Fission gas release behaviour in MOX fuels

International Nuclear Information System (INIS)

Viswanathan, U.K.; Anantharaman, S.; Sahoo, K.C.

2002-01-01

As a part of plutonium recycling programme MOX (U,Pu)O 2 fuels will be used in Indian boiling water reactors (BWR) and pressurised heavy water reactors (PHWR). Based on successful test irradiation of MOX fuel in CIRUS reactor, 10 MOX fuel assemblies have been loaded in the BWR of Tarapur Atomic Power Station (TAPS). Some of these MOX fuel assemblies have successfully completed the initial target average burnup of ∼16,000 MWD/T. Enhancing the burnup target of the MOX fuels and increasing loading of MOX fuels in TAPS core will depend on the feedback information generated from the measurement of released fission gases. Fission gas release behaviour has been studied in the experimental MOX fuel elements (UO 2 - 4% PuO 2 ) irradiated in pressurised water loop (PWL) of CIRUS. Eight (8) MOX fuel elements irradiated to an average burnup of ∼16,000 MWD/T have been examined. Some of these fuel elements contained controlled porosity pellets and chamfered pellets. This paper presents the design details of the experimental set up for studying fission gas release behaviour including measurement of gas pressure, void volume and gas composition. The experimental data generated is compared with the prediction of fuel performance modeling codes of PROFESS and GAPCON THERMAL-3. (author)

Experimental irradiation of UMo fuel: Pie results and modeling of fuel behaviour

International Nuclear Information System (INIS)

Languille, A.; Plancq, D.; Huet, F.; Guigon, B.; Lemoine, P.; Sacristan, P.; Hofman, G.; Snelgrove, J.; Rest, J.; Hayes, S.; Meyer, M.; Vacelet, H.; Leborgne, E.; Dassel, G.

2002-01-01

Seven full-sized U Mo plates containing ca. 8 g/cm 3 of uranium in the fuel meat have been irradiated since the beginning of the French U Mo development program. The first three of them with 20% 235 U enrichment were irradiated at maximum surfacic power under 150 W/cm 2 in the OSIRIS reactor up to 50% burn-up and are under examination. Their global behaviour is satisfactory: no failure and a low swelling. The other four plates were irradiated in the HFR Petten at maximum surfacic power between 150 and 250 W/cm 2 with two enrichments 20 and 35%. The experiment was stopped after two cycles due to a fuel failure. The post- irradiation examinations were completed in 2001 in Petten. Examinations showed a correct behaviour of 20% enriched plates and an abnormal behaviour of the two other plates (35%-enriched) with a clad failure on the plate 4. The fuel failure appears to result from a combination of factors that led to high corrosion cladding and high fuel meat temperatures. (author)

Fuel pins irradiation: experimental devices and analytical behaviour

International Nuclear Information System (INIS)

Lemaignan, C.

1996-01-01

In this text we present the general characteristics of adapted irradiation loops in research reactors and the main results that we can expected with these loops in the behaviour field of PWR and LMFBR fuels( fuel densification, fuel cladding interactions, fission products release, reactor accidents)

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

International Nuclear Information System (INIS)

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

INPR ACPR utilization in fuel behaviour studies under accidental condition

International Nuclear Information System (INIS)

Negut, Gheorghe; Popov, Mircea

1990-01-01

This paper is dedicated to the experimental program, investigating CANDU type fuel behaviour in transient condition, as well as the facilities supporting this program. The tests Reactivity Initiated Accident type. The experiments were performed within TRIGA ACPR facility, installed at INSTITUTE for NUCLEAR POWER REACTORS, Pitesti, ROMANIA. Studies of the safety issues took a great international developement during last years. In USA, Japan, owners of the similar reactors, and USSR there are a big commitment to such programs, intended to establish the nuclear fuel behaviour under RIA-conditions. In our country, too, there are programs aiming a complete testing of the CANDU type fuels. As it is known, RIA is not a CANDU specific accident, but the fuel behaviour in such conditions can give useful informations on the fuel cladding failure threshold and about reflooding post LOCA heat transfer condition. Based on some papers and specific requirements it was initiated and developed a safety research program on CANDU type fuel using the ACPR. The paper describes the reactor,test capsule, instrumentation, fuel samples, tests, post irradiation results. (orig.)

Coupling of channel thermalhydraulics and fuel behaviour in ACR-1000 safety analyses

International Nuclear Information System (INIS)

Huang, F.L.; Lei, Q.M.; Zhu, W.; Bilanovic, Z.

2008-01-01

Channel thermalhydraulics and fuel thermal-mechanical behaviour are interlinked. This paper describes a channel thermalhydraulics and fuel behaviour coupling methodology that has been used in ACR-1000 safety analyses. The coupling is done for all 12 fuel bundles in a fuel channel using the channel thermalhydraulics code CATHENA MOD-3.5d/Rev 2 and the transient fuel behaviour code ELOCA 2.2. The coupling approach can be used for every fuel element or every group of fuel elements in the channel. Test cases are presented where a total of 108 fuel element models are set up to allow a full coupling between channel thermalhydraulics and detailed fuel analysis for a channel containing a string of 12 fuel bundles. An additional advantage of this coupling approach is that there is no need for a separate detailed fuel analysis because the coupling analysis, once done, provides detailed calculations for the fuel channel (fuel bundles, pressure tube, and calandria tube) as well as all the fuel elements (or element groups) in the channel. (author)

Wi-Fi Salvador: mapeamento colaborativo e redes sem fio no Brasil

OpenAIRE

Lemos, André; Pastor, Leonardo; Oliveira, Nelson

2012-01-01

O trabalho faz uma breve demonstração de políticas públicas para inclusão digital por meio da disponibilização do acesso à internet por meio de redes sem fio e discute a forma como as pessoas se relacionam com os lugares, de acordo com a existência ou inexistência de acessibilidade à internet. O objetivo é discutir o acesso à internet pelas redes sem fio no Brasil e, mais especificamente, na cidade de Salvador. Este trabalho é fruto de pesquisa realizada no âmbito do Grupo de Pesquisa em Cibe...

Modelling of phenomena associated with high burnup fuel behaviour during overpower transients

International Nuclear Information System (INIS)

Sills, H.E.; Langman, V.J.; Iglesias, F.C.

1995-01-01

Phenomena of importance to the behaviour of high burnup fuel subjected to conditions of rapid overpower (i.e., LWR RIAs) include the change in cladding material properties due to irradiation, pellet-clad interaction (PCI) and 'rim' effects associated with the periphery of high burnup fuel. 'Rim' effects are postulated to be caused by changes in fuel morphology at high burnup. Typical discharge burnups for CANDU fuel are low compared to LWRs. Maximum linear ratings for CANDU fuel are higher than those for LWRs. However, under normal operating conditions, the Zircaloy-4 clad of the CANDU fuel is collapsed onto the fuel stack. Thus, the CANDU fuel performance codes model the transient behaviour of the fuel-to-clad interface and are capable of assessing the potential for pellet-clad mechanical interaction (PCMI) failures for a wide range of overpower conditions. This report provides a discussion of the modelling of the phenomena of importance to high burnup fuel behaviour during rapid overpower transients. (author)

Characteristics and behaviour of the PHENIX fuel element

International Nuclear Information System (INIS)

Delpeyroux, P.; Balloffet, Y.; Blanchard, P.; Courcon, P.; Jallade, M.; Millet, P.; Rousseau, J.; Carteret, Y.; Coulon, P.

1977-01-01

The Phenix reactor has been in regular industrial operation for two years and has functioned very satisfactorily thanks in particular to the very good behaviour of the fuel element. A brief description is given of the fuel element and the operating conditions which were set for the fuel at the time of start-up (50000 MWd/t). The surveillance scheme is then described with the examinations in the hot laboratory on the basis of which it was possible to achieve the nominal specific burn-up and then to clear the Phenix fuel for a specific burn-up of 60000 MWd/t or 7 at.%. The behaviour of the mixed oxide (U, Pu)O 2 is quite normal and conforms to predictions as regards the heat conditions, swelling and fission gas release. The corrosion reaction between the oxide and the clad is progressing slowly and affects only small thicknesses of cladding. The mechanical integrity of the clad under thermal stresses and the stresses produced by swelling and fission gas pressure do not pose any special problem. The present limitation of the irradiation level is essentially based on the permissible deformations due to swelling and irradiation creep in the fuel pin cladding and in the hexagonal tube. This corresponds to damage to the steel of the order of 80 dpa. The mechanical behaviour of the bundle of pins, its interaction with the hexagonal tube and the thermohydraulic consequences of the deformations are all satisfactory to date. The absence of fuel failures is also worth noting; the only burst can detected to date did not affect either the operation of the fuel assembly or the performance of the reactor [fr

Corrosion behaviour of zircaloy 4 fuel rod cladding in EDF power plants

Energy Technology Data Exchange (ETDEWEB)

Romary, H; Deydier, D [EDF, Direction de l` Equipment SEPTEN, Villeurbanne (France)

1997-02-01

Since the beginning of the French nuclear program, a surveillance of fuel has been carried out in order to evaluate the fuel behaviour under irradiation. Until now, nuclear fuels provided by suppliers have met EDF requirements concerning fuel behaviour and reliability. But, the need to minimize the costs and to increase the flexibility of the power plants led EDF to the definition of new targets: optimization of the core management and fuel cycle economy. The fuel behaviour experience shows that some of these new requirements cannot be fully fulfilled by the present standard fuel due to some technological limits. Particularly, burnup enhancement is limited by the oxidation and the hydriding of the Zircaloy 4 fuel rod cladding. Also, fuel suppliers and EDF need to have a better knowledge of the Zy-4 cladding behaviour in order to define the existing margins and the limiting factors. For this reason, in-reactor fuel characterization programs have been set up by fuel suppliers and EDF for a few years. This paper presents the main results and conclusions of EDF experience on Zy-4 in-reactor corrosion behaviour. Data obtained from oxide layer or zirconia thickness measurements show that corrosion performance of Zy-4 fuel rod cladding, as irradiated until now in EDF reactors, is satisfactory but not sufficient to meet the future needs. The fuel suppliers propose in order to improve the corrosion resistance of fuel rod cladding, low tin Zy-4 cladding and then optimized Zy-4 cladding. Irradiation of these claddings are ongoing. The available corrosion data show the better in-reactor corrosion resistance of optimized Zy-4 fuel rod cladding compared to the standard Zy-4 cladding. The scheduled fuel surveillance program will confirm if the optimized Zy-4 fuel rod cladding will meet the requirements for the future high burnup and high flexibility fuel. (author). 10 refs, 19 figs, 4 tabs.

Graded algebras of the second rank and integration of nonlinear equations Ysub(z)sub(z) = exp(2Y) - exp(-2Y), Ysub(z)sub(z) = 2 exp(Y) - exp(-2Y)

International Nuclear Information System (INIS)

Leznov, A.N.; Smirnov, V.G.

1981-01-01

In the terms of the notions of the theory of infinite-dimensional algebras of finite growth of the second rank, we have derived solutions to the equations Ysub(z)sub(z) = exp(2Y) - exp(-2Y); Ysub(z)sub(z) = 2 exp(Y) - exp(-2Y) dependent on two arbitrary functions. (orig.)

Behaviour of gas cooled reactor fuel under accident conditions

International Nuclear Information System (INIS)

1991-11-01

The Specialists Meeting on Behaviour of Gas Cooled Reactor Fuel under Accident Conditions was convened by the International Atomic Energy Agency on the recommendation of the International Working Group on Gas Cooled Reactors. The purpose of the meeting was to provide an international forum for the review of the development status and for the discussion on the behaviour of gas cooled reactor fuel under accident conditions and to identify areas in which additional research and development are still needed and where international co-operation would be beneficial for all involved parties. The meeting was attended by 45 participants from France, Germany, Japan, Switzerland, the Union of Soviet Socialists Republics, the United Kingdom, the United States of America, CEC and the IAEA. The meeting was subdivided into five technical sessions: Summary of Current Research and Development Programmes for Fuel; Fuel Manufacture and Quality Control; Safety Requirements; Modelling of Fission Product Release - Part I and Part II; Irradiation Testing/Operational Experience with Fuel Elements; Behaviour at Depressurization, Core Heat-up, Power Transients; Water/Steam Ingress - Part I and Part II. 22 papers were presented. A separate abstract was prepared for each of these papers. At the end of the meeting a round table discussion was held on Directions for Future R and D Work and International Co-operation. Refs, figs and tabs

Behaviour of high O/U fuel

International Nuclear Information System (INIS)

Davies, J.H.; Hoshi, E.V.; Zimmerman, D.L.

2000-01-01

Full text: The effect of increased fuel oxygen potential on fuel behaviour has been studied by fabricating and irradiating urania fuel with an average O/U ratio of 2.05. The fuel was fabricated by re-sintering standard urania pellets in a controlled oxygen potential environment and irradiated in a segmented rod bundle in a U.S. BWR. Preirradiation ceramographic characterization of the pellets revealed the well-known Widmanstaetten precipitation of U-409 platelets in the UO 2 matrix. The high O/U fuel pellets were clad in Zircaloy-2 and irradiated to over 20 GWd/MT. Ramp tests were performed in a test reactor and detailed postirradiation examinations of both ramped and nonramped rods have been performed. The cladding inner surface condition, fission gas release and swelling behavior of high O/U fuel have been characterized and compared with standard UO 2 pellets. Although fuel microstructural features in ramp-tested high O/U fuel showed evidence of higher fuel temperatures and/or enhanced transport processes, fission gas release to the fuel rod free space was less than for similarly tested standard UO 2 fuel. However, fuel swelling and cladding strains were significantly greater. In spite of high cladding strains, PCI crack propagation was inhibited in the high O/U fuel I rods. Evidence is presented that the crystallographically oriented etch features often noted in peripheral regions of high burnup fuels are not an indication of higher oxides of uranium. (author)

Mathematical model of thermal and mechanical steady state fuel element behaviour TEDEF

International Nuclear Information System (INIS)

Dinic, N.; Kostic, Z.; Josipovic, M.

1987-01-01

In this paper a numerical model of thermal and thermomechanical behaviour of a cylindrical metal uranium fuel element is described. Presented are numerical method and computer program for solving the stationary temperature field and thermal stresses of a fuel element. The model development is a second phase of analysis of these phenomena, and may as well be used for analysing power nuclear reactor fuel elements behaviour. (author)

Mechanical behaviour of PWR fuel rods during intermediate storage

International Nuclear Information System (INIS)

Bouffioux, P.; Dalmas, R.; Bernaudat, C.

2000-01-01

EDF, which owns the irradiated fuel coming from its NPPs, has initiated studies regarding the mechanical behaviour of a fuel rod and the integrity of its cladding, in the case where the spent fuel is stored for a significant duration. During the phases following in-reactor irradiation (ageing in a water-pool, transport and intermediate storage), many phenomena, which are strongly coupled, may influence the cladding integrity: - residual power and temperature decay; - helium production and release in the free volume of the rod (especially for MOX fuel); - fuel column swelling; - cladding creep-out under the inner gas pressure of the fuel rod; - metallurgical changes due to high temperatures during transportation. In parallel, the quantification of the radiological risk is based on the definition of a cladding integrity criterion. Up to now, this criterion requires that the clad hoop strain due to creep-out does not exceed 1%. A more accurate criterion is being investigated. The study and modelling of all the phenomena mentioned above are included in a R and D programme. This programme also aims at redefining the cladding integrity criterion, which is assumed to be too conservative. The R and D programme will be presented. In order to predict the overall behaviour of the rod during the intermediate storage phases, the AVACYC code has been developed. It includes the models developed in the R and D programme. The input data of the AVACYC code are provided by the results of in-reactor rod behaviour simulations, using the thermal-mechanical CYRANO3 code. Its main results are the evolution vs. time of hoop stresses in the cladding, rod internal pressure and cladding hoop strains. Chained CYRANO-AVACYC calculations have been used to simulate the behaviour of MOX fuel rods irradiated up to 40 GWd/t and stored under air during 100 years, or under water during 50 years. For such fuels, where the residual power remains high, we show that a large part of the cladding strain

Experimental program on fuel rod behaviour under off-normal conditions

International Nuclear Information System (INIS)

Languille, A.; Cecchi, P.

1985-01-01

During LMFBR plant operation, fuel developments are primarily concerned with the fuel pin irradiation behaviour under steady-state conditions up to high burn-up levels. But additional studies under off-normal conditions are necessary in order to assess fuel pin performance and to define operational limits. (author)

Researches of WWER fuel rods behaviour under RIA accident conditions

International Nuclear Information System (INIS)

Nechaeva, O.; Medvedev, A.; Novikov, V.; Salatov, A.

2003-01-01

Unirradiated fuel rod and refabricated fuel rod tests in the BIGR as well as acceptance criteria proving absence of fragmentation and the settlement modeling of refabricated fuel rods thermomechanical behavior in the BIGR-tests using RAPTA-5 code are discussed in this paper. The behaviour of WWER type simulators with E110 and E635 cladding was researched at the BIGR reactor under power pulse conditions simulating reactivity initiated accident. The results of the tests in four variants of experimental conditions are submitted. The behaviour of 12 WWER type refabricated fuel rods was researched in the BIGR reactor under power pulse conditions simulating reactivity initiated accident: burnup 48 and 60 MWd/kgU, pulse width 3 ms, peak fuel enthalpy 115-190 cal/g. The program of future tests in the research reactor MIR with high burnup fuel rod (up to 70 MWd/kgU) under conditions simulating design RIA in WWER-1000 is presented

The modeling of fuel rod behaviour under RIA conditions in the code DYN3D

International Nuclear Information System (INIS)

Rohde, U.

1998-01-01

A description of the fuel rod behaviour and heat transfer model used in the code DYN3D for nuclear reactor core dynamic simulations is given. Besides the solution of heat conduction equations in fuel and cladding, the model comprises detailed description of heat transfer in the gas gap by conduction, radiation and fuel-cladding contact. The gas gap behaviour is modeled in a mechanistic way taking into account transient changes of the gas gap parameters based on given conditions for the initial state. Thermal, elastic and plastic deformations of fuel and cladding are taken into account within 1D approximation. Numerical studies concerning the fuel rod behaviour under RIA conditions in power reactors are reported. Fuel rod behaviour at high pressures and flow rates in power reactors is different from the behaviour under atmospheric pressure and stagnant flow conditions in the experiments. The mechanisms of fuel rod failure for fresh and burned fuel reported from the literature can be qualitatively reproduced by the DYN3D model. (author)

Defect trap model of gas behaviour in UO2 fuel during irradiation

International Nuclear Information System (INIS)

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)

ELOCA: fuel element behaviour during high temperature transients

International Nuclear Information System (INIS)

Sills, H.E.

1979-03-01

The ELOCA computer code was developed to simulate the uniform thermal-mechanical behaviour of a fuel element during high-temperature transients such as a loss-of-coolant accident (LOCA). Primary emphasis is on the diametral expansion of the fuel sheath. The model assumed is a single UO2/zircaloy-clad element with axisymmetric properties. Physical effects considered by the code are fuel expansion, cracking and melting; variation, during the transient, of internal gas pressure; changing fuel/sheath heat transfer; thermal, elastic and plastic sheath deformation (anisotropic); Zr/H 2 O chemical reaction effects; and beryllium-assisted crack penetration of the sheath. (author)

A study on dissolution and leaching behaviour of spent nuclear fuels

International Nuclear Information System (INIS)

Lee, Chang Heon; Im, Hee Jung; Kim, Jong Gu; Park, Yang Soon; Ha, Yeong Keong

2010-12-01

This state of the art report describes a leaching behaviour of spent nuclear fuels which should be considered for safety assessment of spent nuclear fuel disposal in a deep geological repository. A decisive factor of a dissolution of UO 2 , a matrix of the fuel, is chemical characters (redox potential, pH, concentration of inorganic anions, water radiolysis subsequent by radiation field of the fuels) of ground water expected to be in contact with the fuels after the container has failed due to corrosion as well as atmosphere condition of a deep geological repository, which can change the oxidation state of UO 2 . The release rates of radionuclides from UO 2 matrix depend largely on their location within the fuels, that is, the radionuclides fixed in the fuel/cladding gap and grain boundaries are rapidly released. However, the radionuclides within the grains of the fuel are slowly released, and then their release rate is governed by a dissolution behaviour of UO 2

BWR fuel clad behaviour following LOCA

International Nuclear Information System (INIS)

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

1996-01-01

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

Fuel performance and fission product behaviour in gas cooled reactors

International Nuclear Information System (INIS)

1997-11-01

The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

Investigation of the ramp testing behaviour of fuel pins with different diameters

International Nuclear Information System (INIS)

Pott, G.; Herren, M.; Wigger, B.

1979-09-01

The aim of these experiments was the investigation of the influence of different fuel pin diameter on the ramp testing behaviour. Fuel elements with diameter between 10,75 and 15,6 mm and different cladding thickness had been ramptested in the HBWR (Halden Boiling Water Reactor) after preirradiated in the same facility. Fuel pins with the smallest diameter of 10,75 mm failed. This was indicated by fission gas release measurement. Metallographic examination showed these failure were caused by hydride blisters. A systematic influence of fuel pin diameter and cladding thickness on the ramptesting behaviour was not observed. (orig.) [de

The modeling of fuel rod behaviour under RIA conditions in the code DYN3D

International Nuclear Information System (INIS)

Rohde, U.

2001-01-01

A description of the fuel rod behaviour and heat transfer model used in the code DYN3D for nuclear reactor core dynamic simulations is given. Besides the solution of heat conduction equations in fuel and cladding, the model comprises a detailed description of heat transfer in the gas gap by conduction, radiation and fuel-cladding contact. The gas gap behaviour is modeled in a mechanistic way taking into account transient changes of the gas gap parameters based on given conditions for the initial state. Thermal, elastic and plastic deformations of fuel and cladding are taken into account within 1D approximation. A creeping law for time-dependent estimation of plastic deformations is implemented. Metal-water reaction of the cladding material in the high temperature region is considered. The cladding-coolant heat transfer regime map covers the region from one-phase liquid convection to dispersed flow with superheated steam. Special emphasis is put on taking into account the impact of thermodynamic non-equilibrium conditions on heat transfer. For the validation of the model, experiments on fuel rod behaviour during RIAs carried out in Russian and Japanese pulsed research reactors with shortened probes of fresh fuel rods are calculated. Comparisons between calculated and measured results are shown and discussed. It is shown, that the fuel rod behaviour is significantly influenced by plastic deformation of the cladding, post crisis heat transfer with sub-cooled liquid conditions and heat release from the metal-water reaction. Numerical studies concerning the fuel rod behaviour under RIA conditions in power reactors are reported on. It is demonstrated, that the fuel rod behaviour at high pressures and flow rates in power reactors is different from the behaviour under atmospheric pressure and stagnant flow conditions in the experiments. The mechanisms of fuel rod failure for fresh and burned fuel reported from the literature can be qualitatively reproduced by the DYN3D

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

International Nuclear Information System (INIS)

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

Non-linear behaviour of multi-phase MOX fuels: a micro-mechanical approach

International Nuclear Information System (INIS)

Rousette, S.; Gatt, J.M.; Michel, J.C.

2005-01-01

The modelling of mechanical pellet-clad interaction requires knowledge of the thermo-mechanical behaviour of nuclear fuels. Some nuclear fuels such as MOX are composed of several phases. The mechanical properties of these phases, which are elasto-visco-plastic in-pile, are changing in-pile. The objective is to formulate a mechanical behaviour law taking all the physical phenomena into account in the different phases, which can easily be introduced into a fuel rod modelling code. Consequently, Non-uniform Transformation Field Analysis (NTFA) is used on the one hand, to correctly capture the heterogeneity of the anelastic strain in the different phases and, on the other hand, to provide a simple overall constitutive law for computational codes. This method is a good way to describe the behaviour of MOX fuel. Transformation Field Analysis (TFA), which corresponds to piecewise uniform transformation fields, is used to perform a sensitivity study. (authors)

Fuel performance and fission product behaviour in gas cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-11-01

The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

Accuracy of an indirect calorimeter for mechanically ventilated infants and children: the influence of low rates of gas exchange and varying FIO2.

Science.gov (United States)

Joosten, K F; Jacobs, F I; van Klaarwater, E; Baartmans, M G; Hop, W C; Meriläinen, P T; Hazelzet, J A

2000-08-01

To test the accuracy and validity of the Deltatrac II MBM-200 metabolic monitor for use in mechanically ventilated infants and children in the pediatric intensive care unit. Laboratory validation of an indirect calorimeter with a ventilated lung model. The influence of low tidal volumes and low levels of oxygen consumption (V(O2)) and carbon dioxide production (V(CO2)) in combination with different levels of inspired oxygen concentrations (F(IO2)) was investigated. University research laboratory. Low tidal volumes were provided with two intermittent flow types of ventilators, a Servo 300 and a Servo 900C. A butane flame with a V(O2) approximating 20 mL/min and 40 mL/min was ventilated. To investigate the effect of different levels of F(IO2) on the accuracy of V(O2), V(CO2), and respiratory quotient (RQ), measurements were performed at F(IO2) target values of 0.25, 0.40, and 0.60. No significant differences were found between the ventilators regarding V(O2), V(CO2), and RQ measurements. The mean deviation of V(O2) increased significantly with increasing F(IO2) to -7.98% with a V(O2) of 21.0 mL/min and to -8.46% with a V(O2) of 38.9 mL/min (F(IO2), 0.558) with a variability (2 SD) of +/- 4.86% and +/- 6.82%, respectively. The mean deviation and variability of V(CO2) in all tests remained within 8%. The mean deviation of RQ increased significantly with increasing F(IO2) to 5.5% with a V(O2) of 21.0 mL/min and to 5.69% with a V(O2) of 38.9 mL/min (F(IO2), 0.558) with a variability (2 SD) of +/- 5.62% and +/- 5.76%, respectively. The minute to minute delivered F(IO2) fluctuated significantly when increasing the level of F(IO2). The Deltatrac II MBM-200 metabolic monitor appears accurate for low levels of V(O2) and V(CO2) during mechanical ventilation with F(IO2) levels up to 0.390. With increasing F(IO2) to 0.558, the increase in deviation of V(O2) for single measurements can be of clinical relevance for mechanically ventilated infants and children. The increased

Modeling of the thermo-mechanical behaviour of the PWR fuel

International Nuclear Information System (INIS)

Mailhe, P.

2014-01-01

This article reviews the various physical phenomena that take place in an irradiated fuel rod and presents the development of the thermo-mechanical codes able to simulate them. Though technically simple the fuel rod is the place where appear 4 types of process: thermal, gas behaviour, mechanical and corrosion that combine involving 5 elements: the fuel pellet, the fuel clad, the fuel-clad gap, the inside volume and the coolant. For instance the pellet is the place where the following mechanical processes took place: thermal dilatation, elastic deformation, creep deformation, densification, solid swelling, gaseous swelling and cracking. The first industrial code simulating the behaviour of the fuel rod was COCCINEL, it was developed by AREVA teams from the American PAD code that was included in the Westinghouse license. Today the GALILEO code has replaced the COPERNIC code that was developed in the beginning of the 2000 years. GALILEO is a synthesis of the state of the art of the different models used in the codes validated for PWR and BWR. GALILEO has been validated on more than 1500 fuel rods concerning PWR, BWR and specific reactors like Siloe, Osiris, HFR, Halden, Studsvik, BR2/3,...) and also for extended burn-ups. (A.C.)

Effects of salvage logging and pile-and-burn on fuel loading, potential fire behaviour, fuel consumption and emissions

Science.gov (United States)

Morris C. Johnson; Jessica E. Halofsky; David L. Peterson

2013-01-01

We used a combination of field measurements and simulation modelling to quantify the effects of salvage logging, and a combination of salvage logging and pile-and-burn fuel surface fuel treatment (treatment combination), on fuel loadings, fire behaviour, fuel consumption and pollutant emissions at three points in time: post-windstorm (before salvage logging), post-...

Utilidad de la relación Sao2/Fio2 en la evaluación del grado de compromiso pulmonar en pacientes críticos

Directory of Open Access Journals (Sweden)

Alonso Gómez Duque

2002-01-01

Full Text Available Background: Arteriovenous admixture or intrapulmonary shunt and the relation between PaO2 and FI02 (PaO2/FiO2 have been advocated as the main indicators of pulmonary function. Because the Pa02 is a measure of the oxygen given to the tissues, as well as the Sa02, we could expect a close relationship between the Pa02IFI02 index and a new one, the SaO2/FiO2 index, Methods: We conducted a prospective study in a cohortofl07 patients with different pathologies in an Intensive Care Unit, collecting 507 samples of arterial and central venous gases and measuring the Sa02 by pulse oximeter at the same time. PaO2/FiO2 and Sa O2/FiO2, as well as intrapulmonary shunt using invasive and noninvasive methods, were caIculated. The Lung Injury Score (LIS and the Multiple Organ Dysfunction score were calculated too. Results: We found a good correlation between Sa O2/FiO2 and PaO2/FiO2. (R2=0,81,R=0.9, and between the intrapulmonary shuntfraction calculated with invasive and noninvasive methods.(R2=0.75, R=0.86. The data were correlationed with a simple linear regression analysis. Five groups of SaO2/FiO2 were found and a punctuation from Oto 4 was established for them, in a similar approach to the one used by Murray in the Lung Injury Score (LIS and Marshall in the Multiple Organ Dysfunction score. By using this punctuation, we replaced the PaO2/FiO2 by the SaO2/FiO2 in the scale and recalculated the LIS finding a good correlation between the score with PaO2/FiO2 and the one with SaO2/FiO2 (R2=0.94,R=0.96. The same was true for the correlation between the Marshall score with PaO2/FiO2 and the one with SaO2/FiO2. (R2=O.85, R=0.92. The correlation was mantained when a stratified analysis was made according to the degree of severity of the pulmonary injury. Conclusions: SaO2/FiO2 is an useful indicator of the oxygenation in a similar way as the PaO2/FiO2 índex, It could be incorporated into the LIS and Marshall score with similar results. The usefulness of the

Molten fuel-coolant interaction behaviours of various fast reactor fuels (Paper No. HMT-45-87)

International Nuclear Information System (INIS)

Doshi, J.B.

1987-01-01

A parametric computational model of molten fuel-coolant interaction (MFCI) including a particle size distribution is developed and employed to analyse behaviours of various possible reactor fuels, such as oxide, carbide and metal in MFCI scenario. It is observed that while higher thermal conductivity and lower specific heat of carbide compared to oxide is responsible for higher peak pressure and work done per unit mass, the trend is not observed in the metal fuel. The reason for this is the lower operation temperature and latent heat of metallic fuel. (author). 9 refs., 1 fig

Fuel behaviour calculations with version 2.0 of the code FUROM

International Nuclear Information System (INIS)

Kulacsy, K.

2011-01-01

The fuel modelling code FUROM (FUel ROd Model), suitable for calculating the normal operation condition behaviour of PWR and WWER fuels, has been developed at AEKI for several years. In 2010 the new version of the code, FUROM-2.0 was released. Calculations performed with this version and results are presented. (author)

Predicting the behaviour or neptunium during nuclear fuel reprocessing

International Nuclear Information System (INIS)

Drake, V.A.

1988-01-01

Behaviour of Np and its distribution over reprocessing flowsheet is studied due to the necessity of improvement of reprocessing methods of wastes formed during purex-process. Valency states of Np in solutions of reprocessing cycles, Np distribution in organic and acid phases, Np(5) oxidation by nitric acid at the stage of extraction, effect of U and Pu presence on Np behaviour, are considered. Calculation and experimental data are compared; the possibility of Np behaviour forecasting in the process of nuclear fuel reprocessing, provided initial data vay, is shown. 7 refs.; 4 figs.; 1 tab

Cermet fuel behaviour and properties in ADS reactors

International Nuclear Information System (INIS)

Haas, D.; Fernandez, A.; Staicu, D.; Somers, J.; Maschek, W.; Chen, X.

2007-01-01

Within the EUROTRANS Integrated Project co- financed within the 6th Framework Programme of the European commission, the sub-critical Accelerator Driven System (ADS) is now being considered as a potential means to burn long-lived transuranium nuclides. Within the EUROTRANS Programme, the domain AFTRA is responsible to develop and provide the data basis for the fuels to be used in the European Facility for Industrial Transmutation (EFIT). The preferred fuel for such a fast neutron reactor is uranium-free, highly enriched with plutonium and minor actinides. Requirements for ADS transmuter fuels are strongly linked with the core design and safety parameters, the fuel properties and the ease of fabrication and reprocessing. This study concerns the behaviour and properties of fuels with molybdenum as inert matrix. The status of the development work was presented at the last ICENES conference [1]. Since then, the design of the European Facility for Industrial Transmutation (EFIT) was developed and the transmutation capability, the burn-up behaviour, the reactivity swing and power peaking factors, and the safety performance were determined for different cores with inert matrix fuels like MgO and Mo. For the EFIT, the CERMET with the Mo matrix is recommended as the reference fuel and CERCER with the MgO matrix as a back-up solution. The thermal diffusivity and specific heat of the CERMET fuels (loaded with Pu and Am) were measured, and the thermal conductivity was deduced. The thermal conductivity of the CERMET fuels was also predicted using a model proposed in [1], with a microstructure corresponding to a random distribution of spheres, with overlapping. This model microstructure takes into account the negative effects arising from the possible formation of small agglomerates of inclusions in the CERMET fuels. The agreement between the theoretical and calculated values is relatively good (the error is between 0 and 15% of the value of the thermal conductivity

Behaviour of irradiated uranium silicide fuel revisited

International Nuclear Information System (INIS)

Finlay, M. Ross; Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

2002-01-01

Irradiated U 3 Si 2 dispersion fuels demonstrate very low levels of swelling, even at extremely high burn-up. This behaviour is attributed to the stability of fission gas bubbles that develop during irradiation. The bubbles remain uniformly distributed throughout the fuel and show no obvious signs of coalescence. Close examination of high burn-up samples during the U 3 Si 2 qualification program revealed a bimodal distribution of fission gas bubbles. Those observations suggested that an underlying microstructure was responsible for the behaviour. An irradiation induced recrystallisation model was developed that relied on the presence of sufficient grain boundary surface to trap and pin fission gas bubbles and prevent coalescence. However, more recent work has revealed that the U 3 Si 2 becomes amorphous almost instantaneously upon irradiation. Consequently, the recrystallisation model does not adequately explain the nucleation and growth of fission gas bubbles in U 3 Si 2 . Whilst it appears to work well within the range of measured data, it cannot be relied on to extrapolate beyond that range since it is not mechanistically valid. A review of the mini-plates irradiated in the Oak Ridge Research Reactor from the U 3 Si 2 qualification program has been performed. This has yielded a new understanding of U 3 Si 2 behaviour under irradiation. (author)

The decay of $]^{119}$Cd and $\\^{119}$In isomers

CERN Document Server

Scheidemann, O; Patzelt, P

1976-01-01

The decay of /sup 119g, m/Cd and /sup 119g, m/In has been investigated using isotopically separated samples produced by the ISOLDE facility at CERN. The half-lives are 2.69+or-0.02 min, 2.20+or-0.02 min, 2.4+or-0.1 min and 18.0+or-0.3 min, respectively. A total of 33 excited levels have been found in /sup 119/In and 4 excited levels have been found in /sup 119/Sn. The percentage isomeric transition in /sup 119/In has been measured to be 2.5/sub -0.3//sup +0.5/%. The possibility of rotational and vibrational levels in /sup 119/In is discussed. (32 refs).

Molten fuel behaviour during slow overpower transients

International Nuclear Information System (INIS)

Guerin, Y.; Boidron, M.

1985-01-01

In large commercial reactors as Super-Phenix, if we take into account all the uncertainties on the pins and on the core, it is no longer possible to guarantee the absence of fuel melting during incidental events such as slow overpower transients. We have then to explain what happens in the pins when fuel melting occurs and to demonstrate that a limited amount of molten fuel generates no risk of clad failure. For that purpose, we may use the results of a great number of experiments (about 40) that have been performed at C.E.A., most of them in thermal reactor, but some experiments have also been performed in Rapsodie, especially during the last run of this reactor. In a great part of these experiments, fuel melting occurred at beginning of life, but we have also some results at different burnups up to 5 at %. It is not the aim of this paper to describe all these experiments and the results of their post irradiation examination, but to summarize the main conclusions that have been set out of them and that have enabled us to determine the main characteristics of fuel element behaviour when fuel melting occurs

Impact of fission gas on irradiated PWR fuel behaviour at extended burnup under RIA conditions

International Nuclear Information System (INIS)

Lemoine, F.; Schmitz, F.

1996-01-01

With the world-wide trend to increase the fuel burnup at discharge of the LWRs, the reliability of high burnup fuel must be proven, including its behaviour under energetic transient conditions, and in particular during RIAs. Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup. The potential for swelling and transient expansion work under rapid heating conditions characterizes the high burnup fuel behaviour by comparison to fresh fuel. This effect is resulting from the steadily increasing amount of gaseous and volatile fission products retained inside the fuel structure. An attempt is presented to quantify the gas behaviour which is motivated by the results from the global tests both in CABRI and in NSRR. A coherent understanding of specific results, either transient release or post transient residual retention has been reached. The early failure of REP Na1 with consideration given to the satisfactory behaviour of the father rod of the test pin at the end of the irradiation (under load follow conditions) is to be explained both by the transient loading from gas driven fuel swelling and from the reduced clad resistance due to hydriding. (R.P.)

Fast reactor fuel pin behaviour modelling in the UK

Energy Technology Data Exchange (ETDEWEB)

Matthews, J R [UKAEA, Harwell, Didcot, Oxon (United Kingdom); Hughes, H [Springfields Nuclear Power Development Laboratories, Springfields, Salwick, Preston (United Kingdom)

1979-12-01

Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients

Fast reactor fuel pin behaviour modelling in the UK

International Nuclear Information System (INIS)

Matthews, J.R.; Hughes, H.

1979-01-01

Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients

Studies in Phebus reactor of fuel behaviour upon LOCA conditions

International Nuclear Information System (INIS)

Manin, A.; Del Negro, R.; Reocreux, M.

1980-09-01

The fuel behaviour upon LOCA conditions is studied in an in-pile loop, in Phebus reactor. This paper presents: a short description of Phebus reactor; the current program (adjusting the thermohydraulic conditions in order to get cladding failure); the program developments (consequences involved by cladding failure); the fuel test conditions determination [fr

Parametric study of fuel rod behaviour during the RIA using the modified FALCON code

International Nuclear Information System (INIS)

Khvostov, G.; Zimmermann, M.A.; Ledergerber, G.

2010-01-01

Presented in the paper are the results of a parametric study with the use of optimised modules of the FALCON code (FALCON-PSI) that addresses the effects of the selected characteristics of fast thermal transients (e.g., impulse width), fuel rod design (e.g., active fuel attack length) and boundary conditions (e.g., the coolant conditions) on fuel behaviour during a RIA. Specifically, the analysis of the governing processes for the fuel rod behaviour during the RIA events simulated in the experimental facility of the Nuclear Safety Research Reactor (NSRR, Japan) are in the focus of the present study. The results obtained can be useful for a better transfer of the NSRR test results in relation to the corresponding behaviour in LWRs and furthermore might also support the planning of future additional experiments. (authors)

Fuel Behaviour Simulations in Fumex III CRP at NRI

International Nuclear Information System (INIS)

Valach, M.; Klouzal, J.; Dostal, M.; Zymak, J.

2013-01-01

NRI Rez plc took part in the previous coordinated research projects focused on fuel behaviour modelling held by the IAEA - FUMEX-I and FUMEX-II. These were very helpful for the development and validation of various codes used in the Nuclear Research Institute Rez (NRI) for the evaluation of the fuel rod thermomechanical behaviour. Based on the considerations of our needs related to the modeling for Czech NPPs we have performed basic parametric calculations of two LOCA cases (IFA-650.1 and IFA-650.2) and detailed evaluation WWER related cases Kola MIR ramp rods. The AREVA ''Idealized case'' and 16x16 LTA cases were also calculated because of the high burnup reached. Report summarises simulated cases in the frame of FUMEX III Project at the NRI Rez plc. (author)

Effects of fuel load and moisture content on fire behaviour and heating in masticated litter-dominated fuels

Science.gov (United States)

Jesse K. Kreye; Leda N. Kobziar; Wayne C. Zipperer

2013-01-01

Mechanical fuels treatments are being used in fire-prone ecosystems where fuel loading poses a hazard, yetlittle research elucidating subsequent fire behaviour exists, especially in litter-dominated fuelbeds. To address this deficiency, we burned constructed fuelbeds from masticated sites in pine flatwoods forests in northern Florida...

Behaviour of spent fuel assemblies during extended storage

International Nuclear Information System (INIS)

1987-04-01

This report is the final report of the IAEA Co-ordinated Research Programme on Behaviour of Spent Fuel Assemblies During Extended Storage (BEFAST, Phase I, 1981-86). It contains the results on wet and dry spent fuel storage technologies obtained from 11 institutes (10 countries: Austria, Canada, Czechoslovakia, Finland, German Democratic Republic, Hungary, Japan, Sweden, USA and USSR) participating in the BEFAST CRP during the time period 1981-86. Names of participating institutes and chief investigators are given. The interim spent fuel storage has been recognized as an important independent step in the nuclear fuel cycle. Due to the delay in commercial reprocessing of spent fuel in some cases it should be stored up to 30-50 years or more before reprocessing or final disposal. This programme was evaluated by all its participants and observers as very important and helpful for the nuclear community and it was decided to continue it further (1986-91) as BEFAST, Phase II

Simulation model of dynamical behaviour of reactor fuel assemblies

International Nuclear Information System (INIS)

Planchard, J.

1994-01-01

This report briefly describes the homogenized dynamical equations of a tube bundle placed in a perfect irrotational fluid, on case of small displacements. This approach can be used to study the mechanical behaviour of fuel assemblies of PWR reactor submitted to earthquake or depressurization blow-down. The numerical calculations require to define the added mass matrix of the fuel assemblies, for which the principle of computation is presented. (author). 14 refs., 4 figs

COMETHE III-M for transient fuel rod behaviour prediction

International Nuclear Information System (INIS)

Billaux, M.; Vliet, J. van

1983-01-01

The COMETHE III-M version is being developed in order to provide fuel rod behaviour prediction capability both in steady-state and in transient situations. It also allows to estimate the fuel rod enthalpy evolution versus time or burnup which may be important in core-related safety studies. This paper describes the transient heat transfer models, including transient heat conduction inside the fuel rod, and a subchannel model providing transient flow as well as enthalpy calculation capability. Transient fission gas release is also modelled on basis of the change rate of oxide temperature. The models are illustrated by a few calculation examples. (author)

Behaviour of short-lived iodines in operating UO2 fuel elements

International Nuclear Information System (INIS)

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

Statistical treatment of the thermal behaviour of fast reactor fuel

International Nuclear Information System (INIS)

Russo, S.; Truffert, J.; Martella, T.; Marbach, G.

1981-08-01

In a sodium cooled fast reactor, fuel temperature is an important parameter acting on main characteristics of the project on fuel element and core behaviour. This parameter is important to define boundary conditions of fuel element utilisation. A method of statistical evaluation of temperature and of temperature increase higher than a given value is presented. This evaluation is obtained in the FIEVRE code by a combination of incertainties by means of a Monte Carlo optimized method. An application of FIEVRE code is presented in the case of Rapsodie-Fortissimo fuel at the beginning of refueling at nominal conditions without transient [fr

Behaviour of molten reactor fuels under accident conditions

International Nuclear Information System (INIS)

Xavier Swamikannu, A.; Mathews, C.K.

1980-01-01

The behaviour of molten reactor fuels under accident conditions has received considerable importance in recent times. The chemical processes that occur in the molten state among the fuel, the clad components and the concrete of the containment building under the conditions of a core melt down accident in oxide fuelled reactors have been reviewed with the purpose of identifying areas of developmental work required to be performed to assess and minimize the consequences of such an accident. This includes the computation and estimation of vapour pressure of various gaseous species over the fuel, the clad and the coolant, providing of sacrificial materials in the concrete in order to protect the containment building in order to prevent release of radioactive gases into the atmosphere and understanding the distribution and chemical state of fission products in the molten fuel in order to provide for the effective removal of their decay heats. (auth.)

Development of computer models for fuel element behaviour in water reactors

International Nuclear Information System (INIS)

Gittus, J.H.

1987-03-01

Description of fuel behaviour during normal operation transients and accident conditions has always represented a most challenging and important problem. Reliable predictions constitute a basic demand for safety based calculations, for design purposes and for fuel performance. Therefore, computer codes based on deterministic and probabilistic models were developed. Possibility of comprehensive descriptions of the phenomena is precluded in view of the great number of individual processes, involving physical, chemical, thermohydraulical and mechanical parameters, to be considered in a wide range of situations. In case of fast thermal transients predictive capability is limited by the kinetics of evolution of the system and its eventual dynamic behaviour. Evidently, probabilistic approaches are also limited by the sparcity and limited breadth of the impirical data base. Code predictions have to be evaluated against power reactor data, results from simulation experiments and, if possible, include cross validation of different codes and validation of sub-models. Progress on this subject is reviewed in this report, which completes the co-ordinated research programme on 'Development of Computer Models for Fuel Element Behaviour in Water Reactors' (D-COM), initiated under the auspices of the IAEA in 1981

Results of tests under normal and abnormal operating conditions concerning LMFBR fuel element behaviour

International Nuclear Information System (INIS)

Languille, A.; Bergeonneau, P.; Essig, C.; Guerin, Y.

