (TWD) defect does not necessarily pose a threat to either the structural integrity or leaktightness and this paper shows how SG can (and indeed, do) operate safely and reliably while having many tubes affected by deep and even TWD defects

  10. Physics operating experience and fuel management of RAPS-1

    Nakra, A.N.; Purandare, H.D.; Srinivasan, K.R.; Rastogi, B.P.

    1976-01-01

    Rajasthan Atomic Power Station Unit-1 achieved criticality on August 11, 1972. Thereafter the reactor was brought to power, in November, 1972. Due to non-availability of the depleted fuel, the loading of which was necessary to obtain full power to begin with, the core was loaded with all natural uranium fuel and only 70% of the full power could be achieved. During the reactor operation for the last three years, the reactor has seen more than one effective full power year and about 1400 fresh fuel bundles have been loaded in the core. The reactor was subjected to about 150 power cycles resulting in more than 30% variation in operating power level and about 10 fuel bundles have failed. For satisfactory fuel management and refuelling decisions, a three dimensional simulator TRIVENI was developed. This was extensively tested during the start-up experiments and was found to be a satisfactory tool for day to day operation of the plant. In this paper, a brief account of analysis of the start-up experiments, approach to full power, power distortions and flux peaking, fuel management service and analysis of the failed fuel data has been given. (author)

  11. Mutual influences of reactor operation and fuel cycle management

    Lewiner, C.; Schaerer, R.

    1989-01-01

    OPEN (Organisation des Producteurs d'Energie Nucleaire) comprises the electricity producers from seven European countries which now operate or intend to operate nuclear power plants. Its activities include the study of technical, economic and legal subjects related to nuclear electricity. A continuous analysis of the fuel cycle market has been pursued within OPEN for almost 15 years. For the past few years, OPEN has also been concerned with the subject of fuel management in the reactors operated by its members. The purpose of this effort was to obtain an overall picture of possible fuel improvements and to evaluate the effects, in particular the economic ones, of diverse fuel reload managements and of reprocessed uranium and plutonium recycling. The conclusions of this study are as follows: Increase in burn-ups produces notable savings in electricity generating costs. It also permits adaptation of fuel loading mode to the desirable irradiation campaign length. This allows for better management of the country's overall means of electricity generation (nuclear, fossil-fuelled or hydro plants), and adjustment to the electrical demand. These new reload schemes have various impacts on natural uranium consumption and enrichment, but, above all, they affect directly all fuel cycle operations linked to the number of assemblies (fabrication, reprocessing, etc.). 6 figs

  12. Swedish spent fuel management systems, facilities and operating experiences

    Vogt, J.

    1998-01-01

    About 50% of the electricity in Sweden is generated by means of nuclear power from 12 LWR reactors located at four sites and with a total capacity of 10,000 MW. The four utilities have jointly created SKB, the Swedish Nuclear Fuel and Waste Management Company, which has been given the mandate to manage the spent fuel and radioactive waste from its origin at the reactors to the final disposal. SKB has developed a system for the safe handling of all kinds of radioactive waste from the Swedish nuclear power plants. The keystones now in operation of this system are a transport system, a central interim storage facility for spent nuclear fuel (CLAB), a final repository for short-lived, low and intermediate level waste (SFR). The remaining, system components being planned are an encapsulation plant for spent nuclear fuel and a deep repository for encapsulated spent fuel and other long-lived radioactive wastes. (author)

  13. Cask operation and maintenance for spent fuel storage

    Lee, J.S. [International Atomic Energy Agency, Vienna (Austria)

    2004-07-01

    Interim storage is an essential platform for any option to be chosen later as an endpoint for spent fuel management. In view of such a circumstance, the most imminent service required for the spent fuel management worldwide is to provide adequate storage for the future spent fuel inventory arising either from the continued operation of nuclear power plants or from the removal of spent fuel in preparation for plant decommissioning. While the bulk of the global inventory of spent fuel are still stored in AR pools, dry storage has become a prominent alternative especially for newly built AFR facilities, with more than 17,000 t HM already stored in dry storage facilities worldwide. Storage in cask under inert conditions has become the preferred option, given the advantages including passive cooling features and modular mode of capacity increase. In terms of economics, dry storage is particularly propitious for long-term storage in that operational costs are minimized by the passive cooling features. The trend toward dry storage, especially in cask type, is likely to continue with an implication that and the supply will closely follow the increasing demand for storage by incremental additions of casks to the effect of minimizing cost penalty of the idle capacities typical of pool facilities. A variety of storage systems have been developed to meet specific requirements of different reactor fuels and a large number of designs based on these generic technologies are now available for the spent fuel containers (horizontal, vertical etc) and storage facilities. Multi-purpose technologies (i.e. a single technology for storage, transportation and disposal) have also been studied. Recent concern on security measures for protection of spent fuel has prompted a consideration on the possibility of placing storage facility underground. The future evolution of requirements and technologies will bring important impacts on cask operation and maintenance for spent fuel storage.

  14. Cask operation and maintenance for spent fuel storage

    Lee, J.S.

    2004-01-01

    Interim storage is an essential platform for any option to be chosen later as an endpoint for spent fuel management. In view of such a circumstance, the most imminent service required for the spent fuel management worldwide is to provide adequate storage for the future spent fuel inventory arising either from the continued operation of nuclear power plants or from the removal of spent fuel in preparation for plant decommissioning. While the bulk of the global inventory of spent fuel are still stored in AR pools, dry storage has become a prominent alternative especially for newly built AFR facilities, with more than 17,000 t HM already stored in dry storage facilities worldwide. Storage in cask under inert conditions has become the preferred option, given the advantages including passive cooling features and modular mode of capacity increase. In terms of economics, dry storage is particularly propitious for long-term storage in that operational costs are minimized by the passive cooling features. The trend toward dry storage, especially in cask type, is likely to continue with an implication that and the supply will closely follow the increasing demand for storage by incremental additions of casks to the effect of minimizing cost penalty of the idle capacities typical of pool facilities. A variety of storage systems have been developed to meet specific requirements of different reactor fuels and a large number of designs based on these generic technologies are now available for the spent fuel containers (horizontal, vertical etc) and storage facilities. Multi-purpose technologies (i.e. a single technology for storage, transportation and disposal) have also been studied. Recent concern on security measures for protection of spent fuel has prompted a consideration on the possibility of placing storage facility underground. The future evolution of requirements and technologies will bring important impacts on cask operation and maintenance for spent fuel storage

  15. Analysis of WWER-440 fuel performance under normal operating conditions

    Gunduz, Oe; Koese, S; Akbas, T [Atomenerjisi Komisyonu, Ankara (Turkey); Colak, Ue [Ankara Nuclear Research and Training Center (Turkey)

    1994-12-31

    FRAPCON-2 code originally developed for LWR fuel behaviour simulation is used to analyse the WWER-440 fuel rod behaviour at normal operational conditions. The code is capable of utilizing different models for mechanical analysis and gas release calculations. Heat transfer calculations are accomplished through a collocation technique by the method of weighted residuals. Temperature and burnup element properties are evaluated using MATPRO package. As the material properties of Zr-1%Nb used as cladding in WWER-440s are not provided in the code, Zircaloy-4 is used as a substitute for Zr-1%Nb. Mac-Donald-Weisman model is used for gas release calculation. FRACAS-1 and FRACAS-2 models are used in the mechanical calculations. It is assumed that the reactor was operated for 920 days (three consecutive cycles), the burnup being 42000 Mwd/t U. Results of the fuel rod behaviour analysis are given for three axial nodes: bottom node, central node and top node. The variations of the following characteristic fuel rod parameters are studied through the prescribed power history: unmoved gap thickness, gap heat transfer coefficient, fuel axial elongation, cladding axial elongation, fuel centerline temperature and ZrO-thickness at cladding surface. The value of each parameter is calculated as a function of the effective power days for the three nodes by using FRACAS-1 and FRACAS-2 codes for comparison.The results show that calculations with deformable pellet approximation with FRACAS-II model could provide better information for the behaviour of a typical fuel rod. Calculations indicate that fuel rod failure is not observed during the operation. All fuel rod parameters investigated are found to be within the safety limits. It is concluded, however, that for better assessment of reactor safety these calculations should be extended for transient conditions such as LOCA. 1 tab., 10 figs., 4 refs.

  16. Introduction of HTR-PM Operation and Fuel Management System

    Liu Fucheng; Luo Yong; Gao Qiang

    2014-01-01

    There is a big difference between High Temperature Gas-cooled Reactor Pebble-modules Demonstration Project(HTR-PM) and PWR in operation mode. HTR-PM is a continually refuelled reactor, and the operation and fuel management of it, which affect each other, are inseparable. Therefore, the analysis of HTR-PM fuel management needs to be carried out “in real time”. HTR-PM operation and fuel management system is developed for on-power refuelling mode of HTR-PM. The system, which calculates the core neutron flux and power distribution, taking high-temperature reactor physics analysis software-VSOP as a basic tool, can track and predict the core state online, and it has the ability to restructure core power distribution online, making use of ex-core detectors to correct and check tracking calculation. Based on the ability to track and predict, it can compute the core parameters to provide support for the operation of the reactor. It can also predict the operation parameters of the reactor to provide reference information for the fuel management.The contents of this paper include the development purposes, architecture, the main function modules, running process, and the idea of how to use the system to carry out HTR-PM fuel management. (author)

  17. Fuel Consumption and Emissions from Airport Taxi Operations

    Jung, Yoon

    2010-01-01

    Developed a method to calculate fuel consumption and emissions of phases of taxi operations. Results at DFW showed that up to 18% of fuel can be saved by eliminating stop-and-go situations. Developed an energy efficient and environmentally friendly surface concept: Spot and Runway Departure Advisory (SARDA) tool. The SARDA tool has been identified as a potential candidate for a technology transfer to the FAA.

  18. Polymer electrolyte fuel cells physical principles of materials and operation

    Eikerling, Michael

    2014-01-01

    The book provides a systematic and profound account of scientific challenges in fuel cell research. The introductory chapters bring readers up to date on the urgency and implications of the global energy challenge, the prospects of electrochemical energy conversion technologies, and the thermodynamic and electrochemical principles underlying the operation of polymer electrolyte fuel cells. The book then presents the scientific challenges in fuel cell research as a systematic account of distinct components, length scales, physicochemical processes, and scientific disciplines. The main part of t

  19. Effects of fabrication requirements on fuel performance in relation to operating conditions. The views of Electricite de France (EDF)

    Ponticq, M.; Richer, P.; Scribe, G.

    1979-01-01

    Because of the operational constraints relating to fuel behaviour, imperfect knowledge of the behaviour of a defective fuel assembly and, in the near future, the need to adapt reactor power to grid following (load following and remote control), EDF is aiming to reduce the present rate of fuel failure. While the phenomena affecting fuel behaviour have now been listed and analysed, the efforts at reducing their consequences have yet to be completed. This can be achieved, firstly, by reducing or eliminating fabrication defects, which are responsible for failure at the beginning of fuel life, through establishment of a good quality assurance organization and the search for still higher efficiency of quality control and fabrication equipment, and secondly, by developing fabrication techniques minimizing in particular cladding-pellet interactions and the stress corrosion of the cladding, which are responsible for fuel failure as from mid-life. However, the reactor operating conditions likely to apply in the near future may lead, for a given fuel configuration, to a re-evaluation of the fabrication parameters of cladding and UO 2 pellets. (author)

  20. System for controlling the operating temperature of a fuel cell

    Fabis, Thomas R.; Makiel, Joseph M.; Veyo, Stephen E.

    2006-06-06

    A method and system are provided for improved control of the operating temperature of a fuel cell (32) utilizing an improved temperature control system (30) that varies the flow rate of inlet air entering the fuel cell (32) in response to changes in the operating temperature of the fuel cell (32). Consistent with the invention an improved temperature control system (30) is provided that includes a controller (37) that receives an indication of the temperature of the inlet air from a temperature sensor (39) and varies the heat output by at least one heat source (34, 36) to maintain the temperature of the inlet air at a set-point T.sub.inset. The controller (37) also receives an indication of the operating temperature of the fuel cell (32) and varies the flow output by an adjustable air mover (33), within a predetermined range around a set-point F.sub.set, in order to maintain the operating temperature of the fuel cell (32) at a set-point T.sub.opset.

  1. Fuel conditioning facility electrorefiner cadmium vapor trap operation

    Vaden, D. E.

    1998-01-01

    Processing sodium-bonded spent nuclear fuel at the Fuel Conditioning Facility at Argonne National Laboratory-West involves an electrometallurgical process employing a molten LiCl-KCl salt covering a pool of molten cadmium. Previous research has shown that the cadmium dissolves in the salt as a gas, diffuses through the salt layer and vaporizes at the salt surface. This cadmium vapor condenses on cool surfaces, causing equipment operation and handling problems. Using a cadmium vapor trap to condense the cadmium vapors and reflux them back to the electrorefiner has mitigated equipment problems and improved electrorefiner operations

  2. Study of Exhaust Emissions Reduction of a Diesel Fuel Operated Heater During Transient Mode of Operation

    Miklánek Ľubomír

    2014-10-01

    Full Text Available Diesel fuel operated heaters (FOHs are generally used as an independent heat source for any system in which a diesel fuel and battery power is available. Based on the fact that future engines will become even more efficient and thus less waste heat will be available to heat the passenger compartment, independent heat sources will be even more necessary.

  3. Influence of the fuel operational parameters on the aluminium cladding quality of discharged fuel

    Chwaszczewski, S.; Czajkowski, W.; Borek-Kruszewska, E. [Institute of Atomic Energy, Otwock Swierk (POLAND)

    2002-07-01

    In the last two years, the new MR6 type fuel containing 1550 g of U with 36% enrichment has been loaded into MARIA reactor core. Its aluminium cladding thickness is 0,6 mm and typical burnup -about 4080 MWh (as compared to 2880 MWh for the 80% enriched fuel used). However, increased fission product release from these assemblies was observed near the end of its operational time. The results presented earlier [1] show that the corrosion behaviour of aluminium cladding depends on the conditions of fuel operation in the reactor. The corrosion process in the aluminum of fuel cladding proceeds faster then in the aluminum of constructional elements. This tendency was also observed in MR-6/80% and in WWR- SM fuel assemblies. Therefore the visual tests of discharged MR-6/36% fuel elements were performed. Some change of appearance of aluminum cladding was observed, especially in the regions with large energy generation i.e. in the centre of reactor core and in the strong horizontal gradient of neutron flux. In the present paper, the results of visual investigation of discharged fuel assemblies are presented. The results of the investigation are correlated with the operational parameters. (author)

  4. Correlation between defect transition levels and thermoelectric operational temperature of doped CrSi2

    Singh, Abhishek; Pandey, Tribhuwan

    2014-03-01

    The performance of a thermoelectric material is quantified by figure of merit ZT. The challenge in achieving high ZT value requires simultaneously high thermopower, high electrical conductivity and low thermal conductivity at optimal carrier concentration. So far doping is the most versatile approach used for modifying thermoelectric properties. Previous studies have shown that doping can significantly improve the thermoelectric performance, however the tuning the operating temperature of a thermoelectric device is a main issue. Using first principles density functional theory, we report for CrSi2, a linear relationship between thermodynamic charge state transition levels of defects and temperature at which thermopower peaks. We show for doped CrSi2 that the peak of thermopower occurs at the temperature Tm, which corresponds to the position of defect transition level. Therefore, by modifying the defect transition level, a thermoelectric material with a given operational temperature can be designed. The authors thankfully acknowledge support from ADA under NpMASS.

  5. Fission gas release in LWR fuel measured during nuclear operation

    Appelhans, A.D.; Skattum, E.; Osetek, D.J.

    1980-01-01

    A series of fuel behavior experiments are being conducted in the Heavy Boiling Water Reactor in Halden, Norway, to measure the release of Xe, Kr, and I fission products from typical light water reactor design fuel pellets. Helium gas is used to sweep the Xe and Kr fission gases out of two of the Instrumented Fuel Assembly 430 fuel rods and to a gamma spectrometer. The measurements of Xe and Kr are made during nuclear operation at steady state power, and for 135 I following reactor scram. The first experiments were conducted at a burnup of 3000 MWd/t UO 2 , at bulk average fuel temperatures of approx. 850 K and approx. 23 kW/m rod power. The measured release-to-birth ratios (R/B) of Xe and Kr are of the same magnitude as those observed in small UO 2 specimen experiments, when normalized to the estimated fuel surface-to-volume ratio. Preliminary analysis indicates that the release-to-birth ratios can be calculated, using diffusion coefficients determined from small specimen data, to within a factor of approx. 2 for the IFA-430 fuel. The release rate of 135 I is shown to be approximately equal to that of 135 Xe

  6. Performance of Combustion Engineering fuel in operating PWRs

    Andrews, M.G.; Freeburn, H.R.; Wohlsen, W.D.

    1979-01-01

    Performance data on fuel assembly components from seven (7) operating reactors are presented, and discussed in detail where potential problems have occurred and been resolved. Fuel rod performance has continually improved over the last four (4) years with the gradual changeover to the current C-E fuel design. The reliability level is estimated at better than 99.99%, based on activity levels obtained through January 1979 at each plant. Control rod guide tubes have shown various degrees of wear caused by vibration of the control rods in their fully-withdrawn position. The retrofit of wear sleeves within the top portion of the affected guide tubes during routine refueling has permitted the use of these fuel assemblies with no significant loss in performance or safety margins

  7. A Mobile Robot for Emergency Operation of Fuel Exchange Machine

    Seo, Yongchil; Lee, Sunguk; Kim, Changhoi; Shin, Hochul; Jung, Seungho; Choi, Changhwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    A Pressurized Heavy Water Reactor (PHWR) uses a heavy water as the coolant and moderator because it does not attenuate the neutron inside the reactor, which makes it possible to use natural uranium for nuclear fuels. However, since the uranium ratio is too low within the natural uranium, the reactor should be refueled everyday while the reactor is working. For that purpose, there is a fuel exchange machine. However as the time passes by, the durability and reliability become a problem. While the fuel handling machine exchanges the reactor fuel, it can be stuck to the pressure tube attached in the Calandra. Although this kind of situation is rarely happen, it can make the reactor be shutdown for normalizing the operation. Since the refueling is performed while the reactor is working, the radiation level is extremely high and the machine can be located at a high position up to nine meters from the floor, that is, the human worker can not approach the machine, so the fuel handling machine should be released remotely. To cope with this situation, the fuel handling machine has a manual drive mechanism at the rear side of it as shown in the circled images. If the worker can handle these manual drive mechanisms, the fuel handling machine can be released form the pressure tube. The KAERI had developed a long-reach manipulator system with a telescophic mast mechanism which can be deployed in the basement of the reactor room and manipulate the manual lever of the fuel exchange machine. Since the manipulator is located in the basement, there are several problems for its application such that the plug hole should be removed before the operation and the vibration of the mast mechanism make it difficult to locate the end effecter of the manipulator.

  8. A Mobile Robot for Emergency Operation of Fuel Exchange Machine

    Seo, Yongchil; Lee, Sunguk; Kim, Changhoi; Shin, Hochul; Jung, Seungho; Choi, Changhwan

    2007-01-01

    A Pressurized Heavy Water Reactor (PHWR) uses a heavy water as the coolant and moderator because it does not attenuate the neutron inside the reactor, which makes it possible to use natural uranium for nuclear fuels. However, since the uranium ratio is too low within the natural uranium, the reactor should be refueled everyday while the reactor is working. For that purpose, there is a fuel exchange machine. However as the time passes by, the durability and reliability become a problem. While the fuel handling machine exchanges the reactor fuel, it can be stuck to the pressure tube attached in the Calandra. Although this kind of situation is rarely happen, it can make the reactor be shutdown for normalizing the operation. Since the refueling is performed while the reactor is working, the radiation level is extremely high and the machine can be located at a high position up to nine meters from the floor, that is, the human worker can not approach the machine, so the fuel handling machine should be released remotely. To cope with this situation, the fuel handling machine has a manual drive mechanism at the rear side of it as shown in the circled images. If the worker can handle these manual drive mechanisms, the fuel handling machine can be released form the pressure tube. The KAERI had developed a long-reach manipulator system with a telescophic mast mechanism which can be deployed in the basement of the reactor room and manipulate the manual lever of the fuel exchange machine. Since the manipulator is located in the basement, there are several problems for its application such that the plug hole should be removed before the operation and the vibration of the mast mechanism make it difficult to locate the end effecter of the manipulator

  9. Increasing Fuel Efficiency of Direct Methanol Fuel Cell Systems with Feedforward Control of the Operating Concentration

    Youngseung Na

    2015-09-01

    Full Text Available Most of the R&D on fuel cells for portable applications concentrates on increasing efficiencies and energy densities to compete with other energy storage devices, especially batteries. To improve the efficiency of direct methanol fuel cell (DMFC systems, several modifications to system layouts and operating strategies are considered in this paper, rather than modifications to the fuel cell itself. Two modified DMFC systems are presented, one with an additional inline mixer and a further modification of it with a separate tank to recover condensed water. The set point for methanol concentration control in the solution is determined by fuel efficiency and varies with the current and other process variables. Feedforward concentration control enables variable concentration for dynamic loads. Simulation results were validated experimentally with fuel cell systems.

  10. Fuel cell bus operation, system investigation H2 bus

    Anon.

    1992-01-01

    The WP covers two tasks: - Prepartion of Technical Catalogue: In cooperation with ICIL, AR have compiled a technical catalogue, providing the impartial descriptions, both of existing technology and regulations, and the likely future developments of these, as to remedy the first problem faced by a potential hydrogen bus fleet operator viz the absence of an impartial description of the available vehicle and fuels systems together with the absence of a description of regulatory and safety factors which need consideration. - Fuel Cell Bus Operation - System Investigation H 2 Bus: The application of fuel cell electric generation systems to hybrid electrical buses or electrical busses without any storage system on board is considered. The task will cover safety and environmental aspects, a cost estimate and a market evaluation. (orig.)

  11. Operational impacts of low-enrichment uranium fuel conversion on the Ford Nuclear Reactor

    Bernal, F.E.; Brannon, C.C.; Burgard, N.E.; Burn, R.R.; Cook, G.M.; Simpson, P.A.

    1985-01-01

    The University of Michigan Department of Nuclear Engineering and the Michigan Memorial-Phoenix Project have been engaged in a cooperative effort with Argonne National Laboratory to test and analyze low-enrichment fuel in the Ford Nuclear Reactor (FNR). The effort was begun in 1979, as part of the Reduced Enrichment Research and Test Reactor Program, to demonstrate on a whole-core basis the feasibility of enrichment reduction from 93% to <20% in Materials Test Reactor-type fuel designs. The first low-enrichment uranium (LEU) core was loaded into the FNR and criticality was achieved on December 8, 1981. The final LEU core was established October 11, 1984. No significant operational impacts have resulted from conversion of the FNR to LEU fuel. Thermal flux in the core has decreased slightly; thermal leakage flux has increased. Rod worths, temperature coefficient, and void coefficient have changed imperceptibly. Impressions from the operators are that power defect has increased slightly and that fuel lifetime has increased

  12. 14 CFR 121.639 - Fuel supply: All domestic operations.

    2010-01-01

    ... nontransport category airplanes type certificated after December 31, 1964, to fly for 30 minutes at normal... § 121.639 Fuel supply: All domestic operations. No person may dispatch or take off an airplane unless it has enough fuel— (a) To fly to the airport to which it is dispatched; (b) Thereafter, to fly to and...

  13. The nuclear fuel cycle associated with the operation of nuclear ...

    The nuclear power option has been mentioned as an alternative for Ghana but the issue of waste management worries both policy makers and the public. In this paper, the nuclear fuel cycle associated with the operation of nuclear power plants (NPPs) for electric power generation has been extensively reviewed. Different ...

  14. Computerised programming of the Dragon reactor fuel handling operations

    Butcher, P.

    1976-11-01

    Two suites of FORTRAN IV computer programs have been written to produce check lists for the operation of the two remote control fuel handling machines of the Dragon Reactor. This document describes the advantages of these programs over the previous manual system of writing check lists, and provides a detailed guide to the programs themselves. (author)

  15. Fuel Property, Emission Test, and Operability Results from a Fleet of Class 6 Vehicles Operating on Gas-to-Liquid Fuel and Catalyzed Diesel Particle Filters

    Alleman, T. L.; Eudy, L.; Miyasato, M.; Oshinuga, A.; Allison, S.; Corcoran, T.; Chatterjee, S.; Jacobs, T.; Cherrillo, R. A.; Clark, R.; Virrels, I.; Nine, R.; Wayne, S.; Lansing, R.

    2005-11-01

    A fleet of six 2001 International Class 6 trucks operating in southern California was selected for an operability and emissions study using gas-to-liquid (GTL) fuel and catalyzed diesel particle filters (CDPF). Three vehicles were fueled with CARB specification diesel fuel and no emission control devices (current technology), and three vehicles were fueled with GTL fuel and retrofit with Johnson Matthey's CCRT diesel particulate filter. No engine modifications were made.

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

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

    2012-01-01

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

  17. International co-operation in the supply of nuclear fuel and fuel cycle services

    Sievering, N.F. Jr.

    1977-01-01

    Recent changes in the United States' nuclear policy, in recognition of the increased proliferation risk, have raised questions of US intentions in international nuclear fuel and fuel-cycle service co-operation. This paper details those intentions in relation to the key elements of the new policy. In the past, the USA has been a world leader in peaceful nuclear co-operation with other nations and, mindful of the relationships between civilian nuclear technology and nuclear weapon proliferation, remains strongly committed to the Non-Proliferation Treaty, IAEA safeguards and other elements concerned with international nuclear affairs. Now, in implementing President Carter's nuclear initiatives, the USA will continue its leading role in nuclear fuel and fuel-cycle co-operation in two ways, (1) by increasing its enrichment capacity for providing international LWR fuel supplies and (2) by taking the lead in solving the problems of near and long-term spent fuel storage and disposal. Beyond these specific steps, the USA feels that the international community's past efforts in controlling the proliferation risks of nuclear power are necessary but inadequate for the future. Accordingly, the USA urges other similarly concerned nations to pause with present developments and to join in a programme of international co-operation and participation in a re-assessment of future plans which would include: (1) Mutual assessments of fuel cycles alternative to the current uranium/plutonium cycle for LWRs and breeders, seeking to lessen proliferation risks; (2) co-operative mechanisms for ensuring the ''front-end'' fuel supply including uranium resource exploration, adequate enrichment capacity, and institutional arrangements; (3) means of dealing with short-, medium- and long-term spent fuel storage needs by means of technical co-operation and assistance and possibly establishment of international storage or repository facilities; and (4) for reprocessing plants, and related fuel

  18. MELOX fuel fabrication plant: Operational feedback and future prospects

    Hugelmann, D.; Greneche, D.

    2000-01-01

    As of December 1, 1998, 32 Europeans LWRs are loaded with MOX fuel. It clearly means that plutonium recycling in MOX fuels is a mature industry, with successful operational experience in fabrication plants in some European countries, especially in France. Indeed, the recycling of plutonium generated in LWRs is one of the objectives of the full Reprocessing-Conditioning-Recycling (RCR) strategy chosen by France in the 70's. The most impressive results of this strategy, is the fact that 31 of the 32 reactors are loaded with MOX fuels supplied by the COGEMA Group from the same efficient fabrication process, the MIMAS process, improved for the MELOX plant to become the A-MIMAS process. In France, 17 reactors are already loaded and 11 additional reactors are technically suited to do so. Indeed, the EDF MOX program plans to use MOX in 28 of its 57 reactors. An EDF 900 MWe reactor core contains 157 assemblies of 264 rods each. 52 fuel assemblies per year are necessary for a 'UO 2 3-batches-MOX 3-batches' core management. In this case, a third of the UO 2 and a third of the MOX assemblies are replaced yearly, that means 36 UO 2 fuel assemblies and 16 MOX fuel assemblies. Some MOX fuelled reactors have now switched from the previously described core management to a so-called 'hybrid core management'. In this case, a quarter of UO 2 assemblies is replaced yearly. The first EDF reactor loaded with MOX fuel was Saint-Laurent B1, in 1987. The in-core experience, based on several hundred assemblies loaded, with reloading on a 1/3 cycle basis, shows that there is no operational difference between UO 2 and MOX fuels, both in terms of performance and safety. MOX fueling of 900 MWe EDF's PWRs, with a limited in-core MOX ratio of 30%, has needed only minor adaptations, such as addition of control rods, modification of the boron concentration in the cooling system and precaution against radiation exposure, easy to set up (optimisation of the fresh MOX fuel handling process, remote

  19. Evaluation of LWR fuel rod behavior under operational transient conditions

    Nakamura, M.; Hiramoto, K.; Maru, A.

    1984-01-01

    To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding. The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8x8 RJ fuel rod temperatures under power ramp conditions. (orig.)

  20. MOX and UOX PWR fuel performances EDF operating experience

    Provost, Jean-Luc; Debes, Michel

    2005-01-01

    Based on a large program of experimentations implemented during the 90s, the industrial achievement of new FAs designs with increased performances opens up new prospects. The currently UOX fuels used on the 58 EDF PWR units are now authorized up to a maximum FA burn-up of 52 GWd/t with a large experience from 45 to 50 GWd/t. Today, the new products, along with the progress made in the field of calculation methods, still enable to increase further the fuel performances with respect to the safety margins. Thus, the conditions are met to implement in the next years new fuel managements on each NPPs series of the EDF fleet with increased enrichment (up to 4.5%) and irradiation limits (up to 62 GWd/t). The recycling of plutonium is part of EDF's reprocessing/recycling strategy. Up to now, 20 PWR 900 MW reactors are managed in MOX hybrid management. The feedback experience of 18 years of PWR operation with MOX is satisfactory, without any specific problem regarding manoeuvrability or plant availability. EDF is now looking to introduce MOX fuels with a higher plutonium content (up to 8.6%) equivalent to natural uranium enriched to 3.7%. It is the goal of the MOX Parity core management which achieve balance of MOX and UOX fuel performance with a significant increase of the MOX average discharge burn-up (BU max: 52 GWd/t for MOX and UOX). The industrial maturity of new FAs designs, with increased performances, allows the implementation in the next years of new fuel managements on each NPPs series of the EDF fleet. The scheduling of the implementation of the new fuel managements on the PWRs fleet is a great challenge for EDF, with important stakes: the nuclear KWh cost decrease with the improvement of the plant operation performance. (author)

  1. nth roots with Hilbert-Schmidt defect operator of normal contractions

    Duggal, B.P.

    1992-08-01

    Let T be a normal contraction (on a complex separable Hilbert space H into itself) with an nth root A such that the defect operator D A =(1-A*A) 1/2 is of the Hilbert-Schmidt class C 2 . Then either A is normal or A is similar to a normal contraction. In the case in which T is hyponormal, A n =T and D A is an element of C 2 , A is a ''coupling'' of a contraction similar to a normal contraction and a contraction which is the quasi-affine transform of a unilateral shift. These results are applied to prove a (Putnam-Fuglede type) commutatively theorem for operator valued roots of commutative analytic functions and hyponormal contractions T which have an nth root with Hilbert-Schmidt defect operator. 23 refs

  2. The qualification of a new fuel - the operator's perspective

    Koonen, E.

    2001-01-01

    Operators of a research reactor generally have as their primary mission to provide the users with a safe, reliable and economic source of neutrons. They have to assure the availability of that source, while respecting the requirements of the license. The fuel management is one of the major aspects they have to tackle in order to fulfill their mission. This sometimes includes the qualification of a new fuel and the core conversion. The operator has to assure that the whole process is conducted in such manner that the availability of the neutron source is only minimally disturbed, that the costs are kept under control and the characteristics of the neutron source are preserved. This paper gives an overview of the various issues that the operator has to consider. (author)

  3. Commentary on spent fuel storage at Morris operation

    Eger, K.J.; Zima, G.E.

    1979-10-01

    The General Electric Company is providing technical support to Battelle Pacific Northwest Laboratories in the analysis of the design, operation, and maintenance experience in the handling of nuclear fuel at the Independent Spent Fuel Storage Facility. The purpose of this report is to provide a description of spent fuel handling activities and systems, and an analysis of the storage performance as developed over the seven year operational history of the Morris Operation. Design considerations and performance are analyzed for both the basin and key supporting systems. The bases for this analysis are the provisions for containing radioactive by-product materials, for shielding from the radiation they emit, and for preventing the formation of a critical array. These provisions have been met effectively over the history of storage at Morris. The release of radioactive materials is minimized by the protection of the cladding integrity, the containment of the basin water, the removal of radioactive and other contaminants from the water, and by filtering and then dispersing the basin air. Four auxiliary systems are provided to accomplish this, the basin leak detection system, the filter, the coolers, and the building ventilation system. This successful history notwithstanding, action to reduce personnel exposure, to improve fuel handling reliability and to lessen the potential for accidents continues to be taken

  4. Application of the integral method to modelling the oxidation of defected fuel elements

    Kolar, M.

    1995-06-01

    The starting point for this report is the discrepancy reported in previous work between the reaction-diffusion calculations and the CEX-1 experiment, which involves storage of defected fuel elements in air at 150 deg C. This discrepancy is considerably diminished here by a more critical choice of theoretical parameters, and by taking into account the fact that different CEX-1 fuel elements were oxidized at very different rates and that the fuel element used previously for comparison with theoretical calculations actually underwent two limited-oxygen-supply cycles. Much better agreement is obtained here between the theory and the third, unlimited-air, storage period of the CEX-1 experiment. The approximate integral method is used extensively for the solution of the one-dimensional diffusion moving-boundary problems that may describe various storage periods of the CEX-1 experiment. In some cases it is easy to extend this method to arbitrary precision by using higher moments of the diffusion equation. Using this method, the validity of quasi-steady-state approximation is verified. Diffusion-controlled oxidation is also studied. In this case, for the unlimited oxygen supply, the integral method leads to an exact analytical solution for linear geometry, and to a good analytical approximation of the solution for the spherically symmetric geometry. These solutions may have some application in the analysis of experiments on the oxidation of small UO 2 fragments or powders when the individual UO 2 grains may be considered to be approximately spherical. (author). 23 refs., 5 tabs., 11 figs

  5. Spent fuel cask handling at an operating nuclear power plant

    Pal, A.C.

    1988-01-01

    The importance of spent fuel handling at operating nuclear power plants cannot be overstated. Because of its highly radioactive nature, however, spent fuel must be handled in thick, lead-lined containers or casks. Thus, all casks for spent fuel handling are heavy loads by the US Nuclear Regulatory Commission's definition, and any load-drop must be evaluated for its potential to damage safety-related equipment. Nuclear Regulatory Guide NUREG-0612 prescribes the regulatory requirements of alternative heavy-load-handling methodologies such as (a) by providing cranes that meet the requirements of NUREG-0554, which shall be called the soft path, or (b) by providing protective devices at all postulated load-drop areas to prevent any damage to safety-related equipment, which shall be called the hard path. The work reported in this paper relates to cask handling at New York Power Authority's James A. FitzPatrick (JAF) plant

  6. Idling operation apparatus for multicylinder fuel injection engine

    Kanahira, A

    1974-11-20

    A device to cut off the fuel supply to a number of cylinders at idling is described for those engines equipped with multicylinder fuel injection systems. The discontinuation of the fuel gas supply to the cylinders is made by a magnetically operated valve which is related to the accelerator. When the engine is idling, a switch activates the magnetic valve and the tube leading to the cylinder closes while a valve on the tube leading to a dual tank opens, and the pumped gas returns to the tank. This valve is installed on several cylinders, but not on all. Thus, at idling only a certain number of cylinders are firing, which lowers the hydrocarbon levels in the exhaust gas since non-firing cylinders intake and discharge only air.

  7. Hydrazine - hydrate water regime and operation of fuel elements

    Pashevitch, V.I.; Pashevitch, D.V.; Bogancs, J.; Tilky, P.

    1997-01-01

    Water chemistries currently used in WWER reactors are potassium based water chemistry (KOH) to adjust the pH with ammonia or hydrazine as oxygen scavenger. Based on the measurements of Zr 95 which is a corrosion product of the zirconium cladding, it is shown in this paper that the amount of corrosion products accompanying the reactor shutdown is smaller when hydrazine is used. This is particularly obvious on PAKS 1 and 2 when Zr 95 measurements are performed before and after switching the water chemistry from ammonia to hydrazine. It is concluded that the main advantage of using the hydrazine water chemistry is to decrease the thickness of the corrosion product layer formed on the fuel cladding, therefore the fuel temperature can be kept low. It is estimated that the fuel temperature increase due to the layer of corrosion products is 120 deg. C for KOLA 3 which is operated with ammonia water chemistry. (author). 5 figs

  8. Operational experience of the fuel cleaning facility of Joyo

    Mukaibo, R.; Matsuno, Y.; Sato, I.; Yoneda, Y.; Ito, H.

    1978-01-01

    Spent fuel assemblies in 'Joyo', after they are taken out of the core, are taken to the Fuel Cleaning Facility in the reactor service building and sodium removal is done. The cleaning process is done by cooling the assembly with argon gas, steam charging and rinsing by demineralized water. Deposited sodium was 50 ∼ 60 g per assembly. The sodium and steam reaction takes about 15 minutes to end and the total time the fuel is placed in the pot is about an hour. The total number of assemblies cleaned in the facility was 95 as of November 1977. In this report the operational experience together with discussions of future improvements are given. (author)

  9. Operational experience of the fuel cleaning facility of Joyo

    Mukaibo, R; Matsuno, Y; Sato, I; Yoneda, Y; Ito, H [O-arai Engineering Centre, PNC, Ibaraki-ken, Tokio (Japan)

    1978-08-01

    Spent fuel assemblies in 'Joyo', after they are taken out of the core, are taken to the Fuel Cleaning Facility in the reactor service building and sodium removal is done. The cleaning process is done by cooling the assembly with argon gas, steam charging and rinsing by demineralized water. Deposited sodium was 50 {approx} 60 g per assembly. The sodium and steam reaction takes about 15 minutes to end and the total time the fuel is placed in the pot is about an hour. The total number of assemblies cleaned in the facility was 95 as of November 1977. In this report the operational experience together with discussions of future improvements are given. (author)

  10. Hydrazine - hydrate water regime and operation of fuel elements

    Pashevitch, V I; Pashevitch, D V [Pand Co. (Russian Federation); Bogancs, J; Tilky, P [Paks NPP (Hungary)

    1997-02-01

    Water chemistries currently used in WWER reactors are potassium based water chemistry (KOH) to adjust the pH with ammonia or hydrazine as oxygen scavenger. Based on the measurements of Zr 95 which is a corrosion product of the zirconium cladding, it is shown in this paper that the amount of corrosion products accompanying the reactor shutdown is smaller when hydrazine is used. This is particularly obvious on PAKS 1 and 2 when Zr 95 measurements are performed before and after switching the water chemistry from ammonia to hydrazine. It is concluded that the main advantage of using the hydrazine water chemistry is to decrease the thickness of the corrosion product layer formed on the fuel cladding, therefore the fuel temperature can be kept low. It is estimated that the fuel temperature increase due to the layer of corrosion products is 120 deg. C for KOLA 3 which is operated with ammonia water chemistry. (author). 5 figs.

  11. Characteristics of Subfreezing Operation of Polymer Electrolyte Membrane Fuel Cells

    Mishler, Jeffrey Harris

    Polymer Electrolyte Membrane (PEM) Fuel Cells are capable of high efficiency operation, and are free of NOx, SOx, and CO2 emissions when using hydrogen fuel, and ideally suited for use in transportation applications due to their high power density and low operating temperatures. However, under subfreezing conditions which may be encountered during winter seasons in some areas, product water will freeze within the membrane, cathode side catalyst layer and gas diffusion media, leading to voltage loss and operation failure. Experiments were undertaken in order to characterize the amount and location of water during fuel cell operation. First, in-situ neutron radiography was undertaken on the fuel cells at a normal operating temperature for various operating current densities, inlet relative humidities, and diffusion media hydrophobicities. It was found that more hydrophobic cathode microporous layer (MPL) or hydrophilic anode MPL may result in a larger amount of water transporting back to the anode. The water profiles along the channels were measured and the point of liquid water emergence, where two phase flow begins, was compared to previous models. Secondly, under subfreezing temperatures, neutron imaging showed that water ice product accumulates because of lack of a water removal mechanism. Water was observed under both the lands and channels, and increased almost linearly with time. It is found that most ice exists in the cathode side. With evidence from experimental observation, a cold start model was developed and explained, following existing approaches in the literature. Three stages of cold start are explained: membrane saturation, ice storage in catalyst layer pores, and then ice melting. The voltage losses due to temperature change, increased transport resistance, and reduced electrochemical surface area. The ionic conductivity of the membrane at subfreezing temperatures was modeled. Voltage evolution over time for isothermal cold starts was predicted and

  12. Assess How Changes in Fuel Cycle Operation Impact Safeguards

    Tobin, Stephen Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Division; Adigun, Babatunde John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Division; Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Division; Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Division; Sprinkle, James K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Division

    2016-10-31

    Since the beginning of commercial nuclear power generation in the 1960s, the ability of researchers to understand and control the isotopic content of spent fuel has improved. It is therefore not surprising that both fuel assembly design and fuel assembly irradiation optimization have improved over the past 50+ years. It is anticipated that the burnup and isotopics of the spent fuel should exhibit less variation over the decades as reactor operators irradiate each assembly to the optimum amount. In contrast, older spent fuel is anticipated to vary more in burnup and resulting isotopics for a given initial enrichment. Modern fuel therefore should be more uniform in composition, and thus, measured safeguards results should be easier to interpret than results from older spent fuel. With spent fuel ponds filling up, interim and long-­term storage of spent fuel will need to be addressed. Additionally after long periods of storage, spent fuel is no longer self-­protecting and, as such, the IAEA will categorize it as more attractive; in approximately 20 years many of the assemblies from early commercial cores will no longer be considered self-­protecting. This study will assess how more recent changes in the reactor operation could impact the interpretation of safeguards measurements. The status quo for spent fuel assay in the safeguards context is that the overwhelming majority of spent fuel assemblies are not measured in a quantitative way except for those assemblies about to be loaded into a difficult or impossible to access location (dry storage or, in the future, a repository). In other words, when the assembly is still accessible to a state actor, or an insider, when it is cooling in a pool, the inspectorate does not have a measurement database that could assist them in re-­verifying the integrity of that assembly. The spent fuel safeguards regime would be strengthened if spent fuel assemblies were measured from discharge to loading into a difficult or impossible

  13. Operation experiences of JOYO fuel failure detection system

    Tamura, Seiji; Hikichi, Takayoshi; Rindo, Hiroshi.

