WorldWideScience

Sample records for fissionable material behaviour

  1. Material synergism fusion-fission

    International Nuclear Information System (INIS)

    Sankara Rao, K.B.; Raj, B.; Cook, I.; Kohyama, A.; Dudarev, S.

    2007-01-01

    In fission and fusion reactors the common features such as operating temperatures and neutron exposures will have the greatest impact on materials performance and component lifetimes. Developing fast neutron irradiation resisting materials is a common issue for both fission and fusion reactors. The high neutron flux levels in both these systems lead to unique materials problems like void swelling, irradiation creep and helium embitterment. Both fission and fusion rely on ferritic-martensitic steels based on 9%Cr compositions for achieving the highest swelling resistance but their creep strength sharply decreases above ∝ 823K. The use of oxide dispersion strengthened (ODS) alloys is envisaged to increase the operating temperature of blanket systems in the fusion reactors and fuel clad tubes in fast breeder reactors. In view of high operating temperatures, cyclic and steady load conditions and the long service life, properties like creep, low cycle fatigue,fracture toughness and creepfatigue interaction are major considerations in the selection of structural materials and design of components for fission and fusion reactors. Currently, materials selection for fusion systems has to be based upon incomplete experimental database on mechanical properties. The usage of fairly well developed databases, in fission programmes on similar materials, is of great help in the initial design of fusion reactor components. Significant opportunities exist for sharing information on technology of irradiation testing, specimen miniaturization, advanced methods of property measurement, safe windows for metal forming, and development of common materials property data base system. Both fusion and fission programs are being directed to development of clean steels with very low trace and tramp elements, characterization of microstructure and phase stability under irradiation, assessment of irradiation creep and swelling behaviour, studies on compatibility with helium and developing

  2. Activation and Radiation Damage Behaviour of Russian Structural Materials for Fusion Reactors in the Fission and Fusion Reactors

    International Nuclear Information System (INIS)

    Blokhin, A.; Demin, N.; Chernov, V.; Leonteva-Smirnova, M.; Potapenko, M.

    2006-01-01

    Various structural low (reduced) activated materials have been proposed as a candidate for the first walls-blankets of fusion reactors. One of the main problems connected with using these materials - to minimise the production of long-lived radionuclides from nuclear transmutations and to provide with good technological and functional properties. The selection of materials and their metallurgical and fabrication technologies for fusion reactor components is influenced by this factor. Accurate prediction of induced radioactivity is necessary for the development of the fusion reactor materials. Low activated V-Ti-Cr alloys and reduced activated ferritic-martensitic steels are a leading candidate material for fusion first wall and blanket applications. At the present time a range of compositions and an impurity level are still being investigated to better understand the sensitive of various functional and activation properties to small compositional variations and impurity level. For the two types of materials mentioned above (V-Ti-Cr alloys and 9-12 % Cr f/m steels) and manufactured in Russia (Russia technologies) the analysis of induced activity, hydrogen and helium-production as well as the accumulation of such elements as C, N, O, P, S, Zn and Sn as a function of irradiation time was performed. Materials '' were irradiated '' by fission (BN-600, BOR-60) and fusion (Russian DEMO-C Reactor Project) typical neutron spectra with neutron fluency up to 10 22 n/cm 2 and the cooling time up to 1000 years. The calculations of the transmutation of elements and the induced radioactivity were carried out using the FISPACT inventory code, and the different activation cross-section libraries like the ACDAM, FENDL-2/A and the decay data library FENDL-2/D. It was shown that the level of impurities controls a long-term behaviour of induced activity and contact dose rate for materials. From this analysis the concentration limits of impurities were obtained. The generation of gas

  3. Theories of fission gas behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Dias, J W.C. [Companhia Brasileira de Tecnologia Nuclear, Rio de Janeiro (Brazil). Diretoria de Tecnologia e Desenvolvimento; Merckx, K R

    1976-01-01

    A review is presented of the theoretical developments and experimental evidence that have helped to evolve current models used to describe the behavior of inert fission gases created during the irradiation of reactor fuel materials. The phenomena which are stressed relate primarily to steady state behavior of fuel elements but are also relevant to an understanding of transient behavior. The processes considered include gas atom solubility; gas atom diffusivity; bubble nucleation; and bubble growth by bubble coalescence.

  4. Fission product behaviour in severe accidents

    International Nuclear Information System (INIS)

    Jokiniemi, J.; Auvinen, A.; Maekynen, J.; Valmari, T.

    1998-01-01

    The understanding of fission product (FP) behaviour in severe accidents is important for source term assessment and accident mitigation measures. For example in accident management the operator needs to know the effect of different actions on the behaviour and release of fission products. At VTT fission product behaviour have been studied in different national and international projects. In this presentation the results of projects in EU funded 4th framework programme Nuclear Fission Safety 1994-1998 are reported. The projects are: fission product vapour/aerosol chemistry in the primary circuit (FI4SCT960020), aerosol physics in containment (FI4SCT950016), revaporisation of test samples from Phebus fission products (FI4SCT960019) and assessment of models for fission product revaporisation (FI4SCT960044). Also results from the national project 'aerosol experiments in the Victoria facility' funded by IVO PE and VTT Energy are reported

  5. International handling of fissionable material

    International Nuclear Information System (INIS)

    1975-01-01

    The opinion of the ministry for foreign affairs on international handling of fissionable materials is given. As an introduction a survey is given of the possibilities to produce nuclear weapons from materials used in or produced by power reactors. Principles for international control of fissionable materials are given. International agreements against proliferation of nuclear weapons are surveyed and methods to improve them are proposed. (K.K.)

  6. Fission gas behaviour in water reactor fuels

    International Nuclear Information System (INIS)

    2002-01-01

    During irradiation, nuclear fuel changes volume, primarily through swelling. This swelling is caused by the fission products and in particular by the volatile ones such as krypton and xenon, called fission gas. Fission gas behaviour needs to be reliably predicted in order to make better use of nuclear fuel, a factor which can help to achieve the economic competitiveness required by today's markets. These proceedings communicate the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors. (author)

  7. Fission product and aerosol behaviour within the containment

    International Nuclear Information System (INIS)

    Beard, A.M.; Benson, C.G.; Bowsher, B.R.; Dickinson, S.; Nichols, A.L.

    1990-04-01

    Experimental studies have been undertaken to characterise the behaviour of fission products in the containment of a pressurised water reactor during a severe accident. The following aspects of fission product transport have been studied: (a) aerosol nucleation, (b) vapour transport processes, (c) chemical forms of high-temperature vapours, (d) interaction of fission product vapours with aerosols generated from within the reactor core, (e) resuspension processes, (f) chemistry in the containment. Chemical effects have been shown to be important in defining and quantifying fission product source terms in a wide range of accident sequences. Both the chemical forms of the fission product vapours and their interactions with reactor materials aerosols could have a major effect on the magnitude and physicochemical forms of the radioactive emission from a severe reactor accident. Only the main conclusions are presented in this summary document; detailed technical aspects of the work are described in separate reports listed in the annex

  8. Brief description of out-of-pile test facilities for study in corrosion and fission product behaviour in flowing sodium

    International Nuclear Information System (INIS)

    Iizawa, K.; Sekiguchi, N.; Atsumo, H.

    1976-01-01

    The experimental methods to perform tests for study in corrosion and fission products behaviour in flowing sodium are outlined. Flow diagrams for the activated materials and fission products behaviour test loop are given

  9. Fission reactors and materials

    International Nuclear Information System (INIS)

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

  10. 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)

  11. Fission product behaviour in the primary circuit of an HTR

    International Nuclear Information System (INIS)

    Decken, C.B. von der; Iniotakis, N.

    1981-01-01

    The knowledge of fission product behaviour in the primary circuit of a High Temperature Reactor (HTR) is an essential requirement for the estimations of the availability of the reactor plant in normal operation, of the hazards to personnel during inspection and repair and of the potential danger to the environment from severe accidents. On the basis of the theoretical and experimental results obtained at the ''Institute for Reactor Components'' of the KFA Juelich /1/,/2/ the transport- and deposition behaviour of the fission- and activation products in the primary circuit of the PNP-500 reference plant has been investigated thoroughly. Special work had been done to quantify the uncertainties of the investigations and to calculate or estimate the dose rate level at different components of the primary cooling circuit. The contamination and the dose rate level in the inspection gap in the reactor pressure vessel is discussed in detail. For these investigations in particular the surface structure and the composition of the material, the chemical state of the fission products in the cooling gas, the composition of the cooling gas and the influence of dust on the transport- and deposition behaviour of the fission products have been taken into account. The investigations have been limited to the nuclides Ag-110m; Cs-134 and Cs-137

  12. Impact of fuel chemistry on fission product behaviour

    International Nuclear Information System (INIS)

    Poortmans, C.; Van Uffelen, P.; Van den Berghe, S.

    1999-01-01

    The report contains a series of papers presented at SCK-CEN's workshop on the impact of fuel chemistry on fission product behaviour. Contributing authors discuss different processes affecting the behaviour of fission products in different types of spent nuclear fuel. In addition, a number of papers discusses the behaviour of actinides and fission products released from spent fuel and vitrified high-level waste in geological disposal conditions

  13. Induced-Fission Imaging of Nuclear Material

    International Nuclear Information System (INIS)

    Hausladen, Paul; Blackston, Matthew A.; Mullens, James Allen; McConchie, Seth M.; Mihalczo, John T.; Bingham, Philip R.; Ericson, Milton Nance; Fabris, Lorenzo

    2010-01-01

    This paper presents initial results from development of the induced-fission imaging technique, which can be used for the purpose of measuring or verifying the distribution of fissionable material in an unopened container. The technique is based on stimulating fissions in nuclear material with 14 MeV neutrons from an associated-particle deuterium-tritium (D-T) generator and counting the subsequent induced fast fission neutrons with an array of fast organic scintillation detectors. For each source neutron incident on the container, the neutron creation time and initial trajectory are known from detection of the associated alpha particle of the d + t → α + n reaction. Many induced fissions will lie along (or near) the interrogating neutron path, allowing an image of the spatial distribution of prompt induced fissions, and thereby fissionable material, to be constructed. A variety of induced-fission imaging measurements have been performed at Oak Ridge National Laboratory with a portable, low-dose D-T generator, including single-view radiographic measurements and three-dimensional tomographic measurements. Results from these measurements will be presented along with the neutron transmission images that have been performed simultaneously. This new capability may have applications to a number of areas in which there may be a need to confirm the presence or configuration of nuclear materials, such as nuclear material control and accountability, quality assurance, treaty confirmation, or homeland security applications.

  14. Aqueous cutting fluid for machining fissionable materials

    Science.gov (United States)

    Duerksen, Walter K.; Googin, John M.; Napier, Jr., Bradley

    1984-01-01

    The present invention is directed to a cutting fluid for machining fissionable material. The cutting fluid is formed of glycol, water and boron compound in an adequate concentration for effective neutron attenuation so as to inhibit criticality incidents during machining.

  15. Nuclear materials for fission reactors

    International Nuclear Information System (INIS)

    Matzke, H.; Schumacher, G.

    1992-01-01

    This volume brings together 47 papers from scientists involved in the fabrication of new nuclear fuels, in basic research of nuclear materials, their application and technology as well as in computer codes and modelling of fuel behaviour. The main emphasis is on progress in the development of non -oxide fuels besides reporting advances in the more conventional oxide fuels. The two currently performed large reactor safety programmes CORA and PHEBUS-FP are described in invited lectures. The contributions review basic property measurements, as well as the present state of fuel performance modelling. The performance of today's nuclear fuel, hence UO 2 , at high burnup is also reviewed with particular emphasis on the recently observed phenomenon of grain subdivision in the cold part of the oxide fuel at high burnup, the so-called 'rim' effect. Similar phenomena can be simulated by ion implantation in order to better elucidate the underlying mechanism and reviews on high resolution electron microscopy provide further information. The papers will provide a useful treatise of views, ideas and new results for all those scientists and engineers involved in the specific questions of current nuclear waste management

  16. A concise review of Harwell modelling of fission gas behaviour

    International Nuclear Information System (INIS)

    Wood, M.H.; Hayns, M.R.

    1976-07-01

    A review is presented of recent theoretical studies, performed at AERE Harwell, of fission gas behaviour in nuclear fuels. This includes a brief description of the rather sophisticated model approach and a discussion of the application of these models to predicting fission gas release and swelling in both normal operational and transient regimes. These studies have resulted in the derivation of more computationally efficient models which are also described. (author)

  17. METHOD OF JACKETING FISSIONABLE MATERIALS

    Science.gov (United States)

    Foster, L.M.

    1959-02-01

    An improvement is presented in the jacketing of a metal body accomplished by electroplating upon that portion of the metal container to be protected from the bonding material a niatcrial such as Cr which is impermeable to the bonding material. After the bonding operation the electroplate is removed and the metal container surfuce, unimpaired, may be welded to a cap which effects a closure. Generally in such an operation the metal body is U, the metal container is Al and the bonding material is a Zn alloy.

  18. International safeguards of fissionable material

    International Nuclear Information System (INIS)

    Tempus, P.

    1991-01-01

    From the very beginning nuclear fissile materials have been subject to state and - outside nuclear weapon states - also to international monitoring. The latter was a principal task of the International Atomic Energy Agency, a UN affiliated organisation formed in 1957 based in Vienna. The legal, technical and political aspects of its monitoring activity are explained

  19. 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.

  20. 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

  1. Fission product chemistry and aerosol behaviour in the primary circuit of a pressurised water reactor under severe accident conditions

    International Nuclear Information System (INIS)

    Bowsher, B.R.

    1985-09-01

    Three key accident sequences are considered covering a representative range of different environments of pressure, flow, temperature history and degree of zircaloy oxidation, and their principle thermal hydraulic and physical characteristics affecting chemistry behaviour are identified. Inventories, chemical forms and timing of fission product release are summarized together with the major sources of structural materials and their release characteristics. Chemistry of each main fission product species is reviewed from available experimental and/or theoretical data. Studies modelling primary circuit fission product behaviour are reviewed. Requirements for further study are assessed. (UK)

  2. Prediction of fission product and aerosol behaviour during a postulated severe accident in a LWR

    International Nuclear Information System (INIS)

    Guentay, S.; Aeby, F.; Raguin, M.; Passalacqua, R.

    1990-02-01

    Lack of appropriate energy removal causes fuel elements in a reactor core to overheat and may eventually cause core to degrade. Fission products will be emitted from a degraded reactor core. Aerosols are generated when the vapours of various fuel and structural materials reach a cold environment and nucleate. In addition to the fission products release and aerosol generation taking place in the reactor vessel, some more fission products release and aerosol generation will occur when the molten core debris leaves the pressure vessel bottom head and comes in contact with the pedestal concrete floor. Fission products, if they are released to environment from the containment boundary, exert a great danger to public health. A source term is defined as the quantity, timing, and characteristics of the release of radionuclide material to the environment following a postulated severe accident. At PSI a considerable effort hase been spent in investigating and establishing a source term assessment methodology in order to predict the source term for a given Light Water Reactor (LWR) accident scenario. This report introduces the computer programs and the methods associated with the release of the fission products, generation of the aerosols and behaviour of the aerosols in LWR compartments used for a source term assessment analysis at PSI. (author) 4 figs., 5 tabs., 28 refs

  3. Rim formation and fission gas behaviour: some structure remarks

    International Nuclear Information System (INIS)

    Spino, J.; Papaioannou, D.; Ray, I.; Baron, D.