1985-04-01

The objective of this paper is to improve the knowledge on LMFBR fuel element behaviour during protected and unprotected transients in RAPSODIE and PHENIX reactors in order to evaluate its reliability. The range of the tests performed in these reactors is sufficiently large to cover normal and also extreme off normal conditions such as fuel melting. Results of such tests allow to better establish transient design limits for reactor structural components in particular for fuel pin cladding which play a lead role in controlling the accident sequence. Three main topics are emphasized in this paper: fuel melting during slow over-power excursions; influence of the fuel element geometrical evolution on reactivity feedback effects and reactor dynamic behaviour; clad damage evaluation during a transient (essentially very severe loss of flow)

Simulation of the behaviour of nuclear fuel under high burnup conditions

International Nuclear Information System (INIS)

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

2014-01-01

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

Specialists' meeting on theoretical modelling of LMFBR fuel pin behaviour. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1979-12-01

The purpose of the meeting was to provide an opportunity for exchanging views of theoretical modelling of LMFBR fuel pin behaviour and to summarise the IWGFR member countries' knowledge in this field. The special emphasis was placed on normal operating conditions. The technical part of the meeting was divided into six sessions, as follows: An overview of fuel modelling studies; Key factors and basic phenomena relevant to fuel pin behaviour modelling; Application to steady state operation and normal transients; Experimental validation through pins in service and specific irradiation experiments; Advanced fuels; and Brief review of existing codes. During the meeting, papers were presented by the delegates on behalf of their countries or organization. The papers, which are included in this report, were either in the form of a general survey of the subject, or on specific technical subjects. In each subject area presentations appropriate to the subject were made from the submitted papers. The presentations were followed by discussions of the questions raised and summary is made.

Specialists' meeting on theoretical modelling of LMFBR fuel pin behaviour. Summary report

International Nuclear Information System (INIS)

1979-12-01

The purpose of the meeting was to provide an opportunity for exchanging views of theoretical modelling of LMFBR fuel pin behaviour and to summarise the IWGFR member countries' knowledge in this field. The special emphasis was placed on normal operating conditions. The technical part of the meeting was divided into six sessions, as follows: An overview of fuel modelling studies; Key factors and basic phenomena relevant to fuel pin behaviour modelling; Application to steady state operation and normal transients; Experimental validation through pins in service and specific irradiation experiments; Advanced fuels; and Brief review of existing codes. During the meeting, papers were presented by the delegates on behalf of their countries or organization. The papers, which are included in this report, were either in the form of a general survey of the subject, or on specific technical subjects. In each subject area presentations appropriate to the subject were made from the submitted papers. The presentations were followed by discussions of the questions raised and summary is made

WWER water chemistry related to fuel cladding behaviour

Energy Technology Data Exchange (ETDEWEB)

Kysela, J; Zmitko, M [Nuclear Research Inst. plc., Rez (Czech Republic); Vrtilkova, V [Nuclear Fuel Inst., Prague (Czech Republic)

1997-02-01

Operational experience in WWER primary water chemistry and corrosion related to the fuel cladding is reviewed. Insignificant corrosion of fuel cladding was found which is caused by good corrosion resistance of Zr1Nb material and relatively low coolant temperature at WWER-440 reactor units. The differences in water chemistry control is outlined and an attention to the question of compatibility of Zircaloys with WWER water chemistry is given. Some results of research and development in field of zirconium alloy corrosion behaviour are discussed. Experimental facility for in-pile and out-of-pile cladding material corrosion testing is shown. (author). 14 refs, 5 figs, 3 tabs.

Mortality prediction to hospitalized patients with influenza pneumonia: PO2 /FiO2 combined lymphocyte count is the answer.

Science.gov (United States)

Shi, Shu Jing; Li, Hui; Liu, Meng; Liu, Ying Mei; Zhou, Fei; Liu, Bo; Qu, Jiu Xin; Cao, Bin

2017-05-01

Community-acquired pneumonia (CAP) severity scores perform well in predicting mortality of CAP patients, but their applicability in influenza pneumonia is powerless. The aim of our research was to test the efficiency of PO 2 /FiO 2 and CAP severity scores in predicting mortality and intensive care unit (ICU) admission with influenza pneumonia patients. We reviewed all patients with positive influenza virus RNA detection in Beijing Chao-Yang Hospital during the 2009-2014 influenza seasons. Outpatients, inpatients with no pneumonia and incomplete data were excluded. We used receiver operating characteristic curves (ROCs) to verify the accuracy of severity scores or indices as mortality predictors in the study patients. Among 170 hospitalized patients with influenza pneumonia, 30 (17.6%) died. Among those who were classified as low-risk (predicted mortality 0.1%-2.1%) by pneumonia severity index (PSI) or confusion, urea, respiratory rate, blood pressure, age ≥65 year (CURB-65), the actual mortality ranged from 5.9 to 22.1%. Multivariate logistic regression indicated that hypoxia (PO 2 /FiO 2  ≤ 250) and lymphopenia (peripheral blood lymphocyte count pneumonia confirmed a similar pattern and PO 2 /FiO 2 combined lymphocyte count was also the best predictor for predicting ICU admission. In conclusion, we found that PO 2 /FiO 2 combined lymphocyte count is simple and reliable predictor of hospitalized patients with influenza pneumonia in predicting mortality and ICU admission. When PO 2 /FiO 2  ≤ 250 or peripheral blood lymphocyte count pneumonia. © 2015 The Authors. The Clinical Respiratory Journal published by John Wiley & Sons Ltd.

Mechanistic modelling of gaseous fission product behaviour in UO2 fuel by Rtop code

International Nuclear Information System (INIS)

Kanukova, V.D.; Khoruzhii, O.V.; Kourtchatov, S.Y.; Likhanskii, V.V.; Matveew, L.V.

2002-01-01

The current status of a mechanistic modelling by the RTOP code of the fission product behaviour in polycrystalline UO 2 fuel is described. An outline of the code and implemented physical models is presented. The general approach to code validation is discussed. It is exemplified by the results of validation of the models of fuel oxidation and grain growth. The different models of intragranular and intergranular gas bubble behaviour have been tested and the sensitivity of the code in the framework of these models has been analysed. An analysis of available models of the resolution of grain face bubbles is also presented. The possibilities of the RTOP code are presented through the example of modelling behaviour of WWER fuel over the course of a comparative WWER-PWR experiment performed at Halden and by comparison with Yanagisawa experiments. (author)

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

OpenAIRE

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...

An evaluation of gas release modelling approaches as to their applicability in fuel behaviour models

International Nuclear Information System (INIS)

Mattila, L.J.; Sairanen, R.T.

1980-01-01

The release of fission gas from uranium oxide fuel to the voids in the fuel rod affects in many ways the behaviour of LWR fuel rods both during normal operating conditions including anticipated transients and during off-normal and accident conditions. The current trend towards significantly increased discharge burnup of LWR fuel will increase the importance of fission gas release considerations both from the design and safety viewpoints. In the paper fission gas release models are classified to 5 categories on the basis of complexity and physical sophistication. For each category, the basic approach common to the models included in the category is described, a few representative models of the category are singled out and briefly commented in some cases, the advantages and drawbacks of the approach are listed and discussed and conclusions on the practical feasibility of the approach are drawn. The evaluation is based on both literature survey and our experience in working with integral fuel behaviour models. The work has included verification efforts, attempts to improve certain features of the codes and engineering applications. The classification of fission gas release models regarding their applicability in fuel behaviour codes can of course be done only in a coarse manner. The boundaries between the different categories are vague and a model may be well refined in a way which transfers it to a higher category. Some current trends in fuel behaviour research are discussed which seem to motivate further extensive efforts in fission product release modelling and are certain to affect the prioritizing of the efforts. (author)

On the behaviour of intragranular fission gas in UO2 fuel

International Nuclear Information System (INIS)

Loesoenen, Pekka

2000-01-01

Data obtained from the literature concerning the behaviour of intragranular gas in sintered LWR UO 2 fuel are reviewed comprehensively. The characteristics of single gas atoms and bubbles, as a function of irradiation time, temperature, fission rate and burn-up are described, based on the reported experimental data. The relevance of various phenomena affecting gas behaviour is evaluated. The current status of modelling of the behaviour of intragranular gas is considered in light of the present findings. Simple calculations showed that the conventional approximation for the effective diffusion coefficient does not adequately describe the gas behaviour under transient conditions, when bubble coarsening plays a key role in the release. The difference in the release fraction, compared with a more mechanistic approach, could be as large as 30%. A number of recommendations regarding possible defects in the mechanistic approach to modelling of intragranular gas are highlighted. The lack of an effective numerical method for solving the set of relevant non-linear differential equations is shown to be a serious obstacle in implementing the mechanistic models for fission gas release (FGR), in integral fuel performance codes

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

International Nuclear Information System (INIS)

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

Fuel rod behaviour during transients

International Nuclear Information System (INIS)

Bilsby, C.F.; Haste, T.J.; Garlick, A.; Cameron, R.F.

1982-04-01

The clad deformation code CANSWELL-2 is described. This is used, either as a stand-alone code or within MABEL-2, to predict and analyse the results of LOCA simulations in the Halden and NRU reactors and in the KfK and PROPAT rigs. Experimental evidence on fuel behaviour in RIA, PCM and ATWS events is presented with inclusion of certain FRAP-T5 results. Published calculations from the accident codes FRAP-T4 and FRAP-T5 are compared with experimental results in simulated loss of coolant tests in the Power Burst Facility. The limitations of this code in its treatment of RIA, PCM and ATWS events are considered. (U.K.)

Nuclear Fuel Behaviour in Loss-of-coolant Accident (LOCA) Conditions

International Nuclear Information System (INIS)

Pettersson, Kjell; Chung, Haijung; ); Billone, Michael; Fuketa, Toyoshi; Nagase, Fumihisa; Grandjean, Claude; Hache, George; Papin, Joelle; Heins, Lothar; Hozer, Zoltan; In de Betou, Jan; Kelppe, Seppo; Mayer, Ralph; Scott, Harold; Voglewede, John; Sonnenburg, Heinz; Sunder, Sham; Valach, Mojmir; Vrtilkova, Vera; Waeckel, Nicolas; Wiesenack, Wolfgang; Zimmermann, Martin

2009-01-01

The NEA Working Group on Fuel Safety (WGFS) is tasked with advancing the current understanding of fuel safety issues by assessing the technical basis for current safety criteria and their applicability to high burn-up and to new fuel designs and materials. The group aims at facilitating international convergence in this area, including as regards experimental approaches and interpretation and the use of experimental data relevant for safety. In 1986, a working group of the NEA Committee on the Safety of Nuclear Installations (CSNI) issued a state-of-the-art report on water reactor fuel behaviour in design-basis accident (DBA) conditions. The 1986 report was limited to the oxidation, embrittlement and deformation of pressurised water reactor (PWR) fuel in a loss-of-coolant accident (LOCA). Since then, considerable experimental and analytical work has been performed, which has led to a broader and deeper understanding of LOCA-related phenomena. Further, new cladding alloys have been produced, which might behave differently than the previously used Zircaloy-4, both under normal operating conditions and during transients. Compared with 20 years ago, fuel burn-up has been significantly increased, which requires extending the LOCA database in order to cover the high burnup range. There was also a clear need to address LOCA performance for reactor types other than PWRs. The present report has been prepared by the WGFS and covers the following technical aspects: - Description of different LOCA scenarios for major types of reactors: BWRs, PWRs, VVERs and to a lesser extent CANDUs. - LOCA phenomena: ballooning, burst, oxidation, fuel relocation and possible fracture at quench. - Details of high-temperature oxidation behaviour of various cladding materials. - Metallurgical phase change, effect of hydrogen and oxygen on residual cladding ductility. - Methods for LOCA testing, for example two-sided oxidation and ring compression for ductility, and integral quench test for

Behaviour of rock-like oxide fuels under reactivity-initiated accident conditions

International Nuclear Information System (INIS)

Kazuyuki, Kusagaya; Takehiko, Nakamura; Makio, Yoshinaga; Hiroshi, Akie; Toshiyuki, Yamashita; Hiroshi, Uetsuka

2002-01-01

Pulse irradiation tests of three types of un-irradiated rock-like oxide (ROX) fuel - yttria-stabilised zirconia (YSZ) single phase, YSZ and spinel (MgAl 2 O 4 ) homogeneous mixture and particle-dispersed YSZ/spinel - were conducted in the Nuclear Safety Research Reactor to investigate the fuel behaviour under reactivity-initiated accident conditions. The ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m 3 , which was comparable to that of un-irradiated UO 2 fuel. The failure mode of the ROX fuels, however, was quite different from that of the UO 2 fuel. The ROX fuels failed with fuel pellet melting and a part of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below∼12 GJ/m 3 . The YSZ type and homogenous YSZ/spinel type ROX fuels failed by cladding burst when their temperatures peaked, while the particle-dispersed YSZ/spinel type ROX fuel seemed to have failed by cladding local melting. (author)

Approaches to simulate channel and fuel behaviour using CATHENA and ELOCA

International Nuclear Information System (INIS)

Sabourin, G.; Huynh, H.M.

1996-01-01

This paper documents a new approach where the detailed fuel and channel thermalhydraulic calculations are performed by an integrated code. The thermalhydraulic code CATHENA is coupled with the fuel code ELOCA. The scenario used in the simulations is a 100% pump suction break, because its power pulse is large and leads to high sheath temperatures. The results shows that coupling the two codes at each time step can have an important effect on parameters such as the sheath, fuel and pressure tube temperature. In summary, this demonstrates that this original approach can model more adequately the channel and fuel behaviour under postulated large LOCAs. (author)

Parameter study on the influence of prepressurization on LWR fuel rod behaviour during normal operation and hypothetical LOCA

International Nuclear Information System (INIS)

Fuchs, H.P.; Brzoska, B.; Depisch, F.; Sauermann, W.

1978-01-01

To analyse the influence of prepressurization on fuel rod behaviour, a parametric study has been performed considering the effects of the as-fabricated fuel rod internal prepressure on the normal operation and postulated LOCA red behaviour of a 1300 MWe1 KWU standard nuclear power plant pressurized water reactor. A reduction of prepressurization in the analysed range results in a negligible worsened normal operation behaviour whereas the LOCA behaviour is improved significantly. (author)

Modelling spent fuel and HLW behaviour in repository conditions

International Nuclear Information System (INIS)

Esparza, A. M.; Esteban, J. A.

2003-01-01

The aim of this report is to give the reader an overall insight of the different models, which are used to predict the long-term behaviour of the spent fuels and HLW disposed in a repository. The models must be established on basic data and robust kinetics describing the mechanisms controlling spent fuel alteration/dissolution in a repository. The UO2 matrix, or source term, contains embedded in it the , majority of radionuclides of the spent fuel (some are in the gap cladding). For this reason the SF radionuclides release models play a significant role in the performance assessment of radioactive waste disposal. The differences existing between models published in the literature are due to the conceptual understanding of the processes and the degree of the conservatism used with the parameter values, and the boundary conditions. They mainly differ in their level of simplification and their final objective. Sometimes are focused the show compliance with regulatory requirements, other to support decision making, to increase the level of confidence of public and scientific community, could be empirical, semi-empirical or analytical. The models take into account the experimental results from radionuclides releases and their extrapolation to the very long term. Its necessary a great statistics for have a representative dissolution rate, due at the number of experimental results is not very high and many of them show a great scatter, independently of theirs different compositions by axial and radial variations, due to linear power or local burnup. On the other hand, it is difficult to predict the spent fuel behaviour over the long term, based in short term experiments. In this report is given a little description of the radionuclides distribution in the spent fuel and also in the cladding/pellet gap, grain boundary, cracks and rim zones (the matrix rim zone can be considered with an especial characteristics very different to the rest of the spent fuel), and structural

Modelling spent fuel and HLW behaviour in repository conditions

Energy Technology Data Exchange (ETDEWEB)

Esparza, A M; Esteban, J A

2003-07-01

The aim of this report is to give the reader an overall insight of the different models, which are used to predict the long-term behaviour of the spent fuels and HLW disposed in a repository. The models must be established on basic data and robust kinetics describing the mechanisms controlling spent fuel alteration/dissolution in a repository. The UO2 matrix, or source term, contains embedded in it the , majority of radionuclides of the spent fuel (some are in the gap cladding). For this reason the SF radionuclides release models play a significant role in the performance assessment of radioactive waste disposal. The differences existing between models published in the literature are due to the conceptual understanding of the processes and the degree of the conservatism used with the parameter values, and the boundary conditions. They mainly differ in their level of simplification and their final objective. Sometimes are focused the show compliance with regulatory requirements, other to support decision making, to increase the level of confidence of public and scientific community, could be empirical, semi-empirical or analytical. The models take into account the experimental results from radionuclides releases and their extrapolation to the very long term. Its necessary a great statistics for have a representative dissolution rate, due at the number of experimental results is not very high and many of them show a great scatter, independently of theirs different compositions by axial and radial variations, due to linear power or local burnup. On the other hand, it is difficult to predict the spent fuel behaviour over the long term, based in short term experiments. In this report is given a little description of the radionuclides distribution in the spent fuel and also in the cladding/pellet gap, grain boundary, cracks and rim zones (the matrix rim zone can be considered with an especial characteristics very different to the rest of the spent fuel), and structural

The MOX fuel behaviour test IFA-597.4/.5/.6/.7; Summary of in-pile fuel temperature and gas release data

Energy Technology Data Exchange (ETDEWEB)

Koike, Hisashi

2003-11-15

It is considered important to study the in-reactor behaviour of MOX fuel in order to enhance the database on such fuel. For this reason, IFA-597.4/.5/.6/.7 were included in the joint research programme of the Halden Project. The series of tests, containing two MIMAS-MOX fuel rods, both equipped with a fuel centre thermocouple and a pressure bellows transducer, has been irradiated in the Halden Reactor since July 1997 under HBWR conditions. The objectives of the test series were to study the thermal and fission gas release (FGR) behaviour of MOX fuel and to explore potential differences in behaviour between solid and hollow pellets. One of the rods had mainly solid pellets, while the other contained only hollow pellets. Both rods had an initial Pu-fissile enrichment of 6.07%. The cladding outside diameter was 9.50 mm, and the initial fuel-clad gap was 180 mum. In the course of the test, power upratings for FGR studies of the MOX fuel were planned at burnup intervals of about 10 MWd/kg MOX. The power uprating was successfully performed at approx10 MWd/kg MOX, where the estimated fuel peak temperature of the solid pellets exceeded the FGR threshold temperature for UO{sub 2} fuel, while that of the hollow pellets remained below the threshold. For the solid fuel, the temperature at onset of FGR was consistent with the empirical threshold temperature for UO{sub 2} fuel. For the hollow fuel, gas release was observed at temperatures below the threshold. FGRs at the end-of-life were approx17% for the solid pellet rod and approx14% for the hollow pellet rod, respectively. As a result of discussions in HPG meetings, IFA-597.7 was unloaded in January 2002. PIE was carried out to check in-pile pressure measurements and examine fuel structural characteristics. The discharge burn-up of the MOX fuel was 32 MWd/kg MOX as determined from in-pile power data. This report supersedes HWR-712 (June 2002) previously issued on in-pile data from IFA-597.4/5/6/7. (Author)

Computer simulation of the behaviour and performance of a CANDU fuel rod

International Nuclear Information System (INIS)

Marino, A.C.

1997-01-01

At the Argentine Atomic Energy Commission (Comision Nacional de Energia Atomica, CNEA) the BACO code (for 'BArra COmbustible', fuel rod) was developed. It allows the simulation of the thermo-mechanical performance of a cylindrical fuel rod in a Pressurized Heavy Water Reactor (PHWR). The standard present version of the code (2.30), is a powerful tool for a relatively easy and complete evaluation of fuel behaviour predictions. Input parameters and, therefore, output ones may include statistical dispersion. As a demonstration of BACO capabilities we include a review of CANDU fuel applications, and the calculation and a parametric analysis of a characteristic CANDU fuel. (author)

Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

Energy Technology Data Exchange (ETDEWEB)

Barani, T.; Bruschi, E.; Pizzocri, D. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division, Via La Masa 34, I-20156 Milano (Italy); Pastore, G. [Fuel Modeling and Simulation Department, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Van Uffelen, P. [European Commission, Joint Research Centre, Directorate for Nuclear Safety and Security, P.O. Box 2340, 76125 Karlsruhe (Germany); Williamson, R.L. [Fuel Modeling and Simulation Department, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Luzzi, L., E-mail: Lelio.Luzzi@polimi.it [Politecnico di Milano, Department of Energy, Nuclear Engineering Division, Via La Masa 34, I-20156 Milano (Italy)

2017-04-01

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel performance analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. In particular, experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of the burst release process in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of conventional diffusion-based models to introduce the burst release effect. The concept and governing equations of the model are presented, and the sensitivity of results to the newly introduced parameters is evaluated through an analytic sensitivity analysis. The model is assessed for application to integral fuel rod analysis by implementation in two structurally different fuel performance codes: BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D code). Model assessment is based on the analysis of 19 light water reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the quantitative predictions of integral fuel rod FGR and the qualitative representation of the FGR kinetics with the transient model relative to the canonical, purely diffusion-based models of the codes. The overall quantitative improvement of the integral FGR predictions in the two codes is comparable. Moreover, calculated radial profiles of xenon concentration after irradiation are investigated and compared to experimental data, illustrating the underlying representation of the physical mechanisms of burst release

Measurement and behaviour of technetium in fast reactor fuel reprocessing

International Nuclear Information System (INIS)

Ferguson, C.; Kyffin, T.W.

1986-02-01

A method is described for the spectrophotometric measurement of technetium in plant solutions from the reprocessing of fast reactor fuel. The technetium is selectively extracted using tri-iso-octylamine. After back extraction, thiocyanate is added, in the presence of tetrabutyl-ammonium hydroxide, to form the red hexa-thiocyanato anionic complex in a chloroform medium. The concentration of the technetium is then calculated from the spectrophotometric measurement of this complex. This method was applied to bulk samples, collected during a PFR fuel reprocessing campaign, to identify the main routes followed by technetium through the reprocessing plant. In order to understand the probable behaviour of technetium in the process plant streams, an investigation into the influence of plutonium IV nitrate on the extraction of Tc (VII) into 20%v/v tributyl phosphate/odourless kerosene solution from nitric acid solutions, was initiated. The results of this investigation, along with the known distribution coefficient for the extraction of the uranyl/technetium complex U0 2 (N0 3 )(Tc0 4 ).2TBP and the redox chemistry of technetium, are used to predict the probable behaviour of technetium in the process plant streams. This predicted behaviour is compared with the experimental results and reasonable agreement is obtained between experiment and theory, considering the history of the samples analysed. (author)

Effects of cold worked and fully annealed claddings on fuel failure behaviour

International Nuclear Information System (INIS)

Saito, Shinzo; Hoshino, Hiroaki; Shiozawa, Shusaku; Yanagihara, Satoshi

1979-12-01

Described are the results of six differently heat-treated Zircaloy clad fuel rod tests in NSRR experiments. The purpose of the test is to examine the extent of simulating irradiated claddings in mechanical properties by as-cold worked ones and also the effect of fully annealing on the fuel failure bahaviour in a reactivity initiated accident (RIA) condition. As-cold worked cladding does not properly simulated the embrittlement of the irradiated one in a RIA condition, because the cladding is fully annealed before the fuel failure even in the short transient. Therefore, the fuel behaviour such as fuel failure threshold energy, failure mechanism, cladding deformation and cladding oxidation of the fully annealed cladding fuel, as well as that of the as-cold worked cladding fuel, are not much different from that of the standard stress-relieved cladding fuel. (author)

Simulation of fuel rod behaviour during various break LOCAs in PWRs

International Nuclear Information System (INIS)

Gadalla, A.A.; El-Fawal, M.M.

1996-01-01

During loss of coolant accident (LOCAs) course of events, attention focuses on fuel rod cladding temperature behaviour. In this study, the DRUFAN analytical model and LOBI-MOD2 experimental modeling scheme for fuel rod temperature behaviour during C L-Break LOCA in PWRs, are described and discussed. These models are applied for the investigation of fuel rod cladding temperature behaviour during LOCA blowdown phase. A spectrum of selected values representing small, intermediate and large CL- Break sizes are considered in the predictions. The results of the predictions demonstrated that calculated heater rod temperature at steady state as well as the transient period up to 1000 sec are going in good agreement with the measured values. However above 1000 sec the calculated temperatures are higher than the measured values. This indicates that code predictions in this period are conservative. The results indicated also that, in case of small CL-break LOCA (0.01 A and 0.01 and 0.03 A), the heater rod cladding temperature don't rise above saturation temperature. However, on the top of the heater rod, DNB is occurred in case of 0.03 A CL break, while for 0.01 A break, DNB didn't occur. In case of intermediate and large CL-break; (0.05 A, 0.10 A and 1 A), the results showed that, the heater rod cladding temperature exceeded the saturation temperature and DNB prevailed in upper and intermediate sections of the core. 15 figs., 2 tabs

PIN99W, Modelling of VVER and PWR Fuel Rod Thermomechanical Behaviour

International Nuclear Information System (INIS)

Valach, M.; Strizhov, P.; Svoboda, R.

2000-01-01

1 - Description of program or function: The Code is developed to describe fuel rod thermomechanical behaviour in operational conditions. The main goal of this code is to calculate fuel temperature, gap conductivity, fission gas release and inner gas pressure. 2 - Methods: - fuel rod temperature response is solved by using one-dimensional finite element method combined with weighted residuals method; - the code involves models describing physical phenomena typical for the fuel irradiated in Light Water Power Reactors (densification, restructuring, fission gas release, swelling and relocation) ; - this code is updated and improves PIN-micro code. 3 - Restrictions on the complexity of the problem: - simplified mechanistic solution; - only steady-state solution; - no cladding failure criterion; - no model for axial fuel-cladding interaction

A user input manual for single fuel rod behaviour analysis code FEMAXI-III

International Nuclear Information System (INIS)

Saito, Hiroaki; Yanagisawa, Kazuaki; Fujita, Misao.

1983-03-01

Principal objectives of Safety related research in connection with lighr water reactor fuel rods under normal operating condition are mainly addressed 1) to assess fuel integrity under steady state condition and 2) to generate initial condition under hypothetical accident. These assessments have to be relied principally upon steady state fuel behaviour computing code that is able to calculate fuel conditions to tbe occurred in a various manner. To achieve these objectives, efforts have been made to develope analytical computer code that calculates in-reactor fuel rod behaviour in best estimate manner. The computer code developed for the prediction of the long-term burnup response of single fuel rod under light water reactor condition is the third in a series of code versions:FEMAMI-III. The code calculates temperature, rod internal gas pressure, fission gas release and pellet-cladding interaction related rod deformation as a function of time-dependent fuel rod power and coolant boundary conditions. This document serves as a user input manual for the code FEMAMI-III which has opened to the public in year of 1982. A general description of the code input and output are included together with typical examples of input data. A detailed description of structures, analytical submodels and solution schemes in the code shall be given in the separate document to be published. (author)

Study of the chemical behaviour of technetium during irradiated fuels reprocessing

International Nuclear Information System (INIS)

Zelverte, A.

1988-04-01

This paper deals with the preparation of the lower oxidation states +III +IV and +V of technetium in nitric acid and its behaviour during the reprocessing of nuclear fuels (PUREX process). The first part of this work is a bibliographical study of this element in solution without any strong ligand. By chemical and electrochemical technics, pentavalent, tetravalent and trivalent technetium species, were prepared in nitric acid. The following chemical reactions are studied: - trivalent and tetravalent technetium oxidation by nitrate ion. - hydrazine and tetravalent uranium oxidation catalysed by technetium: in those reactions, we point out unequivocally the prominent part of trivalent and tetravalent technetium, - technetium behaviour towards hydroxylamine. Technetium should not cause any disturbance in the steps where hydroxylamine is employed to destroy nitrous acid and hydrazine replacement by hydroxylamine in uranium-plutonium partition could contribute to a best reprocessing of nuclear fuels [fr

'THERMOST' for analysing thermo-structural behaviour of LWR fuel rods under PCI conditions

International Nuclear Information System (INIS)

Nuno, H.; Ogawa, S.; Kobayashi, H.

1983-01-01

As a method for evaluating fuel rod performance under power ramping or load following operations, the combined FROST/ THERMOST system has been developed and brought into practical use. FROST was presented at the IAEA Blackpool Meeting in 1978, and THERMOST is the subject of this paper. The major purpose of THERMOST is to analyse very detailed thermal and structural fuel behaviour in a rather localised part of the fuel rod whereas FROST deals with whole rod general performance. The code handles two-dimensional thermal and structural analyses simultaneously by using a finite element method, in axial section or in lateral section. It consists of a fundamental FEM system of generalised constitution, and a surrounding subroutine system which characterises fuel behaviour, such as temperature distribution, thermal expansion, elastoplasticity, creep, cracking, swelling, growth, etc. Thermal analysis is handled by heat conduction and heat transfer element (six kinds), and structural analysis by axisymmetric ring and lateral plane element (six kinds). Boundary problems such as contact, friction and cracking are treated by gap and crack elements. A sample calculation of PCI performance on a PWR fuel rod under ramping conditions is presented with some in-pile test data. (author)

Structural behaviour of fuel assemblies for water cooled reactors. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2005-07-01

At the invitation of the Government of France and in response to a proposal of the IAEA Technical Working Group on Water Reactor Fuel Performance and Technology (TWGFPT), the IAEA convened a Technical Meeting on Fuel Assembly Structural Behaviour in Cadarache, France, from 22 to 26 November 2004. The meeting was hosted by the CEA Cadarache Centre, AREVA Framatome-ANP and Electricite de France. The meeting aimed to provide in depth technical exchanges on PWR and WWER operational experience in the field of fuel assembly mechanical behaviour and the potential impact of future high burnup fuel management on fuel reliability. It addressed in-service experience and remedial solutions, loop testing experience, qualification and damage assessment methods (analytic or experimental ones), mechanical behaviour of the fuel assembly including dynamic and fluid structure interaction aspects, modelling and numerical analysis methods, and impact of the in-service evolution of the structural materials. Sixty-seven participants from 17 countries presented 30 papers in the course of four sessions. The topics covered included the impact of hydraulic loadings on fuel assembly (FA)performance, FA bow and control rod (CR) drop kinetics, vibrations and rod-to-grid wear and fretting, and, finally, evaluation and modelling of accident conditions, mainly from seismic causes. FA bow, CR drop kinetics and hydraulics are of great importance under conditions of higher fuel duties including burnup increase, thermal uprates and longer fuel cycles. Vibrations and rod-to-grid wear and fretting have been identified as a key cause of fuel failure at PWRs during the past several years. The meeting demonstrated that full-scale hydraulic tests and modelling provide sufficient information to develop remedies to increase FA skeleton resistance to hydraulic loads, including seismic ones, vibrations and wear. These proceedings are presented as a book with an attached CD-ROM. The first part of the CD

Contribution to the communication: European fuel behaviour perspective

International Nuclear Information System (INIS)

Pickmann, D.O.; Marin, J.F.; Weidinger, H.; Junkrans, S.; Bairiot, H.

1981-08-01

The safety and security problems particular to pressurized water reactors are reviewed. These problems are followed up at statutory level by the Service Central de Surete des Installations Nucleaires (Central Department of Nuclear Installation Safety) and at technical level by the Institut de Protection et de Surete Nucleaire (Nuclear Protection and Safety Institute) linked to the CEA. The safety analysis is based on the design standards and the technical specifications of reactor components and nuclear substances. They relate to the behaviour of a reactor under normal or accidental operation. The fuel elements are studied in the reactor and outside it by means of loops and power ramps. This information is embodied in models which describe the behaviour of the various parts of the reactor during the accident [fr

Development and application of the BISON fuel performance code to the analysis of fission gas behaviour

International Nuclear Information System (INIS)

Pastore, G.; Hales, J.D.; Novascone, S.R.; Perez, D.M.; Spencer, B.W.; Williamson, R.L.

2014-01-01

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that has been under development at Idaho National Laboratory (USA) since 2009. The capabilities of BISON comprise implicit solution of the fully coupled thermo-mechanics and diffusion equations, applicability to a variety of fuel forms, and simulation of both steady-state and transient conditions. The code includes multiphysics constitutive behavior for both fuel and cladding materials, and is designed for efficient use on highly parallel computers. This paper describes the main features of BISON, with emphasis on recent developments in modelling of fission gas behaviour in LWR-UO 2 fuel. The code is applied to the simulation of fuel rod irradiation experiments from the OECD/NEA International Fuel Performance Experiments Database. The comparison of the results with the available experimental data of fuel temperature, fission gas release, and cladding diametrical strain during pellet-cladding mechanical interaction is presented, pointing out a promising potential of the BISON code with the new fission gas behaviour model. (authors)

Modelling the release behaviour of cesium during severe fuel degradation

International Nuclear Information System (INIS)

Lewis, B.J.; Andre, B.; Morel, B.

1995-01-01

An analytical model has been applied to describe the diffusional release of fission product cesium from Zircaloy-clad fuel under high-temperature reactor accident conditions. The present treatment accounts for the influence of the atmosphere (i.e., changing oxygen potential) on the state of fuel oxidation and the release kinetics. The effects of fuel dissolution on the volatile release behaviour (under reducing conditions) is considered in terms of earlier crucible experiments and a simple model based on bubble coalescence and transport in metal pools. The model has been used to interpret the cesium release kinetics observed in steam and hydrogen experiments at the Vertical Irradiation (VI) Facility in the Oak Ridge National Laboratory and at the HEVA/VERCORS Facility in the Commissariat a l'Energie Atomique. (author)

Power ramping, cycling and load following behaviour of water reactor fuel

International Nuclear Information System (INIS)

1988-05-01

The present meeting was scheduled by the International Atomic Energy Agency upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology. Sixty-three participants representing 15 countries and one international organization attended the meeting. Twenty papers were presented during three technical sessions, followed by panel discussions which allowed to formulate the conclusions of the meeting and recommendations to the Agency. The objective of this Technical Committee Meeting is to review the ''State-of-the-Art'', make critical comments and recommendations with the aim of improving fuel reliability and assure integrity of the cladding and core materials when subjected to ramping and cycling sequences. The Meeting was organized in three sessions: Session 1. ''Mechanical Behaviour and Fission Gas Release'' (7 papers); Session 2. ''Power Ramping and Power Cycling Demonstration Programmes in Research Reactors'' (5 papers); Session 3. ''Fuel Behaviour in Power Reactors'' (9 papers). Between the sessions, the session chairmen, together with the speakers, prepared and presented reports with summary, conclusions and recommendations of the individual sessions. These reports are added to this summary report. A separate abstract was prepared for each of these 21 presentations. Refs, figs and tabs

Fuel Behaviour at High During RIA and LOCA Accidents

International Nuclear Information System (INIS)

Barrio del Juanes, M. T.; Garcia Cuesta, J. C.; Vallejo Diaz, I.; Herranz Puebla

2001-01-01

Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs

Development, verification and validation of the fuel channel behaviour computer code FACTAR

Energy Technology Data Exchange (ETDEWEB)

Westbye, C J; Brito, A C; MacKinnon, J C; Sills, H E; Langman, V J [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

FACTAR (Fuel And Channel Temperature And Response) is a computer code developed to simulate the transient thermal and mechanical behaviour of 37-element or 28-element fuel bundles within a single CANDU fuel channel for moderate loss of coolant accident conditions including transition and large break LOCA`s (loss of coolant accidents) with emergency coolant injection assumed available. FACTAR`s predictions of fuel temperature and sheath failure times are used to subsequent assessment of fission product releases and fuel string expansion. This paper discusses the origin and development history of FACTAR, presents the mathematical models and solution technique, the detailed quality assurance procedures that are followed during development, and reports the future development of the code. (author). 27 refs., 3 figs.

Model for the behaviour of thorium and uranium fuels at pelletization

International Nuclear Information System (INIS)

Ferreira Neto, Ricardo Alberto

2000-11-01

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

IFPE/IFA-533, Fuel Thermal Behaviour at High Burnup, Halden Reactor

International Nuclear Information System (INIS)

Gyori, Cs.; Turnbull, J.A.

1997-01-01

Description: After twelve years irradiation in the Halden Boiling Water Reactor two fuel rods (Rod 807 and Rod 808) were re-instrumented with fuel centre thermocouples and reloaded into the reactor in order to investigate the fuel thermal behaviour at high burnup. The fuel rods were pre-irradiated with four other rods in the upper cluster of IFA-409 (IFA=Instrumented Fuel Assembly) from May 1973 to June 1985. After base irradiation the four neighbouring rods were re-instrumented with pressure transducers and ramp tested in IFA-535.5 and IFA-535.6 providing useful data about fission gas release (FGR) presented in the Fuel Performance Database as well (Ref. 1). The two rods re-instrumented with fuel centre thermocouples have been irradiated as IFA-533.2 from April 1992. As the irradiation history of IFA-533.2 in the first months was very similar to the history of the ramp tests, the fuel temperature and FGR data measured in the different IFAs can complement each other, although the fuel-cladding gap sizes were slightly different and due to re-instrumentation the internal gas conditions were also dissimilar

A comparison of the metallurgical behaviour of dispersion fuels with uranium silicides and U6Fe as dispersants

International Nuclear Information System (INIS)

Nazare, S.

1984-01-01

In the past few years metallurgical studies have been carried out to develop fuel dispersions with U-densities up to 7.0 Mg U m -3 . Uranium silicides have been considered to be the prime candidates as dispersants; U 6 Fe being a potential alternative on account of its higher U-density. The objective of this paper is to compare the metallurgical behaviour of these two material combinations with regard to the following aspects: (1) preparation of the compounds U 3 Si, U 3 Si 2 and U 6 Fe; (2) powder metallurgical processing to miniature fuel element plates; (3) reaction behaviour under equilibrium conditions in the relevant portions of the ternary U-Si-Al and U-Fe-Al systems; (4) dimensional stability of the fuel plates after prolonged thermal treatment; (5) thermochemical behaviour of fuel plates at temperatures near the melting point of the cladding. Based on this data, the possible advantages of each fuel combination are discussed. (author)

Investigation of large grain and Gd-doped WWER fuels behaviour at BOL in the Halden reactor

International Nuclear Information System (INIS)

Volkov, B.

2008-01-01

In this paper the following issues have been discussed: 1) WWER fuel tests in the HBWR; 2) Main objectives of the test with large grains and Gd-doped WWER fuel; 3) Analysis of of the the data at BOL focus on: Gd-doped fuel thermal behaviour, fuel elongation and dimension stability as well as cladding elongation early in life. At the end authors concluded that: 1) No indication of substantial effect of large grains on fuel thermal performance at BOL; 2) Densification observed in large grain fuel is similar to the ordinary uranium dioxide fuel with 95-96 % of theoretical density; 3) Dimension stability of large grain fuel is similar or even better than that in reference WWER fuel; 4) More stable dimension behaviour of large grain fuel at power could be attributed to its lower creep or densification at high temperature in the centre part of the fuel; 5) Cladding elongation detectors indicated identical early-in-life PCMI in both large grain and reference fuel rods, which reflected an accommodation effect of fuel pellets in claddings during first rise to power; no residual strains in either fuel types were observed; subsequent cladding elongation measurements show a trend to irradiation growth; 6) No clear evidence for densification of Gd-doped WWER fuel is observed during first irradiation cycle

IFPE/TRIBULATION R1, Fuel Rod Behaviour at High Burnup

International Nuclear Information System (INIS)

Turnbull, J.A.

2002-01-01

Description: The TRIBULATION (Tests Relative to High Burnup Limitations Arising Normally in LWRs) International Programme started in July 1980 and was organized jointly by BelgoNucleaire and the Nuclear Energy Centre at Mol (CEN/SCK) with the co-sponsorship of 14 participating organizations. The objectives of the programme were twofold. It was primarily a demonstration programme aimed at assessing the fuel rod behaviour at high burn-up, when an earlier transient had occurred in the power plant. The second objective was to investigate the behaviour of different fuel rod designs and manufacturers when subjected to a steady state irradiation history to high burn-up. The first objective was met by irradiating fuel rods under steady state conditions in the BR3 reactor and under transient conditions in BR2. The effect of the transient was determined by comparing data from 4 identical rods tested as follows: i) BR3 irradiation followed by PIE; ii) BR3 irradiation followed by BR2 transient then PIE; iii) BR3 irradiation followed by BR2 transient and re-irradiated in BR3 before PIE; iv) BR3 irradiation and continued BR3 irradiation to maximum burn-up before PIE. The Database contains data from 19 cases using rods fabricated by BelgoNucleaire (BN) (11) and Brown Boveri Reactor GmbH (BBR) (8)

Behaviour of power and research reactor fuel in wet and dry storage

Energy Technology Data Exchange (ETDEWEB)

Freire-Canosa, J [Nuclear Waste Management Organization (Canada)

2012-07-01

Canada has developed extensive experience in both wet and dry storage of CANDU fuel. Fuel has been stored in water pools at CANDU reactor sites for approximately 45 years, and in dry storage facilities for a large part of the past decade. Currently, Canada has 38 450 t U of spent fuel in storage, of which 8850 t U are in dry storage. In June 2007, the Government of Canada selected the Adaptive Phased Management (APM) approach, recommended by the Nuclear Waste Management Organization (NWMO), for the long-term management of Canada's nuclear-fuel waste. The Canadian utilities and AECL are conducting development work in extended storage systems as well as research on fuel behaviour under storage conditions. Both activities have as ultimate objective to establish a technical basis for assuring the safety of long-term fuel storage.

Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

International Nuclear Information System (INIS)

Khvostionov, V.E.; Egorenkov, P.M.; Malankin, P.V.

2004-01-01

Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO 2 SO 4 ) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

On the behaviour of dissolved fission gases prior to transient testing of fuel pins

International Nuclear Information System (INIS)

Wood, M.H.; Matthews, J.R.

1978-10-01

The TREAT and CABRI series of reactor safety experiments on irradiated fuel require the transfer of fuel pins from the reactor in which the fuel has achieved some burn-up to the test facility. Subsequently, the fuel is restored to power in the test facility for some time before transient heating is initiated. Such pre-test manoeuvres, where the fuel is subjected to changes in the fission rate and temperature, may have important consequences for the fission gas behaviour during the transient experiment. The results of rate theory calculations are used to assess these effects. (author)

Seasonal variation of the global mixed layer depth: comparison between Argo data and FIO-ESM

Science.gov (United States)

Zhang, Yutong; Xu, Haiming; Qiao, Fangli; Dong, Changming

2018-03-01

The present study evaluates a simulation of the global ocean mixed layer depth (MLD) using the First Institute of Oceanography-Earth System Model (FIOESM). The seasonal variation of the global MLD from the FIO-ESM simulation is compared to Argo observational data. The Argo data show that the global ocean MLD has a strong seasonal variation with a deep MLD in winter and a shallow MLD in summer, while the spring and fall seasons act as transitional periods. Overall, the FIO-ESM simulation accurately captures the seasonal variation in MLD in most areas. It exhibits a better performance during summer and fall than during winter and spring. The simulated MLD in the Southern Hemisphere is much closer to observations than that in the Northern Hemisphere. In general, the simulated MLD over the South Atlantic Ocean matches the observation best among the six areas. Additionally, the model slightly underestimates the MLD in parts of the North Atlantic Ocean, and slightly overestimates the MLD over the other ocean basins.

Water chemistry regimes for VVER-440 units: water chemistry influence on fuel cladding behaviour

International Nuclear Information System (INIS)

Zmitko, M.

1999-01-01

In this lecture next problems of water chemistry influence on fuel cladding behaviour for VVER-440 units are presented: primary coolant technologies; water chemistry specification and control; fuel integrity considerations; zirconium alloys cladding corrosion (corrosion versus burn-up; water chemistry effect; crud deposition; hydrogen absorption; axial offset anomaly); alternatives for the primary coolant regimes

IAEA specialists' meeting on power ramping and cycling behaviour of water reactor fuel. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1983-06-01

At its fourth Annual Meeting, the IAEA International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended that the Agency should hold a second Specialists' Meeting on 'Power Ramping and Cycling Behaviour of Water Reactor Fuel'. As research activities related to power ramping and cycling of water reactor fuel have been pursued vigorously, it was the objective of this meeting to review and discuss today's State of the Art and current understanding of water reactor fuel behaviour related to this these. Emphasis should be on practical experience and experimental investigations. The meeting was organised in five sessions: Power ramping and power cycling programs in power and and research reactors; Experimental methods; Power ramping and cycling results; Investigations and results of separate effects, especially related to PCI, defect mechanism, mechanical response, fuel design, and specially related to fission gas release; Operational strategies, recommendations and economic implications. The session chairmen, together with the speakers, prepared and presented reports with summary, conclusions and recommendations of the individual sessions. These reports are added to this summary report.