    1982-01-01

    Monitoring of fuel failure in the experimental fast reactor JOYO is provided by two different methods, which are cover gas monitoring (FFDCGM) by means of a precipitator, and delayed neutron monitoring (FFDDNM) by means of neutron detectors. The interpretation of signals which were obtained during the reactor operation for performance testings, was performed. The countrate of the CGM is approximately 120 cps at 75MW operation, whose sources are due to Ne 23 , Ar 41 , and Na 24 . And the countrate of the DNM is approximately 2300 cps at 75MW operation which is mainly due to leakage neutron from the core. With those background of the systems, alarm level for monitoring was set at several times of each background level. The reactor has been operated for 5 years, the burn-up of the fuel is 40,000 MWD/T at the most. No trace of any fuel failure has been observed. The fact is also proven by the results of cover gas and sodium sampling analysis. In order to evaluate sensitivity of the FFD systems, a preliminary simulation study has been performed. According to the results, a signal level against one pin failure of 0.5 mm 2 hole may exceed the alarm level of the FFDCGM system. (author)

  14. Strategies for Lowering Solid Oxide Fuel Cells Operating Temperature

    Albert Tarancón

    2009-11-01

    Full Text Available Lowering the operating temperature of solid oxide fuel cells (SOFCs to the intermediate range (500–700 ºC has become one of the main SOFC research goals. High operating temperatures put numerous requirements on materials selection and on secondary units, limiting the commercial development of SOFCs. The present review first focuses on the main effects of reducing the operating temperature in terms of materials stability, thermo-mechanical mismatch, thermal management and efficiency. After a brief survey of the state-of-the-art materials for SOFCs, attention is focused on emerging oxide-ionic conductors with high conductivity in the intermediate range of temperatures with an introductory section on materials technology for reducing the electrolyte thickness. Finally, recent advances in cathode materials based on layered mixed ionic-electronic conductors are highlighted because the decreasing temperature converts the cathode into the major source of electrical losses for the whole SOFC system. It is concluded that the introduction of alternative materials that would enable solid oxide fuel cells to operate in the intermediate range of temperatures would have a major impact on the commercialization of fuel cell technology.

  15. Test results for fuel cell operation on anaerobic digester gas

    Spiegel, R. J.; Preston, J. L.

    EPA, in conjunction with ONSI, embarked on a project to define, design, test, and assess a fuel cell energy recovery system for application at anaerobic digester waste water (sewage) treatment plants. Anaerobic digester gas (ADG) is produced at these plants during the process of treating sewage anaerobically to reduce solids. ADG is primarily comprised of methane (57-66%), carbon dioxide (33-39%), nitrogen (1-10%), and a small amount of oxygen (sulfur-bearing compounds (principally hydrogen sulfide) and halogen compounds (chlorides). The project has addressed two major issues: development of a cleanup system to remove fuel cell contaminants from the gas and testing/assessing of a modified ONSI PC25 C fuel cell power plant operating on the cleaned, but dilute, ADG. Results to date demonstrate that the ADG fuel cell power plant can, depending on the energy content of the gas, produce electrical output levels close to full power (200 kW) with measured air emissions comparable to those obtained by a natural gas fuel cell. The cleanup system results show that the hydrogen sulfide levels are reduced to below 10 ppbv and halides to approximately 30 ppbv.

  16. Operational method for demonstrating fuel loading integrity in a reactor having accessible 235U fuel

    Ward, D.R.

    1979-07-01

    The Health Physics Research Reactor is a small pulse reactor at the Oak Ridge National Laboratory. It is desirable for the operator to be able to demonstrate on a routine basis that all the fuel pieces are present in the reactor core. Accordingly, a technique has been devised wherein the control rod readings are recorded with the reactor at delayed critical and corrections are made to compensate for the effects of variations in reactor height above the floor, reactor power, core temperature, and the presence of any massive neutron reflectors. The operator then compares these readings with the values expected based on previous operating experience. If this routine operational check suggests that the core fuel loading might be deficient, a more rigorous follow-up may be made

  17. Efficiency improvement of nuclear power plant operation: the significant role of advanced nuclear fuel technologies

    Van Velde, AA. de; Burtak, F.

    2000-01-01

    In this paper authors deals with nuclear fuel cycle and their economic aspects. At Siemens, the developments focusing on the reduction of fuel cycle costs are currently directed on .further batch average burnup increase, .improvement of fuel reliability, .enlargement of fuel operation margins, .improvement of methods for fuel design and core analysis. These items will be presented in detail in the full paper and illustrated by the global operating experience of Siemens fuel for both PWRs and BWRs. (authors)

  18. Safety related considerations for operation with defected elements in EBR-II

    Fryer, R.M.; Sackett, J.I.; Lambert, J.D.B.

    1976-01-01

    Traditionally, EBR-II has employed the 'shutdown and remove' philosophy when breached fuel elements are encountered. This mode of operation maintained in-plant inventories of fission products at low levels and allowed certain fission product detection systems to be employed as automatic plant shutdown devices. Information from fuel failure propagation studies and fast reactor operation indicates that shutdown under these conditions is unwarranted. Analytical studies, as well as fast reactor experience, further indicate that failure propagation, if it occurs at all, will not cross adjacent subassembly boundaries. Therefore, the 'shutdown and remove' philosophy can be liberalized to allow the demonstration of safety during a run-beyond-clad-breach mode of operation. This mode of operation is essential to the demonstration of the economics of commercial LMFBR systems

  19. Effect of wood fuels on power plant operability

    Orjala, M.; Ingalsuo, R.

    2001-01-01

    The objective of the research is to determine the critical properties of wood fuels on the basis of power plant operability, to determine the optimal conditions for reduction of harmful detriments, and to study how the storage and processing of wood fuels effect on the operability. Both the CFB and BFB technologies are studied. The project started in December 2000 and it will be ended by the end of 2002. Experts of the Fuels and Combustion research field of VTT Energy carry out the main parts of the research. Experts of the research field of Mineral Processing of VTT Chemical Technology, located in Outokumpu, and Kemian tutkimuspalvelut Oy/Oulu University, located in Outokumpu, participate in the analytics, and the research field of Materials and Manufacturing Technology of VTT Manufacturing Technology in Otaniemi participates in the research on material effects. System Technology Laboratory of Oulu University carries out the power plant automation and boiler control technology research under supervision of Professor Urpo Kortela. Co-operation with the materials research unit of EU's JRC, located in Petten, which started in the research 'Combustion of Forest Chips', will be continues in this research. Co-operation will be made with Swedish Vaermeforsk in the field of information exchange on experiences in utilisation of wood fuels in Swedish power plants and possibilities to join in the projects of Vaermeforsk in this research field. Following companies participate in the project: Etelae-Savon Energia Oy, Foster Wheeler Energia Oy, Kvaerner Pulping Oy, Simpele pasteboard factory of M-Real Oyj and Vaermeforsk AB (Sweden). (orig.)

  20. Characteriztion of particulate plutonium released in fuel cycle operations

    Seefeldt, W.B.; Mecham, W.J.; Steindler, M.J.

    1976-05-01

    An estimate of the plutonium source terms is made for the fuel cycles of three reactor types on the basis of currently applied, currently available, and estimated future technology. The three reactors are LWR-U, LWR-Pu, and LMFBR. The source terms are characterized as to quantity, form, and particle size distribution. Historical operating data for existing plants and the state of the art of the technology of air cleaning are reviewed

  1. Polymer electrolyte fuel cells: flow field for efficient air operation

    Buechi, F N; Tsukada, A; Haas, O; Scherer, G G [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1997-06-01

    A new flow field was designed for a polymer electrolyte fuel cell stack with an active area of 200 cm{sup 2} for operation at low air stoichiometry and low air over pressure. Optimum of gas flow and channel dimensions were calculated based on the required pressure drop in the fluid. Single cells and a bi-cell stack with the new flow field show an improved current/voltage characteristic when operated at low air stoichiometries as compared to that of the previous non optimized design. (author) 4 figs., 3 refs.

  2. Advanced operator interface design for CANDU-3 fuel handling system

    Arapakota, D.

    1995-01-01

    The Operator Interface for the CANDU 3 Fuel Handling (F/H) System incorporates several improvements over the existing designs. A functionally independent sit-down CRT (cathode-ray tube) based Control Console is provided for the Fuel Handling Operator in the Main Control Room. The Display System makes use of current technology and provides a user friendly operator interface. Regular and emergency control operations can be carried out from this control console. A stand-up control panel is provided as a back-up with limited functionality adequate to put the F/H System in a safe state in case of an unlikely non-availability of the Plant Display System or the F/H Control System'. The system design philosophy, hardware configuration and the advanced display system features are described in this paper The F/H Operator Interface System developed for CANDU 3 can be adapted to CANDU 9 as well as to the existing stations. (author)

  3. Advanced operator interface design for CANDU-3 fuel handling system

    Arapakota, D [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    The Operator Interface for the CANDU 3 Fuel Handling (F/H) System incorporates several improvements over the existing designs. A functionally independent sit-down CRT (cathode-ray tube) based Control Console is provided for the Fuel Handling Operator in the Main Control Room. The Display System makes use of current technology and provides a user friendly operator interface. Regular and emergency control operations can be carried out from this control console. A stand-up control panel is provided as a back-up with limited functionality adequate to put the F/H System in a safe state in case of an unlikely non-availability of the Plant Display System or the F/H Control System`. The system design philosophy, hardware configuration and the advanced display system features are described in this paper The F/H Operator Interface System developed for CANDU 3 can be adapted to CANDU 9 as well as to the existing stations. (author).

  4. Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor

    Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, CA 94550 (United States); Peixoto, Orpet J.M. [Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials, Avenida Rio Branco, 123/Grupo 515- Centro, CEP: 20040-005, Rio de Janeiro (Brazil); Diaz, Gustavo [National Regulatory Authority - Argentina, Av. Del Libertador 8250, (1429) Buenos Aires (Argentina)

    2015-07-01

    At the Atucha-I pressurized heavy water reactor in Argentina, fuel assemblies in the spent fuel pools are stored by suspending them in two vertically stacked layers. This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Since much of the fuel is very old, Cerenkov viewing devices are often not very useful even for the top layer. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 w% {sup 235}U, and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A spent fuel neutron counting tool consisting of a fission chamber, SFNC, has been used at the site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups to levels up 11,000 MWd/t, the existing signal processing software of the tool was found to fail due to non-linearity of the source term with burnup. A new Graphical User Interface software package based on the LabVIEW platform was developed to predict expected neutron signals covering all ranges of burnups and cooling times and establish maps of expected signals at various pool locations. The algorithm employed in the software uses a set of transfer functions in a 47-energy group structure which are coupled with a 47-energy group neutron source spectrum based on various cooling times and burnups for each of the two enrichment levels. The database of the software consists of these transfer functions for the three different inter-assembly pitches that the fuel is stored in at the site. The transfer functions were developed for a 6 by 6 matrix of fuel assemblies with the detector placed at the center surrounded by four near neighbors, eight next nearest neighbors and so on for the 36 assemblies. These calculations were performed using Monte Carlo radiation transport methods. The basic methodology consisted of starting sources in each of the assemblies and tallying the

  5. Spent Nuclear Fuel Project Cold Vacuum Drying Facility Operations Manual

    IRWIN, J.J.

    1999-01-01

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-553, Spent Nuclear Fuel Project Final Safety Analysis Report Annex B--Cold Vacuum Drying Facility. The HNF-SD-SNF-DRD-002, 1999, (Cold Vacuum Drying Facility Design Requirements), Rev. 4. and the CVDF Final Design Report. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence and references to the CVDF System Design Descriptions (SDDs). This manual has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  6. Calculation of the heat flow peak in case of local defect of the fuel plate of a nuclear reactor

    Fabrega, Serge

    1965-11-01

    The author reports the calculation of the local thermal flow which exits a fuel plate in a nuclear reactor, where a fabrication defect creates a much localized peak of the power density released in the plate. He first reports the development of the problem equations: hypotheses and data, equation elaboration, simplification and resolution. He presents the results of a numeric application to actual cases, and describes how the conduction in the sheath is taken into account (study of the influence of peak width and shape), and gives a synthetic presentation of the formula for the approximate calculation of the heat flow in case of local defect [fr

  7. Evaluation of design and operation of fuel handling systems for 25 MW biomass fueled CFB power plants

    Precht, D.

    1991-01-01

    Two circulating fluidized bed, biomass fueled, 25MW power plants were placed into operation by Thermo Electron Energy Systems in California during late 1989. This paper discusses the initial fuel and system considerations, system design, actual operating fuel characterisitics, system operation during the first year and modifications. Biomass fuels handled by the system include urban/manufacturing wood wastes and agricultural wastes in the form of orchard prunings, vineyard prunings, pits, shells, rice hulls and straws. Equipment utilized in the fuel handling system are described and costs are evaluated. Lessons learned from the design and operational experience are offered for consideration on future biomass fueled installations where definition of fuel quality and type is subject to change

  8. Spent Fuel Test - Climax data acquisition system operations manual

    Nyholm, R.A.

    1983-01-01

    The Spent Fuel Test-Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granite rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the US Department of Energy Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. The multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system (DAS) collects and processes data from more than 900 analog instruments. This report documents the software element of the LLNL developed SFT-C Data Acquisition System. It defines the operating system and hardware interface configurations, the special applications software and data structures, and support software

  9. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs

  10. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs.

  11. Solid oxide fuel cell performance under severe operating conditions

    Koch, Søren; Hendriksen, P.V.; Mogensen, Mogens Bjerg

    2006-01-01

    The performance and degradation of Solid Oxide Fuel Cells (SOFC) were studied under severe operating conditions. The cells studied were manufactured in a small series by ECN, in the framework of the EU funded CORE-SOFC project. The cells were of the anode-supported type with a double layer LSM...... cathode. They were operated at 750 °C or 850 °C in hydrogen with 5% or 50% water at current densities ranging from 0.25 A cm–2 to 1 A cm–2 for periods of 300 hours or more. The area specific cell resistance, corrected for fuel utilisation, ranged between 0.20 Ω cm2 and 0.34 Ω cm2 at 850 °C and 520 m......V, and between 0.51 Ω cm2 and 0.92 Ω cm2 at 750 °C and 520 mV. The degradation of cell performance was found to be low (ranging from 0 to 8%/1,000 hours) at regular operating conditions. Voltage degradation rates of 20 to 40%/1,000 hours were observed under severe operating conditions, depending on the test...

  12. [Recostruction of Extensive Acetabular Defects by Bioactive Glass Ceramics in Re-operations of Total Endoprostheses.].

    Urban, K; Sponer, P

    1998-01-01

    The authors made 37 revisions on account of aseptic loosening of total endoprostheses of the hip joint using bioactive glass ceramics BAS-0 of Lasak Co. Prague. For reconstruction of large defects of the acetabulum they used a combination of different types of anti-protrusion metal baskets and granules from this material. In some instances the glass ceramic material was combined with autologous spongiosa. The longest follow-up period is over 4 years. In no instance loosening of the glass ceramic material occurred or its expulsion. All reconstructed sockets of hip joints were burdened by the patients from the third month after surgery. Harris Hip Score before operation was on average 52. During the last checkups of the patients it reached the level of 86. The authors mention complications associated with the procedure. The advantages and disadvantages of the procedure are discussed. Key words: bioactive glass ceramics, reconstruction of acetabular defect, aseptic loosening of endoprosthesis.

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

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

    2002-01-01

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

  14. Sludge behavior in centrifugal contactor operation for nuclear fuel reprocessing

    Sakamoto, Atsushi; Sano, Yuichi; Takeuchi, Masayuki; Okamura, Nobuo; Koizumi, Kenji

    2015-01-01

    The Japan Atomic Energy Agency (JAEA) has been developing the centrifugal contactor for spent fuel reprocessing. In this study, we investigated the sludge behavior in centrifugal contactors at three different scales. The operational conditions (the flow rate and rotor speed) were varied. Most insoluble particles such as sludge remained in the rotor via centrifugal force. The capture ratio of sludge in the contactor was measured as a function of particle size at various flow rates, rotor speeds, and contactor scales. The sludge adhered and accumulated inside the rotor as the operational time increased, and the operational conditions influenced the capture ratio of the sludge; a lower flow rate and higher rotor speed increased the capture ratio. The results confirmed that Stokes' law can be applied to estimate the experimental result on the behavior of the capture ratio for centrifugal contactors with different scales. (author)

  15. Anodes for Solid Oxide Fuel Cells Operating at Low Temperatures

    Abdul Jabbar, Mohammed Hussain

    An important issue that has limited the potential of Solid Oxide Fuel Cells (SOFCs) for portable applications is its high operating temperatures (800-1000 ºC). Lowering the operating temperature of SOFCs to 400-600 ºC enable a wider material selection, reduced degradation and increased lifetime....... On the other hand, low-temperature operation poses serious challenges to the electrode performance. Effective catalysts, redox stable electrodes with improved microstructures are the prime requisite for the development of efficient SOFC anodes. The performance of Nb-doped SrT iO3 (STN) ceramic anodes...... at 400ºC. The potential of using WO3 ceramic as an alternative anode materials has been explored. The relatively high electrode polarization resistance obtained, 11 Ohm cm2 at 600 ºC, proved the inadequate catalytic activity of this system for hydrogen oxidation. At the end of this thesis...

  16. In reactor performance of defected zircaloy-clad U3Si fuel elements in pressurized and boiling water coolants

    Feraday, M.A.; Allison, G.M.; Ambler, J.F.R.; Chalder, G.H.; Lipsett, J.J.

    1968-05-01

    The results of two in-reactor defect tests of Zircaloy-clad U 3 Si are reported. In the first test, a previously irradiated element (∼5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at ∼270 o C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U 3 Si at 300 o C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  17. Dual fuel mode operation in diesel engines using renewable fuels: Rubber seed oil and coir-pith producer gas

    Ramadhas, A.S.; Jayaraj, S.; Muraleedharan, C. [Department of Mechanical Engineering, National Institute of Technology Calicut, Calicut-673601 (India)

    2008-09-15

    Partial combustion of biomass in the gasifier generates producer gas that can be used as supplementary or sole fuel for internal combustion engines. Dual fuel mode operation using coir-pith derived producer gas and rubber seed oil as pilot fuel was analyzed for various producer gas-air flow ratios and at different load conditions. The engine is experimentally optimized with respect to maximum pilot fuel savings in the dual fuel mode operation. The performance and emission characteristics of the dual fuel engine are compared with that of diesel engine at different load conditions. Specific energy consumption in the dual-fuel mode of operation with oil-coir-pith operation is found to be in the higher side at all load conditions. Exhaust emission was found to be higher in the case of dual fuel mode of operation as compared to neat diesel/oil operation. Engine performance characteristics are inferior in fully renewable fueled engine operation but it suitable for stationary engine application, particularly power generation. (author)

  18. Nuclear fuel in water reactors: manufacturing technology operational experience and development activities in the Federal Republic of Germany

    Holzer, R.; Knoedler, D.

    1977-01-01

    The nuclear fuel industry in the F.R. Germany comprises the full range of manufacturing capabilities for pressurized - boiling- and heavy water reactor technology. The existing manufacturing companies are RBU and Alkem. RBU makes natural and enriched UO 2 -fuel assemblies, starting with powder preparation. Facilites to produce UO 2 -Gadolinia and UO 2 -ThO 2 fuel are also available. Alkem is manufacturing mixed oxide UO 2 /PuO 2 -fuel and -rods. Zircaloy cladding tubes are produced by NRG and MRW. This constitutes the largest single nuclear fuel manufacturing capacity outside the USA. The companies are interested in export and current capacity trends indicate some overcapacity caused by delays in plant schedules. Construction of a new fuel manufacturing plant in the FRG has been announced by Exxon. Supplementary to quality control in manufacturing an integrated quality assurance-system has been established between the reactor vendor KWU, fuel design and -engineering division, and the existing manufacturing companies for fuel and tubing. The operating experience with LWR and HWR fuel dates back to 1964/65 and proves good performance. No generic problems like densification or rod bow were encountered. Possible reasons for the small fraction of defective rods could be quickly identified by a fast feedback system incorporating a close cooperation between KWU and the utilities. KWU combines fuel development, hot-cell and poolside service facilities as well as fuel technology linking to manufacturing in one hand. The common responsibility of KWU for core- and fuel design which enabled an integral optimization was also an important reason for the successful operation and flexibility in design. Development efforts will be concentrated on tests to improve the understanding of power ramping capability under extreme operational and postulated abnormal conditions, on statistical evaluation of safety aspects and on improved economy. The LWR fuel development was sponsored by the

  19. Influence of Cast Iron Structure on the Glassmold Equipment Operational Defects

    I. O. Leushin

    2015-01-01

    Full Text Available The growing demand for glass packaging contributes to the increase in production capacity of glass-container plants. Their equipment (cast iron glass-forming sets operates in continuous mode under complex cyclic thermal loads, which lead to the formation of operational defects on the working surfaces of details: graphite falling, cracks, oxidation, etc. Particular influence on the formation of these defects renders the microstructure of the material at the time of installation of details on the line.The article identifies the causes for formation of operational defects, formulates the ways to remedy them and prevent their occurrence.The authors studied details made from grey cast iron with flake and spherical forms of graphite. It is found that in the process of exploitation of the material is greatly reducing its hardness, strength, resistance to oxidation through of graphitization processes, chemical interaction of glass and iron, shock loads working edges. It is proved that the choice of initial microstructure of cast iron (the metal base, the graphite form, the presence of structural-free cementite exercises a determining influence on the durability of the mold tooling. The article proposes differential (layered arrangement of the graphite phase of cast iron in the alloy matrix (ferrite. This arrangement of high-carbon phase can simultaneously increase the thermal and oxidation resistance of the material. The formation of a layered structure of iron is produced by the intensification of the processes of alloying, modifying and directional freezing the melt.These data can be used to select the material of details by manufacturers glass-molds tooling.

  20. Open semiclassical strings and long defect operators in AdS/dCFT correspondence

    Susaki, Yoshiaki; Takayama, Yastoshi; Yoshida, Kentaroh

    2005-01-01

    We consider defect composite operators in a defect superconformal field theory obtained by inserting an AdS 4 xS 2 -brane in the AdS 5 xS 5 background. The one-loop dilatation operator for the scalar sector is represented by an integrable open spin chain. We give a description to construct coherent states for the open spin chain. Then, by evaluating the expectation value of the Hamiltonian with the coherent states in a long operator limit, a Landau-Lifshitz type of sigma model action is obtained. This action is also derived from the string action and hence we find a complete agreement in both super Yang-Mills and string sides. We see that an SO(3) H pulsating string solution is included in the action and its energy completely agrees with the result calculated in a different method. In addition, we argue that our procedure would be applicable to other AdS-brane cases

  1. Validity limits of fuel rod performance calculations from radiochemical data at operating LWRs

    Zaenker, H.; Nebel, D.

    1986-01-01

    There are various calculational models for the assessment of the fuel rod performance on the basis of the activities of gaseous and volatile fission products in the reactor coolant. The most important condition for the applicability of the calculational models is that a steady state release of the fission products into the reactor coolant takes place. It is well known that the models are not applicable during or shortly after reactor transients. The fact that 'unsteady states' caused by the fuel defection processes themselves can also occur in rare cases at steady reactor operation has not been taken into account so far. A test of validity is suggested with the aid of which the applicability of the calculational models can be checked in any concrete case, and the misleading of the reactor operators by gross misinterpretation of the radiochemical data can be avoided. The criteria of applicability are the fission product total activity, the slope tan α in the relationship lg (R/sub i//B/sub i/) proportional to lg lambda/sub i/ for the gaseous and volatile fission products, and the activity of the nonvolatile isotope 239 Np. (author)

  2. Operation and maintenance manuals for VEGA apparatus on radionuclide release from irradiated fuel

    Hayashida, Retsu; Hidaka, Akihide; Nakamura, Takehiko; Kudo, Tamotsu; Ohtomo, Takashi; Uetsuka, Hiroshi

    2001-03-01

    An experimental program, Verification Experiments of radionuclides Gas/Aerosol release (VEGA), was initiated at JAERI from September 1999 to improve source term predictabilities for hypothetical severe accidents. In the experiment, a short fuel segment taken from LWR fuels irradiated in Japanese power reactors is inductively heated to high temperatures (∼3273K) in a hot cell under high pressure conditions up to 1.0MPa. Particularly, a focus will be placed on the release and transport behaviors of low-volatile fission products (FP), actinides and short-life FP which have not been well investigated in previous studies. This experimental apparatus was completed in February 1999 and three experiments were performed by the end of 2000. Most of these experiments were successfully conducted, but some problems were also found. Especially, in the first VEGA-1 test with the purpose of shakedown and reference data acquisition, there were problems such as flow blockage at the outlet of furnace due to structure melting, malfunction of heaters and so on. Therefore, the design for these defective parts was changed for future experiments. Moreover, the apparatus is not so big but the entire processes are very complicated. Accordingly, the operators should well understand the details of the apparatus including the recent change of design. This report describes outlines of the VEGA apparatus and the procedures for operation and maintenance. (author)

  3. Fundamental Studies of Irradiation-Induced Defect Formation and Fission Product Dynamics in Oxide Fuels

    Stubbins, James

    2012-12-19

    The objective of this research program is to address major nuclear fuels performance issues for the design and use of oxide-type fuels in the current and advanced nuclear reactor applications. Fuel performance is a major issue for extending fuel burn-up which has the added advantage of reducing the used fuel waste stream. It will also be a significant issue with respect to developing advanced fuel cycle processes where it may be possible to incorporate minor actinides in various fuel forms so that they can be 'burned' rather than join the used fuel waste stream. The potential to fission or transmute minor actinides and certain long-lived fission product isotopes would transform the high level waste storage strategy by removing the need to consider fuel storage on the millennium time scale.

  4. Basis for Interim Operation for Fuel Supply Shutdown Facility

    BENECKE, M.W.

    2003-01-01

    This document establishes the Basis for Interim Operation (BIO) for the Fuel Supply Shutdown Facility (FSS) as managed by the 300 Area Deactivation Project (300 ADP) organization in accordance with the requirements of the Project Hanford Management Contract procedure (PHMC) HNF-PRO-700, ''Safety Analysis and Technical Safety Requirements''. A hazard classification (Benecke 2003a) has been prepared for the facility in accordance with DOE-STD-1027-92 resulting in the assignment of Hazard Category 3 for FSS Facility buildings that store N Reactor fuel materials (303-B, 3712, and 3716). All others are designated Industrial buildings. It is concluded that the risks associated with the current and planned operational mode of the FSS Facility (uranium storage, uranium repackaging and shipment, cleanup, and transition activities, etc.) are acceptable. The potential radiological dose and toxicological consequences for a range of credible uranium storage building have been analyzed using Hanford accepted methods. Risk Class designations are summarized for representative events in Table 1.6-1. Mitigation was not considered for any event except the random fire event that exceeds predicted consequences based on existing source and combustible loading because of an inadvertent increase in combustible loading. For that event, a housekeeping program to manage transient combustibles is credited to reduce the probability. An additional administrative control is established to protect assumptions regarding source term by limiting inventories of fuel and combustible materials. Another is established to maintain the criticality safety program. Additional defense-in-depth controls are established to perform fire protection system testing, inspection, and maintenance to ensure predicted availability of those systems, and to maintain the radiological control program. It is also concluded that because an accidental nuclear criticality is not credible based on the low uranium enrichment

  5. Operating experiences in fuel handling system at KGS

    Reddy, G.P.; Nagabhushanam

    2006-01-01

    Refuelling operations were started at KGS in August, 2000. Rich and varied experience was gained during this period through internal discussion/Quality circles/Procedural reviews and analysis of various incidents that have taken place in KGS and other units of NPCIL Some of the unique jobs carried out at KGS include-Development of tools for in-situ replacement of FM front end cover in FM service area (which was done for the first time in NPCIL history), Modification of FM magazine rear end plate mounting screws to avoid the possibility of magazine rotation stalling, The incident of Stalling of B-Ram during installation of upstream shield plug in KGS - 1 has brought out many weakness that were existing in the system in a dormant manner. Review of maintenance procedures was carried out and a special underwater operated sensor was developed and installed in Transfer Magazine to sense the presence and proper positioning of fuel bundles in the Transfer magazine tube during fuel loading operation. Numerous modifications were carried out in the system to increase equipment reliability, ease of operation and maintenance, to reduce man-rem consumption. Most notable among these modifications include -zig saw panel modification, EFCV O-ring modification, Ram BF switch modification, provision for increase in SFSB level provision, snout clamp oil circuit modification, ball valve actuator modification, installation of additional switch for sensing STS carriage UP position etc, This paper focuses on the challenges tackled in achieving near perfect performance, innovations and improvements carried out in the system to strive for this goal and development of procedures for reducing man-rem consumption and life extension of critical components. (author)

  6. Cost comparison of transcatheter and operative closures of ostium secundum atrial septal defects

    O’Byrne, Michael L.; Gillespie, Matthew J.; Shinohara, Russell T.; Dori, Yoav; Rome, Jonathan J.; Glatz, Andrew C.

    2015-01-01

    Background Clinical outcomes for transcatheter and operative closures of atrial septal defects (ASDs) are similar. Economic cost for each method has not been well described. Methods A single-center retrospective cohort study of children and adults cost of operative and transcatheter closures of ASD. A propensity score weight-adjusted multivariate regression model was used in an intention-to-treat analysis. Costs for reintervention and crossover admissions were included in primary analysis. Results A total of 244 subjects were included in the study (64% transcatheter and 36% operative), of which 2% (n = 5) were ≥18 years. Crossover rate from transcatheter to operative group was 3%. Risk of reintervention (P = .66) and 30-day mortality (P = .37) were not significantly different. In a multivariate model, adjusted cost of operative closure was 2012 US $60,992 versus 2012 US $55,841 for transcatheter closure (P cost favoring transcatheter closure were length of stay, medications, and follow-up radiologic and laboratory testing, overcoming higher costs of procedure and echocardiography. Professional costs did not differ. The rate of 30-day readmission was greater in the operative cohort, further increasing the cost advantage of transcatheter closure. Sensitivity analyses demonstrated that costs of follow-up visits influenced relative cost but that device closure remained favorable over a broad range of crossover and reintervention rates. Conclusion For single secundum ASD, cost comparison analysis favors transcatheter closure over the short term. The cost of follow-up regimens influences the cost advantage of transcatheter closure. PMID:25965721

  7. Experimental Assessment of a New Passive Neutron Multiplication Counter for Partial Defect Verification of LWR Fuel Assemblies

    LaFleur, A.; Menlove, H.; Park, S.-H.; Lee, S. K.; Oh, J.-M.; Kim, H.-D.

    2015-01-01

    The development of non-destructive assay (NDA) capabilities to improve partial defect verification of spent fuel assemblies is needed to improve the timely detection of the diversion of significant quantities of fissile material. This NDA capability is important to the implementation of integrated safeguards for spent fuel verification by the International Atomic Energy Agency (IAEA) and would improve deterrence of possible diversions by increasing the risk of early detection. A new NDA technique called Passive Neutron Multiplication Counter (PNMC) is currently being developed at Los Alamos National Laboratory (LANL) to improve safeguards measurements of LightWater Reactor (LWR) fuel assemblies. The PNMC uses the ratio of the fast-neutron emission rate to the thermalneutron emission rate to quantify the neutron multiplication of the item. The fast neutrons versus thermal neutrons are measured using fission chambers (FC) that have differential shielding to isolate fast and thermal energies. The fast-neutron emission rate is directly proportional to the neutron multiplication in the spent fuel assembly; whereas, the thermalneutron leakage is suppressed by the fissile material absorption in the assembly. These FCs are already implemented in the basic Self-Interrogation Neutron Resonance Densitometry (SINRD) detector package. Experimental measurements of fresh and spent PWR fuel assemblies were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using a hybrid PNMC and SINRD detector. The results from these measurements provides valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. (author)

  8. Birth defects in perinatal infants in areas contiguous to Hongyanhe Nuclear Power Plant before its normal operation

    Zhou Ling; Yin Zhihua; Han Zhonghui

    2011-01-01

    Objective: To understand the status of birth defects among the perinatal infants in the areas contiguous to Hongyanhe nuclear power plant before its normal operation, so as to provide background information for the evaluation of the impact of nuclear power plant on birth defects. Methods: From 1 October 1995 to 30 September 2009 the midwifery units at second class and above of Wafangdian City were asked to be in charge of recording the birth defects among the perinatal infants born during this period within the range of 50 km around the Hongyanhe nuclear power plant. Results: The total number of birth defects was 697, and the maternal number Was 83779. The average defect rate Was 83.20/10 4 . There were significant differences in the birth defect rate among different years (χ 2 =39.54, P<0.05), however, without linear trend therein,and among the survey areas (χ 2 =15.36, P<0.05) as well. The top five birth defects were congenital heart disease (148 cases), cleft lip with cleft palate (67 cases), congenital hydrocephalus (63 cases), and spina bifida (37 cases) and cleft lip (36 cases). Conclusions: The birth defect rate within the range of 50 km around the Hongyanhe nuclear power plant is lower than that of the region of Liaoning Province and the national rate of birth defects. (authors)

  9. Fuel operation of EDF nuclear fleet presentation of the centralized organization for operational engineering at the nuclear generation division

    Paulin, Ph.

    2006-01-01

    The main feature of EDF Nuclear Fleet is the standardization, with 'series' of homogeneous plants (same equipment, fuel and operation technical documents). For fuel operation, this standardization is related to the concept of 'fuel management scheme' (typical fuel reloads with fixed number and enrichment of fresh assemblies) for a whole series of plants. The context of the Nuclear Fleet lead to the choice of a centralized organization for fuel engineering at the Nuclear Generation Division (DPN), located at UNIPE (National Department for Fleet Operation Engineering) in Lyon. The main features of this organization are the following: - Centralization of the engineering activities for fuel operation support in the Fuel Branch of UNIPE, - Strong real-time link with the nuclear sites, - Relations with various EDF Departments in charge of design, nuclear fuel supply and electricity production optimization. The purposes of the organization are: - Standardization of operational engineering services and products, - Autonomy with independent methods and computing tools, - Reactivity with a technical assistance for sites (24 hours 'hot line'), - Identification of different levels (on site and off site) to solve core operation problems, - Collection, analysis and valorization of operation feedback, - Contribution to fuel competence global management inside EDF. This paper briefly describes the organization. The main figures of annual engineering production are provided. A selection of examples illustrates the contribution to the Nuclear Fleet performance. (authors)

  10. Operator decision aid for breached fuel operation in liquid metal cooled nuclear reactors

    Gross, K.C.; Hawkins, R.E.; Nickless, W.K.

    1991-01-01

    The purpose of this paper is to report the development of an expert system that provides continuous assessment of the safety significance and technical specification conformance of Delayed Neutron (DN) signals during breached fuel operation. The completed expert system has been parallelized on an innovative distributed-memory network-computing system that enables the computationally intensive kernel of the expert system to run in parallel on a group of low-cost Unix workstations. 1 ref

  11. Low stoichiometry operation of a proton exchange membrane fuel cell employing the interdigitated flow field

    Berning, Torsten; Kær, Søren Knudsen

    2012-01-01

    A multiphase fuel cell model based on computational fluid dynamics is used to investigate the possibility of operating a proton exchange membrane fuel cell at low stoichiometric flow ratios (ξ gases. A case study...

  12. Combustion and exhaust emission characteristics of a dual fuel compression ignition engine operated with pilot Diesel fuel and natural gas

    Papagiannakis, R.G.; Hountalas, D.T.

    2004-01-01

    Towards the effort of reducing pollutant emissions, especially soot and nitrogen oxides, from direct injection Diesel engines, engineers have proposed various solutions, one of which is the use of a gaseous fuel as a partial supplement for liquid Diesel fuel. These engines are known as dual fuel combustion engines, i.e. they use conventional Diesel fuel and a gaseous fuel as well. This technology is currently reintroduced, associated with efforts to overcome various difficulties of HCCI engines, using various fuels. The use of natural gas as an alternative fuel is a promising solution. The potential benefits of using natural gas in Diesel engines are both economical and environmental. The high autoignition temperature of natural gas is a serious advantage since the compression ratio of conventional Diesel engines can be maintained. The present contribution describes an experimental investigation conducted on a single cylinder DI Diesel engine, which has been properly modified to operate under dual fuel conditions. The primary amount of fuel is the gaseous one, which is ignited by a pilot Diesel liquid injection. Comparative results are given for various engine speeds and loads for conventional Diesel and dual fuel operation, revealing the effect of dual fuel combustion on engine performance and exhaust emissions

  13. Inhalation exposures during operations in spent fuel bays

    Dua, S.K.; Maniyan, C.G.; Kotrappa, P.

    1987-01-01

    Radioactive aerosols are generated during operations (transfer, cutting, storage and shipment of fuel) in spent fuel bays. A study has been carried out on the airborne concentration, size distribution and solubility of the aerosol, to evaluate the inhalation exposure of the workers. Personal air samplers were used for the measurement of concentration of airborne radioactivity and Andersen impactor for particle size distribution. Some of the collected sampels were followed for their solubility in lung serum simulant for a period of about 200 days. The observed concentrations of the aerosols and their activity median aerodynamic diameters (AMAD) were 418 ± 443 Bq m -3 (βγ), 1.28 ± 1.478 Bq m -3 (∞) and 6.79 ± 0.4μm respectively. Analysis of the samples revealed the presence of 239 Pu, U, 90 Sr and 137 Cs. The clearance half-life of the aerosols was the same (155 days) for all the isotopes (U, 239 Pu, 90 Sr; 137 Cs). Calculation of effective dose equivalent for the clearance half-life and for 6.79 μm AMAD aerosols was carried out using the method recommended in ICRP-30. Annual effective dose equivalent of the workers, if no respirators were worn, worked out to be 3.2 mSv (320 mrem), the contribution from alpha emitters being about 63 per cent of the total. (author). 5 refs., 2 tables

  14. Inhalation exposures during operations in spent fuel bays

    Dua, S.K.; Maniyan, C.G.; Kotrappa, P.

    1987-01-01

    Radioactive aerosols are generated during operations (transfer, cutting, storage and shipment of fuel) in spent fuel bays. A study has been carried out on the airborne concentration, size distribution and dissolution rate of aerosols to evaluate the inhalation exposure of the workers. Personal air samplers were used for the measurement of concentration of airborne radioactivity and an Andersen impactor for the particle size distribution. The dissolution rates of some of the collected samples in lung serum simulant were followed for a period of about 200 days. Analysis of the samples revealed the presence of 239 Pu, U, 90 Sr and 137 Cs. For all the radionuclides measured ( 239 Pu, U, 90 Sr, 137 Cs) 40% of the activity dissolved rapidly (half-time 1.2d), the remainder with a half-time of 155 d. Calculation of effective dose equivalent for this dissolution half-time and for 6.8μm AMAD aerosols was carried out using the method recommended in ICRP 30 and reported in literature. The annual effective dose equivalent of the workers, if no respirators are worn, worked out to be 8.9 mSv (890 mrem), the contribution from alpha emitters being about 91% of the total. (author)

  15. High-Level Functional and Operational Requirements for the Advanced Fuel Cycle Facility

    Charles Park

    2006-01-01

    This document describes the principal functional and operational requirements for the proposed Advanced Fuel Cycle Facility (AFCF). The AFCF is intended to be the world's foremost facility for nuclear fuel cycle research, technology development, and demonstration. The facility will also support the near-term mission to develop and demonstrate technology in support of fuel cycle needs identified by industry, and the long-term mission to retain and retain U.S. leadership in fuel cycle operations. The AFCF is essential to demonstrate a more proliferation-resistant fuel cycle and make long-term improvements in fuel cycle effectiveness, performance and economy

  16. Dual winch nuclear fuel transfer system providing more reliable fuel transfer during refueling operations

    Meuschke, R.E.; Harper, M.J.; Stefko, D.J.