    2002-01-01

    In high burn-up LWR nuclear fuel an increase of the Xe-mobility is observed in the rim region according to EPMA. This often coincides with an increase of the local porosity and the grain subdivision of the material in regions around the pores. The restructuring does not always imply disappearance of the prior grain boundaries. This seems to occur in a final step. Micro-XRD studies also show a contraction of the fuel lattice in the rim zone, reflecting mainly the release of accumulated stresses during irradiation, via reordering of defects and defect complexes, including sub-grain formation and displacement of Xe traps. The lattice contraction is not measurable when the fraction of restructured areas is low and the prior grain structure still remains. Nevertheless, in such a case, even the Xe signal by EPMA is observed to decrease, anticipating the displacement of Xe inside the grains, probably towards cavities. However, the quantitative proportion of Xe in matrix and pores can not be given by EPMA. This is confirmed by TEM examinations, showing still plenty of gas bubbles inside restructured grains, in spite of the low Xe signal detected by EPMA. An alternative determination therefore appears necessary. The fission gas release (FGR) behaviour of the rim zone seems then to depend basically on the efficiency of gas retention in its porosity. The closed character of these pores and the low percolation probability derived from the high pore to grain size ratio anticipate a low incidence of open porosity. Also, mechanical tests suggest a low pore interconnection probability by microcracking. However, at very high local burn-ups (>150 GWd/tM), too high porosity values are determined compared to the values derived from immersion density and solid swelling, suggesting the potential existence of open channels. Also, abnormally high porosity values by quantitative metallography might arise from grain pullout during sample preparation. Here, a rough estimation of the release

  4. Thermal Energetic Reactor with High Reproduction of Fission Materials

    Directory of Open Access Journals (Sweden)

    Vladimir M. Kotov

    2012-01-01

    On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

  5. Nuclear data for structural materials of fission and fusion reactors

    International Nuclear Information System (INIS)

    Goulo, V.

    1989-06-01

    The document presents the status of nuclear reaction theory concerning optical model development, level density models and pre-equilibrium and direct processes used in calculation of neutron nuclear data for structural materials of fission and fusion reactors. 6 refs

  6. Identification of Fissionable Materials Using the Tagged Neutron Technique

    International Nuclear Information System (INIS)

    Keegan, R.P.; Hurley, J.P.; Tinsley, J.R.; Trainham, R.

    2009-01-01

    This summary describes experiments to detect and identify fissionable materials using the tagged neutron technique. The objective of this work is to enhance homeland security capability to find fissionable material that may be smuggled inside shipping boxes, containers, or vehicles. The technique distinguishes depleted uranium from lead, steel, and tungsten. Future work involves optimizing the technique to increase the count rate by many orders of magnitude and to build in the additional capability to image hidden fissionable materials. The tagged neutron approach is very different to other techniques based on neutron die-away or photo-fission. This work builds on the development of the Associated Particle Imaging (API) technique at the Special Technologies Laboratory (STL). Similar investigations have been performed by teams at the Oak Ridge National Laboratory (ORNL), the Khlopin Radium Institute in Russia, and by the EURITRACK collaboration in the European Union

  7. 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)

  8. Space Fission Reactor Structural Materials: Choices Past, Present and Future

    International Nuclear Information System (INIS)

    Busby, Jeremy T.; Leonard, Keith J.

    2007-01-01

    Nuclear powered spacecraft will enable missions well beyond the capabilities of current chemical, radioisotope thermal generator and solar technologies. The use of fission reactors for space applications has been considered for over 50 years, although, structural material performance has often limited the potential performance of space reactors. Space fission reactors are an extremely harsh environment for structural materials with high temperatures, high neutron fields, potential contact with liquid metals, and the need for up to 15-20 year reliability with no inspection or preventative maintenance. Many different materials have been proposed as structural materials. While all materials meet many of the requirements for space reactor service, none satisfy all of them. However, continued development and testing may resolve these issues and provide qualified materials for space fission reactors.

  9. Organizational Behaviour Study Material

    OpenAIRE

    P. Sreeramana Aithal

    2016-01-01

    An overview of Organizational Behaviour – History of Organisational Behaviour and its emergence as a disciple-emerging perspective Organizational Behaviour. Individual process in organisation – Learning, perception and attribution- Individual differences - Basic concepts of motivation - Advanced concepts of motivation. Group process in Organisation – Group dynamics, leadership theories - Power, politics and conflict - inter- personal communication. Enhancing individu...

  10. Fission gas behaviour in UO2 under steady state and transient conditions

    International Nuclear Information System (INIS)

    Zimmermann, H.

    1980-01-01

    Fission gas behaviour in UO 2 is determined by the limited capacity of the fuel to retain fission gas. This capacity depends primarily on temperature, but also on fission rate, pressure loading, and fuel microstructure. Under steady state irradiation conditions fission gas behaviour can be described qualitatively as follows: At the beginning of the irradiation most of the fission gas remains in the grains in irradiation-induced solution. With increasing gas content in the grains the gas transport to the grain boundaries increases, too. The fission gas release from the grain boundaries occurs primarily by interlinkage of inter-granular bubbles. The fission gas release without noticeable fuel swelling during the short-term heating in the LOCA tests and the powdering of the high burnup UO 2 in the annealing tests can only be accounted for by formation of inter-granular separations, which are caused by the fission gas accumulated in the grain boundaries. Besides this short-term effect there are diffusion-controlled long-term effects, such as growth and coalescence of bubbles and formation of inter-connected porosity, which result in time-dependent fission gas release and fuel swelling

  11. Thermal Energetic Reactor with High Reproduction of Fission Materials

    International Nuclear Information System (INIS)

    Kotov, V.M.

    2012-01-01

    Existing thermal reactors are energy production scale limited because of low portion of raw uranium usage. Fast reactors are limited by reprocessing need of huge mass of raw uranium at the initial stage of development. The possibility of development of thermal reactors with high fission materials reproduction, which solves the problem, is discussed here. Neutron losses are decreased, uranium-thorium fuel with artificial fission materials equilibrium regime is used, additional in-core and out-core neutron sources are used for supplying of high fission materials reproduction. Liquid salt reactors can use dynamic loading regime for this purpose. Preferable construction is channel type reactor with heavy water moderator. Good materials for fuel element shells and channel walls are zirconium alloys enriched by 90Zr. Water cooled reactors with usage 12% of raw uranium and liquid metal cooled reactors with usage 25% of raw uranium are discussed. Reactors with additional neutron sources obtain full usage of raw uranium with small additional energy expenses. On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

  12. EURISOL-DS Multi-MW Target: Thermal Behaviour of the fission target disk arrangement inspired by the MAFF project

    CERN Document Server

    Cyril Kharoua, Yacine Kadi and the EURISOL-DS Task#2 collaboration

    This technical note summarises the design calculations performed within Task #2 of the European Isotope Separation On-Line Radioactive Ion Beam Facility Design Study (EURISOL-DS) [1] for the thermal behaviour of the fission target.A preliminary study was carried out in order to determine the heat deposition within the fissile material and estimate the temperature rise. This new solution takes into account the problems related to effusion/diffusion of radioactive isotopes inside a thick target. To enhance the extraction rates and the thermal behaviour it is proposed to study a solution where the fissile material is split into an arrangement of disks.

  13. Radiation effects in fuel materials for fission reactors

    International Nuclear Information System (INIS)

    Matzke, H.

    1983-01-01

    Physical and chemical changes that occur in fuel materials during fission are described. Emphasis is placed on the fuels used today, or those foreseen for the future, hence oxides and carbides of uranium and plutonium. Examples are given to illustrate the most interesting neutron effects. (author)

  14. 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

  15. 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

  16. Early results utilizing high-energy fission product gamma rays to detect fissionable material in cargo

    International Nuclear Information System (INIS)

    Slaughter, D.R.; Accatino, M.R.; Alford, O.J.; Bernstein, A.; Descalle, M.; Gosnell, T.B.; Hall, J.M.; Loshak, A.; Manatt, D.R.; McDowell, M.R.; Moore, T.L.; Petersen, D.C.; Pohl, B.A.; Pruet, J.A.; Prussin, S.G.

    2004-01-01

    Full text: A concept for detecting the presence of special nuclear material ( 235 U or 239 Pu) concealed in inter modal cargo containers is described. It is based on interrogation with a pulsed beam of 6-8 MeV neutrons and fission events are identified between beam pulses by their β-delayed neutron emission or β -delayed high-energy γ-radiation. The high-energy γ-ray signature is being employed for the first time. Fission product γ-rays above 3 MeV are distinct from natural radioactivity and from nearly all of the induced activity in a normal cargo. High-energy γ-radiation is nearly 10X more abundant than the delayed neutrons and penetrates even thick cargo's readily. The concept employs two large (8x20 ft) arrays of liquid scintillation detectors that have high efficiency for the detection of both delayed neutrons and delayed γ-radiation. Detector backgrounds and potential interferences with the fission signature radiation have been identified and quantified. This information, together with predicted signature strength, has been applied to the estimation of detection probability for the nuclear material and estimation of false alarm rates. This work was performed under the auspices of the U.S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under contract No. W-7405-Eng-48

  17. Continuous fluid bed reactor for fissionable material

    International Nuclear Information System (INIS)

    Ziegler, D.L.

    1975-01-01

    Plutonium (Pu) purification and plutonium hexafluoride (PuF 6 ) formation are achieved on a continuous basis by feeding particulate material into one end of an elongated and horizontally disposed vessel having an upper section with generally converging side walls and a lower section with generally vertical side walls, compartmented throughout its length by transversely disposed baffles, so that particulate material flows through the vessel in vertical generally zigzag fashion, being fluidized by dispersing gas that enters the compartment from a lower narrow compartment and discharges through an upper widened compartment. Vaporous PuF 6 formed from a reaction between the dispersing gas and the particulate material discharges through the upper widened compartment and solid impurities discharge for collection through a port at a far or distal end of the elongated vessel. (U.S.)

  18. Overview of standards subcommittee 8, fissionable materials outside reactors

    International Nuclear Information System (INIS)

    McLaughlin, T.P.

    1996-01-01

    The American Nuclear Society's Standards Subcommittee 8, titled open-quotes Fissionable Materials Outside Reactors,close quotes has worked for the past 35 yr to prepare and promote standards on nuclear criticality safety for the handling, processing, storing, and transportation of fissionable materials outside reactors. The reader is referred to the Transactions of the American Nuclear Society, Vols. 39 (1981) and 64 (1991), for previous papers associated with ANS-8 poster sessions. In addition to discussions on the then-current standards, the reader will find articles on working group efforts that never materialized into standards, such as proposed 8.13, open-quotes Use of the Solid-Angle Method in Nuclear Criticality Safety,close quotes and on applications and critiques of current standards. The paper by McLendon in Vol. 39 is particularly interesting as an overview of the early history of ANS-8 and its standards

  19. Neutron irradiation facilities for fission and fusion reactor materials studies

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1985-01-01

    The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

  20. 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.

  1. Fundamentals of passive nondestructive assay of fissionable material: laboratory workbook

    International Nuclear Information System (INIS)

    Reilly, T.D.; Augustson, R.H.; Parker, J.L.; Walton, R.B.; Atwell, T.L.; Umbarger, C.J.; Burns, C.E.

    1975-02-01

    This workbook is a supplement to LA-5651-M, ''Fundamentals of Passive Nondestructive Assay of Fissionable Material'' which is the text used during the Nondestructive Assay Training Session given by Group A-1 of the Los Alamos Scientific Laboratory. It contains the writeups used during the six laboratory sessions covering basic gamma-ray principles, quantitative gamma-ray measurements, uranium enrichment measurements, equipment holdup measurements, basic neutron principles, and quantitative neutron assay

  2. Fundamentals of passive nondestructive assay of fissionable material: laboratory workbook

    Energy Technology Data Exchange (ETDEWEB)

    Reilly, T.D.; Augustson, R.H.; Parker, J.L. Walton, R.B.; Atwell, T.L.; Umbarger, C.J.; Burns, C.E.

    1975-02-01

    This workbook is a supplement to LA-5651-M, ''Fundamentals of Passive Nondestructive Assay of Fissionable Material'' which is the text used during the Nondestructive Assay Training Session given by Group A-1 of the Los Alamos Scientific Laboratory. It contains the writeups used during the six laboratory sessions covering basic gamma-ray principles, quantitative gamma-ray measurements, uranium enrichment measurements, equipment holdup measurements, basic neutron principles, and quantitative neutron assay.

  3. Materials behaviour in PWRs core

    International Nuclear Information System (INIS)

    Barbu, A.; Massoud, J.P.

    2008-01-01

    Like in any industrial facility, the materials of PWR reactors are submitted to mechanical, thermal or chemical stresses during particularly long durations of operation: 40 years, and even 60 years. Materials closer to the nuclear fuel are submitted to intense bombardment of particles (mainly neutrons) coming from the nuclear reactions inside the core. In such conditions, the damages can be numerous and various: irradiation aging, thermal aging, friction wear, generalized corrosion, stress corrosion etc.. The understanding of the materials behaviour inside the cores of reactors in operation is a major concern for the nuclear industry and its long term forecast is a necessity. This article describes the main ways of materials degradation without and under irradiation, with the means used to foresee their behaviour using physics-based models. Content: 1 - structures, components and materials: structure materials, nuclear materials; 2 - main ways of degradation without irradiation: thermal aging, stress corrosion, wear; 3 - main ways of degradation under irradiation: microscopic damaging - point defects, dimensional alterations, evolution of mechanical characteristics under irradiation, irradiation-assisted stress corrosion cracking (IASCC), synergies; 4 - forecast of materials evolution under irradiation using physics-based models: primary damage - fast dynamics, primary damage annealing - slow kinetics microstructural evolution, impact of microstructural changes on the macroscopic behaviour, insight on modeling methods; 5 - materials change characterization techniques: microscopic techniques - direct defects observation, nuclear techniques using a particle beam, global measurements, mechanical characterizations; 6 - perspectives. (J.S.)