IAEA specialists' meeting on power ramping and cycling behaviour of water reactor fuel. Summary report

International Nuclear Information System (INIS)

1983-06-01

At its fourth Annual Meeting, the IAEA International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended that the Agency should hold a second Specialists' Meeting on 'Power Ramping and Cycling Behaviour of Water Reactor Fuel'. As research activities related to power ramping and cycling of water reactor fuel have been pursued vigorously, it was the objective of this meeting to review and discuss today's State of the Art and current understanding of water reactor fuel behaviour related to this these. Emphasis should be on practical experience and experimental investigations. The meeting was organised in five sessions: Power ramping and power cycling programs in power and and research reactors; Experimental methods; Power ramping and cycling results; Investigations and results of separate effects, especially related to PCI, defect mechanism, mechanical response, fuel design, and specially related to fission gas release; Operational strategies, recommendations and economic implications. The session chairmen, together with the speakers, prepared and presented reports with summary, conclusions and recommendations of the individual sessions. These reports are added to this summary report

Computational analysis of the behaviour of nuclear fuel under steady state, transient and accident conditions

International Nuclear Information System (INIS)

2007-12-01

Accident analysis is an important tool for ensuring the adequacy and efficiency of the provision in the defence in depth concept to cope with challenges to plant safety. Accident analysis is the milestone of the demonstration that the plant is capable of meeting any prescribed limits for radioactive releases and any other acceptable limits for the safe operation of the plant. It is used, by designers, utilities and regulators, in a number of applications such as: (a) licensing of new plants, (b) modification of existing plants, (c) analysis of operational events, (d) development, improvement or justification of the plant operational limits and conditions, and (e) safety cases. According to the defence in depth concept, the fuel rod cladding constitutes the first containment barrier of the fission products. Therefore, related safety objectives and associated criteria are defined, in order to ensure, at least for normal operation and anticipated transients, the integrity of the cladding, and for accident conditions, acceptable radiological consequences with regard to the postulated frequency of the accident, as usually identified in the safety analysis reports. Therefore, computational analysis of fuel behaviour under steady state, transient and accident conditions constitutes a major link of the safety case in order to justify the design and the safety of the fuel assemblies, as far as all relevant phenomena are correctly addressed and modelled. This publication complements the IAEA Safety Report on Accident Analysis for Nuclear Power Plants (Safety Report Series No. 23) that provides practical guidance for establishing a set of conceptual and formal methods and practices for performing accident analysis. Computational analysis of the behaviour of nuclear fuel under transient and accident conditions, including normal operation (e.g. power ramp rates) is developed in this publication. For design basis accidents, depending on the type of influence on a fuel element

Plans d'expérience constructions et analyses statistiques

CERN Document Server

Tinsson, Walter

2010-01-01

Il est souvent n cessaire de r aliser des exp riences afin de mod liser le comportement d un ph nom ne complexe. La m thode des plans d exp rience a pour objectif d obtenir un maximum d information sur le ph nom ne tudi en un minimum d exp riences. Ceci est primordial si l objectif est un gain de temps ou de qualit . Cet ouvrage d taille les fondements th oriques de la m thode math matique des plans d exp rience. Ceci est abord tout au long des quatre parties suivantes. Pr sentation g n rale de la m thode et des outils math matiques. Plans d exp rience pour facteurs quantitatifs: mod le d ordr

Multi-scale modelling of the physicochemical-mechanical coupling of fuel behaviour at high temperature in pressurized water reactors

International Nuclear Information System (INIS)

Julien, Jerome

2008-01-01

Within the frame of the problematic of pellet-sheath interaction in a nuclear fuel rod, a good description of the fuel thermo-mechanical behaviour is required. This research thesis reports the coupling of physics-chemistry (simulation of gas transfers between different cavities) and mechanics (assessment of fuel viscoplastic strains). A new micromechanical model is developed which uses a multi-scale approach to describe the evolution of the double population of cavities (cavities with two different scales) while taking internal pressures as well as the fuel macroscopic viscoplastic behaviour into account. The author finally describes how to couple this micromechanical mode to physics-chemistry models [fr

Probabilistic analysis of fuel pin behaviour during an eventual loss of coolant in PWR reactors

International Nuclear Information System (INIS)

1981-02-01

Brief description of the development of the coolant loss incident in a pressurized water reactor and analysis of its significance for the behaviour of the fuel rods. Description of a probalistic method for estimating the effects of the accident on the fuel rods and results obtained [fr

Thermal behaviour of high burnup PWR fuel under different fill gas conditions

International Nuclear Information System (INIS)

Tverberg, T.

2001-01-01

During its more than 40 years of existence, a large number of experiments have been carried out at the Halden Reactor Project focusing on different aspects related to nuclear reactor fuel. During recent years, the fuels testing program has mainly been focusing on aspects related to high burnup, in particular in terms of fuel thermal performance and fission gas release, and often involving reinstrumentation of commercially irradiated fuel. The paper describes such an experiment where a PWR rod, previously irradiated in a commercial reactor to a burnup of ∼50 MWd/kgUO 2 , was reinstrumented with a fuel central oxide thermocouple and a cladding extensometer together with a high pressure gas flow line, allowing for different fill gas compositions and pressures to be applied. The paper focuses on the thermal behaviour of such LWR rods with emphasis on how different fill gas conditions influence the fuel temperatures and gap conductance. Rod growth rate was also monitored during the irradiation in the Halden reactor. (author)

Measurement and analysis of vibrational behaviour of an SNR-fuel element in sodium flow

International Nuclear Information System (INIS)

Hess, B.F.H.; Ruppert, E.; Schmidt, H.; Vinzens, K.

1975-01-01

Within the framework of SNR-300 fuel element development programme a complete full size fuel element dummy has been tested thoroughly for nearly 3000 hours at 650 0 C system temperature in the AKB sodium loop at Interatom, Bensberg. Investigations of the hydraulic characteristics by measurements of specific pressure losses, flow velocities, leakage flow through the piston rings and investigations of its vibrational behaviour were part of this endurance test at elevated temperatures. The pressure drop versus flow and the leakage measurement are mentioned briefly to confirm the correctness of the test hydraulics. The vibrational behaviour of the element and the approach to analysis is the main object of this report. (Auth.)

Dilatational behaviour of ZrNb1 fuel cans of a WWER-type reactor during a loss-of-coolant accident

International Nuclear Information System (INIS)

Adam, E.; Stephan, M.; Wetzel, L.

1987-01-01

Based on an assessment of various factors of influence on the performance of fuel cans during normal operation and imaginable accidents, the necessity of studying creep and burst behaviour of WWER-type fuel cans of ZrNb1 under simulated LOCA conditions has been proved and an experimental facility designed for this purpose is described. Control of fuel can temperature is accomplished through a minicomputer during the creep and bursts experiments. With this, various temperature loading profiles of the fuel cans can be realized. Experimental results on dilatational behaviour of ZrNb1 fuel cans from isothermal creep and burst experiments in air are presented and compared with values for Zircaloy. (author)

The ''THERMOST'' for analysing thermo-structural behaviour of LWR fuel rod under PCI conditions

International Nuclear Information System (INIS)

Nuno, H.; Ogawa, S.; Kobayashi, H.

1983-01-01

As one of the methods for evaluating the fuel rod performances under power ramping or load following operations, the combined ''FROST'' and ''THERMOST'' system has been developed and being brought into practical use. The former had already been presented at Blackpool Meeting in 1978, and the latter is going to be presented in this paper. The major purpose of the THERMOST is to analyse very detailed thermal and structural fuel behaviours in a rather localized part of fuel rod whereas the FROST deals with whole-rod-wide general performances. The code handles 2-dimensional thermal and structural analyses simultaneously by using finite element method, in axial section wide or in lateral section wide. It consists of a fundamental FEM system of generalized constitution and its surrounding subroutine system which characterizes fuel behaviours such as temperature distribution, thermal expansion, elastoplasticity, creep, cracking, swelling, growth, etc. Thermal analysis is handled by heat conduction and heat transfer elements (6 kinds) and structural analysis by axisymmetric ring and lateral plane elements (6 kinds). Boundary problems such as contact, friction and cracking are treated by gap and crack elements. A sample calculation of PCI performance on a PWR fuel rod under ramping condition is presented with some inpile test data. (author)

Development of a microindentation technique to determine the fuel mechanical behaviour at high burnup

International Nuclear Information System (INIS)

Baron, D.; Leclercq, S.; Spino, J.; Taheri, S.

1998-01-01

One of the major problems that face the conceptors and users of nuclear power plants is the demonstration of the cladding integrity (the Zircaloy clad that contains the fuel pellets), particularly in class I and II operating conditions. A long term collaboration between EDF and the Applied Mechanics Laboratory (LMA) of Besancon (France) has existed for several years, and a unified modelling of the cladding has been developed in this frame. But a good understanding of the cladding response is not of total use if the mechanical solicitation applied to this clad by the fuel pellet is not completely known. The potential evolution and the non-homogeneity of the fuel stiffness was recently demonstrated by Spino (TUI) on Vickers micro-hardness tests at room temperature. Thus, in order to get furthermore data, TUI and EDF decided to build a specific microindentation device able to perform the tests needed by the modelers. After a brief recall of what the effects of irradiation are on the fuel pellet mechanical behaviour, this paper presents the microindentation device to be built, as well as the principles that underline its use. Finally, the way the experimental results will be used to determine the mechanical behaviour of the fuel pellet under irradiation is pointed out. (author)

Characterisation and final disposal behaviour of theoria-based fuel kernels in aqueous phases

International Nuclear Information System (INIS)

Titov, M.

2005-08-01

Two high-temperature reactors (AVR and THTR) operated in Germany have produced about 1 million spent fuel elements. The nuclear fuel in these reactors consists mainly of thorium-uranium mixed oxides, but also pure uranium dioxide and carbide fuels were tested. One of the possible solutions of utilising spent HTR fuel is the direct disposal in deep geological formations. Under such circumstances, the properties of fuel kernels, and especially their leaching behaviour in aqueous phases, have to be investigated for safety assessments of the final repository. In the present work, unirradiated ThO 2 , (Th 0.906 ,U 0.094 )O 2 , (Th 0.834 ,U 0.166 )O 2 and UO 2 fuel kernels were investigated. The composition, crystal structure and surface of the kernels were investigated by traditional methods. Furthermore, a new method was developed for testing the mechanical properties of ceramic kernels. The method was successfully used for the examination of mechanical properties of oxide kernels and for monitoring their evolution during contact with aqueous phases. The leaching behaviour of thoria-based oxide kernels and powders was investigated in repository-relevant salt solutions, as well as in artificial leachates. The influence of different experimental parameters on the kernel leaching stability was investigated. It was shown that thoria-based fuel kernels possess high chemical stability and are indifferent to presence of oxidative and radiolytic species in solution. The dissolution rate of thoria-based materials is typically several orders of magnitude lower than of conventional UO 2 fuel kernels. The life time of a single intact (Th,U)O 2 kernel under aggressive conditions of salt repository was estimated as about hundred thousand years. The importance of grain boundary quality on the leaching stability was demonstrated. Numerical Monte Carlo simulations were performed in order to explain the results of leaching experiments. (orig.)

Comparison with experiment of COMETHE III-L fuel rod behaviour predictions

International Nuclear Information System (INIS)

Vliet, J. van; Billaux, M.

1983-01-01

A comparison is presented between experimental results and COMETHE III-L fuel rod behaviour predictions. The first part of the paper focuses on mechanical aspects, with as main experiments, AECL X-264 and Studsvik Interramp. The second part presents the results of a wide FGR benchmarking campaign, with a reference to previous COMETHE versions. It appears that the variance between experiment and calculation has decreased by a factor four when the III-J version was improved into the III-L version. As conclusion, some COMETHE III-L calculations are presented in order to illustrate its capability of predicting fuel rod performance limits. (author)

Modeling CANDU type fuel behaviour during extended burnup irradiations using a revised version of the ELESIM code

International Nuclear Information System (INIS)

Arimescu, V.I.; Richmond, W.R.

1992-05-01

The high-burnup database for CANDU fuel, with a variety of cases, offers a good opportunity to check models of fuel behaviour, and to identify areas for improvement. Good agreement of calculated values of fission-gas release, and sheath hoop strain, with experimental data indicates that the global behaviour of the fuel element is adequately simulated by a computer code. Using, the ELESIM computer code, the fission-gas release, swelling, and fuel pellet expansion models were analysed, and changes made for gaseous swelling, and diffusional release of fission-gas atoms to the grain boundaries. Using this revised version of ELESIM, satisfactory agreement between measured values of fission-gas release was found for most of the high-burnup database cases. It is concluded that the revised version of the ELESIM code is able to simulate with reasonable accuracy high-burnup as well as low-burnup CANDU fuel

Nuclear fuels

International Nuclear Information System (INIS)

2008-01-01

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

Fuel behaviour in the case of severe accidents and potential ATF designs. Fuel Behavior in Severe Accidents and Potential Accident Tolerance Fuel Designs

International Nuclear Information System (INIS)

Cheng, Bo

2013-01-01

This presentation reviews the conditions of fuel rods under severe loss of coolant conditions, approaches that may increase coping time for plant operators to recover, requirements of advanced fuel cladding to increase tolerance in accident conditions, potential candidate alloys for accident-tolerant fuel cladding and a novel design of molybdenum (Mo) -based fuel cladding. The current Zr-alloy fuel cladding will lose all its mechanical strength at 750-800 deg. C, and will react rapidly with high-pressure steam, producing significant hydrogen and exothermic heat at 700-1000 deg. C. The metallurgical properties of Zr make it unlikely that modifications of the Zr-alloy will improve the behaviour of Zr-alloys at temperatures relevant to severe accidents. The Mo-based fuel cladding is designed to (1) maintain fuel rod integrity, and reduce the release rate of hydrogen and exothermic heat in accident conditions at 1200-1500 deg. C. The EPRI research has thus far completed the design concepts, demonstration of feasibility of producing very thin wall (0.2 mm) Mo tubes. The feasibility of depositing a protective coating using various techniques has also been demonstrated. Demonstration of forming composite Mo-based cladding via mechanical reduction has been planned

28 CFR 0.119 - Organization.

Science.gov (United States)

2010-07-01

... 28 Judicial Administration 1 2010-07-01 2010-07-01 false Organization. 0.119 Section 0.119 Judicial Administration DEPARTMENT OF JUSTICE ORGANIZATION OF THE DEPARTMENT OF JUSTICE 1-Office of Community Oriented Policing Services § 0.119 Organization. The Office of Community Oriented Policing...

Composite fuel behaviour under and after irradiation

International Nuclear Information System (INIS)

Dehaudt, P.; Mocellin, A.; Eminet, G.; Caillot, L.; Delette, G.; Bauer, M.; Viallard, I.

1997-01-01

Two kinds of composite fuels have been irradiated in the SILOE reactor. They are made of UO 2 particles dispersed in a molybdenum metallic (CERMET) or a MgAl 2 O 4 ceramic (CERCER) matrix. The irradiation conditions have allowed to reach a 50000 MWd/t U burn-up in these composite fuels after a hundred equivalent full power days long irradiation. The irradiation is controlled by a continuous measure of the pellet centre line temperature. It allows to have information about the TANOX rods thermal behaviour and the fuels thermal conductivities in comparing the centre line temperature versus linear power curves among themselves. Our results show that the CERMET centre line temperature is much lower than the CERCER and UO 2 ones: 520 deg. C against 980 deg. C at a 300W/cm linear power. After pin puncturing tests the rods are dismantled to recover each fuel pellet. In the CERCER case, the cladding peeling off has revealed that the fuel came into contact with the cladding and that some of the pellets were linked together. Optical microscopy observations show a changing of the MgAl 2 O 4 matrix state around the UO 2 particles at the pellets periphery. This transformation may have caused a swelling and would be at the origin of the pellet-cladding and the pellet-pellet interactions. No specific damage is seen after irradiation. The CERMET pellets are not cracked and remain as they were before irradiation. The CERCER crack network is slightly different from that observed in UO 2 . Kr retention was evaluated by annealing tests under vacuum at 1580 deg. C or 1700 deg. C for 30 minutes. The CERMET fission gas release is lower than the CERCER one. Inter- and intragranular fission gas bubbles are observed in the UO 2 particles after heat treatments. The CERCER pellet periphery has also cracked and the matrix has transformed again around UO 2 particles to present a granular and porous aspect. (author). 4 refs, 6 figs, 2 tabs

Creep behaviour of porous metal supports for solid oxide fuel cells

DEFF Research Database (Denmark)

Boccaccini, Dino; Frandsen, Henrik Lund; Sudireddy, Bhaskar Reddy

2014-01-01

The creep behaviour of porous ironechromium alloy used as solid oxide fuel cell support was investigated, and the creep parameters are compared with those of dense strips of similar composition under different testing conditions. The creep parameters were determined using a thermo......-mechanical analyser with applied stresses in the range from 1 to 15 MPa and temperatures between 650 and 800 _C. The GibsoneAshby and Mueller models developed for uniaxial creep of open-cell foams were used to analyse the results. The influence of scale formation on creep behaviour was assessed by comparing the creep...... data for the samples tested in reducing and oxidising atmospheres. The influence of preoxidation on creep behaviour was also investigated. In-situ oxidation during creep experiments increases the strain rate while pre-oxidation of samples reduces it. Debonding of scales at high stress regime plays...

The behaviour of water-cooled reactor fuel rods in steady state and transient conditions

International Nuclear Information System (INIS)

Strupczewski, A.; Marks, P.

1997-01-01

In this report, the results of temperature field and filling gas pressure calculations by means of contemporary calculational models for a WWER-440 and WWER-1000 type fuel rod at low and high burnup operating under steady-state conditions are presented. A review of in-core temperature and pressure measurements for various types of LWR fuel is also included. Basing on calculational and collected measured data, the behaviour of fuel cladding during large and small break LOCA, is estimated with special emphasis on their oxidation and failure resistance. (author)

Design, Manufacturing and Irradiation Behaviour of Fast Reactor Fuel. Proceedings of a Technical Meeting

International Nuclear Information System (INIS)

2013-04-01

Fast reactors are vital for ensuring the sustainability of nuclear energy in the long term. They offer vastly more efficient use of uranium resources and the ability to burn actinides, which are otherwise the long-lived component of high level nuclear waste. These reactors require development, qualification, testing and deployment of improved and innovative nuclear fuel and structural materials having very high radiation resistance, corrosion/erosion and other key operational properties. Several IAEA Member States have made efforts to advance the design and manufacture of technologies of fast reactor fuels, as well as to investigate their irradiation behaviour. Due to the acute shortage of fast neutron testing and post-irradiation examination facilities and the insufficient understanding of high dose radiation effects, there is a need for international exchange of knowledge and experience, generation of currently missing basic data, identification of relevant mechanisms of materials degradation and development of appropriate models. Considering the important role of nuclear fuels in fast reactor operation, the IAEA Technical Working Group on Fuel Performance and Technology (TWGFPT) proposed a Technical Meeting (TM) on 'Design, Manufacturing and Irradiation Behaviour of Fast Reactors Fuels', which was hosted by the Institute of Physics and Power Engineering (IPPE) in Obninsk, Russian Federation, from 30 May to 3 June 2011. The TM included a technical visit to the fuel production plant MSZ in Elektrostal. The purpose of the meeting was to provide a forum to share knowledge, practical experience and information on the improvement and innovation of fuels for fast reactors through scientific presentations and brainstorming discussions. The meeting brought together 34 specialists from national nuclear agencies, R and D and design institutes, fuel vendors and utilities from 10 countries. The presentations were structured into four sections: R and D Programmes on FR Fuel

Microstructural evolution and Am migration behaviour in Am-containing fuels at the initial stage of irradiation

International Nuclear Information System (INIS)

Tanaka, Kosuke; Miwa, Shuhei; Sato, Isamu; Osaka, Masahiko; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shin-ichi; Yoshimochi, Hiroshi; Tanaka, Kenya

2010-01-01

In order to investigate the effect of americium addition to MOX fuels on the irradiation behaviour, the 'Am-1' programme is being conducted in JAEA. The Am-1 programme consists of two short-term irradiation tests of 10-minute and 24-hour irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post-irradiation examinations (PIE) are in progress. The PIE for Am-containing MOX fuels focused on the microstructural evolution and redistribution behaviour of Am at the initial stage of irradiation and the results to date are reported. The successful development of fabrication technology with remote handling and the evaluation of thermo-chemical properties based on the out-of-pile experiments are described with an emphasis on the effects of Am addition on the MOX fuel properties. (authors)

The behaviour of spherical HTR fuel elements under accident conditions

Energy Technology Data Exchange (ETDEWEB)

Schenk, W; Naoumidis, A [Institute for Reactor Material, KFA Juelich (Germany)

1985-07-01

Hypothetical accidents may lead to significantly higher temperatures in HTR fuel than during normal operation. In order to obtain meaningful statements on fission product behaviour and release, irradiated spherical fuel elements containing a large number of coated particles (20,000-40,000) with burnups between 6 and 16% FIMA were heated at temperatures between 1400 and 2500 deg. C. HTI-pyrocarbon coating retains the gaseous fission products (e.g. Kr) very well up to about 2400 deg. C if the burnup does not exceed the specified value for THTR (11.5%). Cs diffuses through the pyrocarbon significantly faster than Kr and the diffusion is enhanced at higher fuel burnups because of irradiation induced kernel microstructure changes. Below about 1800 deg. C the Cs release rate is controlled by diffusion in the fuel kernel; above this temperature the diffusion in the pyrocarbon coating is the controlling parameter. An additional SiC coating interlayer (TRISO) ensures Cs retention up to 1600 deg. C. However, the release obtained in the examined fuel elements was only by a factor of three lower than through the HTI pyrocarbon. Solid fission products added to UO{sub 2}-TRISO particles to simulate high burnup behave in various ways and migrate to attack the SiC coating. Pd migrates fastest and changes the SiC microstructure making it permeable.

Fuel pin behaviour under conditions of control rod withdrawal accident in CABRI-2 experiments

International Nuclear Information System (INIS)

Papin, Joelle; Lemoine, Francette; Sato, Ikken; Struwe, Dankward; Pfrang, Werner

1994-01-01

Simulation of the control rod withdrawal accident has been performed in the international CABRI-2 experimental programme. The tests realized with industrial pins led to clarification of the influence of the pellet design and have shown the important role of fission products on the solid fuel swelling which promotes early pin failure with solid fuel pellet. With annular pellet design, large fuel swelling combined to low smear density leads to degradation of fuel thermal conductivity and thus reduces power to melt. However, the high margin to deterministic failure is confirmed with hollow pellets. Improvements of the modelling were necessary to describe such behaviours in computer codes as SAS-4A, PAPAS-2S and PHYSURAC. (author)

MOX fuel irradiation behaviour: Results from X-ray microbeam analysis

International Nuclear Information System (INIS)

Walker, C.T.; Goll, W.; Matsumura, T.

1997-01-01

The behaviour of plutonium, xenon and caesium were investigated in two sections of irradiated MOX fuel produced by the OCOM process. In one fuel (OCOM30), the MOX agglomerates contained 18 wt% fissile plutonium, and had a low volume fraction of 0.17; in the other (OCOM15) the agglomerates contained 9 wt% fissile plutonium, and had a high volume fraction of 0.34. Both fuels had been irradiated under normal power reactor conditions to a burn-up of approximately 44 GWd/t. The main aim of the work was to establish whether the above differences in composition affected the percentage fission gas released by the fuels. Since U/Pu interdiffusion did not occurred during the irradiation, both fuels remained inhomogeneous on the microscopic scale. However, the concentration of plutonium in the MOX agglomerates decreases by about 50% as a result of fission, whereas the plutonium content of the UO 2 matrix increased by about a factor of four to approximately 2 wt% due to neutron capture by 238 U. The agglomerates in the OCOM15 fuel generally exhibited a finer structure due to the lower burn-up. More than 80% of the fission gas had been released from the oxide lattice of the MOX agglomerates in both fuels. However, a very high fraction of this gas precipitated and remained in the pore structure of the agglomerates. Consequently, puncturing revealed that for both fuels the percentage of gas released to the rod free volume increased from less than 0.5% at 10 GWd/t to a maximum of 3.5% at 45 GWd/t. The conclusion is that the percentage of gas released by MOX fuel is largely unaffected of the level of inhomogeneity of the fuel. In both fuels caesium showed near complete retention in both the MOX agglomerates and the UO 2 matrix. (author). 8 refs, 11 figs, 3 tabs

The behaviour of irradiated fuel under RIA transients: Interpretation of the CABRI experiments

International Nuclear Information System (INIS)

Papin, J.; Rigat, H.; Breton, J.P.; Schmitz, F.

1996-01-01

Paper presents the results of investigation of highly irradiated PWR fuel behaviour under fast power transients conducted in a sodium loop of CABRI reactor, as well as the results on development and validation of computer code SCANAIR. (author). 8 refs, 9 figs, 2 tabs

Exp-function method for solving Maccari's system

International Nuclear Information System (INIS)

Zhang Sheng

2007-01-01

In this Letter, the Exp-function method is used to seek exact solutions of Maccari's system. As a result, single and combined generalized solitonary solutions are obtained, from which some known solutions obtained by extended sine-Gordon equation method and improved hyperbolic function method are recovered as special cases. It is shown that the Exp-function method provides a very effective and powerful mathematical tool for solving nonlinear evolution equations in mathematical physics

Modeling of the PWR fuel mechanical behaviour and particularly study of the pellet-cladding interaction in a fuel rod

International Nuclear Information System (INIS)

Hourdequin, N.

1995-05-01

In Pressurized Water Reactor (PWR) power plants, fuel cladding constitutes the first containment barrier against radioactive contamination. Computer codes, developed with the help of a large experimental knowledge, try to predict cladding failures which must be limited in order to maintain a maximal safety level. Until now, fuel rod design calculus with unidimensional codes were adequate to prevent cladding failures in standard PWR's operating conditions. But now, the need of nuclear power plant availability increases. That leads to more constraining operating condition in which cladding failures are strongly influenced by the fuel rod mechanical behaviour, mainly at high power level. Then, the pellet-cladding interaction (PCI) becomes important, and is characterized by local effects which description expects a multidimensional modelization. This is the aim of the TOUTATIS 2D-3D code, that this thesis contributes to develop. This code allows to predict non-axisymmetric behaviour too, as rod buckling which has been observed in some irradiation experiments and identified with the help of TOUTATIS. By another way, PCI is influenced by under irradiation experiments and identified with the help of TOUTATIS which includes a densification model and a swelling model. The latter can only be used in standard operating conditions. However, the processing structure of this modulus provides the possibility to include any type of model corresponding with other operating conditions. In last, we show the result of these fuel volume variations on the cladding mechanical conditions. (author). 25 refs., 89 figs., 2 tabs., 12 photos., 5 appends

Influência de aderências peritoneais e fio cirúrgico na tensão de ruptura da parede abdominal em ratos

Directory of Open Access Journals (Sweden)

Roberto Martins Gonçalves

Full Text Available O resultado de operações na cavidade abdominal pode ser influenciado por aderências. Existem muitos conceitos ainda não comprovados sobre os efeitos das aderências na resistência de suturas tanto de vísceras intracavitárias quanto da parede abdominal. O presente trabalho visa a avaliar a influência das aderências peritoneais e vários tipos de fios cirúrgicos na tensão de ruptura da parede abdominal. Em 60 ratos Wistar, realizou-se laparotomia de 5cm de comprimento. A parede muscular e o peritônio foram fechados em plano único, com pontos simples, usando aleatoriamente os fios de náilon monofilamentar, ácido poliglicólico, categute simples e categute cromado, todos 4-0. Os animais foram divididos em três grupos: 1- Controle; 2- INTRODUÇÃO de 0,3g de talco dentro da cavidade abdominal; 3- Acréscimo de carboximetilcelulose sódica (CMC juntamente com o talco. Houve a análise dos grupos com sete e 21 dias. Avaliou-se o grau de aderências e a tensão de ruptura da ferida cirúrgica. A CMC reduziu a formação de aderências provocadas pelo talco (p<0,01. Houve diferença na tensão de ruptura quando comparados os grupos de sete e 21 dias (p<0,05. As aderências proporcionaram uma maior força tênsil à ferida (p<0,01. O tipo de fio utilizado não influenciou na tensão de ruptura a longo prazo. Portanto, as aderências aumentaram a força tênsil dos tecidos e o tipo de fio cirúrgico não influenciou nesse processo.

GPR119 agonists: a promising approach for T2DM treatment? A SWOT analysis of GPR119.

Science.gov (United States)

Kang, Sang-Uk

2013-12-01

Ever since its advent as a promising therapeutic target for type 2 diabetes mellitus (T2DM), G-protein-coupled receptor 119 (GPR119) has received much interest from the pharmaceutical industry. This interest peaked in June 2010, when Sanofi-Aventis agreed to pay Metabolex (Cymabay Therapeutics) US$375 million for MBX-2982, which was a representative orally active GPR119 agonist. However, Sanofi-Aventis opted to terminate the deal in May 2011 and another leading GPR119 agonist, GSK1292263, had a loss of efficacy during its clinical trial. In this review, I discuss the pros and cons of GPR119 through a strengths, weaknesses, opportunities, and threats (SWOT) analysis and propose development strategies for the eventual success of a GPR119 agonist development program. Copyright © 2013 Elsevier Ltd. All rights reserved.

Dynamic behaviour of Li batteries in hydrogen fuel cell power trains

Science.gov (United States)

Veneri, O.; Migliardini, F.; Capasso, C.; Corbo, P.

A Li ion polymer battery pack for road vehicles (48 V, 20 Ah) was tested by charging/discharging tests at different current values, in order to evaluate its performance in comparison with a conventional Pb acid battery pack. The comparative analysis was also performed integrating the two storage systems in a hydrogen fuel cell power train for moped applications. The propulsion system comprised a fuel cell generator based on a 2.5 kW polymeric electrolyte membrane (PEM) stack, fuelled with compressed hydrogen, an electric drive of 1.8 kW as nominal power, of the same typology of that installed on commercial electric scooters (brushless electric machine and controlled bidirectional inverter). The power train was characterized making use of a test bench able to simulate the vehicle behaviour and road characteristics on driving cycles with different acceleration/deceleration rates and lengths. The power flows between fuel cell system, electric energy storage system and electric drive during the different cycles were analyzed, evidencing the effect of high battery currents on the vehicle driving range. The use of Li batteries in the fuel cell power train, adopting a range extender configuration, determined a hydrogen consumption lower than the correspondent Pb battery/fuel cell hybrid vehicle, with a major flexibility in the power management.

On behaviour of fuel elements subject to combined cyclic thermomechanical loads

International Nuclear Information System (INIS)

Hsu, T.R.

1980-01-01

This paper presents detailed finite element formulations on the kinematic hardening rule of plasticity included in an existing thermoelastoplastic stress analysis code primarily designed to predict the thermomechanical behaviour of nuclear reactor fuel elements. The kinematic hardening rule is considered to be important for structures subject to repeated (or cyclic) loads. The start-up/operation/shut-down and various power excursions in a reactor all can be classified as cyclic loadings. In addition to the shifting of material yield surfaces as usually handled by the kinematic hardening rule, the thermal effect and temperature-dependent material properties have also been included in the present work for the first time. A case study related to an in-reactor experiment on a single fuel element indicated that significantly higher cumulative sheath residual strains after two load cycles was obtained by the present scheme than those calculated by the usual isotropic hardening rule. This observation may alert fuel modellers to select proper hardening rules in their analyses. (orig.)

Volatile behaviour of enrichment uranium in the total nuclear fuel price

International Nuclear Information System (INIS)

Arnaiz, J.; Inchausti, J. M.; Tarin, F.

2004-01-01

In this article the historical high volatile behaviour of the total nuclear fuel price is evaluated quantitatively and it is concluded that it has been due mainly to the fluctuations of the price of the principal components of enriched uranium (concentrates and enrichment). In order to avoid the negative effects of this volatiles behaviour as far as possible, a basic strategy in the uranium procurement activities is recommended (union of buyers, diversification of supplier, stock management, optimisation of contract portfolio and suitable currency management that guarantees a reliable uranium supply at reasonable prices. These guidelines are those that ENUSA has been following on behalf of the Spanish Utilities in the Commission of Uranium Procurement (CAU in Spanish). (Author) 11 refs

Behaviour of the reactivity for BWR fuel cells; Comportamiento de la reactividad para celdas de combustible BWR

Energy Technology Data Exchange (ETDEWEB)

Gonzalez, J. A.; Alonso, G.; Delfin, A.; Vargas, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: galonso@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

2011-11-15

In this work the behaviour of the reactivity of a fuel assembly type BWR was studied, the objective is to obtain some expressions that consider the average enrichment of U-235 and the gadolinium concentration like a function of the fuel cells burnt. Also, the applicability of the lineal reactivity model was analyzed for fuel cells type BWR. The analysis was carried out with the CASMO-4 code. (Author)

Fission products and nuclear fuel behaviour under severe accident conditions part 1: Main lessons learnt from the first VERDON test

Science.gov (United States)

Pontillon, Y.; Geiger, E.; Le Gall, C.; Bernard, S.; Gallais-During, A.; Malgouyres, P. P.; Hanus, E.; Ducros, G.

2017-11-01

This paper describes the first VERDON test performed at the end of September 2011 with special emphasis on the behaviour of fission products (FP) and actinides during the accidental sequence itself. Two other papers discuss in detail the post-test examination results (SEM, EPMA and SIMS) of the VERDON-1 sample. The first VERDON test was devoted to studying UO2 fuel behaviour and fission product releases under reducing conditions at very high temperature (∼2883 K), which was able to confirm the very good performance of the VERDON loop. The fuel sample did not lose its integrity during this test. According to the FP behaviour measured by the online gamma station (fuel sight), the general classification of the FP in relation to their released fraction is very accurate, and the burn-up effect on the release rate is clearly highlighted.

Technical Meeting on Design, Manufacturing and Irradiation Behaviour of Fast Reactors Fuels. Presentations

International Nuclear Information System (INIS)

2011-01-01

The purpose of this meeting was to enable a rationalization and advancement of the design and manufacturing processes, a better selection of promising fuels, and a reduction of the time and costs currently required for R and D and testing, as well as to contribute to the improvement of the safety features of fuels under all operational states and accidental conditions. An overview of the status and perspective of the design, manufacturing and irradiation behaviour of fast reactors fuels were provided during this meeting. The main objectives are the following: Ensure sharing and dissemination of knowledge and expertise; Discuss specific features and issues of existing fuels; Improve knowledge and data for the design and engineering of fast reactor fuel and core structural materials; Discuss perspectives on advanced fuels; Consider modern technological, design and testing tools enabling reliable performance of fuels in current and planned operational environments; Establish international consensus in the developmental efforts on advanced fast reactor technologies, including collaborative programs and experiments. Contribute to the preparation and outline of the planned IAEA Coordinated Research Project on 'Examination of advanced fast reactor fuel and core structural materials. Each of the 24 presentations made at the meeting have been indexed separately

Behaviour of short-lived fission products within operating UO2 fuel elements

International Nuclear Information System (INIS)

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)

PHEBUS program: first results on PWR fuel behaviour in LOCA conditions

International Nuclear Information System (INIS)

Del Negro, R.; Reocreux, M.; Pelce, J.; Legrand, B.; Berna, P.

1982-09-01

In the first PHEBUS test with pressurized rods some rods burst and clad temperature reached 1100 0 C in the 25 rods bundle. There is now a lot of valuable experimental results and their analysis is in progress. The phase II on fuel behaviour in case of a large LOCA will start at the beginning of 83. The onset of the SFD program is foreseen to take place on the first months of 85

ESTUDO COMPARATIVO DAS ANASTOMOSES ARTERIAIS COM FIO ABSORVÍVEL E NÃO ABSORVÍVEL EM SUÍNOS EM CRESCIMENTO

Directory of Open Access Journals (Sweden)

IBRAHIM ELIAS KALLÁS

1998-07-01

Full Text Available O objetivo deste trabalho foi observar a evolução, após 12 meses, das artérias anastomosadas com polipropileno e com fios de poliglactina 910, nos suínos em fase de crescimento. A amostra foi composta por 16 fêmeas da raça Landrace, distribuídas em dois grupos : A e B. Os animais devidamente pesados, foram submetidos a secção transversal da aorta abdominalis, abaixo das arteriae renalis dextra et sinistra,que após a medida dos diâmetros internos foram anastomosadas com os fios correspondentes, mediante suturas contínuas. Após um ano os animais foram submetidos à eutanásia e novas medidas realizadas. Em ambos os grupos ficou evidenciado que pesos maiores estão relacionados a diâmetros maiores das artérias. Houve aumento médio de 1027,5 % no peso final dos animais do grupo A e 1242,9 % nos animais do grupo B. Houve aumento médio de 98,6 % no diâmetro final das artérias dos animais do Grupo A e de 102,7% nas animais do grupo B. Para a análise estatística valeu-se dos testes de WILCOXON e MANN-WHITNEY, com cálculo de D %. Os fios de polipropileno permaneceram nas artérias dos animais do grupo A sendo identificados de duas formas: rotos ou em alça, mas sempre fazendo proeminência na luz arterial. Os fios de poliglactina 910 não foram identificados. Concluiu-se que o fios de polipropileno e de poliglactina 910 permitem o crescimento das artérias ao nível das anastomoses término-terminais na aorta abdominalis de suínos.The aim of this study is to observe polypropylene and polyglactine 910 anastomosis, in growing pigs after twelve months. Sixteen female "Landrace" piglets were distributed into two groups : A and B. After being weighed, the animals underwent aorta abdominalis transversal section under arteria renalis dextra and sinistra, diameter measurement, and anastomosis in continuous suture with the above mentioned threads. One year later, the animals were euthanized and new measurement was performed. In both groups

45 CFR 400.119 - Interstate movement.

Science.gov (United States)

2010-10-01

... 45 Public Welfare 2 2010-10-01 2010-10-01 false Interstate movement. 400.119 Section 400.119... Services § 400.119 Interstate movement. After the initial placement of an unaccompanied minor, the same procedures that govern the movement of nonrefugee foster cases to other States apply to the movement of...

Study on the influence of water chemistry on fuel cladding behaviour of LWR in Japan

International Nuclear Information System (INIS)

Mishima, Y.

1983-01-01

This article presents the results of the study on the influence of water chemistry on fuel cladding behaviour, which has been performed for more than ten years on BWRs and PWRs in Japan. The post irradiation examination (P.I.E.) program of commercial reactor fuel assembly which was explained at Tokyo meeting in 1981 includes an investigation of the characteristics and build-up conditions of crud deposited on mainly BWR fuel cladding. This article also provides a summary of the results of the investigation and shows how the results are utilized for establishing effective water chemistry measures

Thermo-mechanical behaviour modelling of particle fuels using a multi-scale approach

International Nuclear Information System (INIS)

Blanc, V.

2009-12-01

Particle fuels are made of a few thousand spheres, one millimeter diameter large, compound of uranium oxide coated by confinement layers which are embedded in a graphite matrix to form the fuel element. The aim of this study is to develop a new simulation tool for thermo-mechanical behaviour of those fuels under radiations which is able to predict finely local loadings on the particles. We choose to use the square finite element method, in which two different discretization scales are used: a macroscopic homogeneous structure whose properties in each integration point are computed on a second heterogeneous microstructure, the Representative Volume Element (RVE). First part of this works is concerned by the definition of this RVE. A morphological indicator based in the minimal distance between spheres centers permit to select random sets of microstructures. The elastic macroscopic response of RVE, computed by finite element has been compared to an analytical model. Thermal and mechanical representativeness indicators of local loadings has been built from the particle failure modes. A statistical study of those criteria on a hundred of RVE showed the significance of choose a representative microstructure. In this perspective, a empirical model binding morphological indicator to mechanical indicator has been developed. Second part of the work deals with the two transition scale method which are based on the periodic homogenization. Considering a linear thermal problem with heat source in permanent condition, one showed that the heterogeneity of the heat source involve to use a second order method to localized finely the thermal field. The mechanical non-linear problem has been treats by using the iterative Cast3M algorithm, substituting to integration of the behavior law a finite element computation on the RVE. This algorithm has been validated, and coupled with thermal resolution in order to compute a radiation loading. A computation on a complete fuel element

Fuel behaviour

International Nuclear Information System (INIS)

Fodor, M.; Matus, L.; Vigassy, J.

1987-11-01

A short summary of the main critical points in fuel performance of nuclear power reactors from chemical and mechanical point of view is given. A schedule for a limited research program is included. (author) 17 refs

46 CFR 119.420 - Engine cooling.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Engine cooling. 119.420 Section 119.420 Shipping COAST... Machinery Requirements § 119.420 Engine cooling. (a) Except as otherwise provided in paragraph (b) of this section, all engines must be water cooled and meet the requirements of this paragraph. (1) The engine head...

46 CFR 119.520 - Bilge pumps.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Bilge pumps. 119.520 Section 119.520 Shipping COAST... Ballast Systems § 119.520 Bilge pumps. (a) Each vessel must be provided with bilge pumps in accordance... have a portable hand bilge pump that must be: (1) Capable of pumping water, but not necessarily...

Transreceptor de faixa larga baseado em impulso de rádio para sensor sem fios de curta distância implementado em tecnologia CMOS

OpenAIRE

Rodrigues, David Manuel Cardoso

2013-01-01

Dissertação para obtenção do Grau de Mestre em Engenharia Electrotécnica e de Computadores O rápido desenvolvimento da microtecnologia e microelectrónica tem vindo a contribuir de forma decisiva para uma crescente utilização de sensores, sejam eles com ou sem fios, o que permitirá interagir mais eficientemente com o meio envolvente, em consequência de mais e melhor medição e atuação. Neste contexto, as redes de sensores sem fios (WSNs) estão a emergir como uma das grandes e mais impo...

46 CFR 119.320 - Water heaters.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Water heaters. 119.320 Section 119.320 Shipping COAST... Machinery § 119.320 Water heaters. (a) A water heater must meet the requirements of Parts 53 and 63 in... electric water heater is also acceptable if it: (1) Has a capacity of not more than 454 liters (120 gallons...

Fuel rod behaviour at high burnup WWER fuel cycles

International Nuclear Information System (INIS)

Medvedev, A.; Bogatyr, S.; Kouznetsov, V.; Khvostov, G.; Lagovsky; Korystin, L.; Poudov, V.

2003-01-01

The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles

Results of the investigations of transient fuel rod behaviour

International Nuclear Information System (INIS)

Fiege, A.

1980-01-01

The aim of the research on the fuel rod behaviour mainly effected in the KFZ Karlsruhe and at the KWU Erlangen as a part of the German reactor safety research program is to investigate the physical and chemical phenomena which are significant when the zircaloy claddings are failing, and to establish mathematical models verified by experiments by means of which the extent of damage in the reactor core in different incidents can be worked out in a realistic way. These mathematical models (program system SSYST) shall replace the conservative assumptions so far used for incident analyses and quantify their safety reserves, respectively. (orig./HP) [de

The MOX Fuel Behaviour Test IFA-597.4: Temperature And Pressure Data To A Burn-Up Of 5.4 MWd/kg MOX

International Nuclear Information System (INIS)

McGrath, M. A.; Teshima, H.

1998-02-01

Characterising the behaviour of MOX fuel is becoming increasingly important as many commercial reactors are or will be operating with this type of fuel. With this as a driving force, a new joint programme experiment, IFA-597.4, has been loaded into the reactor at Halden for the purpose of establishing the fission gas release behaviour of MOX fuel. Both annular and solid pellet fuel is being utilised and the irradiation is being conducted such that the fuel is initially operated below the onset of fission gas release. The fuel will later be subjected to small power up ratings which will be held for short periods of time. These are designed to bring the fuel to just above the temperature threshold for fission gas release thus allowing the FGR behaviour of both solid and annular MOX fuel to be established. The rig contains two fuel rods of active length 220 mm and diameter 8.05 mm. Both fuel rods contain MOX fuel with an initial Pu-fissile content of 6.07% and both are instrumented with a fuel centre thermocouple and a pressure transducer. The test is being performed under HBWR conditions and at the time of the reactor shutdown at the end of 1997 a mean burn-up of 5.4 MWd/kg MOX had been achieved with the rods at an average rating of 30 kW/m. The rod pressure data show that no fission gas had been released up to the shutdown. The fuel centre temperatures of both rods exhibit an initial increase concurrent with a fall in the monitored rod internal pressures as a result of fuel densification. It was estimated that about 1-1.4% fuel densification by volume had occurred in the two rods by a burn-up of about 3 MWd/kg MOX. (author)

The relevance of the IFPE Database to the modelling of WWER-type fuel behaviour

International Nuclear Information System (INIS)

Killeen, J.; Sartori, E.