    1991-01-01

    This paper describes a nuclear power plant having an auxiliary building, a containment building having the wall, a track extending through a transfer tube within the containment wall, and a fuel transfer system for moving fuel assemblies along the track between the auxiliary building side and the containment building side of the containment wall. It comprises: a car having wheels for movement along spaced rails of the track and further having a carrying basket for one or more fuel assemblies; winch means located on the auxiliary building side of the containment wall and above the water level existing over the track during refueling operations to drive the car along the track; first cable means and second cable means extending substantially vertically downward from the winch means to the tack level; first sheave means for directing the first and the second cable means substantially in the horizontal direction along the track; means for securing the first cable means to the car so that winch pulling force on the first cable means drives the car away from the containment building; second sheave means located near the containment end of the transfer tube

  17. Cathodes for Solid Oxide Fuel Cells Operating at Low Temperatures

    Samson, Alfred Junio

    . High performance cathodes were obtained from strontium-doped lanthanum cobaltite (LSC) infiltrated - Ce0.9Gd0.1O1.95 (CGO) ionic conducting backbone. Systematic tuning of the CGO and LSC firing temperatures and LSC loading resulted in a cathode with low polarization resistance, Rp = 0.044 cm2 at 600......This dissertation focuses on the development of nanostructured cathodes for solid oxide fuel cells (SOFCs) and their performance at low operating temperatures. Cathodes were mainly fabricated by the infiltration method, whereby electrocatalysts are introduced onto porous, ionic conducting backbones...... with increasing LSC firing temperature, highlighting the importance of materials compability over higher ionic conductivity. The potential of Ca3Co4O9+delta as an electrocatalyst for SOFCs has also been explored and encouraging results were found i.e., Rp = 0.64 cm2 for a Ca3Co4O9+delta/CGO 50 vol % composite...

  18. Operating method of molten carbonate type fuel cell

    Nakanishi, Tsuneo

    1988-12-06

    Molten carbonate type fuel cell involves a problem of oxidation of anode while the unit is stopped. Although there is a method proposed wherein an inactive gas is supplied to anode during the stoppage, the market-available inactive gas contains a slight amount of oxygen which makes it difficult to prevent the deterioration of the anode. In this invention, at the start and the stop other than the normal operation, a protective gas mixture of an inactive gas with a small amount of hydrogen is supplied to the anode. The inactive gas is a commercial type nitrogen, argon or helium; hydrogen is mixed in amount 0.5 - 2.0% of the inactive gas. By this method, oxygen in air which comes in from the gas-sealed portion of the cell is reduced by hydrogen in the protective gas and is discharged in the form of water. 2 figs.

  19. Extended fuel cycle operation for pressurized water reactor plants

    Silvestri, G.J. Jr.

    1978-01-01

    A nuclear steam turbine power plant system having an arrangement therein for extended fuel cycle operation is described. The power plant includes a turbine connected at its inlet to a source of motive fluid having a predetermined pressure associated therewith. The turbine has also connected thereto an extraction conduit which extracts steam from a predetermined location therein for use in an associated apparatus. A bypass conduit is provided between a point upstream of the inlet and the extraction conduit. A flow control device is provided within the bypass conduit and opens when the pressure of the motive steam supply drops beneath the predetermined pressure as a result of reactivity loss within the nuclear reactor. Opening of the bypass conduit provides flow to the associated apparatus and at the same time provides an increased flow orifice to maintain fluid flow rate at a predetermined level

  20. Fractional release of short-lived noble gases and iodine from HTGR fuel compact containing a fraction of coated fuel particles with through-coating defects

    Ogawa, Toru; Fukuda, Kosaku; Kobayashi, Fumiaki; Kikuchi, Teruo; Tobita, Tsutomu; Kashimura, Satoru; Kikuchi, Hironobu; Yamamoto, Katsumune.

    1986-10-01

    Fractional release (R/B) data of short-lived noble gases and iodine from sweep-gas irradiated HTGR fuel compacts were analyzed. Empirical formulas to predict R/B of 88 Kr as a function of temperature and fraction through-coating defects, and to calculate ratios of R/B's of other shortlived gases to that of 88 Kr were proposed. A method to predict R/B of iodine was also proposed. As for 131 I, a fission product of major safety concern, (R/B) I 131 ≅ (R/B) Xe 133 was predicted. Applying those methods, R/B from OGL-1 fuel element (5th and 6th) was predicted to show a good agreement with observation. (author)

  1. A study of point defects in UO 2+x and their impact upon fuel properties

    Ma , Yue

    2017-01-01

    Uranium dioxide is an oxygen excess, non-stoichiometric, fluorite material which exists over a wide range of compositions. At temperatures and oxygen activities that are relevant to its in-reactor behaviour, it has been reported that various types of point defects and clusters thereof may exist with different charge states and different compositions on both the anion and cation sublattices. These defects have a major influence on certain key engineering properties such as cation self -diffusi...

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

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

    2011-01-01

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

  3. Operation results of 3-rd generation nuclear fuel WWER-440 in initial period

    Adeev, V.; Panov, A.

    2011-01-01

    On unit 4 of Kola NPP trial operation of 3-rd generation's fuel began in 2010. Fuel assemblies of 3-rd generation (FA-3) have a number of design features that provide better operational characteristics. Concise description of a design and the basic advantages of fuel of 3-rd generation are described in articles. Increasing of efficiency of nuclear fuel usage will be achieved by reduction of the parasitic capture of thermal neutrons in constructional materials (weight of zirconium is reduced), optimization of uranium-water relation (increase in fuel elements step), increasing of uranium loading (usage of fuel pellets with increased diameter and without central hole in them). By results of trial operation mass transition to use of given type of assemblies in WWER-440 is possible. This report presents the basic outcomes of the trial operation, a brief survey of the obtained data. The basic characteristics of the reactor core with fuel of 3-rd generation are resulted in work. (authors)

  4. Leaching Studies on ACR-1000{sup R} Fuel Under Reactor Operating Conditions

    Sunder, S. [Atomic Energy of Canada Limited, Fuel and Fuel Channel Safety Branch, Chalk River, Ontario, K0J 1J0 (Canada)

    2009-06-15

    ACR-1000{sup R} is the latest nuclear power reactor being developed by AECL. The ACR-1000 fuel uses a modified CANFLEX{sup R} fuel bundle that contains low-enriched uranium and pellets of burnable neutron absorbers (BNA) in a central element. Dysprosium and gadolinium are used as the burnable neutron absorbers and are present as oxides in a 'fully-stabilized' zirconia matrix. The BNA material in the centre element is designed to limit the coolant void reactivity of the reactor core during postulated loss-of-coolant accidents. As part of the ACR-1000 fuel development, the stability of the BNA material under conditions associated with defects of the Zircaloy sheathing of the BNA central element has been investigated. The results of these tests can be used to demonstrate the phase stability and leaching behaviour of the ACR-1000 fuel under reactor operating conditions. The samples were disks, about 3-4 mm thick, obtained from BNA pellets and Candu fuel (natural uranium UO{sub 2}) pellets (the UO{sub 2} measurements provide a reference point). Leaching tests were carried out in light water at 325 deg. C, above the maximum coolant temperature in an ACR-1000 fuel channel during normal operating conditions (319 deg. C). This temperature also bounds the maximum operating temperature for the current Candu reactors (311 deg. C). The initial pH of the solution (measured at room temperature) used in the leaching tests was 10.3. The leach rates were determined by monitoring the amount of metals leached into solutions. Leaching tests were also carried out with BNA pellet samples in the presence of Zr-2.5%Nb pressure tube coupons to determine the effects, if any, of the presence of pressure tube material on leach rates. Other leaching tests with BNA pellet samples and UO{sub 2} pellets were conducted at 80 deg. C to study the effects of temperature on the leach rates. The temperature of 80 deg. C was selected as representative of typical shutdown temperatures

  5. BWR SFAT, gross-defect verification of spent BWR fuel. Final report on Task FIN A563 on the Finnish Support Programme to IAEA Safeguards including BWR SFAT User Manual

    Tarvainen, M.; Paakkunainen, M.; Tiitta, A.; Sarparanta, K.

    1994-04-01

    A measurement instrument called Spent Fuel Attribute Tester, SFAT, has been designed, fabricated and taken into use by the IAEA in gross defect verification of spent BWR fuel assemblies. The equipment consists of an underwater measurement head connected with cables to a control unit on the bridge of the fuel handling machine as well as to a PMCA for measurement of the gamma spectra. The BWR SFAT is optimized for the AFR interim storage, TVO KPA-STORE, of the TVO Power Company in Olkiluoto, Finland. It has a shape and it is moved like a fuel assembly using the fuel handling machine. No fuel movements are needed. Spent fuel specific radiation from the fission product 137 Cs at the gamma-ray energy of 662 keV is detected above the assemblies in the storage rack using a NaI(Tl) detector. In the design and in licensing the requirements of the IAEA, operator and the safety authority have been taken into account. The BWR SFAT allows modifications for other LWR fuel types with minor changes. The work has been carried out under the task FIN A 563 of the Finnish Support Programme to IAEA Safeguards. (orig.) (9 refs., 22 figs.)

  6. 77 FR 45921 - Alaskan Fuel Hauling as a Restricted Category Special Purpose Flight Operation

    2012-08-02

    ... operations, each aircraft used to transport fuel will be required to receive FAA certification for the... regulations. The special purpose of Alaskan fuel hauling was considered for aircraft type-certificated under... required for this special purpose. The fuel hauling system must be shown to meet the applicable...

  7. Fast reactor parameter optimization taking into account changes in fuel charge type during reactor operation time

    Afrin, B.A.; Rechnov, A.V.; Usynin, G.B.

    1987-01-01

    The formulation and solution of optimization problem for parameters determining the layout of the central part of sodium cooled power reactor taking into account possible changes in fuel charge type during reactor operation time are performed. The losses under change of fuel composition type for two reactor modifications providing for minimum doubling time for oxide and carbide fuels respectively, are estimated

  8. Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

    Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

    1988-12-01

    This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO 2 oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO 2 pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs

  9. Operation of Nuclear Fuel Based on Reprocessed Uranium for VVER-type Reactors in Competitive Nuclear Fuel Cycles

    Troyanov, V.; Molchanov, V.; Tuzov, A. [TVEL Corporation, 49 Kashirskoe shosse, Moscow 115409 (Russian Federation); Semchenkov, Yu.; Lizorkin, M. [RRC ' Kurchatov Institute' (Russian Federation); Vasilchenko, I.; Lushin, V. [OKB ' Gidropress' (Russian Federation)

    2009-06-15

    Current nuclear fuel cycle of Russian nuclear power involves reprocessed low-enriched uranium in nuclear fuel production for some NPP units with VVER-type LWR. This paper discusses design and performance characteristics of commercial nuclear fuel based on natural and reprocessed uranium. It presents the review of results of commercial operation of nuclear fuel based on reprocessed uranium on Russian NPPs-unit No.2 of Kola NPP and unit No.2 of Kalinin NPP. The results of calculation and experimental validation of safe fuel operation including necessary isotope composition conformed to regulation requirements and results of pilot fuel operation are also considered. Meeting the customer requirements the possibility of high burn-up achieving was demonstrated. In addition the paper compares the characteristics of nuclear fuel cycles with maximum length based on reprocessed and natural uranium considering relevant 5% enrichment limitation and necessity of {sup 236}U compensation. The expedience of uranium-235 enrichment increasing over 5% is discussed with the aim to implement longer fuel cycles. (authors)

  10. Operation experience of the advanced fuel assemblies at Unit 1 of Volgodonsk NPP within four fuel cycles

    Vasilchenko, I.; Kobelev, S.; Kushmanov, S.

    2006-01-01

    The first commissioning of Volgodonsk NPP Unit 1 with standard reactor WWER-1000 (project V-320) was in 2001. The reactor core, starting from the first fuel charge, was arranged completely with Advanced Fuel Assemblies (AFAs). In this way, it is possible to obtain the experience in startup and operation of the core, completely arranged with AFAs, and also to get a possibility of performing the comprehensive check for justification of newly commissioned units and justification of design solutions accepted in the design of reactor core for Taiwan NPP, Bushehr NPP and Kudankulam NPP. The first fuel charge of the Volgodonsk NPP Unit 1 is a reference and unified for Tiawan NPP (V-428), Bushehr NPP (V-446), Kudankulam NPP(V-412) with small differences caused by design features of RP V-320. The first core charge of Unit 1 of Volgodonsk NPP was arranged of 163 AFAs, comprising 61 CPS ARs and 42 BAR bundles. The subsequent fuel charges were arranged of AFAs with gadolinium oxide integrated into fuel instead of BAR. By 2005 the results of operation of the core at Unit 1 of Volgodonsk NPP during four fuel cycles showed that AFA is sufficiently reliable and serviceable. The activity of the primary coolant of the Volgodonsk NPP is at stable low level. During the whole time of the core operation of the Volgodonsk NPP Unit 1 no leaky AFAs were revealed. The modifications of the internals, made during pre-operational work, are reasonable and effective to provide for fuel mechanical stability in the course of operation. The modifications, made in AFA structure during operation of the Volgodonsk NPP Unit 1, are aimed at improving the service and operational reliability of its components. Correctness of the solutions taken is confirmed by AFAs operation experience both at the Volgodonsk NPP, and at other operating Russian NPPs

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

    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

  12. Initial operational experience with Gd-2M+ fuel at Dukovany NPP

    Borovička, M.; Zýbal, J.

    2015-01-01

    Trend of continuous development of nuclear fuel and fuel cycle can be observed from the very beginning of Dukovany NPP operation. The results of this development are documented on the one hand by extending the length of the cycle and on the other by significant reduction in the number of fresh FA’s which are loaded into reactor cores. As a continuation of this trend introduces Dukovany NPP evolutional change of nuclear fuel from the fuel Gd-2M to the Gd-2M + . (authors) Keywords: Gd-2M + , fuel assembly, operational experience

  13. On site PWR fuel inspection measurements for operational and design verification

    1996-01-01

    The on-site inspection of irradiated Pressurized Water Reactor (PWR) fuel and Non-Fuel Bearing Components (NFBC) is typically limited to visual inspections during refuelings using underwater TV cameras and is intended primarily to confirm whether the components will continue in operation. These inspections do not normally provide data for design verification nor information to benefit future fuel designs. Japanese PWR utilities and Nuclear Fuel Industries Ltd. designed, built, and performed demonstration tests of on-site inspection equipment that confirms operational readiness of PWR fuel and NFBC and also gathers data for design verification of these components. 4 figs, 3 tabs

  14. Operational report, Formation of the XXVII reactor core, plan of fuel exchange

    Martinc, R.

    1977-01-01

    Plan for fuel exchange for formation of the reactor core No. XXVII is presented. This report includes: the quantity of 80% enriched fuel which is input in the core, description of the fuel 'transfer' through the core within this fuelling scheme. It covers the review of reactor safety operating with the core No. XXVII related to reactivity change, thermal load of the fuel channels and fuel burnup. These data result from the analysis based on the same correlated calculation method which was applied for planning the first regular fuel exchange with 80% enriched fuel (core No. XXVI configuration), which has been approved in february 1977. Based on the enclosed data and the fuel exchange according to the proposed procedure it is expected that the reactor operation with core No. XXVII configuration will be safe [sr

  15. Design and operational behaviour of the SNR-reactor fuel element structure

    Dietz, W.; Toebbe, H.

    1985-01-01

    The fuel element and core concept of a fast breeder reactor is described by the example of the SNR 300 (1st core), and the requirements made on the fuel elements with respect to burnup and neutron dose are listed for existing and projected plants. Irradiation experiments carried out and operational experience gained with fuel elements show that the residence time of the fuel elements is influenced mainly by the stability of shape of the fuel element components. The requirements made with reference to neutron loading for future advanced high-performance fuel elements can not be anticipated from the present state of experience. Besides optimization of fuel element design and checking-out of the limits of operation by PFADFINDERELEMENTE elements, R and D work for the improvement of fuel element materials is also necessary. (orig.) [de

  16. Design considerations and operating experience with wet storage of Ontario Hydro's irradiated fuel

    Frost, C.R.; Naqvi, S.J.; McEachran, R.A.

    1987-01-01

    The characteristics of Ontario Hydro's fuel and at-reactor irradiated fuel storage water pools (or irradiated fuel bays) are described. There are two types of bay known respectively as primary bays and auxiliary bays, used for at-reactor irradiated fuel storage. Irradiated fuel is discharged remotely from Ontario Hydro's reactors to the primary bays for initial storage and cooling. The auxiliary bays are used to receive and store fuel after its initial cooling in the primary bay, and provide additional storage capacity as needed. The major considerations in irradiated fuel bay design, including site-specific requirements, reliability and quality assurance, are discussed. The monitoring of critical fuel bay components, such as bay liners, the development of high storage density fuel containers, and the use of several irradiated fuel bays at each reactor site have all contributed to the safe handling of the large quantities of irradiated fuel over a period of about 25 years. Routine operation of the irradiated fuel bays and some unusual bay operational events are described. For safety considerations, the irradiated fuel in storage must retain its integrity. Also, as fuel storage is an interim process, likely for 50 years or more, the irradiated fuel should be retrievable for downstream fuel management phases such as reprocessing or disposal. A long-term experimental program is being used to monitor the integrity of irradiated fuel in long-term wet storage. The well characterized fuel, some of which has been in wet storage since 1962 is periodically examined for possible deterioration. The evidence from this program indicates that there will be no significant change in irradiated fuel integrity (and retrievability) over a 50 year wet storage period

  17. Feasibility studies of computed tomography in partial defect detection of spent BWR fuel

    Levai, F.; Tikkinen, J.; Tarvainen, M.; Arlt, R.

    1990-10-01

    Feasibility studies were made for tomographic reconstruction of a cross-sectional activity distribution of a spent nuclear fuel assembly. The purpose was to determine the number of fuel rods (pins) and localize the positisons where pins are missing. The activity distribution map showing the locations of fuel rods in the assembly was reconstructed. The theoretical part of this work consists of simulation of image reconstruction based on theoretically calculated data from a reference assembly model. Evaluation of different image reconstruction techniques was made. Measurements were made in real facility conditions. Gamma radiation from an irradiated 8 x 8 - 1 BWR fuel assembly was measured through a narrow custom made collimator from different angles and positions. The measured data set was used as projections for reconstructing the activity profile of the assembly in cross-sectional plane

  18. Effect of water chemistry and fuel operation parameters on Zr + 1% Nb cladding corrosion

    Kritsky, V G; Petrik, N G; Berezina, I G; Doilnitsina, V V [VNIPIET, St. Petersburg (Russian Federation)

    1997-02-01

    In-pile corrosion of Zr + 1%Nb fuel cladding has been studied. Zr-oxide and hydroxide solubilities at various temperatures and pH values have been calculated and correlations obtained between post-transition corrosion and the solubilities nodular corrosion and fuel operation parameters, as well as between the rate of fuel cladding degradation and water chemistry. Extrapolations of fuel assemblies behaviour to higher burnups have also performed. (author). 12 refs, 11 figs.

  19. Development of dynamic simulation code for fuel cycle of fusion reactor. 1. Single pulse operation simulation

    Aoki, Isao; Seki, Yasushi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1997-11-01

    A dynamic simulation code for the fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during a single pulse operation. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the function of fuel burn, exhaust, purification, and supply. The processing constants of subsystem for the steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using the code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  20. Operating performance and reliability of CANDU PHWR fuel channels in Canada

    Strachan, B.; Brown, D.R.

    1983-03-01

    CANDU nuclear plants use many small-diameter high-pressure fuel channels. Good operating performance from the CANDU fuel channels has made a major contribution to the world-leading operating record of the CANDU nuclear power plants. As of 1982 December 31, there were 7,480 fuel channels installed in 18 CANDU reactors over 500 MW(e) in size. Eight of these reactors have been declared in-service and have accumulated 24,000 fuel channel-years of operation. The only significant operating problems with fuel channels have been the occurrence of leaking cracks in 70 fuel channels and a larger amount of axial creep on the early reactors than was originally provided for in the design. Both of these problems have been corrected on all CANDU reactors built since the Bruce GS 'A' station and the newer reactors should exhibit even better performance

  1. Efficiency improvement of nuclear power plant operation: the significant role of advanced nuclear fuel technologies

    Velde Van de, A.; Burtak, F.

    2001-01-01

    Due to the increased liberalisation of the power markets, nuclear power generation is being exposed to high cost reduction pressure. In this paper we highlight the role of advanced nuclear fuel technologies to reduce the fuel cycle costs and therefore increase the efficiency of nuclear power plant operation. The key factor is a more efficient utilisation of the fuel and present developments at Siemens are consequently directed at (i) further increase of batch average burnup, (ii) improvement of fuel reliability, (iii) enlargement of fuel operation margins and (iv) improvement of methods for fuel design and core analysis. As a result, the nuclear fuel cycle costs for a typical LWR have been reduced during the past decades by about US$ 35 million per year. The estimated impact of further burnup increases on the fuel cycle costs is expected to be an additional saving of US$10 - 15 million per year. Due to the fact that the fuel will operate closer to design limits, a careful approach is required when introducing advanced fuel features in reload quantities. Trust and co-operation between the fuel vendors and the utilities is a prerequisite for the common success. (authors)

  2. PWR Fuel licensing in France - from design to reprocessing: licensing of nuclear PWR fuel rod design to satisfy with criteria for normal and abnormal fuel operation in France

    Beraha, R.

    1999-01-01

    In this lecture are presented: French regulatory context; Current fuel management methods; Request from the french operator EdF; Most recent actions of the french Nuclear safety authority; Fuel assemblies deformations (impact of high burn-up; investigations during reactor's exploitation; control rods drop off times)

  3. Administrative and managerial controls for the operation of nuclear fuel reprocessing plants

    Anon.

    1979-01-01

    Guidelines are provided for the administrative and managerial controls necessary for the safe and efficient operation of nuclear fuel reprocessing plants. Topics covered include: administrative organization; review and audit; facility administrative policies and procedures; and tests and inspections. Recognizing that administrative practices vary among organizations operating nuclear fuel reprocessing plants, the standard incorporates flexibility that provides for compliance by any organization

  4. Operation of CANDU power reactor in thorium self-sufficient fuel cycle

    This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the ...

  5. The Effect of the Existence of Defective Items in Assembly Operations

    Eben-Chaime , Moshe

    2014-01-01

    Part 1: Knowledge-Based Performance Improvement; International audience; Quality is a principle issue in production management (PM). No process is perfect and the production of defective items is unavoidable. Very few studies regard the effect of the existence of defective items (EEDI) in production processes. Further, quality has been studied in isolation to high extent, of other PM domains. In this study, defect rates together with the assembly ratios of the bill of material are embedded in...

  6. Thermal behaviour of fuel: influence on the behavior of fuel elements in nominal and incidental operating conditions

    Languille, A.

    1984-02-01

    The behaviour of the oxide, in normal conditions as well as in incidental conditions is an important care at the fuel element design level in a fast reactor. In nominal operating conditions, the probability of melt to core of the pellet is very low and even for high burnup. The behaviour in incidental operating conditions is also satisfying, especially for inadvertent rod ejections [fr

  7. Low-temperature fuel cells operating with contaminated feedstock

    Wingelaar, P.J.H.

    2007-01-01

    This work concerns the analysis and modeling of the dynamic and static behavior of Polymer Electrolyte Membrane Fuel Cells (PEMFC). Three fundamentally different measurement methods are used to determine the static, the large-signal, and the small-signal dynamic behavior of a fuel cell system. By

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

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

    1986-01-01

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

  9. Near-optimal operation of dual-fuel launch vehicles

    Ardema, M.D.; Chou, H.C.; Bowles, J.V.

    1994-01-01

    Current studies of single-stage-to-orbit (SSTO) launch vehicles are focused on all-rocket propulsion systems. One option for such vehicles is the use of dual-fuel (liquid hydrocarbon and liquid hydrogen (LH 2 )), for a portion of the mission. As compared with LH 2 , hydrocarbon fuel has higher density and produces higher thrust-to-weight, but has lower specific impulse. The advantages of hydrocarbon fuel are important early in the ascent trajectory, and its use may be expected to lead to reduced vehicle size and weight. Because LH 2 is also needed for cooling purposes, in the early portion of the trajectory both fuels must be burned simultaneously. Later in the ascent, when vehicle weight is lower, specific impulse is the key parameter, indicating single-fuel LH 2 use

  10. Feasibility and incentives for the consideration of spent fuel operating histories in the criticality analysis of spent fuel shipping casks

    Sanders, T.L.; Westfall, R.M.; Jones, R.H.

    1987-08-01

    Analyses have been completed that indicate the consideration of spent fuel histories (''burnup credit'') in the design of spent fuel shipping casks is a justifiable concept that would result in cost savings and public risk benefits in the transport of spent nuclear fuel. Since cask capacities could be increased over those of casks without burnup credit, the number of shipments necessary to transport a given amount of fuel could be reduced. Reducing the number of shipments would increase safety benefits by reducing public and occupational exposure to both radiological and nonradiological risks associated with the transport of spent fuel. Economic benefits would include lower in-transit shipping, reduced transportation fleet capital costs, and reduced numbers of cask handling operations at both shipping and receiving facilities. 44 refs., 66 figs., 28 tabs

  11. Physical security in multinational nuclear-fuel-cycle operations

    Willrich, M.

    1977-01-01

    Whether or not multinationalization will reduce or increase risks of theft or sabotage will depend on the form and location of the enterprise, the precise nature of the physical security arrangements applied to the enterprise, and the future course of crime and terrorism in the nuclear age. If nuclear operations are multinationalized, the host government is likely to insist on physical security measures that are at least as stringent as those for a national or private enterprise subject to its jurisdiction. At the same time, the other participants will want to be sure the host government, as well as criminal groups, do not steal nuclear material from the facility. If designed to be reasonably effective, the physical security arrangements at a multinational nuclear enterprise seem likely to reduce the risk that any participating government will seek to divert material from the facility for use in a nuclear weapons program. Hence, multinationalization and physical security will both contribute to reducing the risks of nuclear weapons proliferation to additional governments. If economic considerations dominate the timing, scale and location of fuel-cycle facilities, the worldwide nuclear power industry is likely to develop along lines where the problems of physical security will be manageable. If, however, nuclear nationalism prevails, and numerous small-scale facilities become widely dispersed, the problem of security against theft and sabotage may prove to be unmanageable. It is ironic, although true, that in attempting to strengthen its security by pursuing self-sufficiency in nuclear power, a nation may be reducing its internal security against criminal terrorists

  12. Turbojet Performance and Operation at High Altitudes with Hydrogen and Jp-4 Fuels

    Fleming, W A; Kaufman, H R; Harp, J L , Jr; Chelko, L J

    1956-01-01

    Two current turbojet engines were operated with gaseous-hydrogen and JP-4 fuels at very high altitudes and a simulated Mach number of 0.8. With gaseous hydrogen as the fuel stable operation was obtained at altitudes up to the facility limit of about 90,000 feet and the specific fuel consumption was only 40 percent of that with JP-4 fuel. With JP-4 as the fuel combustion was unstable at altitudes above 60,000 to 65,000 feet and blowout limits were reached at 75,000 to 80,000 feet. Over-all performance, component efficiencies, and operating range were reduced considerable at very high altitudes with both fuels.

  13. Ab initio Defect Energetics in LaBO3 Perovskite Solid Oxide Fuel Cell Materials

    Lee, Yueh-Lin; Morgan, Dane; Kleis, Jesper

    2009-01-01

    Perovskite materials of the form ABO3 are a promising family of compounds for use in solid oxide fuel cell (SOFC) cathodes. Study of the physics of these compounds under SOFC conditions with ab initio methods is particularly challenging due to high temperatures, exchange of oxygen with O2 gas...

  14. FFTF fuel pin design procedure verification for transient operation

    Baars, R.E.

    1975-05-01

    The FFTF design procedures for evaluating fuel pin transient performance are briefly reviewed, and data where available are compared with design procedure predictions. Specifically, burst conditions derived from Fuel Cladding Transient Tester (FCTT) tests and from ANL loss-of-flow tests are compared with burst pressures computed using the design procedure upon which the cladding integrity limit was based. Failure times are predicted using the design procedure for evaluation of rapid reactivity insertion accidents, for five unterminated TREAT experiments in which well characterized fuel failures were deliberately incurred. (U.S.)

  15. Design and Operation of an Electrochemical Methanol Concentration Sensor for Direct Methanol Fuel Cell Systems

    Narayanan, S. R.; Valdez, T. I.; Chun, W.

    2000-01-01

    The development of a 150-Watt packaged power source based on liquid feed direct methanol fuel cells is being pursued currently at the Jet propulsion Laboratory for defense applications. In our studies we find that the concentration of methanol in the fuel circulation loop affects the electrical performance and efficiency the direct methanol fuel cell systems significantly. The practical operation of direct methanol fuel cell systems, therefore, requires accurate monitoring and control of methanol concentration. The present paper reports on the principle and demonstration of an in-house developed electrochemical sensor suitable for direct methanol fuel cell systems.

  16. Performance Evaluation and Suggestion of Upgraded Fuel Handling Equipment for Operating OPR1000

    Chang, Sang Gyoon; Hwang, Jeung Ki; Choi, Taek Sang; Na, Eun Seok; Lee, Myung Lyul; Baek, Seung Jin; Kim, Man Su; Kunik, Jack

    2011-01-01

    The purpose of this study is to evaluate the performance of upgraded FHE (Fuel Handling Equipment) for operating OPR 1000 (Optimized Power Reactor) by using data measured during the fuel reloading, and make some suggestions on enhancing the performance of the FHE. The fuel handling equipment, which serves critical processes in the refueling outage, has been upgraded to increase and improve the operational availability of the plant. The evaluation and suggestion of this study can be a beneficial tool related to the performance of the fuel handling equipment

  17. Safety standards, legislation and codes of practice for fuel cell manufacture and operation

    Wilcox, C.P.

    1999-07-01

    This report examines safety standards, legislation and codes of practice for fuel cell manufacture and operation in the UK, Europe and internationally. Management of health and safety in the UK is discussed, and the characteristics of phosphoric acid (PAFC), proton exchange membrane (PEM), molten carbonate (MCFC), solid oxide (SOFC) fuel cells are described. Fuel cell power plant standards and manufacture in the UK, design and operational considerations, end of life disposal, automotive fuel cell system, and fuelling and vehicular concerns are explored, and standards, legislation and codes of practice are explained in the appendix.

  18. Impact of Zr + 2.5% Nb alloy corrosion upon operability of RBMK-1000 fuel channels

    Kovyrshin, V.; Zaritsky, N.

    1999-01-01

    The basic components of RBMK-1000 core (fuel channels, bimetal adapters, claddings of fuel elements, etc.) are of zirconium alloys. Their corrosion is one of factors influencing upon fuel channels operability. Dynamics of channel tubes nodular corrosion development is presented by the results of in-reactor investigation at ChNPP. Radiation-induced mechanism of corrosion damage of tubes surface in contact with coolant was formulated and substantiated by data of post-reactor studies. Within the certain time period of operation corrosion of zirconium alloy of lower bimetal adapter along with removal from there of corrosion products are predominant within the whole process of reactor elements corrosion. The experimental and calculating method was proposed and substantiated to predict time duration up to loss of fuel channels leak tightness. The approaches were generalized to control state of fuel channels material to assess their operability under operation of RBMK-1000 reactors. (author)

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

    Bodey, Isaac T [ORNL

    2014-05-01

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

  20. EURATOM work on standard defects and dimensional measurements in neutron radiography of nuclear fuel elements

    Domanus, J.C.

    1981-10-01

    In 1979 a working group on neutron radiography was formed at Euratom. The purpose of this group is the standardization of neutron radiographic methods in the field of nuclear fuel. First priority was given to the development of image quality indicators and standard objects for the determination of accuracy of dimensional measurements from neutron radiographs. For that purpose beam purity and sensitivity indicators as well as a calibration fuel pin were designed and fabricated at Risoe. All the Euratom neutron radiography centers have recieved the above items for comparative neutron radiography. The measuring results obtained, using various measuring apparatus, will form the basis to formulate conclusions about the best measuring methods and instruments to be used in that field. (author)

  1. The Effect of Uncertainties on the Operating Temperature of U-Mo/Al Dispersion Fuel

    Sweidana, Faris B.; Mistarihia, Qusai M.; Ryu Ho Jin [KAIST, Daejeon (Korea, Republic of); Yim, Jeong Sik [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, uncertainty and combined uncertainty studies have been carried out to evaluate the uncertainty of the parameters affecting the operational temperature of U-Mo/Al fuel. The uncertainties related to the thermal conductivity of fuel meat, which consists of the effects of thermal diffusivity, density and specific heat capacity, the interaction layer (IL) that forms between the dispersed fuel and the matrix, fuel plate dimensions, heat flux, heat transfer coefficient and the outer cladding temperature were considered. As the development of low-enriched uranium (LEU) fuels has been pursued for research reactors to replace the use of highly-enriched uranium (HEU) for the improvement of proliferation resistance of fuels and fuel cycle, U-Mo particles dispersed in an Al matrix (UMo/Al) is a promising fuel for conversion of the research reactors that currently use HEU fuels to LEUfueled reactors due to its high density and good irradiation stability. Several models have been developed for the estimation of the thermal conductivity of U–Mo fuel, mainly based on the best fit of the very few measured data without providing uncertainty ranges. The purpose of this study is to provide a reasonable estimation of the upper bounds and lower bounds of fuel temperatures with burnup through the evaluation of the uncertainties in the thermal conductivity of irradiated U-Mo/Al dispersion fuel. The combined uncertainty study using RSS method evaluated the effect of applying all the uncertainty values of all the parameters on the operational temperature of U-Mo/Al fuel. The overall influence on the value of the operational temperature is 16.58 .deg. C at the beginning of life and it increases as the burnup increases to reach 18.74 .deg. C at a fuel meat fission density of 3.50E+21 fission/cm{sup 3}. Further studies are needed to evaluate the behavior more accurately by including other parameters uncertainties such as the interaction layer thermal conductivity.

  2. Change in geometrical parameters of WWER high burnup fuel rods under operational conditions and transient testing

    Kanashov, B.; Amosov, S.; Lyadov, G.; Markov, D.; Ovchinnikov, V; Polenok, V.; Smirnov, A.; Sukhikh, A.; Bek, E.; Yenin, A.; Novikov, V.

    2001-01-01

    The paper discusses changes in fuel rods geometric parameters as result of operation conditions and burnups. The degree of geometry variability of fuel rods, cladding and column is one of the most important characteristics affecting fuel serviceability. On the other hand, changes in fuel rod geometric parameters influence fuel temperature, fission gas release, fuel-to-cladding stress strained state as well as the degree of interaction with FA skeleton elements and skeleton rigidity. Change in fuel-to-cladding gap is measured using compression technique. The axial distribution of fuel-to-cladding gap demonstrates the largest decrease of the gap in the region 500 to 2000 mm from the bottom of the fuel rod (WWER-440) and in the region of 500 to 3000 mm for WWER-1000. The cladding material creep in WWER fuel rods together with the radiation growth results in fuel rod cladding elongation. A set of transient tests for spent WWER-440 and WWER-1000 fuel rods carried out in SSC RIAR during a period 1995-1999, with the aim to estimate the changes in geometric parameters of FRs. The estimation of changes in outer diameter of cladding and fuel column and fuel-to-cladding gap are performed in transient conditions (changes in linear power range of 180 to 400 W/cm) for both WWER-440 and WWER-1000. WWER-440 fuel rods having the same burnup and close fuel-cladding contact before testing are subjected to considerable hoop cladding strain in testing up to 300 W/cm. But the hoop strain does not grow due to the structural changes in fuel column and decrease in central hole diameter occurred when the power is higher

  3. Improvement of the environmental and operational characteristics of vehicles through decreasing the motor fuel density.

    Magaril, Elena

    2016-04-01

    The environmental and operational characteristics of motor transport, one of the main consumers of motor fuel and source of toxic emissions, soot, and greenhouse gases, are determined to a large extent by the fuel quality which is characterized by many parameters. Fuel density is one of these parameters and it can serve as an indicator of fuel quality. It has been theoretically substantiated that an increased density of motor fuel has a negative impact both on the environmental and operational characteristics of motor transport. The use of fuels with a high density leads to an increase in carbonization within the engine, adversely affecting the vehicle performance and increasing environmental pollution. A program of technological measures targeted at reducing the density of the fuel used was offered. It includes a solution to the problem posed by changes in the refining capacities ratio and the temperature range of gasoline and diesel fuel boiling, by introducing fuel additives and adding butanes to the gasoline. An environmental tax has been developed which allows oil refineries to have a direct impact on the production of fuels with improved environmental performance, taking into account the need to minimize the density of the fuel within a given category of quality.

  4. Trial operation of a phosphoric acid fuel cell (PC25) for CHP applications in Europe

    Uhrig, M.; Droste, W.; Wolf, D. [Ruhrgas AG, Dorsten (Germany)

    1996-12-31

    In Europe, ten 200 kW phosphoric acid fuel cells (PAFCs) produced by ONSI (PC25) are currently in operation. Their operators collaborate closely in the European Fuel Cell Users Group (EFCUG). The experience gained from trial operation by the four German operators - HEAG, HGW/HEW, Thyssengas and Ruhrgas - coincides with that of the other European operators. This experience can generally be regarded as favourable. With a view to using fuel cells in combined heat and power generation (CHP), the project described in this report, which was carried out in cooperation with the municipal utility of Bochum and Gasunie of the Netherlands, aimed at gaining experience with the PC 25 in field operation under the specific operating conditions prevailing in Europe. The work packages included heat-controlled operation, examination of plant behavior with varying gas properties and measurement of emissions under dynamic load conditions. The project received EU funding under the JOULE programme.

  5. Analysis of long-time operation of micro-cogeneration unit with fuel cell

    Patsch Marek

    2015-01-01

    Full Text Available Micro-cogeneration is cogeneration with small performance, with maximal electric power up to 50 kWe. On the present, there are available small micro-cogeneration units with small electric performance, about 1 kWe, which are usable also in single family houses or flats. These micro-cogeneration units operate on principle of conventional combustion engine, Stirling engine, steam engine or fuel cell. Micro-cogeneration units with fuel cells are new progressive developing type of units for single family houses. Fuel cell is electrochemical device which by oxidation-reduction reaction turn directly chemical energy of fuel to electric power, secondary products are pure water and thermal energy. The aim of paper is measuring and evaluation of operation parameters of micro-cogeneration unit with fuel cell which uses natural gas as a fuel.

  6. Mechanical behaviour of PEM fuel cell catalyst layers during regular cell operation

    Maher A.R. Sadiq Al-Baghdadi

    2010-01-01

    Damage mechanisms in a proton exchange membrane fuel cell are accelerated by mechanical stresses arising during fuel cell assembly (bolt assembling), and the stresses arise during fuel cell running, because it consists of the materials with different thermal expansion and swelling coefficients. Therefore, in order to acquire a complete understanding of the mechanical behaviour of the catalyst layers during regular cell operation, mechanical response under steady-state hygro-thermal stresses s...

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

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

    1998-01-01

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

  8. Analytical and experimental justification of safe operation of fuel loads of VVER reactors at Rovno NPP

    Andrianov, A.; Zagrebelny, L.

    2011-01-01

    The main task during the nuclear fuel operation us ensuring of the larger fuel burnup and as a result - reduction of the fuel constituent in the electricity production cost. The neutron-physic calculations are performed using the qualified codes BIPR-7A and PERMAK developed by Kurchatov Institute. The limits for calculated parameters are set by the Ukrainian regulations. Calculation results are documented in reports subject for independent expert review requested by the regulatory authority. In this report the following item have presented: 1) metrological check and calibration of measuring channels; 2) fuel cycles at Rivne NPP; 3) determination of experimental values of thermal-physic and neutron-physic parameters; 4) ICIS equipment check, execution of the program confirming correct connection of the temperature and neutron flux monitoring sensors; 5) monitoring of the core TPh and NPh parameters in all operating modes; 6) monitoring of the fuel condition in the core and 7) FE leak tightness monitoring at the operating reactor

  9. Alternative concepts for spent fuel storage basin expansion at Morris Operation

    Graf, W.A. Jr.; King, C.E.; Miller, G.P.; Shadel, F.H.; Sloat, R.J.