  4. Material challenges for the next generation of fission reactor systems

    International Nuclear Information System (INIS)

    Buckthorpe, Derek

    2010-01-01

    The new generation of fission reactor systems wil require the deployment and construction of a series of advanced water cooled reactors as part of a package of measures to meet UK and European energy needs and to provide a near term non-fossil fuel power solution that addresses CO 2 emission limits. In addition new longer term Generation IV reactor tye systems are being developed and evaluated to enhance safety, reliability, sustainability economics and proliferation resistance requirements and to meet alternative energy applications (outside of electricity generation) such as process heat and large scale hydrogen generation. New fission systems will impose significant challenges on materials supply and development. In the near term, because of the need to 'gear up' to large scale construction after decades of industrial hibernation/contraction and, in the longer term, because of the need for materials to operate under more challenging environments requiring the deployment and development of new alternative materials not yet established to an industrial stage. This paper investigates the materials challenges imposed by the new Generation III+ and Generation IV systems. These include supply and fabrication issues, development of new high temperature alloys and non-metallic materials, the use of new methods of manufacture and the best use of currently available resources and minerals. Recommendations are made as to how these materials challenges might be met and how governments, industry, manufacturers and researchers can all play their part. (orig.)

  5. Material challenges for the next generation of fission reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Buckthorpe, Derek [AMEC, Knutsford, Cheshire (United Kingdom)

    2010-07-01

    The new generation of fission reactor systems wil require the deployment and construction of a series of advanced water cooled reactors as part of a package of measures to meet UK and European energy needs and to provide a near term non-fossil fuel power solution that addresses CO{sub 2} emission limits. In addition new longer term Generation IV reactor tye systems are being developed and evaluated to enhance safety, reliability, sustainability economics and proliferation resistance requirements and to meet alternative energy applications (outside of electricity generation) such as process heat and large scale hydrogen generation. New fission systems will impose significant challenges on materials supply and development. In the near term, because of the need to 'gear up' to large scale construction after decades of industrial hibernation/contraction and, in the longer term, because of the need for materials to operate under more challenging environments requiring the deployment and development of new alternative materials not yet established to an industrial stage. This paper investigates the materials challenges imposed by the new Generation III+ and Generation IV systems. These include supply and fabrication issues, development of new high temperature alloys and non-metallic materials, the use of new methods of manufacture and the best use of currently available resources and minerals. Recommendations are made as to how these materials challenges might be met and how governments, industry, manufacturers and researchers can all play their part. (orig.)

  6. Structural materials issues for the next generation fission reactors

    Science.gov (United States)

    Chant, I.; Murty, K. L.

    2010-09-01

    Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

  7. 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

  8. COMEDIE BD1 experiment: Fission product behaviour during depressurization transients

    International Nuclear Information System (INIS)

    Gillet, R.; Brenet, D.; Hanson, D.L.; Kimball, O.F.

    1996-01-01

    An experimental program in the CEA COMEDIE loop has been carried out to obtain integral test data to validate the methods and transport models used to predict fission product release from the core and plate-out in the primary coolant circuit of the Modular High Temperature Gas Cooled Reactor (MHTGR) during normal operation and liftoff, and during rapid depressurization transients. The loop consists of an in-pile section with the fuel element, deposition section (heat exchanger), filters for collecting condensible Fission Productions (FP) during depressurization tests and an out-of-pile section devoted to chemical composition control of the gas and on-line analysis of gaseous FP. After steady state irradiation, the loop was subjected to a series of in-situ blowdowns at shear ratios (ratio of the wall shear stress during blowdown to that during steady state operation) ranging from 0.7 to 5.6. The results regarding the FP profiles on the plate-out section, before and after blowdowns are given. It appears that: the plate-out profiles depend on the FP chemistry; the depressurization phases have led to significant desorption of I 131, but on the contrary, they have almost no effect for the other FP such as Ag 110m, Cs 134, Cs 137 and Te 132. (author). 1 ref., 15 figs

  9. 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)

  10. 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)

  11. Precalculation of the fission gas behaviour in the MOL 7C/6 experiment with the LAKU model

    International Nuclear Information System (INIS)

    Vaeth, L.

    1988-03-01

    The fission gas behaviour in the planned experiment MOL 7C/6 is simulated with the Karlsruhe model LAKU, employing temperatures calculated with the pin behaviour model TRANSURANUS. Two different modes of experimental flow blockage simulation are investigated and compared to an estimated fission gas behaviour during a realistic blockage build-up. The results indicate, that the start-up procedure leading to greatly reduced fission gas content is the more realistic one. Details of the calculations and their results are presented in the report

  12. Criticality safety margins for mixtures of fissionable materials

    International Nuclear Information System (INIS)

    Williamson, T.G.; Mincey, J.F.

    1992-01-01

    In the determination of criticality safety margins, approximations for combinations of fissile and fissionable isotopes are sometimes used that go by names such as the rule of fractions or equivalency relations. Use of the rule of fractions to ensure criticality safety margins was discussed in an earlier paper. The purpose of this paper is to correct errors and to clarify some of the implications. Deviations of safety margins from those calculated by the rule of fractions are still noted; however, the deviations are less severe. Caution in applying such rules is still urged. In general, these approximations are based on American National Standard ANSI/ANS-8.15, Sec. 5.2. This section allows that ratios of material masses to their limits may be summed for fissile nuclides in aqueous solutions. It also allows the addition of nonfissile nuclides if an aqueous moderator is present and addresses the effects of infinite water or equivalent reflector. Water-reflected binary combinations of aqueous solutions of fissile materials, as well as binary combinations of fissile and fissionable metals, were considered. Some combinations were shown to significantly decrease the margin of subcriticality compared to the single-unit margins. In this study, it is confirmed that some combinations of metal units in an optimum geometry may significantly decrease the margin of subcriticality. For some combinations of aqueous solutions of fissile materials, the margin of subcriticality may also be reduced by very small amounts. The conclusion of Ref. 1 that analysts should be careful in applying equivalency relations for combining materials remains valid and sound advice. The ANSI/ANS standard, which allows the use of ratios of masses to their limits, applies to aqueous, fully water-reflected, single-unit solutions. Extensions to other situations should be considered with extreme care

  13. The mass transfer mechanism of fissile material due to fission

    International Nuclear Information System (INIS)

    Shafrir, N.H.

    1975-01-01

    A thin 252 Cf source of a mean thickness of an approXimately mono-atomic layer was used as an experimental model for the study of the basic mechanism of the knock-on process taking place in fissile material. Because of the thinness of the source it can be assumed that mainly primary knock-ons are formed. The ejection rate of knock-ons created by direct collisions between fission fragments and source atoms was measured as follows: the ejected atoms were collected in high vacuum on a catcher foil and 252 Cf determined by alpha spectroscopy using a silicon surface barrier detector. The number of 252 Cf ejected from the source in unit time could thus be determined while considering the anisotropy of ejection, geometry and counting efficiency. Taking into account the chemical composition of the source, eta(theor.) = 252 Cf atoms/fission was obtained. This result can be considered in reasonable agreement with experiment confirming that under the experimental conditions described, practically no knock-on cascade is formed. (B.G.)

  14. Disposal of fissionable material from dismantled nuclear weapons

    International Nuclear Information System (INIS)

    Taylor, J.J.

    1991-01-01

    The reduction in tensions between the United States and the Soviet Union has improved the prospects for nuclear disarmament, making it more likely that significant numbers of nuclear warheads will be dismantled by the United States and USSR in the foreseeable future. Thus, the question becomes more urgent as to the disposition of the weapons materials, highly enriched uranium and plutonium. It is timely, therefore, to develop specific plans for such disposal. The overall process for disposal of weapons materials by the burnup option involves the following steps: (1) removing the weapons material from the warheads, (2) converting the material to a fuel form suitable for power reactors, (3) burning it up as a power reactor fuel, and (4) removing the spent fuel and placing it in a permanent repository. This paper examines these four steps with the purpose of answering the following questions. What facilities would be appropriate for the disposal process? Do they need to be dedicated facilities, or could industrial facilities be used? What is the present projection of the economics of the burnup process, both the capital investment and the operating costs? How does one assure that fissionable materials will not be diverted to military use during the disposal process? Is the spent fuel remaining from the burnup process proliferation resistant? Would the disposal of spent fuel add an additional burden to the spent fuel permanent repository? The suggested answers are those of the author and do not represent a position by the Electric Power Research Institute

  15. Preliminary results utilizing high-energy fission product γ-rays to detect fissionable material in cargo

    Science.gov (United States)

    Slaughter, D. R.; Accatino, M. R.; Bernstein, A.; Church, J. A.; Descalle, M. A.; Gosnell, T. B.; Hall, J. M.; Loshak, A.; Manatt, D. R.; Mauger, G. J.; Moore, T. L.; Norman, E. B.; Pohl, B. A.; Pruet, J. A.; Petersen, D. C.; Walling, R. S.; Weirup, D. L.; Prussin, S. G.; McDowell, M.

    2005-12-01

    A concept for detecting the presence of special nuclear material (235U or 239Pu) concealed in intermodal cargo containers is described. It is based on interrogation with a pulsed beam of 7 MeV neutrons that produce fission events and their β-delayed neutron emission or β-delayed high-energy γ radiation between beam pulses provide the detection signature. Fission product β-delayed γ-rays above 3 MeV are nearly 10 times more abundant than β-delayed neutrons and are distinct from natural radioactivity and from nearly all of the induced activity in a normal cargo. Detector backgrounds and potential interferences with the fission signature radiation have been identified and quantified.

  16. Fission product behaviour - in particular Cs-137 - in HTR-TRISO-coated particle fuel

    International Nuclear Information System (INIS)

    Allelein, H.J.

    1980-12-01

    This work is performed between 1977 and 1979. The main task is to determine a temperature dependent diffusion coefficient of the fission product Cs-137 in the silicon carbide interlayer of HTR particles. The raw material is laso presented as the used measuring techniques and computer codes. The results are discussed in detail and some critical remarks are made about the efficiency of the silicon carbide interlayer to retent fission products including Ag-110m, Sr-90, and Ru-106, which temperature dependent diffusion coefficient is also been determined. (orig.) [de

  17. 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

  18. 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.)

  19. 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.)

  20. Irradiation effects and behaviour of fission products in zirconia and spinel

    International Nuclear Information System (INIS)

    Gentils, A.

    2003-10-01

    Crystalline oxides, such as zirconia (ZrO 2 ) and spinel (MgAl 2 O 4 ), are promising inert matrices for the transmutation of plutonium and minor actinides. This work deals with the study of the physico-chemical properties of these matrices, more specifically their behaviour under irradiation and their capacity to retain fission products. Irradiations at low energy and incorporation of stable analogs of fission products (Cs, I, Xe) into yttria-stabilized zirconia and magnesium-aluminate spinel single crystals were performed by using the ion implanter IRMA (CSNSM-Orsay). Irradiations at high energy were made on several heavy ion accelerators (GANIL-Caen, ISL-Berlin, HIL-Warsaw). The damage induced by irradiation and the release of fission products were monitored by in situ Rutherford Backscattering Spectrometry experiments. Transmission electron microscopy was also used in order to determine the nature of the damage induced by irradiation. The results show that irradiation of ZrO 2 and MgAl 2 O 4 with heavy ions (about hundred keV and about hundred MeV) induces a huge structural damage in crystalline matrices. Total disorder (amorphization) is however never reached in zirconia, contrary to what is observed in the case of spinel. The results also emphasize the essential role played by the concentration of implanted species on their retention capacity. A dramatic release of fission products was observed when the concentration exceeds a threshold of a few atomic percent. Irradiation of implanted samples with medium-energy noble-gas ions leads to an enhancement of the fission product release. The exfoliation of spinel crystals implanted at high concentration of Cs ions is observed after a thermal treatment at high temperature. (author)

  1. 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)

  2. Fission product behaviour during operation of the second Peach Bottom core

    International Nuclear Information System (INIS)

    Malinauskas, A.P.; Nordwall, H.J. de; Dyer, F.F.; Wichner, R.P.; Martin, W.J.; Kolb, J.O.

    1976-01-01

    The Peach Bottom high-temperature, gas-cooled reactor began operation on 1 June 1967 and continued power production until 9 October 1969, accumulating 452 equivalent full power days (EFPD) operation. After reload, power production with Core 2 began 14 July 1970 and terminated 31 October 1974 after 897 EFPD operation. Surveillance of fission product release and behaviour was intensified during Core 2 operation to permit a wider range of measurements to be made. In addition to monitoring the noble gas content of the fuel element purge system and the coolant circuit, the programme was extended to include measurements of radioactive and other condensible species (including dust) entering or exiting the core and steam generator, and of surface concentrations of gamma-emitting nuclides deposited on the primary coolant surfaces. These data, which were obtained over the operating period April 1971 - October 1974, are summarized and discussed. The data demonstrate that caesium behaviour in the coolant circuit during the first two-thirds of Core 2 life was primarily governed by caesium released during Core 1 operation. The data also indicate that whereas the steam generator surfaces attenuate molecular caesium concentrations in the coolant, the dust-borne component is remarkably persistent. Driver fuel elements were removed from the reactor after 385 EFPD, 701 EFPD, and at end-of-life. These fuel elements are at various stages of an intensive post-irradiation examination. Some of the axial and radial concentration profiles of fission products which have been obtained are likewise presented. Although these profiles indicate varied fission product behaviour, the observations can in general be qualitatively described on the basis of the operational histories of the fuel elements. (author)

  3. Organization of customs control of fissionable and other radioactive materials

    International Nuclear Information System (INIS)

    Ukhlinov, L.; Bojko, V.

    2001-01-01

    Among the routine inspection tasks of the Sheremetyevo customs office are tasks stemming from international commitments of Russia to prevent proliferation of nuclear weapons and material that can be used for making these weapons. These tasks are: radiation monitoring of all vehicles, passengers, their luggage and goods crossing the state border; inspection of fissionable and radioactive materials (FRM) legally transported by participants in the foreign trade activities with a view to checking that the declared data fully correspond to the actual radioactive cargo. Organizational measures and technical measures at the Sheremetyevo customs office are described in detail. The efficiency of the scheme is illustrated by the following figures. In 1997, when appropriate technical means and trained personnel were lacking, there were only 2 events of detecting items with a rather high radioactivity level in the luggage. In 1999, after the entire radiation monitoring system was fully deployed (i.e. the flight checkpoint was equipped with technical means of radiation monitoring, personnel was trained, special technologies and algorithms were developed), there were 61 events of radiation detection, and in 2000 there have been 90 events, including breaches of legal FRM traffic regulations through disagreement of declared and actual parameters. We believe that the above-considered organization of radiation monitoring allows effective and quite reliable control of and adequate response to possible illicit transport of FRM through the airport Sheremetyevo to other countries, including CIS. In the near future we plan to increase the efficiency of the radiation monitoring by integrating the currently operational customs-used stationary FRM detection systems into a single information network capable of providing simultaneous video-aided continuous nuclear monitoring at three terminals (Sheremetyevo-1, Sheremetyevo-2, Sheremetyevo-Cargo) with display of information at the workstation

  4. 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.

  5. 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

  6. Local system for control by console-mobile crane for russian depository of fissionable materials

    International Nuclear Information System (INIS)

    Troshchenko, V.G.; Kapustin, V.N.; Zinina, N.V.; Derbyshev, S.A.