2006-01-01

The aim of the International Fuel Performance Experimental Database (IFPE Database) is to provide, in the public domain, a comprehensive and well-qualified database on zircaloy-clad UO 2 fuel for model development and code validation. The data encompass both normal and off-normal operation and include prototypic commercial irradiations as well as experiments performed in Material Testing Reactors. To date, the Database contains over 800 individual cases, providing data on fuel centreline temperatures, dimensional changes and FGR either from in-pile pressure measurements or PIE techniques, including puncturing, Electron Probe Micro Analysis (EPMA) and X-ray Fluorescence (XRF) measurements. This work in assembling and disseminating the Database is carried out in close co-operation and co-ordination between OECD/NEA and the IAEA. The majority of data sets are dedicated to fuel behaviour under LWR irradiation, and every effort has been made to obtain data representative of BWR, PWR and WWER conditions. In each case, the data set contains information on the pre-characterisation of the fuel, cladding and fuel rod geometry, the irradiation history presented in as much detail as the source documents allow, and finally any in-pile or PIE measurements that were made. The purpose of this paper is to highlight data that are relevant specifically to WWER application. To this end, the NEA and IAEA have been successful in obtaining appropriate data for both WWER-440 and WWER-1000-type reactors. These are: 1) Twelve (12) rods from the Finnish-Russian co-operative SOFIT programme; 2) Kola-3 WWER-440 irradiation; 3) MIR ramp tests on Kola-3 rods; 4) Zaporozskaya WWER-1000 irradiation; 5) Novovoronezh WWER-1000 irradiation. Before reviewing these data sets and their usefulness, the paper touches briefly on recent, more novel additions to the Database and on progress made in the use of the Database for the current IAEA FUMEX II Project. Finally, the paper describes the Computer

Dynamic behaviour of solvent contactors in fuel reprocessing plants- an analysis

Energy Technology Data Exchange (ETDEWEB)

Raju, R P; Siddiqui, H R [Nuclear Waste Management Group, Bhabha Atomic Research Centre, Mumbai (India); Murthy, K K; Kansra, V P [Fuel Reprocessing Group, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

Fuel reprocessing plants carry out separation of useful fissile and fertile materials from spent nuclear fuels by isolating highly radioactive fission products using solvent extraction method. In the fuel reprocessing step of nuclear fuel cycle, optimisation of process parameters in the PUREX flowsheet design is of great importance particularly on account of the need to realize high degree of recovery of fissile and fertile materials and to ensure proper control on concentrations of fissile element in process streams for avoidance of criticality. In counter-current solvent contactors of PUREX flowsheet there are a variety of processes conditions which may cause plutonium accumulations that requires attention to ascertain safe Pu concentrations within the contactors. A study was carried out using the PUREX process mathematical model Solvent Extraction Program Having Interacting Solutes (SEPHIS) for pulsed solvent contactors in PREFRE-1, Tarapur and PREFRE-2, Kalpakkam flowsheets for optimising the process parameters in plutonium purification cycles. The study was extended to predict the behaviour of contactors handling plutonium bearing solutions under certain anticipated deviations in the process parameters. Modifications wherever necessary were carried out to the original SEPHIS code. This paper discusses the results obtained during this analysis. (author). 2 figs., 2 tabs.

Assessing fuel spill risks in polar waters: Temporal dynamics and behaviour of hydrocarbons from Antarctic diesel, marine gas oil and residual fuel oil.

Science.gov (United States)

Brown, Kathryn E; King, Catherine K; Kotzakoulakis, Konstantinos; George, Simon C; Harrison, Peter L

2016-09-15

As part of risk assessment of fuel oil spills in Antarctic and subantarctic waters, this study describes partitioning of hydrocarbons from three fuels (Special Antarctic Blend diesel, SAB; marine gas oil, MGO; and intermediate grade fuel oil, IFO 180) into seawater at 0 and 5°C and subsequent depletion over 7days. Initial total hydrocarbon content (THC) of water accommodated fraction (WAF) in seawater was highest for SAB. Rates of THC loss and proportions in equivalent carbon number fractions differed between fuels and over time. THC was most persistent in IFO 180 WAFs and most rapidly depleted in MGO WAF, with depletion for SAB WAF strongly affected by temperature. Concentration and composition remained proportionate in dilution series over time. This study significantly enhances our understanding of fuel behaviour in Antarctic and subantarctic waters, enabling improved predictions for estimates of sensitivities of marine organisms to toxic contaminants from fuels in the region. Copyright © 2016 Elsevier Ltd. All rights reserved.

13 CFR 119.2 - Definitions.

Science.gov (United States)

2010-01-01

... entrepreneurs, such as, but not limited to, assistance intended to enhance business planning, marketing, management, financial management skills, business operations, or assistance for the purpose of increasing... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Definitions. 119.2 Section 119.2...

High Burnup Fuel Behaviour under LOCA Conditions as Observed in Halden Reactor Experiments

International Nuclear Information System (INIS)

Kolstad, E.; Wiesenack, W.; Oberlander, B.; Tverberg, T.

2013-01-01

In the context of assessing the validity of safety criteria for loss of coolant accidents with high burnup fuel, the OECD Halden Reactor Project has implemented an integral in-pile LOCA test series. In this series, fuel fragmentation and relocation, axial gas communication in high burnup rods as affected by gap closure and fuel- clad bonding, and secondary cladding oxidation and hydriding are of major interest. In addition, the data are being used for code validation as well as model development and verification. So far, nine tests with irradiated fuel segments (burnup 40-92 MW.d.kg -1 ) from PWR, BWR and VVER commercial nuclear power plants have been carried out. The in-pile measurements and the PIE results show a good repeatability of the experiments. The paper describes the experimental setup as well as the principal features and main results of these tests. Fuel fragmentation and relocation have occurred to varying degrees in these tests. The paper compares the conditions leading to the presence or absence of fuel fragmentation, e.g., burnup and loss of constraint. Axial gas flow is an important driving force for clad ballooning, fuel relocation and fuel expulsion. The experiments have provided evidence that such gas flow can be impeded in high burnup fuel with a potential impact on the ballooning and fuel dispersal. Although the results of the Halden LOCA tests are, to some extent, amplified by conditions and features deliberately introduced into the test series, the fuel behaviour identified in the Halden tests has an impact on the safety assessment of high burnup fuel and should give rise to improvements of the predictive capabilities of LOCA modelling codes. (author)

Improvement of Computer Codes Used for Fuel Behaviour Simulation (FUMEX-III). Report of a Coordinated Research Project 2008-2012

International Nuclear Information System (INIS)

2013-03-01

It is fundamental to the future of nuclear power that reactors can be run safely and economically to compete with other forms of power generation. As a consequence, it is essential to develop the understanding of fuel performance and to embody that knowledge in codes to provide best estimate predictions of fuel behaviour. This in turn leads to a better understanding of fuel performance, a reduction in operating margins, flexibility in fuel management and improved operating economics. The IAEA has therefore embarked on a series of programmes addressing different aspects of fuel behaviour modelling with the following objectives: - To assess the maturity and prediction capabilities of fuel performance codes, and to support interaction and information exchange between countries with code development and application needs (FUMEX series); - To build a database of well defined experiments suitable for code validation in association with the OECD Nuclear Energy Agency (OECD/NEA); - To transfer a mature fuel modelling code to developing countries, to support teams in these countries in their efforts to adapt the code to the requirements of particular reactors, and to provide guidance on applying the code to reactor operation and safety assessments; - To provide guidelines for code quality assurance, code licensing and code application to fuel licensing. This report describes the results of the coordinated research project on the ''Improvement of computer codes used for fuel behaviour simulation (FUMEX-III)''. This programme was initiated in 2008 and completed in 2012. It followed previous programmes on fuel modelling: D-COM 1982-1984, FUMEX 1993-1996 and FUMEX-II 2002-2006. The participants used a mixture of data derived from commercial and experimental irradiation histories, in particular data designed to investigate the mechanical interactions occurring in fuel during normal, transient and severe transient operation. All participants carried out calculations on priority

49 CFR 41.119 - DOT regulated buildings.

Science.gov (United States)

2010-10-01

... 49 Transportation 1 2010-10-01 2010-10-01 false DOT regulated buildings. 41.119 Section 41.119 Transportation Office of the Secretary of Transportation SEISMIC SAFETY § 41.119 DOT regulated buildings. (a) Each DOT Operating Administration with responsibility for regulating the structural safety of buildings...

Molten salt burner fuel behaviour and treatment

International Nuclear Information System (INIS)

Ignatiev, V.V.; Zakirov, R.Y.; Grebenkine, K.F.

2001-01-01

The objective of this paper is to discuss the feasibility of molten salt reactor technology for treatment of Pu, minor actinides and fission products, when the reactor and fission product clean-up unit are planned as an integral system. This contribution summarises the available R and D which led to selection of the fuel compositions for the molten salt reactor of the TRU burner type (MSB). Special characteristics of behaviour of TRUs and fission products during power operation of MSB concepts are presented. The present paper briefly reviews the processing developments underlying the prior molten salt reactor programmes and relates them to the separation requirements of the MSB concept, including the permissible range of processing cycle times and removal times. Status and development needs in the thermodynamic properties of fluorides, fission product clean-up methods and container materials compatibility with the working fluids for the fission product clean-up unit are discussed. (authors)

Behaviour in air at 175-400 degrees C of irradiated UO2 fuel

International Nuclear Information System (INIS)

Hastings, I.J.; McCracken, D.

1984-09-01

The authors extended their study of irradiated, defected UO 2 fuel elements to 200 and 400 degrees C. At 200 degrees C there was no diametral change, but at 400 degrees C we observed swelling and severe sheath splitting. Neither short-lived fission products, nor Cs-134, Cs-137 or Ru-106 above background, were detected. Maximum Kr-85 release was 4 Bq ( -6 Ci). Discharge time was 2.5 years. UO 2 fragment studies were extended to 400 degrees C. The oxidation process for unirradiated and irradiated fuel up to 300 degrees C was characterized by activation energies of 140 +- 10 and 120 +- 10 kJ/mol, respectively; enhancement of oxidation rate was confirmed in the irradiated samples. There is an apparent reduction of activation energy above about 300 degrees C. Fuel elements with artificial and natural defects showed similar oxidation and dimensional response at 250 degrees C. Behaviour of fuel fragments from the defect area of a naturally-defected element is consistent with that for fragments from intact elements when prior oxidation during the defect period is considered

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

Energy Technology Data Exchange (ETDEWEB)

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.

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

International Nuclear Information System (INIS)

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.

In-core instrumentation and in-situ measurement in connection with fuel behaviour. Working material

International Nuclear Information System (INIS)

1996-01-01

The subject of this meeting has been touched on briefly in most of the Specialist's and topical meetings related to fuel behaviour. On the basis of the conclusions and recommendations of these meetings the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended the Agency to organize a dedicated Specialist's Meeting on the subject. The twenty one papers covered the instrumentation, sensors, methods and computer codes currently used in Material Test Reactor (MTR) and power reactors as well as improved instrumentation and methods. The meeting acknowledged the fast development of fuel modelling and therefore the growing need of dedicated high burnup fuel experiments carried out in MTR reactors on refabricated rods from power reactors. In order to reduce safety margins in power reactors, thus improving economics, the necessity to develop more sophisticated on-line calculations, based on improved sensors, was recognized, although this development is limited by insufficient knowledge of the mechanisms involved. Refs, figs, tabs

17 CFR 248.103-248.119 - [Reserved

Science.gov (United States)

2010-04-01

... 17 Commodity and Securities Exchanges 3 2010-04-01 2010-04-01 false [Reserved] 248.103-248.119 Section 248.103-248.119 Commodity and Securities Exchanges SECURITIES AND EXCHANGE COMMISSION (CONTINUED) REGULATIONS S-P AND S-AM Regulation S-AM: Limitations on Affiliate Marketing §§ 248.103-248.119 [Reserved] ...

Starting Point, Keys and Milestones of a Computer Code for the Simulation of the Behaviour of a Nuclear Fuel Rod

Directory of Open Access Journals (Sweden)

Armando C. Marino

2011-01-01

Full Text Available The BaCo code (“Barra Combustible” was developed at the Atomic Energy National Commission of Argentina (CNEA for the simulation of nuclear fuel rod behaviour under irradiation conditions. We present in this paper a brief description of the code and the strategy used for the development, improvement, enhancement, and validation of a BaCo during the last 30 years. “Extreme case analysis”, parametric (or sensitivity, probabilistic (or statistic analysis plus the analysis of the fuel performance (full core analysis are the tools developed in the structure of BaCo in order to improve the understanding of the burnup extension in the Atucha I NPP, and the design of advanced fuel elements as CARA and CAREM. The 3D additional tools of BaCo can enhance the understanding of the fuel rod behaviour, the fuel design, and the safety margins. The modular structure of the BaCo code and its detailed coupling of thermo-mechanical and irradiation-induced phenomena make it a powerful tool for the prediction of the influence of material properties on the fuel rod performance and integrity.

Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

International Nuclear Information System (INIS)

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

Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

Energy Technology Data Exchange (ETDEWEB)

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.

Modelling Spent Fuel and HLW Behaviour in Repository Conditions. A review of the state of the art

International Nuclear Information System (INIS)

Martinez-Esparza, A.; Esteban, J. A.; Quinones, J.; Pablo, J. de; Casas, I.; Gimenez, J.; Clarens, F.; Rovira, M.; Merino, J.; Cera, E.; Bruno, J.; Ripoll, S.

2002-01-01

The SFS (Spent Fuel Stability) project that is being carried out as part of the European Union's 5th Framework Programme has a dual objective, technical and sociologic. The technical objectives consists of developing a model of the behaviour of irradiated fuel and of the radionuclides contained therein, under the conditions of a deep geological disposal facility, incorporating the experimental part of this and previous. European projects. The sociological objectives is to develop a common scientific and technical opinion throughout the European Union, for consensus to be reached regarding the evolution of a deep geological disposal facility for high level wastes. With a view to achieving this dual objective, and as a project activity, a Seminar was organised in Avila in June 2002 (the presentations made of this Seminar will be the subject of another publication), the aim being to establish the bases for a new spent fuel behaviour model with and ample experimental basis and the consensus of the European countries participating in the project (France, Switzerland, Germany, Sweden, Belgium and Spain. (Author)

Efeito da tenacidade da fibra sobre propriedades tecnológicas do fio de algodão Effect of the cotton fiber strength on yarn properties

Directory of Open Access Journals (Sweden)

Nelson Paulieri Sabino

1995-01-01

Full Text Available Consideraram-se três variedades de algodão com valores de tenacidade da fibra variando de 20,5 a 22,2 g/Tex: IAC 16, IAC 13-1 e IAC 17, classificadas, respectivamente, como de alta, média e baixa tenacidade. Tais variedades apresentaram características tecnológicas semelhantes quanto a comprimento, uniformidade de comprimento, índice de finura Micronaire e maturidade. As amostras foram processadas em estabelecimentos industriais, da maneira convencional, produzindo, cada uma, fios de títulos Ne20, Ne30 e Ne40. Para cada título, empregaram-se sete coeficientes de torção, representados pelas constantes 3,4, 3,6, 3,8, 4,0, 4,2, 4,5 e 4,7. Efetuaram-se as análises da variância dos resultados, de acordo com o delineamento fatorial 3 x 3 x 7, representado pelas três variedades, pelos três títulos e pelos sete níveis de coeficientes de torção. Mediante os resultados, conclui-se que fibras de algodão com alta tenacidade produzem fios mais resistentes e elásticos do que aquelas de baixa tenacidade, para qualquer título ou torção. A quantidade de torções requeridas para a obtenção de máxima resistência dos fios de algodão é pouco afetada pela tenacidade da fibra. Os fios de títulos mais altos têm os menores valores de tenacidade e elongação. A variedade IAC 16 apresentou fios com os maiores valores de tenacidade, seguida da 'IAC 13-1' e da 'IAC 17', e fios mais elásticos, acompanhada da 'IAC 17' e da 'IAC 13-1'.Three cottons with fiber strength of 20.5, 20.9 and 22.2 g/Tex and having other important fiber properties approximately equal were selected. The cottons were processed on conventional processing equipment into 20/1, 30/1 and 40/1 yarn counts, using a range of twist multipliers of 3.4, 3.6, 3.8, 4.0, 4.2, 4.5 and 4.7. Yarn strength and elongation determinations were made on a pendulum-type tester of 150-300 lbs capacity. It was found that: 1 - High strength cotton produced stronger yarns than low strength for any

SSYST, Modular System for Transient Fuel Rod Behaviour Under Accident Condition

International Nuclear Information System (INIS)

Gulden, W.; Meyder, R.; Borgwaldt, H.

1987-01-01

1 - Description of problem or function: SSYST is a code system for analyzing transient fuel rod behaviour under off-normal conditions, developed jointly by the Institut fuer Kernenergetik und Energie-systeme (IKE), Stuttgart, and Kernforschungszentrum Karlsruhe (KfK) under contract for the Projekt Nukleare Sicherheit (PNS) at KfK. Main differences versus codes with similar applications are: (1) an open-ended modular code organisation; (2) a preference for simple models, wherever possible. While feature (1) makes SSYST a very flexible tool, easily adapted to changing requirements, feature (2) leads to short execution times. The analysis of transient rod behaviour under LOCA boundary conditions takes 2 minutes CPU time on IBM 3033, so that extensive parametric studies are feasible. Main differences between SSYST-3 and previous versions are related to a general clean-up of the code system, which reduces the implementation effort: - advanced modules for cladding deformation and oxidation and reflooding conditions are included; - an input processor thoroughly checks all input data

COMETHE III J a computer code for predicting mechanical and thermal behaviour of a fuel pin

International Nuclear Information System (INIS)

Verbeek, P.; Hoppe, N.

1976-01-01

The design of fuel pins for power reactors requires a realistic evaluation of their thermal and mechanical performances throughout their irradiation life. This evaluation involves the knowledge of a number of parameters, very intricate and interconnected, for example, the temperature, the restructuring and the swelling rates of the fuel pellets, the dimensions, the stresses and the strains in the clad, the composition and the properties of gases, the inner gas pressure etc. This complex problem can only be properly handled by a computer programme which analyses the fuel pin thermal and mechanical behaviour at successive steps of its irradiation life. This report presents an overall description of the COMETHE III-J computer programme, designed to calculate the integral performance of oxide fuel pins with cylindrical metallic cladding irradiated in thermal or fast flux. (author)

Investigation of Solitary wave solutions for Vakhnenko-Parkes equation via exp-function and Exp(-ϕ(ξ))-expansion method.

Science.gov (United States)

Roshid, Harun-Or; Kabir, Md Rashed; Bhowmik, Rajandra Chadra; Datta, Bimal Kumar

2014-01-01

In this paper, we have described two dreadfully important methods to solve nonlinear partial differential equations which are known as exp-function and the exp(-ϕ(ξ)) -expansion method. Recently, there are several methods to use for finding analytical solutions of the nonlinear partial differential equations. The methods are diverse and useful for solving the nonlinear evolution equations. With the help of these methods, we are investigated the exact travelling wave solutions of the Vakhnenko- Parkes equation. The obtaining soliton solutions of this equation are described many physical phenomena for weakly nonlinear surface and internal waves in a rotating ocean. Further, three-dimensional plots of the solutions such as solitons, singular solitons, bell type solitary wave i.e. non-topological solitons solutions and periodic solutions are also given to visualize the dynamics of the equation.

Implementação de sistemas eólicos para redes de sensores sem fios

OpenAIRE

Mendonça, Fábio Rúben Silva

2015-01-01

O trabalho tem por objetivo principal estudar a utilização de aerogeradores para microprodução de energia (menor que 1 W) de forma a poderem alimentar os nós das redes de sensores sem fios. Para tal, analisaram-se dois tipos de aerogeradores, os de eixo vertical e os de eixo horizontal, tendo-se efetuado simulações de CFD (Computational Fluid Dynamics) para verificar o desempenho das turbinas, de forma a determinar qual é o mais adequado. Foi realizado o projeto dos dois sistemas e anali...

Roteamento em redes em malha sem fio com balanceamento de carga e caminhos mais curtos

OpenAIRE

Mello, Micael Oliveira Massula Carvalho de

2014-01-01

Redes em Malha Sem Fio - Wireless Mesh Networks (WMNs) são infraestruturas com propriedades autonômicas, como auto-organização e autorrecuperação, que podem ser implementadas com tecnologias amplamente disponíveis e de custo acessível. Além de suas aplicações atuais, como redes comunitárias e redes de acesso à Internet, as WMNs podem auxiliar na comunicação de Internet das Coisas e constituir infraestruturas robustas para redes inteligentes de energia, dentre outros usos. No en...

7 CFR 57.119 - Political activity.

Science.gov (United States)

2010-01-01

... Political activity. Federal inspectors may participate in certain political activities, including management and participation in political campaigns as allowed by Federal regulation and AMS directives... 7 Agriculture 3 2010-01-01 2010-01-01 false Political activity. 57.119 Section 57.119 Agriculture...

MOX fuel effective behaviour modeling by a micro-mechanical nonuniform transformation field analysis

International Nuclear Information System (INIS)

Largenton, R.

2012-01-01

The objective of this research thesis is to develop a modelling by scale change, based on the NTFA approach (Non uniform Transformation Field Analysis). These developments have been achieved on three-dimensional structures which are representative of the MOX fuel, and for local visco-elastic ageing behaviour with free deformations. First, the MOX fuel is represented by using existing methods to process and segment 2D experimental images. 2D information has been upgraded in 3D by a stereo-logic Saltykov method. Tools have been developed to represent and discretize (periodic 3D grid generator) a particulate multiphase composite representative of MOX. Developments made on the NTFA model and on the three-phase particulate composite have been theoretically and numerically studied. The model has then been validated by comparison with reference calculations performed in full field for the effective behaviour as well as for local fields for different test types (imposed strain rate, creep, relaxation, rotating). The approach is then compared with a recently developed homogenisation method: the semi-analytical 'incremental Mori-Tanka' model. Theoretical similarities are outlined. These methods are very fast in terms of CPU time, but the NTFA method remains the one giving the most information, and the most precise, but requires a more important preliminary work (mode identification) [fr

Seismic behaviour of fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Song, Heuy Gap; Jhung, Myung Jo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1993-11-01

A general approach for the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced from earthquake. The dynamic responses such as fuel assembly shear force, bending moment and displacement, and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed. (Author) 9 refs., 24 figs., 1 tab.

Quorum sensing in the plant pathogen Erwinia carotovora subsp. carotovora: the role of expR(Ecc).

Science.gov (United States)

Andersson, R A; Eriksson, A R; Heikinheimo, R; Mäe, A; Pirhonen, M; Kõiv, V; Hyytiäinen, H; Tuikkala, A; Palva, E T

2000-04-01

The production of the main virulence determinants of the plant pathogen Erwinia carotovora subsp. carotovora, the extracellular cell wall-degrading enzymes, is partly controlled by the diffusible signal molecule N-(3-oxohexanoyl)-L-homoserine lactone (OHHL). OHHL is synthesized by the product of the expI/carI gene. Linked to expI we found a gene encoding a putative transcriptional regulator of the LuxR-family. This gene, expR(Ecc), is transcribed convergently to the expI gene and the two open reading frames are partially overlapping. The ExpR(Ecc) protein showed extensive amino acid sequence similarity to the repressor EsaR from Pantoea stewartii subsp. stewartii (formerly Erwinia stewartii subsp. stewartii) and to the ExpR(Ech) protein of Erwinia chrysanthemi. Inactivation of the E. carotovora subsp. carotovora expR(Ecc) gene caused no decrease in virulence or production of virulence determinants in vitro. In contrast, there was a slight increase in the maceration capacity of the mutant strain. The effects of ExpR(Ecc) were probably mediated by changes in OHHL levels. Inactivation of expR(Ecc) resulted in increased OHHL levels during early logarithmic growth. In addition, overexpression of expR(Ecc) caused a clear decrease in the production of virulence determinants and part of this effect was likely to be caused by OHHL binding to ExpR(Ecc). ExpR(Ecc) did not appear to exhibit transcriptional regulation of expI, but the effect on OHHL was apparently due to other mechanisms.

Exact solutions for nonlinear evolution equations using Exp-function method

International Nuclear Information System (INIS)

Bekir, Ahmet; Boz, Ahmet

2008-01-01

In this Letter, the Exp-function method is used to construct solitary and soliton solutions of nonlinear evolution equations. The Klein-Gordon, Burger-Fisher and Sharma-Tasso-Olver equations are chosen to illustrate the effectiveness of the method. The method is straightforward and concise, and its applications are promising. The Exp-function method presents a wider applicability for handling nonlinear wave equations

SSYST, a code-system for analysing transient LWR fuel rod behaviour under off-normal conditions

International Nuclear Information System (INIS)

Borgwaldt, H.; Gulden, W.

1983-01-01

SSYST is a code-system for analysing transient fuel rod behaviour under off-normal conditions, developed conjointly by the Institut fuer Kernenergetik und Energiesysteme (IKE), Stuttgart, and Kernforschungszentrum Karlsruhe (KfK) under contract of Projek Nukleare Sicherheit (PNS) at KfK. The main differences between SSYST and similar codes are (1) an open-ended modular code organisation, and (2) a preference for simple models, wherever possible. While the first feature makes SSYST a very flexible tool, easily adapted to changing requirements, the second feature leads to short execution times. The analysis of transient rod behaviour under LOCA boundary conditions takes 2 min cpu-time (IBM-3033), so that extensive parametric studies become possible. This paper gives an outline of the overall code organisation and a general overview of the physical models implemented. Besides explaining the routine application of SSYST in the analysis of loss-of-coolant accidents, examples are given of special applications which have led to a satisfactory understanding of the decisive influence of deviations from rotational symmetry on the fuel rod perimeter. (author)

SSYST: A code-system for analyzing transient LWR fuel rod behaviour under off-normal conditions

International Nuclear Information System (INIS)

Borgwaldt, H.; Gulden, W.

1983-01-01

SSYST is a code-system for analyzing transient fuel rod behaviour under off-normal conditions, developed conjointly by the Institut fur Kernenergetik und Energiesysteme (IKE), Stuttgart, and Kernforschungszentrum Karlsruhe (KfK) under contract of Projekt Nukleare Sicherheit (PNS) at KfK. The main differences between SSYST and similar codes are an open-ended modular code organization, and a preference for simple models, wherever possible. While the first feature makes SSYST a very flexible tool, easily adapted to changing requirements, the second feature leads to short execution times. The analysis of transient rod behaviour under LOCA boundary conditions takes 2 min cpu-time (IBM-3033), so that extensive parametric studies become possible. This paper gives an outline of the overall code organisation and a general overview of the physical models implemented. Besides explaining the routine application of SSYST in the analysis of loss-of-coolant accidents, examples are given of special applications which have led to a satisfactory understanding of the decisive influence of deviations from rotational symmetry on the fuel rod perimeter

Research problems of fission product behaviour in fuels of nuclear power plants and ways of their solution

International Nuclear Information System (INIS)

Sulaberidze, V.Sh.

1988-01-01

The most important problems of studying behaviour of fission products in fuel elements of maneouvrable nuclear power plants units are formulated. In-pile and out-of-pile investigation methods solving these problems are characterized in brief. 12 refs.; 2 figs

GPR119 as a fat sensor

DEFF Research Database (Denmark)

Hansen, Harald S; Rosenkilde, Mette M; Holst, Jens Juul

2012-01-01

acting through, for example, GPR40, but is also probably mediated in large part through the luminal formation of 2-monoacylglycerol acting on the 'fat sensor' GPR119. In the pancreas GPR119 may also be stimulated by 2-monoacylglycerol generated from local turnover of pancreatic triacylglycerol. Knowledge...

9 CFR 590.119 - Political activity.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Political activity. 590.119 Section 590.119 Animals and Animal Products FOOD SAFETY AND INSPECTION SERVICE, DEPARTMENT OF AGRICULTURE EGG... party or candidate, except as authorized by law or regulation of the Department, is prohibited. This...

Studies on the safety and transmutation behaviour of innovative fuels for light water reactors

International Nuclear Information System (INIS)

Schitthelm, Oliver

2012-01-01

Nuclear power plants contribute a substantial part to the energy demand in industry. Today the most common fuel cycle uses enriched uranium which produces plutonium due to its 238 U content. With respect to the long-term waste disposal Plutonium is an issue due to its heat production and radiotoxicity. This thesis consists of three main parts. In the first part the development and validation of a new code package MCBURN for spatial high resolution burnup simulations is presented. In the second part several innovative uranium-free and plutonium-burning fuels are evaluated on assembly level. Candidates for these fuels are a thorium/plutonium fuel and an inert matrix fuel consisting of plutonium dispersed in an enriched molybdenum matrix. The performance of these fuels is evaluated against existing MOX and enriched uranium fuels considering the safety and transmutation behaviour. The evaluation contains the boron efficiency, the void coefficient, the doppler coefficient and the net balances of every radionuclide. In the third part these innovative fuels are introduced into a German KONVOI reactor core. Considering todays approved usage of MOX fuels a partial loading of one third of innovative fuels and two third of classical uranium fuels was analysed. The efficiency of the plutonium depletion is determined by the ratio of the production of higher isotopes compared to the plutonium depletion. Todays MOX-fuels transmutate about 25% to 30% into higher actinides as Americium or Curium. In uranium-free fuels this ratio is about 10% due to the lack of additional plutonium production. The analyses of the reactor core have shown that one third of MOX fuel is not capable of a net reduction of plutonium. On the other hand a partial loading with thorium/plutonium fuel incinerates about half the amount of plutonium produced by an uranium only core. If IMF is used the ratio increases to about 75%. Considering the safety behavior all fuels have shown comparable results.

Seismic behaviour of PWR fuel assemblies model and its validation

International Nuclear Information System (INIS)

Queval, J.C.; Gantenbein, F.; Brochard, D.; Benjedidia, A.

1991-01-01

The validity of the models simulating the seismic behaviour of PWR cores can only be exactly demonstrated by seismic testing on groups of fuel assemblies. Shake table seismic tests of rows of assembly mock-ups, conducted by the CEA in conjunction with FRAMATOME, are presented in reference /1/. This paper addresses the initial comparisons between model and test results for a row of five assemblies in air. Two models are used: a model with a single beam per assembly, used regularly in accident analyses, and described in reference /2/, and a more refined 2-beam per assembly model, geared mainly towards interpretation of test results. The 2-beam model is discussed first, together with parametric studies used to characterize it, and the study of the assembly row for a period limited to 2 seconds and for different excitation levels. For the 1-beam model assembly used in applications, the row is studied over the total test time, i.e twenty seconds, which covers the average duration of the core seismic behaviour studies, and for a peak exciting acceleration value at 0.4 g, which corresponds to the SSE level of the reference spectrum

The Effect of Lactate, Albumin, C-reactive Protein, PaO2/FiO2 and Glucose Levels of Trauma Patients at the Time of Administration to Intensive Care

Directory of Open Access Journals (Sweden)

Eren Yılmaz

2014-12-01

Full Text Available Objective: Blood analyses are preferred in the observation of cases requiring intensive care unit (ICU following a trauma. The purpose of this study was to examine the relationship of albumin, C-reactive protein (CRP, PaO2/FiO2 and glucose levels of trauma patients at time of admission with mortality. Material and Method: The patients who were admitted into ICU following a trauma between the years of 2010 and 2012 were retrospectively evaluated. 200 trauma cases were included in the study. Their demographic data, APACHE II scores, Glasgow Coma Scales (GCS, and arterial blood gas in the lactate and PaO2/FiO2 ratio, CRP, glucose and albumin levels in the first collected arterial blood gas, as well as, the presence of thoracic, cardiac, renal, abdominal and head trauma, length of ICU stay and mortality were recorded. Results: Of the patients included in the study 84% were male, with an average age of 38.3 and an average APACHE II score of 16.6. 64% suffered from head trauma and the average GCS was calculated to be 11.2. The patients were observed in the ICU for an average of 18.7 days and the rate of mortality was 33.5%. GCS, PaO2/FiO2, age and elevated lactate levels increased mortality as independent risk factors. Conclusion: It has been concluded that parameters like age and the first GCS, lactate, glucose, albumin and PaO2/FiO2 at time of acceptance into the ICU were found to be related with mortality.

Peak Pressures and PaO2/FiO2 Ratios Are Associated With Adverse Outcomes in Patients on Mechanical Ventilators.

Science.gov (United States)

Whiting, Jeremy; Edriss, Hawa; Yang, Shengping; Nugent, Kenneth

2016-06-01

Patients requiring mechanical ventilation can have complications related to their underlying diseases and hospital-related events. It is possible that easily obtained information early in the course of mechanical ventilation can provide information about important outcomes. Medical records from 281 episodes of mechanical ventilation in the medical intensive care unit were reviewed to collect information on patient demographics, admitting diagnoses, laboratory tests, duration of mechanical ventilation, the development of ventilator-associated events and mortality. Ventilator pressures from day 2 were analyzed for this study. Most patients (72.7%) were ≥50 years, 53.8% were men and 66.3% had a body mass index (BMI) ≥ 25kg/m(2).The mean Acute Physiology and Chronic Healthy Evaluation II score was 13.6 ± 5.9. The median initial PaO2/FiO2 was 240 with interquartile range of 177-414. The median duration of ventilation was 4 days (interquartile range: 2-9 days). A PaO2/FiO2 ratio 500, and a BMI > 30kg/m(2) was associated with decreased mortality compared with normal BMIs. A PaO2/FiO2 ratio 30kg/m(2) were all associated with having a ventilator-associated event. There was a positive correlation between peak pressure (day 2) and the duration of ventilation (r = 0.263, P = 0.007). Easily available information collected on day 2 of mechanical ventilation can help identify patients at risk for poor outcomes, including the duration of mechanical ventilation, the development of ventilator-associated complications and mortality. Prospective studies measuring peak pressures are needed to evaluate the utility of this simple measurement in the management of patients requiring mechanical ventilation. Published by Elsevier Inc.

Controle dos vasos renais usando clips vasculares e fio cirúrgico em nefrectomias vídeo-assistidas de doadores vivos

Directory of Open Access Journals (Sweden)

Alcides José Branco Filho

Full Text Available OBJETIVO: A nefrectomia laparoscópica em doadores vivos para transplante renal vem assumindo um papel importante na era das cirurgias minimamente invasivas, acarretando menor morbidade aos doadores, e resultados semelhantes à técnica aberta no que se refere ao enxerto renal. O objetivo do presente artigo é relatar a experiência do nosso serviço utilizando a técnica de controle dos vasos renais usando fio cirúrgico e clips vasculares. MÉTODO: Foram realizadas 45 nefrectomias utilizando a técnica vídeo-assistida, com ligadura dos vasos renais com clips de titânio (LT-300 e fio cirúrgico. As variáveis analisadas foram tempo cirúrgico, perda sangüínea, tempo de isquemia quente, permanência hospitalar, necessidade de conversão e complicações. RESULTADOS: O procedimento foi realizado com sucesso em todos os casos. O tempo cirúrgico médio foi de 118 minutos, com perda sangüínea estimada em 84ml e tempo de isquemia quente de 4,3 minutos. Dois casos de íleo prolongado, uma lesão de veia gonadal, um escape de artéria renal e uma necrose de ureter foram observados. A permanência hospitalar média foi de 3,7 dias. O uso de clips vasculares e fio cirúrgico reduziu a perda de tecido venoso comparado à técnica com staplers e gerou redução de custos. CONCLUSÕES: A nefrectomia vídeo-assistida com a técnica descrita é factível e mostrou ser efetiva na contenção de gastos e na redução de tecido venoso perdido.

Reactivity And Neutron Flux At Silicide Fuel Element In The Core Of RSG-GAS

International Nuclear Information System (INIS)

Hamzah, Amir

2000-01-01

In order to 4.8 and 5.2 gr U/cm exp 3 loading of U 3 Si 2 --Al fuel plates characterization, he core reactivity change and neutron flux depression had been done. Control rod calibration method was used to reactivity change measurement and neutron flux distribution was measured using foil activation method. Measurement of insertion of A-type of testing fuel element with U-loading above cannot be done due to technical reason, so the measurement using full type silicide fuel element of 2.96 gr U/cm exp 3 loading. The reactivity change measurement result of insertion in A-9 and C-3 is + 2.67 cent. The flux depression at silicide fuel in A-9 is 1.69 times bigger than oxide and in C-3 is 0.68 times lower than oxide

Fios ortodônticos: conhecer para otimizar a aplicação clínica Orthodontic wires: knowledge to optimize clinical application

Directory of Open Access Journals (Sweden)

Cátia Cardoso Abdo Quintão

2009-12-01

Full Text Available A grande variedade de fios ortodônticos presente no mercado pode gerar dúvidas quanto à melhor escolha para situações clínicas. Assim, o conhecimento das propriedades mecânicas dos mesmos facilita a escolha para aplicação do movimento ortodôntico na dependência da fase em que o tratamento se encontra. A evolução da tecnologia de manufatura dos fios e a elaboração de novas técnicas ortodônticas geraram a busca por uma melhor qualidade das ligas, a fim de torná-los biologicamente mais efetivos no que diz respeito aos dentes e tecidos de suporte. O presente artigo resume as principais características dos fios utilizados em Ortodontia, em relação ao histórico, propriedades mecânicas e aplicação clínica, de acordo com fases específicas de tratamento.The huge variety of orthodontic wires brands available in market might generate confusion as regard to the best choice for clinical application. Therefore mechanical properties knowledge about wires would help the professional to apply the best orthodontic technique depending on the treatment phase. The wires manufacturing evolution and the new orthodontic techniques proposed guided the market into the search for better quality alloys, in order to make them biologically more effective to teeth and support tissues. This paper aims to summarize some main characteristics of orthodontic wires related to their history, mechanical properties and clinical application as regard to individual phase of treatment.

SSYST-1. A computer code system to analyse the fuel rod behaviour during a loss of coolant accident

International Nuclear Information System (INIS)

Gulden, W.

1977-08-01

The modules of the SSYST program system allow the detailed analysis of an LWR fuel rod in the course of a postulated loss-of-coolant accident. They provide a tool for considering the interaction between the heat conduction in the fuel rod, heat transfer in the gap, fuel and cladding tube deformation, pressure in the coolant, as well as thermal and fluid dynamics in the cooling channel and for calculating the time and location of ballooning and rod failure, respectively. They can be used both to precalculate the behaviour of fuel rods during LWR accidents and in support of the design of experiments. Depending on the problem to be solved, the individual modules can be easily combined. (orig.) [de

Uncoordinated (UNC)119: coordinating the trafficking of myristoylated proteins.

Science.gov (United States)

Constantine, Ryan; Zhang, Houbin; Gerstner, Cecilia D; Frederick, Jeanne M; Baehr, Wolfgang

2012-12-15

The mechanism by which myristoylated proteins are targeted to specific subcellular membrane compartments is poorly understood. Two novel acyl-binding proteins, UNC119A and UNC119B, have been shown recently to function as chaperones/co-factors in the transport of myristoylated G protein α-subunits and src-type tyrosine kinases. UNC119 polypeptides feature an immunoglobulin-like β-sandwich fold that forms a hydrophobic pocket capable of binding lauroyl (C12) and myristoyl (C14) side chains. UNC119A in rod photoreceptors facilitates the transfer of transducin α subunits (Tα) from inner segment to outer segment membranes by forming an intermediate diffusible UNC119-Tα complex. Similar complexes are formed in other sensory neurons, as the G proteins ODR-3 and GPA-13 in Caenorhabditis elegans unc-119 mutants traffic inappropriately. UNC119B knockdown in IMCD3 cells prevents trafficking ofmyristoylated nephrocystin-3 (NPHP3), a protein associated with nephronophthisis, to cilia. Further, UNC119A was shown to transport myristoylated src-type tyrosine kinases to cell membranes and to affect T-cell receptor (TCR) and interleukin-5 receptor (IL-5R) activities. These interactions establish UNC119 polypeptides as novel lipid-binding chaperones with specificity for a diverse subset of myristoylated proteins. Copyright © 2012 Elsevier Ltd. All rights reserved.

Aprediction study for the behaviour of fuel cell membrane subjected to hygro and thermal stresses in running PEM fuel cell

OpenAIRE

Maher A.R. Sadiq Al-Baghdadi

2016-01-01

A three-dimensional, multi–phase, non-isothermal computational fluid dynamics model of a proton exchange membrane fuel cell has been used and developed to investigate the hygro and thermal stresses in polymer membrane, which developed during the cell operation due to the changes of temperature and relative humidity. The behaviour of the membrane during operation of a unit cell has been studied and investigated under real cell operating conditions. The results show that the non-uniform distrib...

Fricção em braquetes gerada por fios de aço inoxidável, superelásticos com IonGuard e sem IonGuard Friction force on brackets generated by stainless steel wire and superelastic wires with and without IonGuard

Directory of Open Access Journals (Sweden)

Luiz Carlos Campos Braga

2011-08-01

Full Text Available OBJETIVO: o objetivo deste estudo foi verificar a fricção no braquete (Roth, Composite, 10.17.005, 3,2mm, largura 0,022" x 0,030", Torque -2° e angulação +13°, Morelli®, Brasil, utilizando fios ortodônticos retangulares de 0,019" x 0,025" de aço inoxidável (Morelli®, Brasil e de níquel-titânio superelásticos Bioforce com IonGuard e sem IonGuard (Bioforce, GAC®, EUA. MÉTODOS: foram utilizados 24 conjuntos braquetes/segmento de fio, divididos em 3 grupos de acordo com o fio. Cada conjunto braquete/segmento de fio foi testado 3 vezes e obtida uma média. Os ensaios foram realizados em máquina universal de ensaios EMIC DL2000®. Os dados foram submetidos à Análise de Variância com significância de 95%. RESULTADOS: o fio retangular Bioforce com IonGuard apresentou fricção significativamente menor que o Bioforce sem IonGuard, porém sem diferença do fio de aço inoxidável. Entretanto, o coeficiente de variação dos fios Bioforce com e sem IonGuard foi menor que o do fio de aço inoxidável. CONCLUSÃO: os fios retangulares de 0,019" x 0,025" Bioforce com IonGuard apresentam menor fricção que o fio Bioforce sem IonGuard, sem diferença para o fio de aço inoxidável.OBJECTIVE: The aim of this study was to evaluate the friction forces on brackets (Roth, Composite, 10.17.005, 3.2 mm, width 0.022" x 0.030 ", Torque -2° and angulation +13°, Morelli®, Brazil, with stainless steel orthodontic rectangular wire (Morelli®, Brazil and nickel titanium superelastic Bioforce wires with and without IonGuard (Bioforce, GAC®, USA. MATERIAL AND METHODS: Twenty-four brackets/wire segment combinations were used, distributed into three groups according to the orthodontic wire. Each bracket/wire segment combination was tested three times. The tests were performed in a universal testing machine Emic DL2000®. The data was submitted to ANOVA one way followed by Tukey's post hoc test (p<0.05. RESULTS: The rectangular orthodontic Bioforce wire

Exp2 polymorphisms associated with variation for fiber quality properties in cotton (Gossypium spp.

Directory of Open Access Journals (Sweden)

Daohua He

2014-10-01

Full Text Available Plant expansins are a group of extracellular proteins thought to affect the quality of cotton fibers. Previous expression profile analysis revealed that six Expansin A genes are present in cotton, of which two (GhExp1 and GhExp2 produce transcripts that are specific to the developing cotton fiber. To identify the phenotypic function of Exp2, and to determine whether nucleotide variation among alleles of Exp2 affects fiber quality, candidate gene association mapping was conducted. Gene-specific primers were designed to amplify the Exp2 gene. By amplicon sequencing, the nucleotide diversity of Exp2 was investigated across 92 accessions (including 7 Gossypium arboreum, 74 Gossypium hirsutum, and 11 Gossypium barbadense accessions with different fiber qualities. Twenty-six SNPs and seven InDels including 14 from the coding region of Exp2 were detected, forming twelve distinct haplotypes in the cotton collection. Among the 14 SNPs in the coding region, five were missense mutations and nine were synonymous nucleotide changes. The average SNP/InDel per nucleotide ratio was 2.61% (one SNP per 39 bp, with 1.81 and 3.87% occurring in coding and non-coding regions, respectively. Nucleotide and haplotype diversity across the entire Exp2 region was 0.00603 (π and 0.844, respectively, and diversity in non-coding regions was higher than that in coding regions. For linkage disequilibrium (LD, the mean r2 value for all polymorphism loci pairs was 0.48, and LD did not decay over 748 bp. Based on 132 simple sequence repeat (SSR loci evenly covering 26 chromosomes, the population structure was estimated, and the accessions were divided into seven groups that agreed well with their genomic origin and evolutionary history. A general linear model was used to calculate the Exp2-wide diversity–trait associations of 5 fiber quality traits, considering population structure (Q. Four SNPs in Exp2 were associated with at least one of the fiber quality traits, but not with

46 CFR 119.430 - Engine exhaust pipe installation.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Engine exhaust pipe installation. 119.430 Section 119... INSTALLATION Specific Machinery Requirements § 119.430 Engine exhaust pipe installation. (a) The design of all... an exhaust pipe. (b) Exhaust gas must not leak from the piping or any connections. The piping must be...

Fuel accumulation and forest structure change following hazardous fuel reduction treatments throughout California

Science.gov (United States)

Nicole M. Vaillant; Erin K. Noonan-Wright; Alicia L. Reiner; Carol M. Ewell; Benjamin M. Rau; Josephine A. Fites-Kaufman; Scott N. Dailey

2015-01-01

Altered fuel conditions coupled with changing climate have disrupted fire regimes of forests historically characterised by high-frequency and low-to-moderate-severity fire. Managers use fuel treatments to abate undesirable fire behaviour and effects. Short-term effectiveness of fuel treatments to alter fire behaviour and effects is well documented; however, long-term...

The electrolyte challenge for a direct methanol-air polymer electrolyte fuel cell operating at temperatures up to 200 C

Science.gov (United States)

Savinell, Robert; Yeager, Ernest; Tryk, Donald; Landau, Uziel; Wainright, Jesse; Gervasio, Dominic; Cahan, Boris; Litt, Morton; Rogers, Charles; Scherson, Daniel

1993-01-01

Novel polymer electrolytes are being evaluated for use in a direct methanol-air fuel cell operating at temperatures in excess of 100 C. The evaluation includes tests of thermal stability, ionic conductivity, and vapor transport characteristics. The preliminary results obtained to date indicate that a high temperature polymer electrolyte fuel cell is feasible. For example, Nafion 117 when equilibrated with phosphoric acid has a conductivity of at least 0.4 Omega(exp -1)cm(exp -1) at temperatures up to 200 C in the presence of 400 torr of water vapor and methanol vapor cross over equivalent to 1 mA/cm(exp 2) under a one atmosphere methanol pressure differential at 135 C. Novel polymers are also showing similar encouraging results. The flexibility to modify and optimize the properties by custom synthesis of these novel polymers presents an exciting opportunity to develop an efficient and compact methanol fuel cell.