    1980-08-01

    Alternative concepts for increasing basin capabilities for storage of spent fuel at the Morris Operation have been defined in a series of simplified flow diagrams and equipment schematics. Preliminary concepts have been outlined for (1) construction alternatives for an add-on basin, (2) high-density baskets for storage of fuel bundles or possible consolidated fuel rods in the existing or add-on basins, (3) modifications to the existing facility for increasing cask handling and fuel receiving capabilities and (4) accumulation, treatment and disposal of radwastes from storage operations. Preliminary capital and operating costs have been prepared and resource and schedule requirements for implementing the concepts have been estimated. The basin expansion alternatives would readily complement potential dry storage projects at the site in an integrated multi-stage program that could provide a total storage capacity of up to 7000 tonnes of spent fuel

  10. Handling and transfer operations for partially-spent nuclear fuel

    Ibrahim, J K [PUSPATI, Kuala Lumpur (Malaysia)

    1983-12-01

    This project involved the handling and transfer of partially-spent reactor fuel from the Oregon State University TRIGA Reactor in Corvallis, Oregon to Hanford Engineering Development Laboratory in Richland, Washington. The method of handling is dependent upon the burn-up history of the fuel elements. Legal constraints imposed by standing U.S. nuclear regulations determine the selection of transport containers, transportation procedures, physical security arrangements in transit and nuclear material accountability documentation. Results of in-house safety evaluations of the project determine the extent of involvement of pertinent nuclear regulatory authorities. The actual handling activities and actual radiation dose rates are also presented.

  11. Bruce B fuelling-with-flow operations: fuel damage investigation

    Manzer, A.M. [CANTECH Associates Ltd., Burlington, Ontario (Canada); Morikawa, D. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Hains, A.J.; Cichowlas, W.M. [Nuclear Safety Solutions Limited, Toronto, Ontario (Canada); Roberts, J.G.; Wylie, J. [Bruce Power, Ontario (Canada)

    2005-07-01

    This paper summarizes the fuel bundle damage characterization done by Nuclear Safety Solutions Limited (NSS) and the out-reactor flow visualization tests done at Atomic Energy of Canada Limited (AECL) to reproduce the damage observed on irradiated fuel bundles. The bearing pad damage mechanism was identified and the tests showed that a minor change to the fuelling sequence would eliminate the mechanical interaction. The change was implemented in January 2005. Since then, the bearing pad damage appears to have been greatly reduced based on the small number of discharged bundles inspected to date. (author)

  12. Bruce B fuelling-with-flow operations: fuel damage investigation

    Manzer, A.M.; Morikawa, D.; Hains, A.J.; Cichowlas, W.M.; Roberts, J.G.; Wylie, J.

    2005-01-01

    This paper summarizes the fuel bundle damage characterization done by Nuclear Safety Solutions Limited (NSS) and the out-reactor flow visualization tests done at Atomic Energy of Canada Limited (AECL) to reproduce the damage observed on irradiated fuel bundles. The bearing pad damage mechanism was identified and the tests showed that a minor change to the fuelling sequence would eliminate the mechanical interaction. The change was implemented in January 2005. Since then, the bearing pad damage appears to have been greatly reduced based on the small number of discharged bundles inspected to date. (author)

  13. Handling and transfer operations for partially-spent nuclear fuel

    Ibrahim, J.K.

    1983-01-01

    This project involved the handling and transfer of partially-spent reactor fuel from the Oregon State University TRIGA Reactor in Corvallis, Oregon to Hanford Engineering Development Laboratory in Richland, Washington. The method of handling is dependent upon the burn-up history of the fuel elements. Legal constraints imposed by standing U.S. nuclear regulations determine the selection of transport containers, transportation procedures, physical security arrangements in transit and nuclear material accountability documentation. Results of in-house safety evaluations of the project determine the extent of involvement of pertinent nuclear regulatory authorities. The actual handling activities and actual radiation dose rates are also presented (author)

  14. The source regime for irradiation plant operated with fuel elements

    Suckow, W.

    1976-11-01

    The rapid and irregular decay of the gamma radiation from reactor fuel elements requires the establishment of an optimal source regime in order to utilise reactor fuel elements as radiation sources on a technological basis. Critical values have been derived which enable the determination of optimal conditions. In this context all technologically interesting types of source regimes have been examined. Methods to achieve a high gamma yield and a satisfactory dose consistency with time have been developed and important values for these two aspects have been derived. The conditions for optimal radiation source regimes are described in the final conclusions. (author)

  15. Spent Nuclear Fuel (SNF) Startup Plan to Operations

    GREGORY, J.R.

    2000-01-01

    This plan defines the approach that will be used to ensure the transition from initial startup to normal operations of the SNF operations--are performed in a safe, controlled, and deliberate manner. It provides a phased approach that bridges the operations between the completion of the ORR and the return to normal operations. This plan includes management oversight and administrative controls to be implemented and then reduced in a controlled manner until normal operations are authorized by SNF Management

  16. Recent advances and achievements in WWER-1000 fuel design performance and operation

    Vasilchenko, I.; Molchanov, V.

    2009-01-01

    In this paper the main results of TVS-2 (TVS-2M) basic design operation like TVS-2 - 1216 pcs., including 687 in operation, TVS-2M - 66 pcs. in operation (Balakovo-1); reliability -1,6·10-6 in 2008 and the Balakovo NPP capacity factor - 90% are presented. Average efficient operating time of the Balakovo power units, calculated capacity factor for the Balakovo NPP at transition to fuel cycle 3x1,5, contemporary operational requirements for fuel and their realization as well as SPND axial position variation in the core with TVS - 2M and CPS AR usage are shown

  17. Core design options for high conversion BWRs operating in Th–233U fuel cycle

    Shaposhnik, Y.; Shwageraus, E.; Elias, E.

    2013-01-01

    Highlights: • BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. • Seed blanket optimization that includes assembly size array and axial dimensions. • Fully coupled MC with fuel depletion and thermo-hydraulic feedback modules. • Thermal-hydraulic analysis includes MCPR observation. -- Abstract: Several options of fuel assembly design are investigated for a BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. The designs rely on an axially heterogeneous fuel assembly structure consisting of a single axial fissile zone “sandwiched” between two fertile blanket zones, in order to improve fertile to fissile conversion ratio. The main objective of the study was to identify the most promising assembly design parameters, dimensions of fissile and fertile zones, for achieving net breeding of 233 U. The design challenge, in this respect, is that the fuel breeding potential is at odds with axial power peaking and the core minimum critical power ratio (CPR), hence limiting the maximum achievable core power rating. Calculations were performed with the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly having reflective radial boundaries was modeled applying simplified restrictions on the maximum centerline fuel temperature and the CPR. It was found that axially heterogeneous fuel assembly design with a single fissile zone can potentially achieve net breeding, while matching conventional BWR core power rating under certain restrictions to the core loading pattern design

  18. RESULTS OF AIRPLANE TU-204-300 OPERATED BY "VLADIVOSTOK AVIA" COMPANY FUEL FLOW MONITIRING

    G. E. Maslennikova

    2014-01-01

    Full Text Available This article describes the results obtained from continuous monitoring of fuel flow on the airplanes Tu-204-300, operated by the aircompany "Vladivostok Avia". The reasons for the change in the cost of each copy of airplane and changes in fuel characteristics due to pilots manner of flying.

  19. Experience with fuel damage caused by abnormal conditions in handling and transporting operations

    Bailey, W.J.

    1983-01-01

    Pacific Northwest Laboratory (PNL) conducted a study to determine the expected condition of spent USA light-water reactor (LWR) fuel upon arrival at interim storage or fuel reprocessing facilities or, if fuel is declared a waste, at disposal facilities. Initial findings were described in an earlier PNL paper at PATRAM '80 and in a report. Updated findings are described in this paper, which includes an evaluation of information obtained from the literature and a compilation of cases of known or suspected damage to fuel as a result of handling and/or transporting operations. To date, PNL has evaluated 123 actual cases (98 USA and 25 non-USA). Irradiated fuel was involved in all but 10 of the cases. From this study, it is calculated that the frequency of unusual occurrences involving fuel damage from handling and transporting operations has been low. The damage that did occur was generally minor. The current base of experience with fuel handling and transporting operations indicates that nearly all of these unusual occurrences had only a minor or negligible effect on spent fuel storage facility operations

  20. Experiences with fuels B30 and B 100 in haulage, railway operation and agricultural machinery

    Matejovsky, V. [QMS Consulting, Prague (Czech Republic); Hendrych, K.; Mares, V. [PREOL, Lovosice (Czech Republic)

    2013-06-01

    High prices of diesel fuel have increased an interest in cheaper biodiesel, especially for vehicles with high fuel consumption and not only for haulage vehicle parks but also for railway vehicles and heavy agricultural machinery. When price difference between standard diesel B7 and cheaper biodiesel B100 reached more than 10% it was a sufficient benefit for operators to use biodiesel but this fuel had not been approved for all vehicles types by their manufacturers. Despite this problem, some operators have begun to use biodiesel also for vehicles not having the approval. To prevent operational problems and misgiving of engines damage, the transition to alternative fuel was organized as field tests of one or more vehicles from the operator's fleet. The tests usually started with B30 fuel and if no operational problems occurred the second stage continued with B100. The tested vehicles were under permanent surveillance at least during one year of operation and once a month and later once in a quarter a deeper inspections were made including engine diagnostics, emissions testing, engine oil sampling for laboratory examination, injectors tenting and filters and fuel hoses condition evaluation. The presentation includes the results of vehicles inspections and the measures that had to be done to prevent engines failure and to ensure trouble-free operation of vehicles using biofuels. (orig.)

  1. Nuclear fuels

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

    2009-01-01

    , Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO 2 ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO 2 and MOX ceramics (Chromium oxide-doped UO 2 fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The nature of spent nuclear

  2. Nuclear fuels

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

    2009-07-01

    irradiation, Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO{sub 2} ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO{sub 2} and MOX ceramics (Chromium oxide-doped UO{sub 2} fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The

  3. Thermodynamic analysis of solid oxide fuel cell gas turbine systems operating with various biofuels

    Patel, H.C.; Woudstra, T.; Aravind, P.V. [Process and Energy Laboratory, Delft University of Technology, Section Energy Technology, Leeghwaterstraat 44, 2628 CA Delft (Netherlands)

    2012-12-15

    Solid oxide fuel cell-gas turbine (SOFC-GT) systems provide a thermodynamically high efficiency alternative for power generation from biofuels. In this study biofuels namely methane, ethanol, methanol, hydrogen, and ammonia are evaluated exergetically with respect to their performance at system level and in system components like heat exchangers, fuel cell, gas turbine, combustor, compressor, and the stack. Further, the fuel cell losses are investigated in detail with respect to their dependence on operating parameters such as fuel utilization, Nernst voltage, etc. as well as fuel specific parameters like heat effects. It is found that the heat effects play a major role in setting up the flows in the system and hence, power levels attained in individual components. The per pass fuel utilization dictates the efficiency of the fuel cell itself, but the system efficiency is not entirely dependent on fuel cell efficiency alone, but depends on the split between the fuel cell and gas turbine powers which in turn depends highly on the nature of the fuel and its chemistry. Counter intuitively it is found that with recycle, the fuel cell efficiency of methane is less than that of hydrogen but the system efficiency of methane is higher. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  4. Emission and operating performance of a biomethane tractor with dual fuel engine

    Mautner, Sebastian; Emberger, Peter; Thuneke, Klaus; Remmele, Edgar

    2016-01-01

    The use of biomethane as fuel for agricultural machinery with dual fuel technology is contributing to climate protection and ensures safe fuel supply. So far, hardly any documented operational experiences are known. The aim of the project, funded by the Bavarian Ministry of Economic Affairs and Media, Energy and Technology, was to investigate practicability for daily use and the emission behaviour of a Valtra N101 prototype tractor (exhaust stage IIIA). The retrofitted dual-fuel technology of the former conventional diesel tractor simultaneously uses biomethane or natural gas and diesel as ignition fuel. During the field test over 590 working hours, the tractor showed overall high reliability. On average the operating range in dual-fuel mode with one complete filling of the gas tanks was about 11.5 hours. On the tractor test bench a significant improvement of the exhaust emissions could be observed, since the gas ECU had been optimized and changed by the manufacturer. For dual-fuel operation, nitrogen oxides (NO x ) are lower, whereas carbon monoxide (CO), hydrocarbons (HC) and particulate matter emissions (PM) are higher compared to solely diesel operation. In particular, HC emissions exceed the proposed limiting value, submitted by the European Commission. This is due to incomplete gas combustion and insufficient conversion by the exhaust after-treatment-system (methane slip). A big potential for optimization is expected by adjusting the operating point-specific gasdiesel ratio and improving the exhaust gas aftertreatment system.

  5. Environmental and ventilation benefits for underground mining operations using fuel cell powered production equipment

    Kocsis, C.; Hardcastle, S.

    2007-01-01

    The benefits of replacing diesel engines with fuel cells in mine production equipment were discussed. The paper was part of a multi-year feasibility study conducted to evaluate the use of hydrogen fuel cell-powered equipment to replace diesel engine powered equipment in underground mining operations. The feasibility study demonstrated that fuel cells are capable of eliminating the unwanted by-products of combustion engines. However, the use of fuel cells also reduced the amount of ventilation that mines needed to supply, thereby further reducing energy consumption. This study examined the benefits of replacing diesel engines with fuel cells, and discussed the mitigating qualifiers that may limit ventilation energy savings. Solutions to retaining and maintaining additional ventilation in the event of hydrogen leaks from fuel cell stacks were also investigated. The analyses were conducted on 6 operating mines. Current operating costs were compared with future operating conditions using fuel cell powered production vehicles. Operating costs of the primary ventilation system were established with a mine ventilation simulator. The analysis considered exhaust shaft velocities, heating system air velocities, and levels of silica exposure. Canadian mine design criteria were reviewed. It was concluded that appropriate safeguards are needed along hydrogen distribution lines to lower the impacts of hydrogen leaks. Large financial commitments may also be required to ensure a spark-free environment. 20 refs., 6 tabs., 3 figs

  6. Simulated first operating campaign for the Integral Fast Reactor fuel cycle demonstration

    Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Park, K.H.; Ackerman, J.P.

    1993-01-01

    This report discusses the Integral Fast Reactor (IFR) which is an innovative liquid-metal-cooled reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid-metal cooling to offer significant improvements in reactor safety, operation, fuel cycle-economics, environmental protection, and safeguards. Over the next few years, the IFR fuel cycle will be demonstrated at Argonne-West in Idaho. Spent fuel from the Experimental Breeder Reactor II (EBR-II) win be processed in its associated Fuel Cycle Facility (FCF) using a pyrochemical method that employs molten salts and liquid metals in an electrorefining operation. As part of the preparation for the fuel cycle demonstration, a computer code, PYRO, was developed at Argonne to model the electrorefining operation using thermodynamic and empirical data. This code has been used extensively to evaluate various operating strategies for the fuel cycle demonstration. The modeled results from the first operating campaign are presented. This campaign is capable of processing more than enough material to refuel completely the EBR-II core

  7. Solid Oxide Fuel Cell Based Upon Colloidal Deposition of Thin Films for Lower Temperature Operation (Preprint)

    Reitz, T. L; Xiao, H

    2006-01-01

    In order to reduce the operating temperature of solid oxide fuel cells (SOFCs), anode-supported cells incorporating thin film electrolytes in conjunction with anode/electrolyte and cathode/electrolyte interlayers were studied...

  8. Safety and operations of hydrogen fuel infrastructure in northern climates : a collaborative complex systems approach.

    2010-10-07

    "This project examined the safety and operation of hydrogen (H2) fueling system infrastructure in : northern climates. A multidisciplinary team lead by the University of Vermont (UVM), : combined with investigators from Zhejiang and Tsinghua Universi...

  9. Optimal design and operation of solid oxide fuel cell systems for small-scale stationary applications

    Braun, Robert Joseph

    The advent of maturing fuel cell technologies presents an opportunity to achieve significant improvements in energy conversion efficiencies at many scales; thereby, simultaneously extending our finite resources and reducing "harmful" energy-related emissions to levels well below that of near-future regulatory standards. However, before realization of the advantages of fuel cells can take place, systems-level design issues regarding their application must be addressed. Using modeling and simulation, the present work offers optimal system design and operation strategies for stationary solid oxide fuel cell systems applied to single-family detached dwellings. A one-dimensional, steady-state finite-difference model of a solid oxide fuel cell (SOFC) is generated and verified against other mathematical SOFC models in the literature. Fuel cell system balance-of-plant components and costs are also modeled and used to provide an estimate of system capital and life cycle costs. The models are used to evaluate optimal cell-stack power output, the impact of cell operating and design parameters, fuel type, thermal energy recovery, system process design, and operating strategy on overall system energetic and economic performance. Optimal cell design voltage, fuel utilization, and operating temperature parameters are found using minimization of the life cycle costs. System design evaluations reveal that hydrogen-fueled SOFC systems demonstrate lower system efficiencies than methane-fueled systems. The use of recycled cell exhaust gases in process design in the stack periphery are found to produce the highest system electric and cogeneration efficiencies while achieving the lowest capital costs. Annual simulations reveal that efficiencies of 45% electric (LHV basis), 85% cogenerative, and simple economic paybacks of 5--8 years are feasible for 1--2 kW SOFC systems in residential-scale applications. Design guidelines that offer additional suggestions related to fuel cell

  10. The time course of an X-ray picture following operation for the ventricular septal defect

    Khenynya, R.L.; Latsis, A.T.; Rubene, M.Ya.

    1983-01-01

    An x-ray picture was studied over time in 80 pediatric patients by hemodynamic groUps in accordance with the classification adopted in the A.I.Bakulev Institute of Cardiovascular Surgery, USSR AMS. Of them in 15 patients in addition to the ventricUlar septal defect, stenosis of the pulmonary artery was revealed, in 11 concomitant insUfficiency of the tricuspid valve. The best results after radical correction of the defect were achieved in patients with insignificant or moderate disorder of the pulmonary hemdynamics; positive results in patients with a high pulmonary hypertension were observed over time later. X-ray studies on the regression of disorders of the pulmonary hemodynamics emphasize the necessity of early diagnosis of the ventricular septal defect followed by surgical correction

  11. Time course of an X-ray picture following operation for the ventricular septal defect

    Khenynya, R.L.; Latsis, A.T.; Rubene, M.Ya.

    1983-01-01

    An x-ray picture was studied over time in 80 pediatric patients by hemodynamic groups in accordance with the classification adopted in the A.I. Bakulev Institute of Cardiovascular Surgery, USSR AMS. Of them in 15 patients in addition to the ventricular septal defect, stenosis of the pulmonary artery was revealed, in 11 concomitant insufficiency of the tricuspid valve. The best results after radical correction of the defect were achieved in patients with insignificant or moderate disorder of the pulmonary hemodynamics; positive results in patients with a high pulmonary hypertension were observed over time later. X-ray studies on the regression of disorders of the pulmonary hemodynamics emphasize the necessity of early diagnosis of the ventricular septal defect followed by surgical correction.

  12. HTGR fuel reprocessing pilot plant: results of the sequential equipment operation

    Strand, J.B.; Fields, D.E.; Kergis, C.A.

    1979-05-01

    The second sequential operation of the HTGR fuel reprocessing cold-dry head-end pilot plant equipment has been successfully completed. Twenty standard LHGTR fuel elements were crushed to a size suitable for combustion in a fluid bed burner. The graphite was combusted leaving a product of fissile and fertile fuel particles. These particles were separated in a pneumatic classifier. The fissile particles were fractured and reburned in a fluid bed to remove the inner carbon coatings. The remaining products are ready for dissolution and solvent extraction fuel recovery

  13. Future trends in nuclear fuels

    Guitierrez, J.E.

    2006-01-01

    This series of transparencies presents: the fuel management cycle and key areas (security of supplies, strategies and core management, reliability, spent fuel management), the world nuclear generating capacity, concentrate capacity, enrichment capacity, and manufacturing capacity forecasts, the fuel cycle strategies and core management (longer cycles, higher burnups, power up-rates, higher enrichments), the Spanish nuclear generation cost, the fuel reliability (no defects, robust designs, operational margins, integrated fuel and core design), spent fuel storage (design and safety criteria, fuel performance and integrity). (J.S.)

  14. Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

    Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A. [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

  15. Experimental investigation on dual fuel operation of acetylene in a DI diesel engine

    Lakshmanan, T. [Department of Mechanical Engineering, Rajarajeswari Engineering College, Adayalampattu, Chennai, 600095 (India); Nagarajan, G. [Internal Combustion Engineering Division, College of Engineering, Anna University, Chennai, 600025 (India)

    2010-05-15

    Depletion of fossils fuels and environmental degradation have prompted researchers throughout the world to search for a suitable alternative fuel for diesel engine. One such step is to utilize renewable fuels in diesel engines by partial or total replacement of diesel in dual fuel mode. In this study, acetylene gas has been considered as an alternative fuel for compression ignition engine, which has excellent combustion properties. Investigation has been carried out on a single cylinder, air cooled, direct injection (DI), compression ignition engine designed to develop the rated power output of 4.4 kW at 1500 rpm under variable load conditions, run on dual fuel mode with diesel as injected primary fuel and acetylene inducted as secondary gaseous fuel at various flow rates. Acetylene aspiration resulted in lower thermal efficiency. Smoke, HC and CO emissions reduced, when compared with baseline diesel operation. With acetylene induction, due to high combustion rates, NO{sub x} emission significantly increased. Peak pressure and maximum rate of pressure rise also increased in the dual fuel mode of operation due to higher flame speed. It is concluded that induction of acetylene can significantly reduce smoke, CO and HC emissions with a small penalty on efficiency. (author)

  16. Standard for assessment of fuel integrity under anticipated operational occurrences in BWR power plant:2002

    Mishima, Kaichiro; Suzuki, Riichiro; Komura, Seiichi; Kudo, Yoshiro; Yamanaka, Akihiro; Oomizu, Satoru; Kitamura, Hideya; Nagata, Yoshifumi

    2003-01-01

    To secure fuel integrity, a Light Water Reactor (LWR) core is designed so that no boiling transition (BT) should take place in fuel assemblies and excessive rise in fuel cladding temperature due to deteriorated that transfer should be avoided in Anticipated Operational Occurrences (AOO). In some AOO in a Boiling Water Reactor (BWR), however, the rise in reactor power could be limited by SCRAM or void reactivity effect. Recent studies have provided accumulated knowledge that even if BT takes place in fuel assemblies, the rise in fuel cladding temperature could be so small that it will not threat to fuel integrity, as long as the BT condition terminates within a short period of time. In addition, appropriate methods have been developed to evaluate the cladding temperature during dryout. This standard provides requirements in the assessment of fuel integrity under AOO in which limited-BT condition is temporarily reached and the propriety of reusing a fuel assembly that has experienced limited-BT condition. The standard has been approved by the Atomic Energy Society of Japan following deliberation by impartial members for two and half years. It is now expected that this standard will provide an effective measure for the rational expansion of fuel design and operational margin. (author)

  17. Mechanical stress analysis for a fuel rod under normal operating conditions

    Pino, Eddy S.; Giovedi, Claudia; Serra, Andre da Silva; Abe, Alfredo Y.

    2013-01-01

    Nuclear reactor fuel elements consist mainly in a system of a nuclear fuel encapsulated by a cladding material subject to high fluxes of energetic neutrons, high operating temperatures, pressure systems, thermal gradients, heat fluxes and with chemical compatibility with the reactor coolant. The design of a nuclear reactor requires, among a set of activities, the evaluation of the structural integrity of the fuel rod submitted to different loads acting on the fuel rod and the specific properties (dimensions and mechanical and thermal properties) of the cladding material and coolant, including thermal and pressure gradients produced inside the rod due to the fuel burnup. In this work were evaluated the structural mechanical stresses of a fuel rod using stainless steel as cladding material and UO 2 with a low degree of enrichment as fuel pellet on a PWR (pressurized water reactor) under normal operating conditions. In this sense, tangential, radial and axial stress on internal and external cladding surfaces considering the orientations of 0 deg, 90 deg and 180 deg were considered. The obtained values were compared with the limit values for stress to the studied material. From the obtained results, it was possible to conclude that, under the expected normal reactor operation conditions, the integrity of the fuel rod can be maintained. (author)

  18. High pressure operation of tubular solid oxide fuel cells and their intergration with gas turbines

    Haynes, C.; Wepfer, W.J. [Georgia Institute of Technology, Atlanta, GA (United States)

    1996-12-31

    Fossil fuels continue to be used at a rate greater than that of their natural formation, and the current byproducts from their use are believed to have a detrimental effect on the environment (e.g. global warming). There is thus a significant impetus to have cleaner, more efficient fuel consumption alternatives. Recent progress has led to renewed vigor in the development of fuel cell technology, which has been shown to be capable of producing high efficiencies with relatively benign exhaust products. The tubular solid oxide fuel cell developed by Westinghouse Electric Corporation has shown significant promise. Modeling efforts have been and are underway to optimize and better understand this fuel cell technology. Thus far, the bulk of modeling efforts has been for operation at atmospheric pressure. There is now interest in developing high-efficiency integrated gas turbine/solid oxide fuel cell systems. Such operation of fuel cells would obviously occur at higher pressures. The fuel cells have been successfully modeled under high pressure operation and further investigated as integrated components of an open loop gas turbine cycle.

  19. The Cost of SOx Limits to Marine Operators; Results from Exploring Marine Fuel Prices

    Orestis Schinas

    2013-06-01

    Full Text Available Marine operators are confronted with the new air emissions regulations that determine the limits of sulfur content in marine fuels. The low-sulfur (LS marine fuels have a higher price, and their fluctuation is almost similar to the fluctuation of high-sulfur (HS fuels. The price difference between HS and LS might also determine the decision of operators for alternative technical means, such as scrubbers, in order to comply with the new limits. This paper aims to provide a thorough statistical analysis of the currently available LS and HS marine fuels time series, as well as to present the analysis of the differential of the HS and LS fuel prices. The paper concludes with suggestions for further research.

  20. Economic assessment of nuclear power plant operation with regard to effective use of nuclear fuel

    Svec, P.; Raninec, S.; Mizov, J.

    1988-01-01

    The essential preconditions are discussed for the better utilization of fuel in nuclear power plants. The MORNAP program which models the operation of the reactor is used for assessing the consequences of various fuel utilization strategies on technical and economic parameters of WWER-440 nuclear power plant operation. Some results of model calculation are given for the third and fourth units of the Jaslovske Bohunice nuclear power plant. The calculations have served for the economic assessment of the transition of part of the nuclear fuel from a three-campaign to a four-campaign cycle. This transition reduces fuel costs by 1.7%. The implementation of this strategy on a larger scale is expected to save 7 to 9% of fuel costs. (Z.M.). 2 tabs., 7 refs

  1. Development of materials for use in solid oxid fuel cells anodes using renewable fuels in direct operation

    Lima, D.B.P.L. de; Florio, D.Z. de; Bezerra, M.E.O.

    2016-01-01

    Fuel cells produce electrical current from the electrochemical combustion of a gas or liquid (H2, CH4, C2H5OH, CH3OH, etc.) inserted into the anode cell. An important class of fuel cells is the SOFC (Solid Oxide Cell Fuel). It has a ceramic electrolyte that transports protons (H +) or O-2 ions and operating at high temperatures (500-1000 °C) and mixed conductive electrodes (ionic and electronic) ceramics or cermets. This work aims to develop anodes for fuel cells of solid oxide (SOFC) in order to direct operations with renewable fuels and strategic for the country (such as bioethanol and biogas). In this context, it becomes important to study in relation to the ceramic materials, especially those that must be used in high temperatures. Some types of double perovskites such as Sr2MgMoO6 (or simply SMMO) have been used as anodes in SOFC. In this study were synthesized by the polymeric precursor method, analyzed and characterized different ceramic samples of families SMMO, doped with Nb, this is: Sr2 (MgMo)1-xNbxO6 with 0 ≤ x ≤ 0.2. The materials produced were characterized by various techniques such as, thermal analysis, X-ray diffraction and scanning electron microscopy, and electrical properties determined by dc and ac measurements in a wide range of temperature, frequency and partial pressure of oxygen. The results of this work will contribute to a better understanding of advanced ceramic properties with mixed driving (electronic and ionic) and contribute to the advancement of SOFC technology operating directly with renewable fuels. (author)

  2. Performance Analysis of Air Breathing Proton Exchange Membrane Fuel Cell Stack (PEMFCS) At Different Operating Condition

    Sunil, V.; Venkata siva, G.; Yoganjaneyulu, G.; Ravikumar, V. V.

    2017-08-01

    The answer for an emission free power source in future is in the form of fuel cells which combine hydrogen and oxygen producing electricity and a harmless by product-water. A proton exchange membrane (PEM) fuel cell is ideal for automotive applications. A single cell cannot supply the essential power for any application. Hence PEM fuel cell stacks are used. The effect of different operating parameters namely: type of convection, type of draught, hydrogen flow rate, hydrogen inlet pressure, ambient temperature and humidity, hydrogen humidity, cell orientation on the performance of air breathing PEM fuel cell stack was analyzed using a computerized fuel cell test station. Then, the fuel cell stack was subjected to different load conditions. It was found that the stack performs very poorly at full capacity (runs only for 30 min. but runs for 3 hours at 50% capacity). Hence, a detailed study was undertaken to maximize the duration of the stack’s performance at peak load.

  3. Commissioning and Operational Experience in Power Reactor Fuel Reprocessing Plant

    Pradhan, S., E-mail: spradhan@barctara.gov.in [Tarapur Based Reprocessing Plant, Bhabha Atomic Research Centre, Tarapur (India)

    2014-10-15

    After completing design, construction, commissioning, operation and maintenance experience of the reprocessing plants at Tarapur, Mumbai and Kalpakkam a new reprocessing plant is commissioned and put into operation at BARC, Tarapur since 2011. Subsequent to construction clearance, commissioning of the plant is taken in many steps with simultaneous review by design and safety committees. In spite of vast experience, all the staff was retrained in various aspects of process and utility operations and in operation of innovative changes incorporated in the design. Operating personnel are licensed through an elaborate procedure consisting of various check lists followed by personnel interview. Commissioning systems were divided in sub-systems. Sub-systems were commissioned independently and later integrated testing was carried out. For commissioning, extreme operating conditions were identified in consultation with designers and detailed commissioning procedures were made accordingly. Commissioning was done in different conditions to ensure safety, smooth operation and maintainability. Few modifications were carried out based on commissioning experience. Technical specifications for operation of the plant are made in consultation with designers and reviewed by safety committees. Operation of the plant was carried out after successful commissioning trials with Deep Depleted Uranium (DDU). Emergency operating procedures for each design basis accident were made. Performance of various systems, subsystems are quite satisfactory and the plant has given very good capacity factor. (author)

  4. Design and operation of equipment used to develop remote coating capability for HTGR fuel particles

    Suchomel, R.R.; Stinton, D.P.; Preston, M.K.; Heck, J.L.; Bolfing, B.J.; Lackey, W.J.

    1978-12-01

    Refabrication of HTGR fuels is a manufacturing process that consists of preparation of fuel kernels, application of multiple layers of pyrolytic carbon and silicon carbide, preparation of fuel rods, and assembly of fuel rods into fuel elements. All the equipment for refabrication of 233 U-containing fuel must be designed for completely remote operation and maintenance in hot-cell facilities. Equipment to remotely coated HTGR fuel particles has been designed and operated. Although not all of the equipment development needed for a fully remote coating system has been completed, significant progress has been made. The most important component of the coating furnace is the gas distributor, which must be simple, reliable, and easily maintainable. Techniques for loading and unloading the coater and handling microspheres have been developed. An engineering-scale system, currently in operation, is being used to verify the workability of these concepts. Coating crucible handling components are used to remove the crucible from the furnace, remove coated particles, and exchange the crucible, if necessary. After the batch of particles has been unloaded, it is transferred, weighed, and sampled. The components used in these processes have been tested to ensure that no particle breakage or holdup occurs. Tests of the particle handling system have been very encouraging because no major problems have been encountered. Instrumentation that controls the equipment performed very smoothly and reliably and can be operated remotely

  5. In-ground operation of Geothermic Fuel Cells for unconventional oil and gas recovery

    Sullivan, Neal; Anyenya, Gladys; Haun, Buddy; Daubenspeck, Mark; Bonadies, Joseph; Kerr, Rick; Fischer, Bernhard; Wright, Adam; Jones, Gerald; Li, Robert; Wall, Mark; Forbes, Alan; Savage, Marshall

    2016-01-01

    This paper presents operating and performance characteristics of a nine-stack solid-oxide fuel cell combined-heat-and-power system. Integrated with a natural-gas fuel processor, air compressor, reactant-gas preheater, and diagnostics and control equipment, the system is designed for use in unconventional oil-and-gas processing. Termed a ;Geothermic Fuel Cell; (GFC), the heat liberated by the fuel cell during electricity generation is harnessed to process oil shale into high-quality crude oil and natural gas. The 1.5-kWe SOFC stacks are packaged within three-stack GFC modules. Three GFC modules are mechanically and electrically coupled to a reactant-gas preheater and installed within the earth. During operation, significant heat is conducted from the Geothermic Fuel Cell to the surrounding geology. The complete system was continuously operated on hydrogen and natural-gas fuels for ∼600 h. A quasi-steady operating point was established to favor heat generation (29.1 kWth) over electricity production (4.4 kWe). Thermodynamic analysis reveals a combined-heat-and-power efficiency of 55% at this condition. Heat flux to the geology averaged 3.2 kW m-1 across the 9-m length of the Geothermic Fuel Cell-preheater assembly. System performance is reviewed; some suggestions for improvement are proposed.

  6. Experimental investigation and modeling of an aircraft Otto engine operating with gasoline and heavier fuels

    Saldivar Olague, Jose

    A Continental "O-200" aircraft Otto-cycle engine has been modified to burn diesel fuel. Algebraic models of the different processes of the cycle were developed from basic principles applied to a real engine, and utilized in an algorithm for the simulation of engine performance. The simulation provides a means to investigate the performance of the modified version of the Continental engine for a wide range of operating parameters. The main goals of this study are to increase the range of a particular aircraft by reducing the specific fuel consumption of the engine, and to show that such an engine can burn heavier fuels (such as diesel, kerosene, and jet fuel) instead of gasoline. Such heavier fuels are much less flammable during handling operations making them safer than aviation gasoline and very attractive for use in flight operations from naval vessels. The cycle uses an electric spark to ignite the heavier fuel at low to moderate compression ratios, The stratified charge combustion process is utilized in a pre-chamber where the spray injection of the fuel occurs at a moderate pressure of 1200 psi (8.3 MPa). One advantage of fuel injection into the combustion chamber instead of into the intake port, is that the air-to-fuel ratio can be widely varied---in contrast to the narrower limits of the premixed combustion case used in gasoline engines---in order to obtain very lean combustion. Another benefit is that higher compression ratios can be attained in the modified cycle with heavier fuels. The combination of injection into the chamber for lean combustion, and higher compression ratios allow to limit the peak pressure in the cylinder, and to avoid engine damage. Such high-compression ratios are characteristic of Diesel engines and lead to increase in thermal efficiency without pre-ignition problems. In this experimental investigation, operations with diesel fuel have shown that considerable improvements in the fuel efficiency are possible. The results of

  7. Modeling and operation optimization of a proton exchange membrane fuel cell system for maximum efficiency

    Han, In-Su; Park, Sang-Kyun; Chung, Chang-Bock

    2016-01-01

    Highlights: • A proton exchange membrane fuel cell system is operationally optimized. • A constrained optimization problem is formulated to maximize fuel cell efficiency. • Empirical and semi-empirical models for most system components are developed. • Sensitivity analysis is performed to elucidate the effects of major operating variables. • The optimization results are verified by comparison with actual operation data. - Abstract: This paper presents an operation optimization method and demonstrates its application to a proton exchange membrane fuel cell system. A constrained optimization problem was formulated to maximize the efficiency of a fuel cell system by incorporating practical models derived from actual operations of the system. Empirical and semi-empirical models for most of the system components were developed based on artificial neural networks and semi-empirical equations. Prior to system optimizations, the developed models were validated by comparing simulation results with the measured ones. Moreover, sensitivity analyses were performed to elucidate the effects of major operating variables on the system efficiency under practical operating constraints. Then, the optimal operating conditions were sought at various system power loads. The optimization results revealed that the efficiency gaps between the worst and best operation conditions of the system could reach 1.2–5.5% depending on the power output range. To verify the optimization results, the optimal operating conditions were applied to the fuel cell system, and the measured results were compared with the expected optimal values. The discrepancies between the measured and expected values were found to be trivial, indicating that the proposed operation optimization method was quite successful for a substantial increase in the efficiency of the fuel cell system.

  8. Design considerations, operating and maintenance experience with wet storage of Ontario Hydro's used fuel

    Frost, C.R.

    1989-01-01

    The characteristics of Ontario Hydro's fuel and at-reactor used fuel storage water pools (or used fuel bays) are described. There are two types of bay, known respectively as primary bays and auxiliary bays, used for at- reactor used fuel storage. Used fuel is discharged remotely from Ontario Hydro's reactors to the primary bays for initial storage and cooling. The auxiliary bays are used to receive and store fuel after its initial cooling in the primary bay, and provide additional storage capacity as needed. With on- power fueling of reactors, each reactor of greater than 500 MW(e) net discharges an average of 10 or more used fuel bundles to bay storage every full power day. The logistics of handling such large quantities of used fuel bundles (corresponding to about 300 te/year of uranium for a 4 unit station) present a challenge to designers and operators. The major considerations in used fuel bay design, including site- specific requirements, reliability and quality assurance, are discussed

  9. Transporting spent fuel and reactor waste in Sweden experience from 5 years of operation

    Dybeck, P.; Gustafsson, B.

    1990-01-01

    This paper reports that since the Final Repository for Reactor Waste, SFR, was taken into operation in 1988, the SKB sea transportation system is operating at full capacity by transporting spent fuel and now also reactor waste from the 12 Swedish reactors to CLAB and SFR. Transports from the National Research Center, Studsvik to the repository has recently also been integrated in the system. CLAB, the central intermediate storage for spent fuel, has been in operation since 1985. The SKB Sea Transportation System consists today of the purpose built ship M/s Sigyn, 10 transport casks for spent fuel, 2 casks for spent core components, 27 IP-2 shielded steel containers for reactor waste and 5 terminal vehicles. During an average year about 250 tonnes of spent fuel and 3 -- 4000 m 3 of reactor waste are transported to CLAB and SFR respectively, corresponding to around 30 sea voyages

  10. Effectiveness of paper-structured catalyst for the operation of biodiesel-fueled solid oxide fuel cell

    Quang-Tuyen, Tran; Kaida, Taku; Sakamoto, Mio; Sasaki, Kazunari; Shiratori, Yusuke

    2015-06-01

    Mg/Al-hydrotalcite (HDT)-dispersed paper-structured catalyst (PSC) was prepared by a simple paper-making process. The PSC exhibited excellent catalytic activity for the steam reforming of model biodiesel fuel (BDF), pure oleic acid methyl ester (oleic-FAME, C19H36O2) which is a mono-unsaturated component of practical BDFs. The PSC exhibited fuel conversion comparable to a pelletized catalyst material, here, conventional Ni-zirconia cermet anode for solid oxide fuel cell (SOFC) with less than one-hundredth Ni weight. Performance of electrolyte-supported cell connected with the PSC was evaluated in the feed of oleic-FAME, and stable operation was achieved. After 60 h test, coking was not observed in both SOFC anode and PSC.

  11. Operating Point Optimization of a Hydrogen Fueled Hybrid Solid Oxide Fuel Cell-Steam Turbine (SOFC-ST Plant

    Juanjo Ugartemendia

    2013-09-01

    Full Text Available This paper presents a hydrogen powered hybrid solid oxide fuel cell-steam turbine (SOFC-ST system and studies its optimal operating conditions. This type of installation can be very appropriate to complement the intermittent generation of renewable energies, such as wind generation. A dynamic model of an alternative hybrid SOFC-ST configuration that is especially suited to work with hydrogen is developed. The proposed system recuperates the waste heat of the high temperature fuel cell, to feed a bottoming cycle (BC based on a steam turbine (ST. In order to optimize the behavior and performance of the system, a two-level control structure is proposed. Two controllers have been implemented for the stack temperature and fuel utilization factor. An upper supervisor generates optimal set-points in order to reach a maximal hydrogen efficiency. The simulation results obtained show that the proposed system allows one to reach high efficiencies at rated power levels.

  12. Tests to determine the release of short-lived fission products from UO2 fuel operating at linear powers of 45 and 60 kW/m

    Hastings, I.J.; Hunt, C.E.L.; Lipsett, J.J.; MacDonald, R.D.