    2005-01-01

    Description of crane of console-mobile type used for transportation of fissionable materials in depository with local control system is represented. Local control system realizes program control in real time [ru

  7. 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)

  8. Modelling of elasto-plastic material behaviour

    International Nuclear Information System (INIS)

    Halleux, J.P.

    1981-01-01

    The present report describes time-independent elasto-plastic material behaviour modelling techniques useful for implementation in fast structural dynamics computer programs. Elasto-plastic behaviour is characteristic for metallic materials such as steel and is thus of particular importance in the study of reactor safety-related problems. The classical time-independent elasto-plastic flow theory is recalled and the fundamental incremental stress-strain relationships are established for strain rate independent material behaviour. Some particular expressions useful in practice and including reversed loading are derived and suitable computational schemes are shwon. Modelling of strain rate effects is then taken into account, according to experimental data obtained from uniaxial tension tests. Finally qualitative strain rate history effects are considered. Applications are presented and illustrate both static and dynamic material behaviour

  9. Fission product behaviour in the Peach Bottom and Fort St. Vrain HTGRs

    International Nuclear Information System (INIS)

    Hanson, D.L.; Baldwin, N.L.; Strong, D.E.

    1981-01-01

    Actual operating data from the Peach Bottom (PB) and Fort St. Vrain (FSV) High-Temperature Gas-Cooled Reactors (HTGRs) have been compared with code predictions to assess the validity of the methods used to predict the behaviour of fission products in the primary coolant circuit. For both reactors the measured circuit activities were significantly below design values, and the observations generally verify the codes used for large HTGR design. The PB primary circuit after seven years of operation was exceptionally clean. A fuel element purge system virtually eliminated the release of fission gases into the primary coolant circuit. Extensive examinations at end-of-life revealed that only Cs and trace amounts of Sr had plated out in the circuit. Their plateout distributions were in excellent agreement with PAD code predictions. Most of the deposited activity was associated with carbonaceous surface films which resulted from occasional small inleakages of lubricating oil. Primary circuit activities in FSV during the first cycle were also very low. Noble gas activity was about 1% of the design limit; and the circulating iodines were at least one order of magnitude below the limit, although the measurement uncertainties are significant. The plateout per pass of the iodine isotopes increased with decreasing half-life (the value for I-131 is about 1% per pass) as predicted with the PADLOC code. Gamma scanning of two helium circulators indicated very low plateout activities. Iodine-131 was the principal fission product observed, along with small amounts of Cs-134, Cs-137, and Ba/La-140. (author)

  10. Study of the behaviour of tetracycline as fission products extracting agent

    International Nuclear Information System (INIS)

    Cunha, I.I.L.

    1983-01-01

    Both spectrophotometric and potentiometric titration techniques were used to show the formation of complexes between tetracycline and the elements: zirconium, uranium, molybdenum, strontium, barium and ruthenium. It has been verified that tetracycline does not form complexes with cesium, tellurium and iodine. Those techniques have also been used to determine the sites on the tetracycline molecule at which ions may be bound. The behaviour of tetracycline as an extracting agent for those elements, as well as for niobium and technetium has been studied and the influence of the acidity of the aqueous phase upon extraction of the elements mentioned has been considered. Extraction experiments were carried out in the presence of chloride, perchlorate, nitrate and sulfate ions. Studies have been made to determine whether or not the complex extracted into organic phase is really the complex formed between tetracycline and the elements considered as well as to determine the time of shaking necessary so that the equilibrium between the phases is attained. Based on all information obtained from extraction experiments made for uranium and the fission products Zr-95, Nb-95, Ce-141, La-140, Ru-103, Ba-140 and Cs-137, the possibility of using tetracycline for separating those fission products from each other and from uranium has been studies and a scheme for simultaneous separation of those elements has been proposed. The same study has been made for I-131, Tc-99m, Mo-99, Te-132, Np-239 and uranium. The method described is applicable to the separation of some fission products existing in solutions at tracer levels, and not to be used in nuclear fuel reprocessing or any other industrial application. (Author) [pt

  11. Thermochemical data for reactor materials and fission products

    International Nuclear Information System (INIS)

    Cordfunke, E.H.P.; Konings, R.J.M.

    1990-01-01

    This volume presents a collection of critically assessed data on inorganic compounds which are of special interest in nuclear reactor safety studies. Thermodynamic equilibrium calculations are an important and widely used instrument in the understanding of the chemical behavior and release of fission products in the course of nuclear reactor accidents. The reliability of such calculations is, nevertheless, limited by the availability of accurate input data for relevant compounds

  12. Materials compatibility considerations for a fusion-fission hybrid reactor design

    International Nuclear Information System (INIS)

    DeVan, J.H.; Tortorelli, P.F.

    1983-01-01

    The Tandem Mirror Hybrid Reactor is a fusion reactor concept that incorporates a fission-suppressed breeding blanket for the production of 233 U to be used in conventional fission power reactors. The present paper reports on compatibility considerations related to the blanket design. These considerations include solid-solid interactions and liquid metal corrosion. Potential problems are discussed relative to the reference blanket operating temperature (490 0 C) and the recycling time of breeding materials (<1 year)

  13. 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.)

  14. 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

  15. Modelling of buffer material behaviour

    International Nuclear Information System (INIS)

    Boergesson, L.

    1988-12-01

    Some material models of smectite rich buffer material suited for nuclear waste isolation are accounted for in the report. The application of these models in finite element calculations of some scenarios and performance are also shown. The rock shear scenario has been closely studied with comparisons between calculated and measured results. Sensitivity analyses of the effect of changing the density of the clay and the rate of shear have been performed as well as one calculation using a hollow steel cylinder. Material models and finite element calculations of canister settlement, thermomechanical effects and swelling are also accounted for. The report shows the present state of the work to establish material models and calculation tools which can be used at the final design of the repository. (31 illustrations)

  16. Studies on fission tracks and distributions of uranium and rare earths in granite materials

    International Nuclear Information System (INIS)

    Matsuda, Hiroshi; Sakanoue, Masanobu

    1987-01-01

    Many materials contain fossil records of the slow spontaneous fission of uranium they contain as an impurity. Fission fragments, heavy charged particles released in each fission event, produce microscopic trails of radiation damage that may persist over geological times and may be developed to a size observable under an optical microscope by a suitable etching treatment. Such tracks are also produced by fissions induced by thermal neutron irradiation of the uranium. When the material is heated sufficiently, it anneals and the the microscopic trails become shorter and narrower. The track density decreases, because the chemical etchant will not reach some of the shortened tracks. Measurements of track densities before and after annealing can be used, along with laboratory studies of annealing rates, to determine the annealing temperature. Also, the track density of induced fissions is related to the concentration of uranium and the fluence of neutrons to which it was exposed. If the track density due to induced fissions can be distinguished from that due to fossil tracks, estimates of either the concentration or the fluence can be made if the other is known. Two such materials (one a fragment of a granite paving stone, the other a piece of stained glass from a cathedral window) that had been exposed to the atomic bomb at Nagasaki were used in the present work. The fossil record in zircons in the granite was used to estimate the temperature to which it had been exposed in the bombing. Induced fissions were used to estimate the concentration of uranium in the zircons. Nonuniform heating and cooling and nearly uniform exposure to the neutrons make the granite sample unsuitable for determining the neutron fluence from the bomb. Induced fissions in the stained glass were used to estimate the concentration of uranium and the thermal neutron fluence from the A-bomb. Annealing of tracks in glass was also studied

  17. Irradiation effects and behaviour of fission products in zirconia and spinel; Effets d'irradiation et comportement des produits de fission dans la zircone et le spinelle

    Energy Technology Data Exchange (ETDEWEB)

    Gentils, A

    2003-10-01

    Crystalline oxides, such as zirconia (ZrO{sub 2}) and spinel (MgAl{sub 2}O{sub 4}), are promising inert matrices for the transmutation of plutonium and minor actinides. This work deals with the study of the physico-chemical properties of these matrices, more specifically their behaviour under irradiation and their capacity to retain fission products. Irradiations at low energy and incorporation of stable analogs of fission products (Cs, I, Xe) into yttria-stabilized zirconia and magnesium-aluminate spinel single crystals were performed by using the ion implanter IRMA (CSNSM-Orsay). Irradiations at high energy were made on several heavy ion accelerators (GANIL-Caen, ISL-Berlin, HIL-Warsaw). The damage induced by irradiation and the release of fission products were monitored by in situ Rutherford Backscattering Spectrometry experiments. Transmission electron microscopy was also used in order to determine the nature of the damage induced by irradiation. The results show that irradiation of ZrO{sub 2} and MgAl{sub 2}O{sub 4} with heavy ions (about hundred keV and about hundred MeV) induces a huge structural damage in crystalline matrices. Total disorder (amorphization) is however never reached in zirconia, contrary to what is observed in the case of spinel. The results also emphasize the essential role played by the concentration of implanted species on their retention capacity. A dramatic release of fission products was observed when the concentration exceeds a threshold of a few atomic percent. Irradiation of implanted samples with medium-energy noble-gas ions leads to an enhancement of the fission product release. The exfoliation of spinel crystals implanted at high concentration of Cs ions is observed after a thermal treatment at high temperature. (author)

  18. HAC and fission reactors

    International Nuclear Information System (INIS)

    Fujiwara, I.; Moriyama, H.; Tachikawa, E.

    1984-01-01

    In the fission process, newly formed fission products undergo hot atom reactions due to their energetic recoil and abnormal positive charge. The hot atom reactions of the fission products are usually accompanied by secondary effects such as radiation damage, especially in condensed phase. For reactor safety it is valuable to know the chemical behaviour and the release behaviour of these radioactive fission products. Here, the authors study the chemical behaviour and the release behaviour of the fission products from the viewpoint of hot atom chemistry (HAC). They analyze the experimental results concerning fission product behaviour with the help of the theories in HAC and other neighboring fields such as radiation chemistry. (Auth.)

  19. Exploiting Fission Chain Reaction Dynamics to Image Fissile Materials

    Science.gov (United States)

    Chapman, Peter Henry

    Radiation imaging is one potential method to verify nuclear weapons dismantlement. The neutron coded aperture imager (NCAI), jointly developed by Oak Ridge National Laboratory (ORNL) and Sandia National Laboratories (SNL), is capable of imaging sources of fast (e.g., fission spectrum) neutrons using an array of organic scintillators. This work presents a method developed to discriminate between non-multiplying (i.e., non-fissile) neutron sources and multiplying (i.e., fissile) neutron sources using the NCAI. This method exploits the dynamics of fission chain-reactions; it applies time-correlated pulse-height (TCPH) analysis to identify neutrons in fission chain reactions. TCPH analyzes the neutron energy deposited in the organic scintillator vs. the apparent neutron time-of-flight. Energy deposition is estimated from light output, and time-of-flight is estimated from the time between the neutron interaction and the immediately preceding gamma interaction. Neutrons that deposit more energy than can be accounted for by their apparent time-of-flight are identified as fission chain-reaction neutrons, and the image is reconstructed using only these neutron detection events. This analysis was applied to measurements of weapons-grade plutonium (WGPu) metal and 252Cf performed at the Nevada National Security Site (NNSS) Device Assembly Facility (DAF) in July 2015. The results demonstrate it is possible to eliminate the non-fissile 252Cf source from the image while preserving the fissileWGPu source. TCPH analysis was also applied to additional scenes in which theWGPu and 252Cf sources were measured individually. The results of these separate measurements further demonstrate the ability to remove the non-fissile 252Cf source and retain the fissileWGPu source. Simulations performed using MCNPX-PoliMi indicate that in a one hour measurement, solid spheres ofWGPu are retained at a 1sigma level for neutron multiplications M -˜ 3.0 and above, while hollowWGPu spheres are

  20. Thorium determination in water and biological materials by fission track

    International Nuclear Information System (INIS)

    Melo Ferreira, A.C. de.

    1989-01-01

    As a segment of a research programme on the study of bioaccumulation of radionuclides, in animals and vegetables from Morro do Ferro, Pocos de Caldas, MG, a fission track method for the determination of low levels of thorium in environmental samples was developed as an alternative for alpha spectroscopy. The study was carried out in early alpha spectroscopy samples, containing high levels of 228 Th activity, which makes difficult the 232 Th determination. A dry way method for thorium evaluation was developed. Pieces of membrane filters, containing La F 3 (Th), coupled to Makrofol detectors, were irradiated in the core of a research reactor, IEA-R1 (IPEN). (author)

  1. 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.

  2. Contribution to the study of the behaviour, in the urban environment, during the runoff of rainwater, of the fission products emitted during a nuclear accident

    International Nuclear Information System (INIS)

    Pioch, M.

    1993-01-01

    In the context of research into the environmental consequences of a serious accident occurring on a pressurized water reactor, this paper concerns the experimental study of behaviour of five fission products (caesium, strontium, iodine, ruthenium and tellurium) in the urban environment under the action of rainwater. Stable or radioactive multiple-element aerosols were produced. Their physicochemical characteristics and their solubility in rainwater were studied. Caesium and rubidium forms solutions totally and quickly, while strontium is partially soluble (approximately 50 %) and iodine is only slightly soluble. The behaviour of fission products on five urban surfaces was then studied. Batch experiments showed that the retention of dissolved forms of radioelements varied according to the material. The reactions involved are ion exchange reactions. The presence of certain ions in water (in particular NH 4 + ) increase the desorption of radioelements. Using a laboratory rainfall simulator, the re-entrainment of fission products by rainwater was examined. Two modes of deposition and two intensities of rainfall were simulated. The desorption of radioelements is greater after wet deposition and remobilization is reduced by an increase in intensity of rainfall. An addition of NH 4 + in water is especially effective in the case of wet depositions. Suggestions are made in order to improve experimental protocols and continue the research. (author). 75 refs., 51 figs., 69 tabs., 14 appends

  3. First wall material damage induced by fusion-fission neutron environment

    Energy Technology Data Exchange (ETDEWEB)

    Khripunov, Vladimir, E-mail: Khripunov_VI@nrcki.ru

    2016-11-01

    Highlights: • The highest damage and gas production rates are experienced within the first wall materials of a hybrid fusion-fission system. • About ∼2 times higher dpa and 4–5 higher He appm are expected compared to the values distinctive for a pure fusion system at the same DT-neutron wall loading. • The specific nuclear heating may be increased by a factor of ∼8–9 due to fusion and fission neutrons radiation capture in metal components of the first wall. - Abstract: Neutronic performance and inventory analyses were conducted to quantify the damage and gas production rates in candidate materials when used in a fusion-fission hybrid system first wall (FW). The structural materials considered are austenitic SS, Cu-alloy and V- alloys. Plasma facing materials included Be, and CFC composite and W. It is shown that the highest damage rates and gas particles production in materials are experienced within the FW region of a hybrid similar to a pure fusion system. They are greatly influenced by a combined neutron energy spectrum formed by the two-component fusion-fission neutron source in front of the FW and in a subcritical fission blanket behind. These characteristics are non-linear functions of the fission neutron source intensity. Atomic displacement damage production rate in the FW materials of a subcritical system (at the safe subcriticality limit of ∼0.95 and the neutron multiplication factor of ∼20) is almost ∼2 times higher compared to the values distinctive for a pure fusion system at the same 14 MeV neutron FW loading. Both hydrogen (H) and helium (He) gas production rates are practically on the same level except of about ∼4–5 times higher He-production in austenitic and reduced activation ferritic martensitic steels. A proper simulation of the damage environment in hybrid systems is required to evaluate the expected material performance and the structural component residence times.