A comparative analysis of the effect of gaseous fission products release on the thermal behaviour of oxide fuel rods

International Nuclear Information System (INIS)

Totev, T.L.; Kolev, I.G.

1992-01-01

Four different models of gaseous fission product release are compared in order to assess the relative effect of thermal characteristics of the fuel rods. The results show that the use of Weisman and EPRI models at a high burnup (over 50000 MW.d/tU) leads to almost the same figures of maximum fuel temperature and gas gap thermal conductivity. The use of Beyer-Hann (Betelle) and Pazdera-Valach (Rzez) models leads to under prediction of the fuel element thermal characteristics. A conclusion has been made that the Weisman model is the most suitable for the WWER-type fuel elements behaviour prediction. 10 refs., 7 figs

32 CFR 701.119 - Privacy and the web.

Science.gov (United States)

2010-07-01

... 32 National Defense 5 2010-07-01 2010-07-01 false Privacy and the web. 701.119 Section 701.119... THE NAVY DOCUMENTS AFFECTING THE PUBLIC DON Privacy Program § 701.119 Privacy and the web. DON activities shall consult SECNAVINST 5720.47B for guidance on what may be posted on a Navy Web site. ...

Near-ambient solid polymer fuel cell

Science.gov (United States)

Holleck, G. L.

1993-01-01

Fuel cells are extremely attractive for extraterrestrial and terrestrial applications because of their high energy conversion efficiency without noise or environmental pollution. Among the various fuel cell systems the advanced polymer electrolyte membrane fuel cells based on sulfonated fluoropolymers (e.g., Nafion) are particularly attractive because they are fairly rugged, solid state, quite conductive, of good chemical and thermal stability and show good oxygen reduction kinetics due to the low specific adsorption of the electrolyte on the platinum catalyst. The objective of this program is to develop a solid polymer fuel cell which can efficiently operate at near ambient temperatures without ancillary components for humidification and/or pressurization of the fuel or oxidant gases. During the Phase 1 effort we fabricated novel integral electrode-membrane structures where the dispersed platinum catalyst is precipitated within the Nafion ionomer. This resulted in electrode-membrane units without interfacial barriers permitting unhindered water diffusion from cathode to anode. The integral electrode-membrane structures were tested as fuel cells operating on H2 and O2 or air at 1 to 2 atm and 10 to 50 C without gas humidification. We demonstrated that cells with completely dry membranes could be self started at room temperature and subsequently operated on dry gas for extended time. Typical room temperature low pressure operation with unoptimized electrodes yielded 100 mA/cm(exp 2) at 0.5V and maximum currents over 300 mA/cm(exp 2) with low platinum loadings. Our results clearly demonstrate that operation of proton exchange membrane fuel cells at ambient conditions is feasible. Optimization of the electrode-membrane structure is necessary to assess the full performance potential but we expect significant gains in weight and volume power density for the system. The reduced complexity will make fuel cells also attractive for smaller and portable power supplies and as

In-reactor measurements of thermo mechanical behaviour and fission gas release of water reactor fuel

International Nuclear Information System (INIS)

Kolstad, E.; Vitanza, C.

1983-01-01

the fuel performance during and after a power ramp can be investigated by direct in-pile measurements related to the thermal, mechanical and fission gas release behaviour. The thermal response is examined by thermocouples placed at the centre of the fuel. Such measurements allow the determination of thermal feedback effects induced by the simultaneous liberation of fission gases. The thermal feedback effect is also being separately studied out-of-pile in a specially designed rod where the fission gas release is simulated by injecting xenon in known quantities at different axial positions within the rod. Investigations on the mechanical behaviour are based on axial and diametral cladding deformation measurements. This enables the determination of the amount of local cladding strain and ridging during ramping, the extent of relaxation during the holding time and the amount of residual (plastic) deformation. Gap width measurements are also performed in operating fuel rods using a cladding deflection technique. Fission gas release data are obtained, besides from post-irradiation puncturing, by continuous measurements of the rod internal pressure. This type of measurement leads to the description of the kinetics of the fission gas release process at different powers. The data tend to indicate that the time-dependent release can be reasonably well described by simple diffusion. The paper describes measuring techniques developed and currently in use in Halden, and presents and discusses selected experimental results obtained during various power ramps and transients. (author)

checkCIF/PLATON report Datablock: exp_2

Indian Academy of Sciences (India)

THIS REPORT IS FOR GUIDANCE ONLY. IF USED AS PART OF A REVIEW PROCEDURE. FOR PUBLICATION, IT SHOULD NOT REPLACE THE EXPERTISE OF AN EXPERIENCED. CRYSTALLOGRAPHIC REFEREE. No syntax errors found. CIF dictionary Interpreting this report. Datablock: exp_2. Bond precision:.

checkCIF/PLATON report Datablock: exp_3042

Indian Academy of Sciences (India)

THIS REPORT IS FOR GUIDANCE ONLY. IF USED AS PART OF A REVIEW PROCEDURE. FOR PUBLICATION, IT SHOULD NOT REPLACE THE EXPERTISE OF AN EXPERIENCED. CRYSTALLOGRAPHIC REFEREE. No syntax errors found. CIF dictionary Interpreting this report. Datablock: exp_3042. Bond precision:.

checkCIF/PLATON report Datablock: exp_2816

Indian Academy of Sciences (India)

THIS REPORT IS FOR GUIDANCE ONLY. IF USED AS PART OF A REVIEW PROCEDURE. FOR PUBLICATION, IT SHOULD NOT REPLACE THE EXPERTISE OF AN EXPERIENCED. CRYSTALLOGRAPHIC REFEREE. No syntax errors found. CIF dictionary Interpreting this report. Datablock: exp_2816. Bond precision:.

checkCIF/PLATON report Datablock: exp_3274

Indian Academy of Sciences (India)

THIS REPORT IS FOR GUIDANCE ONLY. IF USED AS PART OF A REVIEW PROCEDURE. FOR PUBLICATION, IT SHOULD NOT REPLACE THE EXPERTISE OF AN EXPERIENCED. CRYSTALLOGRAPHIC REFEREE. No syntax errors found. CIF dictionary Interpreting this report. Datablock: exp_3274. Bond precision:.

Parametric study of the behaviour of a pre irradiated BWR fuel rod under conditions of LOCA simulated in the halden in pile test system with the FALCON code

Energy Technology Data Exchange (ETDEWEB)

Khvostov, G.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland); Kolstad, E. [Institute for Energy Technology - OECD Halden Reactor Project, Halden (Norway); Montgomery, R. O. [Anatech Corporation, San Diego (United States)

2008-10-15

A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn up with burst of the cladding expected to occur at a temperature of about 1050.deg.C, which is essentially higher than in the preceding experiments. The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of 1150 .deg. C (peak local). The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG 0630, as well as consistency with the data from Halden LOCA testing available so far. Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn up.

Accidental behaviour of nuclear fuel in a warehousing site under air: investigation of the nuclear ceramic oxidation and of fission gas release

International Nuclear Information System (INIS)

Desgranges, L.

2006-12-01

After a brief presentation of the context of his works, i.e. the nuclear fuel, its behaviour in a nuclear reactor, and studies performed in high activity laboratory, the author more precisely presents its research topic: the behaviour of defective nuclear fuel in air. Then, he describes the researches performed in three main directions: firstly, the characterization and understanding of fission gas localisation (experimental localisation, understanding of the bubble forming mechanisms), secondly, the determination of mechanisms related to oxidation (atomic mechanisms related to UO 2 oxidation, oxidation of fragments of irradiated fuel, the CROCODILE installation). He finally presents his scientific project which notably deals with fission gas release (from UO 2 to U 3 O 7 , and from U 3 O 7 to U 3 O 8 ), and with further high activity laboratory experiments

14 CFR 119.9 - Use of business names.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 3 2010-01-01 2010-01-01 false Use of business names. 119.9 Section 119.9... COMMERCIAL OPERATORS General § 119.9 Use of business names. (a) A certificate holder under this part may not operate an aircraft under part 121 or part 135 of this chapter using a business name other than a business...

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

International Nuclear Information System (INIS)

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)

Nuclear fuel behaviour modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

2001-07-01

of MOX behaviour up to that for UO 2 fuel. There appears to be a good consensus on how MOX fuel performance differs from UO 2 , and on the issues that need to be addressed to achieve higher burnups. The final sessions of the TCM considered the current status of integrated fuel behaviour codes and the challenges for higher burnup modelling. The meeting provided a valuable forum for a review of the state-of-the-art. Presentations were given on a number of existing codes and others under development, covering PWR, WWER, BWR and CANDU fuel performance. Some specialised methods for specific advanced fuel types were also discussed. Recommendations on future work in the area of fission gas release; clad modelling; and MOX fuel modelling are included

38 CFR 4.119 - Schedule of ratings-endocrine system.

Science.gov (United States)

2010-07-01

... minute), eye involvement, muscular weakness, loss of weight, and sympathetic nervous system...-endocrine system. 4.119 Section 4.119 Pensions, Bonuses, and Veterans' Relief DEPARTMENT OF VETERANS AFFAIRS SCHEDULE FOR RATING DISABILITIES Disability Ratings The Endocrine System § 4.119 Schedule of ratings...

7 CFR 3550.119 - WWD eligibility requirements.

Science.gov (United States)

2010-01-01

....119 Agriculture Regulations of the Department of Agriculture (Continued) RURAL HOUSING SERVICE... 306C Water and Waste Disposal Grants § 3550.119 WWD eligibility requirements. In addition to the... residing in the household that is below the most recent poverty income guidelines established by the...

Perinatal and chronic hypothyroidism impair behavioural development in male and female rats.

NARCIS (Netherlands)

Wijk, van N.; Rijntjes, E.; Heijning, van de B.J.

2008-01-01

Perinatal and chronic hypothyroidism impair behavioural development in male and female rats. EXP PHYSIOL 00(0) 000-000, 0000. - A lack of thyroid hormone, i.e. hypothyroidism, during early development results in multiple morphological and functional alterations in the developing brain. In the

Mapa de energia baseado em predição para redes de sensores sem fio

OpenAIRE

Raquel Aparecida de Freitas Mini

2004-01-01

O objetivo deste trabalho é construir o mapa de energia de uma rede de sensores sem fio utilizando técnicas de predição. Em mapas de energia baseados em predição, cada nó sensor utiliza algoritmos de predição para modelar sua dissipação de energia com o objetivo de prever seu consumo de energia no futuro. Nesta situação, ao invés de enviar para o nó sorvedouro apenas sua quantidade de energia disponível, cada sensor envia também os parâmetros do modelo que descreve seu consumo de energia. O n...

The MOX fuel behaviour test IFA-597.4/.5. Temperature and pressure data to a burn-up of 15 MWd/kg MOX

International Nuclear Information System (INIS)

Takano, K.

1999-04-01

The behaviour of MOX fuel should be investigated in detail for more effective use in the future, especially concerning its thermal performance and fission gas release. IFA-597.4 and IFA-597.5, containing two MOX fuel rods each with a fuel centre thermocouple and a pressure transducer, have been irradiated in the Halden Reactor to study the temperature threshold of fission gas release for MOX fuel and to explore potential differences in the thermal and fission gas release behaviour between solid and hollow pellets. The two rods of MOX fuel with an initial Pu-fissile content of 6.07 percent have solid pellets and hollow pellets respectively, and with an active length of about 220 mm. The diameter of the pellets is 8.05 mm with 180μm of diametral gap to the cladding. For the purpose of the test, power ramp operation, in which estimated peak temperature of the MOX pellets increases and decreases above and below the threshold for fission gas release in UO 2 fuel, is planned every 10 MWd/kgMOX of burn-up. The first ramp operation has been successfully performed at 10 MWd/kgMOX. When the estimated peak temperature of the fuel gets close to but below the threshold of UO 2 , fission gas release was observed at around 28 kW/m of power. Densification of the MOX pellets could be estimated to about 1.2 percent for the solid pellets and about 2,3 percent for the hollow pellets from normalised internal rod pressure. After 13.5 MWd/kgMOX the average assembly power has been operated low enough to observe swelling rate of MOX fuel pellets and behaviour after significant fission gas release. The burn-up had reached 15.5 MWd/kgMOX as of the end of 1998. The target burn-up of this MOX test is 60 MWd/kgMOX (author) (ml)

Symphonie « pragmatique » de l’expérience musicale.

Directory of Open Access Journals (Sweden)

François Debruyne

2011-03-01

Full Text Available User d’une métaphore musicale éculée, celle de la symphonie , pour commenter une œuvre sociologique qui a pour thème « l’expérience musicale » peut sembler accessoire pour décrire un ouvrage, dont le titre et le sous-titre originaux – Écologie sociale de l’oreille. Enquêtes sur l’expérience musicale  – disent plus précisément ce dont il est question. Néanmoins, surtitrer Symphonie « pragmatique » de l’expérience musicale permet à la fois de signifier ...

Remote and Virtual Labs @ exp.at’11

Directory of Open Access Journals (Sweden)

Maria Teresa Restivo

2012-03-01

Full Text Available exp.at’11, the first event of Experiment@, a new International Conference series devoted to online experimentation, had as scope to contribute to the world capabilities in online experimentation and in particular in remote and virtual labs, fostering the collaborative work in emergent technologies.

Glutathione oxidation correlates with one-lung ventilation time and PO2/FiO2 ratio during pulmonary lobectomy.

Science.gov (United States)

García-de-la-Asunción, José; García-Del-Olmo, Eva; Galan, Genaro; Guijarro, Ricardo; Martí, Francisco; Badenes, Rafael; Perez-Griera, Jaume; Duca, Alejandro; Delgado, Carlos; Carbonell, Jose; Belda, Javier

2016-09-01

During lung lobectomy, the operated lung completely collapses with simultaneous hypoxic pulmonary vasoconstriction, followed by expansion and reperfusion. Here, we investigated glutathione oxidation and lipoperoxidation in patients undergoing lung lobectomy, during one-lung ventilation (OLV) and after resuming two-lung ventilation (TLV), and examined the relationship with OLV duration. We performed a single-centre, observational, prospective study in 32 patients undergoing lung lobectomy. Blood samples were collected at five time-points: T0, pre-operatively; T1, during OLV, 5 minutes before resuming TLV; and T2, T3, and T4, respectively, 5, 60, and 180 minutes after resuming TLV. Samples were tested for reduced glutathione (GSH), oxidized glutathione (GSSG), glutathione redox potential, and malondialdehyde (MDA). GSSG and MDA blood levels increased at T1, and increased further at T2. OLV duration directly correlated with marker levels at T1 and T2. Blood levels of GSH and glutathione redox potential decreased at T1-T3. GSSG, oxidized glutathione/total glutathione ratio, and MDA levels were inversely correlated with arterial blood PO2/FiO2 at T1 and T2. During lung lobectomy and OLV, glutathione oxidation, and lipoperoxidation marker blood levels increase, with further increases after resuming TLV. Oxidative stress degree was directly correlated with OLV duration, and inversely correlated with arterial blood PO2/FiO2.

18 CFR 401.119 - Disclosure to Congress.

Science.gov (United States)

2010-04-01

... 18 Conservation of Power and Water Resources 2 2010-04-01 2010-04-01 false Disclosure to Congress. 401.119 Section 401.119 Conservation of Power and Water Resources DELAWARE RIVER BASIN COMMISSION... Disclosure to Congress. All records of the Commission shall be disclosed to Congress upon an authorized...

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

International Nuclear Information System (INIS)

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

Um novo fio de aço inoxidável para aplicações ortodônticas A new stainless steel wire for orthodontic purposes

Directory of Open Access Journals (Sweden)

André Itman Filho

2011-08-01

Full Text Available OBJETIVO: desenvolver uma metodologia para fabricação de fios ortodônticos de aço inoxidável austeno-ferrítico SEW 410 Nr. 14517 por meio dos processos convencionais de laminação e trefilação. MÉTODOS: o aço austeno-ferrítico foi elaborado em um forno elétrico de indução. A qualidade dos fios foi avaliada por ensaios de tração e medidas de microdureza. A ductilidade e a manuseabilidade foram analisadas por meio da confecção de componentes ortodônticos. RESULTADOS E CONCLUSÕES: os valores encontrados mostraram que os fios de aço inoxidável austeno-ferrítico atenderam às normas BS 3507:1976 e ISO 5832-1, e apresentaram ótima ductilidade para confecção de componentes ortodônticos com dobras complexas.OBJECTIVE: To develop a method to manufacture austenitic-ferritic stainless steel orthodontic wires (SEW 410 Nr. 14517 using conventional rolling and wiredrawing processes. METHODS: Austenitic-ferritic steel was produced in an induction furnace. Traction trials and microhardness measurements were used to evaluate wire quality. Orthodontic parts were fabricated to assess ductility and malleability. RESULTS AND CONCLUSIONS: Austenitic-ferritic stainless steel wires meet the BS 3507:1976 and ISO 5832-1 norms and have excellent ductility for the fabrication of orthodontic parts with complex folds.

IMRT Commissioning: application of the AAPM's TG-119; Comissionamento de IMRT: aplicacao do TG-119 da AAPM

Energy Technology Data Exchange (ETDEWEB)

Zeppellini, Caroline; Furnari, Laura, E-mail: laurafurnari@hotmail.com [Universidade de Sao Paulo (USP), Sao Paulo, SP (Brazil). Fac. de Medicina. Inst. de Radiologia

2013-08-15

In order to verify the commissioning of the planning of intensity-modulated radiation therapy system (IMRT), the TG-119 of the American Association of Physicists in Medicine (AAPM) was applied. Using pre defined targets and normal structures, plans were realized, absolute and relative dose were measured with an ionizing chamber and films, and the results were compared with planned values. The maximum deviation of the measurements with the ionization chamber was 3,6%, but, in the total eleven measurements, only two were bigger than the tolerance limit of 3%, recommended by TG-119. The number of points which passed criteria gamma 3% to 3 mm ranged between 96.36% and 99.92%, all measurements were within the recommended 95%. The confidence limits found for both film and for chamber were lower than those achieved in the TG-119. Our results showed a good concordance with TG-119, what means that the system is adequate for clinical applications. (author)

Rupture behaviour of nuclear fuel cladding during loss-of-coolant accident

Energy Technology Data Exchange (ETDEWEB)

Suman, Siddharth [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Khan, Mohd Kaleem, E-mail: mkkhan@iitp.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Pathak, Manabendra [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Singh, R.N.; Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

2016-10-15

Highlights: • Modelling of nuclear fuel cladding during loss-of-coolant accident transient. • Phase transformation, corrosion, and creep combined to evaluate burst criterion. • Effect of oxygen concentration on burst stress and burst strain. • Effect of heating rate, internal pressure fluctuation, shear modulus incorporated. - Abstract: A burst criterion model accounting the simultaneous phenomena of corrosion, solute-strengthening effect of oxygen, oxygen concentration based non-isothermal phase transformation, and thermal creep has been developed to predict the rupture behaviour of zircaloy-4 nuclear fuel cladding during the loss-of-coolant accident transients. The present burst criterion model has been validated using experimental data obtained from single-rod transient burst tests performed in steam environment. The predictions are in good agreement with the experimental results. A detailed computational analysis has been performed to assess the role of different parameters in the rupture of zircaloy cladding during loss-of-coolant accidents. This model reveals that at low temperatures, lower heating rates produce higher burst strains as oxidation effect is nominal. For high temperatures, the lower heating rates produce less burst strains, whereas higher heating rates yield greater burst strains.

expVIP: a Customizable RNA-seq Data Analysis and Visualization Platform.

Science.gov (United States)

Borrill, Philippa; Ramirez-Gonzalez, Ricardo; Uauy, Cristobal

2016-04-01

The majority of transcriptome sequencing (RNA-seq) expression studies in plants remain underutilized and inaccessible due to the use of disparate transcriptome references and the lack of skills and resources to analyze and visualize these data. We have developed expVIP, an expression visualization and integration platform, which allows easy analysis of RNA-seq data combined with an intuitive and interactive interface. Users can analyze public and user-specified data sets with minimal bioinformatics knowledge using the expVIP virtual machine. This generates a custom Web browser to visualize, sort, and filter the RNA-seq data and provides outputs for differential gene expression analysis. We demonstrate expVIP's suitability for polyploid crops and evaluate its performance across a range of biologically relevant scenarios. To exemplify its use in crop research, we developed a flexible wheat (Triticum aestivum) expression browser (www.wheat-expression.com) that can be expanded with user-generated data in a local virtual machine environment. The open-access expVIP platform will facilitate the analysis of gene expression data from a wide variety of species by enabling the easy integration, visualization, and comparison of RNA-seq data across experiments. © 2016 American Society of Plant Biologists. All Rights Reserved.

A summary of the assessment of fuel behaviour, fission product release and pressure tube integrity following a postulated large loss-of-coolant accident

International Nuclear Information System (INIS)

Langman, V.J.; Weaver, K.R.

1984-05-01

The Ontario Hydro analyses of fuel and pressure tube temperatures, fuel behaviour, fission product release and pressure tube integrity for large break loss-of-coolant accidents in Bruce A or Pickering A have been critically reviewed. The determinations of maximum fuel temperatures and fission product release are very uncertain, and pressure tube integrity cannot be assured where low steam flows are predicted to persist for times on the order of minutes

Application of the Exp-function method to the equal-width wave equation

International Nuclear Information System (INIS)

Biazar, J; Ayati, Z

2008-01-01

In this paper, the Exp-function method is used to find an exact solution of the equal-width wave (EW) equation. The method is straightforward and concise, and its applications are promising. It is shown that the Exp-function method, with the help of symbolic computation, provides a very effective and powerful mathematical tool for solving the EW equation.

Fractional Complex Transform and exp-Function Methods for Fractional Differential Equations

Directory of Open Access Journals (Sweden)

Ahmet Bekir

2013-01-01

Full Text Available The exp-function method is presented for finding the exact solutions of nonlinear fractional equations. New solutions are constructed in fractional complex transform to convert fractional differential equations into ordinary differential equations. The fractional derivatives are described in Jumarie's modified Riemann-Liouville sense. We apply the exp-function method to both the nonlinear time and space fractional differential equations. As a result, some new exact solutions for them are successfully established.

Lifetime measurements and the nonaxial deformation in 119I

International Nuclear Information System (INIS)

Srebrny, J.; Droste, Ch.; Morek, T.; Starosta, K.; Juutinen, S.; Piiparinen, M.

2000-01-01

Lifetimes of negative parity levels in four bands in 119 I were determined by RDM and DSAM. The 119 I nuclei were produced in the 109 Ag( 13 C,3n) reaction. Calculations in the frame of the CQPC model show that the model of γ-soft nucleus better describes 119 I than the γ-rigid one. (author)

Fios que Tecem o Tempo Escolar: do ritmo padrão à simultaneidade

Directory of Open Access Journals (Sweden)

Ana Sueli Teixeira de Pinho

2017-09-01

Full Text Available Este texto tem por objetivo tecer os fios que, de modo complexo, atam e desatam, de forma silenciosa, a constituição do tempo em duas escolas com classes multisseriadas da Ilha de Maré, problematizando a ideia de ritmo padrão e apontando a simultaneidade como possibilidade para o tempo escolar. Do ponto de vista metodológico foi adotada a narrativa (autobiográfica, e a técnica de pesquisa selecionada foi a entrevista narrativa. O estudo realizado aponta que o tempo escolar, das duas escolas pesquisadas, é convidado a reconhecer a existência e a legitimidade de outros tempos, para além do seu: o tempo da maré, o tempo do trabalho, o tempo das práticas simbólicas, o tempo livre e as temporalidades dos sujeitos.

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

International Nuclear Information System (INIS)

Freund, D.; Geithoff, D.

1988-04-01

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

TIME-DEPENDENT MOSSBAUER-SPECTROSCOPY AND 119MTE-IMPLANTED GAAS

NARCIS (Netherlands)

MO, D; ZHANG, GL; NIESEN, L; Waard , de Hendrik

1991-01-01

A new type of time-dependent Mossbauer spectroscopy is proposed and realized on the basis of using the two-step decay (119m)Te --> 113Sb --> Sn-119. For the GaAs samples, implanted with a dose of 110-keV (119m)Te + 10(15) stable Te/cm2 and annealed at 600-degrees-C, the relative intensities of

In-reactor behaviour of centrifugally atomized U3Si dispersion fuel irradiated at high temperature in HANARO

International Nuclear Information System (INIS)

Kim, Ki Hwan; Park, Jong Man; Yoo, Byeong Ok; Park, Dae Kyu; Lee, Choong Sung; Kim, Chang Kyu

2002-01-01

The irradiation test on full-size U 3 Si dispersion fuel elements, prepared by centrifugal atomization and conventional comminution method, has been performed up to about 77 at.% U-235 in maximum burn-up at CT hole position having the highest power condition in the HANARO reactor, in order to examine the irradiation performance of the atomized U 3 Si for the driver fuels of HANARO. The in-reactor interaction of the atomized U 3 Si dispersion fuel meats is generally assumed to be acceptable with the range of 5-15 μm in average thickness. The atomized spherical particles have more uniform and thinner reaction layer than the comminuted irregular particles. The U 3 Si particles have relatively fine and uniform size distribution of fission gas bubbles, irrespective of the powdering method. The bubble population in the atomized particles appears to be finer and more homogeneous with the characteristics of narrower bubble size distribution than that of the comminuted fuel. The atomized U 3 Si dispersion fuel elements exhibit sound swelling behaviours of 5 % in ΔV/V m even at ∼77 at.% U-235 burn-up, which meets with the safety criterion of the fuel rod, 20vol.% for HANARO. The atomized U3Si dispersion fuel elements show smaller swelling than the comminuted fuel elements

47 CFR 69.119 - Basic service element expedited approval process.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 3 2010-10-01 2010-10-01 false Basic service element expedited approval process. 69.119 Section 69.119 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) COMMON CARRIER SERVICES (CONTINUED) ACCESS CHARGES Computation of Charges § 69.119 Basic service element...

High-level radioactive wastes storage characterization and behaviour of spent fuels in long-term

International Nuclear Information System (INIS)

Diaz Arocas, P.; Cobos, J.; Quinones, J.; Rodriguez Almazan, J. L.; Serrano, J.

2001-01-01

In order to understand the long term spent fuel dissolution under repository this report shows the study performed by considering spent fuel as a part of the multi barriers containment system. The study takes into account that the oxidative alteration/dissolution of spent fuel matrix is influenced by the intrinsic spent fuel physicochemical characteristics and the repository environmental parameters. Experimental and modelling results for granite and saline repositories are reported. Parameters considered in this work were pH, pCO 2 , S/V ratio, redox conditions and the influence of the container material in the redox conditions. The influence of alpha, beta and gamma radiation and the resulting radiolytic products formed remains as one of the main uncertainties to quantify the spent fuel behaviour under repository conditions. It was studied in a first approach through dose calculations, modelling of radiolytic products formation and leaching experiments in the presence of external gamma irradiation source and leaching experiments of alpha doped UO 2 pellets. Materials considered are LWR spent fuel (UO 2 and MOX fuel) and their chemical analogues non irradiated UO 2 , SIMFUEL and alpha doped UO 2 . Lea chants were granite groundwater, synthetic granite groundwater, synthetic granite groundwater saturated in bentonite, and high concentrated saline solutions. The matrix dissolution rate and release rate of key radionuclides (i. e. actinides and fission products) obtained through the several experimental techniques and methodologies (dissolution, co-dissolution, precipitation and co-precipitation) together with modelling studies supported in geochemical codes are proposed. Moreover, secondary phases formed that could control release and retention of key nuclides are identified. Maximum concentration values for these radionuclides are reported. The data provided by this study were used in ENRESA-2000 performance assessment. (Author)

LWR fuel rod testing facilities in high flux reactor (HFT) Petten for investigation of power cycling and ramping behaviour

Energy Technology Data Exchange (ETDEWEB)

Markgraf, J; Perry, D; Oudaert, J [Commission of the European Communities, Joint Reserach Centre, Petten Establishment, Petten (Netherlands)

1983-06-01

LWR fuel rod irradiation experiments are being performed in HFR's Pool Side Facility (PSF) by means of pressurized boiling water capsules (BWFC). For more than 6 years the major application of these devices has been in performing irradiation programs to investigate the power ramp behaviour of PWR and BWR rods which have been pre-irradiated in power reactors. Irradiation devices with various types of monitoring instrumentation are available, e.g. for fuel rod length, fuel stack length, fuel rod internal pressure and fuel rod central temperature measurements. The application scope covers PWR and BWR fuel rods, with burn-up levels up to 45 MWd/kg(U), max. linear heat generation rates up to 700 W/cm and simulation of constant power change rates between 0.05 and 1000 W/cm.min. The paper describes the different designs of LWR fuel rod testing facilities and associated non-destructive testing devices in use at the HFR Petten. It also addresses the new test capabilities that will become available after exchange of the HFR vessel in 1983. Furthermore it shows some typical results. (author)

Synthesis of [119mSn]-mesoporphyrin IX dichloride

International Nuclear Information System (INIS)

Denissen, J.F.

1990-01-01

Tin mesoporphyrin IX dichloride (Sn-MPCl 2 ) is a heme oxygenase inhibitor of current clinical interest for the treatment of neonatal hyperbilirubinemia. The synthesis of [ 119m Sn]-MPCl 2 for drug metabolism and disposition studies is reported. [ 119m Sn]-MPCl 2 was prepared in 60% radiochemical yield by metalation of the porphyrin nucleus of mesoporphyrin IX dihydrochloride with tin(II)-119m acetate. The product had a specific activity of 43.4 mCi/mmol and a radiochemical purity of 99%, as determined by radio-HPLC analysis. (author)

Global Energy Issues and Alternate Fueling

Science.gov (United States)

Hendricks, Robert C.

2007-01-01

This viewgraph presentation describes world energy issues and alternate fueling effects on aircraft design. The contents include: 1) US Uses about 100 Quad/year (1 Q = 10(exp 15) Btu) World Energy Use: about 433 Q/yr; 2) US Renewable Energy about 6%; 3) Nuclear Could Grow: Has Legacy Problems; 4) Energy Sources Primarily NonRenewable Hydrocarbon; 5) Notes; 6) Alternate Fuels Effect Aircraft Design; 7) Conventional-Biomass Issue - Food or Fuel; 8) Alternate fuels must be environmentally benign; 9) World Carbon (CO2) Emissions Problem; 10) Jim Hansen s Global Warming Warnings; 11) Gas Hydrates (Clathrates), Solar & Biomass Locations; 12) Global Energy Sector Response; 13) Alternative Renewables; 14) Stratospheric Sulfur Injection Global Cooling Switch; 15) Potential Global Energy Sector Response; and 16) New Sealing and Fluid Flow Challenges.

Exp-function method for constructing exact solutions of Sharma-Tasso-Olver equation

International Nuclear Information System (INIS)

Erbas, Baris; Yusufoglu, Elcin

2009-01-01

In this paper we use the Exp-function method for the analytic treatment of Sharma-Tasso-Olver equation. New solitonary solutions are formally derived. Change of parameters, which drastically changes the characteristics of the equations, is examined. It is shown that the Exp-function method provides a powerful mathematical tool for solving high-dimensional nonlinear evolutions in mathematical physics. The proposed schemes are reliable and manageable.

checkCIF/PLATON report Datablock: exp_2b

Indian Academy of Sciences (India)

THIS REPORT IS FOR GUIDANCE ONLY. IF USED AS PART OF A REVIEW PROCEDURE. FOR PUBLICATION, IT SHOULD NOT REPLACE THE EXPERTISE OF AN EXPERIENCED. CRYSTALLOGRAPHIC REFEREE. No syntax errors found. CIF dictionary Interpreting this report. Datablock: exp_2b. Bond precision:.

checkCIF/PLATON report Datablock: exp_1b

Indian Academy of Sciences (India)

THIS REPORT IS FOR GUIDANCE ONLY. IF USED AS PART OF A REVIEW PROCEDURE. FOR PUBLICATION, IT SHOULD NOT REPLACE THE EXPERTISE OF AN EXPERIENCED. CRYSTALLOGRAPHIC REFEREE. No syntax errors found. CIF dictionary Interpreting this report. Datablock: exp_1b. Bond precision:.

Fuel Behaviour in Transport after Dry Storage: a Key Issue for the Management of used Nuclear Fuel

International Nuclear Information System (INIS)

Issard, Herve

2014-01-01

Interim used fuel dry storage has been developed in many countries providing an intermediate solution while waiting for evaluation and decisions concerning future use (such as recycling) or disposal sites. There is an important industrial experience feedback and excellent safety records. It appears that the duration of interim storage may become longer than initially expected. At the start of storage operations 40 years was considered sufficiently long to make a decision on either recycling or direct disposal of used nuclear fuel. Now it is said that storage time may have to be extended. Whatever the choice for the management of used fuel, it will finally have to be transported from the storage facility to another location, for recycling or final disposal. Bearing in mind the important principle that radioactive waste shall be managed in such a way that undue burdens will not be imposed on future generations, there is no guarantee that the fuel characteristics can be maintained in perpetuity. On the other hand, transport accident conditions from applicable regulation (IAEA SSR-6) are very severe for irradiated materials. Therefore, in compliance with transport regulations, the safety analysis of the fuel in transport after storage is mandatory. This paper will give an overview of the current situation related to the used fuel behaviour in transport after dry storage. On this matter there are some elements of information already available as well as some gaps of knowledge. Several national R and D programs and international teams are presently addressing these gaps. A lot of R and D work has already been done. An objective of these R and D projects is to aid decision makers. It is important to fix a limit and not to multiply intermediate operations because it means higher costs and more uncertainties. The identified gaps concern the following issues especially for high burn-up (HBU) fuels: thermal model for casks, degradation process of fuel material, cladding creep

Impact of fuels on diesel exhaust emissions

International Nuclear Information System (INIS)

Westerholm, R.

1991-09-01

This report presents an investigation of the emissions from eight diesel fuels with different sulphur and aromatic content. A bus and a truck were used in the investigation. Chemical analysis and biological testing have been performed. The aim of this project was to find a 'good' diesel fuel which can be used in urban areas. Seven of the fuels were meant to be such fuels. It has been confirmed in this study that there exists a quantifiable relationship between the variables of the diesel fuel blends and the variables of the chemical emissions and their biological effects. 119 figs., 12 tabs., approx. 100 refs

The behaviour under irradiation of molybdenum matrix for inert matrix fuel containing americium oxide (CerMet concept)

Energy Technology Data Exchange (ETDEWEB)

D' Agata, E., E-mail: elio.dagata@ec.europa.eu [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, 1755 ZG Petten (Netherlands); Knol, S.; Fedorov, A.V. [Nuclear Research and Consultancy Group, P.O. Box 25, 1755 ZG Petten (Netherlands); Fernandez, A.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Klaassen, F. [Nuclear Research and Consultancy Group, P.O. Box 25, 1755 ZG Petten (Netherlands)

2015-10-15

Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors or Accelerator Driven System (ADS, subcritical reactors dedicated to transmutation) of long-lived nuclides like {sup 241}Am is therefore an option for the reduction of radiotoxicity of waste packages to be stored in a repository. In order to safely burn americium in a fast reactor or ADS, it must be incorporated in a matrix that could be metallic (CerMet target) or ceramic (CerCer target). One of the most promising matrix to incorporate Am is molybdenum. In order to address the issues (swelling, stability under irradiation, gas retention and release) of using Mo as matrix to transmute Am, two irradiation experiments have been conducted recently at the High Flux Reactor (HFR) in Petten (The Netherland) namely HELIOS and BODEX. The BODEX experiment is a separate effect test, where the molybdenum behaviour is studied without the presence of fission products using {sup 10}B to “produce” helium, the HELIOS experiment included a more representative fuel target with the presence of Am and fission product. This paper covers the results of Post Irradiation Examination (PIE) of the two irradiation experiments mentioned above where molybdenum behaviour has been deeply investigated as possible matrix to transmute americium (CerMet fuel target). The behaviour of molybdenum looks satisfying at operating temperature but at high temperature (above 1000 °C) more investigation should be performed.

14 CFR 91.119 - Minimum safe altitudes: General.

Science.gov (United States)

2010-01-01

... than 500 feet to any person, vessel, vehicle, or structure. (d) Helicopters. Helicopters may be... 14 Aeronautics and Space 2 2010-01-01 2010-01-01 false Minimum safe altitudes: General. 91.119 Section 91.119 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION...

Contribution of thermodynamics in the understanding of the physico-chemical behaviour of fuels at high temperature

International Nuclear Information System (INIS)

Gueneau, C.; Chatain, S.; Gosse, S.; Dumas, J.C.; Defoort, F.

2006-01-01

The thermodynamic approach for studying the physico-chemical behaviour of nuclear fuels at high temperature is presented. For instance is shown how the thermodynamic study of the uranium-oxygen-zirconium-iron system has contributed to improve the understanding of the scenario considered in studies on serious accidents for PWR reactors. Concerning the fuels of the future high temperature reactors, has been developed a thermodynamic data base 'fuelbase' (U-Pu-O-C-N-Si-Zr-Ti-Mo-Cr) using the Calphad method in parallel with experimental studies. In the framework of the studies on high temperature reactors, experimental works on the study of the interaction between the uranium dioxide and graphite are presented. This interaction leads to the formation of gaseous CO and CO 2 which can potentially be prejudicial to the thermomechanical resistance of the fuel in reactor. In this framework, the thermodynamic properties of the uranium-oxygen-carbon system are studied. (O.M.)

Application of Exp-function method for (2 + 1)-dimensional nonlinear evolution equations

International Nuclear Information System (INIS)

Bekir, Ahmet; Boz, Ahmet

2009-01-01

In this paper, the Exp-function method is used to construct solitary and soliton solutions of (2 + 1)-dimensional nonlinear evolution equations. (2 + 1)-dimensional breaking soliton (Calogero) equation, modified Zakharov-Kuznetsov and Konopelchenko-Dubrovsky equations are chosen to illustrate the effectiveness of the method. The method is straightforward and concise, and its applications are promising. The Exp-function method presents a wider applicability for handling nonlinear wave equations.

Creep behaviour of ZrNb1 fuel cans in argon and steam

International Nuclear Information System (INIS)

Adam, E.; Stephan, M.; Wetzel, L.

1988-01-01

The paper is concerned with experimental investigations on the creep behaviour of fuel cans made of the ZrNb1 alloy. The isobaric-isothermal creep tests were performed in the range of temperatures from 990 K to 1290 K and with differential pressures over the can between 1.0 MPa and 2.5 MPa. They were characterized by linear heating of the test cans with 2 K/s until a given temperature was reached, followed by maintaining the cans at a constant temperature (Δ = ± 3 K) and loading it with purified argon produced internal pressure. The experiments were carried out in both an argon atmosphere surrounding the cans from outside and steam. (author)

40 CFR Appendix Viii to Part 600 - Fuel Economy Label Formats

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Fuel Economy Label Formats VIII... POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Pt. 600, App. VIII Appendix VIII to Part 600—Fuel Economy Label Formats EC01MY92.117 EC01MY92.118 EC01MY92.119 EC01MY92.120...

Overview of fuel behaviour and core degradation, based on modelling analyses. Overview of fuel behaviour and core degradation, on the basis of modelling results

International Nuclear Information System (INIS)

Massara, Simone

2013-01-01

Since the very first hours after the accident at Fukushima-Daiichi, numerical simulations by means of severe accident codes have been carried out, aiming at highlighting the key physical phenomena allowing a correct understanding of the sequence of events, and - on a long enough timeline - improving models and methods, in order to reduce the discrepancy between calculated and measured data. A last long-term objective is to support the future decommissioning phase. The presentation summarises some of the available elements on the role of the fuel/cladding-water interaction, which became available only through modelling because of the absence of measured data directly related to the cladding-steam interaction. This presentation also aims at drawing some conclusions on the status of the modelling capabilities of current tools, particularly for the purpose of the foreseen application to ATF fuels: - analyses with MELCOR, MAAP, THALES2 and RELAP5 are presented; - input data are taken from BWR Mark-I Fukushima-Daiichi Units 1, 2 and 3, completed with operational data published by TEPCO. In the case of missing or incomplete data or hypotheses, these are adjusted to reduce the calculation/measurement discrepancy. The behaviour of the accident is well understood on a qualitative level (major trends on RPV pressure and water level, dry-wet and PCV pressure are well represented), allowing a certain level of confidence in the results of the analysis of the zirconium-steam reaction - which is accessible only through numerical simulations. These show an extremely fast sequence of events (here for Unit 1): - the top of fuel is uncovered in 3 hours (after the tsunami); - the steam line breaks at 6.5 hours. Vessel dries at 10 hours, with a heat-up rate in a first moment driven by the decay heat only (∼7 K/min) and afterwards by the chemical heat from Zr-oxidation (over 30 K/min), associated with massive hydrogen production. It appears that the level of uncertainty increases with

ExpEdit: a webserver to explore human RNA editing in RNA-Seq experiments.

Science.gov (United States)

Picardi, Ernesto; D'Antonio, Mattia; Carrabino, Danilo; Castrignanò, Tiziana; Pesole, Graziano

2011-05-01

ExpEdit is a web application for assessing RNA editing in human at known or user-specified sites supported by transcript data obtained by RNA-Seq experiments. Mapping data (in SAM/BAM format) or directly sequence reads [in FASTQ/short read archive (SRA) format] can be provided as input to carry out a comparative analysis against a large collection of known editing sites collected in DARNED database as well as other user-provided potentially edited positions. Results are shown as dynamic tables containing University of California, Santa Cruz (UCSC) links for a quick examination of the genomic context. ExpEdit is freely available on the web at http://www.caspur.it/ExpEdit/.

Fast reactor fuel design and development

International Nuclear Information System (INIS)

Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

1977-01-01

Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

International Nuclear Information System (INIS)

Khvostov, G.; Wiesenack, W.; Zimmermann, M.A.; Ledergerber, G.

2011-01-01

Highlights: → A model for the dynamics of axial gas redistribution in fuel rods during the LOCA is developed and coupled to the FALCON fuel behaviour code. → The first verification of the model is carried out using the data of the selected Halden LOCA tests. → According to calculation, the short rods used in the Halden tests show a small effect of the delayed gas redistribution during the clad ballooning. → The predicted effect is significant in the full length rods, eventually resulting in a considerable delay of the predicted moment of cladding rupture. → The predicted delay of cladding burst may be large enough to eventually affect the efficiency of the emergency core cooling system. - Abstract: A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code. The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from

Estudo da ação dos fios de categute cromado e poliglecaprone 25, na ileocistoplastia em ratos, destacando a formação de cálculos

Directory of Open Access Journals (Sweden)

Schauffert Marli Doroti

2000-01-01

Full Text Available Realizou-se um estudo da ação dos fios de categute cromado 6-0 e de poliglecaprone 25, 5-0, para determinar o fio ideal na ileocistoplastia em ratas. O maior objetivo, foi a profilaxia litiásica. Utilizaram-se 51 ratos de Wistar, que participaram de dois grupos experimentais: grupo do plano piloto (27 ratos, que foi o grupo categute (GC e grupo da tese do mestrado (24 ratos, que foi o grupo poliglecaprone (GP. Sob anestesia intraperitonial com pentobarbital sódico a 3%, todos os animais foram submetidos a ileocistoplastia, após laparotomia mediana longitudinal. Nos ratos do GC, a anastomose da bexiga urinária (aberta 0,5 cm sagitalmente no seu ápice, com a extremidade distal do segmento ileal ,era realizada por meio de pontos separados em plano único, com fio de categute cromado 6-0, e nas ratas do GP, com fio de poliglecaprone 25, 5-0. Seguiu-se a síntese da parede abdominal e recuperação anestésica. O estudo era realizado em 27, 42 e 57 dias nos animais do GC, que eram divididos em 3 subgrupos de 9 ratos. No GP, os animais eram analisados em 28 e 84 dias, por subgrupos de 12 animais. Transcorrido o tempo determinado para cada subgrupo, eram reoperados e observados os aspectos macroscópicos da cicatrização, aderências e formação de cálculos. A bexiga urinária ampliada pelo segmento ileal, era ressecada, aberta, lavada em solução salina isotônica, fixada no Líquido de Boüin, e processada , para histologia. A eutanásia era consumada com dose mínima letal anestésica. Os resultados apresentados, foram 29,6% de litíase, nas ratas do GC: 5,4% em 27dias; 10% em 42 dias e 14,25 em 57 dias do pós operatório (p.o.. Nas ratas do GP, não foram encontrados cálculos. Os estudos microscópicos, foram submetidos à análise estatística (a £ a 0,05, e utilizados para outro trabalho científico. Conclui-se: na ileocistoplastia com categute cromado 6-0, há 29,6% de cálculos, do 27o ao 57o dia p.o. e com o uso do fio de

49 CFR 215.119 - Defective freight car truck.

Science.gov (United States)

2010-10-01

... 49 Transportation 4 2010-10-01 2010-10-01 false Defective freight car truck. 215.119 Section 215... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD FREIGHT CAR SAFETY STANDARDS Freight Car Components Suspension System § 215.119 Defective freight car truck. A railroad may not place or continue in service a...

Nuclear fuel

International Nuclear Information System (INIS)

D Hondt, P.