    1985-09-01

    Experiments have been carried out using a 'sweep gas' technique to determine the behaviour of short-lived fission products within operating, intact UO 2 fuel elements. The Zircaloy-4-clad elements were 600 mm long and contained fuel of density 10.65 - 10.71 Mg/m 3 . A He-2% H 2 carrier gas swept gaseous or volatile fission products out of the operating fuel element past a gamma spectrometer for measurement. We outline our loop model and give full details of calculational procedures. In tests at linear powers of 45 (FIO-122) and 60 kW/m (FIO-124) to a maximum burnup of 80 MW.h/kg U, the species measured directly at the spectrometer during normal operation were generally the short-lived xenons and kryptons. Iodines were not observed during normal operation. The behaviour of I-133 and I-135 was deduced from the decay of Xe-133 and Xe-135 during reactor shutdowns. Plots of R/B (released/born) against λ (decay constant) or effective λ for all isotopes observed at 45 and 60 kW/m show that a line of slope -0.5, corresponding with diffusion kinetics, is a good fit to the measured xenon and krypton data. The inferred release of iodine fits the same line. From this we can extrapolate to an R/B for I-131 of about 5 x 10 -4 at 45 kW/m, and 3 x 10 -3 at 60 kW/m. Both tests were terminated by defects. Under defect conditions, R/B dependence on λ was about 0.6. I-131 release under defect conditions was 5 Ci and 60 mCi for FIO-122 and FI0-124, respectively. 22 refs

  13. Studies on dual fuel operation of rubber seed oil and its bio-diesel with hydrogen as the inducted fuel

    Edwin Geo, V.; Nagalingam, B. [Department of Mechanical Engineering, KCG College of Technology, Chennai, Tamil Nadu 600097 (India); Nagarajan, G. [Department of Mechanical Engineering, IC Engineering Division, Anna University, Chennai, Tamil Nadu 600025 (India)

    2008-11-15

    The main problems with the use of neat vegetable oils in diesel engines are higher smoke levels and lower thermal efficiency as compared to diesel. The problem can be tackled by inducting a gaseous fuel in the intake manifold along with air. In this investigation, hydrogen is used as the inducted fuel and rubber seed oil (RSO), rubber seed oil methyl ester (RSOME) and diesel are used as main fuels in a dual fuel engine. A single cylinder diesel engine with rated output of 4.4 kW at 1500 rpm was converted to operate in the dual fuel mode. Dual fuel operation of varying hydrogen quantity with RSO and RSOME results in higher brake thermal efficiency and significant reduction in smoke levels at high outputs. The maximum brake thermal efficiency is 28.12%, 29.26% and 31.62% with RSO, RSOME and diesel at hydrogen energy share of 8.39%, 8.73% and 10.1%, respectively. Smoke is reduced from 5.5 to 3.5 BSU with RSOME and for RSO it is from 6.1 to 3.8 BSU at the maximum efficiency point. The peak pressure and maximum rate of pressure rise increase with hydrogen induction. Heat release rate indicates an increase in the combustion rate with hydrogen induction. On the whole it is concluded that hydrogen can be inducted along with air in order to reduce smoke levels and improve thermal efficiency of RSO and its bio-diesel fuelled diesel engines. (author)

  14. Heat and fuel coupled operation of a high temperature polymer electrolyte fuel cell with a heat exchanger methanol steam reformer

    Schuller, G.; Vázquez, F. Vidal; Waiblinger, W.; Auvinen, S.; Ribeirinha, P.

    2017-04-01

    In this work a methanol steam reforming (MSR) reactor has been operated thermally coupled to a high temperature polymer electrolyte fuel cell stack (HT-PEMFC) utilizing its waste heat. The operating temperature of the coupled system was 180 °C which is significantly lower than the conventional operating temperature of the MSR process which is around 250 °C. A newly designed heat exchanger reformer has been developed by VTT (Technical Research Center of Finland LTD) and was equipped with commercially available CuO/ZnO/Al2O3 (BASF RP-60) catalyst. The liquid cooled, 165 cm2, 12-cell stack used for the measurements was supplied by Serenergy A/S. The off-heat from the electrochemical fuel cell reaction was transferred to the reforming reactor using triethylene glycol (TEG) as heat transfer fluid. The system was operated up to 0.4 A cm-2 generating an electrical power output of 427 Wel. A total stack waste heat utilization of 86.4% was achieved. It has been shown that it is possible to transfer sufficient heat from the fuel cell stack to the liquid circuit in order to provide the needed amount for vaporizing and reforming of the methanol-water-mixture. Furthermore a set of recommendations is given for future system design considerations.

  15. A comparison of low-pressure and supercharged operation of polymer electrolyte membrane fuel cell systems for aircraft applications

    Werner, C.; Preiß, G.; Gores, F.; Griebenow, M.; Heitmann, S.

    2016-08-01

    Multifunctional fuel cell systems are competitive solutions aboard future generations of civil aircraft concerning energy consumption, environmental issues, and safety reasons. The present study compares low-pressure and supercharged operation of polymer electrolyte membrane fuel cells with respect to performance and efficiency criteria. This is motivated by the challenge of pressure-dependent fuel cell operation aboard aircraft with cabin pressure varying with operating altitude. Experimental investigations of low-pressure fuel cell operation use model-based design of experiments and are complemented by numerical investigations concerning supercharged fuel cell operation. It is demonstrated that a low-pressure operation is feasible with the fuel cell device under test, but that its range of stable operation changes between both operating modes. Including an external compressor, it can be shown that the power demand for supercharging the fuel cell is about the same as the loss in power output of the fuel cell due to low-pressure operation. Furthermore, the supercharged fuel cell operation appears to be more sensitive with respect to variations in the considered independent operating parameters load requirement, cathode stoichiometric ratio, and cooling temperature. The results indicate that a pressure-dependent self-humidification control might be able to exploit the potential of low-pressure fuel cell operation for aircraft applications to the best advantage.

  16. Indigenous development of system integration for proton exchange membrane fuel cell operation

    Hussain, S.; Arshad, M.; Anjum, A.R.

    2011-01-01

    System integration was developed for fuel cell to control various parameters including voltage, current, power, temperature, pressure of gas (H/sub 2/), humidification, etc. The compact software has also been developed for monitoring different parameters of fuel cell system. System integrated was installed on fuel cell stack to manipulate these parameters. The compact software has been linked with the integrated system for visual monitoring of different parameters of fuel cell system during operation on PC. The installation of software and integrated system on fuel cell stack is the key achievement for the safe operation of fuel cell stack and for the provision of requisite power to any electric device for optimum performance. The compact software was developed for micro controller in KIEL. Control card and driver card are controlled by software-driven micro controller. A communication protocol was designed and developed. PC software has been developed to control and watch the values of all parameters of fuel cell such as voltage, current, power, temperature, pressure of hydrogen, pressure of oxygen, operational times and performance of the system on computer screen. (author)

  17. Simulation of facility operations and materials accounting for a combined reprocessing/MOX fuel fabrication facility

    Coulter, C.A.; Whiteson, R.; Zardecki, A.

    1991-01-01

    We are developing a computer model of facility operations and nuclear materials accounting for a facility that reprocesses spent fuel and fabricates mixed oxide (MOX) fuel rods and assemblies from the recovered uranium and plutonium. The model will be used to determine the effectiveness of various materials measurement strategies for the facility and, ultimately, of other facility safeguards functions as well. This portion of the facility consists of a spent fuel storage pond, fuel shear, dissolver, clarifier, three solvent-extraction stages with uranium-plutonium separation after the first stage, and product concentrators. In this facility area mixed oxide is formed into pellets, the pellets are loaded into fuel rods, and the fuel rods are fabricated into fuel assemblies. These two facility sections are connected by a MOX conversion line in which the uranium and plutonium solutions from reprocessing are converted to mixed oxide. The model of the intermediate MOX conversion line used in the model is based on a design provided by Mike Ehinger of Oak Ridge National Laboratory (private communication). An initial version of the simulation model has been developed for the entire MOX conversion and fuel fabrication sections of the reprocessing/MOX fuel fabrication facility, and this model has been used to obtain inventory difference variance estimates for those sections of the facility. A significant fraction of the data files for the fuel reprocessing section have been developed, but these data files are not yet complete enough to permit simulation of reprocessing operations in the facility. Accordingly, the discussion in the following sections is restricted to the MOX conversion and fuel fabrication lines. 3 tabs

  18. Spent nuclear fuel project cold vacuum drying facility operations manual

    IRWIN, J.J.

    1999-01-01

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998) and, the HNF-SD-SNF-DRD-002, 1997, Cold Vacuum Drying Facility Design Requirements, Rev. 3a. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence, and has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  19. Comparison of defects in ProTaper hand-operated and engine-driven instruments after clinical use.

    Cheung, G S P; Bian, Z; Shen, Y; Peng, B; Darvell, B W

    2007-03-01

    To compare the type of defects and mode of material failure of engine-driven and hand-operated ProTaper instruments after clinical use. A total of 401 hand-operated and 325 engine-driven ProTaper instruments were discarded from an endodontic clinic over 17 months. Those that had fractured were examined for plastic deformation in lateral view and remounted for fractographical examination in scanning electron microscope. The mode of fracture was classified as 'fatigue' or 'shear' failure. The lengths of fractured segments in both instruments were recorded. Any distortion in hand instrument was noted. Data were analysed using chi-square, Fisher's exact or Student's t-test, where appropriate. Approximately 14% of all discarded hand-operated instruments and 14% of engine-driven instruments were fractured. About 62% of hand instruments failed because of shear fracture, compared with approximately 66% of engine-driven instruments as a result of fatigue (P hand instruments were affected by shear, and either remained intact or was fractured, compared with 5% of engine-driven instruments (P hand versus engine-driven group (P hand instruments were discarded intact but distorted (rarely for engine-driven instruments); all were in the form of unscrewing of the flutes. The location of defects in hand Finishing instruments was significantly closer to the tip than that for Shaping instruments (P ProTaper engine-driven and hand-operated instruments appeared to be different, with shear failure being more prevalent in the latter.

  20. Analysis of defects in ProTaper hand-operated instruments after clinical use.

    Shen, Ya; Bian, Zhuan; Cheung, Gary Shun-pan; Peng, Bin

    2007-03-01

    The purpose of this study was to analyze the type and location of defects observed in ProTaper for Hand Use (PHU) instruments after routine clinical use. We analyzed a total of 401 PHUs discarded from an endodontic clinic over a 17-month period. Those failed instruments were examined on the lateral and fractographic surface by scanning electron microscope. Of the 86 PHUs that showed discernible defects, 28 were intact but partially unwound, and 58 were fractured (36 because of shear and 22 from fatigue failure). The primary characteristic of shear failure was the presence of a skewed dimple and/or tear ridge, a typical pattern developed because of a combination of various loads. Nearly 74% of the instruments with defects exhibited shear damage. About three-quarters of the instrument fractures occurred in the apical one-third of the canal, mostly in molars. The results of this study indicated that most PHU instruments fail because of either shear or fatigue.

  1. High Performance Nano-Crystalline Oxide Fuel Cell Materials. Defects, Structures, Interfaces, Transport, and Electrochemistry

    Barnett, Scott [Northwestern Univ., Evanston, IL (United States); Poeppelmeier, Ken [Northwestern Univ., Evanston, IL (United States); Mason, Tom [Northwestern Univ., Evanston, IL (United States); Marks, Lawrence [Northwestern Univ., Evanston, IL (United States); Voorhees, Peter [Northwestern Univ., Evanston, IL (United States)

    2016-09-07

    This project addresses fundamental materials challenges in solid oxide electrochemical cells, devices that have a broad range of important energy applications. Although nano-scale mixed ionically and electronically conducting (MIEC) materials provide an important opportunity to improve performance and reduce device operating temperature, durability issues threaten to limit their utility and have remained largely unexplored. Our work has focused on both (1) understanding the fundamental processes related to oxygen transport and surface-vapor reactions in nano-scale MIEC materials, and (2) determining and understanding the key factors that control their long-term stability. Furthermore, materials stability has been explored under the “extreme” conditions encountered in many solid oxide cell applications, i.e, very high or very low effective oxygen pressures, and high current density.

  2. HTGR Fuel performance basis

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-05-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents

  3. Operation of real landfill gas fueled solid oxide fuel cell (SOFC) using internal dry reforming

    Langnickel, Hendrik; Hagen, Anke

    2017-01-01

    Biomass is one renewable energy source, which is independent from solar radiation and wind effect. Solid oxide fuel cells (SOFC’s) are able to convert landfill gas derived from landfill directly into electricity and heat with a high efficiency. In the present work a planar 16cm2 SOFC cell...... was necessary to prevent poisoning and thereby to decrease the degradation rate....

  4. Summary of NRC LWR safety research programs on fuel behavior, metallurgy/materials and operational safety

    Bennett, G.L.

    1979-09-01

    The NRC light-water reactor safety-research program is part of the NRC regulatory program for ensuring the safety of nuclear power plants. This paper summarizes the results of NRC-sponsored research into fuel behavior, metallurgy and materials, and operational safety. The fuel behavior research program provides a detailed understanding of the response of nuclear fuel assemblies to postulated off-normal or accident conditions. Fuel behavior research includes studies of basic fuel rod properties, in-reactor tests, computer code development, fission product release and fuel meltdown. The metallurgy and materials research program provides independent confirmation of the safe design of reactor vessels and piping. This program includes studies on fracture mechanics, irradiation embrittlement, stress corrosion, crack growth, and nondestructive examination. The operational safety research provides direct assistance to NRC officials concerned with the operational and operational-safety aspects of nuclear power plants. The topics currently being addressed include qualification testing evaluation, fire protection, human factors, and noise diagnostics

  5. Conversion of diesel engines to dual fuel (propane/diesel) operations

    Pepper, S W; DeMaere, D A

    1984-02-01

    A device to convert a diesel engine to dual fuel (propane/diesel) operation was developed and evaluated. Preliminary experimentation has indicated that as much as 30% of the diesel fuel consumed in diesel engines could be displaced with propane, accompanied by an improvement in fuel efficiency, engine maintenance and an overall reduction in emission levels. Dual fuel operations in both transportation and stationary applications would then project a saving of ca 90,000 barrels of diesel fuel per day by the year 1990. A turbo-charged 250 hp diesel engine was directly coupled to a dynamometer under laboratory conditions, and operated at speeds between 500 and 2500 rpm and at various torque levels. At each rpm/torque point the engine first operated on diesel fuel alone, and then increasing quantities of propane were induced into the air intake until detonation occured. Results indicate that the proportion of propane that can be safely induced into a diesel engine varies considerably with rpm and torque so that a sophisticated metering system would be required to maximize diesel oil displacement by propane. Conversion is not cost effective at 1983 price levels.

  6. Implementation and operational experience of an integrated fuel information service at the BNFL THORP facility

    Robson, D.N.; Ramsden, P.N.

    1995-01-01

    BNFL's THORP Plant, which started active operations early in 1994, has contracts to reprocess 7000t(U) of fuel belonging to 33 customers in 9 countries in the UK, Europe and Japan during its first 10 years of operation. Contracts are in place or being negotiated, and further business sought after, with the expectation of extending THORP's operations well beyond the initial 10 years. An integrated data management service, for the fuel storage areas of BNFL's THORP Division, is being implemented to replace several, independent, systems. This Fuel Information Service (FIS) will bring the Nuclear Materials Accountancy and Safeguards Records together with the Operating Records into one database from which all Safeguards Reports will be made. BNFL's contractual and commercial data and technical data on the stored fuel, required to support the reprocessing business, will also be brought into the common database. FIS is the first stage in a project to integrate the Materials Management systems throughout the THORP nuclear recycling business including irradiated fuel receipt and storage, reprocessing and storage of products, mixed oxide fuel manufacture and the conditioning and storage of wastes

  7. Operational considerations in specifying legal weight vehicles for the highway transport of spent nuclear fuel

    Hill, C.V.; Rutenkroger, E.O.; Ratledge, J.E.

    1990-01-01

    This paper presents the results of a research project in which tractor manufacturers and carrier companies were interviewed to gather information on operational concerns in specifying a tractor to haul legal weight spent fuel casks. The interaction between driver fatigue,safety, and equipment is discussed, as are innovative operating strategies that could save weight

  8. 14 CFR 121.643 - Fuel supply: Nonturbine and turbo-propeller-powered airplanes: Supplemental operations.

    2010-01-01

    ... operating nontransport category airplanes type certificated after December 31, 1964, to fly for 30 minutes...-powered airplanes: Supplemental operations. 121.643 Section 121.643 Aeronautics and Space FEDERAL AVIATION... Flight Release Rules § 121.643 Fuel supply: Nonturbine and turbo-propeller-powered airplanes...

  9. Operating experience feedback report: Assessment of spent fuel cooling. Volume 12

    Ibarra, J.G.; Jones, W.R.; Lanik, G.F.; Ornstein, H.L.; Pullani, S.V.

    1997-02-01

    This report documents the results of an independent assessment by a team from the Office of Analysis and Evaluation of Operational Data of spent-fuel-pool (SFP) cooling in operating nuclear power plants. The team assessed the likelihood and consequences of an extended loss of SFP cooling and suggested corrective actions, based on their findings

  10. Processing biogas to obtain motor fuel - Operational experience

    Seifert, M.

    2008-01-01

    This article takes a look at how raw biogas can be processed in order to remove carbon dioxide and corrosive substances and thus bring it up to natural gas quality. The ecological advantages of using biogas as a fuel are discussed and the situation in Europe and Switzerland is examined. Also, feeding biogas into the normal natural gas mains is discussed and the technologies necessary for the cleaning and preparation of the biogas are described. These include absorption and adsorption processes as well as membrane systems that are used to remove excessive carbon dioxide. The costs involved are discussed on the basis of experience gained in Sweden and Switzerland. Finally, the environmental aspects of methane losses are discussed.

  11. Hydrogen Fuel as Ecological Contribution to Operation of the Existing Coal-Fired Thermal Power Plants

    Cosic, D.

    2009-01-01

    The analysis is carried out of the application of a new hydrogen based alternative fuel as ecological contribution of the coal thermal power plants operation. Given the fact that coal thermal power plants are seen as the largest producers, not only of CO 2 , but of all others harmful gases, the idea is initiated to use the new alternative fuel as an additive to the coal which would result in much better performance of the coal power plants from an ecological point of view. It is possible to use such a fuel in relation of 10-30% of former coal use. The positive influence of such an application is much bigger than relative used quantity. This lecture has a goal to incite potential investors to create conditions for industrial testing of the new fuel. It will be very interesting to animate investors for large-scale production of the new fuel, too.(author).

  12. Operational experience gained with the failed fuel rod detection system in nuclear power plants

    Boehm, H.H.; Forch, H.

    1985-01-01

    Brown Boveri Reaktor GmbH together with Krautkramer Company developed such a FAILED FUEL ROD DETECTION SYSTEM (FFRDS) which allows to located defective fuel rods without dismantling the fuel assembly or pulling of individual rods. Since 1979 the FFRDS is employed successfully in various nuclear power plants in Europe, USA, Japan, and Korea. The short inspection time and the high reliability of the method make the FFRDS a true competitor to the sipping method. In this paper the authors discuss the method and the design of the system, the equipment set-up, its features and the experience gained so far. The system has been performed and automated to such an extent that within a short installation period series of fuel assemblies can be tested with relatively short intervals of time (5 minutes for BWR and 7 minutes for PWR fuel assemblies per side). The ability of the system for deployment under various conditions and the experience gained during the past six years have made this system universally applicable and highly sensitive to the requirements of NDT during outages and for transport of FAs to intermediate storage facilities. Comparison of FFRDS to conventional sipping has indicated in several instances that the FFRDS is superior to the latter technique

  13. Effect of reactor chemistry and operating variables on fuel cladding corrosion in PWRs

    Park, Moon Ghu; Lee, Sang Hee

    1997-01-01

    As the nuclear industry extends the fuel cycle length, waterside corrosion of zircaloy cladding has become a limiting factor in PWR fuel design. Many plant chemistry factors such as, higher lithium/boron concentration in the primary coolant can influence the corrosion behavior of zircaloy cladding. The chemistry effect can be amplified in higher duty fuel, particularlywhen surface boiling occurs. Local boiling can result in increased crud deposition on fuel cladding which may induce axial power offset anomalies (AOA), recently reported in several PWR units. In this study, the effect of reactor chemistry and operating variables on Zircaloy cladding corrosion is investigated and simulation studies are performed to evaluate the optimal primary chemistry condition for extended cycle operation. (author). 8 refs., 3 tabs., 16 figs

  14. In-core fuel management for the course on operational physics of power reactors

    Levine, S.H.

    1982-01-01

    The heart of a nuclear power station is the reactor core producing power from the fissioning of uranium or plutonium fuel. Expertise in many different technical fields is required to provide fuel for continuous economical operation of a nuclear power plant. In general, these various technical disciplines can be dichotomized into ''Out-of-core'' and ''In-core'' fuel management. In-core fuel management is concerned, as the name implies, with the reactor core itself. It entails calculating the core reactivity, power distribution, and isotopic inventory for the first and subsequent cores of a nuclear power plant to maintain adequate safety margins and operating lifetime for each core. In addition, the selection of reloading schemes is made to minimize energy costs

  15. A microfluidic-structured flow field for passive direct methanol fuel cells operating with highly concentrated fuels

    Wu, Q X; Zhao, T S; Chen, R; Yang, W W

    2010-01-01

    Conventional direct methanol fuel cells (DMFCs) have to operate with excessively diluted methanol solutions to limit methanol crossover and its detrimental consequences. Operation with such diluted methanol solutions not only results in a significant penalty in the specific energy of the power pack, limiting the runtime of this type of fuel cell, but also lowers the cell performance and operating stability. In this paper, a microfluidic-structured anode flow field for passive DMFCs with neither liquid pumps nor gas compressors/blowers is developed. This flow field consists of plural micro flow passages. Taking advantage of the liquid methanol and gas CO 2 two-phase counter flow, the unique fluidic structure enables the formation of a liquid–gas meniscus in each flow passage. The evaporation from the small meniscus in each flow passage can lead to an extremely large interfacial mass-transfer resistance, creating a bottleneck of methanol delivery to the anode CL. The fuel cell tests show that the innovative flow field allows passive DMFCs to achieve good cell performance with a methanol concentration as high as 18.0 M, increasing the specific energy of the DMFC system by about five times compared with conventional designs.

  16. Impact of alternative fuels on the operational and environmental performance of a small turbofan engine

    Gaspar, R.M.P.; Sousa, J.M.M.

    2016-01-01

    Highlights: • A wide range of alternative fuels is studied in a small two-spool turbofan engine. • Impact of fuel properties on flame temperature and droplet evaporation considered. • Performance and pollutant emissions at different operating conditions are analyzed. • Use of alternative fuels generally leads to an improved performance of the engine. • Mostly reductions in soot emissions, but also cuts in NO_x and CO, are obtained. - Abstract: A wide range of alternative jet fuels is studied in this work for use in a small two-spool turbofan engine. These embrace the five production pathways currently approved by the American Society for Testing and Materials. Both neat products and blends (within certified limits) have been considered. The present analysis is based on a 0-D thermodynamic modeling of the aero-engine for off-design and transient simulations. In addition, the selected approach incorporates fuel effects on combustion and the impact of fuel properties on the flame temperature, as well as on the droplet evaporation rate. Predicted performance and pollutant emission outputs for the alternative fuels are presented at different operating conditions, namely: take-off, top of climb, cruise, low power and ground idle. The results are discussed and comprehensively compared with data available in the literature. It was concluded that the combustion of alternative fuels generally leads to enhancements in engine performance with respect to the use of conventional kerosene. Reductions in pollutant emissions occur mostly in soot, but also in nitrogen oxides and carbon monoxide, depending on the fuel and operating conditions. In contrast, increased emissions of unburned hydrocarbons are generally observed. Concerns about the aero-engine dynamic response are raised only in very few cases, involving the use of neat products.

  17. EFFECT SIGNIFICANCE ASSESSMENT OF THE THERMODYNAMICAL FACTORS ON THE SOLID OXIDE FUEL CELL OPERATION

    V. A. Sednin

    2015-01-01

    Full Text Available Technologies of direct conversion of the fuel energy into electrical power are an upcoming trend in power economy. Over the last decades a number of countries have created industrial prototypes of power plants on fuel elements (cells, while fuel cells themselves became a commercial product on the world energy market. High electrical efficiency of the fuel cells allows predictting their further spread as part of hybrid installations jointly with gas and steam turbines which specifically enables achieving the electrical efficiency greater than 70 %. Nevertheless, investigations in the area of increasing efficiency and reliability of the fuel cells continue. Inter alia, research into the effects of oxidizing reaction thermodynamic parameters, fuel composition and oxidation reaction products on effectiveness of the solid oxide fuel cells (SOFC is of specific scientific interest. The article presents a concise analysis of the fuel type effects on the SOFC efficiency. Based on the open publications experimental data and the data of numerical model studies, the authors adduce results of the statistical analysis of the SOFC thermodynamic parameters effect on the effectiveness of its functioning as well as of the reciprocative factors of these parameters and gas composition at the inlet and at the outlet of the cell. The presented diagrams reflect dimension of the indicated parameters on the SOFC operation effectiveness. The significance levels of the above listed factors are ascertained. Statistical analysis of the effects of the SOFC functionning process thermodynamical, consumption and concentration parameters demonstrates quintessential influence of the reciprocative factors (temperature – flow-rate and pressure – flow-rate and the nitrogen N2 and oxygen O2 concentrations on the operation efficiency in the researched range of its functioning. These are the parameters to be considered on a first-priority basis while developing mathematical models

  18. Nuclear reactor internals construction and failed fuel rod detection system

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A system is provided for determining during operation of a nuclear reactor having fluid pressure operated control rod mechanisms the exact location of a fuel assembly with a defective fuel rod. The construction of the reactor internals is simplified in a manner to facilitate the testing for defective fuel rods and the reduce the cost of producing the upper internals of the reactor. 13 claims, 10 drawing figures

  19. International co-operation in the supply of nuclear fuel cycle services

    Allday, C.

    1977-01-01

    The paper draws on British Nuclear Fuels' (BNFL) wide experience of international collaboration in nuclear fuel process activities to examine the pros and cons of international agreements. Initially, the factors that influence the need to co-operate, the extent of possible co-operation and the alternative types of agreement, are reviewed. Next, the benefits, problems and risks associated with each function, such as management, financial R and D, marketing and operations that could be covered within the scope of an international agreement, are examined in detail. The paper continues by calling upon specific experience obtained by BNFL in the co-operation with other organizations over several years in both major and much smaller agreements, illustrating the rationale behind the co-operation, the resolution of 'teething' troubles and the present status of these organizations. In conclusion, the paper comments upon the effectiveness of collaboration agreements and identifies several requirements for international co-operation to succeed. (author)

  20. Status report - expert knowledge of operators in fuel reprocessing plants, enrichment plants and fuel fabrication plants

    Preuss, W.; Kramer, J.; Wildberg, D.

    1987-01-01

    The necessary qualifications of the responsible personnel and the knowledge required by personnel otherwise employed in nuclear plants are among the requirements for licensing laid down in paragraph 7 of the German Atomic Energy Act. The formal regulations for nuclear power plants are not directly applicable to plants in the fuel cycle because of the differences in the technical processes and the plant and work organisation. The aim of the project was therefore to establish a possible need for regulations for the nuclear plants with respect to the qualification of the personnel, and to determine a starting point for the definition of the required qualifications. An extensive investigation was carried out in the Federal Republic of Germany into: the formal requirements for training; the plant and personnel organisation structures; the tasks carried out by the responsible and otherwise employed personnel; and the state of training. For this purpose plant owners and managers were interviewed and the literature and plant specific documentation (e.g. plant rules) were reviewed. On the basis of literature research, foreign practices were determined and used to make comparative evaluations. The status report is divided into three separate parts for the reprocessing, the uranium enrichment, and the manufacture of the fuel elements. On the basis of the situation for reprocessing plants (particularly that of the WAK) and fuel element manufacturing plants, the development of a common (not uniform) regulation for all the examined plants in the fuel cycle was recommended. The report gives concrete suggestions for the content of the regulations. (orig.) [de

  1. Palliative effects of H2 on SOFCs operating with carbon containing fuels

    Reeping, Kyle W.; Bohn, Jessie M.; Walker, Robert A.

    2017-12-01

    Chlorine can accelerate degradation of solid oxide fuel cell (SOFC) Ni-based anodes operating on carbon containing fuels through several different mechanisms. However, supplementing the fuel with a small percentage of excess molecular hydrogen effectively masks the degradation to the catalytic activity of the Ni and carbon fuel cracking reaction reactions. Experiments described in this work explore the chemistry behind the "palliative" effect of hydrogen on SOFCs operating with chlorine-contaminated, carbon-containing fuels using a suite of independent, complementary techniques. Operando Raman spectroscopy is used to monitor carbon accumulation and, by inference, Ni catalytic activity while electrochemical techniques including electrochemical impedance spectroscopy and voltammetry are used to monitor overall cell performance. Briefly, hydrogen not only completely hides degradation observed with chlorine-contaminated carbon-containing fuels, but also actively removes adsorbed chlorine from the surface of the Ni, allowing for the methane cracking reaction to continue, albeit at a slower rate. When hydrogen is removed from the fuel stream the cell fails immediately due to chlorine occupation of methane/biogas reaction sites.

  2. Operation and maintenance of spent fuel storage and transportation casks/containers

    2007-01-01

    Member States have a growing need for casks for spent fuel storage and transportation. A variety of casks has been developed and is in use at an increasing number of sites. This has resulted in an accumulation of experience that will provide valuable information for other projects in spent fuel management. This publication provides a comprehensive review of information on the cask operation and maintenance associated with spent fuel storage. It draws upon generic knowledge from industrial experience and applications and is intended to serve as a basis for better planning and implementation in future projects

  3. High Burnup Fuel: Implications and Operational Experience. Proceedings of a Technical Meeting

    2016-08-01

    This publication reports on the outcome of a technical meeting on high burnup fuel experience and economics, held in Buenos Aires, Argentina in 2013. The purpose of the meeting was to revisit and update the current operational experience and economic conditions associated with high burnup fuel. International experts with significant experience in experimental programmes on high burnup fuel discussed and evaluated physical limitations at pellet, cladding and structural component levels, with a wide focus including fabrication, core behaviour, transport and intermediate storage for most types of commercial nuclear power plants

  4. The effect of test configuration on the true operating conditions of PEM fuel cells. Paper no. IGEC-1-124

    Simpson, T.; Li, X.

    2005-01-01

    The operating conditions of a single PEM fuel cell can be significantly affected by the configuration in which the fuel cell test is setup. This study investigates the effect on the gas dewpoint temperature of not insulating the inlet fittings to a PEM fuel cell and the effect of non-optimal stack control thermocouple placement on fuel cell stack operating temperature. Both of these setup configurations can significantly affect fuel cell membrane humidification conditions, especially in a single fuel cell as demonstrated through the sample test conditions presented in this paper. (author)

  5. Radiological safety experience in nuclear fuel cycle operations at Bhabha Atomic Research Center, Trombay, Mumbai, India

    Pushparaja; Gopalakrishnan, R.K.; Subramaniam, G.

    2000-01-01

    Activities at Bhabha Atomic Research Centre (BARC), Mumbai, cover nuclear fuel cycle operations based on natural uranium as the fuel. The facilities include: plant for purification and production of nuclear grade uranium metal, fuel fabrication, research reactor operation, fuel reprocessing and radioactive waste management in each stage. Comprehensive radiation protection programmes for assessment and monitoring of radiological impact of these operations, both in occupational and public environment, have been operating in BARC since beginning. These programmes, based on the 1990 ICRP Recommendations as prescribed by national regulatory body, the Atomic Energy Regulatory Board (AERB), are being successfully implemented by the Health, Safety and Environment Group, BARC. Radiation Hazards Control Units attached to the nuclear fuel cycle facilities provide radiation safety surveillance to the various operations. The radiation monitoring programme consists of measurement and control of external exposures by thermoluminescent dosimeters (TLDs), hand-held and installed instruments, and internal exposures by bioassay and direct whole body counting using shadow shield counter for beta gamma emitters and phoswich detector based system for plutonium. In addition, an environmental monitoring programme is in place to assess public exposures resulting from the operation of these facilities. The programme involves analysis of various matrices in the environment such as bay water, salt, fish, sediment and computation of resulting public exposures. Based on the operating experience in these plants, improved educating and training programmes for plant operators, have been designed. This, together with the application of new technologies have brought down individual as well as average doses of occupational workers. The environmental releases remain a small fraction of the authorised limits. The operating health physics experience in some of these facilities is discussed in this paper

  6. Dual fuel operation of used transformer oil with acetylene in a DI diesel engine

    Behera, Pritinika; Murugan, S.; Nagarajan, G.

    2014-01-01

    Highlights: • Utilisation of Used transformer oil (UTO) as a fuel in a diesel engine. • UTO with acetylene in a diesel engine, on a dual fuel mode technique. • Analysis of combustion characteristics of the diesel engine. • Analysis of performance and emission characteristics of the diesel engine. - Abstract: Used transformer oil (UTO) is a waste oil obtained from power transformers and welding transformers. It possesses considerable heating value and properties similar to diesel fuel. A preliminary investigation on the utilization of the UTO in a single cylinder, four stroke small powered direct injection (DI) diesel engine revealed that at an optimum injection timing of 20°CA the engine exhibited lower nitric oxide (NO) and higher smoke emissions, compared to that of diesel operation. In order to improve the performance and reduce the smoke emission, a dual fuel operation was attempted in the present investigation. Acetylene was inducted as a primary fuel at four different flow rates viz 132 g/h, 198 g/h, 264 g/h and 330 g/h along with the air, to study the combustion, performance and emission behavior of a four-stroke, 4.4 kW diesel engine, while the UTO was injected as pilot fuel with the optimized injection timing. The experimental results were compared with diesel-acetylene dual fuel operation in the same engine. Acetylene aspiration reduced the ignition delay and maximum cylinder pressure by about 3°CA, and 25% respectively at full load in comparison with the sole UTO operation. Higher thermal efficiency and lower exhaust gas were also observed at full load. Smoke was reduced by about 13.7%, in comparison with the UTO operation at full load

  7. FREC-4A: a computer program to predict fuel rod performance under normal reactor operation

    Harayama, Yasuo; Izumi, Fumio

    1981-10-01

    The program FREC-4A (Fuel Reliability Evaluation Code-version 4A) is used for predicting fuel rod performance in normal reactor operation. The performance is calculated in accordance with the irradiation history of fuel rods. Emphasis is placed on the prediction of the axial elongation of claddings induced by pellet-cladding mechanical interaction, including the influence of initially preloaded springs inserted in fuel rod lower plenums. In the FREC-4A, an fuel rod is divided into axial segments. In each segment, it is assumed that the temperature, stress and strain are axi-symmetrical, and the axial strain in constant in fuel pellets and in a cladding, though the values in the pellets and in the cladding are different. The calculation of the contact load and the clearance along the length of a fuel rod and the stress and strain in each segment is explained. The method adopted in the FREC-4A is simple, and suitable to predict the deformation of fuel rods over their full length. This report is described on the outline of the program, the method of solving the stiffness equations, the calculation models, the input data such as irradiation history, output distribution, material properties and pores, the printing-out of input data and calculated results. (Kako, I.)

  8. Operational Readiness Review Final Report for K Basin Fuel Transfer System

    DAVIES, T.H.

    2002-01-01

    An Operational Readiness Review (ORR) was conducted by the U.S. Department of Energy (DOE), Richland Operations Office (RL) to verify that an adequate state of readiness had been achieved for startup of the K Basin Fuel Transfer System (FTS). The DOE ORR was conducted during the period November 6-18, 2002. The DOE ORR team concluded that the K Basin Fuel Transfer System is ready to start operations, subject to completion and verification of identified pre-start findings. The ORR was conducted in accordance with the Spent Nuclear Fuel (SNF) K Basin Fuel Transfer System (FTS) Operational Readiness Review (ORR) Plan of Action and the Operational Readiness Review Implementation Plan for K Basin Fuel Transfer System. Review activities consisted of staff interviews, procedure and document reviews, and observations of normal facility operations, operational upset conditions, and an emergency drill. The DOE ORR Team also reviewed and assessed the adequacy of the contractor ORR3 and the RL line management review. The team concurred with the findings and observations identified in these two reports. The DOE ORR for the FTS evaluated the contractor under single-shift operations. Of concern to the ORR Team was that SNF Project management intended to change from a single-shift FTS operation to a two-shift operation shortly after the completion of the DOE ORR. The ORR team did not assess two-shift FTS operations and the ability of the contractor to conduct a smooth transition from shift to shift. However, the DOE ORR team did observe an operational upset drill that was conducted during day shift and carried over into swing shift; during this drill, swing shift was staffed with fewer personnel as would be expected for two-shift operations. The facility was able to adequately respond to the event with the reduced level of staff. The ORR Team was also able to observe a Shift Manager turnover meeting when one shift manager had to be relieved during the middle of the day. The ORR

  9. Operation strategy for solid oxide fuel cell systems for small-scale stationary applications

    Liso, Vincenzo; Nielsen, Mads Pagh; Kær, Søren Knudsen

    2009-01-01

    Solid oxide fuel cell micro cogeneration systems have the potential to reduce domestic energy consumption by providing both heat and power on site without transmission losses. The high grade heat produced during the operation of the power causes high thermal transients during startup/shutdown pha......Solid oxide fuel cell micro cogeneration systems have the potential to reduce domestic energy consumption by providing both heat and power on site without transmission losses. The high grade heat produced during the operation of the power causes high thermal transients during startup....../shutdown phases and degrades the fuel cells. To counteract the degradation, the system has not to be stressed with rapid load variation during the operation. The analysis will consider an average profile for heat and power demand of a family house. Finally data analysis and power system limitations will be used...

  10. Current density distribution mapping in PEM fuel cells as an instrument for operational measurements

    Geske, M.; Heuer, M.; Heideck, G.; Styczynski, Z. A. [Otto-von-Guericke University Magdeburg, Chair Electric Power Networks and Renewable Energy Sources, Magdeburg (Germany)

    2010-07-01

    A newly developed measurement system for current density distribution mapping has enabled a new approach for operational measurements in proton exchange membrane fuel cells (PEMFC). Taking into account previously constructed measurement systems, a method based on a multi layer printed circuit board was chosen for the development of the new system. This type of system consists of a sensor, a special electronic device and the control and visualization PC. For the acquisition of the current density distribution values, a sensor device was designed and installed within a multilayer printed circuit board with integrated shunt resistors. Varying shunt values can be taken into consideration with a newly developed and evaluated calibration method. The sensor device was integrated in a PEM fuel cell stack to prove the functionality of the whole measurement system. A software application was implemented to visualize and save the measurement values. Its functionality was verified by operational measurements within a PEMFC system. Measurement accuracy and possible negative reactions of the sensor device during PEMFC operation are discussed in detail in this paper. The developed system enables operational measurements for different operating phases of PEM fuel cells. Additionally, this can be seen as a basis for new opportunities of optimization for fuel cell design and operation modes. (author)

  11. Current Density Distribution Mapping in PEM Fuel Cells as An Instrument for Operational Measurements

    Martin Geske

    2010-04-01

    Full Text Available A newly developed measurement system for current density distribution mapping has enabled a new approach for operational measurements in proton exchange membrane fuel cells (PEMFC. Taking into account previously constructed measurement systems, a method based on a multi layer printed circuit board was chosen for the development of the new system. This type of system consists of a sensor, a special electronic device and the control and visualization PC. For the acquisition of the current density distribution values, a sensor device was designed and installed within a multilayer printed circuit board with integrated shunt resistors. Varying shunt values can be taken into consideration with a newly developed and evaluated calibration method. The sensor device was integrated in a PEM fuel cell stack to prove the functionality of the whole measurement system. A software application was implemented to visualize and save the measurement values. Its functionality was verified by operational measurements within a PEMFC system. Measurement accuracy and possible negative reactions of the sensor device during PEMFC operation are discussed in detail in this paper. The developed system enables operational measurements for different operating phases of PEM fuel cells. Additionally, this can be seen as a basis for new opportunities of optimization for fuel cell design and operation modes.

  12. Conversion of a gasoline internal combustion engine to operate on hydrogen fuel

    Bates, M.; Dincer, I.

    2009-01-01

    This study deals with the conversion of a gasoline spark ignition internal combustion engine to operate on hydrogen fuel while producing similar power, economy and reliability as gasoline. The conversion engine will have the fuel system redesigned and ignition and fuel timing changed. Engine construction material is of great importance due to the low ignition energy of hydrogen, making aluminum a desirable material in the intake manifold and combustion chamber. The engine selected to convert is a 3400 SFI dual over head cam General Motors engine. Hydrogen reacts with metals causing hydrogen embrittlement which leads to failure due to cracking. There are standards published by American Society of Mechanical Engineers (ASME) to avoid such a problem. Tuning of the hydrogen engine proved to be challenging due to the basic tuning tools of a gasoline engine such as a wide band oxygen sensor that could not measure the 34:1 fuel air mixture needed for the hydrogen engine. Once the conversion was complete the engine was tested on a chassis dynamometer to compare the hydrogen horsepower and torque produced to that of a gasoline engine. Results showed that the engine is not operating correctly. The engine is not getting the proper amount of fuel needed for complete combustion when operated in a loaded state over 3000 rpm. The problem was found to be the use of the stock injector driver that could not deliver enough power for the proper operation of the larger CM4980 injectors. (author)

  13. Performance and emissions of a dual-fuel pilot diesel ignition engine operating on various premixed fuels

    Yousefi, Amin; Birouk, Madjid; Lawler, Benjamin; Gharehghani, Ayatallah

    2015-01-01

    and the higher content of oxygen with increased methanol, soot and CO emissions are significantly reduced for richer premixed methanol mixtures. According to these findings, a dual-fuel engine operating on methanol and diesel performs better at rich conditions, whereas the performance of hydrogen and diesel is superior to that of natural gas/diesel and methanol/diesel mixtures at lean conditions.

  14. Studying some regimes of the WWER-440 type reactor failed fuel element operation

    Aksenov, N.A.; Samsonov, B.V.; Sulaberidze, V.Sh.; Frej, A.K.