  4. 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.

  5. 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

  6. The study of two, three and four dimensional nonlinear dynamics of nuclear fission reactors and effective parameters on its behaviour

    International Nuclear Information System (INIS)

    Tajik, M.; Ghasemizad, A.

    2008-01-01

    In this research, new physical fission reactor parameters which have very sensitive effects on the qualitative behavior of a reactor, are introduced. Therefore, the two, the nonlinear dynamics of two, three and four dimensional, considering almost the effective parameters are formulated for describing nuclear fission reactor systems. Using both analytical and numerical methods, the stability and instability of the given dynamical equations and the conditions of stability are studied in these systems. We have shown that the two parameters of the mean energy residence time in fuel and coolant and also their ratios have the most qualitative effects on the dynamical behaviour of a typical nuclear fission reactor. Increasing or decreasing of these parameters from a captain limit can lead to stability or un stability in a given system

  7. Structural stability and fission product behaviour in U{sub 3}Si

    Energy Technology Data Exchange (ETDEWEB)

    Middleburgh, S.C., E-mail: simon.middleburgh@hotmail.co.uk [IME, Australian Nuclear Science and Technology Organisation, Lucas Heights, New South Wales (Australia); Westinghouse Electric Sweden AB, SE-72163 Västerås (Sweden); Burr, P.A. [Department of Materials, Imperial College London, South Kensington, London SW7 2AZ (United Kingdom); King, D.J.M.; Edwards, L.; Lumpkin, G.R. [IME, Australian Nuclear Science and Technology Organisation, Lucas Heights, New South Wales (Australia); Grimes, R.W. [Department of Materials, Imperial College London, South Kensington, London SW7 2AZ (United Kingdom)

    2015-11-15

    The crystalline and amorphous structures of U{sub 3}Si have been investigated using density functional theory techniques for the first time. The effects of disorder and the impact of fission products has been separated to understand the swelling characteristics of U{sub 3}Si in both crystalline and amorphous U{sub 3}Si. Initially, the stability of the three experimentally observed polymorphs of U{sub 3}Si were explored. Subsequently, we modelled the amorphous U{sub 3}Si system and conclude that initial increase in volume observed experimentally at low temperature corresponds well with the volume change that occurs with the observed amorphisation of the material. The solubility of Xe and Zr into both the crystalline and amorphous systems was subsequently investigated.

  8. Heat and fission product transport in molten core material pool with crust

    International Nuclear Information System (INIS)

    Yun, J.I.; Suh, K.Y.; Kang, C.S.

    2005-01-01

    Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the reactor vessel during a severe accident. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool is estimated by product of the mass concentration and energy conversion factor of each fission product. Twenty-nine elements are chosen and classified by their chemical properties to calculate heat generation rate in the pool. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis is performed for heat and fission product transport in a molten core material pool during the Three Mile Island Unit 2 (TMI-2) accident. The pool is assumed to be a partially filled hemisphere, whose change in geometry is neglected during the numerical calculation. Calculated results indicate that the peak temperature in the molten pool is significantly lowered, since a substantial amount of the volatile fission products is released from the molten pool during progression of the accident. The results may directly be applied to the existing severe accident analysis codes to more mechanistically determine the thermal load to the reactor vessel lower head during the in-vessel retention

  9. Thermochemical data for reactor materials and fission products: The ECN database

    International Nuclear Information System (INIS)

    Cordfunke, E.H.P.; Konings, R.J.M.

    1993-02-01

    The activities of the authors regarding the compilation of a database of thermochemical properties for reactor materials and fission products is reviewed. The evaluation procedures and techniques are outlined and examples are given. In addition, examples of the use of thermochemical data for the application in the field of Nuclear Technology are given. (orig.)

  10. On the fissionable materials management system in the process of nuclear disarmament

    International Nuclear Information System (INIS)

    Vikharev, S.S.; Mikijchuk, N.B.; Pinaev, V.S.; Sudarushkin, I.S.; Yuferev, V.I.

    1994-01-01

    Various scenarios of nuclear weapons proliferation and goals of fissionable material accounting and control system (FMACS) are considered. Ways of improving FMACS in Russia under a complicated social situation are discussed. This improvement should follow two directions: introduction of non-destructive control methods and accounting and control process automation

  11. Study of the behaviour of cesium fission product in uranium dioxide by the ab initio method

    International Nuclear Information System (INIS)

    Gupta, Florence

    2008-01-01

    The knowledge of the behaviour of fission products in the nuclear fuel is very important for safety considerations and for understanding the evolution of the fuel properties under irradiation. In this work, we focussed mainly on the behaviour of caesium in UO 2 through ab initio studies of its solubility at point defects in the matrix, its diffusion and its contribution to the formation of solid phases in the fuel. The role of electronic correlation effects of the f electrons of uranium on these properties and on the description of the defect free crystal, is assessed. The formation energies of the main point defects are calculated and their concentration as a function of fuel stoichiometry and temperature is estimated. The migration barriers and migration paths for the self-diffusion of oxygen and uranium vacancies and oxygen interstitials in UO 2 are discussed. The solubility of Cs is found to be very low in UO 2 in agreement with experimental findings. The most favourable trapping sites are determined as a function of oxygen concentration in the fuel. Our results show that in the hyper-stoichiometric regime, the diffusion of Cs from its most favourable trapping site is limited by the uranium vacancy diffusion mechanism. We also considered the formation of the main solid phases of caesium resulting from its oxidation (Cs 2 O, Cs 2 O 2 , CsO 2 ) and from its interaction with the fuel (Cs 2 UO 4 ), with molybdenum (Cs 2 MoO 4 ) and with the zirconium of the clad (Cs 2 ZrO 3 ), since the formation of such phases, their solubility and their interdependence will affect the release of caesium. (author)

  12. 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)

  13. A method of surface area measurement of fuel materials by fission gas release at low temperature

    International Nuclear Information System (INIS)

    Kaimal, K.N.G.; Naik, M.C.; Paul, A.R.; Venkateswarlu, K.S.

    1989-01-01

    The present report deals with the development of a method for surface area measurement of nuclear fuel as well as fissile doped materials by fission gas release study at low temperature. The method is based on the evaluation of knock-out release rate of fission 133 Xe from irradiated fuel after sufficient cooling to decay the short lived activity. The report also describes the fabrication of an ampoule breaker unit for such study. Knock-out release rate of 133 Xe has been studied from UO 2 powders having varying surface area 'S' ranging from 270 cm 2 /gm to 4100 cm 2 /gm at two fissioning rates 10 12 f/cm 3 . sec. and 3.2x10 10 f/cm.sec. A relation between K and A has been established and discussed in this report. (author). 6 refs

  14. Methodology of long term behaviour study of containment materials

    International Nuclear Information System (INIS)

    Vernaz, E.; Godon, N.

    1994-01-01

    Here is the presentation of the papers shown in the colloquium on environment and ceramics; the Atomic Energy Commissariat (Cea) have been working for fifteen years on the long term behaviour of fission products glasses on very long periods, about several millions years. The method of studies is detailed. 2 refs

  15. Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems

    International Nuclear Information System (INIS)

    Was, Gary S.

    2007-01-01

    The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems

  16. Comparison of material irradiation conditions for fusion, spallation, stripping and fission neutron sources

    International Nuclear Information System (INIS)

    Vladimirov, P.; Moeslang, A.

    2004-01-01

    Selection and development of materials capable of sustaining irradiation conditions expected for a future fusion power reactor remain a big challenge for material scientists. Design of other nuclear facilities either in support of the fusion materials testing program or for other scientific purposes presents a similar problem of irradiation resistant material development. The present study is devoted to an evaluation of the irradiation conditions for IFMIF, ESS, XADS, DEMO and typical fission reactors to provide a basis for comparison of the data obtained for different material investigation programs. The results obtained confirm that no facility, except IFMIF, could fit all user requirements imposed for a facility for simulation of the fusion irradiation conditions

  17. Local behaviour of negative thermal expansion materials

    International Nuclear Information System (INIS)

    Fornasini, P.; Dalba, G.; Grisenti, R.; Purans, J.; Vaccari, M.; Rocca, F.; Sanson, A.

    2006-01-01

    EXAFS can represent a powerful probe of the local behaviour of negative thermal expansion (NTE) materials, thanks to the possibility of measuring the expansion of selected inter-atomic bonds and the perpendicular relative atomic displacements. The effectiveness of EXAFS for NTE studies is illustrated by a comparison of results recently obtained on germanium, CuCl and the cuprites Cu 2 O and Ag 2 O

  18. Water reactor fuel behaviour and fission products release in off-normal and accident conditions

    International Nuclear Information System (INIS)

    1987-09-01

    The present meeting was scheduled by the International Atomic Energy Agency upon the proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology and held at the IAEA Headquarters in Vienna from 10 to 13 November 1986. Thirty participants from 17 countries and an international organization attended the meeting. Eighteen papers were presented from 13 countries and one international organization. The meeting was composed of four sessions and covered subjects related to: physico-chemical properties of core materials under off-normal conditions, and their interactions up to and after melt-down (5 papers); core materials deformation, relocation and core coolability under (severe) accident conditions (4 papers); fission products release: including experience, mechanisms and modelling (5 papers); power plant experience (4 papers). A separate abstract was prepared for each of these 18 papers. Four working groups covering the above-mentioned topics were held to discuss the present status of the knowledge and to develop recommendations for future activities in this field. Refs, figs and tabs

  19. LASL analytical chemistry program for fissionable materials safeguards

    International Nuclear Information System (INIS)

    Jackson, D.D.; Marsh, S.F.

    1979-01-01

    Gas-solid reactions at elevated temperature, used previously to convert uranium in refractory forms to species readily soluble in acid, are being applied to thorium materials. A microgram-sensitive spectrophotometric method was developed for determining uranium and the LASL Automated Spectrophotometer has been modified to use it. The instrument now is functional for determining milligram amounts of plutonium, and milligram and microgram amounts of uranium. Construction of an automated controlled-potential-coulometric analyzer has been completed. It is giving design performance of 0.1% relative standard deviation for the determination of plutonium using a method developed especially for the instrument. A method has been developed for the microcomplexometric titration of uranium in its stable (VI) oxidation state. A color probe analyzer assembled for this titration also has been used for microcomplexometric titration of thorium. The present status of reference materials prepared for NBS and for the SALE program, as well as examples of working reference materials prepared for use with nondestructive analyzers, is given. The interlaboratory measured value of the 239 Pu half-life is 24,119 y. Just completed measurement of the half life of 241 Pu is 14.38 y. Measurement of the 240 Pu half life is in progress

  20. Multiplicity Analysis during Photon Interrogation of Fissionable Material

    International Nuclear Information System (INIS)

    Clarke, Shaun D.; Pozzi, Sara A.; Padovani, Enrico; Downar, Thomas J.

    2007-01-01

    Simulation of multiplicity distributions with the Monte Carlo method is difficult because each history is treated individually. In order to accurately model the multiplicity distribution, the intensity and time width of the interrogation pulse must be incorporated into the calculation. This behavior dictates how many photons arrive at the target essentially simultaneously. In order to model the pulse width correctly, a Monte Carlo code system consisting of modified versions of the codes MCNPX and MCNP-PoliMi has been developed in conjunction with a post-processing algorithm to operate on the MCNP-PoliMi output file. The purpose of this subroutine is to assemble the interactions into groups corresponding to the number of interactions which would occur during a given pulse. The resulting multiplicity distributions appear more realistic and capture the higher-order multiplets which are a product of multiple reactions occurring during a single accelerator pulse. Plans are underway to gather relevant experimental data to verify and validate the methodology developed and presented here. This capability will enable the simulation of a large number of materials and detector geometries. Analysis of this information will determine the feasibility of using multiplicity distributions as an identification tool for special nuclear material.

  1. Induced fission track distribution from highly radioactive particles in fallout materials

    International Nuclear Information System (INIS)

    Hashimoto, Tetsuo; Okada, Tatemichi

    1987-01-01

    Some highly radioactive fallout particles (GPs) from the 19th Chinese nuclear detonation were followed to the neutron irradiation in a reactor after sandwiched with mica detectors. The interesting star-like fission track patterns were revealed on the etched surface of the mica detectors. The simple chemical separation procedure for the GPs was applied for the separation of U and Pu as fissile elements and the both resultant fractions were examined with the similar high sensitive fission tracking detection. Subsequently, a representative track pattern from a black spherical particle was subjected to the determination of fissile nuclide content; comparing the total fission events evaluated on the basis of the numerical calculation of track densities with the total thermal neutron fluence. The results implied that the uranium is responsible for the main fissile nuclide remaining within a particle as unfissioned fractions and should be certainly enriched with respect to U-235 within such small fallout particles. This sophisticated method was also applied to determine the dead GPs, which have been highly radioactive particles just after the detonations, in the rain and snow-residual materials. Many induced star-like fission tracks verified certainly that there remains a lot of dead particles in the atmosheric environment till nowadays. (author)

  2. Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

    Science.gov (United States)

    Norman, Eric B [Oakland, CA; Prussin, Stanley G [Kensington, CA

    2009-01-27

    A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

  3. Detection of fissionable materials in cargoes using monochromatic photon radiography

    Science.gov (United States)

    Danagoulian, Areg; Lanza, Richard; O'Day, Buckley; LNSP Team

    2015-04-01

    The detection of Special Nuclear Materials (e.g. Pu and U) and nuclear devices in the commercial cargo traffic is one of the challenges posed by the threat of nuclear terrorism. Radiography and active interrogation of heavily loaded cargoes require ~ 1 - 10MeV photons for penetration. In a proof-of-concept system under development at MIT, the interrogating monochromatic photon beam is produced via a 11B(d , nγ) 12C reaction. To achieve this, a boron target is used along with the 3 MeV d+ RFQ accelerator at MIT-Bates. The reactions results in the emission of very narrow 4.4 MeV and 15.1 MeV gammas lines. The photons, after traversing the cargo, are detected by an array of NaI(Tl) detectors. A spectral analysis of the transmitted gammas allows to independently determine the areal density and the atomic number (Z) of the cargo. The proposed approach could revolutionize cargo inspection, which, in its current fielded form has to rely on simple but high dose bremsstrahlung sources. Use of monochromatic sources would significantly reduce the necessary dose and allow for better determination of the cargo's atomic number. The general methodology will be described and the preliminary results from the proof-of-concept system will be presented and discussed. Supported by NSF/DNDO Collaborative Research ARI-LA Award ECCS-1348328.