1998-01-01

The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

Simulation of integral local tests with high-burnup fuel

International Nuclear Information System (INIS)

Gyori, G.

2011-01-01

The behaviour of nuclear fuel under LOCA conditions may strongly depend on the burnup-dependent fuel characteristics, as it has been indicated by recent integral experiments. Fuel fragmentation and the associated fission gas release can influence the integral fuel behaviour, the rod rupture and the radiological release. The TRANSURANUS fuel performance code is a proper tool for the consistent simulation of burnup-dependent phenomena during normal operation and the thermo-mechanical behaviour of the fuel rod in a subsequent accident. The code has been extended with an empirical model for micro-cracking induced FGR and fuel fragmentation and verified against integral LOCA tests of international projects. (author)

Nuclear fuel activities in Canada

Energy Technology Data Exchange (ETDEWEB)

Cox, D S [Fuel Development Branch, Chalk River Labs., AECL (Canada)

1997-12-01

Nuclear fuel activities in Canada are considered in the presentation on the following directions: Canadian utility fuel performance; CANDU owner`s group fuel programs; AECL advanced fuel program (high burnup fuel behaviour and development); Pu dispositioning (MOX) activities. 1 tab.

UMo nuclear fuels behaviour under heavy ion irradiation: a μ-XAS study

International Nuclear Information System (INIS)

Palancher, H.; Martin, P.; Dubois, S.; Valot, C.; Sabathier, C.; Palancher, H.; Nassif, V.; Proux, O.; Hazemann, J.L.; Wieschalla, N.; Petry, W.; Jarousse, C.

2007-01-01

Full text of publication follows. A worldwide program encourages the use of low enriched uranium (LEU, 235 U 235 U concentration up to 93 wt. %). Due to the decrease in 235 U enrichment for the conversion to LEU, the total density of uranium atoms in the fuel must be increased accordingly. To preserve the neutron flux, metallic uranium alloys could be the best fuel material. The fuel, which consists of UMo alloy spherical particles surrounded by an Al matrix (cf. Figure 1-A), is rolled between two aluminium claddings. Post-irradiation examinations of U-7 wt%Mo demonstrated its strong potentialities as fuel but they also pointed out its interaction with aluminium (cf. Figure 1-B). In certain cases this interaction can cause a break-away swelling of the plate. The aim of this project is the understanding of: - the phenomena driving the growth of the interaction layer. - the influence on interaction layer composition of limited adjunction of elements (silicon...) to the Al matrix. To overcome the difficulties inherent to the in-pile irradiated samples, an out-of-pile methodology (collaboration between CEA, FRM II and CERCA) has been developed based on heavy ion irradiation. This methodology enables to simulate the fission fragment damages using a 80 MeV iodine beam at the Maier Leibnitz laboratory (Garching, Germany). After irradiation, samples are characterised at micrometer scale by microscopy (SEM coupled with EDX) and X-Ray techniques (XRD and XAS). The irradiation (final dose: 2 x 10 17 at/cm 2 ) of undoped U-7 wt%Mo fuel plates leads to the formation of an interaction layer surrounding each fuel particles (cf. Figure 1-C). μ-XRD analysis performed at the ESRF (ID18f) showed only the presence of UAl 3 phase in the interaction layer. Same results have been obtained on in-pile irradiated fuel by Sears et al using neutron diffraction confirming the interest of the developed methodology. However the behaviour of the Mo atoms in the interaction layer could not be

Equipment for detach the fuel elements of the irradiated candu fuel bundle

International Nuclear Information System (INIS)

Cojocaru, V.; Dinuta, G.

2013-01-01

Monitoring the behaviour of the fuel bundles during their combustion provides useful information for the operation of the nuclear power plant as well as for the fuel manufacturer. Before placing it inside the reactor, the fuel bundle is inspected visually, dimensionally and, during combustion in the reactor, its radioactive behaviour is monitored. The purpose of the presented equipment is to allow the visual external inspection of the damaged fuel bundle in order to identify visible defects and to detach the fuel element by breaking the welded connection between the cap and grid. These devices are operated using the handler devices already existing in the hot cells Post-Irradiation Examination Laboratory (LEPI). This equipment has been used successfully in the LEPI laboratory at SCN Pitesti to inspect the damaged fuel from Cernavoda NPP, in March 2013. (authors)

Preliminary results of the BTF-104 experiment: an in-reactor test of fuel behaviour and fission-product release and transport under LOCA/LOECC conditions

Energy Technology Data Exchange (ETDEWEB)

Dickson, L W; Elder, P H; Devaal, J W; Irish, J D; Yamazaki, A R [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

1996-12-31

The BTF-104 experiment is one of a series of in-reactor tests being performed to measure fuel behaviour and fission-product release from nuclear fuel subjected to accident conditions. The primary objective of the BTF-104 experiment was to measure fission-product releases from a CANDU-sized fuel element under combined Loss-of-Coolant Accident (LOCA) and Loss-of-Emergency-Core-Cooling (LOECC) conditions at an average fuel temperature of about 1550 deg C. The preliminary results of the BTF-104 experiment are presented in this paper. (author). 6 refs., 12 figs.

High level radioactive wastes storage characterization and long-term behaviour of spent fuels

International Nuclear Information System (INIS)

Diaz Arocas, P.P.; Garcia Serrano, J.; Mendez Martin, F.J.; Quinones Diez, J.; Rodriguez Almazan, J.L.; Serrano Agejas, J.A.; Esteban Hernandez, J.A.

1997-04-01

The knowledge of long term spent fuel behaviour in a repository is one of the main goals in the waste management assessment due to its influence on repository design topics and on the performance assessment. At the moment, Spain has not selected a geological formation for a final repository. Therefore, R AND D activities are performed by considering granite, salt and clay as candidate options. This report summarises the activities carried out in CIEMAT from 1991 to 1995 in the frame of the Agreement between CIEMAT and ENRESA in the Area of spent fuel direct disposed. Experimental activities include leaching experiments of spent fuel, UO 2 and SIMFUEL and co-precipitation/solubility experiments of relevant secondary solid phases expected under repository conditions. The objective of leaching studies is to understand the processes which will occur when the underground water accede to the source term and to provide leaching rates of spent fuel and the influence of several variables as pH, Eh, etc. The co-precipitation/solubility experiments are focused on the knowledge of the formation conditions of relevant secondary phases, to characterise these phases and to determine their solubility, which could control the leaching of spent fuel. One of the main items to carry out the objectives before indicated in both leaching and co-precipitation/solubility experiments is to perform a extensive solid phases characterisation in order to facilitate the understanding of the processes involved. This report is structured in three parts, the first include experimental procedures, characterisation techniques and solid and solution analyses. The second shows the leaching results obtained by considering the effect of pH, complex formation, redox conditions, surface/volume ratio, etc. The third supply the results of the co-precipitation/solubility studies. The conclusions obtained in this work are considered as the start point of going on and more extensive studies on the mechanisms

21 CFR 201.119 - In vitro diagnostic products.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 4 2010-04-01 2010-04-01 false In vitro diagnostic products. 201.119 Section 201...) DRUGS: GENERAL LABELING Exemptions From Adequate Directions for Use § 201.119 In vitro diagnostic products. (a) “In vitro diagnostic products” are those reagents, instruments and systems intended for use...

Behaviour and effects of prescribed fire in masticated fuelbeds

Science.gov (United States)

Eric Knapp; J. Morgan Varner; Matt Busse; Carl Skinner; Carol Shestak

2011-01-01

Mechanical mastication converts shrub and small tree fuels into surface fuels, and this method is being widely used as a treatment to reduce fire hazard. The compactness of these fuelbeds is thought to moderate fire behaviour, but whether standard fuel models can accurately predict fire behaviour and effects is poorly understood. Prescribed burns were conducted in...

Steady state behaviour of gaseous fission products in UO2 nuclear fuel at low temperature

International Nuclear Information System (INIS)

Rao, C.B.; Raj, Baldev

1980-01-01

Theoretical modelling studies have been performed on steady state fission gas behaviour in UO 2 fuels at temperatures in the range 1073deg K to 1473deg K. The concentrations of gas atoms in the matrix and in the bubbles are determined. Fraction of total generated gas atoms migrating to and forming bubbles at grain boundaries is calculated. Contributions of intragranular and intergranular bubbles to the swelling are also computed. The various assumptions made to simplify computer calculations and their validity are discussed at length. Effects of changes in the fission rate, the resolution parameter, bubble concentration, gas atom diffusivity and grain radius on swelling and gas release are studied. The results of this model are compared to other theoretical models and experimental results available in literature. Possibility of extending the present model to advanced carbide and nitride fuels is discussed. (auth.)

ROBÔ PARA INSPEÇÃO DE ÁREAS CLASSIFICADAS E DE DIFÍCIL ACESSO COM TRANSMISSÃO DE IMAGENS SEM FIO

Directory of Open Access Journals (Sweden)

Jonathan Pereira

2010-01-01

Full Text Available Mediante a necessidade de monitoramento de áreas de difícil acesso ao homem ou áreas que levem risco a sua integridade física, esta pesquisa tem por objetivo construir um robô que faça a captura de imagens por um sistema totalmente sem fio e controlado por computador. Para isso fez-se necessário o projeto e implementação mecânica e eletrônica de um protótipo composto de servomotores, um microcontrolador, um módulo de rádio para a comunicação entre robô e computador e um sistema de captação e transmissão de imagem sem fio. Para a implementação do software que proporciona a execução de comandos remotos, desenvolveu-se uma aplicação utilizando o ambiente C++ Builder®. Em toda a implementação fez-se uso de conceitos visto nas disciplinas de programação de computadores, eletrônica analógica e digital, microcontroladores, sistema de transmissão de dados e mecânica. PALAVRAS-CHAVE: Monitoramento, Robô, microcontrolador.

Geração de solicitação de serviço para inspeção e manutenção em máquinas industriais utilizando redes sem fio

OpenAIRE

Ornelas, Fernando César de

2004-01-01

Dissertação (mestrado) - Universidade Federal de Santa Catarina, Centro Tecnológico. Programa de Pós-graduação em Ciência da Computação Nesta dissertação apresentamos um estudo de caso de aplicação real de rede sem fio, cujo objetivo é a melhoria nas atividades de inspeção e a identificação da necessidade de manutenção em máquinas e equipamentos na área industrial. Fazendo utilização de redes sem fios e de unidades móveis, que se conectam periodicamente a uma rede estruturada, permitimos a...

Quantifying physical characteristics of wildland fuels using the fuel characteristic classification system.

Science.gov (United States)

Cynthia L. Riccardi; Susan J. Prichard; David V. Sandberg; Roger D. Ottmar

2007-01-01

Wildland fuel characteristics are used in many applications of operational fire predictions and to understand fire effects and behaviour. Even so, there is a shortage of information on basic fuel properties and the physical characteristics of wildland fuels. The Fuel Characteristic Classification System (FCCS) builds and catalogues fuelbed descriptions based on...

21 CFR 801.119 - In vitro diagnostic products.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false In vitro diagnostic products. 801.119 Section 801...) MEDICAL DEVICES LABELING Exemptions From Adequate Directions for Use § 801.119 In vitro diagnostic products. A product intended for use in the diagnosis of disease and which is an in vitro diagnostic...

Can age make a difference? A moderated model of altruistic organizational citizenship behaviour antecedents

OpenAIRE

Silvia Profili; Alessia Sammarra; Laura Innocenti

2016-01-01

This paper utilizes lifespan approaches to examine how the effects of fun at work, work-life balance, and perceived supervisor support on altruistic Organizational Citizenship Behaviour (OCB) are moderated by age. Based on multilevel analysis of a large sample of 6,182 employees in 37 companies, fun at work significantly predicted altruism towards co-workers for young employees only, while work-life balance predicted altruistic behaviours for mid- and old-age group employees. Contrary to expe...

Advances in direct oxidation methanol fuel cells

Science.gov (United States)

Surampudi, S.; Narayanan, S. R.; Vamos, E.; Frank, H.; Halpert, G.; Laconti, Anthony B.; Kosek, J.; Prakash, G. K. Surya; Olah, G. A.

1993-01-01

Fuel cells that can operate directly on fuels such as methanol are attractive for low to medium power applications in view of their low weight and volume relative to other power sources. A liquid feed direct methanol fuel cell has been developed based on a proton exchange membrane electrolyte and Pt/Ru and Pt catalyzed fuel and air/O2 electrodes, respectively. The cell has been shown to deliver significant power outputs at temperatures of 60 to 90 C. The cell voltage is near 0.5 V at 300 mA/cm(exp 2) current density and an operating temperature of 90 C. A deterrent to performance appears to be methanol crossover through the membrane to the oxygen electrode. Further improvements in performance appear possible by minimizing the methanol crossover rate.

Canadian power reactor fuel

International Nuclear Information System (INIS)

Page, R.D.

1976-03-01

The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

Safety assessment for the CANFLEX-NU fuel bundles with respect to the 37-element fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Suk, H. C.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-11-01

The KAERI and AECL have jointly developed an advanced CANDU fuel, called CANFLEX-NU fuel bundle. CANFLEX 43-element bundle has some improved features of increased operating margin and enhanced safety compared to the existing 37-element bundle. Since CANFLEX fuel bundle is designed to be compatible with the CANDU-6 reactor design, the behaviour in the thermalhydraulic system will be nearly identical with 37-element bundle. But due to different element design and linear element power distribution between the two bundles, it is expected that CANFLEX fuel behaviour would be different from the behaviour of the 37-element fuel. Therefore, safety assessments on the design basis accidents which result if fuel failures are performed. For all accidents selected, it is observed that the loading of CANFLEX bundle in an existing CANDU-6 reactor would not worsen the reactor safety. It is also predicted that fission product release for CANFLEX fuel bundle generally is lower than that for 37-element bundle. 3 refs., 2 figs., 2 tabs. (Author)

Spent fuel behaviour during dry storage - a review

International Nuclear Information System (INIS)

Shivakumar, V.; Anantharaman, K.

1997-09-01

One of the strategies employed for management of spent fuel prior to their final disposal/reprocessing is their dry storage in casks, after they have been sufficiently cooled in spent fuel pools. In this interim storage, one of the main consideration is that the fuel should retain its integrity to ensure (a) radiological health hazard remains minimal and (b) the fuel is retrievable for down steam fuel management processes such as geological disposal or reprocessing. For dry storage of spent fuel in air, oxidation of the exposed UO 2 is the most severe of phenomena affecting the integrity of fuel. This is kept within acceptable limits for desired storage time by limiting the fuel temperature in the storage cask. The limit on the fuel temperature is met by having suitable limits on maximum burn-up of fuel, minimum cooling period in storage pool and optimum arrangement of fuel bundles in the storage cask from heat removal considerations. The oxidation of UO 2 by moist air has more deleterious effects on the integrity of fuel than that by dry air. The removal of moisture from the storage cask is therefore a very important aspect in dry storage practice. The kinetics of the oxidation phenomena at temperatures expected during dry storage in air is very slow and therefore the majority of the existing data is based on extrapolation of data obtained at higher fuel temperatures. This and the complex effects of factors like fission products in fuel, radiolysis of storage medium etc. has necessitated in having a conservative limiting criteria. The data generated by various experimental programmes and results from the on going programmes have shown that dry storage is a safe and economical practice. (author)

32 CFR 536.119 - Scope for maritime claims.

Science.gov (United States)

2010-07-01

... 32 National Defense 3 2010-07-01 2010-07-01 true Scope for maritime claims. 536.119 Section 536... CLAIMS AGAINST THE UNITED STATES Maritime Claims § 536.119 Scope for maritime claims. The AMCSA applies...) Damage that is maritime in nature and caused by tortious conduct of U.S. military personnel or federal...

7 CFR 28.119 - Fee when request for classification is withdrawn.

Science.gov (United States)

2010-01-01

....119 Section 28.119 Agriculture Regulations of the Department of Agriculture AGRICULTURAL MARKETING SERVICE (Standards, Inspections, Marketing Practices), DEPARTMENT OF AGRICULTURE COMMODITY STANDARDS AND... Cotton Standards Act Fees and Costs § 28.119 Fee when request for classification is withdrawn. When the...

40 CFR 81.119 - Western Tennessee Intrastate Air Quality Control Region.

Science.gov (United States)

2010-07-01

... Quality Control Region. 81.119 Section 81.119 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY... Air Quality Control Regions § 81.119 Western Tennessee Intrastate Air Quality Control Region. The Western Tennessee Intrastate Air Quality Control Region consists of the territorial area encompassed by...

Mechanistic modelling of the corrosion behaviour of copper nuclear fuel waste containers

Energy Technology Data Exchange (ETDEWEB)

King, F; Kolar, M

1996-10-01

A mechanistic model has been developed to predict the long-term corrosion behaviour of copper nuclear fuel waste containers in a Canadian disposal vault. The model is based on a detailed description of the electrochemical, chemical, adsorption and mass-transport processes involved in the uniform corrosion of copper, developed from the results of an extensive experimental program. Predictions from the model are compared with the results of some of these experiments and with observations from a bronze cannon submerged in seawater saturated clay sediments. Quantitative comparisons are made between the observed and predicted corrosion potential, corrosion rate and copper concentration profiles adjacent to the corroding surface, as a way of validating the long-term model predictions. (author). 12 refs., 5 figs.

International Standard problem ISP 14: behaviour of a fuel bundle simulator during a specified heatup and flooding period (Rebeka experiment): results of post-test analyses: final comparison report

International Nuclear Information System (INIS)

Karwat, H.

1985-02-01

The test consisted in investigating the non-steady material behaviour of a bundle of electrically heated fuel rod simulators with respect to local fuel temperatures, cladding strain, time to burst and local strain at location of burst, together with the thermal hydraulic boundary conditions. The original aim has not been fully achievable. The applied codes for mechanical fuel behaviour largely demonstrated their capabilities for pretest predictions when certain local fluid dynamic parameters are well known to the code users. The difficulties expected with proper analysis of thermal hydraulics of the test were confirmed, caused by the coupling between pin cooling conditions, rod upper plenum calculations and the feedback to clad deformation and burst simulation

Fuel performance evaluation through iodine activity monitoring

International Nuclear Information System (INIS)

Anantharaman, K.; Chandra, R.

1995-01-01

The objective of the failed fuel detection system is to keep a watch on fuel behaviour during operation. This paper describes the evaluation of fuel behaviour by monitoring the activities of various isotopes of iodine both during steady state and during a reactor shutdown. The limitations of this approach also has been explained. The monitoring of tramp uranium for different types of release, namely fixed contamination and continuous release from fuel, is also presented. (author)

GenExp: an interactive web-based genomic DAS client with client-side data rendering.

Directory of Open Access Journals (Sweden)

Bernat Gel Moreno

Full Text Available BACKGROUND: The Distributed Annotation System (DAS offers a standard protocol for sharing and integrating annotations on biological sequences. There are more than 1000 DAS sources available and the number is steadily increasing. Clients are an essential part of the DAS system and integrate data from several independent sources in order to create a useful representation to the user. While web-based DAS clients exist, most of them do not have direct interaction capabilities such as dragging and zooming with the mouse. RESULTS: Here we present GenExp, a web based and fully interactive visual DAS client. GenExp is a genome oriented DAS client capable of creating informative representations of genomic data zooming out from base level to complete chromosomes. It proposes a novel approach to genomic data rendering and uses the latest HTML5 web technologies to create the data representation inside the client browser. Thanks to client-side rendering most position changes do not need a network request to the server and so responses to zooming and panning are almost immediate. In GenExp it is possible to explore the genome intuitively moving it with the mouse just like geographical map applications. Additionally, in GenExp it is possible to have more than one data viewer at the same time and to save the current state of the application to revisit it later on. CONCLUSIONS: GenExp is a new interactive web-based client for DAS and addresses some of the short-comings of the existing clients. It uses client-side data rendering techniques resulting in easier genome browsing and exploration. GenExp is open source under the GPL license and it is freely available at http://gralggen.lsi.upc.edu/recerca/genexp.

Application of Exp-function method to potential Kadomtsev-Petviashvili equation

International Nuclear Information System (INIS)

Xian Daquan; Dai Zhengde

2009-01-01

Exact periodic kink-wave solution, periodic soliton and doubly periodic solutions for the potential Kadomtsev-Petviashvii (PKP) equation are obtained using Exp-function method with the help of Maple computation.

Behaviour of conductivity improvers in jet fuel

Energy Technology Data Exchange (ETDEWEB)

Dacre, B.; Hetherington, J.I. [Cranfield Univ., Wiltshire (United Kingdom)

1995-05-01

Dangerous accumulation of electrostatic charge can occur due to high speed pumping and microfiltration of fuel. This can be avoided by increasing the electrical conductivity of the fuel using conductivity improver additives. However, marked variations occur in the conductivity response of different fuels when doped to the same level with conductivity improver. This has been attributed to interactions of the conductivity improver with other fuel additives or fuel contaminants. The present work concentrates on the effects of fuel contaminants, in particular polar compounds, on the performance of the conductivity improver. Conductivity is the fuel property of prime interest. The conductivity response of model systems of the conductivity improver STADIS 450 in dodecane has been measured and the effect on this conductivity of additions of model polar contaminants sodium naphthenate, sodium dodecyl benzene sulphonate, and sodium phenate have been measured. The sodium salts have been found to have a complex effect on the performance of STADIS 450, reducing the conductivity at low concentrations to a minimum value and then increasing the conductivity at high concentrations of sodium salts. This work has focused on characterising this minimum in the conductivity values and on understanding the reason for its occurrence. The effects on the minimum conductivity value of the following parameters are investigated: (a) time, (b) STADIS 450 concentration, (c) sodium salt concentration, (d) mixed sodium salts, (e) experimental method, (f) a phenol, (g) individual components of STADIS 450. The complex conductivity response of the STADIS 450 to sodium salt impurities is discussed in terms of possible inter-molecular interactions.

Between and within laboratory reliability of mouse behaviour recorded in home-cage and open-field.

Science.gov (United States)

Robinson, Lianne; Spruijt, Berry; Riedel, Gernot

2018-04-15

Reproducibility of behavioural findings between laboratories is difficult due to behaviour being sensitive to environmental factors and interactions with genetics. The objective of this study was to investigate reproducibility of behavioural data between laboratories using the PhenoTyper home cage observation system and within laboratory reproducibility using different lighting regimes. The ambulatory activity of C57BL/6 and DBA/2 mice was tested in PhenoTypers in two laboratories under near identical housing and testing conditions (Exp. 1). Additionally activity and anxiety were also assessed in the open-field test. Furthermore, testing in either a normal or inverted light/dark cycle was used to determine effects of lighting regime in a within-laboratory comparison in Aberdeen (Exp. 2). Using the PhenoTyper similar circadian rhythms were observed across laboratories. Higher levels of baseline and novelty-induced activity were evident in Aberdeen compared to Utrecht although strain differences were consistent between laboratories. Open field activity was also similar across laboratories whereas strain differences in anxiety were different. Within laboratory analysis of different lighting regimes revealed that behaviour of the mice was sensitive to changes in lighting. Utilisation of a home cage observation system facilitates the reproducibility of activity but not anxiety-related behaviours across laboratories by eliminating environmental factors known to influence reproducibility in standard behavioural tests. Standardisation of housing/test conditions resulted in reproducibility of home cage and open field activity but not anxiety-related phenotypes across laboratories with some behaviours more sensitive to environmental factors. Environmental factors include lighting and time of day. Copyright © 2017 Elsevier B.V. All rights reserved.

Instrumentation of fuel elements and fuel plates

International Nuclear Information System (INIS)

Durand, J.P.; Fanjas, Y.

1993-01-01

When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have led to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

Instrumentation of fuel elements and fuel plates

International Nuclear Information System (INIS)

Durand, J.P.; Fanjas, Y.

1994-01-01

When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have lead to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

HTR fuel modelling with the ATLAS code. Thermal mechanical behaviour and fission product release assessment

International Nuclear Information System (INIS)

Guillermier, Pierre; Daniel, Lucile; Gauthier, Laurent

2009-01-01

To support AREVA NP in its design on HTR reactor and its HTR fuel R and D program, the Commissariat a l'Energie Atomique developed the ATLAS code (Advanced Thermal mechanicaL Analysis Software) with the objectives: - to quantify, with a statistical approach, the failed particle fraction and fission product release of a HTR fuel core under normal and accidental conditions (compact or pebble design). - to simulate irradiation tests or benchmark in order to compare measurements or others code results with ATLAS evaluation. These two objectives aim at qualifying the code in order to predict fuel behaviour and to design fuel according to core performance and safety requirements. A statistical calculation uses numerous deterministic calculations. The finite element method is used for these deterministic calculations, in order to be able to choose among three types of meshes, depending on what must be simulated: - One-dimensional calculation of one single particle, for intact particles or particles with fully debonded layers. - Two-dimensional calculations of one single particle, in the case of particles which are cracked, partially debonded or shaped in various ways. - Three-dimensional calculations of a whole compact slice, in order to simulate the interactions between the particles, the thermal gradient and the transport of fission products up to the coolant. - Some calculations of a whole pebble, using homogenization methods are being studied. The temperatures, displacements, stresses, strains and fission product concentrations are calculated on each mesh of the model. Statistical calculations are done using these results, taking into account ceramic failure mode, but also fabrication tolerances and material property uncertainties, variations of the loads (fluence, temperature, burn-up) and core data parameters. The statistical method used in ATLAS is the importance sampling. The model of migration of long-lived fission products in the coated particle and more

Final Report. Fumex-III. Improvement of Models Used for Fuel Behaviour Simulation

International Nuclear Information System (INIS)

Kulacsy, Katalin

2013-01-01

The FUMEX-III coordinated research programme organised by the IAEA was the first FUMEX exercise in which AEKI (Hungarian Academy of Sciences KFKI Atomic Energy Research Institute) took part with the partial support of Paks NPP. The aim of the participation was to test the code FUROM developed at AEKI against not only measurements but also other fuel behaviour simulation codes, to share and discuss modelling experience and issues, and to establish acquaintance with fuel modellers in other countries. Among the numerous cases proposed for the programme, AEKI chose to simulate normal operation up to high burn-up and ramp tests, with special interest in VVER rods and PWR rods with annular pellets. The US PWR 16x16, the SPC RE GINNA, the Kola3-MIR, the IFA-519.9 cases and the AREVA idealised rod were thus selected. The present Final Report gives a short description of the FUROM models relevant to the selected cases, presents the results for the 5 cases and summarises the conclusions of the FUMEX-III programme. The input parameters used for the simulations can be found in the Appendix at the end of the Report. Observations concerning the IFPE datasets are collected for each dataset in their respective Sections for possible use in the IFPE database. (author)

7 CFR 1221.119 - Refunds.

Science.gov (United States)

2010-01-01

... Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (MARKETING AGREEMENTS... INFORMATION ORDER Sorghum Promotion, Research, and Information Order Sorghum Promotion, Research, and Information Board § 1221.119 Refunds. Any producer or importer from whom an assessment is collected and...

Boiling and fragmentation behaviour during fuel-sodium interactions

International Nuclear Information System (INIS)

Schins, H.; Gunnerson, F.S.

1986-01-01

A selection of the results and subsequent analysis of molten fuel-sodium interaction experiments conducted within the JRC BETULLA I and II facilities are reported. The fuels were copper and stainless steel, at initial temperatures far above their melting points; or urania and alumina, initially at their melting points. For each test, the molten fuel masses were in lower kilogram range and the subcooled pool mass was either 160 or 4 kg. The sodium pool was instrumented continually monitor the system temperature and pressure. Post-test examination results of the fragmented fuel debris sizes, shape and crystalline structure are given. The results of this study suggest the following: Transition boiling is the dominant boiling mode for the tested fuels in subcooled sodium. Two fragmentation mechanisms, vapour bubble formation/collapse and thermal stress shrinkage cracking prevailed for the oxide fuels. This was evidenced by the presence of both smooth and fractured particulate. In contrast, all metal fuel debris was smooth, suggesting fragmentation by the vapour bubble formation/collapse mechanism only during the molten state and for each test, there was no evidence of an energetic fuel-coolant interaction. (orig.)

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

International Nuclear Information System (INIS)

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.)

37 CFR 11.9 - Limited recognition in patent matters.

Science.gov (United States)

2010-07-01

... 37 Patents, Trademarks, and Copyrights 1 2010-07-01 2010-07-01 false Limited recognition in patent matters. 11.9 Section 11.9 Patents, Trademarks, and Copyrights UNITED STATES PATENT AND TRADEMARK OFFICE, DEPARTMENT OF COMMERCE REPRESENTATION OF OTHERS BEFORE THE UNITED STATES PATENT AND TRADEMARK OFFICE...

Fuel loads and fuel type mapping

Science.gov (United States)

Chuvieco, Emilio; Riaño, David; Van Wagtendonk, Jan W.; Morsdof, Felix; Chuvieco, Emilio

2003-01-01

Correct description of fuel properties is critical to improve fire danger assessment and fire behaviour modeling, since they guide both fire ignition and fire propagation. This chapter deals with properties of fuel that can be considered static in short periods of time: biomass loads, plant geometry, compactness, etc. Mapping these properties require a detail knowledge of vegetation vertical and horizontal structure. Several systems to classify the great diversity of vegetation characteristics in few fuel types are described, as well as methods for mapping them with special emphasis on those based on remote sensing images.

Use of advanced simulations in fuel performance codes

International Nuclear Information System (INIS)

Van Uffelen, P.

2015-01-01

The simulation of the cylindrical fuel rod behaviour in a reactor or a storage pool for spent fuel requires a fuel performance code. Such tool solves the equations for the heat transfer, the stresses and strains in fuel and cladding, the evolution of several isotopes and the behaviour of various fission products in the fuel rod. The main equations along with their limitations are briefly described. The current approaches adopted for overcoming these limitations and the perspectives are also outlined. (author)

Protocolo para a distribuição de informações de canais em redes em malha sem fio

Directory of Open Access Journals (Sweden)

Roni Francis Shigueta

2014-04-01

Full Text Available Este artigo apresenta um protocolo para a distribuição de informações de canais em uma rede em malha sem fio. O canal corresponde à faixa de frequência na qual dois dispositivos sem fio (nós da rede podem se comunicar. O protocolo apresentado permite criar um repositório que contém os canais que estão disponíveis em um dispositivo para que se comunique com o seu nó vizinho. A partir das informações desse repositório, os dispositivos podem tomar decisões de alocação de canais, baseadas em algum critério, como por exemplo, a menor interferência gerada na rede. Dependendo da quantidade de canais disponíveis, é possível atribuir canais diferentes aos enlaces da rede (par de nós, permitindo transmissões simultâneas. A simulação do protocolo é realizada por meio de canais de frequência do padrão IEEE 802.11b. Apesar das simulações serem realizadas nesse padrão, o protocolo pode ser adaptado para operar em outras faixas de frequência. Os resultados demonstram que a partir dos valores dos canais transmitidos através de mensagens de hello, é possível encontrar os canais comuns entre dois nós, realizando a intersecção entre a lista de canais do nó vizinho com a lista do nó local.

Grain boundary sweeping and dissolution effects on fission product behaviour under severe fuel damage accident conditions

International Nuclear Information System (INIS)

Rest, J.

1986-01-01

The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory of grain boundary sweeping of gas bubbles, gas bubble behavior during fuel liquefaction (destruction of grain boundaries due to formation of a U-rich melt phase), and during U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. The grain-boundary-sweeping theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges. The theory of the effects of fuel liquefaction and U-Zr eutectic melting on fission product behaviour considers the migration and coalescence of fission gas bubbles in either molten uranium, or a Zircaloy-Uranium eutectic melt. Results of the analyses demonstrate that intragranular fission product behavior during the tests can be interpreted in terms of a grain-growth/grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. Whereas fuel liquefaction leads to an enhanced release of fission products in trace-irradiated fuel, the occurrence of fuel liquefaction in normally-irradiated fuel can degrade fission product release. This phenomenon is due in part to reduced gas-bubble mobilities in a viscous medium as compared to vapor transport, and in part to a degradation of grain growth rates and the subsequent decrease in grain-boundary sweeping of intragranular fission products into the liquified lamina. The analysis shows that total UO 2 dissolution due to eutectic melting leads to increased release for both trace-irradiated and normally-irradiated fuel. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in normally

CFD analysis of aircraft fuel tanks thermal behaviour

Science.gov (United States)

Zilio, C.; Longo, G. A.; Pernigotto, G.; Chiacchio, F.; Borrelli, P.; D'Errico, E.

2017-11-01

This work is carried out within the FP7 European research project TOICA (Thermal Overall Integrated Conception of Aircraft, http://www.toica-fp7.eu/). One of the tasks foreseen for the TOICA project is the analysis of fuel tanks as possible heat sinks for future aircrafts. In particular, in the present paper, commercial regional aircraft is considered as case study and CFD analysis with the commercial code STAR-CCM+ is performed in order to identify the potential capability to use fuel stored in the tanks as a heat sink for waste heat dissipated by other systems. The complex physical phenomena that characterize the heat transfer inside liquid fuel, at the fuel-ullage interface and inside the ullage are outlined. Boundary conditions, including the effect of different ground and flight conditions, are implemented in the numerical simulation approach. The analysis is implemented for a portion of aluminium wing fuel tank, including the leading edge effects. Effect of liquid fuel transfer among different tank compartments and the air flow in the ullage is included. According to Fuel Tank Flammability Assessment Method (FTFAM) proposed by the Federal Aviation Administration, the results are exploited in terms of exponential time constants and fuel temperature difference to the ambient for the different cases investigated.

Not all mice are equal: welfare implications of behavioural habituation profiles in four 129 mouse substrains.

Directory of Open Access Journals (Sweden)

Hetty Boleij

Full Text Available Safeguarding the welfare of animals is an important aim when defining housing and management standards in animal based, experimental research. While such standards are usually defined per animal species, it is known that considerable differences between laboratory mouse strains exist, for example with regard to their emotional traits. Following earlier experiments, in which we found that 129P3 mice show a lack of habituation of anxiety related behaviour after repeated exposure to an initially novel environment (non-adaptive profile, we here investigated four other 129 inbred mouse substrains (129S2/SvPas, 129S2/SvHsd (exp 1; 129P2 and 129X1 (exp 2 on habituation of anxiety related behaviour. Male mice of each strain were repeatedly placed in the modified hole board test, measuring anxiety-related behaviour, exploratory and locomotor behaviour. The results reveal that all four substrains show a lack of habituation behaviour throughout the period of testing. Although not in all of the substrains a possible confounding effect of general activity can be excluded, our findings suggest that the genetic background of the 129 substrains may increase their vulnerability to cope with environmental challenges, such as exposure to novelty. This vulnerability might negatively affect the welfare of these mice under standard laboratory conditions when compared with other strains. Based on our findings we suggest to consider (substrain-specific guidelines and protocols, taking the (substrain-specific adaptive capabilities into account.

Modelling the actual behaviour of the MOX fuel by a micromechanical analysis in non-uniform transformation fields

International Nuclear Information System (INIS)

Largenton, R.

2012-01-01

This research thesis aimed at developing a model based on scale change to assess more precisely the distribution of local thermo-mechanical fields within a heterogeneous medium as MOX fuel. The analysis method is a non-uniform transformation field analysis (NTFA) which is adapted to the problem of scale change in presence of a coupling between dissipative and elastic effects. More precisely, the author addressed the development of a NTFA model based on specific three-phase and three-dimensional microstructures which are typical of the MOX fuel in an in-service operation. The first part proposes an overview of knowledge and use of MOX. It recalls the context and the industrial problematic associated with this fuel: operating principles for a 900 MWe PWR, fuel fabrication processes, fuel morphologies and structural and microstructural consequences. It addresses local mechanisms within each phase during irradiation, and presents the approach methodology regarding scale change. The second part reports the representation and analysis in complete fields of multiphase particle-based composites (MOX type) in order to determine the representative elementary volume and the local behaviour of each phase. The third part reports the extension of the NTFA approach to 3D aspects, free deformations, ageing and optimization. The last part compares the NTFA approach with the incremental two-phase and three-phase Mori-Tanaka models

Further analysis of extended storage of spent fuel. Final report of a co-ordinated research programme on the behaviour of spent fuel assemblies during extended storage (BEFAST-III) 1991-1996

International Nuclear Information System (INIS)

1997-05-01

Considerable quantities of spent fuel continue to be produced and to accumulate in a number of countries. Although some new reprocessing facilities have been constructed, many countries are investigating the option of extended spent fuel storage prior to reprocessing or fuel disposal. Wet storage continues to predominate as an established technology. However, dry storage is becoming increasingly used with many countries considering dry storage for the longer term. This Technical Document is the final report of the IAEA Co-ordinated Research Programme on the Behaviour of Spent Fuel Assemblies During Extended Storage (BEFAST-III, 1991-1996). It contains analyses of wet and dry spent fuel storage technologies obtained from 16 organizations representing 13 countries (Canada, Finland, France, Germany, Hungary, the Republic of Korea, Japan, the Russian Federation, Slovakia, Spain, Sweden, the United Kingdom and the USA) which participated in the co-ordinated research programme as participants or observers. The report contains information presented during the three Research Co-ordination meetings and also data which were submitted by the participants in response to request by the Scientific Secretary. 48 refs, 4 tabs

Pellet-clad interaction in water reactor fuels

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

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

Pellet-clad interaction in water reactor fuels

International Nuclear Information System (INIS)

2004-01-01

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

Technology developments for Japanese BWR MOX fuel utilization

International Nuclear Information System (INIS)

Oguma, M.; Mochida, T.; Nomata, T.; Asahi, K.

1997-01-01

The Long-Term Program for Research, Development and Utilization of Nuclear Energy established by the Atomic Energy Commission of Japan asserts that Japan will promote systematic utilization of MOX fuel in LWRs. Based on this Japanese nuclear energy policy, we have been pushing development of MOX fuel technology aimed at future full scale utilization of this fuel in BWRs. In this paper, the main R and D topics are described from three subject areas, MOX core and fuel design, MOX fuel irradiation behaviour, and MOX fuel fabrication technology. For the first area, we explain the compatibility of MOX fuel with UO 2 core, the feasibility of the full MOX core, and the adaptability of MOX design methods based on a mock-up criticality experiment. In the second, we outline the Tsuruga MOX irradiation program and the DOMO program, and suggest that MOX fuel behaviour is comparable to ordinary BWR UO 2 fuel behaviour. In the third, we examine the development of a fully automated MOX bundle assembling apparatus and its features. (author). 14 refs, 11 figs, 3 tabs

Automated FiO2-SpO2 control system in neonates requiring respiratory support: a comparison of a standard to a narrow SpO2 control range.

Science.gov (United States)

Wilinska, Maria; Bachman, Thomas; Swietlinski, Janusz; Kostro, Maria; Twardoch-Drozd, Marta

2014-05-28

Managing the oxygen saturation of preterm infants to a target range has been the standard of care for a decade. Changes in target ranges have been shown to significantly impact mortality and morbidity. Selecting and implementing the optimal target range are complicated not only by issues of training, but also the realities of staffing levels and demands. The potential for automatic control is becoming a reality. Results from the evaluation of different systems have been promising and our own experience encouraging. This study was conducted in two tertiary level newborn nurseries, routinely using an automated FiO2-SpO2 control system (Avea-CLiO2, Yorba Linda CA, USA). The aim of this study was to compare the performance of the system as used routinely (set control range of 87-93% SpO2), to a narrower higher range (90-93%). We employed a 12-hour cross-over design with the order of control ranges randomly assigned for each of up to three days. The primary prospectively identified end points were time in the 87-93% SpO2 target range, time at SpO2 extremes and the distribution of the SpO2 exposure. Twenty-one infants completed the study. The infants were born with a median EGA of 27 weeks and studied at a median age of 17 days and weight of 1.08 kg. Their median FiO2 was 0.32; 8 were intubated, and the rest noninvasively supported (7 positive pressure ventilation and 6 CPAP). The control in both arms was excellent, and required less than 2 manual FiO2 adjustments per day. There were no differences in the three primary endpoints. The narrower/higher set control range resulted in tighter control (IQR 3.0 vs. 4.3 p < 0.001), and less time with the SpO2 between 80-86 (6.2% vs. 8.4%, p = 0.006). We found that a shift in the median of the set control range of an automated FiO2-SpO2 control system had a proportional effect on the median and distribution of SpO2 exposure. We found that a dramatic narrowing of the set control range had a disproportionally smaller impact. Our

Experimental simulation of irradiation effects on thermomechanical behaviour of UO2 fuel: Impact of solid and gaseous fission products

International Nuclear Information System (INIS)

Balland, J.

2007-12-01

Predictive simulation of thermomechanical behaviour of nuclear fuel has to take into account irradiation effects. Fission Products (FP) can modify the thermomechanical behaviour of UO 2 . During this thesis, differentiation was made between fission products which create a solid solution with UO 2 and gaseous products, generating pressurized bubbles. SIMFUELS containing gadolinium oxide and pressurized argon bubbles were manufactured, respectively by conventional process and by Gas Pressure Sintering. Brittle and ductile behaviour of UO 2 was investigated, under experimental conditions representative of Pellet-Cladding Interaction (PCI), respectively with 3 points bending tests and compressive creep tests. Investigation of brittle behaviour of UO 2 showed that fracture is mainly controlled by natural defects, like porosities, acting like starting points for cracks propagation. Addition of simulates fission products increase the brittle-to-ductile transition temperature of UO 2 , up to 400-500 C regarding FP in solid solution, and up to 200 C for gaseous products. Fission products although reduce fracture stresses, by a factor between 1.5 and 4, respectively for gas bubbles and solid solutions. Decrease of fracture stress is linked to an increase of microstructural defects due the solid solution and to pressurized bubbles located at grain boundaries. Pellets were tested under compressive solicitation at high temperatures. Experimental results of creep tests are well represented by Norton laws. Creep controlling mechanisms are evidenced by microstructural analysis performed on pellets at different strains. On the basis of calculations made for fuels having the same microstructures than the SIMFUELs, a creep factor is determined. It revealed a strong hardening effect of the solid solution, due to the fact that the added elements anchor the dislocations, whereas pressurized bubbles showed a coupling between hardening and softening effects. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Helfer, Th

2006-03-15

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

Fuel mechanical design as a boundary condition for fuel management optimization

International Nuclear Information System (INIS)

Wunderlich, F.; Aisch, F.W.; Heins, L.

1988-01-01

The incentive to reduce fuel cycle costs as well as the amount of active waste requires, among others, measures to optimize fuel management. Improved fuel management in this sense calls, e.g., for reduction of parasitic neutron absorption, for reduction of neutron leakage, and particularly for burnup extension. Such measures result in increased demands for fuel mechanical design. In the first part of this paper their impact on fuel mechanical behaviour is described. In the second part, some examples of practical importance for the interaction between fuel management optimization and fuel mechanical design are discussed. (orig.) [de

19 CFR 11.9 - Special marking on certain articles.

Science.gov (United States)

2010-04-01

... 19 Customs Duties 1 2010-04-01 2010-04-01 false Special marking on certain articles. 11.9 Section... OF THE TREASURY PACKING AND STAMPING; MARKING Marking § 11.9 Special marking on certain articles. (a... of additional U.S. Note 4, Chapter 91. If any article so required to be marked is found not to be...

Lifetime measurements and the nonaxial deformation in 119I

International Nuclear Information System (INIS)

Srebrny, J.; Droste, Ch.; Morek, T.; Starosta, K.; Juutinen, S.; Piiparinen, M.; Toermaenen, S.; Virtanen, A.

2000-01-01

Complete text of publication follows. Lifetimes in four negative parity bands of 119 I were measured using DSAM and RDM. 119 I nuclei were produced in the 109 Ag( 13 C,3n) reaction, γγ coincidences were collected using the NORDBALL array. The detailed description of experiment is given in (1,2,3). Information about electromagnetic properties of four negative parity bands, originating from the h 11/2 quasiproton coupled to an axially asymmetric core, was obtained. The lifetimes of 31 negative parity levels were determined. That is one of the largest sets of electromagnetic transition probabilities for an odd - A nucleus from the 50 119 I nucleus. We see that the 53-rd proton added to the 118 Te nucleus, through the polarisation effect, changes the properties of the even-even core. The β-deformation becomes at least as large as that of 120 Xe (β ∼ 0.28), whereas the γ-deformation is around 30 deg. Comparison of experimental data with calculation within Core Quasiparticle Coupling Model indicates the advantage of the γ- soft model over the γ-rigid one in the description of h 11/2 band structure in 119 I. One can see, that the most valuable information concerning the shape of 119 I is based on the properties of the unfavoured states, especially those belonging to band 9, with their regular energy spacing and fast intraband transitions. (author)

Preliminary analysis of the creep behaviour of nuclear fuel-waste container materials

International Nuclear Information System (INIS)

Dutton, R.; Leitch, B.W.; Crosthwaite, J.L.; Kasprick, G.R.