    1981-01-01

    The results of investigating the serviceability of experimental fuel elements close by type to that of the WWER-440 type reactor in the cans of which untightness in the form of small opening are made. The tests are carried out in the SM-2 reactor high temperature water loop at the temperature of 473 K, pressure of (1-2)x10 4 kPa, coolant flow rate of 3.7-5.5 m 3 /h. The analysis of the obtained results shows that the character of changes in the fission product (FP) activity in the circuit in a considerable extent is determined bt the thermal-optical conditions of the fuel element operation. If water in the gap between fuel and can does not boil, activity changes smoothly and bursts caused by increased FP release are observed only under transient conditions of reactor operation. In the presence of water boiling in the gap the FP release has of impulse character with the frequency determined besides the untightness dimension by free volume inside the fuel element can (with its increase the pulsation frequency increases). FP release from fuel is connected with their direct escape from an open surface. When water in the gap the FP release from the fuel element occurs practically immediately. Without boiling the FP delay in the gap is determined by their diffusion in a layer of water. The conclusion is drawn that the FP release from failed fuel elements may be reduced by eliminating the water boiling in the gap between the fuel and the can by means of the fuel element power or coolant temperature decrease

  15. Reliability of fast reactor mixed-oxide fuel during operational transients

    Boltax, A.; Neimark, L.A.; Tsai, Hanchung; Katsuragawa, M.; Shikakura, S.

    1991-07-01

    Results are presented from the cooperative DOE and PNC Phase 1 and 2 operational transient testing programs conducted in the EBR-2 reactor. The program includes second (D9 and PNC 316 cladding) and third (FSM, AST and ODS cladding) generation mixed-oxide fuel pins. The irradiation tests include duty cycle operation and extended overpower tests. the results demonstrate the capability of second generation fuel pins to survive a wide range of duty cycle and extended overpower events. 15 refs., 9 figs., 4 tabs

  16. Implementation of ICRP recommendation in nuclear fuel cycle operations: challenges and achievements

    Gupta, V.K.

    1999-01-01

    The operating experience with regard to occupational exposure and environmental releases in Nuclear Fuel Cycle Facilities are described. The achievements of Nuclear Fuel Cycle Facilities in adhering to the revised radiological protection standards are highlighted, with particular reference to Nuclear Power Plants (NPPs). The downward trend of occupational and public doses due to nuclear power plant operation is emphasised. Some of the important radiologically significant jobs executed at NPPs are listed. With the vast experiences in the field of radiological protection, vis-a-vis stringent regulatory requirements, and design modifications envisaged in future facilities the radiological impact, both in the occupational and public domain is bound to be minimum. (author)

  17. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    Sanders, T.L.; Lake, W.H.

    1989-01-01

    It is possible to develop an optimal strategy for implementing burnup credit in spent fuel transport casks. For transport, the relative risk is rapidly reduced if additional pre-transport controls such as a cavity dryness verifications are conducted prior to transport. Some other operational and design features that could be incorporated into a burnup credit cask strategy are listed. These examples represent many of the system features and alternatives already available for use in developing a broadly based criticality safety strategy for implementing burnup credit in the design and operation of spent fuel transport casks. 4 refs., 1 tab

  18. One year of operation of the Belgonucleaire (Dessel) plutonium fuel fabrication plant

    Leblanc, J.M.

    1975-01-01

    Based on experience with plutonium since 1958, Belgonucleaire has successively launched a pilot plant and then a fuel fabrication plant for mixed uranium and plutonium oxides in 1968 and 1973 respectively. After describing briefly the plants and the most important stages in the planning, construction and operation of the Dessel plant, the present document describes the principal problems which were met during the course of operation of the plant and their direct incidence on the capacity and quality of the production of fuel elements

  19. The differential radiological impact of plutonium recycle in the light-water reactor fuel cycle: effluent discharges during normal operation

    Bouville, A.; Guetat, P.; Jones, J.A.; Kelly, G.N.; Legrand, J.; White, I.F.

    1980-01-01

    The radiological impact of a light-water reactor fuel cycle utilizing enriched uranium fuel may be altered by the recycle of plutonium. Differences in impact may arise during various operations in the fuel cycle: those which arise from effluents discharged during normal operation of the various installations comprising the fuel cycle are evaluated in this study. The differential radiological impact on the population of the European Communities (EC) of effluents discharged during the recycling of 10 tonnes of fissile plutonium metal is evaluated. The contributions from each stage of the fuel cycle, i.e. fuel fabrication, reactor operation and fuel reprocessing and conversion, are identified. Separate consideration is given to airborne and liquid effluents and account is taken of a wide range of environmental conditions, representative of the EC, in estimating the radiological impact. The recycle of plutonium is estimated to result in a reduction in the radiological impact from effluents of about 30% of that when using enriched uranium fuel

  20. Development of four-year fuel cycle based on the advanced fuel assembly with uranium-gadolinium fuel and its implementation to the operating WWER-440 units

    Lunin, G.; Novikov, A.; Pavlov, V.; Pavlovichev, P.; Filimonov, P.

    2000-01-01

    Over the past few years in Russia the investigations aimed at the increase of the reliability, safety and efficiency of operation of the WWER-1000 reactors as well as of its competitiveness in the world market were carried out. In the frame of these investigations the four-year fuel cycle, based on advanced fuel assemblies with zirconium alloy spacer grids and guide tubes and with fuel pellet having a reduced diameter of the central hole (1,5 mm), has been developed. For the compensation of a part of excess reactivity, Gd 2 O 3 integrated burnable absorbers are used. CPS absorbing rods contain a combine absorber (B 4 C + Dy 2 O 3 *TiO 2 ). A part of depleted fuel is located on the core periphery. The algorithms controlling the reactor power and power distribution have been updated. For checking of the solutions adopted and for verification of code package developed at the RRC 'Kurchatov Institute' the wide-scale experimental operation of advanced FA and its individual components is carried out. (Authors)

  1. REVA Advanced Fuel Design and Codes and Methods - Increasing Reliability, Operating Margin and Efficiency in Operation

    Frichet, A.; Mollard, P.; Gentet, G.; Lippert, H. J.; Curva-Tivig, F.; Cole, S.; Garner, N.

    2014-07-01

    Since three decades, AREVA has been incrementally implementing upgrades in the BWR and PWR Fuel design and codes and methods leading to an ever greater fuel efficiency and easier licensing. For PWRs, AREVA is implementing upgraded versions of its HTP{sup T}M and AFA 3G technologies called HTP{sup T}M-I and AFA3G-I. These fuel assemblies feature improved robustness and dimensional stability through the ultimate optimization of their hold down system, the use of Q12, the AREVA advanced quaternary alloy for guide tube, the increase in their wall thickness and the stiffening of the spacer to guide tube connection. But an even bigger step forward has been achieved a s AREVA has successfully developed and introduces to the market the GAIA product which maintains the resistance to grid to rod fretting (GTRF) of the HTP{sup T}M product while providing addition al thermal-hydraulic margin and high resistance to Fuel Assembly bow. (Author)

  2. Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

    Stefanova, S; Vitkova, M; Simeonova, V; Passage, G; Manolova, M [Institute for Nuclear Research and Nuclear Energy, Sofia (Bulgaria); Haralampieva, Z [National Electric Company Ltd., Kozloduy (Bulgaria); Scheglov, A; Proselkov, V [Institute of Nuclear Reactors, RSC Kurchatov Inst., Moscow (Russian Federation)

    1997-08-01

    The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab.

  3. Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

    Stefanova, S.; Vitkova, M.; Simeonova, V.; Passage, G.; Manolova, M.; Haralampieva, Z.; Scheglov, A.; Proselkov, V.

    1997-01-01

    The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab

  4. CANDU reactor experience: fuel performance

    Truant, P.T.; Hastings, I.J.

    1985-07-01

    Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

  5. Real-world fuel use and gaseous emission rates for flex fuel vehicles operated on E85 versus gasoline.

    Delavarrafiee, Maryam; Frey, H Christopher

    2018-03-01

    Flex fuel vehicles (FFVs) typically operate on gasoline or E85, an 85%/15% volume blend of ethanol and gasoline. Differences in FFV fuel use and tailpipe emission rates are quantified for E85 versus gasoline based on real-world measurements of five FFVs with a portable emissions measurement system (PEMS), supplemented chassis dynamometer data, and estimates from the Motor Vehicle Emission Simulator (MOVES) model. Because of inter-vehicle variability, an individual FFV may have higher nitrogen oxide (NO x ) or carbon monoxide (CO) emission rates on E85 versus gasoline, even though average rates are lower. Based on PEMS data, the comparison of tailpipe emission rates for E85 versus gasoline is sensitive to vehicle-specific power (VSP). For example, although CO emission rates are lower for all VSP modes, they are proportionally lowest at higher VSP. Driving cycles with high power demand are more advantageous with respect to CO emissions, but less advantageous for NO x . Chassis dynamometer data are available for 121 FFVs at 50,000 useful life miles. Based on the dynamometer data, the average difference in tailpipe emissions for E85 versus gasoline is -23% for NO x , -30% for CO, and no significant difference for hydrocarbons (HC). To account for both the fuel cycle and tailpipe emissions from the vehicle, a life cycle inventory was conducted. Although tailpipe NO x emissions are lower for E85 versus gasoline for FFVs and thus benefit areas where the vehicles operate, the life cycle NO x emissions are higher because the NO x emissions generated during fuel production are higher. The fuel production emissions take place typically in rural areas. Although there are not significant differences in the total HC emissions, there are differences in HC speciation. The net effect of lower tailpipe NO x emissions and differences in HC speciation on ozone formation should be further evaluated. Reported comparisons of flex fuel vehicle (FFV) tailpipe emission rates for E85 versus

  6. Optimization of combustion chamber geometry and operating conditions for compression ignition engine fueled with pre-blended gasoline-diesel fuel

    Lee, Seokhwon; Jeon, Joonho; Park, Sungwook

    2016-01-01

    Highlights: • Pre-blended gasoline-diesel fuel was used with direct injection system. • KIVA-CHEMKIN code modeled dual-fuel fuel spray and combustion processes with discrete multi-component model. • The characteristics of Combustion and emission on pre-blended fuel was investigated with various fuel reactivities. • Optimization of combustion chamber shape improved combustion performance of the gasoline-diesel blended fuel engine. - Abstract: In this study, experiments and numerical simulations were used to improve the fuel efficiency of compression ignition engine using a gasoline-diesel blended fuel and an optimization technology. The blended fuel is directly injected into the cylinder with various blending ratios. Combustion and emission characteristics were investigated to explore the effects of gasoline ratio on fuel blend. The present study showed that the advantages of gasoline-diesel blended fuel, high thermal efficiency and low emission, were maximized using the numerical optimization method. The ignition delay and maximum pressure rise rate increased with the proportion of gasoline. As the gasoline fraction increased, the combustion duration and the indicated mean effective pressure decreased. The homogeneity of the fuel-air mixture was improved due to longer ignition delay. Soot emission was significantly reduced up to 90% compared to that of conventional diesel. The nitrogen oxides emissions of the blended fuel increased slightly when the start of injection was retarded toward top dead center. For the numerical study, KIVA-CHEMKIN multi-dimensional CFD code was used to model the combustion and emission characteristics of gasoline-diesel blended fuel. The micro genetic algorithm coupled with the KIVA-CHEMKIN code were used to optimize the combustion chamber shape and operating conditions to improve the combustion performance of the blended fuel engine. The optimized chamber geometry enhanced the fuel efficiency, for a level of nitrogen oxides

  7. Czech interim spent fuel storage facility: operation experience, inspections and future plans

    Fajman, V.; Bartak, L.; Coufal, J.; Brzobohaty, K.; Kuba, S.

    1999-01-01

    The paper describes the situation in the spent fuel management in the Czech Republic. The interim Spent Fuel Storage Facility (ISFSF) at Dukovany, which was commissioned in January 1997 and is using dual transport and storage CASTOR - 440/84 casks, is briefly described. The authors deal with their experience in operating and inspecting the ISFSF Dukovany. The structure of the basic safety document 'Limits and Conditions of Normal Operation' is also mentioned, including the experience of the performance. The inspection activities focused on permanent checking of the leak tightness of the CASTOR 440/84 casks, the maximum cask temperature and inspections monitoring both the neutron and gamma dose rate as well as the surface contamination. The results of the inspections are mentioned in the presentation as well. The operator's experience with re-opening partly loaded and already dried CASTOR-440/84 cask, after its transport from NPP Jaslovske Bohunice to the NPP Dukovany is also described. The paper introduces briefly the concept of future spent fuel storage both from the NPP Dukovany and the NPP Temelin, as prepared by the CEZ. The preparatory work for the Central Interim Spent Nuclear Fuel Storage Facility (CISFSF) in the Czech Republic and the information concerning the planned storage technology for this facility is discussed in the paper as well. The authors describe the site selection process and the preparatory steps concerning new spent fuel facility construction including the Environmental Impact Assessment studies. (author)

  8. Dynamic modeling and evaluation of solid oxide fuel cell - combined heat and power system operating strategies

    Nanaeda, Kimihiro; Mueller, Fabian; Brouwer, Jacob; Samuelsen, Scott

    Operating strategies of solid oxide fuel cell (SOFC) combined heat and power (CHP) systems are developed and evaluated from a utility, and end-user perspective using a fully integrated SOFC-CHP system dynamic model that resolves the physical states, thermal integration and overall efficiency of the system. The model can be modified for any SOFC-CHP system, but the present analysis is applied to a hotel in southern California based on measured electric and heating loads. Analysis indicates that combined heat and power systems can be operated to benefit both the end-users and the utility, providing more efficient electric generation as well as grid ancillary services, namely dispatchable urban power. Design and operating strategies considered in the paper include optimal sizing of the fuel cell, thermal energy storage to dispatch heat, and operating the fuel cell to provide flexible grid power. Analysis results indicate that with a 13.1% average increase in price-of-electricity (POE), the system can provide the grid with a 50% operating range of dispatchable urban power at an overall thermal efficiency of 80%. This grid-support operating mode increases the operational flexibility of the SOFC-CHP system, which may make the technology an important utility asset for accommodating the increased penetration of intermittent renewable power.

  9. Proceedings of the specialist meeting on nuclear fuel and control rods: operating experience, design evolution and safety aspects

    1997-01-01

    Design and management of nuclear fuel has undergone a strong evolution process during past years. The increase of the operating cycle length and of the discharge burnup has led to the use of more advanced fuel designs, as well as to the adoption of fuel efficient operational strategies. The analysis of recent operational experience highlighted a number of issues related to nuclear fuel and control rod events raising concerns about the safety aspects of these new designs and operational strategies, which led to the organisation of this Specialists Meeting on fuel and control rod issues. The meeting was intended to provide a forum for the exchange of information on lessons learned and safety concern related to operating experience with fuel and control rods (degradation, reliability, experience with high burnup fuel, and others). After an opening session 6 papers), this meeting was subdivided into four sessions: Operating experience and safety concern (technical session I - 6 papers), Fuel performance and operational issues (technical session II - 7 papers), Control rod issues (technical session III - 9 papers), Improvement of fuel design (technical session IV.A - 4 papers), Improvement on fuel fabrication and core management (technical session IV.B - 6 papers)

  10. Degradation Mechanism in a Direct Carbon Fuel Cell Operated with Demineralised Brown Coal

    Rady, Adam C.; Giddey, Sarbjit; Kulkarni, Aniruddha; Badwal, Sukhvinder P.S.; Bhattacharya, Sankar

    2014-01-01

    Graphical abstract: - Highlights: • Degradation mechanism studied for demineralised coal in a direct carbon fuel cell. • Diffusion limited processes dominate the electrode polarisation losses in pure N 2 . • Major fuel cell performance loss occurred due to loss of carbon/anode contacts. • The anode retained its phase structure with minor other phases formed in operation. - Abstract: The performance of a demineralised and devolatilised coal from the Morwell mine in the Latrobe Valley, Victoria, has been investigated in a direct carbon fuel cell (DCFC) operated at 850 °C. The focus of the investigation has been on understanding degradation issues as a function of time involving a sequence of electrochemical impedance spectroscopy and voltage-current characteristic. Diffusion limited processes dominate the electrode polarisation losses in pure N 2 atmosphere, however, these decrease substantially in the presence of CO 2 as the anode chamber purge gas, due to in situ generation of fuel species by the reaction of CO 2 with carbon. Post-mortem analysis of anode by SEM and XRD revealed only a minor degradation due to its reduction, particle agglomeration as well as the formation of small quantity of new phases. However, major fuel cell performance degradation (increase of ohmic resistive and electrode polarisation losses) occurred due to loss of carbon/anode contacts and a reduction in the electron-conducting pathways as the fuel was consumed. The investigations revealed that the demineralised coal char can be used as a viable fuel for DCFC, however, further developments on anode materials and fuel feed mechanism would be required to achieve long-term sustained performance

  11. Behaviour of short-lived iodines in operating UO2 fuel elements

    Lipsett, J.J.; Hastings, I.J.; Hunt, C.E.L.

    1984-11-01

    Sweep gas experiments have been done to determine the behaviour of short-lived fission products within operating UO 2 fuel elements at linear powers of 45, 54, and 60 KW/m, and to burnups of 70, 80, and 50 MWh/kgU respectively. Although radioiodine transport was not observed directly during normal operation, equilibrium gap inventories for I-131 were deduced from the shutdown decay behaviour of the fission gases. These inventories were a strong function of fuel power and ranged from 10 GBq (0.27 Ci) to 100 GBq (2.7 Ci) over the range tested. We conclude that the iodine inventory was adsorbed onto the fuel and/or sheath surfaces with a volatile fraction of less than 10 -2 and a charcoal-filter-penetrating fraction of less than 2x10 -4

  12. Vibrational effects of fuel elements detected during KNK II power operation

    Mitzel, F.; Vaeth, W.; Ansari, S.

    1982-08-01

    The reactivity signal of the KNK II reactor shows almost harmonic reactivity oscillations of Δρ≤0.5 cent. Sensitive correlation measurements, made during the regular plant operation with the normal out-of-core plant instrumentation, revealed that they are associated with individual fuel elements. Auxiliary measurements under various operational conditions and theoretical considerations showed that the oscillations are caused by flow-induced mechanical vibrations. Similar characteristics with respect to the frequencies of these oscillations have obviously not yet been observed for fuel element vibrations in other reactors and tests in out-of-core loops. Therefore efforts were made to classify the phenomenon and to identify the excitation mechanism by using only the normal plant instrumentation. It seems to be most likely a flow-induced vibration of whole fuel elements by vortex shedding or jet switching. This model can explain all observations without exception [de

  13. Failed fuel diagnosis during WWER reactor operation using the RTOP-CA code

    Likhanskii, V.; Afanasieva, E.; Sorokin, A.; Evdokimov, I.; Kanukova, V.; Khromov, A.

    2006-01-01

    The mechanistic code RTOP-CA is developed for objectives of failed fuel diagnosis during WWER reactor operation. The RTOP-CA code enables to solve a direct problem: modelling the failed fuel behavior and prediction of primary coolant activity if characteristics of failures in the reactor core are known. Results of verification of the RTOP-CA code are presented. Separate physical models were verified on small-scale in-pile and out-of-pile experiments. Integral verification cases included data obtained at research reactors and at nuclear power plants. The RTOP-CA code is used for development of a neural-network approach to the inverse problem: detection of failure characteristics on the base of data on primary coolant activity during reactor operation. Preliminary results of application of the neural-network approach for evaluation of fuel failure characteristics are presented. (authors)

  14. Improvement of operation efficiency for WWER-440 and WWER-1000 for TRIGON fuel assembly design features

    Silberstein, A [European WWER Fuels GmbH, Lyon (France)

    1994-12-31

    TRIGON 440 and TRIGON 1000 fuel assemblies and their assembly matching counterparts are described. Their role in increasing the efficiency of WWER reactors is stressed. Special attention is paid to their design features as well as calibrated means of predicting behaviour under irradiation from light water reactor core operation. They reduce the fuel cycle cost as a result of the reduced need for natural uranium which have to be enriched and of the smaller number of fuel assemblies which have to be fabricated, stored or reprocessed. The improved control assemblies bring comfort to the plant operator due to intrinsic progress in safety with respect to accidental situation, trouble-free behaviour and long time utilization in the reactor. 14 figs.

  15. The NFI TOKAI SD System - management of the capabilities of operators in fuel fabrication plants

    Fukushima, T.

    2008-01-01

    Since the JCO criticality accident occurred in 1999, even more emphasis has been placed on the management of nuclear safety in Japan. This is particularly true for the education of operators and the observance of operational procedures. Even prior to this accident, Nuclear Fuel Industries, Ltd., NFI, regarded the education and development of skilled operators very seriously and we have developed an education system, called the SD system (Skill Development system), to assure the careful education of the operators and the improvement of their skill in order to prevent human error events. Our education system in the Tokai works, is explained. (author)

  16. Internal Nozzle Flow Simulations of Gasoline-Like Fuels under Diesel Operating Conditions

    Torelli, R.; Som, S.; Pei, Y.; Zhang, Yu; Traver, Michael

    2017-05-15

    Spray formation in internal combustion engines with direct injection is strictly correlated with internal nozzle flow characteristics, which are in turn influenced by fuel physical properties and injector needle motion. This paper pre-sents a series of 3D simulations that model the in-nozzle flow in a 5-hole mini-sac diesel injector. Two gasoline-like naphtha fuels, namely full-range and light naphtha, were tested under operating conditions typical of diesel applica-tions and were compared with n-dodecane, selected from a palette used as diesel surrogates. Validated methodolo-gies from our previous work were employed to account for realistic needle motion. The multi-phase nature of the problem was described by the mixture model assumption with the Volume of Fluid method. Cavitation effects were included by means of the Homogeneous Relaxation Model and turbulence closure was achieved with the Standard k-ε model in an Unsteady Reynolds-Averaged Navier-Stokes formulation. The results revealed that injector perfor-mance and propensity to cavitation are influenced by the fuel properties. Analyses of several physical quantities were carried out to highlight the fuel-to-fuel differences in terms of mass flow rate, discharge coefficients, and fuel vapor volume fraction inside the orifices. A series of parametric investigations was also performed to assess the fuel response to varied fuel injection temperature, injection pressure, and cross-sectional orifice area. For all cases, the strict correlation between cavitation magnitude and saturation pressure was confirmed. Owing to their higher volatil-ity, the two gasoline-like fuels were characterized by higher cavitation across all the simulated conditions. Occur-rence of cavitation was mostly found at the needle seat and at the orifice inlets during the injection event’s transient, when very small gaps exist between the needle and its seat. This behavior tended to disappear at maximum needle lift, where cavitation was

  17. Production and Maintenance Planning for a Deteriorating System with Operation-Dependent Defectives

    Héctor Rivera-Gómez

    2018-01-01

    Full Text Available This paper provides new insights to the area of sustainable manufacturing systems at analyzing the novel paradigm of integrated production logistics, quality, and maintenance design. For this purpose, we investigate the optimal production and repair/major maintenance switching strategy of an unreliable deteriorating manufacturing system. The effects of the deterioration process are mainly observed on the failure intensity and on the quality of the parts produced, where the rate of defectives depends on the production rate. When unplanned failures occur, either a minimal repair or a major maintenance could be conducted. The integration of availability and quality deterioration led us to propose a new stochastic dynamic programming model where optimality conditions are derived through the Hamilton-Jacobi-Bellman equations. The model defined the joint production and repair/major maintenance switching strategies minimizing the total cost over an infinite planning horizon. In the results, the influence of the deterioration process were evident in both the production and maintenances control parameters. A numerical example and an extensive sensitivity analysis were conducted to illustrate the usefulness of the results. Finally, the proposed control policy was compared with alternative strategies based on common assumptions of the literature in order to illustrate its efficiency.

  18. Dry vault for spent fuel depository, basic outsets, operating results and safety of the ''CASCAD'' plant

    Mercier, J.P.; Bardelle, P.

    1994-01-01

    Reprocessing and recycling of fissile materials is the preferred approach to spent fuel management in France. However, a number of spent fuel elements from prototype and experimental nuclear reactors cannot be reprocessed in the existing industrial facilities, either because such facilities are booked to full capacity, or due to technical factors such as non standard nature of fuel or limited series of fuel. The CEA therefore built a facility in which spent fuel can be stored for a few decades (50 years), until favourable conditions prevail for its disposal. The main features of this project consist of a dry depository, which presents a low cost of operation, against a wet on which is more expensive due to the circulation and the continuous controls of the water. Therefore, this is a fair solution because the experimental fuels will present a rather low residual heat power after decay in the nuclear reactor. At this stage, it becomes possible to cool the fuel elements by a fully passive air circulation. This process allows a good efficiency without mechanical equipment and works all the better as the amount of heat to exhaust is great, in the limits of the design. However, we will see that this concept may be extended to a depository of standard spent fuel elements. This facility, known as ''CASCAD'' (shortening for CASemate (=vault) CADarache) started up in 1990, and received its first canister of fuel on May 29th 1990. This paper reviews the basic design data of the facility, outlines the main techniques used for its construction, draws the safety concepts and presents the first results determined by a looking-back over 4 years of working. (author). 1 fig

  19. Low stoichiometry operation of a polymer electrolyte membrane fuel cell employing the interdigitated flow field design

    Berning, Torsten; Odgaard, Madeleine; Kær, Søren Knudsen

    2011-01-01

    Fuel cell operation on dry reactant gases under low stoichiometry conditions employing the interdigitated flow field is investigated using a multi-fluid model. It is assumed that the MEA contains a water uptake layer which facilitates water absorption to the membrane and hence prevents the anode...

  20. General Atomic HTGR fuel reprocessing pilot plant: results of initial sequential equipment operation

    1978-09-01

    In September 1977, the processing of 20 large high-temperature gas-cooled reactor (LHTGR) fuel elements was completed sequentially through the head-end cold pilot plant equipment. This report gives a brief description of the equipment and summarizes the results of the sequential operation of the pilot plant. 32 figures, 15 tables

  1. The operational and logistic experience on transportation of Brazilian spent fuel to USA

    Maiorino, Jose Rubens; Frajndlich, Roberto; Mandlae, Martin; Bensberg, Werner; Renger, August; Grabow, Karsten

    2000-01-01

    A shipment of 127 spent MTR fuel assemblies was made from IEA-R1 Research Reactor located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Sao Paulo, Brazil to Savannah River Site Laboratory in the United States. This paper describes the operational and logistic experience on this transportation made by IPEN staff and the Consortium NCS/GNS. (author)

  2. Balance and behavior of gaseous radionuclides released during initial fast reactor fuel reprocessing operations

    Leudet, A.; Goumondy, J.P.; Charrier, G.

    1985-10-01

    Five pins from the fast reactor Phenix are cut and dissolved in a specially designed cell for the accurate determination of gas released during the operation. Amount and activity of gaseous radionuclides: Kr, Xe, Kr-85, I, I-129, H-3 and C-14 are determined in the fuel pins and also their distribution between shearing and dissolution [fr

  3. Aging and defect characterization of motor-operated valves: progress based on NPAR strategy

    Eissenberg, D.M.

    1984-01-01

    The Nuclear Plant Aging Research (NPAR) program strategy is directed at carrying out comprehensive aging assessments in order to define and resolve issues related to aging (including service wear) of electrical and mechanical components and structures at operating reactor facilities and their possible impact on plant safety. This paper describes work recently completed at Oak Ridge National Laboratory which applied the NPAR strategy to motor-operated valves (MOVs). The objective of the work was primarily to develop an understanding of the operating history and conditions and the failure modes of MOVs in nuclear plant service as a preliminary to identifying and recommending methods for trending aging degradation. A second objective was to demonstrate, using MOVs as an example, that the NPAR strategy can be applied to many electrical and mechanical components of nuclear power plants

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

    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)

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

    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)

  6. Experimental Analysis of a Small Generator set Operating on Dual Fuel Diesel-Ethanol

    Marcel Alex Vailatti

    2017-08-01

    Full Text Available This work aims to analyze the operation of a generator set on single fuel mode with diesel oil, and on dual fuel mode using diesel–ethanol blends. The engine used to realize the experimental analysis was a diesel cycle model, single cylinder, direct injection, air refrigerated and coupled to a three-phase electric generator, whose set capacity was 8.0 kVA. The generated electric energy was dissipated in electrical resistances inside a reservoir with running water. Fuels were blended in different volumetric ratios, using a small portion of vegetable castor oil to promote the homogenization. The percentages of substitutions of diesel oil were by 10% to 50%, increasing by 10% the replacement for each sample. Also, the engine was operated with 100% substitution of diesel oil, i.e., for this condition, the samples were composed of ethanol/castor oil 90/10 (volume/volume, 80/20 and 75/25. The blends of diesel and ethanol did not obtain good performance, mainly in taxes of substitution above 40%, causing combustion failures, operational instability, and increase of fuel consumption, although it has achieved a greatly reduction on opacity percentages. The blends with 100% of substitution of diesel oil obtained good performance except to blend with 90% ethanol, where occurred combustion failures, which caused operational instability. To these conditions, the results achieved are increase of consumption by 17%, decrease of opacity by 79%, decrease of exhaust gas temperature by 3.5% and increase of engine thermal efficiency by 1.3%. At the ethanol – castor oil blends there was a decrease in the percentage of opacity by 96%, decrease of exhaust gas temperature by 17.6%, with a minimum of operational irregularities, although fuel consumption has increased by 52.4% and the engine thermal efficiency has decreased almost 1.7%.

  7. Combustion performance of an aluminum melting furnace operating with liquid fuel

    Nieckele, Angela Ourivio; Naccache, Monica Feijo; Gomes, Marcos Sebastiao de P. [Pontificia Universidade Catolica (PUC-Rio), Rio de Janeiro, RJ (Brazil). Dept. de Engenharia Mecanica], E-mails: nieckele@puc-rio.br, naccache@puc-rio.br, mspgomes@puc-rio.br

    2010-10-15

    The characteristics associated with the delivery of the fuel to be used as the energy source in any industrial combustion equipment are of extreme importance, as for example, in improving the performance of the combustion process and in the preservation of the equipment. A clean and efficient combustion may be achieved by carefully selecting the fuel and oxidant, as well as the operational conditions of the delivery system for both. In the present work, numerical simulations were carried out using the commercial code FLUENT for analyzing some of the relevant operational conditions inside an aluminum reverb furnace employing liquid fuel and air as the oxidant. Different fuel droplets sizes as well as inlet droplet stream configurations were examined. These characteristics, associated with the burner geometry and the fuel dispersion and delivery system may affect the flame shape, and consequently the temperature and the heat flux distribution within the furnace. Among the results obtained in the simulations, it was shown the possible damages to the equipment, which may occur as a result of the combustion process, if the flame is too long or too intense and concentrated. (author)

  8. IMPROVEMENT OF PERFORMANCE OF DUAL FUEL ENGINE OPERATED AT PART LOAD

    N. Kapilan

    2010-12-01

    Full Text Available Rising petroleum prices, an increasing threat to the environment from exhaust emissions, global warming and the threat of supply instabilities has led to the choice of inedible Mahua oil (MO as one of the main alternative fuels to diesel oil in India. In the present work, MO was converted into biodiesel by transesterification using methanol and sodium hydroxide. The cost of Mahua oil biodiesel (MOB is higher than diesel. Hence liquefied petroleum gas (LPG, which is one of the cheapest gaseous fuels available in India, was fumigated along with the air to reduce the operating cost and to reduce emissions. The dual fuel engine resulted in lower efficiency and higher emissions at part load. Hence in the present work, the injection time was varied and the performance of the dual fuel engine was studied. From the engine tests, it is observed that an advanced injection time results in higher efficiency and lower emissions. Hence, advancing the injection timing is one of the ways of increasing the efficiency of LPG+MOB dual fuel engine operated at part load.

  9. Spent fuel storage cask testing and operational experience at the Idaho National Engineering Laboratory

    Eslinger, L.E.; Schmitt, R.C.

    1989-01-01

    Spent-fuel storage cask research, development, and demonstration activities are being performed for the U.S. Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) as a part of the storage cask testing program. The cask testing program at federal sites and other locations supports the Nuclear Waste Policy Act (NWPA) and DOE objectives for cooperative demonstrations with the cask vendors and utilities for development of at-reactor dry cask storage capabilities for spent nuclear fuel assemblies. One research and development program for the storage cask performance testing of metal storage cask was initiated through a cooperative agreement between Virginia Power and DOE in 1984. The performance testing was conducted for the DOE and the Electric Power Research Institute by the Pacific Northwest laboratory, operated for DOE by Battelle Memorial Institute, and the Idaho National Engineering Laboratory (INEL), operated for DOE by EG ampersand G Idaho, Inc. In 1988 a cooperative agreement was entered into by DOE with Pacific Sierra Nuclear Associates (PSN) for performance testing of the PSN concrete Ventilated Storage Cask. Another closely related activity involving INEL is a transportable storage cask project identified as the Nuclear Fuel Services Spent-Fuel Shipping/Storage Cask Demonstration Project. The purpose of this project is to demonstrate the feasibility of packing, transporting, and storing commercial spent fuel in dual-purpose transport/storage casks

  10. Operation of molten carbonate fuel cells with different biogas sources: A challenging approach for field trials

    Trogisch, S.; Hoffmann, J.; Daza Bertrand, L.

    In the past years research in the molten carbonate fuel cells (MCFC) area has been focusing its efforts on the utilisation of natural gas as fuel (S. Geitmann, Wasserstoff- & Brennstoffzellen-Projekte, 2002, ISBN 3-8311-3280-1). In order to increase the advantages of this technology, an international consortium has worked on the utilisation of biogas as fuel in MCFC. During the 4 years lasting RTD project EFFECTIVE two different gas upgrading systems have been developed and constructed together with two mobile MCFC test beds which were operated at different locations for approximately 2.000-5.000 h in each run with biogas from different origins and quality. The large variety of test locations has enabled to gather a large database for assessing the effect of the different biogas qualities on the complete system consisting of the upgrading and the fuel cell systems. The findings are challenging. This article also aims at giving an overview of the advantages of using biogas as fuel for fuel cells.

  11. Evaluation of spectral shift controlled reactors operating on the uranium fuel cycle. Final report

    Matzie, R.A.; Sider, F.M.

    1979-08-01

    The performance of the spectral shift controlled reactor (SSCR) operating on uranium fuel cycles was evaluated and compared with the conventional pressurized water reactor (PWR). In order to analyze the SSCR, the PSR design methodology was extended to include systems moderated by mixtures of light water and heavy water and these methods were validated by comparison with experimental results. Once the design methods had been formulated, the resouce requirements and power costs were determined for the uranium-fueled SSCR. The ore requirements of the UO 2 once-through fuel cycle and the UO 2 fuel cycle with self-generated recycle (SGR) of plutonium were found to be 10% and 19% less than those of similarly fueled PWRs, respectively. A fuel cycle optimization study was performed for the UO 2 once-through SSCR and the SGR SSCR. By individually altering lattice parameters, discharge exposure or number of in-core batches, savings of less than 8% in resource requirements and less than 1% in power costs were obtained

  12. Prospect and problems of self-supporting nuclear fuel cycle operation in Korea beyond the year 2000

    Kim, Chang Hyo

    1988-01-01

    Korea ranks tenth in the free world in the installed capacity of nuclear power plants. Since she lacks in indigenous uranium resources and built-in domestic fuel cycle technology, however, she had to rely heavily upon overseas uranium markets and fuel cycle services. Despite the fact that localization of fuel fabrication technology is nearly achieved, such issues as the timely and stable procurement of front-end fuel cycle services including uranium ore, interim storage and ultimate fuel cycle services including uranium ore, interim storage and ultimate disposal of ever-accumulating spent fuels, disposal of radioactive waste, and decommissioning of the retiring nuclear power plants has yet fuel cycle technology beyond the year 2000. The purpose of this report is to identify these issues of fuel cycle operation and to suggest mobilizing nation's research potential for establishing self-reliant fuel cycle technology beyond the year 2000

  13. Comparison of problems and experience of core operation with distorted fuel element assemblies in VVER-1000 and PWR reactors

    Afanas'ev, A.

    1999-01-01

    The main reactors leading to distortion of fuel element assemblies during reactor operation were studied. A series of actions which compensate this effect was proposed. Criteria of operation limitation in VVER-1000 and PWR reactors are described

  14. Analysed a defective of the machine for a cap-tube nuclear fuel element ME-27 from its electricity point of view

    Achmad Suntoro

    2009-01-01

    It has been analysed a defective of the machine for a cap-tube nuclear fuel element ME-27 from its electricity point of view. The machine uses magnetic force resistance welding technique. A short circuit was happened within the machine because the nut for tightening high voltage cable for welding transformer was broken so that the cable touched the machine body and produced the short circuit. This condition made both the primary circuit breaker in the building down and produced high voltage pulse induction to the electronic circuit within the machine so that one of its electronic components was defective. This case becomes warnings on how important of tightening a nut according to its strength specification (using wrench torque) and the necessity of voltage transient limitation circuit to be installed. Both of the warnings are necessary for any equipment consuming high electric current oriented such as the ME-27 machine. (author)

  15. Performance and endurance of a PEMFC operated with synthetic reformate fuel feed

    Sishtla, C; Koncar, G; Platon, R [Institute of Gas Technology, Des Plaines, IL (United States); Gamburzev, S; Appleby, A J [Texas Engineering Experimental Station, Texas A and M Univ. System, College Station, TX (United States). Center for Electrochemical Systems and Hydrogen Research; Velev, O A [AeroVironment, Inc., Monrovia, CA (United States)

    1998-03-15

    Widespread implementation of polymer electrolyte membrane fuel cell (PEMFC) powerplants for stationary and vehicular applications will be dependent in the near future on using readily available hydrocarbon fuels as the source of the hydrogen fuel. Methane and propane are ideal fuels for stationary applications, while methanol, gasoline, and diesel fuel are better suited for vehicular applications. Various means of fuel processing are possible to produce a gaseous fuel containing H{sub 2}, CO{sub 2} and CO. CO is a known electrocatalyst poison and must be reduced to low (10`s) ppm levels and CO{sub 2} is said to cause additional polarization effects. Even with no CO in the feed gas a H{sub 2}/CO{sub 2}/H{sub 2}O gas mixture will form some CO. Therefore, as a first step of developing a PEMFC that can operate for thousands of hours using a reformed fuel, we used an anode gas feed of 80% H{sub 2} and 20% CO{sub 2} to simulate the reforming of CH{sub 4}. To investigate the effect of reformate on cell performance and endurance, a single cell with an active area of 58 cm{sup 2} was assembled with a membrane electrode assembly (MEA) furnished by Texas A and M University using IGT`s internally manifolded heat exchange (IMHEX{sup TM}) design configuration. The MEA consisted of a Nafion 112 membrane with anode and cathode Pt catalyst loadings of 0.26 and 1.46 mg/cm{sup 2}, respectively. The cell was set to operate on a synthetic reformate - air at 60 C and 1 atm and demonstrated over 5000 h of endurance with a decay rate of less than 1%/1000 h of operation. The cell also underwent four successful thermal cycles with no appreciable loss in performance. The stable performance is attributed to a combination of the IGT IMHEX plate design with its inherent uniform gas flow distribution across the entire active area and MEA quality. The effects of temperature, gas composition, fuel utilization (stoics) and thermal cycle on cell performance are described. (orig.)

  16. Example of an application of an assessment procedure for defects in plant operating in the creep range

    Ainsworth, R.; Coleman, M.C.

    1986-10-01

    A worked example is presented to illustrate the application of a CEGB procedure for the assessment of defects in plant that operates within the creep range. The example chosen is a large cylindrical pressure vessel with a fully circumferential crack for which experimental results are available. The example demonstrates that it is straightforward to perform the calculations required in the procedure for the time for structural failure by continuum damage mechanisms, the time for crack incubation and the time for crack growth. It is predicted that crack incubation in the pressure vessel occurs at about one-third of life with the remaining life being taken up by crack growth. The predictions are in close agreement with the experimentally observed incubation, growth and final failure times. (UK)

  17. Performance and specific emissions contours throughout the operating range of hydrogen-fueled compression ignition engine with diesel and RME pilot fuels

    Shahid Imran

    2015-09-01

    Full Text Available This paper presents the performance and emissions contours of a hydrogen dual fueled compression ignition (CI engine with two pilot fuels (diesel and rapeseed methyl ester, and compares the performance and emissions iso-contours of diesel and rapeseed methyl ester (RME single fueling with diesel and RME piloted hydrogen dual fueling throughout the engines operating speed and power range. The collected data have been used to produce iso-contours of thermal efficiency, volumetric efficiency, specific oxides of nitrogen (NOX, specific hydrocarbons (HC and specific carbon dioxide (CO2 on a power-speed plane. The performance and emission maps are experimentally investigated, compared, and critically discussed. Apart from medium loads at lower and medium speeds with diesel piloted hydrogen combustion, dual fueling produced lower thermal efficiency everywhere across the map. For diesel and RME single fueling the maximum specific NOX emissions are centered at the mid speed, mid power region. Hydrogen dual fueling produced higher specific NOX with both pilot fuels as compared to their respective single fueling operations. The range, location and trends of specific NOX varied significantly when compared to single fueling cases. The volumetric efficiency is discussed in detail with the implications of manifold injection of hydrogen analyzed with the conclusions drawn.