  4. 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

  5. The behaviour of transport from the fission products caesium and strontium in coated particles for high temperature reactors under irradiation conditions

    International Nuclear Information System (INIS)

    Zoller, P.

    1976-07-01

    At first survey is given about existing knowledge of the behaviour of caesium and strontium fission product transport in coated particles. In order to describe the complicated fission product transport mechanisms under irradiation conditions a suitable calculating model (SLIPPER) is taken over and modified to the special problems of an irradiation experiment. Fundamentally, the fission product transport is represented by the two contributions of diffusion and recoil, at which the diffusion is described by effective diffusion coefficients. In difference of that the possibility of a two-phase-diffusion is examined for the Cs diffusion in the fuel kernel. The model application on measuring results from irradiation experiments of KFA-Juelich and Mol-Belgien allowed the explanation from the characteristic of fission product transport in coated particles under irradiation conditions and produced effective diffusion coefficients for the fission products Cs and Sr. (orig.) [de

  6. Impact of material thicknesses on fission observables obtained with the FALSTAFF experimental setup

    Directory of Open Access Journals (Sweden)

    Thulliez L.

    2017-01-01

    Full Text Available In the past years, the fission studies have been mainly focused on thermal fission because most of the current nuclear reactors work in this energy domain. With the development of GEN-IV reactor concepts, mainly working in the fast energy domain, new nuclear data are needed. The FALSTAFF spectrometer under development at CEA-Saclay, France, is a two-arm spectrometer which will provide mass yields before (2V method and after (EV method neutron evaporation and consequently will have access to the neutron multiplicity as a function of mass. The axial ionization chamber, in addition to the kinetic energy value, will measure the energy loss profile of the fragment along its track. This energy loss profile will give information about the fragment nuclear charge. This paper will focus on recent developments on the FALSTAFF design. A special attention will be paid to the impact of the detector material thickness on the uncertainty of different observables.

  7. Fissionable material

    International Nuclear Information System (INIS)

    Schuuring, C.; Tuininga, E.-J.; Turkenburg, W.

    1983-01-01

    This book is a presentation of controversies surrounding nuclear energy discussions in the Netherlands and aims to show that there are serious arguments against nuclear energy. Chapters on the following topics are included: the various dimensions of the energy discussion, the background to the existence of controversies in the nuclear energy discussion, the relation between nuclear energy and prosperity, different opinions concerning the cost of producing electricity from nuclear energy, radioactive waste, the consequences of a large scale accident, the relation between the peaceful use of nuclear energy and the proliferation of nuclear weapons, the effects of low radiation doses, the relation between nuclear energy and developments in the Third World, the effect of nuclear energy on democracy. The authors of these chapters, themselves critical of nuclear energy, have described the particular controversy and have given the viewpoints of both advocates and adversaries, followed by their own opinion. The conclusions from each chapter are recapitulated in a summary and the various components of the nuclear energy cycle are presented in an appendix. (C.F.)

  8. Thermally and Chemically responsive nanoporous materials for efficient capture of fission product gases.

    Energy Technology Data Exchange (ETDEWEB)

    Stroeve, Pieter; Faller, Roland

    2018-04-24

    The objective of this project was to develop robust, high-efficiency materials for capture of fission product gases such as He, Xe and Kr in scenarios relevant for both reactor fuels and reprocessing operations. The relevant environments are extremely harsh, encompassing temperatures up to 1500 °C, high levels of radiation, as well as potential exposures to highly-reactive chemicals such as nitric acid and organic solvents such as kerosene. The requirement for nanostructured capture materials is driven in part by the very short (few micron) diffusion distances for product gases in nuclear fuel.1-2 We achieved synthesis, characterization and detailed modeling of the materials. Although not all materials reviewed in this report will be feasible for the ultimate goal of integration in nuclear fuel, nevertheless each material studied has particular properties which will enable an optimized material to be efficiently developed and characterized.

  9. The opportunity to limit and reduce inventories of fissionable weapon materials

    International Nuclear Information System (INIS)

    Hebel, L.C.

    1991-01-01

    As the United States and the Soviet government agree on major reductions in nuclear weapon delivery systems, they need to address the disposal of the nuclear warheads and bombs for those systems. Such measures could be strongly reinforced if the two nations also institute restraints and reductions in the total amount of fissionable materials available for weapons. Many metric tonnes of such materials would be made surplus by the reductions in strategic nuclear weapons due to the Strategic Arms Reduction Treaty (START-I). Equally large reductions in short-range (theater) nuclear weapons are expected in the wake of the recent Treaty on Conventional Forces in Europe (CFE)

  10. 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

  11. Technology for Fissionable Materials Detection by Use of 100 MeV Variable Linac

    CERN Document Server

    Karasyov, Sergey P; Dovbnja, Anatoliy N; Eran, L; Kiryukhin, Nikolay M; Melnik, Yu M; Ran'iuk, Yu; Shlyakhov, Il'ya N; Trubnikov, Sergiy V

    2005-01-01

    A new concept for a two-step facility to increase the accuracy/reliability of detecting heavily shielded fissionable materials (FM) in marine containers is presented. The facility will detect FM in two steps. An existing dual-view; dual-energy X-ray scanner, which is based on 7 MeV electron accelerator, will select the suspicious places inside container. The linac with variable energy (up to 100 MeV) will be used for the second step. The technology will detect fissionable nuclei by gamma induced fission reactions and delayed neutron registration. A little-known Ukrainian experimental data obtained in Chernobil' clean-up program will be presented to ground proposed concept. The theoretical calculations of neutron fluxes scale these results to marine container size. Modified GEANT code for electron/gamma penetration and authors' own software for neutron yield/penetration are used for these calculations. Available facilities (X-ray scanners; linac; detectors), which will be used for concept proof, are described....

  12. Helium and fission gas behaviour in magnesium aluminate spinel and zirconia for actinide transmutation

    NARCIS (Netherlands)

    Damen, P.M.G.

    2003-01-01

    In order to reduce the long-term radiotoxicity of spent nuclear fuel, many studies are performed on partitioning and transmutation of actinides. In such a scenario, the long-lived radio-isotopes (mostly actinides) are partitioned from the nuclear waste, and subsequently transmuted or fissioned in a

  13. Asymmetric fission and evaporation of Cr+60 (r = 2-4) fullerene ions in ion-C60 collisions: III. Universal behaviour of fission

    International Nuclear Information System (INIS)

    Bordenave-Montesquieu, D; Bordenave-Montesquieu, A; Rentenier, A; Moretto-Capelle, P

    2005-01-01

    The behaviour of the asymmetrical fission (AF) scheme (correlated ion distributions) against the collision conditions is investigated using H + x (x = 1-3) and He + projectiles in the 1-130 keV collision energy range. The present work is an extension of our recent publications on this topic using 11 keV protons (Rentenier et al 2004 J. Phys. B: At. Mol. Opt. Phys. 37 2429 and 2455). The threshold for AF is observed at 2 keV proton energy corresponding to a maximum deposited energy equal to about 41 eV. The main result concerns the fragment distributions resulting from AF of C r+ 60 ions, and secondary dissociation of even-n C + n fragments, which are both found to remain independent of the projectile species and collision velocity. These findings indicate that they are insensitive to the internal energy distributions of the parent ions. In addition, a contribution of binary collisions between the projectile and individual carbon atoms of the C 60 molecule to AF is identified in the C + 1 production at the lowest collision velocities, the so-called impulsive fragmentation

  14. Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors (Workshop Report)

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, RE

    2004-07-15

    The ''Workshop on Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors'' was convened to determine the degree to which an increased effort in modeling and simulation could help bridge the gap between the data that is needed to support the implementation of these advanced nuclear technologies and the data that can be obtained in available experimental facilities. The need to develop materials capable of performing in the severe operating environments expected in fusion and fission (Generation IV) reactors represents a significant challenge in materials science. There is a range of potential Gen-IV fission reactor design concepts and each concept has its own unique demands. Improved economic performance is a major goal of the Gen-IV designs. As a result, most designs call for significantly higher operating temperatures than the current generation of LWRs to obtain higher thermal efficiency. In many cases, the desired operating temperatures rule out the use of the structural alloys employed today. The very high operating temperature (up to 1000 C) associated with the NGNP is a prime example of an attractive new system that will require the development of new structural materials. Fusion power plants represent an even greater challenge to structural materials development and application. The operating temperatures, neutron exposure levels and thermo-mechanical stresses are comparable to or greater than those for proposed Gen-IV fission reactors. In addition, the transmutation products created in the structural materials by the high energy neutrons produced in the DT plasma can profoundly influence the microstructural evolution and mechanical behavior of these materials. Although the workshop addressed issues relevant to both Gen-IV and fusion reactor materials, much of the discussion focused on fusion; the same focus is reflected in this report. Most of the physical models and computational methods

  15. Simple and effective method of determining multiplicity distribution law of neutrons emitted by fissionable material with significant self -multiplication effect

    International Nuclear Information System (INIS)

    Yanjushkin, V.A.

    1991-01-01

    At developing new methods of non-destructive determination of plutonium full mass in nuclear materials and products being involved in uranium -plutonium fuel cycle by its intrinsic neutron radiation, it may be useful to know not only separate moments but the multiplicity distribution law itself of neutron leaving this material surface using the following as parameters - firstly, unconditional multiplicity distribution laws of neutrons formed in spontaneous and induced fission acts of the given fissionable material corresponding nuclei and unconditional multiplicity distribution law of neutrons caused by (α,n) reactions at light nuclei of some elements which compose this material chemical structure; -secondly, probability of induced fission of this material nuclei by an incident neutron of any nature formed during the previous fissions or(α,n) reactions. An attempt to develop similar theory has been undertaken. Here the author proposes his approach to this problem. The main advantage of this approach, to our mind, consists in its mathematical simplicity and easy realization at the computer. In principle, the given model guarantees any good accuracy at any real value of induced fission probability without limitations dealing with physico-chemical composition of nuclear material

  16. Critical survey of the neutron-induced creep behaviour of steel alloys for the fusion reactor materials programme

    International Nuclear Information System (INIS)

    Hausen, H.

    1985-01-01

    The differences between the irradiation environment of a fission reactor and that of a fusion reactor are respectively described in relation to the radiation damage found and expected in the two types of nuclear reactor. It is shown that the microstructure developing for instance in stainless steel alloys is almost invariant to whether the production rate of helium is high or low. The finding is valid up to neutron doses corresponding to about 60 dpa. For this reason, irradiation creep data obtained in fission reactors may be used, with caution, for predicting creep behaviour in fusion reactors.It was further recognized that irradiation creep performed with high energy particles from an accelerator, yields results which are comparable to those obtained in fission reactors. For this reason, simulation creep experiments are found to be valuable for the development of irradiation creep resistant materials using, for example, high energy electrons or protons. Such kind of experiments are performed in many laboratories. For irradiation doses larger than 60 dpa, predictions with respect to creep rates in fission and fusion reactors are difficult. In end-of-life tests, which concern swelling, ductility, tensile properties, rupture, fatigue and embrittlement, the presence of helium, due to its production rate being much higher in most materials exposed to 14 MeV neutrons than to fission neutrons, may be of great importance

  17. Study on behaviour in long term of vitrified materials

    International Nuclear Information System (INIS)

    Vernaz, E.

    1993-01-01

    In collaboration with EDF (Electricite de France), after testing fusion of Refiom (Residus d'Epuration des Fumees d'Incineration d'Ordures Menageres), residues from purification of incineration smokes of household rubbish, realised at Porcheville and at the Laboratory of Renardieres with experimental processing of vitrification by plasma, CEA (Centre d'Etudes Atomiques), atomic center of research, began study on resistance in long term of vitrified products. From about thirty five years, CEA carries out research to confine radioactive waste of high activity in stable materials. Glass was the first best one which allowed to incorporate about thirty different chemical elements found in fission products solutions into a stable die with a good chemical durability; three vitrification shops raised, one at Marcoule ('AVM', 1978) in the south of France, the two other ones at La Hague ('R7', 1989 and 'T7', 1992) in Normandy. To determine a possible impact of a deep radioactive waste disposal on human and environment, several studies began. In particular, studies on aqueous corrosion of glasses to determine behaviour in long term of glass package (first barrier of confinement) and to estimate kinetics of releasing confined toxical elements on periods of several thousands years. Principal results are exposed in this conference. Experience shows that safety analysis cannot be based on long term extrapolation of a simple lixiviation result. This analysis must include: a sufficient knowledge in basic mechanisms of alteration to predict the kinetic evolution in a long term. To take in account environment conditions with a normal or accidental scheme (acidity, clay, organic compounds,...). This knowledge broadly developed by CEA for nuclear glasses seems to be easily transposable to different wastes (industrial ones or from hospitals) and takes place in a contract of research CEA/EDF to valorize vitrified products. 9 figs. 4 refs

  18. Behaviour of fission products 90Sr, 137Cs and 144Ce in soil-plant system

    International Nuclear Information System (INIS)

    Zhu Yongyi; Qiu tongcai

    1988-11-01

    A small quantity of radioonuclides, such as fission products 90 Sr, 137 Cs and 144 Ce etc., generally may leak out from nuclear inductry system and may be disseminated on soul and plant cover. The accumulation and distribution of the radionuclides in spring wheat planted in the contaminated soil are described. The factors as nuclide chemical forms, soil agrochemical properties, growing stages of the plant and fertilizing etc., which affect the accumulation and the distribution were discussed. Possible approches were supposed to eliminate or clean the radionuclides from contaminated soil, which include planting adaptable herbage, applying some fertilizers and scraping regolith etc

  19. CSER 00-008 use of PFP Glovebox HC-18BS for Storage and Transport of Fissionable Material

    International Nuclear Information System (INIS)

    ERICKSON, D.G.

    2000-01-01

    This CSER addresses the feasibility of increasing the allowed number of open containers and permitting the transfer and storage of fissionable material in Glovebox HC-18BS without regard to form or density (metal, oxide having an H/X (le) 20, material having unrestricted moderation and plutonium hydroxide having a plutonium density of 0.2 g/cm 3 )

  20. Decree of 8 October 1969, Stb. 471, concerning the implementation of Sections 13 and 14 of the Nuclear Energy Act (Fissionable Materials and Ores (Registration))

    International Nuclear Information System (INIS)

    1969-01-01

    This Decree lays down the system for registration and notification of fissionable materials and ores in accordance with the Nuclear Energy Act. The register must list the quantities of fissionable materials and ores available in the Netherlands and their location. This procedure applies only to materials and ores subject to licensing. (NEA) [fr

  1. Contribution to the study of the behaviour, in the urban environment, during the runoff of rainwater, of the fission products emitted during a nuclear accident; Contribution a l`etude du devenir, en milieu urbain, pendant le ruissellement des eaux pluviales, des produits de fission emis en cas d`accident nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Pioch, M

    1993-05-24

    In the context of research into the environmental consequences of a serious accident occurring on a pressurized water reactor, this paper concerns the experimental study of behaviour of five fission products (caesium, strontium, iodine, ruthenium and tellurium) in the urban environment under the action of rainwater. Stable or radioactive multiple-element aerosols were produced. Their physicochemical characteristics and their solubility in rainwater were studied. Caesium and rubidium forms solutions totally and quickly, while strontium is partially soluble (approximately 50 %) and iodine is only slightly soluble. The behaviour of fission products on five urban surfaces was then studied. Batch experiments showed that the retention of dissolved forms of radioelements varied according to the material. The reactions involved are ion exchange reactions. The presence of certain ions in water (in particular NH{sub 4}{sup +}) increase the desorption of radioelements. Using a laboratory rainfall simulator, the re-entrainment of fission products by rainwater was examined. Two modes of deposition and two intensities of rainfall were simulated. The desorption of radioelements is greater after wet deposition and remobilization is reduced by an increase in intensity of rainfall. An addition of NH{sub 4}{sup +} in water is especially effective in the case of wet depositions. Suggestions are made in order to improve experimental protocols and continue the research. (author). 75 refs., 51 figs., 69 tabs., 14 appends.