1996-12-01

In the Canadian Nuclear Fuel Waste Management Program, it is proposed that nuclear fuel waste be placed in a durable container and disposed of in a deep underground vault. Consideration of various disposal-container designs has identified either titanium or copper as the material suitable for constructing the container shell. As part of the R and D program to examine the structural integrity of the container, creep tests are being conducted on commercially pure titanium and oxygen-free copper. This report presents the preliminary data obtained. It also describes the evaluation of various constitutive equations to represent the creep curves, thus providing the basis for extrapolation of the creep behaviour over the design lifetime of the container. In this regard, a specific focus is placed on equations derived from the 0-Projection Concept. Recognizing that the container lifetime will be determined by the onset of tertiary creep leading to creep rupture, we present the results of the metallographic examination of creep damage. This shows that the tertiary stage in titanium is associated with the formation of transgranular cavities within the region of localized necking of the creep specimens. In contrast, creep damage in copper is in the form of intergranular cavities uniformly distributed throughout the gauge length. These results are analyzed within the context of the extant literature, and their implications for future container design are discussed. (author)

Third international conference on CANDU fuel

International Nuclear Information System (INIS)

Boczar, Peter

1992-01-01

These proceedings contain full texts of all 49 papers from the ten sessions and the banquet address. The sessions were on the following subjects: International experience and programs; Fuel behaviour and operating experience; Fuel modelling; Fuel design; Advanced fuel and fuel cycle technology; AECL's concept for the disposal of nuclear fuel waste. The individual papers have been abstracted separately

Emotional and behavioural symptoms, risk behaviours and academic success in Chilean Mapuche and non-Mapuche adolescents.

Science.gov (United States)

Martínez-González, Agustín Ernesto; Rodríguez-Jiménez, Tíscar; Piqueras, José Antonio; Vera-Villarroel, Pablo; Torres-Ortega, Jorge

2018-02-28

There is controversy over the real existence of differences in mental health and academic performance between the Mapuche ethnic minority male adolescents and the male adolescents not belonging to this ethnicity in Chile. In consequence, the aim of this study was to investigate the differences in emotional and behavioural symptoms, risky behaviours and academic success on the Chilean Mapuche and non-Mapuche adolescents. The sample consisted of 233 adolescents of which 119 were Mapuche adolescents and 114 were non-Mapuche adolescents. The results showed that the Mapuche adolescents do not have more anxiety problems and depression than the non-Mapuche adolescents. Furthermore, the Mapuche adolescents present less drug consumption and behavioural problems. Moreover, there were no differences in academic performance. This study provides social interest data of the adolescents' mental health, which can be useful for the country's socio-sanitary and political decisions. Future studies should investigate these and other variables related to the mental health of minorities in greater depth.

Etude numérique et expérimentale de l'evaporation d'une ou plusieurs gouttes de mélange de carburants dans un écoulement chauffé

Science.gov (United States)

Daïf, A.; Ali Chérif, A.; Bresson, J.; Sarh, B.

1995-10-01

The vaporization of one or two multi-component fuel droplets in hot air-stream is presented. A thermal wind tunnel with experimental channel has been designed to develop an experimental process. Firstly, the comparison between experimental results and numerical data is presented for the case of an isolated multi-component droplet. The numerical method is based on the resolution of heat and mass transfer equations between the droplet and the gas stream. This model includes the effect of Stephan flow, the effect of variable thermophysical properties of the components, and the non-unitary Lewis number in the gas film. The experimental results show the micro-explosion phenomenon observed in the liquid phase of multi-component droplet at low temperature. The experimental case of two pure or multi-component droplets in interaction is also presented. On présente un article de synthèse sur l'évaporation d'une ou deux gouttes de carburants à plusieurs composants dans un écoulement d'air chaud. Un dispositif expérimental constitué d'une soufflerie thermique, avec veine d'expérimentation, est réalisé pour permettre cette étude. Pour le cas d'une goutte isolée, une comparaison expérience-calcul est entreprise. Le principe de la méthode numerique consiste en la résolution des équations de transfert de masse et de chaleur entre la goutte et l'écoulement. Ce modèle prend en compte les effets de l'écoulement de Stephan, les variations des propriétés thermophysiques des composants dans les deux phases et la valeur du nombre de Lewis différente de l'unité dans le film de vapeur. Outre l'analyse plus approfondie qu'apporte la confrontation entre le calcul et l'expérience, les résultats expérimentaux montrent le phénomène de micro-explosion observé à l'intérieur de la goutte liquide. Le cas expérimental de deux gouttes en interaction est abordé qu'il s'agisse de gouttes de carburant pur ou de mélange.

Correlations between fuel pins irradiated in fast and thermal fluxes using the frump fuel pin modelling program

International Nuclear Information System (INIS)

Hayns, M.R.; Adam, J.

1975-08-01

There is no experimental facilities in which a fuel pin can be irradiated in a fast environment under well defined conditions of over power or flow run down. Consequently most of the infor mation which is being accumulated on the behaviour of fuel pins under severe conditions is obtained from either capsule or loop rigs in thermal reactors. It is the purpose of this paper to highlight the differences between the behaviour of fuel pins irradiated in a thermal flux and a fast flux. A typical set of conditions is taken from an overpower experiment in a thermal flux and the behaviour of the system is analysed using the fuel modelling program FRUMP. A second numerical experiment is then performed in which the same conditions prevail, except that a fast flux is assumed, the criterion for comparison being that the total power input to the system is the same in both cases. From the many possible correlations which result from such an exercise the fuel tempreature has been selected to highlight various important features of the two irradiations. It is demonstrated that the flux depression can cause differences in the pin behaviour, even to altering the order of events in a transient. For example fuel melting will occur at different times and at different positions in the fuel in the two cases. It is concluded that the techniques of fuel modelling, as typified in the program FRUMP can provide a very useful tool indeed for the analysis of such experiments and for guiding the establishment of the appropriate correlations for the extrapolation to the fast flux case. (author)

A methodology to analyze the creep behaviour of nuclear fuel waste containers

Energy Technology Data Exchange (ETDEWEB)

Dutton, R [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

1995-12-01

The concept for the disposal of used-fuel waste from CANDU reeactors operating in Canada comprises a system of natural and engineered barriers surrounding the waste in a mined vault situated at a depth of 500 - 1000 m in plutonic rock of the Canadian Shield. The fuel would be packaged in a highly durable metal container, within a matrix of compacted particulate. The design of the container takes into account that it would be subjected to an external hydrostatic pressure. Consideration of the rate of radioactive decay of the radionuclides contained in the fuel, suggests that the lifetime of the container should be at least 500 years. Consequently, the role of creep deformation, and the possibility of creep rupture of the container shell, must be included in the assessment of time-dependent mechanical integrity. This report describes an analytical approach that can be used to quantify the long-term creep properties of the container material and facilitate the engineering design. The overall objective is to formulate a constitutive creep equation that provides the required input for a finite element computer model being developed to analyze the elastic-plastic behaviour of the container. Alternative forms of such equations are reviewed. It is shown that the capability of many of these equations to extrapolate over long time scales is limited by their empirical nature. Thus, the recommended equation is based on current mechanistic understanding of creep deformation and creep rupture. A criterion for determining the onset of material failure by creep rupture, that could be used in the design of containers with extended structural integrity, is proposed. Interpretation and extrapolation will be supported by the complementary Deformation and Fracture Mechanism Maps. (author) 103 refs., 2 tabs., 54 figs.

A methodology to analyze the creep behaviour of nuclear fuel waste containers

International Nuclear Information System (INIS)

Dutton, R.

1995-12-01

The concept for the disposal of used-fuel waste from CANDU reeactors operating in Canada comprises a system of natural and engineered barriers surrounding the waste in a mined vault situated at a depth of 500 - 1000 m in plutonic rock of the Canadian Shield. The fuel would be packaged in a highly durable metal container, within a matrix of compacted particulate. The design of the container takes into account that it would be subjected to an external hydrostatic pressure. Consideration of the rate of radioactive decay of the radionuclides contained in the fuel, suggests that the lifetime of the container should be at least 500 years. Consequently, the role of creep deformation, and the possibility of creep rupture of the container shell, must be included in the assessment of time-dependent mechanical integrity. This report describes an analytical approach that can be used to quantify the long-term creep properties of the container material and facilitate the engineering design. The overall objective is to formulate a constitutive creep equation that provides the required input for a finite element computer model being developed to analyze the elastic-plastic behaviour of the container. Alternative forms of such equations are reviewed. It is shown that the capability of many of these equations to extrapolate over long time scales is limited by their empirical nature. Thus, the recommended equation is based on current mechanistic understanding of creep deformation and creep rupture. A criterion for determining the onset of material failure by creep rupture, that could be used in the design of containers with extended structural integrity, is proposed. Interpretation and extrapolation will be supported by the complementary Deformation and Fracture Mechanism Maps. (author) 103 refs., 2 tabs., 54 figs

ParaExp Using Leapfrog as Integrator for High-Frequency Electromagnetic Simulations

Science.gov (United States)

Merkel, M.; Niyonzima, I.; Schöps, S.

2017-12-01

Recently, ParaExp was proposed for the time integration of linear hyperbolic problems. It splits the time interval of interest into subintervals and computes the solution on each subinterval in parallel. The overall solution is decomposed into a particular solution defined on each subinterval with zero initial conditions and a homogeneous solution propagated by the matrix exponential applied to the initial conditions. The efficiency of the method depends on fast approximations of this matrix exponential based on recent results from numerical linear algebra. This paper deals with the application of ParaExp in combination with Leapfrog to electromagnetic wave problems in time domain. Numerical tests are carried out for a simple toy problem and a realistic spiral inductor model discretized by the Finite Integration Technique.

Moderation of the 119mSn isomer radioactive decay

International Nuclear Information System (INIS)

Godovikov, S.K.

1999-01-01

The evaluation of the constant of the braked 119m Sn nuclei decay in the Moessbauer source, being for a long time in contact with a resonance shield, is carried out. The high stability of these nuclei relative to decay is established. The 119m Sn subjected to prolonged impact of the standing electromagnetic wave field become resistant to radioactive decay [ru

Models for fuel rod behaviour at high burnup

Energy Technology Data Exchange (ETDEWEB)

Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

2004-12-01

This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

Effects of alpha-decay on spent fuel corrosion behaviour

International Nuclear Information System (INIS)

Wiss, T.; Rondinella, V.V.; Cobos, J.; Wegen, D.H.; Amme, M.; Ronchi, C.

2004-01-01

An overview of results in the area of spent fuel characterization as nuclear waste is presented. These studies are focused on primary aspects of spent fuel corrosion, by considering different fuel compositions and burn ups, as well as a wide set of environmental conditions. The key parameter is the storage time of the fuel e.g. in view of spent fuel retrieval or in view of its final disposal. To extrapolate data obtainable from a laboratory-acceptable timescale to those expected after storage periods of interest have elapsed (amounting in the extreme case to geological ages) is a tough challenge. Emphasis is put on key aspects of fuel corrosion related to fuel properties at a given age and environmental conditions expected in the repository: e.g. the fuel activity (radiolysis effects), the effects of helium build-up and of groundwater composition. A wide range of techniques, from traditional leaching experiments to advanced electrochemistry, and of materials, including spent fuel with different compositions/burnups and analogues like the so-called alpha-doped UO 2 , are employed for these studies. The results confirm the safety of European underground repository concepts. (authors)

Eletroerosão por fio em metal duro para ferramentas de estampagem de lâminas de motores elétricos

OpenAIRE

Hespanhol, Heber de Carvalho

2009-01-01

Dissertação (mestrado) - Universidade Federal de Santa Catarina, Centro Tecnológico. Programa de Pós-Graduação em Engenharia Mecânica. O processo de eletroerosão por fio é utilizado atualmente na fabricação da maioria das matrizes e punções de metal duro (carbeto de tungstênio), usados em ferramentas de estampagem de aços para motores elétricos. As indústrias de eletrodomésticos, motores elétricos industriais, transformadores e compressores herméticos, utilizam ferramentas progressivas de...

Long term wet spent nuclear fuel storage

International Nuclear Information System (INIS)

1987-04-01

The meeting showed that there is continuing confidence in the use of wet storage for spent nuclear fuel and that long-term wet storage of fuel clad in zirconium alloys can be readily achieved. The importance of maintaining good water chemistry has been identified. The long-term wet storage behaviour of sensitized stainless steel clad fuel involves, as yet, some uncertainties. However, great reliance will be placed on long-term wet storage of spent fuel into the future. The following topics were treated to some extent: Oxidation of the external surface of fuel clad, rod consolidation, radiation protection, optimum methods of treating spent fuel storage water, physical radiation effects, and the behaviour of spent fuel assemblies of long-term wet storage conditions. A number of papers on national experience are included

Reactive-transport model for the prediction of the uniform corrosion behaviour of copper used fuel containers

International Nuclear Information System (INIS)

King, F.; Kolar, M.; Maak, P.

2008-01-01

Used fuel containers in a deep geological repository will be subject to various forms of corrosion. For containers made from oxygen-free, phosphorus-doped copper, the most likely corrosion processes are uniform corrosion, underdeposit corrosion, stress corrosion cracking, and microbiologically influenced corrosion. The environmental conditions within the repository are expected to evolve with time, changing from warm and oxidizing initially to cool and anoxic in the long-term. In response, the corrosion behaviour of the containers will also change with time as the repository environment evolve. A reactive-transport model has been developed to predict the time-dependent uniform corrosion behaviour of the container. The model is based on an experimentally-based reaction scheme that accounts for the various chemical, microbiological, electrochemical, precipitation/dissolution, adsorption/desorption, redox, and mass-transport processes at the container surface and in the compacted bentonite-based sealing materials within the repository. Coupling of the electrochemical interfacial reactions with processes in the bentonite buffer material allows the effect of the evolution of the repository environment on the corrosion behaviour of the container to be taken into account. The Copper Corrosion Model for Uniform Corrosion predicts the time-dependent corrosion rate and corrosion potential of the container, as well as the evolution of the near-field environment

Calibration of Sn-119 isomer shift using ab initio wave function methods

NARCIS (Netherlands)

Kurian, Reshmi; Filatov, Michael

2009-01-01

The isomer shift for the 23.87 keV M1 resonant transition in the Sn-119 nucleus is calibrated with the help of ab initio calculations. The calibration constant alpha(Sn-119) obtained from Hartree-Fock (HF) calculations (alpha(HF)(Sn-119)=(0.081 +/- 0.002)a(0)(-3) mm/s) and from second-order

Fuel behavior modeling using the MARS computer code

International Nuclear Information System (INIS)

Faya, S.C.S.; Faya, A.J.G.

1983-01-01

The fuel behaviour modeling code MARS against experimental data, was evaluated. Two cases were selected: an early comercial PWR rod (Maine Yankee rod) and an experimental rod from the Canadian BWR program (Canadian rod). The MARS predictions are compared with experimental data and predictions made by other fuel modeling codes. Improvements are suggested for some fuel behaviour models. Mars results are satisfactory based on the data available. (Author) [pt

Reactor based plutonium disposition - physics and fuel behaviour benchmark studies of an OECD/NEA experts group

International Nuclear Information System (INIS)

D'Hondt, P.; Gehin, J.; Na, B.C.; Sartori, E.; Wiesenack, W.

2001-01-01

One of the options envisaged for disposing of weapons grade plutonium, declared surplus for national defence in the Russian Federation and Usa, is to burn it in nuclear power reactors. The scientific/technical know-how accumulated in the use of MOX as a fuel for electricity generation is of great relevance for the plutonium disposition programmes. An Expert Group of the OECD/Nea is carrying out a series of benchmarks with the aim of facilitating the use of this know-how for meeting this objective. This paper describes the background that led to establishing the Expert Group, and the present status of results from these benchmarks. The benchmark studies cover a theoretical reactor physics benchmark on a VVER-1000 core loaded with MOX, two experimental benchmarks on MOX lattices and a benchmark concerned with MOX fuel behaviour for both solid and hollow pellets. First conclusions are outlined as well as future work. (author)

Fuel rod behaviour during transients

International Nuclear Information System (INIS)

Hughes, H.; Haste, T.J.; Cameron, R.F.; Sinclair, J.E.

1982-04-01

The fuel pin performance code SLEUTH, the transient codes FRAP-T5 and TRAFIC and the clad deformation code CANSWEL-2 are described. It is shown how the codes treat gas release, pin cooling, cladding deformation and interaction, gap conductance etc. The materials properties used are indicated. (author)

Some Remarks on Exp-Function Method and Its Applications

International Nuclear Information System (INIS)

Aslan Ismail; Marinakis Vangelis

2011-01-01

Recently, many important nonlinear partial differential equations arising in the applied physical and mathematical sciences have been tackled by a popular approach, the so-called Exp-function method. In this paper, we present some shortcomings of this method by analyzing the results of recently published papers. We also discuss the possible improvement of the effectiveness of the method. (general)

Petição para confirmação do direito do consumidor de utilização de programa de comunicações pela internet em redes sem fio e de conexão de dispositivos em tais redes

Directory of Open Access Journals (Sweden)

Christopher Libertelli

2011-05-01

Full Text Available À medida que a indústria de telecomunicações sem fio amadurece, a consolidação e o relacionamento enFtre as empresas detentoras de infraestrutura de transporte [carriers] e os produtores de aparelhos celulares [handsets] têm revelado práticas de mercado que levantam questionamentos substanciais sobre se os consumidores estariam desfrutando de todos os benefícios possíveis oriundos da competição nas telecomunicações sem fio. Por exemplo, empresas detentoras de infraestrutura de transporte de telecomunicações têm começado a influenciar agressivamente o design de programas [software] e produtos em detrimento do consumidor. Com o amadurecimento do mercado de telecomunicações sem fio e o reconhecimento de que os aparelhos celulares tornaram-se um componente indispensável para muitos americanos, empresas detentoras de infraestrutura de transporte têm utilizado de sua considerável influência sobre o uso e design destes aparelhos para manutenção do controle e dos limites ao direito dos assinantes de executar aplicativos comunicacionais de sua escolha. Ao invés de transportarem as mensagens dos assinantes independentemente do conteúdo, empresas detentoras de infraestrutura de transporte de telecomunicações têm exercido cada vez mais controle sobre a forma como os consumidores acessam a internet móvel. Ao darem preferência a serviços próprios e ao afastarem seus rivais, ditas empresas têm desabilitado ou inutilizado interfaces amigáveis ao consumidor [consumer-friendly features] em dispositivos móveis. Além disso, as empresas detentoras de infraestrutura de transporte de telecomunicações têm implementado tais práticas em violação ao princípio Carterfone da FCC e às restrições constantes da decisão da Comissão em sua formulação original referente à autorização de venda conjugada [bundling] do terminal de acesso individual e do serviço celular. A Comissão deve agir agora para fazer valer a decis

A model describing intra-granular fission gas behaviour in oxide fuel for advanced engineering tools

Science.gov (United States)

Pizzocri, D.; Pastore, G.; Barani, T.; Magni, A.; Luzzi, L.; Van Uffelen, P.; Pitts, S. A.; Alfonsi, A.; Hales, J. D.

2018-04-01

The description of intra-granular fission gas behaviour is a fundamental part of any model for the prediction of fission gas release and swelling in nuclear fuel. In this work we present a model describing the evolution of intra-granular fission gas bubbles in terms of bubble number density and average size, coupled to gas release to grain boundaries. The model considers the fundamental processes of single gas atom diffusion, gas bubble nucleation, re-solution and gas atom trapping at bubbles. The model is derived from a detailed cluster dynamics formulation, yet it consists of only three differential equations in its final form; hence, it can be efficiently applied in engineering fuel performance codes while retaining a physical basis. We discuss improvements relative to previous single-size models for intra-granular bubble evolution. We validate the model against experimental data, both in terms of bubble number density and average bubble radius. Lastly, we perform an uncertainty and sensitivity analysis by propagating the uncertainties in the parameters to model results.

A Small Molecule that Targets r(CGG)exp and Improves Defects in Fragile X-Associated Tremor Ataxia Syndrome

Science.gov (United States)

Disney, Matthew D.; Liu, Biao; Yang, Wang-Yong; Sellier, Chantal; Tran, Tuan; Charlet-Berguerand, Nicolas; Childs-Disney, Jessica L.

2012-01-01

The development of small molecule chemical probes or therapeutics that target RNA remains a significant challenge despite the great interest in such compounds. The most significant barrier to compound development is a lack of knowledge of the chemical and RNA motif spaces that interact specifically. Herein, we describe a bioactive small molecule probe that targets expanded r(CGG) repeats, or r(CGG)exp , that causes Fragile X-associated Tremor Ataxia Syndrome (FXTAS). The compound was identified by using information on the chemotypes and RNA motifs that interact. Specifically, 9-hydroxy-5,11-dimethyl-2-(2-(piperidin-1-yl)ethyl)-6H-pyrido[4,3-b]carbazol-2-ium, binds the 5’CGG/3’GGC motifs in r(CGG)exp and disrupts a toxic r(CGG)exp -protein complex in vitro. Structure-activity relationships (SAR) studies determined that the alkylated pyridyl and phenolic side chains are important chemotypes that drive molecular recognition to r(CGG)exp . Importantly, the compound is efficacious in FXTAS model cellular systems as evidenced by its ability to improve FXTAS-associated pre-mRNA splicing defects and to reduce the size and number of r(CGG)exp -protein aggregates. This approach may establish a general strategy to identify lead ligands that target RNA while also providing a chemical probe to dissect the varied mechanisms by which r(CGG)exp promotes toxicity. PMID:22948243

The behaviour of water-cooled reactor fuel rods in steady state and transient conditions; Zachowanie sie pretow paliwowych reaktorow chlodzonych woda w stanach ustalonych i nieustalonych

Energy Technology Data Exchange (ETDEWEB)

Strupczewski, A.; Marks, P. [Institute of Atomic Energy, Otwock-Swierk (Poland)

1997-12-31

In this report, the results of temperature field and filling gas pressure calculations by means of contemporary calculational models for a WWER-440 and WWER-1000 type fuel rod at low and high burnup operating under steady-state conditions are presented. A review of in-core temperature and pressure measurements for various types of LWR fuel is also included. Basing on calculational and collected measured data, the behaviour of fuel cladding during large and small break LOCA, is estimated with special emphasis on their oxidation and failure resistance. (author) 38 refs, 40 figs, 15 tabs

Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

Energy Technology Data Exchange (ETDEWEB)

Barrio del Juanes, M T; Garcia Cuesta, J C; Vallejo Diaz, I; Puebla, Herranz

2001-07-01

Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

Energy Technology Data Exchange (ETDEWEB)

Barrio del Juanes, M.T.; Garcia Cuesta, J.C.; Vallejo Diaz, I.; Herranz Puebla

2001-07-01

Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

Solitary wave solutions of the fourth order Boussinesq equation through the exp(-Ф(η))-expansion method.

Science.gov (United States)

Akbar, M Ali; Hj Mohd Ali, Norhashidah

2014-01-01

The exp(-Ф(η))-expansion method is an ascending method for obtaining exact and solitary wave solutions for nonlinear evolution equations. In this article, we implement the exp(-Ф(η))-expansion method to build solitary wave solutions to the fourth order Boussinesq equation. The procedure is simple, direct and useful with the help of computer algebra. By using this method, we obtain solitary wave solutions in terms of the hyperbolic functions, the trigonometric functions and elementary functions. The results show that the exp(-Ф(η))-expansion method is straightforward and effective mathematical tool for the treatment of nonlinear evolution equations in mathematical physics and engineering. 35C07; 35C08; 35P99.

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

International Nuclear Information System (INIS)

2000-10-01

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

Computer modelling of water reactor fuel element performance and life time

International Nuclear Information System (INIS)

Bibilashvili, Yu.K.; Golovnin, I.S.; Elesin, V.F.

1983-01-01

Well calibrated models and methods of calculation permit the confident prediction of fuel element behaviour under most different operational conditions; based on the prediction of this kind one can improve designs and fuel element behaviour. Therefore, in the Soviet Union in the development of reactor cores for NPP one of the leading parts is given to design problems associated with computer modelling of fuel element performance and reliability. Special attention is paid to methods of calculation that permit the prediction of fuel element behaviour under conditions which either make experimental studies very complicated (practically impossible) or require laborious and expensive in-pile tests. Primarily it concerns accidents of different types, off-normal conditions, transients, fuel element behaviour at high burn-up, when an accumulation of a great amount of fission fragments is accompanied by changes in physical and mechanical properties as induced by irradiation damage, mechanical fatigue, physical and chemical reactions with a coolant, fission products etc. Some major computer modelling programs for the prediction of water reactor fuel behaviour are briefly described below and tendencies in the further development of work in this area are summarized

Fuel performance-REP, Seminars on nuclear fuel performance based on basic underlining phenomena, proceedings

International Nuclear Information System (INIS)

2008-01-01

Description: The need for further improving the understanding of basic phenomena underlying nuclear fuel behaviour has been recognised both by fuel vendors, experts in fuel research in the different laboratories and committees and working groups coordinating international activities. The OECD/NEA Nuclear Science Committee has established an Experts Group addressing this issue. This has led to establishing an International Fuel Performance Experiments Database (IFPE) that should help model evaluation and validation. Many years ago the IAEA established an International Working Group on Fuel Performance and Technology (IWGFPT) that led to the FUMEX-I and FUMEX-II (Fuel Modelling Exercise) which has had an important impact on code improvements. Both international organisations, with the support of national organisations, co-operate in establishing and maintaining the Database and to build confidence in the predictive power of the models through international comparison exercises. But above all the different parties have agreed that seminars focussed on specific phenomena would be beneficial to exchange current knowledge, identify outstanding problems and agree on common action that would lead to improved understanding of the phenomena. A series of three seminars has been initiated by the Commissariat a l'Energie Atomique (CEA), Electricite de France (EdF), Framatome and Cogema under the aegis of the OECD/NEA and the IAEA. 1. Thermal Performance of High Burn-Up LWR Fuel at Cadarache, France, from 3 to 6 of March 1998. Thermal performance occupies the most important aspect of the fuel performance modelling. Not only is it extremely important from a safety point of view, but also many of the material properties of interest and behaviour, such as transport properties like fuel creep and fission gas release are thermally activated processes. Thus, in order to model these processes correctly, it is critical to calculate temperatures and their distribution as accurately as

Investigation of WWER fuel behaviour under MIR power ramps

International Nuclear Information System (INIS)

Bibilashvili, Yu.K.; Novikov, V.V.; Agafonov, S.N.

1996-01-01

The paper discusses results of experimental WWER fuel investigation under power ramps. Specificity of using the research reactor ''MIR'' to accomplish scheduled power rating of fuel is considered. The paper presents the methodology of experiments using irradiation facility ''TEST''. Reactor experiments were performed at burn-up ∼ 10000 MW.day/t UO 2 using standard fuel pins and the ones having backfitted fuel and cladding. (author). 7 figs, 1 tab

Exp-function method for solving fractional partial differential equations.

Science.gov (United States)

Zheng, Bin

2013-01-01

We extend the Exp-function method to fractional partial differential equations in the sense of modified Riemann-Liouville derivative based on nonlinear fractional complex transformation. For illustrating the validity of this method, we apply it to the space-time fractional Fokas equation and the nonlinear fractional Sharma-Tasso-Olver (STO) equation. As a result, some new exact solutions for them are successfully established.

Family influences on physical activity and sedentary behaviours in Chinese junior high school students: a cross-sectional study.

Science.gov (United States)

Wang, Xin; Liu, Qing-Min; Ren, Yan-Jun; Lv, Jun; Li, Li-Ming

2015-03-25

Family influence plays an important role in a child's physical activity (PA). This study aimed to describe the level of moderate to vigorous intensity physical activity (MVPA) and sedentary behaviours among Chinese junior high school students and examine the associations between different types of family influence and MVPA or sedentary behaviours. Participants of two independent cross-sectional surveys, conducted in 2009 and 2011, were students in Grade 7 and 9 from all junior high schools in Hangzhou, China. The daily duration and frequency of MVPA, amount of sedentary time and frequency of family support were self-reported. Multi-level mixed-effects logistic regression was used to examine the associations between different types or levels of family influence and MVPA or sedentary behaviours. A total of 7286 students were analysed finally. Overall, only 9.0% of the students participated in MVPA at least 60 minutes/day; 63.9% spent no more than 2 hours/day in sedentary behaviours. Frequent verbal encouragement and watching were associated with less leisure-time sedentary behaviours. The multivariate-adjusted odds ratios (ORs) for verbal encouragement and watching were 1.29 (95% CI, 1.08 to 1.55) and 1.19 (95% CI, 0.97 to 1.45) for 5-7 days per week. The involvement of family in the children's activity in most days of the week was associated with both higher level of MVPA and less leisure-time sedentary behaviours. The respective ORs among students who reported familial support 5-7 days per week, were 1.50 (95% CI, 1.21 to 1.86) for engaging in seven days of MVPA per week, 1.67 (95% CI, 1.19 to 2.32) for at least 60 minutes of MVPA daily, and 1.48 (95% CI, 1.19 to 1.84) for no more than 2 hours of leisure-time sedentary behaviours daily. This study found that less than 10.0% of urban Chinese adolescents engaged in MVPA at least 60 minutes/day. Family involving themselves in the children's activity exerted the most significant influence on children's behaviours

Approche expérimentale de l'utilisation de glyphosate dans le ...

African Journals Online (AJOL)

Approche expérimentale de l'utilisation de glyphosate dans le contrôle de Melaleuca quinquenervia (Myrtaceae), une espèce envahissante dans la réserve communautaire de la forêt d'Analalava-Foulpointe (Madagascar)

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

Energy Technology Data Exchange (ETDEWEB)

Ferreira Neto, Ricardo Alberto

2000-11-15

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

Transition probabilities in negative parity bands of the 119I nucleus

International Nuclear Information System (INIS)

Srebrny, J.; Droste, Ch.; Morek, T.; Starosta, K.; Wasilewski, A.A.; Pasternak, A.A.; Podsvirova, E.O.; Lobach, Yu.N.; Hagemann, G.H.; Juutinen, S.; Piiparinen, M.; Toermaenen, S.; Virtanen, A.

2001-01-01

Lifetimes in four negative-parity bands of 119 I were measured using DSAM and RDM. 119 I nuclei were produced in the 109 Ag( 13 C,3n) reaction, γγ coincidences were collected using the NORDBALL array. RDDSA - a new method of RDM analysis - is described. This method allowed for the self-calibration of stopping power. From 31 measured lifetimes, 39 values of B(E2) were established. Calculations in the frame of the Core Quasi Particle Coupling (CQPC) model were focused on the problem of susceptibility of the nucleus to γ-deformation. It was established that nonaxial quadrupole deformation of 119 I plays on important role. The Wilets-Jean model of a γ-soft nucleus describes the 119 I nucleus in a more consistent way then the Davydov-Filippov model of a γ-rigid nucleus

RIA tests in CABRI with MOX fuel

International Nuclear Information System (INIS)

Schmitz, F.; Papin, J.; Gonnier, C.

2000-01-01

Three MOX-fuel tests have been successfully performed within the framework of the CABRI REP-Na test program. From the experimental findings which are presently available, no evidence for thermal effects resulting from the heterogeneous nature of the fuel can be given. There are very clear hints however that fission gas effects are enhanced with regard to the behaviour of UO 2 . The clad rupture observed in REP-Na 7 is of different nature than the failures observed in Cabri tests with UO 2 fuel. Failures of UO 2 fuel rods only occurred when the clad mechanical properties were severely affected by the presence of hydride blisters, while in REP-Na 7 a clear indication is made that the loading potential of the MOX fuel pellets was high enough to break a sound cladding. Concerning the transient fuel behaviour after reaching the critical heat-flux under reactor typical conditions (pressure, temperature and flow), no data base could be provided by the tests in the present sodium test loop (as for the UO 2 fuel behaviour). The IPSN project to implement into the Cabri reactor a pressurised water loop which will allow to simulate the complete RIA accident sequence under PWR reactor typical conditions, aims at providing this missing data base. (author)

CANDU fuel bundle deformation modelling with COMSOL multiphysics

International Nuclear Information System (INIS)

Bell, J.S.; Lewis, B.J.

2012-01-01

Highlights: ► The deformation behaviour of a CANDU fuel bundle was modelled. ► The model has been developed on a commercial finite-element platform. ► Pellet/sheath interaction and end-plate restraint effects were considered. ► The model was benchmarked against the BOW code and a variable-load experiment. - Abstract: A model to describe deformation behaviour of a CANDU 37-element bundle has been developed under the COMSOL Multiphysics finite-element platform. Beam elements were applied to the fuel elements (composed of fuel sheaths and pellets) and endplates in order to calculate the bowing behaviour of the fuel elements. This model is important to help assess bundle-deformation phenomena, which may lead to more restrictive coolant flow through the sub-channels of the horizontally oriented bundle. The bundle model was compared to the BOW code for the occurrence of a dry-out patch, and benchmarked against an out-reactor experiment with a variable load on an outer fuel element.

Flow behaviour in a CANDU horizontal fuel channel from stagnant subcooled initial conditions

International Nuclear Information System (INIS)

Caplan, M.Z.; Gulshani, P.; Holmes, R.W.; Wright, A.C.D.

1984-01-01

The flow behaviour in a CANDU primary system with horizontal fuel channels is described following a small inlet header break. With the primary pumps running, emergency coolant injection is in the forward direction so that the channel outlet feeders remain warmer than the inlet thereby promoting forward natural circulation. However, the break force opposes the forward driving force. Should the primary pumps run down after the circuit has refilled, there is a break size for which the natural circulation force is balanced by the break force and channels could, theoretically, stagnate. Result of visualization and of full-size channel tests on channel flow behaviour from an initially stagnant channel condition are discussed. After a channel stagnation, the decay power heats the coolant to saturation. Steam is then formed and the coolant stratifies. The steam expands into the subcooled water in the end fitting in a chugging type of flow regime due to steam condensation. After the end fitting reaches the saturation temperature, steam is able to penetrate into the vertical feeder thereby initiating a large buoyancy induced flow which refills the channel. The duration of stagnation is shown to be sensitive to small asymmetries in the initial conditions. A small initial flow can significantly shorten the occurrence and/or duration of boiling as has been confirmed by reactor experience. (author)

ExpI and PhzI are descendants of the long lost cognate signal synthase for SdiA.

Directory of Open Access Journals (Sweden)

Anice Sabag-Daigle

Full Text Available SdiA of E. coli and Salmonella is a LuxR homolog that detects N-acyl homoserine lactones (AHLs. Most LuxR homologs function together with a cognate AHL synthase (a LuxI homolog, but SdiA does not. Instead, SdiA detects AHLs produced by other bacterial species. In this report, we performed a phylogenetic analysis of SdiA. The results suggest that one branch of the Enterobacteriaceae obtained a rhlR/rhlI pair by horizontal transfer. The Erwinia and Pantoea branches still contain the complete pair where it is known as expR/expI and phzR/phzI, respectively. A deletion event removed the luxI homolog from the remainder of the group, leaving just the luxR homolog known as sdiA. Thus ExpR and PhzR are SdiA orthologs and ExpI and PhzI are descendants of the long lost cognate signal synthase of SdiA.

Driving Behaviour and Sustainable Mobility—Policies and Approaches Revisited

Directory of Open Access Journals (Sweden)

Ali Keyvanfar

2018-04-01

Full Text Available Climate change is receiving increasing attention in recent years. The transportation sector contributes substantially to increased fuel consumption, greenhouse gas (GHG emissions, and poor air quality, which imposes a serious respiratory health hazard. Road transport has made a significant contribution to this effect. Consequently, many countries have attempted to mitigate climate change using various strategies. This study analysed and compared the number of policies and other approaches necessary to achieve reduced fuel consumption and carbon emission. Frequency aggregation indicates that the mitigation policies associated with driving behaviours adopted to curtail this consumption and decrease hazardous emissions, as well as a safety enhancement. Furthermore, car-sharing/carpooling was the least investigated approach to establish its influence on mitigation of climate change. Additionally, the influence of such driving behaviours as acceleration/deceleration and the compliance to speed limits on each approach was discussed. Other driving behaviours, such as gear shifting, compliance to traffic laws, choice of route, and idling and braking style, were also discussed. Likewise, the influence of aggression, anxiety, and motivation on driving behaviour of motorists was highlighted. The research determined that driving behaviours can lead to new adaptive driving behaviours and, thus, cause a significant decrease of vehicle fuel consumption and CO2 emissions.

The modeling experience of fuel element units operation under MSC.MARC and MENTAT 2008R1

International Nuclear Information System (INIS)

Kulakov, G.; Kashirin, B.; Kosaurov, A.; Konovalov, Y.; Kuznetsov, A.; Medvedev, A.; Novikov, V.; Vatulin, A.

2009-01-01

MSC Software is leading developer of CAE-software in the world, so behaviour of fuel elements modeling with MSC.MARC use is of great practical importance. Behaviour of fuel elements usually is modeled in the elastic-viscous-plastic statement with account on fuel swelling during irradiation. For container type fuel elements contact interaction between fuel pellets and cladding or other parts of fuel element in top and bottom plugs must be in account. Results of simulated behaviour of various type fuel elements - container type fuel elements for PWR and RBMK reactors, dispersion type fuel elements for research reactors are presented. (authors)

Fuel corrosion processes under waste disposal conditions

International Nuclear Information System (INIS)

Shoesmith, D.W.

2000-01-01

The release of the majority of radionuclides from spent nuclear fuel under permanent disposal conditions will be controlled by the rate of dissolution of the UO 2 fuel matrix. In this manuscript the mechanism of the coupled anodic (fuel dissolution) and cathodic (oxidant reduction) reactions which constitute the overall fuel corrosion process is reviewed, and the many published observations on fuel corrosion under disposal conditions discussed. The primary emphasis is on summarizing the overall mechanistic behaviour and establishing the primary factors likely to control fuel corrosion. Included are discussions on the influence of various oxidants including radiolytic ones, pH, temperature, groundwater composition, and the formation of corrosion product deposits. The relevance of the data recorded on unirradiated UO 2 to the interpretation of spent fuel behaviour is included. Based on the review, the data used to develop fuel corrosion models under the conditions anticipated in Yucca Mountain (NV, USA) are evaluated

The Synergy between Scuba Diving and Household Behaviour: Testing Plastic and Food Waste "The use of natural habitats for tourism education"

OpenAIRE

Soares Mota, Luís Cândido

2014-01-01

The activity of scuba diving is used for studying behaviours of U.S. visitors to a popular tourist destination in Mexico. The impact created by human activity can produce marine debris and therefore affect the marine environment. The subpopulation of 181 divers was tested for their current household practices regarding discarding plastic and food waste, providing quantitative statistics for divers’ referential behaviour. Prior to partaking in scuba diving, certified, trainee, and “one-day-exp...

Radionuclides and isotopes release of spent fuel matrix. Conceptual and mathematical models of wastes behaviour

International Nuclear Information System (INIS)

Cera, E.; Merino, J.; Bruno, J.

2000-01-01

We have developed a conceptual and numerical model to calculate release of selected radionuclides from spent fuel under repository condition. This has been done in the framework of the Enresa 2000 performance assessment exercise. The model has been developed based on kinetic mass balance equations in order to study the evolution of the spent fuel water interface as a function of time. Several processes have been kinetically modelled: congruent dissolution, radioactive decay, ingrowth and water turnover in the gap. The precipitation/redissolution of secondary solid phases has been taken into account from a thermodynamic point of view. Both approaches have been coupled and the resulting equations solved for a number of radionuclides in both, a conservative and realistic approach. The results show three distinct groups of radionuclides based on their release behaviour: a first group is composed of radioisotopes of highly insoluble elements (e. g., Pu, Am, Pd) whose concentration in the gap is mainly controlled by their solubility and therefore their evolution is identical in both cases. Secondly, a set of radionuclides from soluble elements under these conditions (e. g., I, Cs, Ra) show concentrations kinetically controlled, decreasing with time following the congruent dissolution trend. Their release concentrations are one order of magnitude larger in the conservative case than in the realistic case. Finally, a third group has been identified (e. g., Se, Th, Cm) where a mixed behaviour takes place: initially their solubility limiting phases control their concentration in the gap but the situation reverts to a kinetic control as the chemical conditions change and the secondary precipitates become totally dissolved. The fluxes of the different radionuclides are also given as an assessment of the source term in the performance assessment. (Author)

Fission gas behaviour modelling in plate fuel during a power transient

International Nuclear Information System (INIS)

Portier, S.

2003-01-01

This thesis is dedicated to the identification and modelization of the phenomena which are at the origin of the release of the fission gas formed in UO 2 plate fuels during the irradiation in a power transient. In the first experimental part, samples of plate fuels, irradiated at 36 GWj/tU, have been annealed to temperatures from 1100 C to 1500 C in a device that enabled the measurement of gas release in real time. At 1300 C, post-annealing observations demonstrated a link between the measured gas releases to a rapid formation of labyrinths at the grain surface. These labyrinths, which were formed by intergranular bubble interconnection, create release paths for the gas atoms which reach the grain surface. At this stage, the available experimental results (annealing and observations) were interpreted considering that it is the spreading of the gas atoms from the grains to the grain boundaries that is at the origin of the observed releases. This interpretation generates the hypothesis that a) at the end of the basic irradiation, the gas is at the atomic state and b) during the annealing, the spreading is reduced by the intragranular bubbles of the gas atoms. The last part of the work is dedicated to the modelization of the main phenomena at the origin of the gas release. The model developed, based on the model of the gas behaviour in MARGARET PWR, highlighted the great influence of the irradiation conditions on the gas distribution at the end of the irradiation and also its influence on the fission gas release during the power transient. (author) [fr

Attitudes toward suicidal behaviour among professionals at mental health outpatient clinics in Stavropol, Russia and Oslo, Norway

OpenAIRE

Norheim, Astrid Berge; Grimholt, Tine K.; Loskutova, Ekaterina; Ekeberg, Oivind

2016-01-01

Background Attitudes toward suicidal behaviour can be essential regarding whether patients seek or are offered help. Patients with suicidal behaviour are increasingly treated by mental health outpatient clinics. Our aim was to study attitudes among professionals at outpatient clinics in Stavropol, Russia and Oslo, Norway. Methods Three hundred and forty-eight (82?%) professionals anonymously completed a questionnaire about attitudes. Professionals at outpatient clinics in Stavropol (n?=?119; ...

29 CFR 794.119 - Dependence of exemption on sales volume of the enterprise.

Science.gov (United States)

2010-07-01

... 29 Labor 3 2010-07-01 2010-07-01 false Dependence of exemption on sales volume of the enterprise. 794.119 Section 794.119 Labor Regulations Relating to Labor (Continued) WAGE AND HOUR DIVISION... Act Annual Gross Volume of Sales § 794.119 Dependence of exemption on sales volume of the enterprise...

S Tank Farm SL-119 saltwell piping failure analysis report

International Nuclear Information System (INIS)

Carlos, W.C.

1994-01-01

On January 24, 1992, while pressure testing saltwell line SL-119 in the 241-S Tank Farm, water was observed spraying out of heat trace enclosure. The SL-115, SL-116, SN-215, and SN-216 saltwell lines also recently failed pressure testing because of leaks. This study documents the pertinent facts about the SL-119 line and discusses the cause of the failures. The inspection of the SL-119 failure revealed two through-the-wall holes in the top center of the pipeline. The inspection also strongly suggests that the heat tracing system is directly responsible for causing the SL-119 failure. Poor design of the heat tracing system allowed water to enter, condense, and collect in the electric metallic tubing (EMT) carbon steel conduits. Water flowed to the bottom of the elbow of the conduit and corroded out the elbow. The design also allowed drifting desert sand to enter into the conduit and fall to the bottom (elbow) of the conduit. The sand became wet and aided in the corrosion of the elbow of the conduit. After the EMT conduits corroded though, the water dripped from the corroded ends of the EMT conduits onto the top of the saltwell pipe, corroding the two holes into the top of the line. If the heat tracing hot splice box had not allowed moisture to enter the EMT conduits, the saltwell piping would not have corroded and caused SL-119 to fail

Application of 119Sn Moessbauer spectroscopy to investigations of RET2-X2-type phases

International Nuclear Information System (INIS)

Goerlich, E.

1991-01-01

Moessbauer spectroscopy of 119 Sn is an effective tool to study nuclear hyperfine interactions which in turn are important as a source of information complementary to that supplied by ''classical'' methods. However, particularly when the effects are subtle, experimental conditions as well as data analysis should be handled with care. The attention is devoted primarily to the finite absorber thickness effects. As examples serve our recent results obtained from investigations of ternary tetragonal phases of RET 2 X 2 -type. In EuPd 2 Si 2 mixed valent system 119 Sn was used as probe which detects the influence of Eu-valency change at a distant site of Si. Electrical resistivity measurements in CeNi 2 Sn 2 may indicate a Kondo-type behaviour while Moessbauer effect leads to the conclusion in favour of a magnetically ordered state. The analysis of resonance absorption spectra of CeAg 2 Sn 2 using a full transmission integral indicates the presence of a metallic β-tin in the sample, while data fitting within the thin absorber approximation has lead to the opposite conclusion. Neither in CeNi 2 Sn 2 nor in ErNi 2 Sn 2 the temperature dependence of the recoil free factors alone cannot explain the observed broadening of spectra at low temperatures. (author). 14 refs, 13 figs

Spatial Graduation of Fuel Taxes

Energy Technology Data Exchange (ETDEWEB)

Rietveld, P.; Van Vuuren, D. [Tinbergen Institute, Labor, Region and Environment, Amsterdam/Rotterdam (Netherlands); Bruinsma, F. [Department of Spatial Economics, Faculty of Economics and Econometrics, Vrije Universiteit Amsterdam, Amsterdam (Netherlands)

1999-06-01

Substantial differences exist among fuel taxes in various countries. These differences represent a form of fiscal competition that has undesirable side effects because it leads to cross-border fuelling and hence to extra kilometres driven. One possible way of solving the problem of low fuel taxes in neighbouring countries is to introduce a spatial differentiation of taxes: low near the border and higher further away. This paper contains an empirical analysis of the consequences of such a spatial graduation of fuel taxes for the Netherlands. We will analyse impacts on fuelling behaviour, vehicle kilometres driven, tax receipts, and sales by owners of gas stations. The appropriate slope of the graduation curve is also discussed. Our conclusion is that in a small country such as the Netherlands, a spatial graduation of fuel taxes will lead to substantial changes in fuelling behaviour, even when the graduation curve is not steep. Depending on the graduation profile implemented, the spatial differentiation of fuel tax will give rise to substantial problems for owners of gas stations in areas with decreasing fuel sales. 9 refs.