  18. Increasing the flexibility of base-load generating units in operation on fossil fuel

    Girshfel' d, V Ya; Khanaev, V A; Volkova, E D; Gorelov, V A; Gershenkroi, M L

    1979-01-01

    Increasing the flexibility of base-load generating units operating on fossil fuel by modifying them is a necessary measure. The highest economic effect is attained with modification of gas- and oil-fired generating units in the Western United Power Systems of the European part of the SPSS. On the basis of available experience, 150- and 200-MW units can be extensively used to regulate the power in the European part of the SPSS through putting them into reserve for the hours of the load dip at night. The change under favorable conditions of 150- and 200-MW units operating on coal to a district-heating operating mode does not reduce the possibilities for flexible operation of these units because it is possible greatly to unload the turbines while the minimum load level of the pulverized fuel fired boiler is retained through transferring a part of the heat load to the desuperheater. It is necessary to accumulate and analyze experience with operation of generating units (especially of supercritical units) with regular shutdowns and starts of groups of units and to solve the problems of modification of generating units, with differentiation with respect to types of fuel and to the united power supply system.

  19. Sinusoidal potential cycling operation of a direct ethanol fuel cell to improving carbon dioxide yields

    Majidi, Pasha; Pickup, Peter G.

    2014-12-01

    A direct ethanol fuel cell has been operated under sinusoidal (AC) potential cycling conditions in order to increase the yield of carbon dioxide and thereby increase cell efficiency relative to operation at a fixed potential. At 80 °C, faradaic yields of CO2 as high as 25% have been achieved with a PtRu anode catalyst, while the maximum CO2 production at constant potential was 13%. The increased yields under cycling conditions have been attributed to periodic oxidative stripping of adsorbed CO. These results will be important in the optimization of operating conditions for direct ethanol fuel cells, where the benefits of potential cycling are projected to increase as catalysts that produce CO2 more efficiently are implemented.

  20. Analysis of the critical and first full power operating cores for PARR using leu oxide fuel

    Khan, L.A.; Qazi, M.K.; Bokhari, I.H.; Fazal, R.

    1989-10-01

    This paper explains the analysis for determining the first full power operating core for PARR using LEU oxide fuel. The core configuration selected for this first full power operation contains about 6.13 kg of U-235 distributed in 19 standard and five control fuel elements. The neutron flux level is doubled when core is shifted from 5MW to 10 MW. Total nuclear power peaking factor of the core is 2.03. The analysis shows that the core can be operated safely at 5 MW with a flow rate of 520 meter cube per hour and at 10 MW with a flow rate of 900 meter cube per hour. (A.B.). 10 figs

  1. Acid-doped Polybenzimidazole Membranes as Electrolyte for Fuel Cells Operating Above 100°C

    Qingfeng, Li; Jensen, Jens Oluf; He, Ronhuan

    2003-01-01

    The technical achievement and challenges for the PEMFC technology based on perfluorosulfonic acid (PFSA) polymer membranes (e.g. Nafion®) are briefly discussed. The newest development in the field is alternative polymer electrolytes for operation above 100°C. As one of the successful approaches...... to high operational temperatures, the development and evaluation of acid doped PBI membranes are reviewed, covering polymer synthesis, membrane casting, acid doping, physiochemical characterization and fuel cell tests. A high temperature PEMFC system operational at up to 200°C is demonstrated with no gas...... humidification and high CO-tolerance up to 10 vol%. This high CO tolerance allows for a direct use of reformed hydrogen without further CO removal, which opens the possibility for an integrated reformer-fuel cell system. The content of this review is to a large extent based on research performed by the authors...

  2. Carbon Tolerant Fuel Electrodes for Reversible Sofc Operating on Carbon Dioxide

    Papazisi Kalliopi Maria

    2017-01-01

    Full Text Available A challenging barrier for the broad, successful implementation of Reversible Solid Oxide Fuel Cell (RSOFC technology for Mars application utilizing CO2 from the Martian atmosphere as primary reactant, remains the long term stability by the effective control and minimization of degradation resulting from carbon built up. The perovskitic type oxide material La0.75Sr0.25Cr0.9Fe0.1O3-δ (LSCF has been developed and studied for its performance and tolerance to carbon deposition, employed as bi-functional fuel electrode in a Reversible SOFC operating on the CO2 cycle (Solid Oxide Electrolysis Cell/SOEC: CO2 electrolysis, Solid Oxide Fuel Cell/SOFC: power generation through the electrochemical reaction of CO and oxygen. A commercial state-of-the-art NiO-YSZ (8% mol Y2O3 stabilized ZrO2 cermet was used as reference material. CO2 electrolysis and fuel cell operation in 70% CO/CO2 were studied in the temperature range of 900-1000°C. YSZ was used as electrolyte while LSM-YSZ/LSM (La0.2Sr0.8MnO3 as oxygen electrode. Results showed that LSCF had high and stable performance under RSOFC operation.

  3. Operational experience using the OSTR flip fuel self-protection program

    Dodd, B.; Ringle, J.C.; Anderson, T.V.; Johnson, A.G.

    1982-01-01

    Recent changes in NRC Physical Security regulations make it highly desirable for a small number of TRIGA research reactor establishments to maintain each of the fuel elements in their reactor core above the self-protection dose rate criterion. OSTR operations personnel have written a computer program (SPOOF) which calculates the exposure rate (in Rhr -1 ) from an irradiated fuel element at 3 feet in air using the actual operating history of the reactor. The purpose of this current paper is to describe the operational experience gained over the last year and a half while using the SPOOF computer program, and while performing the quarterly dose rate measurements needed to confirm the continuing accuracy of the program, and, most importantly, the self-protection status of the OSTR fuel. The computer program in association with the quarterly dose rate measurements have been accepted by the NRC, and allow the OSTR to take credit for self-protecting FLIP fuel under the current physical security regulations

  4. Balance and behavior of gaseous radionuclides released during initial PWR fuel reprocessing operations

    Leudet, A.; Miquel, P.; Goumondy, P.J.; Charrier, G.

    1982-08-01

    Five fuel pins, taken from a PWR fuel assembly with 32000 MWD/t burn-up were chopped and dissolved in leak-proof equipment designed for accurate determination of the composition and quantity of gaseous elements released in these operations. Analytical methods were specially developped to determine directly the noble gases, tritium and gaseous carbon compounds in the gas phase. Volatile iodine was kept as close as possible to the source by cold traps, then transferred to a caustic solution for quantitative analysis. The quantities and activities of gaseous fission products thus determined were compared with predicted values obtained through computation. Very good agreement was generally observed

  5. Balance and behavior of gaseous radionuclides released during initial PWR fuel reprocessing operations

    Leudet, A.; Miquel, P.; Goumondy, P.J.; Charrier, G.

    1983-01-01

    Five fuel pins, taken from a PWR fuel assembly with 32,000 MwD/t burn-up were chopped and dissolved in leak-proof equipment designed for accurate determination of the composition and quantity of gaseous elements released in these operations. Analytical methods were specially developed to determine directly the noble gases, tritium and gaseous carbon compounds in the gas phase. Volatile iodine was kept as close as possible to the source by cold traps, then transferred to a caustic solution for quantitative analysis. The quantities and activities of gaseous fission products thus determined were compared with predicted values obtained through computation. Very good agreement was generally observed

  6. Direct ethanol solid oxide fuel cell operating in gradual internal reforming

    Nobrega, S. D.; Galesco, M. V.; Girona, K.; de Florio, D. Z.; Steil, M. C.; Georges, S.; Fonseca, F. C.

    2012-09-01

    An electrolyte supported solid oxide fuel cell (SOFC) using standard electrodes, doped-lanthanum manganite cathode and Ni-cermet anode, was operated with direct (anhydrous) ethanol for more than 100 h, delivering essentially the same power output as running on hydrogen. A ceria-based layer provides the catalytic activity for the gradual internal reforming, which uses the steam formed by the electrochemical oxidation of hydrogen for the decomposition of ethanol. Such a concept opens up the way for multi-fuel SOFCs using standard components and a catalytic layer.

  7. Lessons learned from MELOX plant operation and support to design of new MOX fuel fabrication plants

    Tourre, Joel; Gattegno, Robert; Guay, Philippe; Bariteau, Jean-Pierre

    2005-01-01

    AREVA is participating in the design of the US MOX Fuel Fabrication Facility (MFFF). To support this project and allow the U.S. Department of Energy (DOE) client to reap full benefit from the MELOX operating experience, AREVA, through COGEMA and its engineering subsidiary SGN have implemented a rigorous process to prudently apply MELOX Lessons Learned to the MFFF design. This paper describes the Lessons Learned process, how the process supports the advancement of fuel fabrication technology and, how the results of the process are benefiting the client. (author)

  8. Operational experience for the latest generation of ATRIUM trademark 10 fuel assemblies

    Schoss, Volker; Hoffmann, Petra Britt; Schaefer, Jens

    2011-01-01

    AREVA NP's ATRIUM trademark 10 product family was first introduced to the BWR market in 1992. Lead test campaigns confirmed the outstanding product performance and justified introduction of reload quantities. Further development of particular product features was demonstrated and implemented in the fuel design to meet highest expectations for reliability and fuel economics. The latest generation called ATRIUM trademark 10XP and subsequently ATRIUM trademark 10XM was introduced in 2002 and 2005, respectively. The first lead test assemblies completed their operation successfully after seven cycles. (orig.)

  9. Experimental study on the impact of operating conditions and fuel composition on PCCI combustion

    Leermakers, C.A.J.

    2010-03-15

    Premixed Charge Compression Ignition (PCCI) is a combustion concept that holds the promise of combining emission levels of a spark-ignition (SI) engine with the efficiency of a compressionignition (CI) engine. In a short term scenario, PCCI combustion will be used in the low load part of the engine operating range only. This would guarantee low engine-out emission levels at operating conditions where exhaust temperatures are too low for effective NOx reduction through catalytic after treatment. At higher loads, the engine would run in conventional CI combustion mode, with emission requirements met through catalytic NOx reduction. Implicit with this scenario is that engine hardware design would be very close to that of current modern diesel engines. Compression ratio could be made load dependent through implementation of variable valve actuation. The PCCI experiments presented here have been performed using a modified 6 cylinder 12.6 liter heavy duty DI DAF XE 355 C engine. Experiments are conducted in one dedicated cylinder, which is equipped with a stand-alone fuel injection system, EGR circuit, and air compressor. For the low to medium load operating range the compression ratio has been lowered to 12:1 by means of a thicker head gasket. As engine hardware should - in the short term - preferably remain close to current diesel engines, optimizing operating conditions should focus on parameters like EGR level, intake temperature, intake pressure and injection timing. While past work in the Combustion Technology group has focused on low load PCCI combustion, in this report the effects on engine performance and emission behavior are investigated for both low and medium load PCCI combustion, up to 40% of full load. In the interpretation of experimental results, emphasis lies on the effect on combustion phasing and maximum pressure rise rate, which are inherent challenges to enable viable PCCI combustion. As in the short term scenario fuels will be used that are not too

  10. Design of vertical thermosiphon reboilers for operation under vacuum conditions application in nuclear fuel reprocessing

    Moore, M.J.C.; Keys, M.H.; Plumb, G.R.

    1988-01-01

    Reprocessing of nuclear fuel requires concentration of uranium, plutonium and other active effluent streams at various stages for purification, storage or solidification. This is usually achieved by evaporation and in U.K. plant such processes are often carried out under reduced pressure. For high throughput streams, there are considerable advantages in using vertical thermosiphon systems for evaporation and for recovery of nitric acid. However, data for such systems at reduced pressure is limited and the development by John Brown E and C Ltd of a computer program for reliable prediction of thermosiphon performance was carried out on behalf of British Nuclear Fuels Plc using data from operating plant. (author)

  11. The safety of operations in the Dragon fuel element production building during the manufacture of thorium fuel for the first charge of the Dragon Reactor experiment

    Beutler, H.; Gardham, B.; Holliday, J.

    1965-04-01

    The first charge of fuel and fuel elements for the Dragon Reactor has been completed without significant difficulty. This report covers the safety of operations during the production of the 10 thorium elements together with the final 2 driver elements needed to complete the 37 element charge. (author)

  12. Data base for a CANDU-PHW operating on a once-through, natural uranium fuel cycle

    1979-07-01

    This report, prepared for INFCE, describes a standard 600 MW(e) CANDU-PHW reactor operating on a once-through natural uranium fuel cycle. Subsequently, data are given for an extrapolated 1000 MW(e) design (the nominal capacity adopted for the INFCE study) operating on the same fuel cycle. (author)

  13. Installation, maintenance and operating manual for the Lucas-type fuel injection system of the 3 B rotary engine

    1985-01-01

    The installation procedure, maintenance, adjustment and operation of a Lucas type fuel injection system for 13B rotary racing engine is outlined. Components of the fuel injection system and installation procedure and notes are described. Maintenance, adjustment, and operation are discussed.

  14. Study by electronic structure calculations of the radiation damage in the UO2 nuclear fuel: behaviour of the point defects and fission gases

    Vathonne, Emerson

    2014-01-01

    Uranium dioxide (UO 2 ) is worldwide the most widely used fuel in nuclear plants in the world and in particular in pressurized water reactors (PWR). In-pile the fission of uranium nuclei creates fission products and point defects in the fuel. The understanding of the evolution of these radiation damages requires a multi-scale modelling approach of the nuclear fuel, from the scale of the pellet to the atomic scale. We used an electronic structure calculation method based on the density functional theory (DFT) to model radiation damage in UO 2 at the atomic scale. A Hubbard-type Coulomb interaction term is added to the standard DFT formalism to take into account the strong correlations of the 5f electrons in UO 2 . This method is used to study point defects with various charge states and the incorporation and diffusion of krypton in uranium dioxide. This study allowed us to obtain essential data for higher scale models but also to interpret experimental results. In parallel of this study, three ways to improve the state of the art of electronic structure calculations of UO 2 have been explored: the consideration of the spin-orbit coupling neglected in current point defect calculations, the application of functionals allowing one to take into account the non-local interactions such as van der Waals interactions important for rare gases and the use of the Dynamical Mean Field Theory combined to the DFT method in order to take into account the dynamical effects in the 5f electron correlations. (author) [fr

  15. Performance of diagonal control structures at different operating conditions for polymer electrolyte membrane fuel cells

    Serra, Maria; Husar, Attila; Feroldi, Diego; Riera, Jordi [Institut de Robotica i Informatica Industrial, Universitat Politecnica de Catalunya, Consejo Superior de Investigaciones Cientificas, C. Llorens i Artigas 4, 08028 Barcelona (Spain)

    2006-08-25

    This work is focused on the selection of operating conditions in polymer electrolyte membrane fuel cells. It analyses efficiency and controllability aspects, which change from one operating point to another. Specifically, several operating points that deliver the same amount of net power are compared, and the comparison is done at different net power levels. The study is based on a complex non-linear model, which has been linearised at the selected operating points. Different linear analysis tools are applied to the linear models and results show important controllability differences between operating points. The performance of diagonal control structures with PI controllers at different operating points is also studied. A method for the tuning of the controllers is proposed and applied. The behaviour of the controlled system is simulated with the non-linear model. Conclusions indicate a possible trade-off between controllability and optimisation of hydrogen consumption. (author)

  16. Impact of UO{sub 2} Enrichment of Fuel Zoning Rods in Long Cycle Operation of PWR

    Shin, Ho Cheol; Lee, Deokjung [KHNP CRI, Daejeon (Korea, Republic of); Jeong, Eun; Choe, Jiwon [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    Extending the cycle length can not only increase the energy production, but also bring down outage costs by reducing the number of refueling outages during the lifetime of a nuclear power plant. It is reasonable that more fresh fuels are loaded for long cycle operation. However, minimizing the number of fresh fuels is essential in aspect of fuel economics. This can cause high power peaking near the water holes, due to increased thermalization of neutrons in those regions. To prevent this, special fuel zoning rods are used and surround the water holes. These rods use lower-enriched uranium (they have an enrichment rate lower than the other fuel rods). If we adjust the enrichment rate of fuel zoning rods, we can reduce power peaking and moreover increase cycle length. In this paper, we designed a core suitable for long cycle operation and we conducted sensitivity tests of fuel cycle length on UO2 enrichment rate in fuel zoning region in order to extend the cycle length while using the same number of fresh fuels. The correlations between the fuel zoning enrichment and cycle length, peaking factor, CBC and shutdown margin were analyzed. The more the enrichment rate in fuel zoning region increases, the more the fuel cycle length increases. At the same time, CBC, Fq and shutdown margin do not change significantly. Increasing the fuel zoning enrichment rate presents the right property of increasing the fuel cycle length without causing a large change to CBC, Fq and shutdown margin. In conclusion, by increasing the uranium enrichment rate in fuel zoning region, fuel cycle length can be increased and the safety margins can be maintained for long cycle operation of cores.

  17. ORGAN PRESERVED RECONSTRUCTIVE OPERATIONS OF THE REPLACEMENT OF DEFECTS OF BONES, FORMING KNEE JOINT

    A. V. Balberkin

    2011-01-01

    Full Text Available New modular endoprosthesis CITO – MATI was developed and introduced in the clinical practice on the basis of the exploration of results of reconstructive interventions on knee joint area with the usage of different Russian and foreign endoprosthesis. The aim of our work was to compare the functional results and survival of CITO – MATI and Beznoska endoprosthesis. To the moment of the study among 59 patients, who had implanted Beznoska endoprosthesis, 16 (27% were secondly operated due to aseptic instability and the destruction of endoprosthesis components. Functional results of the Beznoska endoprosthesis application were good and excellent in 70,7%, and satisfactory in 29,3%. Among 107 patients with implanted CITO – MATI endoprosthesis 3 (2,8% were operated because of the aseptic instability. The destruction of endoprosthesis components in the observed period of time was not revealed. Functional results of the CITO – MATI endoprosthesis application were good and excellent in 86,6% of cases, and satisfactory in 13,4%. It was revealed, that survival and functional results of the CITO – MATI endoprosthesis application are higher, than those of the Beznoska endoprothesis.

  18. Detection of leak-defective fuel rods using the circumferential Lamb waves excited by the resonance backscattering of ultrasonic pulses

    Choi, M.S.; Yang, M.S.; Kim, H.C.

    1992-01-01

    A new ultrasonic technique for detecting the infiltrated water in leaked fuel rods is developed. Propagation characteristics of the circumferential Lamb waves in the cladding tubes are estimated by the resonance scattering theory. The Lamb waves are excited by the resonance backscattering of ultrasonic pulses. In sound fuel rods, the existence of the Lamb waves is revealed by a series of periodic echoes. In leaked fuel rods, however, the Lamb waves are perturbed strongly by the scattered waves from the surface of fuel pellets, thus the periodic echoes are not observed. (author)

  19. Confirming competence of operators - A regulatory approach to fuel cycle facilities

    Vesely, M.; Sigetich, J.

    2013-01-01

    For the past 40 years the Canadian Nuclear Safety Commission (CNSC), formerly the Atomic Energy Control Board, has certified workers in nuclear facilities. The requirement for certified personnel has ensured that workers assigned to positions that have a direct impact on the safe operation of the facility are fully qualified to perform their duties. This certification regime is defined in the regulatory framework under which the CNSC operates. Traditionally, this certification regime has been applied to Reactor Operators, Shift Supervisors and Health Physicists in Nuclear Power Plants and research reactors as well as to Exposure Device Operators who use nuclear substances for the purposes of industrial radiography. Stemming from progress made in implementing risk-informed regulatory oversight activities as well as a formal suggestion from the International Atomic Energy Agency - International Regulatory Review Service (IRRS) conducted on the CNSC in 2009, a regulatory approach to confirming the competence of Operators at Fuel Cycle Facilities has been initiated by CNSC staff. In the first stage of the implementation of this new regulatory approach, the CNSC had Cameco Corporation implement a formal internal qualification programme for the UF6 Operators at its Port Hope Conversion Facility (PHCF) in Port Hope, Ontario. In the future, following a review of the results of the qualification programme at the PHCF, the CNSC staff will evaluate the need for the application of a similar regulatory approach to confirm the competence of the Operators at other Fuel Cycle Facilities in Canada. (authors)

  20. The influence of the types of marine fuel over the Energy Efficiency Operational Index

    Acomi, Nicoleta; Acomi, Ovidiu

    2014-05-01

    One of the main concerns of our society is certainly the environment protection. The international efforts for maintaining the environment clean are various and this paper refers to the efforts in the maritime transport field. Marine pollution consists of the water pollution and also the air pollution. Regardless of the delay in recognizing the later type of pollution, it rapidly gains many organizations to argue on it. The first step was including a dedicated annex (Annex VI) in the International Convention for the Prevention of Pollution from Ships, in 1997, which seeks to minimize the airborne emissions from ships. In order to control and minimize the air pollution, the International Maritime Organization has also developed a series of measures for monitoring the emissions. These measures are grouped in three main directions: technical, operational and management related. The subject of our study is the concept of Energy Efficiency Operational Index (EEOI), developed to provide ship-owners with assistance in the process of establishing the emissions from ships in operation, and to suggest the methods for achieving their reduction. As a monitoring tool, EEOI represents the mass of CO2 emitted per unit of transport work. The actual CO2 emission from combustion of fuel on board a ship during each voyage is calculated by multiplying total fuel consumption for each type of fuel (e.g. diesel oil, gas oil, light fuel oil, heavy fuel oil, liquefied petroleum gas, liquefied natural gas) with the carbon to CO2 conversion factor for the fuel in question. The performed transport work is calculated by multiplying mass of cargo (tonnes, number of TEU/cars, or number of passengers) with the distance in nautical miles corresponding to the transport work done. Using the software developed by the author it will be emphasized the variation of the EEOI value for one vessel using different types of fuel for the voyage's legs (distance to discharge port, distance to loading port, the

  1. Durability study and lifetime prediction of baseline proton exchange membrane fuel cell under severe operating conditions

    Marrony, M.; Quenet, S.; Aslanides, A. [European Institute for Energy Research, Emmy-Noether Strasse 11, 76131 Karlsruhe (Germany); Barrera, R.; Ginocchio, S.; Montelatici, L. [Edison, Via Giorgio La Pira 2, 10028 Trofarello (Italy)

    2008-08-01

    Comparative studies of mechanical and electrochemical properties of Nafion{sup registered} - and sulfonated polyetheretherketone polymer-type membranes are carried out under severe fuel cell conditions required by industrials, within stationary and cycling electric load profiles. These membranes are proposed to be used in PEM between 70 and 90 C as fluorinated or non-fluorinated baseline membranes, respectively. Thus, though the performance of both membranes remains suitable, Nafion{sup registered} backbone brought better mechanical properties and higher electrochemical stabilities than sulfonated polyetheretherketone backbone. The performance stability and the mechanical strength of the membrane-electrode assembly were shown to be influenced by several intrinsic properties of the membrane (e.g., thermal pre-treatment, thickness) and external conditions (fuel cell operating temperature, relative humidity). Finally, a lifetime prediction for membranes under stationary conditions is proposed depending on the operation temperature. At equivalent thicknesses (i.e. 50 {mu}m), Nafion{sup registered} membranes were estimated able to operate into the 80-90 C range while sulfonated polyetheretherketone would be limited into the 70-80 C range. This approach brings baseline information about the capability of these types of polymer electrolyte membrane under fuel cell critical operations. Finally, it is revealed as a potential tool for the selection of the most promising advanced polymers for the ensuing research phase. (author)

  2. Operational programs for national radioactive waste and spent fuel management programme in Slovenia

    Zeleznik, Nadja; Kralj, Metka; Mele, Irena

    2007-01-01

    The first separate National Radioactive Waste and Spent Fuel Management Programme (National Programme) was prepared in Slovenia in 2005 as a supplementary part of the National Environmental Action Programme and was adopted in February 2006 by the Slovenian Parliament. The new National Programme includes all topics being relevant for the management of the radioactive waste and spent fuel which are produced in Slovenia, from the legislation and identification of different waste streams, to the management of radioactive waste and spent fuel, the decommissioning of nuclear facilities and management of (TE)NORM in the near future from 2006 up to the 2015. The National Programme identified the existing and possible future problems and proposed the technical solutions and action plans for two distinctive periods: 2006-2009 and 2010- 2015. According to the requirement of Act on Protection against Ionising Radiation and Nuclear Safety the national Agency for Radwaste Management (ARAO) prepared the operational programmes for the four year period with technical details on implementation of the National programme. ARAO gained the detailed plans of different involved holders and proposed 9 operational programmes with aims, measures, individual organizations in charge, expenses and resources for each of the programmes. The Operational programmes were already reviewed by the Ministry of Environment and Physical Planning and are under acceptance. The orientation of the radioactive waste management according to the National Programme and operational activities within additional limitations based on the strategical decisions of Slovenian Government is presented in the paper. (authors)

  3. The agnion Heatpipe-Reformer - operating experiences and evaluation of fuel conversion and syngas composition

    Gallmetzer, Georg; Ackermann, Pascal [Highterm Research GmbH, Hettenshausen (Germany); Schweiger, Andreas; Kienberger, Thomas [Highterm Research GmbH, Graz (Austria); Groebl, Thomas; Walter, Heimo [Technische Universitaet Wien, Institut fuer Energietechnik und Thermodynamik, Wien (Austria); Zankl, Markus; Kroener, Martin [Agnion Technologies GmbH, Hettenshausen (Germany)

    2012-09-15

    Fluidized bed gasification of solid fuels is considered as one of the core technologies for future sustainable energy supply. Whereas autothermal oxygen-driven gasification is applied in large-scale substitute natural gas (SNG) and Fischer-Tropsch (FT) plants or small-scale combined heat and power (CHP) plants, the allothermal steam-reforming process of the agnion Heatpipe-Reformer is designed for cost- and fuel-efficient syngas generation at small scales for distributed applications. The Heatpipe-Reformer's pressurized syngas generation provides a number of benefits for SNG, biomass to liquid (BTL) and CHP applications. A modified gas engine concept uses the pressurized and hydrogen-rich syngas for increased performance and tar tolerance at decreased capital expenses. Agnion has installed and operated a 500-kW thermal input pilot plant in Pfaffenhofen, Germany, over the last 2 years, showing stable operation over a variety of operating points. The syngas composition has been measured at values expected by thermodynamic models. An influence of the steam-to-fuel ratio and reformer temperature was observed. Tar and sulphur contents have been monitored and correlated to operation parameters, showing influences on stoichiometry and carbon conversion. The mass and energy streams of the plant were balanced. One of the main observations in the monitoring programme is the fact that syngas output, efficiency and syngas quality correlate to high values if the carbon conversion is high. Carbon conversion rates and cold gas efficiencies are comparably high in respect to today's processes, promising economic and fuel-efficient operation of the Heatpipe-Reformer applications. (orig.)

  4. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    Sanders, T.L.; Lake, W.H.

    1989-01-01

    The traditional assumption used in evaluating criticality safety of spent fuel cask is that the spent fuel is as reactive as when it was fresh (new). This is known as the fresh fuel assumption. It avoids a number of calculational and verification difficulties, but could take a heavy toll in decreased efficiency. The alternative to the fresh fuel assumption is called burnup credit. That is, the reduced reactivity of spent fuel that comes about from depletion of fissile radionuclides and net increase in neutron absorbers (poisons) is taken into account. It is recognizable that the use of burnup credit will in fact increase the percentage of unacceptable or non-specification fuel available for misloading. This could reduce individual cask safety margins if current practices with respect to loading procedures are maintained. As such, additional operational, design, analysis, and validation requirements should be established that, as a minimum, compensate for any potential reduction in fuel loading safety margin. This method is based on a probabilistic (PRA) approach and is called a relative risk comparison. The method assumes a linear risk model, and uses a selected probability function to compare the system of interest and an acceptable reference system by varying the features of each to assess effects on system safety. While risk is the product of an event probability and its consequence, the consequences of criticality in a cask are considered to be both unacceptable and the same, regardless of the initiating sequence. Therefore, only the probability of the event is considered in a relative risk evaluation

  5. Ground measurements of fuel and fuel consumption from experimental and operational prescribed fires at Eglin Air Force Base, Florida

    Roger D. Ottmar; Robert E. Vihnanek; Clinton S. Wright; Andrew T. Hudak

    2014-01-01

    Ground-level measurements of fuel loading, fuel consumption, and fuel moisture content were collected on nine research burns conducted at Eglin Air Force Base, Florida in November, 2012. A grass or grass-shrub fuelbed dominated eight of the research blocks; the ninth was a managed longleaf pine (Pinus palustrus) forest. Fuel loading ranged from 1.7 Mg ha-1 on a...

  6. Dual-fuel HCCI operation with DME/LPG/gasoline/hydrogen

    Bae, C.

    2009-01-01

    The advantages of homogeneous charge compression ignition (HCCI) engines include usage of the different type of fuels, ultra low nitrogen oxide and particulate matter emissions and improved fuel economy. Disadvantages include an excessive combustion rate, engine noise, and hydrocarbon and carbon emissions. An experiment on dual-fuel HCCI operation with dimethyl ether (DME)/liquefied petroleum gas (LPG)/gasoline/hydrogen was presented. The advantages and disadvantages were first presented and the dual-fuel HCCI combustion engine was illustrated through an experimental apparatus. The experimental conditions were also presented in terms of engine speed, DME injection quantity, LPC injection quantity, and LPC composition. Experimental results were discussed for output performance and indicated mean effective pressure (IMEP). It was concluded that the effect of LPG composition in a DME-LPG dual-fueled HCCI engine at various injection quantity and injective timing were observed. Specifically, it was found that propane was a more effective way to increase IMEP in this study, and that in a DME HCCI engine, higher load limit was extended by using LPG as an ignition inhibitor. tabs., figs.

  7. Special considerations on operating a fuel cell power plant using natural gas with marginal heating value

    Moses, L. Ng; Chien-Liang Lin [Industrial Technology Research Institute, Taiwan (China); Ya-Tang Cheng [Power Research Institute, Taiwan (China)

    1996-12-31

    In realizing new power generation technologies in Taiwan, a phosphoric acid fuel cell power plant (model PC2513, ONSI Corporation) has been installed in the premises of the Power Research Institute of the Taiwan Power Company in Taipei County of Taiwan. The pipeline gas supplying to the site of this power plant has a high percentage of carbon dioxide and thus a slightly lower heating value than that specified by the manufacturer. Because of the lowering of heating value of input gas, the highest Output power from the power plant is understandably less than the rated power of 200 kW designed. Further, the transient response of the power plant as interrupted from the Grid is also affected. Since this gas is also the pipeline gas supplying to the heavily populated Taipei Municipal area, it is conceivable that the success of the operations of fuel cells using this fuel is of vital importance to the promotion of the use of this power generation technology in Taiwan. Hence, experiments were set up to assess the feasibility of this fuel cell power plant using the existing pipeline gas in this part of Taiwan where fuel cells would most likely find useful.

  8. Design and operational experience of the NUHOMS-24P spent fuel storage system

    McConaghy, W.J.; Lehnert, R.A.; Rasmussen, R.W.

    1991-01-01

    The NUHOMS spent fuel storage system provides a safe and economical method for the dry storage of spent fuel assemblies either at an independent spent fuel storage installation (ISFSI) at reactor or at a centralized storage facility away from reactor. The system consists of three major safety-related components: a dry shielded canister (DSC) which provides a high integrity containment boundary and a controlled storage environment for the fuel; a reinforced concrete horizontal storage module (HSM) which houses the stored DSCs and provides radiation shielding, protection against natural phenomena and an efficient means for decay heat removal; and a transfer cask which provides for the safe shielded transfer of DSCs from a plant spent fuel pool to a HSM. The NUHOMS system is designed and licensed to the requirements of 10 CFR 72 and ANS/ANSI 57.9 for ISFSIs. The NUHOMS concept was developed in early 1980s, and in 1987, a larger version of the NUHOMS system, 24P, was developed. The operational features of NUHOMS and the loading experience at Oconee are reported. (K.I.)

  9. Regulated and unregulated emissions from an internal combustion engine operating on ethanol-containing fuels

    Poulopoulos, S. G.; Samaras, D. P.; Philippopoulos, C. J.

    In the present work, the effect of ethanol addition to gasoline on regulated and unregulated emissions is studied. A 4-cylinder OPEL 1.6 L internal combustion engine equipped with a hydraulic brake dynamometer was used in all the experiments. For exhaust emissions treatment a typical three-way catalyst was used. Among the various compounds detected in exhaust emissions, the following ones were monitored at engine and catalyst outlet: methane, hexane, ethylene, acetaldehyde, acetone, benzene, 1,3-butadiene, toluene, acetic acid and ethanol. Addition of ethanol in the fuel up to 10% w/w had as a result an increase in the Reid vapour pressure of the fuel, which indicates indirectly increased evaporative emissions, while carbon monoxide tailpipe emissions were decreased. For ethanol-containing fuels, acetaldehyde emissions were appreciably increased (up to 100%), especially for fuel containing 3% w/w ethanol. In contrast, aromatics emissions were decreased by ethanol addition to gasoline. Methane and ethanol were the most resistant compounds to oxidation while ethylene was the most degradable compound over the catalyst. Ethylene, methane and acetaldehyde were the main compounds present at engine exhaust while methane, acetaldehyde and ethanol were the main compounds in tailpipe emissions for ethanol fuels after the catalyst operation.

  10. Evaluation of gap heat transfer model in ELESTRES for CANDU fuel element under normal operating conditions

    Lee, Kang Moon; Ohn, Myung Ryong; Im, Hong Sik; Choi, Jong Hoh; Hwang, Soon Taek

    1995-01-01

    The gap conductance between the fuel and the sheath depends strongly on the gap width and has a significant influence on the amount of initial stored energy. The modified Ross and Stoute gap conductance model in ELESTRES is based on a simplified thermal deformation model for steady-state fuel temperature calculations. A review on a series of experiments reveals that fuel pellets crack, relocate, and are eccentrically positioned within the sheath rather than solid concentric cylinders. In this paper, the two recently-proposed gap conductance models (offset gap model and relocated gap model) are described and are applied to calculate the fuel-sheath gap conductances under experimental conditions and normal operating conditions in CANDU reactors. The good agreement between the experimentally-inferred and calculated gap conductance values demonstrates that the modified Ross and Stoute model was implemented correctly in ELESTRES. The predictions of the modified Ross and Stoute model provide conservative values for gap heat transfer and fuel surface temperature compared to the offset gap and relocated gap models for a limiting power envelope. 13 figs., 3 tabs., 16 refs. (Author)

  11. Design, fabrication and operating experience of Monju ex-vessel fuel storage tank

    Yokota, Yoshio; Yamagishi, Yoshiaki; Kuroha, Mitsuo; Inoue, Tatsuya

    1995-01-01

    In FBRs there are two methods of storing and cooling the spent fuel - the in-vessel storage and the ex-vessel storage. Because of the sodium leaks through the tank at the beginning of pre-operation, the utilization of the ex-vessel fuel storage tank (EVST) of some FBR plant has been changed from the ex-vessel fuel storage to the interim fuel transfer tank. This led to reactor designers focusing on the material, structure and fabrication of the carbon steel sodium storage tanks worldwide. The Monju EVST was at the final stage of the design, when the leaks occurred. The lesson learned from that experience and the domestic fabrication technology are reflected to the design and fabrication of the Monju EVST. This paper describes the design, fabrication and R and D results for the tank, and operating experience in functional test. The items to be examined are as follows: (1) Overall structure of the tank and design philosophy on the function, (2) Structure of the cover shielding plug and its design philosophy, (3) Structures of the rotating rack and its bearings, and their design philosophy, (4) Cooling method and its design philosophy, (5) Structure and fabrication of the cooling coil support inside EVST with comparison of leaked case, (6) R and D effort for items above. The fabrication of the Monju EVST started in August 1986 and it was shipped to the site in March 1990. Installation was completed in November 1990, and sodium fill after pre-heating started in 1991. The operation has been continued since September 1992. In 1996 when the first spent fuel is stored, its total functions will be examined. (author)

  12. Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor

    Gonzalez C, J.; Martin del Campo M, C.

    2003-01-01

    This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

  13. Performance of a diesel engine operating on raw coal-diesel fuel and solvent refined coal-diesel fuel slurries. Final report

    Marshall, H.P.

    1980-03-01

    Performance tests using an 11 kW single cylinder diesel engine were made to determine the effects of three different micronized coal-fuel oil slurries being considered as alternative fuels. Slurries containing 20, 32, and 40%-wt micronized raw coal in No. 2 fuel oil were used. Results are presented indicating the changes in the concentrations of SO/sub X/ and NO/sub X/ in the exhaust, exhaust opacity, power and efficiency, and in wear rates relative to operation on fuel oil No. 2. The engine was operated for 10 h at full load and 1400 rpm on al fuels except the 40%-wt slurry. This test was discontinued because of extremely poor performance.

  14. Radiological impact of plutonium recycle in the fuel cycle of LWR type reactors: professional exposure during mormal operation

    White, I.F.; Kelly, G.N.

    1983-01-01

    The radiological impact of the fuel cycle of light water type reactors using enriched uranium may be changed by plutonium recycle. The impact on human population and on the persons professionally exposed may be different according to the different steps of the fuel cycle. This report analyses the differential radiological impact on the different types of personnel involed in the fuel cycle. Each step of the fuel cycle is separately studied (fuel fabrication, reactor operation, fuel reprocessing), as also the transport of the radioactive materials between the different steps. For the whole fuel cycle, one estimates that, with regard to the fuel cycle using enriched uranium, the plutonium recycle involves a small increase of the professional exposure

  15. Integrated anode structure for passive direct methanol fuel cells with neat methanol operation

    Wu, Huijuan; Zhang, Haifeng; Chen, Peng; Guo, Jing; Yuan, Ting; Zheng, Junwei; Yang, Hui

    2014-02-01

    A microporous titanium plate based integrated anode structure (Ti-IAS) suitable for passive direct methanol fuel cells (DMFCs) fueled with neat methanol is reported. This anode structure incorporates a porous titanium plate as a methanol mass transfer barrier and current collector, pervaporation film for passively vaporizing methanol, vaporous methanol cavity for evenly distributing fuel, and channels for carbon dioxide venting. With the effective control of methanol delivery rate, the Ti-IAS based DMFC allows the direct use of neat methanol as the fuel source. In the meantime, the required water for methanol-oxidation reaction at the anode can also be fully recovered from the cathode with the help of the highly hydrophobic microporous layer in the cathode. DMFCs incorporating this new anode structure exhibit a power density as high as 40 mW cm-2 and a high volumetric energy density of 489 Wh L-1 operating with neat methanol and at 25 °C. Importantly, no obvious performance degradation of the passive DMFC system is observed after more than 90 h of continuous operation. The experimental results reveal that the compact DMFC based on the Ti-IAS exhibits a substantial potential as power sources for portable applications.

  16. Change in nuclear fuel material processing operation at Tokai Plant of Mitsubishi Atomic Fuel Co., Ltd. (report)

    1987-01-01

    This report, compiled by the Nuclear Safety Commission to be submitted to the Prime Minister, deals with studies on a proposed change in the operation of processing nuclear fuel substances at the Tokai Plant of Mitsubishi Atomic Fuel Co., Ltd. The conclusions of and principles for the examination and evaluation are described. It is concluded that part of the proposed change is appropriate with respect to required technical capability and that part of the change will not have adverse effects on the safety of the plant. The studies carried out are focused on the safety of the facilities. The study on the earthquake resistance reveals that anti-earthquake design for the new buildings is properly developed. The buildings are of fireproof construction and the systems and equipment to be installed are made of incomustible materials to ensure the prevention of fire and explosion. It is confirmed that criticality control (for each unit and for the group of units) will be performed appropriately and that the waste (gaseous waste, liquid waste, solid waste) treatment systems are designed appropriately. A study is also made on the radiation control methods (working condition control, individual exposure control, surrounding environment control). In addition accident evaluation is carried out to confirm the safety of the residents around the plant. (Nogami, K.)

  17. Change in nuclear fuel material processing operation at Tokai Plant of Mitsubishi Atomic Fuel Co. , Ltd. (report)

    1987-09-01

    This report, compiled by the Nuclear Safety Commission to be submitted to the Prime Minister, deals with studies on a proposed change in the operation of processing nuclear fuel substances at the Tokai Plant of Mitsubishi Atomic Fuel Co., Ltd. The conclusions of and principles for the examination and evaluation are described. It is concluded that part of the proposed change is appropriate with respect to required technical capability and that part of the change will not have adverse effects on the safety of the plant. The studies carried out are focused on the safety of the facilities. The study on the earthquake resistance reveals that anti-earthquake design for the new buildings is properly developed. The buildings are of fireproof construction and the systems and equipment to be installed are made of incomustible materials to ensure the prevention of fire and explosion. It is confirmed that criticality control (for each unit and for the group of units) will be performed appropriately and that the waste (gaseous waste, liquid waste, solid waste) treatment systems are designed appropriately. A study is also made on the radiation control methods (working condition control, individual exposure control, surrounding environment control). In addition accident evaluation is carried out to confirm the safety of the residents around the plant. (Nogami, K.).