  2. Joint ICFRM-14 (14. international conference on fusion reactor materials) and IAEA satellite meeting on cross-cutting issues of structural materials for fusion and fission applications. PowerPoint presentations

    International Nuclear Information System (INIS)

    2009-01-01

    The Conference was devoted to the challenges in the development of new materials for advanced fission, fusion and hybrid reactors. The topics discussed include fuels and materials research under the high neutron fluence; post-irradiation examination; development of radiation resistant structural materials utilizing fission research reactors; core materials development for the advanced fuel cycle initiative; qualification of structural materials for fission and fusion reactor systems; application of charged particle accelerators for radiation resistance investigations of fission and fusion structural materials; microstructure evolution in structural materials under irradiation; ion beams and ion accelerators

  3. An integrated circuit/microsystem/nano-enhanced four species radiation sensor for inexpensive fissionable material detection

    Science.gov (United States)

    Waguespack, Randy Paul

    2011-12-01

    Small scale radiation detectors sensitive to alpha, beta, electromagnetic, neutron radiation are needed to combat the threat of nuclear terrorism and maintain national security. There are many types of radiation detectors on the market, and the type of detector chosen is usually determined by the type of particle to be detected. In the case of fissionable material, an ideal detector needs to detect all four types of radiation, which is not the focus of many detectors. For fissionable materials, the two main types of radiation that must be detected are gamma rays and neutrons. Our detector uses a glass or quartz scintillator doped with 10B nanoparticles to detect all four types of radiation particles. Boron-10 has a thermal neutron cross section of 3,840 barns. The interaction between the neutron and boron results in a secondary charge particle in the form of an alpha particle to be emitted, which is detectable by the scintillator. Radiation impinging on the scintillator matrix produces varying optical pulses dependent on the energy of the particles. The optical pulses are then detected by a photomultiplier (PM) tube, creating a current proportional to the energy of the particle. Current pulses from the PM tube are differentiated by on-chip pulse height spectroscopy, allowing for source discrimination. The pulse height circuitry has been fabricated with discrete circuits and designed into an integrated circuit package. The ability to replace traditional PM tubes with a smaller, less expensive photomultiplier will further reduce the size of the device and enhance the cost effectiveness and portability of the detector.

  4. 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.

  5. Materials-of-Construction Radiation Sensitivity for a Fission Surface Power Convertor

    Science.gov (United States)

    Bowman, Cheryl L.; Geng, Steven M.; Niedra, Janis M.; Sayir, Ali; Shin, Eugene E.; Sutter, James K.; Thieme, Lanny G.

    2007-01-01

    A fission reactor combined with a free-piston Stirling convertor is one of many credible approaches for producing electrical power in space applications. This study assumes dual-opposed free-piston Stirling engines/linear alternators that will operate nominally at 825 K hot-end and 425 K cold-end temperatures. The baseline design options, temperature profiles, and materials of construction discussed here are based on historical designs as well as modern convertors operating at lower power levels. This notional design indicates convertors primarily made of metallic components that experience minimal change in mechanical properties for fast neutron fluences less than 10(sup 20) neutrons per square centimeter. However, these radiation effects can impact the magnetic and electrical properties of metals at much lower fluences than are crucial for mechanical property integrity. Moreover, a variety of polymeric materials are also used in common free-piston Stirling designs for bonding, seals, lubrication, insulation and others. Polymers can be affected adversely by radiation doses as low as 10(sup 5) - 10(sup 10) rad. Additionally, the absorbing dose rate, radiation hardness, and the resulting effect (either hardening or softening) varies depending on the nature of the particular polymer. The classes of polymers currently used in convertor fabrication are discussed along possible substitution options. Thus, the materials of construction of prototypic Stirling convertor engines have been considered and the component materials susceptible to damage at the lowest neutron fluences have been identified.

  6. Current status of modeling fission gas behaviour in the Karlsruhe code LANGZEIT/KURZZEIT

    International Nuclear Information System (INIS)

    Vaeth, L.

    1980-12-01

    The programme LANGZEIT/KURZZEIT has been recently extended to describe intragranular bubble coalescence and volume equilibration, to model intergranular gas behaviour and transient release from closed porosity. The model is described and the results of some comparisons with transient experiments are discussed. Further necessary refinements of the model are outlined. (orig.) [de

  7. Survey of Materials for Fusion Fission Hybrid Reactors Vol 1 Rev. 0

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, Joseph Collin [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Chemistry Materials and Life Sciences Directorate

    2007-07-03

    Materials for fusion-fission hybrid reactors fall into several broad categories, including fuels, blanket and coolant materials, cladding, structural materials, shielding, and in the specific case of inertial-confinement fusion systems, laser and optical materials. This report surveys materials in all categories of materials except for those required for lasers and optics. Preferred collants include two molten salt mixtures known as FLIBE (Li2BeF4) and FLINABE (LiNaBeF4). In the case of homogenous liquid fuels, UF4 can be dissolved in these molten salt mixtures. The transmutation of lithium in this coolant produces very corrosive hydrofluoric acid species (HF and TF), which can rapidly degrade structural materials. Broad ranges of high-melting radiation-tolerant structural material have been proposed for fusion-fission reactor structures. These include a wide variety of steels and refractory alloys. Ferritic steels with oxide-dispersion strengthening and graphite have been given particular attention. Refractory metals are found in Groups IVB and VB of the periodic table, and include Nb, Ta, Cr, Mo, and W, as serve as the basis of refractory alloys. Stable high-melting composites and amorphous metals may also be useful. Since amorphous metals have no lattice structure, neutron bombardment cannot dislodge atoms from lattice sites, and the materials would be immune from this specific mode of degradation. The free energy of formation of fluorides of the alloying elements found in steels and refractory alloys can be used to determine the relative stability of these materials in molten salts. The reduction of lithium transmutation products (H+ and T+) drives the electrochemical corrosion process, and liberates aggressive fluoride ions that pair with ions formed from dissolved structural materials. Corrosion can be suppressed through the use of metallic Be and Li, though the molten salt becomes laden with colloidal suspensions of Be and Li corrosion

  8. THAI test facility for experimental research on hydrogen and fission product behaviour in light water reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, S., E-mail: gupta@becker-technologies.com [Becker Technologies GmbH, Koelner Strasse 6, 65760 Eschborn (Germany); Schmidt, E.; Laufenberg, B. von; Freitag, M.; Poss, G. [Becker Technologies GmbH, Koelner Strasse 6, 65760 Eschborn (Germany); Funke, F. [AREVA GmbH, P.O. Box 1109, 91001 Erlangen (Germany); Weber, G. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Forschungszentrum, Boltzmannstraße 14, 85748 Garching (Germany)

    2015-12-01

    hydraulics, gas distribution) and ISP-49 (hydrogen combustion), EU-SARNET/SARNET2 code-benchmark exercises involving THAI data on iodine/surface interactions, iodine mass transfer, passive autocatalytic recombiner performance, iodine oxide behaviour and iodine transport in multi-compartment behaviour. The present paper provides an overview of the THAI experiments related to hydrogen and fission products issues performed in the frame of national and international projects. From the comprehensive THAI experimental database, a selection of typical results is presented to illustrate the multi-functionality of the THAI facility and the broad variety of the experimental investigations.

  9. Mechanical Behaviour of Materials Volume 1 Micro- and Macroscopic Constitutive Behaviour

    CERN Document Server

    François, Dominique; Zaoui, André

    2012-01-01

    Advances in technology are demanding ever-increasing mastery over the materials being used: the challenge is to gain a better understanding of their behaviour, and more particularly of the relations between their microstructure and their macroscopic properties.   This work, of which this is the first volume, aims to provide the means by which this challenge may be met. Starting from the mechanics of deformation, it develops the laws governing macroscopic behaviour – expressed as the constitutive equations – always taking account of the physical phenomena which underlie rheological behaviour. The most recent developments are presented, in particular those concerning heterogeneous materials such as metallic alloys, polymers and composites. Each chapter is devoted to one of the major classes of material behaviour.   As the subtitles indicate, Volume 1 deals with micro- and macroscopic constitutive behaviour and Volume 2 with damage and fracture mechanics. A third volume will be devoted to exercises and the...

  10. Fission and corrosion product behaviour in liquid metal fast breeder reactors (LMFBRs)

    International Nuclear Information System (INIS)

    1993-02-01

    It is intended that this review will be useful not only to scientists but also to those concerned with design, day-to-day operation of plant, with liquid metal fast breeder reactors (LMFBRs), safety and decommissioning. Because of this, the review has been widened to include not only the mass transfer behaviour of the various radionuclides in experimental and operating systems, but also the monitoring of the various species, the methods of measurement and the development of methods to control the build-up of the more important long half-life species in operating plants. The information used in the review has been taken from open literature sources to provide an up-to-date presentation of the behaviour of the various isotopes in LMFBRs. 172 refs, 14 figs, 22 tabs

  11. The behaviour of materials in fast reactors

    International Nuclear Information System (INIS)

    Matthews, J.R.

    1977-01-01

    Fast neutron damage in fast reactors can limit the life of structural components through the growth voids. The main features of the current theory of point defect production and condensation are surveyed. The role of metallurgical structures and radiation produced extended defects is outlined and used to demonstrate the development of volume swelling and radiation hardening. Mechanisms of radiation creep are described in the context of the preceding treatment of point defect behaviour. Finally, future trends in the field are briefly explored. (author)

  12. Gas turbine blades and disks. Materials and component behaviour

    International Nuclear Information System (INIS)

    1990-01-01

    This progress report summarizes the research results obtained by the special research programme 339 in the years 1988 and 1989. Emphasis is given to the following aspects and problems: Optimisation of structure, protective coatings, connection between structure parameters and mechanical materials behaviour, tribologic materials and component behaviour, impacts of overall loads, and of stress and deformation state in the inelastic regime under mechanical and thermal load, and impacts of the manufacturing process on component behaviour, quality assurance. Eleven of the fifteen papers of the report have been separately analysed for the ENERGY database, and thirteen for the DELURA database. (orig./MM) With 191 figs., 13 tabs [de

  13. The experience of Russian Federation in organization of customs control of fissionable and other radioactive materials

    International Nuclear Information System (INIS)

    Podchishaev, A.

    2001-01-01

    Among the routine inspection tasks of customs offices are tasks stemming from international commitments of Russia to prevent proliferation of nuclear weapons and material that can be used for making these weapons. These tasks are: radiation monitoring of all vehicles, passengers, their luggage and goods crossing the state border; inspection of fissionable and radioactive materials (FRM) legally transported by participants in the foreign trade activities with a view to checking that the declared data fully correspond to the actual radioactive cargo. Organizational measures involve the Sheremetyevo customs office has a department whose personnel is specially trained in radiation monitoring and can operate radiometric and spectrometric instruments. These specialists are included in shifts on duty responsible for customs clearing and inspection and carry out continuous radiation monitoring of passengers and their luggage, vehicles and goods crossing the border. They work on the 24-hour basis, which allows quickly and skillfully localizing the detected radiation source and avoiding direct contact of customs, officers, airport personnel, and passengers with the radioactive item. Technical measures include provision and everyday use of radiation monitoring instrumentation, classified as: stationary equipment of primary radiation monitoring (SEPRM); hand-held instruments for additional radiation monitoring (RM); spectrometric equipment for control of legal FRM transport. The customs procedure for monitoring of fissionable and radioactive materials is divided into three stages. Stage I, primary RM is carried out by stationary FRM detection systems Yantar for customs applications installed on the customs inspection line next to the X-ray inspection equipment (XIE). Stage II, additional RM is carried out by officer who uses hand-held instruments to check the passenger's luggage for surface contamination; to perform primary identification of the detected radioactive source

  14. Fission meter

    Science.gov (United States)

    Rowland, Mark S [Alamo, CA; Snyderman, Neal J [Berkeley, CA

    2012-04-10

    A neutron detector system for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly into information that a first responder can use to discriminate materials. The system comprises counting neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source.

  15. Influence of transmutation and high neutron exposure on materials used in fission-fusion correlation experiments

    International Nuclear Information System (INIS)

    Garner, F.A.

    1990-07-01

    This paper explores the response of three different materials to high fluence irradiation as observed in recent fusion-related experiments. While helium at fusion-relevant levels influences the details of the microstructure of Fe--Cr--Ni alloys somewhat, the resultant changes in swelling and tensile behavior are relatively small. Under conditions where substantially greater-than-fusion levels of helium are generated, however, an extensive refinement of microstructure can occur, leading to depression of swelling at lower temperatures and increased strengthening at all temperatures studied. The behavior of these alloys is dominated by their tendency to converge to saturation microstructures which encourage swelling. Irradiations of nickel are dominated by its tendency to develop a different type of saturation microstructure that discourages further void growth. Swelling approaches saturation levels that are remarkably insensitive to starting microstructure and irradiation temperature. The rate of approach to saturation is very sensitive to variables such as helium, impurities, dislocation density and displacement rate, however. Copper exhibits a rather divergent response depending on the property measured. Transmutation of copper to nickel and zinc plays a large role in determining electrical conductivity but almost no role in void swelling. Each of these three materials offers different challenges in the interpretation of fission-fusion correlation experiments

  16. 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

  17. Synthesis of Actinide Materials for the Study of Basic Actinide Science and Rapid Separation of Fission Products

    Energy Technology Data Exchange (ETDEWEB)

    Dorhout, Jacquelyn Marie [Univ. of Nevada, Las Vegas, NV (United States)

    2017-11-28

    This dissertation covers several distinct projects relating to the fields of nuclear forensics and basic actinide science. Post-detonation nuclear forensics, in particular, the study of fission products resulting from a nuclear device to determine device attributes and information, often depends on the comparison of fission products to a library of known ratios. The expansion of this library is imperative as technology advances. Rapid separation of fission products from a target material, without the need to dissolve the target, is an important technique to develop to improve the library and provide a means to develop samples and standards for testing separations. Several materials were studied as a proof-of-concept that fission products can be extracted from a solid target, including microparticulate (< 10 μm diameter) dUO2, porous metal organic frameworks (MOFs) synthesized from depleted uranium (dU), and other organicbased frameworks containing dU. The targets were irradiated with fast neutrons from one of two different neutron sources, contacted with dilute acids to facilitate the separation of fission products, and analyzed via gamma spectroscopy for separation yields. The results indicate that smaller particle sizes of dUO2 in contact with the secondary matrix KBr yield higher separation yields than particles without a secondary matrix. It was also discovered that using 0.1 M HNO3 as a contact acid leads to the dissolution of the target material. Lower concentrations of acid were used for future experiments. In the case of the MOFs, a larger pore size in the framework leads to higher separation yields when contacted with 0.01 M HNO3. Different types of frameworks also yield different results.