Spatial Graduation of Fuel Taxes

International Nuclear Information System (INIS)

Rietveld, P.; Van Vuuren, D.; Bruinsma, F.

1999-06-01

Substantial differences exist among fuel taxes in various countries. These differences represent a form of fiscal competition that has undesirable side effects because it leads to cross-border fuelling and hence to extra kilometres driven. One possible way of solving the problem of low fuel taxes in neighbouring countries is to introduce a spatial differentiation of taxes: low near the border and higher further away. This paper contains an empirical analysis of the consequences of such a spatial graduation of fuel taxes for the Netherlands. We will analyse impacts on fuelling behaviour, vehicle kilometres driven, tax receipts, and sales by owners of gas stations. The appropriate slope of the graduation curve is also discussed. Our conclusion is that in a small country such as the Netherlands, a spatial graduation of fuel taxes will lead to substantial changes in fuelling behaviour, even when the graduation curve is not steep. Depending on the graduation profile implemented, the spatial differentiation of fuel tax will give rise to substantial problems for owners of gas stations in areas with decreasing fuel sales. 9 refs

Fuel and fission product behaviour in early phases of a severe accident. Part I: Experimental results of the PHEBUS FPT2 test

Energy Technology Data Exchange (ETDEWEB)

Barrachin, M., E-mail: marc.barrachin@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Gavillet, D. [Paul Scherrer Institute, Würenlingen and Villigen, CH-5232 Villigen PSI (Switzerland); Dubourg, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); De Bremaecker, A. [Institute for Nuclear Materials Sciences, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

2014-10-15

One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO{sub 2} fuel test section and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 and 900 mm) of the 1-m long test section are presented in this paper. Material interactions leading to local corium formation were identified: firstly between fuel and Zircaloy-4 cladding, notably at 823 mm, where the cladding melting temperature was reached, and secondly between fuel and stainless steel oxides. Regarding fission products, molybdenum left so-called metallic precipitates mainly composed of ruthenium. Xenon and caesium behave similarly whereas barium and molybdenum often seems to be associated in precipitates.

Estudo de propriedades dinâmico-mecânicas de um compósito aeronáutico de CFRP com fios de ligas com memória de forma embebidos

Directory of Open Access Journals (Sweden)

Zoroastro Torres Vilar

2016-01-01

Full Text Available Resumo Este trabalho trata-se de um estudo experimental do comportamento de propriedades termomecânicas de um compósito ativo, obtido através da introdução de diferentes frações volumétricas de fios de uma liga NiTi com efeito de memoria de forma em uma matriz polimérica reforçada com fibras de carbono (CFRP. Para tal estudo, inicialmente fez-se uma análise do módulo elástico de um pré preg de CFRP aeronáutico com diferentes posições de ângulo de fibras de carbono, a fim de determinar a matriz mais adequada para a introdução de fios de NiTi para obter compósito ativo. Verificou-se que a matriz com fibras alinhadas a 15º apresenta-se como uma boa opção para a obtenção de compósitos ativos, sendo utilizada para a fabricação de amostras de CFRP/NiTi. As amostras tiveram seu potencial de ativação avaliado através da análise da variação do módulo de elasticidade com o aumento da temperatura usando a técnica de Análise Dinâmico-Mecânica (DMA. Através dos resultados obtidos, verificou-se a capacidade de ativação dos compósitos CFRP/NiTi, os quais apresentam variação positiva de módulo de elasticidade ao serem aquecidos acima da temperatura de transformação dos fios NiTi.

7 CFR 1710.119 - Loan processing priorities.

Science.gov (United States)

2010-01-01

... and Basic Policies § 1710.119 Loan processing priorities. (a) Generally loans are processed in... in effect at the time the facilities were originally constructed; (3) To finance the capital needs of...

New application of Exp-function method for improved Boussinesq equation

Energy Technology Data Exchange (ETDEWEB)

Abdou, M.A. [Theoretical Research Group, Physics Department, Faculty of Science, Mansoura University, 35516 Mansoura (Egypt); Department of Physics, Faculty of Education for Girls, Science Departments, King Khalid University, Bisha (Saudi Arabia)], E-mail: m_abdou_eg@yahoo.com; Soliman, A.A. [Department of Mathematics, Faculty of Education (AL-Arish) Suez Canal University, AL-Arish 45111 (Egypt); Department of Mathematics, Teacher' s College (Bisha), King Khalid University, Bisha, PO Box 551 (Saudi Arabia)], E-mail: asoliman_99@yahoo.com; El-Basyony, S.T. [Theoretical Research Group, Physics Department, Faculty of Science, Mansoura University, 35516 Mansoura (Egypt)

2007-10-01

The Exp-function method is used to obtain generalized solitary solutions and periodic solutions for nonlinear evolution equations arising in mathematical physics with the aid of symbolic computation method, namely, the improved Boussinesq equation. The method is straightforward and concise, and its applications is promising for other nonlinear evolution equations in mathematical physics.

Fuels and auxiliary materials

International Nuclear Information System (INIS)

Svab, V.

A brief survey is given of the problems of fuels, fuel cans, absorption and moderator materials proceeding from the papers presented at the 1971 4th Geneva Conference on the Peaceful Uses of Nuclear Energy and the 1970 IAEA Conference in New York. Attention is focused on the behaviour of fuel and fuel can materials for thermal and fast reactors during irradiation, radiation stability of absorption materials and the effects of radiation on concrete and on moderator materials. (Z.M.)

Fuel Services

International Nuclear Information System (INIS)

Silberstein, A.

1982-09-01

FRAGEMA has developed most types of inspection equipments to work on irradiated fuel assemblies and on single fuel rods during reactor outages with an efficiency compatible with the utilities operating priorities. In order to illustrate this statement, two specific examples of inspection equipments are shortly described: the on-site removable fuel rod assembly examination stand, and the fuel assembly multiple examination device. FRAGEMA has developed techniques for the identifiction of the leaking fuel rods in the fuel assembly and the tooling necessary to perform the replacement of the faulted element. These examples of methods, techniques and equipments described and the experience accumulated through their use allow FRAGEMA to qualify for offering the supply of the corresponding software, hardware or both whenever an accurate understanding of the fuel behaviour is necessary and whenever direct intervention on the assembly and associated components is necessary due to safety, operating or economical reasons

Agglomeration and Deposition Behaviour of Solid Recovered Fuel

DEFF Research Database (Denmark)

Pedersen, Morten Nedergaard; Jensen, Peter Arendt; Nielsen, Mads

2015-01-01

Waste derived fuels such as Solid Recovered Fuel (SRF) are increasingly being used in the cement industry as a means to reduce cost [1]. SRF is produced by separating the combustible fraction from industrial or municipal solid waste (MSW). The recovered fraction has a higher content of combustibl...

Mechanical behaviour and failure of fuel cladding zirconium alloys in nuclear power plants under accidental RIA-type situation

International Nuclear Information System (INIS)

Doan, D.T.

2009-01-01

In French Nuclear Pressurized Water Reactors (PWRs), most of structural parts of the fuel assembly consist of zirconium alloy tubes and plates. Optimizing the management of fuel in nuclear power plants led to the increase in the duration of fuel cycles and power. The use of high fuel burnups requires drastic changes in the rules for reactor design in the nuclear safety. The evaluation of nuclear reactors in accident situations is based on reference accident scenarios. One of these hypothetical accidents, examined in this study, is the 'Reactivity Initiated Accident'. In order to assess the structural integrity of these parts it is necessary to characterize both the plastic flow and fracture behaviour of the materials at various stages of the life cycle, (i.e. at increasing levels of hydriding, irradiation, oxidation or thermal mechanical loading). The purpose of this work is to provide experimental data and to develop a model of the thermo-mechanical behaviour and to propose a design analysis method in the case of non-irradiated clads, in RIA-type situations. Mechanical tests were conducted on Cold-Worked-Stress-Relieved and on Recrystallized Zircaloy-4 sheets using various kinds of samples including smooth and notched tensile specimens and small punch tests. Temperature was set to 25, 250 and 600 C with hydrogen contents between 0 and 1000 ppm. The model is based on a simplified description of a Zircaloy polycrystal in which scalar isotropic ductile damage including void nucleation and growth is added. The model is also physically based to easily transfer parameters determined for one material state to another (e.g. transfer between sheet and tube or between different levels of irradiation). The model was implemented in the Finite Element software Zebulon using either an explicit or an implicit time integration scheme. Uniaxial tension tests were used to tune the model parameters for both materials, considering various values of temperature and hydrogen levels

Computer simulation of fuel element performance

Energy Technology Data Exchange (ETDEWEB)

Sukhanov, G I

1979-01-01

The review presents reports made at the Conference on the Bahaviour and Production of Fuel for Water Reactors on March 13-17, 1979. Discussed at the Conference are the most developed and tested calculation models specially evolved to predict the behaviour of fuel elements of water reactors. The following five main aspects of the problem are discussed: general conceptions and programs; mechanical mock-ups and their applications; gas release, gap conductivity and fuel thermal conductivity; analysis of nonstationary processes; models of specific phenomena. The review briefly describes the physical principles of the following models and programs: the RESTR, providing calculation of the radii of zones of columnar and equiaxial grains as well as the radius of the internal cavity of the fuel core; programs for calculation of fuel-can interaction, based on the finite elements method; a model predicting the behaviour of the CANDU-PHW fuel elements in transient conditions. General results are presented of investigations of heat transfer through a can-fuel gap and thermal conductivity of UO/sub 2/ with regard for cracking and gas release of the fuel. Many programs already suit the accepted standards and are intensively tested at present.

Highlights from the IAEA coordinated research programme on fuel performance and fission product data

International Nuclear Information System (INIS)

Nabielek, H.; Schenk, W.; Verfondern, K.

1996-01-01

Seven countries are cooperating with the objectives (i) to document the status of the experimental data base and of the predictive methods for Gas-Cooled Reactor fuel performance and fission product behaviour; (ii) to verify and validate methods in fuel performance and fission product retention prediction. These countries are China, France, Germany, Japan, Russia, USA and the UK. Duration of the programme is 1993-96. The technology areas addressed in this IAEA Coordinated Research Programme are: Fuel design and manufacture, Normal operation fuel performance and fission product behaviour, Accident condition fuel performance and fission product behaviour, -core heatup, -fast transients, -oxidising conditions (water and air ingress), Plateout, re-entrainment of plateout, fission product behaviour in the reactor building, and Performance of advanced fuels. Work performed so far has generated a 300-page draft document with important information for normal operations (Germany, Japan, China, Russia) and accident conditions (USA, Japan, Germany, Russia) and, additionally, a special chapter on advanced fuels (Japan). (author)

Integral approach to innovative fuel and material investigations in the Halden reactor

International Nuclear Information System (INIS)

Volkov, B.

2009-01-01

Integral approach used for fuel and material investigations in the Halden reactor can be used in support of qualification and certification of fuel to be introduced in commercial NPPs. This approach has been partly used for WWER fuel investigation in the Halden Reactor in a series of irradiation tests. In-pile fuel performance tests with reliable measurements provided by Halden instrumentation under different conditions can be used for validation of the WWER fuel behaviour models and verification of fuel performance codes. These models and codes can be used for qualification of innovative fuel behaviour under extended conditions

Recent results from CEC cost sharing research programme on LWR fuel behaviour under accident conditions

International Nuclear Information System (INIS)

Fairbairn, S.A.

1983-01-01

The present structure and intentions of the CEC sponsored cost sharing programme for LWR safety research are outlined. Detailed results are reported for two projects from this programme. The first project concerns experimental data on the thermohydraulic effects of flow diversion around ballooned fuel rods. Data are presented on single and two phase heat transfer in an electrically heated rod bundle. Detailed photographic data on droplet behaviour are also given. The second project is an investigation of the effects of zircaloy oxidation on rewetting during reflood. It is shown that as oxide thickness increases from 1μm to 76μm that rewet rates can increase by up to 40%. A systematic effect of oxidation on rewet temperatures is also noted. (author)

Exp-function method for solving Fisher's equation

Energy Technology Data Exchange (ETDEWEB)

Zhou, X-W [Department of Mathematics, Kunming Teacher' s College, Kunming, Yunnan 650031 (China)], E-mail: km_xwzhou@163.com

2008-02-15

There are many methods to solve Fisher's equation, but each method can only lead to a special solution. In this paper, a new method, namely the exp-function method, is employed to solve the Fisher's equation. The obtained result includes all solutions in open literature as special cases, and the generalized solution with some free parameters might imply some fascinating meanings hidden in the Fisher's equation.

The sphere-PAC fuel code 'SPHERE-3'

International Nuclear Information System (INIS)

Wallin, H.

2000-01-01

Sphere-PAC fuel is an advanced nuclear fuel, in which the cladding tube is filled with small fuel spheres instead of the more usual fuel pellets. At PSI, the irradiation behaviour of sphere-PAC fuel is calculated using the computer code SPHERE-3. The paper describes the present status of the SPHERE-3 code, and some results of the qualification process against experimental data. (author)

The sphere-pac fuel code 'SPHERE-3'

International Nuclear Information System (INIS)

Wallin, H.; Nordstroem, L.A.; Hellwig, C.

2001-01-01

Sphere-pac fuel is an advanced nuclear fuel, in which the cladding tube is filled with small fuel spheres instead of the more usual fuel pellets. At PSI, the irradiation behaviour of sphere-pac fuel is calculated using the computer code SPHERE-3. The paper describes the present status of the SPHERE-3 code, and some results of the qualification process against experimental data. (author)

27 Etude comparative de la compacité expérimentale et ...

African Journals Online (AJOL)

Moussa

objectif de notre recherche ... Par définition la compacité virtuelle est inaccessible à l'expérience et à partir de la formule (1) on remarque .... l'optimisation granulaire par des modèles théoriques constitue un bon outil d'étude des empilements.

Fuel behavior aspects of the interpretation of the SCARABEE fast reactor safety experiments

International Nuclear Information System (INIS)

Schmitz, F.; Matthews, J.R.

1980-01-01

The main conclusions of the fuel behaviour analysis of 16 single pin and 8 seven-pin bundle experiments of the SCARABEE programme are presented as result of the tripartite interpretation agreement between CEA, UKAEA and KfK. From all partners it is stated that existing fuel behaviour codes calculate with adequate precison the temperature, structure and geometry under steady state conditions. The state of the SCARABEE fuel at the beginning of the transient phase (which determines the subsequent transient behaviour) can be considered to be well known. For the transient phase of the experiments a fairly good description is given for overpower conditions with single phase coolant flow. In and beyond two phase flow region the understanding of the fuel pin behaviour remained difficult. Failure prediction either by mechanical rupture or by clad melting is strongly linked to the thermohydraulic behaviour and dependent on failure criteria. (orig.)

Computer code SICHTA-85/MOD 1 for thermohydraulic and mechanical modelling of WWER fuel channel behaviour during LOCA and comparison with original version of the SICHTA code

International Nuclear Information System (INIS)

Bujan, A.; Adamik, V.; Misak, J.

1986-01-01

A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)

Espaces virtuels et pré-expérience de l’espace géographique

Directory of Open Access Journals (Sweden)

Jérémie Valentin

2013-05-01

Full Text Available Les représentations virtuelles de l’espace se sont largement démocratisées lors des dernières années. La libéralisation du marché des images satellites et l’augmentation des capacités de transfert de données sur les réseaux filaires et non filaires participent largement à la mise en place de mondes miroirs désormais connectés au cyberespace.L’homme a de plus en plus recours aux espaces virtuels pour appréhender l’espace. Que ce soit pour l’élaboration d’un itinéraire ou la découverte d’un lieu, la carte papier et le guide touristique s’effacent au profit du monde miroir et du téléphone portable.Dans ce contexte de transfert d’usage, les représentations virtuelles de l’espace semblent, au même titre que l’usage des GPS, offrir une perception biaisée de l’espace. Car malgré leurs qualités intrinsèques communes, en particulier pour la navigation dans un espace inconnu, leurs usages modifieraient notre relation à l’espace et pourraient dans certains cas altérer son apprentissage. L’objectif de cette communication pluridisciplinaire est de mettre au jour la popularité des mondes miroirs tout en interrogeant leurs usages en condition de pré-expérience de l’espace, car les dispositifs actuels permettent d’expérimenter virtuellement l’espace avant d’en faire physiquement l’expérience. Nous verrons à travers une expérience comparative, que dans le cadre d’un parcours complexe, les sujets ayant eu une pré-expérience virtuelle de l’espace s’avèrent moins efficaces que les sujets qui ont consulté un plan papier 2D et moins efficaces que les sujets ayant effectué une visite guidée. L’analyse des traces GPS des sujets nous permettra de tirer des conclusions sur les typologies d’échecs et d’erreurs.

Development of advanced spent fuel management process

International Nuclear Information System (INIS)

Ro, Seung Gy; Shin, Y. J.; Do, J. B.; You, G. S.; Seo, J. S.; Lee, H. G.

1998-03-01

This study is to develop an advanced spent fuel management process for countries which have not yet decided a back-end nuclear fuel cycle policy. The aims of this process development based on the pyroreduction technology of PWR spent fuels with molten lithium, are to reduce the storage volume by a quarter and to reduce the storage cooling load in half by the preferential removal of highly radioactive decay-heat elements such as Cs-137 and Sr-90 only. From the experimental results which confirm the feasibility of metallization technology, it is concluded that there are no problems in aspects of reaction kinetics and equilibrium. However, the operating performance test of each equipment on an engineering scale still remain and will be conducted in 1999. (author). 21 refs., 45 tabs., 119 figs

On an integral representation of the function Tr(exp(A-lambdaB))

Energy Technology Data Exchange (ETDEWEB)

Mehta, M L [CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France). Service de Physique Theorique; Kumar, K

1976-02-01

The conjecture that Tr(exp(A-lambdaB)) can be written as a Laplace transform with a positive measure is proved for a certain class of matrices A and B. A few remarks are made about the undecided cases.

New Traveling Wave Solutions of the Higher Dimensional Nonlinear Partial Differential Equation by the Exp-Function Method

Directory of Open Access Journals (Sweden)

Hasibun Naher

2012-01-01

Full Text Available We construct new analytical solutions of the (3+1-dimensional modified KdV-Zakharov-Kuznetsev equation by the Exp-function method. Plentiful exact traveling wave solutions with arbitrary parameters are effectively obtained by the method. The obtained results show that the Exp-function method is effective and straightforward mathematical tool for searching analytical solutions with arbitrary parameters of higher-dimensional nonlinear partial differential equation.

Soluções de redução de consumo energético para redes de sensores sem fio (RSSFs aplicadas à ambientes florestais

Directory of Open Access Journals (Sweden)

Paulo César Sedrez Moncks

2016-07-01

Full Text Available Redes de sensores sem fio (RSSFs têm sido utilizadas para aplicações de monitoramento nos mais diversos cenários, como controle industrial, gerenciamento de tráfego, segurança pública, automação residencial, saúde e também monitoramento ambiental. Estas redes são compostas de sensores com restrição de recursos onde a eficiência energética é parte essencial para sua real aplicabilidade. É apresentada neste artigo a sistematização feita das características de três técnicas para redução do consumo energético das redes de sensores sem fio, e como principal objetivo desta sistematização espera-se fornecer elementos para avaliar a aplicabilidade de RSSFs nas tarefas de predição de risco e monitoramento de incêndios florestais. Foi possível concluir que o uso de RSSFs aplicadas a ambientes florestais ainda é uma frente de pesquisa em aberto, sobretudo no que se refere à durabilidade da vida útil da rede. Ainda que as técnicas de redução de consumo energético propostas nos trabalhos avaliados apresentem ganhos, é necessário aprofundar as pesquisas para alcançar um tempo maior de duração das baterias e com isso tornar viável a instalação de nodos em florestas.

Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

International Nuclear Information System (INIS)

Mella, R.; Wenman, M.R.

2013-01-01

Thermo-mechanical contributions to pellet–clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS’s well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used

48 CFR 18.119 - Use of patented technology under the North American Free Trade Agreement.

Science.gov (United States)

2010-10-01

... under the North American Free Trade Agreement. 18.119 Section 18.119 Federal Acquisition Regulations... Available Acquisition Flexibilities 18.119 Use of patented technology under the North American Free Trade Agreement. Requirement to obtain authorization prior to use of patented technology may be waived in...

Fission products and nuclear fuel behaviour under severe accident conditions part 3: Speciation of fission products in the VERDON-1 sample

Science.gov (United States)

Le Gall, C.; Geiger, E.; Gallais-During, A.; Pontillon, Y.; Lamontagne, J.; Hanus, E.; Ducros, G.

2017-11-01

Qualitative and quantitative analyses on the VERDON-1 sample made it possible to obtain valuable information on fission product behaviour in the fuel during the test. A promising methodology based on the quantitative results of post-test characterisations has been implemented to assess the release fraction of non γ-emitter fission products. The order of magnitude of the estimated release fractions for each fission product was consistent with their class of volatility.

The flexfuel tractor. Invesigations on the combustion behaviour of vegetable oil fuels and on the discernability of fossil and biogenic fuels; Der Flexfuel Traktor. Untersuchungen zum Verbrennungsverhalten von Pflanzenoelkraftstoffen und zur Unterscheidbarkeit fossiler und biogener Kraftstoffe

Energy Technology Data Exchange (ETDEWEB)

Dieringer, Stefanie

2012-07-01

Increasing energy prices, especially for fossil fuels, as well as the necessity to reduce CO{sub 2} emissions are emphasizing the advantages of self-produced vegetable oil fuels in agriculture. Monetary advantages are depending on basic conditions like farm size or tax legislation, which can be changing locally as well as temporarily. Due to the differing properties of diesel and vegetable oil fuel, engines have to be adapted to each fuel to fulfil performance requirements as well as emission limits and reliability. Knowing that there are advantages of vegetable oil compared to diesel fuel, though not always and everywhere present, it becomes obvious that the well known flexible fuel concept of passenger cars should be adapted for diesel engines of agricultural machines. So called flexfuel engines imply the detection of the fuel type and an automated adjustment of the engine control parameters without any manual action of an operator. Therefore, the first step consists of the evaluation of the combustion properties of rapeseed, sunflower, jatropha and false flax oil compared to diesel fuel. The tested vegetable oils showed very similar behaviour in the tested common rail diesel engine. Especially the limited emissions were met with the same engine control software with all vegetable oils. In consequence it is possible to realize a flexfuel engine using the two engine control maps available at the moment, one for diesel and the other one for vegetable oil fuels. For further investigations one oil type, namely rapeseed oil was selected to test the combustion behaviour of fuel blends made of diesel and vegetable oil. The goal was to determine the blend ratio of vegetable oil and diesel fuel at which the engine control software has to be changed from the diesel to the vegetable oil map automatically. If the fuel consists of 40% or more vegetable oil, the vegetable oil engine control map has to be selected in order to fulfil legal emission limits. Finally the

Income inequality and happiness in 119 nations

NARCIS (Netherlands)

M.C. Berg (Maarten); R. Veenhoven (Ruut)

2010-01-01

textabstractINCOME INEQUALITY AND HAPPINESS IN 119 NATIONS All modern nations reduce income differences to some extent, and as a result there is an ongoing discussion about what degree of income inequality is acceptable. In this discussion libertarians oppose egalitarians and a principled consensus

Materials in the environment of the fuel in dry storage

Energy Technology Data Exchange (ETDEWEB)

Issard, H [TN International (Cogema Logistics) (France)

2012-07-01

Spent nuclear fuel has been stored safely in pools or dry systems in over 30 countries. The majority of IAEA Member States have not yet decided upon the ultimate disposition of their spent nuclear fuel: reprocessing or direct disposal. Interim storage is the current solution for these countries. For developing the technological knowledge data base, a continuation of the IAEA's spent fuel storage performance assessment was achieved. The objectives are: Investigate the dry storage systems and gather basic fuel behaviour assessment; Gather data on dry storage environment and cask materials; Evaluate long term behaviour of cask materials.

Comment on : "A novel approach for solving the Fisher equation using Exp-function method" [Phys. Lett. A 372 3836

NARCIS (Netherlands)

Kudryashov, Nikolai A.

2009-01-01

Using Exp-function method Ozis and Koroglu [T. Ozis, C. Koroglu, Phys. Lett. A 372 (2008) 3836] have found exact "solutions" of the Fisher equation. In this comment we demonstrate that all these solutions do not satisfy the Fisher equation. The efficiency of application of Exp-function method to

Desenvolvimento de uma rede de sensores sem fio aplicada no monitoramento da variabilidade térmica em casas de vegetação

OpenAIRE

Barbosa, Rogério Zanarde [UNESP

2015-01-01

Este é um trabalho de tecnologia computacional aplicada na área agrícola, cujo objetivo principal do trabalho é desenvolver uma rede de sensores sem fio, que envolve aspectos de software e hardware, para o monitoramento térmico no interior de uma casa de vegetação. Além da rede propriamente dita, o trabalho também inclui a sua aplicação no levantamento quantitativo da variabilidade térmica na casa de vegetação o que pode ser aplicado em diversas atividades agrícolas a serem desenvolvidas no i...

La colonisation de l’expérience politique

Directory of Open Access Journals (Sweden)

Marin Ledun

2005-10-01

Full Text Available En prenant pour objet les projets expérimentaux publics ou industriels en matière de TIC, dans ou à la marge de la sphère politique française, dès la fin des années soixante, cette recherche analyse, du point de vue des Sciences de l’Information et de la Communication, les enjeux liés à l’extension du dispositif des TIC dans les activités politiques et à la redéfinition de la place du sujet politique dans les procédures démocratiques.

A CAREM type fuel element dynamic analysis

International Nuclear Information System (INIS)

Magoia, J.E.

1990-01-01

A first analysis on the dynamic behaviour of a fuel element designed for the CAREM nuclear reactor (Central Argentina de Elementos Modulares) was performed. The model used to represent this dynamic behaviour was satisfactorily evaluated. Using primary estimations for some of its numerical parameters, a first approximation to its natural vibrational modes was obtained. Results obtained from fuel elements frequently used in nuclear power plants of the PWR (Pressurized Water Reactors) type, are compared with values resulting from similar analysis. (Author) [es

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

International Nuclear Information System (INIS)

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

Modelling of WWER-1000 fuel: state and prospects

International Nuclear Information System (INIS)

Medvedev, A.; Bibilashvili, Yu.; Bogatyr, S.; Khvostov, G.

1994-01-01

The role of START-3 code in studying and computerized modelling of post-irradiation behaviour of standard fuel rods in real operation conditions of WWER-1000 reactors is described. The models used in the code are based on experimental study of material properties, processes and post irradiation research on standard and experimental fuel pins. The code capability is verified by comparison with data from experiments on WWER test rods performed in MR reactor, the Russia-Finland tests SOFIT and the international program FUMEX. The comparison performed and the results thus obtained demonstrate the satisfactory ability of START-3 code to simulate fuel rod behaviour in normal operation condition. The calculations confirm the experimentally observed evidence of an essential margin on serviceability of WWER-1000 fuel pin with three year operation cycle permitting an increase in design fuel burnup. 2 tabs., 18 figs

Modelling of WWER-1000 fuel: state and prospects

Energy Technology Data Exchange (ETDEWEB)

Medvedev, A; Bibilashvili, Yu; Bogatyr, S; Khvostov, G [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

1994-12-31

The role of START-3 code in studying and computerized modelling of post-irradiation behaviour of standard fuel rods in real operation conditions of WWER-1000 reactors is described. The models used in the code are based on experimental study of material properties, processes and post irradiation research on standard and experimental fuel pins. The code capability is verified by comparison with data from experiments on WWER test rods performed in MR reactor, the Russia-Finland tests SOFIT and the international program FUMEX. The comparison performed and the results thus obtained demonstrate the satisfactory ability of START-3 code to simulate fuel rod behaviour in normal operation condition. The calculations confirm the experimentally observed evidence of an essential margin on serviceability of WWER-1000 fuel pin with three year operation cycle permitting an increase in design fuel burnup. 2 tabs., 18 figs.

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Modeling steady state and transient fission gas behaviour with the Karlsruhe code LAKU

International Nuclear Information System (INIS)

Vaeth, L.

1984-08-01

The programme LAKU models the behaviour of gaseous fission products in reactor fuel under steady state and transient conditions, including molten fuel. A presentation of the full model is given, starting with gas behaviour in the grains and on grain faces and including the treatment of release from porosity. The results of some recent calculations are presented. (orig.) [de

Operational monitoring of the release behaviour of the AVR core

International Nuclear Information System (INIS)

Wawrzik, U.; Ivens, G.

1985-01-01

The AVR reactor has been used for the mass testing of spherical HTR fuel elements for more than 17 years. To date 14 fuel element types have been used, some of which differ considerably with reference to the heavy metal content, fuel coating and chemical fuel composition. The aim of the measurements at the reactor and of a comprehensive post-irradiation examination programme for fuel elements is to check and evaluate the behaviour of these fuel elements under real reactor conditions. This paper considers only those measurements which are of interest for reactor operation. The integral release behaviour of the fuel elements is continuously monitored by measuring the noble fission gas activity in the primary system. This value is directly determined by the heavy metal contamination. If any significant particle defects occur during operation, these are immediately indicated by a considerable increase in the noble gas activity. The integral release of the solid fission products is monitored by means of filter samples both in the hot and in the cold gas area; this examination, however, is performed intermittently and with a time delay. As these integral measurements only allow one to draw limited conclusions about the behaviour of single fuel element charges or types, they are supplemented by the systematic extraction of fuel elements. These elements are then subjected to standardized annealing tests (KFA) to determine the individual noble gas release, and to examinations of the fuel-free shell to establish the distribution of the solid fission products in it (AVR). The latter method, in particular, has proved to be practicable, as particle defects are detected at an early stadium. During operation to date, only one fuel element charge exhibited incipient particle defects shortly before reaching its final burnup. It was possible to limit the activity release by altering the charging strategy, which resulted in lower fuel element temperatures, and by systematically

An analysis of confidence limit calculations used in AAPM Task Group No. 119

International Nuclear Information System (INIS)

Knill, Cory; Snyder, Michael

2011-01-01

Purpose: The report issued by AAPM Task Group No. 119 outlined a procedure for evaluating the effectiveness of IMRT commissioning. The procedure involves measuring gamma pass-rate indices for IMRT plans of standard phantoms and determining if the results fall within a confidence limit set by assuming normally distributed data. As stated in the TG report, the assumption of normally distributed gamma pass rates is a convenient approximation for commissioning purposes, but may not accurately describe the data. Here the authors attempt to better describe gamma pass-rate data by fitting it to different distributions. The authors then calculate updated confidence limits using those distributions and compare them to those derived using TG No. 119 method. Methods: Gamma pass-rate data from 111 head and neck patients are fitted using the TG No. 119 normal distribution, a truncated normal distribution, and a Weibull distribution. Confidence limits to 95% are calculated for each and compared. A more general analysis of the expected differences between the TG No. 119 method of determining confidence limits and a more time-consuming curve fitting method is performed. Results: The TG No. 119 standard normal distribution does not fit the measured data. However, due to the small range of measured data points, the inaccuracy of the fit has only a small effect on the final value of the confidence limits. The confidence limits for the 111 patient plans are within 0.1% of each other for all distributions. The maximum expected difference in confidence limits, calculated using TG No. 119's approximation and a truncated distribution, is 1.2%. Conclusions: A three-parameter Weibull probability distribution more accurately fits the clinical gamma index pass-rate data than the normal distribution adopted by TG No. 119. However, the sensitivity of the confidence limit on distribution fit is low outside of exceptional circumstances.

The relationship between gambling expenditure, socio-demographics, health-related correlates and gambling behaviour-a cross-sectional population-based survey in Finland.

Science.gov (United States)

Castrén, Sari; Kontto, Jukka; Alho, Hannu; Salonen, Anne H

2018-01-01

To investigate gambling expenditure and its relationship with socio-demographics, health-related correlates and past-year gambling behaviour. Cross-sectional population survey. Population-based survey in Finland. Finnish people aged 15-74 years drawn randomly from the Population Information System. The participants in this study were past-year gamblers with gambling expenditure data available (n = 3251, 1418 women and 1833 men). Expenditure shares, means of weekly gambling expenditure (WGE, €) and monthly gambling expenditure as a percentage of net income (MGE/NI, %) were calculated. The correlates used were perceived health, smoking, mental health [Mental Health Inventory (MHI)-5], alcohol use [Alcohol Use Disorders Identification Test (AUDIT)-C], game types, gambling frequency, gambling mode and gambling severity [South Oaks Gambling Screen (SOGS)]. Gender (men versus women) was found to be associated significantly with gambling expenditure, with exp(β) = 1.40, 95% confidence interval (CI) = 1.29, 1.52 and P gambling behaviour correlates were associated significantly with WGE and MGE/NI: gambling frequency (several times a week versus once a month/less than monthly, exp(β) = 30.75, 95% CI = 26.89, 35.17 and P gambling severity (probable pathological gamblers versus non-problem gamblers, exp(β) = 2.83, 95% CI = 2.12, 3.77 and P gambling (on-line and land-based versus land-based only, exp(β) = 1.35, 95% CI = 1.24, 1.47 and P gambling expenditure and monthly gambling expenditure related to net income. People in Finland with lower incomes contribute proportionally more of their income to gambling compared with middle- and high-income groups. © 2017 The Authors. Addiction published by John Wiley & Sons Ltd on behalf of Society for the Study of Addiction.

DESENVOLVIMENTO E AVALIAÇÃO DE UM ANEMÔMETRO DE FIO QUENTE OPERANDO À TEMPERATURA CONSTANTE

Directory of Open Access Journals (Sweden)

Carlos Augusto de P. Sampaio

1998-08-01

Full Text Available RESUMO Esta pesquisa teve como objetivo desenvolver um anemômetro de fio quente para operar à temperatura constante usando-se, como elemento sensível, filamento de tungstênio de lâmpada incandescente, como braço da ponte de Wheatstone. Um circuito eletrônico foi construído e testado, permitindo controlar a tensão no sensor e restaurar o equilíbrio da ponte de Wheatstone em processos de troca de calor pelo sensor. Os resultados mostraram que a corrente elétrica e a temperatura de operação do sensor, mediante compromisso entre vida útil e sensibilidade, foram de 160mA e 140°C, respectivamente; a calibração com o sensor no circuito ponte de Wheatstone mostrou sensibilidade para medir a velocidade do ar no intervalo de 0,00 e 5,00m.s-1 e que o circuito eletrônico apresentou saturação para velocidades do ar inferiores a 0,50m.s-1.

29 CFR 570.119 - Fourteen-year minimum.

Science.gov (United States)

2010-07-01

... Regulations Relating to Labor (Continued) WAGE AND HOUR DIVISION, DEPARTMENT OF LABOR REGULATIONS CHILD LABOR REGULATIONS, ORDERS AND STATEMENTS OF INTERPRETATION General Statements of Interpretation of the Child Labor Provisions of the Fair Labor Standards Act of 1938, as Amended Oppressive Child Labor § 570.119 Fourteen-year...

Fission gas release behaviour of a 103 GWd/t{sub HM} fuel disc during a 1200 °C annealing test

Energy Technology Data Exchange (ETDEWEB)

Noirot, J., E-mail: jean.noirot@cea.fr [CEA, DEN, DEC, Cadarache, F-13108 St. Paul Lez Durance (France); Pontillon, Y. [CEA, DEN, DEC, Cadarache, F-13108 St. Paul Lez Durance (France); Yagnik, S. [EPRI, P.O. Box 10412, Palo Alto, CA 94303-0813 (United States); Turnbull, J.A. [Independent Consultant (United Kingdom); Tverberg, T. [IFE, P.O. Box 173, NO-1751 Halden (Norway)

2014-03-15

Within the Nuclear Fuel Industry Research (NFIR) program, several fuel variants, in the form of thin circular discs, were irradiated in the Halden Boiling Water Reactor (HBWR) to a range of burn-ups ∼100 GWd/t{sub HM}. The design of the assembly was similar to that used in other HBWR programs: the assembly contained several rods with fuel discs sandwiched between Mo discs, which limited temperature gradients within the fuel discs. One such rod contained standard grain UO{sub 2} discs (3D grain size = 18 μm) reaching a burn-up of 103 GWd/t{sub HM}. After the irradiation, the gas release upon rod puncturing was measured to be 2.9%. Detailed characterizations of one of these irradiated UO{sub 2} discs, using electron probe microanalysis (EPMA), scanning electron microscopy (SEM) and secondary ion mass spectrometry (SIMS), were performed in a CEA Cadarache hot laboratory. Examination revealed the high burn-up structure (HBS) formation throughout the whole of the disc, also the fission gas distribution within this HBS, with a very high proportion of the gas in the HBS bubbles. A sibling disc was submitted to a temperature transient up to 1200 °C in the out-of-pile (OOP) annealing test device “Merarg” at a relatively low temperature ramp rate (0.2 °C/s). In addition to the total gas release during this annealing test, the release peaks throughout the temperature range were monitored. The fuel was then characterized with the same microanalysis techniques as before the annealing test to investigate the effects of this test on the microstructure of the fuel and on the fission gases. It provided valuable insights into fission gas localization and the release behaviour in UO{sub 2} fuel with high burn-up structure (HBS)

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

International Nuclear Information System (INIS)

1973-01-01

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

LWR-core behaviour project

International Nuclear Information System (INIS)

Paratte, J.M.

1982-07-01

The LWR-Core behaviour project concerns the mathematical simulation of a light water reactor in normal operation (emergency situations excluded). Computational tools are assembled, i.e. programs and libraries of data. These computational tools can likewise be used in nuclear power applications, industry and control applications. The project is divided into three parts: the development and application of calculation methods for quantisation determination of LWR physics; investigation of the behaviour of nuclear fuels under radiation with special attention to higher burnup; simulation of the operating transients of nuclear power stations. (A.N.K.)

Studies on the safety and transmutation behaviour of innovative fuels for light water reactors; Untersuchungen zum Sicherheits- und Transmutationsverhalten innovativer Brennstoffe fuer Leichtwasserreaktoren

Energy Technology Data Exchange (ETDEWEB)

Schitthelm, Oliver

2012-07-01

Nuclear power plants contribute a substantial part to the energy demand in industry. Today the most common fuel cycle uses enriched uranium which produces plutonium due to its {sup 238}U content. With respect to the long-term waste disposal Plutonium is an issue due to its heat production and radiotoxicity. This thesis consists of three main parts. In the first part the development and validation of a new code package MCBURN for spatial high resolution burnup simulations is presented. In the second part several innovative uranium-free and plutonium-burning fuels are evaluated on assembly level. Candidates for these fuels are a thorium/plutonium fuel and an inert matrix fuel consisting of plutonium dispersed in an enriched molybdenum matrix. The performance of these fuels is evaluated against existing MOX and enriched uranium fuels considering the safety and transmutation behaviour. The evaluation contains the boron efficiency, the void coefficient, the doppler coefficient and the net balances of every radionuclide. In the third part these innovative fuels are introduced into a German KONVOI reactor core. Considering todays approved usage of MOX fuels a partial loading of one third of innovative fuels and two third of classical uranium fuels was analysed. The efficiency of the plutonium depletion is determined by the ratio of the production of higher isotopes compared to the plutonium depletion. Todays MOX-fuels transmutate about 25% to 30% into higher actinides as Americium or Curium. In uranium-free fuels this ratio is about 10% due to the lack of additional plutonium production. The analyses of the reactor core have shown that one third of MOX fuel is not capable of a net reduction of plutonium. On the other hand a partial loading with thorium/plutonium fuel incinerates about half the amount of plutonium produced by an uranium only core. If IMF is used the ratio increases to about 75%. Considering the safety behavior all fuels have shown comparable results.

Review of the IAEA Nuclear Fuel Cycle Materials Section activities related to WWER fuel

International Nuclear Information System (INIS)

Killeen, J.

2003-01-01

The IAEA Nuclear Fuel Cycle Programme, designated as Programme B, has the main objective of supporting Member States in policy making, strategic planning, developing technology and addressing issues with respect to safe, reliable, economically efficient, proliferation resistant and environmentally sound nuclear fuel cycle. This paper is concentrated on describing the work within Sub-programme B.2 'Fuel Performance and Technology'. Two Technical Working Groups assist in the preparation of the IAEA programme in the nuclear fuel cycle area - Technical Working Group on Water Reactor Fuel Performance and Technology and Technical Working Group on Nuclear Fuel Cycle Options. The activities of the Unit within the Nuclear Fuel Cycle and Materials Section working on Fuel Performance and Technology are given, based on the sub-programme structure of the Agency programme and budget for 2002-2003. Within the framework of Co-ordinated Research Projects a study of the delayed hydride cracking (DHC) of the zirconium alloys used in pressurised heavy water reactors (PHWR) involving 10 countries has been completed. It achieved very effective transfer of know-how at the laboratory level in three technologically important areas: 1) Controlled hydriding of samples to predetermined levels; 2) Accurate measurement of hydrogen concentrations at the relatively low levels found in pressure tubes and RBMK channel tubes; and 3) In the determination of DHC rates under various conditions of temperature and stress. A new project has been started on the 'Improvement of Models used for Fuel Behaviour Simulation' (FUMEX II) to assist Member States in improving the predictive capabilities of computer codes used in modelling fuel behaviour for extended burnup. The IAEA also collaborates with organisations in the Member States to support activities and meetings on nuclear fuel cycle related topics

Analysis of WWER-440 fuel performance under normal operating conditions

Energy Technology Data Exchange (ETDEWEB)

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.

Reparação do ligamento cruzado cranial de cães por tendão homólogo conservado em glicerina e associado a fio de náilon

Directory of Open Access Journals (Sweden)

Oliveira Simone Tostes de

2003-01-01

Full Text Available A ruptura do ligamento cruzado cranial (LCC é uma das patologias de maior prevalência no joelho de cães e sua correção é um problema atual. Após a ruptura experimental do LCC, em cães, implantou-se o tendão do músculo extensor longo dos dedos, homólogo, conservado em glicerina a 98%, associado ao fio de mononáilon, utilizando-se técnica intra-articular. Foram utilizados 18 cães com peso médio de 16,5kg, adultos, separados em três grupos (A, B e C e avaliados por 45, 80 e 120 dias de pós-operatório, respectivamente. Foi realizada avaliação clínica, radiográfica, macroscópica e histológica. A técnica demonstrou-se eficiente para o retorno funcional do membro promovendo estabilidade articular, apesar de o implante biológico ter sido reabsorvido na maioria dos animais, e do fio de náilon ter perdido sua tensão inicial.

Sources de photons uniques et expérience à choix retardé de Wheeler : la dualité onde corpuscule à l'épreuve de l'expérience

Science.gov (United States)

Jacques, V.

ésentations classiques de la réalité physique incompatibles. Le travail décrit dans cet ouvrage s'inscrit dans l'étude de la dualité onde-corpuscule pour un photon unique à l'aide d'une source de photons uniques fondée sur l'excitation impulsionnelle d'un centre coloré NV individuel dans un nanocristal de diamant. Nous présentons dans un premier temps une expérience d'interférence à un photon très proche dans sa conception de celle des trous d'Young. Cette expérience nous permettra de discuter de la complémentarité entre interférence et connaissance du chemin suivi par la particule dans l'interféromètre. Pour aller un peu plus au cÅ`ur des problèmes conceptuels soulevés par la dualité onde-corpuscule, nous décrivons par la suite la réalisation expérimentale de l'expérience de pensée dite de “choix retardé” proposée par Wheeler au début des années soixante-dix. Dans cette expérience, la décision de fermer ou non l'interféromètre, et donc d'observer soit les interférences (associé à une propriété ondulatoire) soit le chemin suivi par le photon dans l'interféromètre (associé à une propriété de type corpusculaire), n'est prise qu'une fois que le photon a franchi l'élément séparateur à l'entrée de l'interféromètre. Les résultats de cette expérience nous permettront de conclure qu'aucune réalité physique classique ne saurait être attribuée au photon indépendamment de l'appareil de mesure, comme le stipule le principe de complémentarité.

Fuel pin integrity assessment under large scale transients

International Nuclear Information System (INIS)

Dutta, B.K.

2006-01-01

The integrity of fuel rods under normal, abnormal and accident conditions is an important consideration during fuel design of advanced nuclear reactors. The fuel matrix and the sheath form the first barrier to prevent the release of radioactive materials into the primary coolant. An understanding of the fuel and clad behaviour under different reactor conditions, particularly under the beyond-design-basis accident scenario leading to large scale transients, is always desirable to assess the inherent safety margins in fuel pin design and to plan for the mitigation the consequences of accidents, if any. The severe accident conditions are typically characterized by the energy deposition rates far exceeding the heat removal capability of the reactor coolant system. This may lead to the clad failure due to fission gas pressure at high temperature, large- scale pellet-clad interaction and clad melting. The fuel rod performance is affected by many interdependent complex phenomena involving extremely complex material behaviour. The versatile experimental database available in this area has led to the development of powerful analytical tools to characterize fuel under extreme scenarios