  18. Pre-operative evaluation with MR in tetralogy of Fallot and pulmonary atresia with ventricular septal defect

    Holmqvist, C.; Hochbergs, P.; Bjoerkhem, G.; Brockstedt, S.; Laurin, S.

    2000-01-01

    To assess whether MR imaging could replace angiography in pre-operative evaluation of patients with tetralogy of Fallot and pulmonary atresia with ventricular septal defect (VSD), especially since the surgical correction was done earlier than was previously the rule. Fourteen patients with tetralogy of Fallot (n=10) or pulmonary atresia with VSD (n=4), mean age 7.5±4.4 months, were evaluated with angiocardiography and MR before definitive surgical correction. There was good diagnostic agreement between the two modalities when evaluating right ventricular outflow obstruction; 86% for valvular and 93% for supra valvular stenosis, but the agreement was somewhat lower for the sub valvular obstruction (57%). Surgery findings, however, were in favour of MR in 5 patients concerning the sub valvular right ventricular outflow tract obstruction. MR images identified all stenoses in the right and left pulmonary arteries, but overlooked one stenosis in the main pulmonary artery. MR could evaluate patency in all palliative shunts. Even in this young age group, MR imaging offers a good alternative to angiocardiography for the pre-operative evaluation of the right ventricular outflow tract, the main pulmonary artery and the proximal right and left pulmonary arteries, before definitive surgical correction of tetralogy of Fallot and pulmonary atresia with VSD

  19. Pre-operative evaluation with MR in tetralogy of Fallot and pulmonary atresia with ventricular septal defect

    Holmqvist, C.; Hochbergs, P. [Univ. Hospital, Lund (Sweden). Dept of Diagnostic Radiology; Bjoerkhem, G. [Univ. Hospital, Lund (Sweden). Dept of Paediatrics; Brockstedt, S.; Laurin, S. [Univ. Hospital, Lund (Sweden). Dept of Diagnostic Radiology

    2000-01-01

    To assess whether MR imaging could replace angiography in pre-operative evaluation of patients with tetralogy of Fallot and pulmonary atresia with ventricular septal defect (VSD), especially since the surgical correction was done earlier than was previously the rule. Fourteen patients with tetralogy of Fallot (n=10) or pulmonary atresia with VSD (n=4), mean age 7.5{+-}4.4 months, were evaluated with angiocardiography and MR before definitive surgical correction. There was good diagnostic agreement between the two modalities when evaluating right ventricular outflow obstruction; 86% for valvular and 93% for supra valvular stenosis, but the agreement was somewhat lower for the sub valvular obstruction (57%). Surgery findings, however, were in favour of MR in 5 patients concerning the sub valvular right ventricular outflow tract obstruction. MR images identified all stenoses in the right and left pulmonary arteries, but overlooked one stenosis in the main pulmonary artery. MR could evaluate patency in all palliative shunts. Even in this young age group, MR imaging offers a good alternative to angiocardiography for the pre-operative evaluation of the right ventricular outflow tract, the main pulmonary artery and the proximal right and left pulmonary arteries, before definitive surgical correction of tetralogy of Fallot and pulmonary atresia with VSD.

  20. 49 CFR 1242.67 - Switch crews; controlling operations; yard and terminal clerical; locomotive fuel; electric power...

    2010-10-01

    ... 49 Transportation 9 2010-10-01 2010-10-01 false Switch crews; controlling operations; yard and terminal clerical; locomotive fuel; electric power purchased/produced for motive power; operating switches... SERVICE FOR RAILROADS 1 Operating Expenses-Transportation § 1242.67 Switch crews; controlling operations...

  1. Performance of an accountability measurement system at an operating fuel reprocessing facility

    Wade, M.A.; Spraktes, F.W.; Hand, R.L.; Baldwin, J.M.; Filby, E.E.; Lewis, L.C.

    1978-01-01

    The ICPP has been engaged for 25 years in the recovery of uranium from spent reactor fuels. In concert with the reprocessing activity, an accountability measurements system has been operated throughout the history of the ICPP. The structure and functions of the accountability measurements system are presented. Its performance is evaluated in order to illustrate the relation of analytical methodology to the overall measurements system. 6 figures, 5 tables

  2. Operation of Atucha I nuclear power plant with 25 cooling channels without fuel elements

    Perez, R.A.; Sidelnik, J.I.; Salom, G.F.

    1987-01-01

    In view of the need of removing the irradiation probes from the reactor of Atucha I nuclear power plant, a study about the consequences of operating with 25 channels without their respective fuel elements was performed. This condition was simulated by means of the code PUMA symmetry I and the consequences were analyzed. From the study resulted a program of stepped power reduction of the nuclear plant that would take place during the process of channel emptying. (Author)

  3. Safety analysis of ventilation and inspection operations on barrels in which spent fuel elements are stored

    Pavlovic, R.; Todorovic, D.; Plecas, I.

    2000-01-01

    Safety analysis of possible accident during performing ventilation and inspection operations on barrels in which spent fuel elements are stored has been proposed. Investigation is confined to a case when primary contamination is localized to the disposal room and controlled release of gaseous effluents to the atmosphere. 85 Kr is the most important radionuclide considered risk estimation due to release of gaseous fission products (author) [sr

  4. Increasing the operation temperature of polymer electrolyte membranes for fuel cells: From nanocomposites to hybrids

    Licoccia, Silvia; Traversa, Enrico

    Among the possible systems investigated for energy production with low environmental impact, polymeric electrolyte membrane fuel cells (PEMFCs) are very promising as electrochemical power sources for application in portable technology and electric vehicles. For practical applications, operating FCs at temperatures above 100 °C is desired, both for hydrogen and methanol fuelled cells. When hydrogen is used as fuel, an increase of the cell temperature produces enhanced CO tolerance, faster reaction kinetics, easier water management and reduced heat exchanger requirement. The use of methanol instead of hydrogen as a fuel for vehicles has several practical benefits such as easy transport and storage, but the slow oxidation kinetics of methanol needs operating direct methanol fuel cells (DMFCs) at intermediate temperatures. For this reason, new membranes are required. Our strategy to achieve the goal of operating at temperatures above 120 °C is to develop organic/inorganic hybrid membranes. The first approach was the use of nanocomposite class I hybrids where nanocrystalline ceramic oxides were added to Nafion. Nanocomposite membranes showed enhanced characteristics, hence allowing their operation up to 130 °C when the cell was fuelled with hydrogen and up to 145 °C in DMFCs, reaching power densities of 350 mW cm -2. The second approach was to prepare Class II hybrids via the formation of covalent bonds between totally aromatic polymers and inorganic clusters. The properties of such covalent hybrids can be modulated by modifying the ratio between organic and inorganic groups and the nature of the chemical components allowing to reach high and stable conductivity values up to 6.4 × 10 -2 S cm -1 at 120 °C.

  5. Fuel consumption in an air blower for agricultural use under different operating conditions

    Silva, Robson L. da

    2017-01-01

    ABSTRACT Evaluation of fuel consumption in internal combustion engines (ICE) of agricultural machinery and equipment is important in determining the performance under various operating conditions, especially when using biofuels. This study consisted of experimental evaluation of the gasoline (petrol)/ethanol consumption in a two-stroke 1-cylinder ICE, Otto cycle, functioning as an air blower for agriculture and related applications. A methodology for tests of non-automotive ICE, based on ABNT...

  6. Behaviour of short-lived fission products within operating UO2 fuel elements

    Hastings, I.J.; Hunt, C.E.L.; Lipsett, J.J.

    1983-01-01

    We have carried out experiments using a ''sweep gas'' technique to determine the behaviour of short-lived fission products within operating, intact UO 2 fuel elements. The Zircaloy-4-clad elements were 500 mm long and contained fuel of density 10.65-10.71 Mg/m 3 . A He-2% H 2 carrier gas swept gaseous or volatile fission products out of the operating fuel element past a gamma spectrometer for measurement. In tests at linear powers of 45 and 60 kW/m to maximum burnups of 70 MW.h/kg U, the species measured directly at the spectrometer were generally the short-lived xenons and kryptons. We did not observe iodine or bromine during normal operation. However, we have deduced the behaviour of I-133 and I-135 from the decay of Xe-133 and Xe-135 during reactor shutdowns. Plots of R/B (released/born) against lambda (decay constant) or effective lambda for all isotopes observed at 45 and 60 kW/m show that a line of slope -0.5, corresponding with diffusion kinetics, is a good fit to the measured xenon and krypton data. Our inferred release of iodine fits the same line. From this we can extrapolate to an R/B for I-131 of about 5x10 -3 . The ANS 5.4 release correlation gives calculated results in good agreement with our measurements. (author)

  7. Analyses of the Anticipated Operational Occurrences for the HANARO Fuel Test Loop

    Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, C. Y.; Ahn, S. H

    2007-12-15

    The analyses of anticipated operational occurrences of the HANARO fuel test loop have been carried out by using the MARS/FTL{sub A} code, which is a modified version of the MARS code. A critical heat flux correlation on the three rods with triangular array was implemented in the MARS/FTL{sub A} code. The correlation was obtained from the critical heat fluxes measured at a test section, which is the same geometry of the in-pile test section of the HANARO fuel test loop. The anticipated operational occurrences of the HANARO fuel test loop are the inadvertent closure of the isolation valves, the over-power transient of the HANARO, the stuck open of the safety valves, and the loss of HANARO class IV power. A minimum DNBR (Departure from Nucleate Boiling Ratio) was predicted in the inadvertent closure of the isolation valves. It is indicated that the minimum DNBR of 1.85 is greater than the design limit DNBR of 1.39. The maximum coolant pressure calculated in the anticipated operational occurrences is also less than the 110 percents of the design pressure.

  8. Analyses of the Anticipated Operational Occurrences for the HANARO Fuel Test Loop

    Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, C. Y.; Ahn, S. H.

    2007-12-01

    The analyses of anticipated operational occurrences of the HANARO fuel test loop have been carried out by using the MARS/FTL A code, which is a modified version of the MARS code. A critical heat flux correlation on the three rods with triangular array was implemented in the MARS/FTL A code. The correlation was obtained from the critical heat fluxes measured at a test section, which is the same geometry of the in-pile test section of the HANARO fuel test loop. The anticipated operational occurrences of the HANARO fuel test loop are the inadvertent closure of the isolation valves, the over-power transient of the HANARO, the stuck open of the safety valves, and the loss of HANARO class IV power. A minimum DNBR (Departure from Nucleate Boiling Ratio) was predicted in the inadvertent closure of the isolation valves. It is indicated that the minimum DNBR of 1.85 is greater than the design limit DNBR of 1.39. The maximum coolant pressure calculated in the anticipated operational occurrences is also less than the 110 percents of the design pressure

  9. In reactor performance of defected zircaloy-clad U{sub 3}Si fuel elements in pressurized and boiling water coolants

    Feraday, M A; Allison, G M; Ambler, J F.R.; Chalder, G H; Lipsett, J J

    1968-05-15

    The results of two in-reactor defect tests of Zircaloy-clad U{sub 3}Si are reported. In the first test, a previously irradiated element ({approx}5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at {approx}270{sup o}C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U{sub 3}Si at 300{sup o}C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  10. Environmental aspects based on operation performance of nuclear fuel fabrication facilities

    2001-07-01

    This publication was prepared within the framework of the IAEA Project entitled Development and Upgrading of Guidelines, Databases and Tools for Integrating Comparative Assessment into Energy System Analysis and Policy Making, which included the collection, review and input of data into a database on health and environmental impacts related to operation of nuclear fuel cycle facilities. The objectives of the report included assembling environmental data on operational performance of nuclear fabrication facilities in each country; compiling and arranging the data in a database, which will be easily available to experts and the public; and presenting data that may be of value for future environmental assessment of nuclear fabrication facilities

  11. 48{sup th} Annual meeting on nuclear technology (AMNT 2017). Key topic / Enhanced safety and operation excellence. Technical session: Operation and safety of nuclear installations, fuel

    Hollands, Thorsten [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Bereich Reaktorsicherheitsforschung

    2017-12-15

    The sessions Fuel and Materials and Containment and SFP, as part of the Technical Sessions Operation and Safety of Nuclear Installations, Fuel implemented in the Key Topic Enhanced Safety and Operation Excellence were chaired by Dr. Thorsten Hollands (Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH) and Dr. Erwin Fischer (PreussenElektra GmbH) who was the keynote coordinator for the Technical Sessions. Both sessions consist of a keynote lecture followed by technical presentations.

  12. Spent Nuclear Fuel (SNF) Cold Vacuum Drying (CVD) Facility Operations Manual; FINAL

    IRWIN, J.J.

    1999-01-01

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-553, Spent Nuclear Fuel Project Final Safety Analysis Report Annex B-Cold Vacuum Drying Facility. The HNF-SD-SNF-DRD-002, 1999, Cold Vacuum Drying Facility Design Requirements, Rev. 4, and the CVDF Final Design Report. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence and references to the CVDF System Design Descriptions (SDDs). This manual has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  13. Spent Nuclear Fuel (SNF) Cold Vacuum Drying (CVD) Facility Operations Manual

    IRWIN, J.J.

    1999-07-02

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-553, Spent Nuclear Fuel Project Final Safety Analysis Report Annex B--Cold Vacuum Drying Facility. The HNF-SD-SNF-DRD-002, 1999, Cold Vacuum Drying Facility Design Requirements, Rev. 4, and the CVDF Final Design Report. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence and references to the CVDF System Design Descriptions (SDDs). This manual has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  14. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    IRWIN, J.J.

    2000-02-03

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of the Processing Systems (Garvin 1998) and, the HNF-SD-SNF-DRD-002, 1997, Cold Vacuum Drying Facility Design Requirements, Rev. 3a. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence, and has been developed for the spent nuclear fuel project (SNFP) Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  15. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    IRWIN, J.J.

    2000-01-01

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of the Processing Systems (Garvin 1998) and, the HNF-SD-SNF-DRD-002, 1997, Cold Vacuum Drying Facility Design Requirements, Rev. 3a. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence, and has been developed for the spent nuclear fuel project (SNFP) Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  16. Operational considerations in specifying legal weight vehicles for the highway transport of spent nuclear fuel

    Hill, C.V.; Rutenkroger, E.O.; Ratledge, J.E.

    1990-01-01

    This paper presents the results of a research project in which tractor manufacturers and carrier companies were interviewed to gather information on operational concerns in specifying a tractor to haul legal weight spent fuel casks. The system was assumed to operate very close to the 80,000 pound legal weight limit. Safety, performance, reliability, and maintainability of equipment were factors given particular attention. The interaction between driver fatigue, safety, and equipment was also discussed. Innovative operating strategies that could save weight were discussed. The paper concluded that operational considerations require that planners working with standard off-the-shelf tractor equipment should allow at least 17,350 pounds for the weight of the tractor as a starting point from which further weight reduction analysis can proceed. 4 refs., 3 figs

  17. An innovative multivariate tool for fuel consumption and costs estimation of agricultural operations

    Mirko Guerrieri

    2016-12-01

    Full Text Available The estimation of operating costs of agricultural and forestry machineries is a key factor in both planning agricultural policies and farm management. Few works have tried to estimate operating costs and the produced models are normally based on deterministic approaches. Conversely, in the statistical model randomness is present and variable states are not described by unique values, but rather by probability distributions. In this study, for the first time, a multivariate statistical model based on Partial Least Squares (PLS was adopted to predict the fuel consumption and costs of six agricultural operations such as: ploughing, harrowing, fertilization, sowing, weed control and shredding. The prediction was conducted on two steps: first of all few initial selected parameters (time per surface-area unit, maximum engine power, purchase price of the tractor and purchase price of the operating machinery were used to estimate the fuel consumption; then the predicted fuel consumption together with the initial parameters were used to estimate the operational costs. Since the obtained models were based on an input dataset very heterogeneous, these resulted to be extremely efficient and so generalizable and robust. In details the results show prediction values in the test with r always ≥ 0.91. Thus, the approach may results extremely useful for both farmers (in terms of economic advantages and at institutional level (representing an innovative and efficient tool for planning future Rural Development Programmes and the Common Agricultural Policy. In light of these advantages the proposed approach may as well be implemented on a web platform and made available to all the stakeholders.

  18. An innovative multivariate tool for fuel consumption and costs estimation of agricultural operations

    Guerrieri, M.; Fedrizzi, M.; Antonucci, F.; Pallottino, F.; Sperandio, G.; Pagano, M.; Figorilli, S.; Menesatti, P.; Costa, C.

    2016-01-01

    The estimation of operating costs of agricultural and forestry machineries is a key factor in both planning agricultural policies and farm management. Few works have tried to estimate operating costs and the produced models are normally based on deterministic approaches. Conversely, in the statistical model randomness is present and variable states are not described by unique values, but rather by probability distributions. In this study, for the first time, a multivariate statistical model based on Partial Least Squares (PLS) was adopted to predict the fuel consumption and costs of six agricultural operations such as: ploughing, harrowing, fertilization, sowing, weed control and shredding. The prediction was conducted on two steps: first of all few initial selected parameters (time per surface-area unit, maximum engine power, purchase price of the tractor and purchase price of the operating machinery) were used to estimate the fuel consumption; then the predicted fuel consumption together with the initial parameters were used to estimate the operational costs. Since the obtained models were based on an input dataset very heterogeneous, these resulted to be extremely efficient and so generalizable and robust. In details the results show prediction values in the test with r always ≥ 0.91. Thus, the approach may results extremely useful for both farmers (in terms of economic advantages) and at institutional level (representing an innovative and efficient tool for planning future Rural Development Programmes and the Common Agricultural Policy). In light of these advantages the proposed approach may as well be implemented on a web platform and made available to all the stakeholders.

  19. An innovative multivariate tool for fuel consumption and costs estimation of agricultural operations

    Guerrieri, M.; Fedrizzi, M.; Antonucci, F.; Pallottino, F.; Sperandio, G.; Pagano, M.; Figorilli, S.; Menesatti, P.; Costa, C.

    2016-07-01

    The estimation of operating costs of agricultural and forestry machineries is a key factor in both planning agricultural policies and farm management. Few works have tried to estimate operating costs and the produced models are normally based on deterministic approaches. Conversely, in the statistical model randomness is present and variable states are not described by unique values, but rather by probability distributions. In this study, for the first time, a multivariate statistical model based on Partial Least Squares (PLS) was adopted to predict the fuel consumption and costs of six agricultural operations such as: ploughing, harrowing, fertilization, sowing, weed control and shredding. The prediction was conducted on two steps: first of all few initial selected parameters (time per surface-area unit, maximum engine power, purchase price of the tractor and purchase price of the operating machinery) were used to estimate the fuel consumption; then the predicted fuel consumption together with the initial parameters were used to estimate the operational costs. Since the obtained models were based on an input dataset very heterogeneous, these resulted to be extremely efficient and so generalizable and robust. In details the results show prediction values in the test with r always ≥ 0.91. Thus, the approach may results extremely useful for both farmers (in terms of economic advantages) and at institutional level (representing an innovative and efficient tool for planning future Rural Development Programmes and the Common Agricultural Policy). In light of these advantages the proposed approach may as well be implemented on a web platform and made available to all the stakeholders.

  20. Assessment of causes for degrading fuel performance at Darlington NGS

    Judah, J.; Goodchild, S.

    2013-01-01

    Fuel performance at the Darlington nuclear generating station has historically been excellent. Until recently, the majority of these few fuel defects have been attributed to fretting by heat transport system debris. The minority have been linked to manufacturing issues. Recently, Darlington has experienced an increase in the number of fuel defects. Although the defect rate remains low with respect to industry standards, this defect experience is considered to be unacceptable given current industry expectations and the OPG zero defect policy. Nine fuel defects have been discharged since 2007 from the four Darlington reactors. This represents a fuel defect rate of just 0.35 defects per year per reactor. At the time of this writing three additional defects are suspected to be in core. Although a definitive defect cause has yet to be identified, these fuel performance issues appear to be due to the coincidental degradation of manufacturing and operational factors, thereby decreasing the margins to fuel failure due to fuelling power ramps. All of the confirmed defected bundles have been long bundles and all experienced a relatively high power ramp when shifted from Position 2 to Position 6. High bundle uranium masses and low internal clearances are thought to be significant contributing factors. Bundle burnups at the time of the power ramps were low and these bundles were not identified by existing power ramp defect predictive tools. Our assessment has resulted in a number of recommendations which are designed to mitigate these adverse conditions by restoring the margins to power ramp failures. These recommendations impact broadly across a number of organizations including reactor physics, fuel design, fuel manufacturing, reactor design, inspections and PIE. (author)

  1. Optimised fuel cell operation management. Final report; Optimiertes Betriebsmanagement von Brennstoffzellen. Schlussbericht

    NONE

    2003-11-05

    Investigations were carried out on two fuel cell systems of the type ONSI PC 25C (large fuel cells) and three fuel cell heaters of the type HXS 1000 PREMIERE by Sulzer HEXIS (small fuel cells) for long periods of time and in different modes of operation. Information was obtained on efficiencies and performance, pollutant emissions, degradation and availability. The status of the project is reported and compared with the goals stated initially. (orig.) [German] Das Hauptziel des Projektes bestand darin, Kriterien fuer den optimalen Betrieb (hohe Verfuegbarkeit, geringe Anzahl von Abschaltungen, hohe Energienutzung etc.) von Brennstoffzellen im Zusammenspiel mit den peripheren Komponenten der Gebaeudetechnik zu ermitteln. Hierzu wurden Untersuchungen an zwei Brennstoffzellen-Anlagen vom Typ ONSI PC 25C (grosse BZ) und an drei Brennstoffzellen-Heizgeraeten HXS 1000 PREMIERE von Sulzer HEXIS (kleine BZ) ueber laengere Zeitraeume und bei weitestgehend unterschiedlichen Betriebsweisen durchgefuehrt. In Auswertung dieser Messkomplexe wurden Aussagen zum Betriebsverhalten der Brennstoffzellen, wie z.B. zu Wirk- und Nutzungsgraden, Schadstoffemissionen, Verlauf der Degradation sowie zur Verfuegbarkeit getroffen. Im vorliegenden Bericht wird der erreichte Sachstand detailliert dargestellt und mit den Zielsetzungen des FuE-Antrages verglichen. (orig.)

  2. Hydrogen Fuel System Design Trades for High-Altitude Long-Endurance Remotely- Operated Aircraft

    Millis, Marc G.; Tornabene, Robert T.; Jurns, John M.; Guynn, Mark D.; Tomsik, Thomas M.; VanOverbeke, Thomas J.

    2009-01-01

    Preliminary design trades are presented for liquid hydrogen fuel systems for remotely-operated, high-altitude aircraft that accommodate three different propulsion options: internal combustion engines, and electric motors powered by either polymer electrolyte membrane fuel cells or solid oxide fuel cells. Mission goal is sustained cruise at 60,000 ft altitude, with duration-aloft a key parameter. The subject aircraft specifies an engine power of 143 to 148 hp, gross liftoff weight of 9270 to 9450 lb, payload of 440 lb, and a hydrogen fuel capacity of 2650 to 2755 lb stored in two spherical tanks (8.5 ft inside diameter), each with a dry mass goal of 316 lb. Hydrogen schematics for all three propulsion options are provided. Each employs vacuum-jacketed tanks with multilayer insulation, augmented with a helium pressurant system, and using electric motor driven hydrogen pumps. The most significant schematic differences involve the heat exchangers and hydrogen reclamation equipment. Heat balances indicate that mission durations of 10 to 16 days appear achievable. The dry mass for the hydrogen system is estimated to be 1900 lb, including 645 lb for each tank. This tank mass is roughly twice that of the advanced tanks assumed in the initial conceptual vehicle. Control strategies are not addressed, nor are procedures for filling and draining the tanks.

  3. WWER expert system for fuel failure analysis using the RTOP-CA code

    Likhanskii, V.; Evdokimov, I.; Sorokin, A.; Khromov, A.; Kanukova, V.; Apollonova, O.; Ugryumov, A.

    2008-01-01

    The computer expert system for fuel failure analysis of WWER during operation is presented. The diagnostics is based on the measurement of specific activity of reference nuclides in reactor primary coolant and application of a computer code for the data interpretation. The data analysis includes an evaluation of tramp uranium mass in reactor core, detection of failures by iodine and caesium spikes, evaluation of burnup of defective fuel. Evaluation of defective fuel burnup was carried out by applying the relation of caesium nuclides activity in spikes and relations of activities of gaseous fission products for steady state operational conditions. The method of burnup evaluation of defective fuel by use of fission gas activity is presented in details. The neural-network analysis is performed for determination of failed fuel rod number and defect size. Results of the expert system application are illustrated for several fuel campaigns on operating WWER NPPs. (authors)

  4. An independent spent-fuel storage installation at Surry Station: Design and operation

    McKay, H.S.; Wakeman, B.H.; Pickworth, J.M.; Routh, S.D.; Hopkins, W.C.

    1989-07-01

    Design and licensing of the Surry Power Station Independent Spent Fuel Storage Installation (ISFSI) was initiated in 1982 by Virginia Power as part of a comprehensive strategy to increase spent fuel storage capacity at the Station. Designed to use large, metal dry storage casks, the Surry ISFSI will accommodate 84 such casks with a total storage capacity of 811 MTU of spent PWR fuel assemblies. The ISFSI is located at the Surry Station in a wooded area approximately 1000 meters (3300 feet) east of the reactor facilities. Construction of the first of three reinforced concrete storage pads and its associated support systems was completed in March 1986. The operating license and Technical Specifications were issued by the US NRC on July 2, 1986. Initial loading operations of a General Nuclear Systems, Inc., CASTOR V/21 storage cask began in September 1986. The first two CASTOR V/21 casks were placed in storage at the ISFSI in December 1986. 16 refs., 33 figs., 16 tabs

  5. Operational feedback on internal structure vibration in 54 French PWRs during 300 fuel cycles

    Trenty, A.

    1995-01-01

    EDF has acquired extensive feedback on vibration of reactor vessel internals by analysing ex-core neutron noise on its 54 pressurized water reactors during the course of over 300 fuel cycles. This feedback has been built up by processing more than 3,000 vibratory signatures acquired since the startup of its reactors. These signatures are now centralized for the whole of France in the ''SINBAD'' data base. Signature processing has enabled: distinguishing between mechanical phenomena and signature variation linked to unit operation: in particular, the impact on signature level of unit operating parameters such as initial fuel enrichment and burn-up rate was assessed; among the purely mechanical phenomena, pointing up slight changes in position of vessel internals and the first signs of structural wear; relaxation (in the hold-down spring and fuel rod assemblies) and wear on surfaces of contact between internals and reactor vessel were detected; lastly and most importantly, automatic recognition of the various types of vibratory behavior of internals. It was consequently possible to draw up user requirement specifications for automated monitoring of internals, which should soon be integrated in PSAD, a system which groups several reactor monitoring functions. (author)

  6. Neutron radiography characterization of an operating proton exchange membrane fuel cell with localized current distribution measurements

    Gagliardo, J.J.; Owejan, J.P.; Trabold, T.A.; Tighe, T.W.

    2009-01-01

    Neutron radiography has proven to be a powerful tool to study and understand the effects of liquid water in an operating fuel cell. In the present work, this experimental method is coupled with locally resolved current and ohmic resistance measurements, giving additional insight into water management and fuel cell performance under a variety of conditions. The effects of varying the inlet humidification level and the current density of the 50 cm 2 cell are studied by simultaneously monitoring electrochemical performance with a 10x10 matrix of current sensors, and liquid water volumes are measured using the National Institute of Standards and Technology (NIST) neutron imaging facility. A counter flow, straight channel proton exchange membrane (PEM) fuel cell is used to demonstrate localized performance loss corresponds to water-filled channels that impede gas transport to the catalyst layer, thereby creating an area that has low current density. Furthermore, certain operating conditions causing excess water accumulation in the channels can result in localized proton resistance increase, a result that can only be accurately observed with combined radiography and distributed electrochemical measurements.

  7. A selective electrocatalyst-based direct methanol fuel cell operated at high concentrations of methanol.

    Feng, Yan; Liu, Hui; Yang, Jun

    2017-06-01

    Owing to the serious crossover of methanol from the anode to the cathode through the polymer electrolyte membrane, direct methanol fuel cells (DMFCs) usually use dilute methanol solutions as fuel. However, the use of high-concentration methanol is highly demanded to improve the energy density of a DMFC system. Instead of the conventional strategies (for example, improving the fuel-feed system, membrane development, modification of electrode, and water management), we demonstrate the use of selective electrocatalysts to run a DMFC at high concentrations of methanol. In particular, at an operating temperature of 80°C, the as-fabricated DMFC with core-shell-shell Au@Ag 2 S@Pt nanocomposites at the anode and core-shell Au@Pd nanoparticles at the cathode produces a maximum power density of 89.7 mW cm -2 at a methanol feed concentration of 10 M and maintains good performance at a methanol concentration of up to 15 M. The high selectivity of the electrocatalysts achieved through structural construction accounts for the successful operation of the DMFC at high concentrations of methanol.

  8. A selective electrocatalyst–based direct methanol fuel cell operated at high concentrations of methanol

    Feng, Yan; Liu, Hui; Yang, Jun

    2017-01-01

    Owing to the serious crossover of methanol from the anode to the cathode through the polymer electrolyte membrane, direct methanol fuel cells (DMFCs) usually use dilute methanol solutions as fuel. However, the use of high-concentration methanol is highly demanded to improve the energy density of a DMFC system. Instead of the conventional strategies (for example, improving the fuel-feed system, membrane development, modification of electrode, and water management), we demonstrate the use of selective electrocatalysts to run a DMFC at high concentrations of methanol. In particular, at an operating temperature of 80°C, the as-fabricated DMFC with core-shell-shell Au@Ag2S@Pt nanocomposites at the anode and core-shell Au@Pd nanoparticles at the cathode produces a maximum power density of 89.7 mW cm−2 at a methanol feed concentration of 10 M and maintains good performance at a methanol concentration of up to 15 M. The high selectivity of the electrocatalysts achieved through structural construction accounts for the successful operation of the DMFC at high concentrations of methanol. PMID:28695199

  9. Towards a Future of District Heating Systems with Low-Temperature Operation together with Non-Fossil Fuel Heat Sources

    Tol, Hakan; Dinçer, Ibrahim; Svendsen, Svend

    2012-01-01

    This study focused on investigation of non-fossil fuel heat sources to be supplied to low-energy district heating systems operating in low temperature such as 55 C and 25 C in terms of, respectively, supply and return. Vast variety of heat sources classed in categories such as fossil fuel...

  10. Buckling resistance calculation of Guide Thimbles for the mechanical design of fuel assembly type PWR under normal reactor operating conditions

    Cruz, C.B.L.

    1990-01-01

    The calculations demonstrate the fulfillment of one of the mechanical design criteria for the Fuel Assembly Structure under normal reactor operating conditions. The calculations of stresses in the Guide Thimbles are performed with the aid of the program ANSYS. This paper contains program parameters and modelling of a typical Fuel Assembly for a Reactor similar to ANGRA II. (author)

  11. Dependence of the time-constant of a fuel rod on different design and operational parameters

    Elenkov, D.; Lassmann, K.; Schubert, A.; Laar, J. van de

    2001-01-01

    The temperature response during a reactor shutdown has been measured for many years in the OECD-Halden Project. It has been shown that the complicated shutdown processes can be characterized by a time constant τ which depends on different fuel design and operational parameters, such as fuel geometry, gap size, fill gas pressure and composition, burnup and linear heat rate. In the paper the concept of a time constant is analyzed and the dependence of the time constant on various parameters is investigated analytically. Measured time constants for different designs and conditions are compared with those derived from calculations of the TRANSURANUS code. Employing standard models results in a systematic underprediction of the time constant, i.e. the heat transfer during shutdown is overestimated. (author)

  12. Operating results and simulations on a fuel cell for residential energy systems

    Hamada, Yasuhiro; Goto, Ryuichiro; Nakamura, Makoto; Kubota, Hideki; Ochifuji, Kiyoshi

    2006-01-01

    This paper describes the performance evaluation of a polymer electrolyte fuel cell (PEFC) prototype and demonstration experiments of the electric power and domestic hot water system using it from a pragmatic view-point. Three types of demonstration experiments were carried out applying standard electric power and hot water demands. It was shown that the primary energy reduction rate of this system as compared to the conventional system reached up to 24% under double daily start and stop (DSS) operation. The amount of primary energy reduction in experiments using the energy demand of a household in Sapporo in winter exceeded the experimental results of the standard energy demand, demonstrating that the effects of the introduction of a fuel cell in cold regions could be considerable, in particular, during the winter season

  13. Operational experience in the spent fuel receipt and storage facility at the Tokai Reprocessing Plant

    Nakashima, S.; Yamaguchi, Y.; Iimura, I.; Yamamura, O.; Ogata, Y.

    1992-01-01

    The development of the double containment system led to the reduction of labor time for the cask decontamination to one-tenth compared to the original manner. And also it led to the great decrease of floor contamination in the receipt and storage facility. The decrease permitted as many as about 20,000 visitors to take tours in the fuel receipt and storage facility in the past three years without contamination trouble with the visitors. Different types of spent fuels can be easily handled and stored by the specially designed tools in the pool water. The exchange of the cooling water in the transport cask before unloading and the use of the storage container keep contamination of the pool water to a minimum. The pool water treatment system has been successfully operated. As result, the pool water condition has been well-controlled

  14. International co-operation in the supply of nuclear fuel cycle services

    Allday, C.

    1977-01-01

    The paper draws on B.N.F.L.'s wide experience of international collaboration in nuclear fuel process activities to examine the pros and cons of international agreements. Initially, the factors that influence the need to co-operate, the extent of possible co-operation and the alternative types of agreement are reviewed. Next, the benefits, problems and risks associated with each function, such as managmenet, financial, R and D, marketing and operations that could be covered within the scope of an international agreement, are examined in detail. The paper continues by calling upon specific experience obtained by B.N.F.L. in co-operation with other organisations over several years in operating both major and much smaller agreements illustrating the rationale behind the co-operation, the resolution of 'teething' troubles and the current status of these organisations. In conclusion, the paper comments upon the effectiveness of collaboration agreements and identifies several requirements for internation co-operation to succeed

  15. Strategic optimisation of microgrid by evolving a unitised regenerative fuel cell system operational criterion

    Bhansali, Gaurav; Singh, Bhanu Pratap; Kumar, Rajesh

    2016-09-01

    In this paper, the problem of microgrid optimisation with storage has been addressed in an unaccounted way rather than confining it to loss minimisation. Unitised regenerative fuel cell (URFC) systems have been studied and employed in microgrids to store energy and feed it back into the system when required. A value function-dependent on line losses, URFC system operational cost and stored energy at the end of the day are defined here. The function is highly complex, nonlinear and multi dimensional in nature. Therefore, heuristic optimisation techniques in combination with load flow analysis are used here to resolve the network and time domain complexity related with the problem. Particle swarm optimisation with the forward/backward sweep algorithm ensures optimal operation of microgrid thereby minimising the operational cost of the microgrid. Results are shown and are found to be consistently improving with evolution of the solution strategy.

  16. Cycle to Cycle Variation Study in a Dual Fuel Operated Engine

    Pasunurthi, Shyamsundar

    2017-03-28

    The standard capability of engine experimental studies is that ensemble averaged quantities like in-cylinder pressure from multiple cycles and emissions are reported and the cycle to cycle variation (CCV) of indicated mean effective pressure (IMEP) is captured from many consecutive combustion cycles for each test condition. However, obtaining 3D spatial distribution of all the relevant quantities such as fuel-air mixing, temperature, turbulence levels and emissions from such experiments is a challenging task. Computational Fluid Dynamics (CFD) simulations of engine flow and combustion can be used effectively to visualize such 3D spatial distributions. A dual fuel engine is considered in the current study, with manifold injected natural gas (NG) and direct injected diesel pilot for ignition. Multiple engine cycles in 3D are simulated in series like in the experiments to investigate the potential of high fidelity RANS simulations coupled with detailed chemistry, to accurately predict the CCV. Cycle to cycle variation (CCV) is expected to be due to variabilities in operating and boundary conditions, in-cylinder stratification of diesel and natural gas fuels, variation in in-cylinder turbulence levels and velocity flow-fields. In a previous publication by the authors [1], variabilities in operating and boundary conditions are incorporated into several closed cycle simulations performed in parallel. Stochastic variations/stratifications of fuel-air mixture, turbulence levels, temperature and internal combustion residuals cannot be considered in such closed cycle simulations. In this study, open cycle simulations with port injection of natural gas predicted the combined effect of the stratifications on the CCV of in-cylinder pressure. The predicted Coefficient of Variation (COV) of cylinder pressure is improved compared to the one captured by closed cycle simulations in parallel.

  17. Probability analysis of WWER-1000 fuel elements behavior under steady-state, transient and accident conditions of reactor operation

    Tutnov, A.; Alexeev, E.

    2001-01-01

    'PULSAR-2' and 'PULSAR+' codes make it possible to simulate thermo-mechanical and thermo-physical parameters of WWER fuel elements. The probabilistic approach is used instead of traditional deterministic one to carry out a sensitive study of fuel element behavior under steady-state operation mode. Fuel elements initial parameters are given as a density of the probability distributions. Calculations are provided for all possible combinations of initial data as fuel-cladding gap, fuel density and gas pressure. Dividing values of these parameters to intervals final variants for calculations are obtained . Intervals of permissible fuel-cladding gap size have been divided to 10 equal parts, fuel density and gas pressure - to 5 parts. Probability of each variant realization is determined by multiplying the probabilities of separate parameters, because the tolerances of these parameters are distributed independently. Simulation results are turn out in the probabilistic bar charts. The charts present probability distribution of the changes in fuel outer diameter, hoop stress kinetics and fuel temperature versus irradiation time. A normative safety factor is introduced for control of any criterion realization and for determination of a reserve to the criteria failure. A probabilistic analysis of fuel element behavior under Reactivity Initiating Accident (RIA) is also performed and probability fuel element depressurization under hypothetical RIA is presented

  18. Analysis, operation and maintenance of a fuel cell/battery series-hybrid bus for urban transit applications

    Bubna, Piyush; Brunner, Doug; Gangloff, John J. Jr.; Advani, Suresh G.; Prasad, Ajay K. (Center for Fuel Cell Research, Department of Mechanical Engineering, University of Delaware, Newark, DE 19716 United States)

    2010-06-15

    The fuel cell hybrid bus (FCHB) program was initiated at the University of Delaware in 2005 to demonstrate the viability of fuel cell vehicles for transit applications and to conduct research and development to facilitate the path towards their eventual commercialization. Unlike other fuel cell bus programs, the University of Delaware's FCHB design features a battery-heavy hybrid which offers multiple advantages in terms of cost, performance and durability. The current fuel cell hybrid bus is driven on a regular transit route at the University of Delaware. The paper describes the baseline specifications of the bus with a focus on the fuel cell and the balance of plant. The fuel cell/battery series-hybrid design is well suited for urban transit routes and provides key operational advantages such as hydrogen fuel economy, efficient use of the fuel cell for battery recharging, and regenerative braking. The bus is equipped with a variety of sensors including a custom-designed cell voltage monitoring system which provide a good understanding of bus performance under normal operation. Real-time data collection and analysis have yielded key insights for fuel cell bus design optimization. Results presented here illustrate the complex flow of energy within the various subsystems of the fuel cell hybrid bus. A description of maintenance events has been included to highlight the issues that arise during general operation. The paper also describes several modifications that will facilitate design improvements in future versions of the bus. Overall, the fuel cell hybrid bus demonstrates the viability of fuel cells for urban transit applications in real world conditions. (author)

  19. Analysis, operation and maintenance of a fuel cell/battery series-hybrid bus for urban transit applications

    Bubna, Piyush; Brunner, Doug; Gangloff, John J.; Advani, Suresh G.; Prasad, Ajay K.

    The fuel cell hybrid bus (FCHB) program was initiated at the University of Delaware in 2005 to demonstrate the viability of fuel cell vehicles for transit applications and to conduct research and development to facilitate the path towards their eventual commercialization. Unlike other fuel cell bus programs, the University of Delaware's FCHB design features a battery-heavy hybrid which offers multiple advantages in terms of cost, performance and durability. The current fuel cell hybrid bus is driven on a regular transit route at the University of Delaware. The paper describes the baseline specifications of the bus with a focus on the fuel cell and the balance of plant. The fuel cell/battery series-hybrid design is well suited for urban transit routes and provides key operational advantages such as hydrogen fuel economy, efficient use of the fuel cell for battery recharging, and regenerative braking. The bus is equipped with a variety of sensors including a custom-designed cell voltage monitoring system which provide a good understanding of bus performance under normal operation. Real-time data collection and analysis have yielded key insights for fuel cell bus design optimization. Results presented here illustrate the complex flow of energy within the various subsystems of the fuel cell hybrid bus. A description of maintenance events has been included to highlight the issues that arise during general operation. The paper also describes several modifications that will facilitate design improvements in future versions of the bus. Overall, the fuel cell hybrid bus demonstrates the viability of fuel cells for urban transit applications in real world conditions.

  20. Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

    Rastogi, B.P.

    1989-01-01

    This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)