  18. Fission products and nuclear fuel behaviour under severe accident conditions part 2: Fuel behaviour in the VERDON-1 sample

    Science.gov (United States)

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

    2017-11-01

    Within the framework of the International Source Term Programme (ISTP), the VERDON programme aims at quantifying the source term of radioactive materials in case of a hypothetical severe accident in a light water reactor (LWR). Tests were performed in a new experimental laboratory (VERDON) built in the LECA-STAR facility (CEA Cadarache). The VERDON-1 test was devoted to the study of a high burn-up UO2 fuel and FP releases at very high temperature (≈2873 K) in a reducing atmosphere. Post-test qualitative and quantitative characterisations of the VERDON-1 sample led to the proposal of a scenario explaining the phenomena occurring during the experimental sequence. Hence, the fuel and the cladding may have interacted which led to the melting of UO2-ZrO2 alloy. Although no relocation was observed during the test, it may have been imminent.

  19. Power deposition distribution in liquid lead cooled fission reactors and effects on the reactor thermal behaviour

    International Nuclear Information System (INIS)

    Cevolani, S.; Nava, E.; Burn, K. W.

    2001-01-01

    In the framework of an ADS study (Accelerator Driven System, a reactor cooled by a lead bismuth alloy) the distribution of the deposited energy between the fuel, coolant and structural materials was evaluated by means of Monte Carlo calculations. The energy deposition in the coolant turned out to be about four percent of the total deposited energy. In order to study this effect, further calculations were performed on water and sodium cooled reactors. Such an analysis showed, for both coolant materials, a much lower heat deposition, about one percent. Based on such results, a thermohydraulic analysis was performed in order to verify the effect of this phenomenon on the fuel assembly temperature distribution. The main effect of a significant fraction of energy deposition in the coolant is concerned with the decrease of the fuel pellet temperature. As a consequence, taking into account this effect allows to increase the possibilities of optimization at the disposal of the designer [it

  20. Behaviour of organic materials in radiation environment

    CERN Document Server

    Tavlet, M

    2000-01-01

    Radiation effects in polymers are reminded together with the ageing factors. Radiation-ageing results are mainly discussed about thermosetting insulators, structural composites and cable-insulating materials. Some hints are given about high-voltage insulations, cooling fluids, organic scintillators and light-guides. Some parameters to be taken into account for the estimate of the lifetime of components in radiation environment are also shown. (23 refs).

  1. Feynman variance for neutrons emitted from photo-fission initiated fission chains - a systematic simulation for selected speacal nuclear materials

    Energy Technology Data Exchange (ETDEWEB)

    Soltz, R. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Danagoulian, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sheets, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Korbly, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hartouni, E. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2013-05-22

    Theoretical calculations indicate that the value of the Feynman variance, Y2F for the emitted distribution of neutrons from ssionable exhibits a strong monotonic de- pendence on a the multiplication, M, of a quantity of special nuclear material. In 2012 we performed a series of measurements at the Passport Inc. facility using a 9- MeV bremsstrahlung CW beam of photons incident on small quantities of uranium with liquid scintillator detectors. For the set of objects studies we observed deviations in the expected monotonic dependence, and these deviations were later con rmed by MCNP simulations. In this report, we modify the theory to account for the contri- bution from the initial photo- ssion and benchmark the new theory with a series of MCNP simulations on DU, LEU, and HEU objects spanning a wide range of masses and multiplication values.

  2. Behaviour of organic materials in radiation environment

    International Nuclear Information System (INIS)

    Tavlet, M.; Ilie, S.

    1999-01-01

    An extensive radiation damage test program has been carried out in CERN for decades and many results have yet been published. Over the years, EPR/EPDM-based rubbers and polyolefin-based compounds used for cable insulation have been tested. Polyolefin-based compounds usually present an important dose-rate effect. This is related to the presence of oxygen, it may be combined with a temperature effect. On the other hand, it appears from many results that the degradation of cable insulations does not depend on the radiation type. Tests of insulating and structural materials after irradiation at cryogenic temperature have shown that there is no significant influence of the irradiation temperature on the radiation degradation of thermo-sets and composites, while the degradation of plastic films is even less severe as they are protected against oxidation. Some experiments about the synergy between irradiation and mechanical stress have shown that rubbers and composites under stress are more sensitive to radiation and degrade faster. Very strong synergetic effects between radiation and other parameters are observed in organic optical materials such as scintillators and optical fibres. For fluorocarbon cooling fluids, a special care must be paid to alkanes and hydro-fluoro-alkanes, which are usually present as impurities, and of which the C-H bonds content opens the way to the reactive hydrofluoric acid evolution during the radiolytic process

  3. Fission-product behaviour in irradiated TRISO-coated particles: Results of the HFR-EU1bis experiment and their interpretation

    International Nuclear Information System (INIS)

    Barrachin, M.; Dubourg, R.; Groot, S. de; Kissane, M.P.; Bakker, K.

    2011-01-01

    Highlights: → The microstructure and FPs in UO 2 TRISO particles (10% FIMA, 1573 K) were studied. → Very large porosities (>10 μm) were observed in the high temperature particles. → Significant Xe and Cs releases from the kernel were observed. → Mo and Ru are mainly present in the metallic precipitates in the kernel. - Abstract: It is important to understand fission-product (FP) and kernel micro-structure evolution in TRISO-coated fuel particles. FP behaviour, while central to severe-accident evaluation, impacts: evolution of the kernel oxygen potential governing in turn carbon oxidation (amoeba effect and pressurization); particle pressurization through fission-gas release from the kernel; and coating mechanical resistance via reaction with some FPs (Pd, Cs, Sr). The HFR-Eu1bis experiment irradiated five HTR fuel pebbles containing TRISO-coated UO 2 particles and went beyond current HTR specifications (e.g., central temperature of 1523 K). This study presents ceramographic and EPMA examinations of irradiated urania kernels and coatings. Significant evolutions of the kernel (grain structure, porosity, metallic-inclusion size, intergranular bubbles) as a function of temperature are shown. Results concerning FP migration are presented, e.g., significant xenon, caesium and palladium release from the kernel, molybdenum and ruthenium mainly present in metallic precipitates. The observed FP and micro-structural evolutions are interpreted and explanations proposed. The effect of high flux rate and high temperature on fission-gas behaviour, grain-size evolution and kernel swelling is discussed. Furthermore, Cs, Mo and Zr behaviour is interpreted in connection with oxygen-potential. This paper shows that combining state-of-the-art post-irradiation examination and state-of-the-art modelling fundamentally improves understanding of HTR fuel behaviour.

  4. Mechanical Behaviour of Materials Volume II Fracture Mechanics and Damage

    CERN Document Server

    François, Dominique; Zaoui, André

    2013-01-01

    Designing new structural materials, extending lifetimes and guarding against fracture in service are among the preoccupations of engineers, and to deal with these they need to have command of the mechanics of material behaviour. This ought to reflect in the training of students. In this respect, the first volume of this work deals with elastic, elastoplastic, elastoviscoplastic and viscoelastic behaviours; this second volume continues with fracture mechanics and damage, and with contact mechanics, friction and wear. As in Volume I, the treatment links the active mechanisms on the microscopic scale and the laws of macroscopic behaviour. Chapter I is an introduction to the various damage phenomena. Chapter II gives the essential of fracture mechanics. Chapter III is devoted to brittle fracture, chapter IV to ductile fracture and chapter V to the brittle-ductile transition. Chapter VI is a survey of fatigue damage. Chapter VII is devoted to hydogen embrittlement and to environment assisted cracking, chapter VIII...

  5. Fission-product behaviour during irradiation of TRISO-coated particles in the HFREU1bis experiment - HTR2008-58125

    International Nuclear Information System (INIS)

    De Groot, S.; Bakker, K.; Barrachin, M.; Dubourg, R.; Kissane, M.

    2008-01-01

    The irradiation experiment HFR-EU1bis, coordinated by the European Joint Research Centre - Inst. for Energy, was performed in the High Flux Reactor (HFR) at Petten to test five spherical HTR fuel pebbles of former German production with TRISO coated particles in conditions beyond the specifications of current HTR reactor designs (central temperature of 1250 deg. C). In this paper, the behaviour of the fission products (FPs) and kernel micro-structure evolution during the test are investigated. While FP behaviour is a key issue for potential source term evaluation it also determines the evolution of the oxygen potential in the oxide kernel which in turn is important for formation of carbon oxides (amoeba effect and pressurization). Fission-gas release from the kernel can induce additional mechanical loading and finally some FPs (Ag, Cs, Sr) might alter the mechanical integrity of the coatings. This study is based on post- irradiation examinations (ceramography + EPMA) performed both on UO 2 kernels and on coatings. Significant evolutions of the kernel as a function of temperature are shown (grain structure, porosity, size of metallic inclusions). The quality of the ceramography results allows characteristics of the intergranular bubbles in the kernel (and estimation of swelling) to be determined. Remarkable results considering FP release from the kernel have been observed and will be presented. Examples are the significant release of Cs out of the kernel as well as Pd, whereas Zr remains trapped. Mo and Ru are mainly incorporated in metallic precipitates. These observations are interpreted and mechanisms for FP and micro-structural evolutions are proposed. These results are coupled to the results of calculations performed with the mechanistic code MFPR (Module for Fission Product Release) and the thermodynamic database MEPHISTA (Multiphase Equilibria in Fuels via Standard Thermodynamic Analysis). The effect of high flux rate and high temperature on fission gas

  6. Properties of container and backfill materials for the final disposal of highly radioactive fission products

    International Nuclear Information System (INIS)

    Mirschinka, V.

    1983-11-01

    The qualifications of six metallic alloys to serve as canister materials for an in-can glass smelting process were studied. These alloys are: N 6 1.4864 (X 12NiCrSi3616, Thermax 16/36), No. 2.4816 (NiCr15Fe, Inconel 600), No. 2.4610 (Hastelloy C4), No. 2.4778 (UMCO50), No. 1.5415 (15MO3), No. 1.1005 (ZSH-Spezial). The mechanical properties of any of the six materials at high temperatures were found to be sufficient. The chemical interactions between glass and metal were investigated by glass smelting tests and electron microprobe analyses, showing that chromium as an alloying element of the crucible material may affect the quality of the glass product by causing inhomogeneities and a violent blistering in the glass matrix. The resistance against corrosion by concentrated salt solutions under elevated pressure and temperature similar to final depository conditions was tested showing that the presence of a bentonite suspension in the salt solution reduces the corrosion attack of the metal significantly. Diffusion experiments of salt solutions doted with radioactive isotopes Na-22 and Cl-36 as tracer substances were made to show the retardation behaviour of salt ions in compacted bentonite. However, a long-term barrier effect of the bentonite against salt ion diffusion could not be verified. (orig./HOE)

  7. Decree of 4 September 1969, Stb. 405, concerning the implementation of Sections 16, 19, paragraph 1, 21, 29, 30, paragraph 2 and 32 of the Nuclear Energy Act (Fissionable Materials, Ores and Radioactive Materials (Transport))

    International Nuclear Information System (INIS)

    1969-01-01

    The regulations governing the transport of fissionable materials, ores and radioactive materials are embodied in this Decree, together with the regulations concerning operations involving their movements into and out of the Netherlands and their storage incidental to transport. (NEA) [fr

  8. Study of the prompt gamma ray signal from fissions in special nuclear materials induced using an associated particle neutron generator

    International Nuclear Information System (INIS)

    Koltick, D. S.; Kane, S. Z.

    2009-01-01

    More than 42 million cargo containers entered the United States in 2005. To search for a few kilograms of special nuclear material (SNM) within this vast stream of cargo, an inspection system based on neutron-induced fission followed by the coincident detection of multiple prompt fission gamma rays is investigated using MCNP-Polimi code. The system utilizes two deuterium-tritium (DT) associated particle neutron generators, each capable of 10 9 neutrons/s at 14.1 MeV, with sub-nanosecond timing resolution ZnO:Ga alpha detectors internal to the generator. Because prompt fission signals are approximately 100 times stronger than the delayed signals, the neutron flux is greatly reduced compared to 10 11-12 neutrons/s required for systems based on delayed signals such as the 'nuclear car wash' [4]. In addition the system utilizes 30 cm deep liquid krypton (LKr) noble gas detectors having 94% detection efficiency for 1 MeV gamma rays, high solid angle coverage (∼ 50% of the total solid angle), and sub-nanosecond timing resolution (∼ 600 ps). An algorithm for distinguishing U-235 from U-238 is presented. (authors)

  9. Modelling of the high temperature behaviour of metallic materials

    International Nuclear Information System (INIS)

    Mohr, R.

    1999-01-01

    The design of components of metallic high-temperature materials by the finite element method requires the application of phenomenological viscoplastic material models. The route from the choice of a convenient model, the numerical integration of the equations and the parameter identification to the design of components is described. The Chaboche-model is used whose evolution equations are explicitly integrated. The parameters are determined by graphical and numerical methods in order to use the material model for describing the deformation behaviour of a chromium steel and an intermetallic titanium aluminide alloy. (orig.)

  10. Delayed fission

    Energy Technology Data Exchange (ETDEWEB)

    Hatsukawa, Yuichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-07-01

    Delayed fission is a nuclear decay process that couples {beta} decay and fission. In the delayed fission process, a parent nucleus undergoes {beta} decay and thereby populates excited states in the daughter. If these states are of energies comparable to or greater than the fission barrier of the daughter, then fission may compete with other decay modes of the excited states in the daughter. In this paper, mechanism and some experiments of the delayed fission will be discussed. (author)

  11. Behavioural response of Phytoseiulus persimilisin inert materials for technical application.

    Science.gov (United States)

    Wendorf, Dennis; Sermann, Helga; Katz, Peter; Lerche, Sandra; Büttner, Carmen

    2009-01-01

    A large scale application of the predatory mite Phytoseiulus persimilis Athias-Henriot for use in the biological control of spider mites in the field requires testing the behaviour of Phytoseiulus persimilis in inert materials, like millet pelts and Vermiculite (1-3 mm). In laboratory studies, the distribution of the individuals in such materials, the time of remaining in the material were proved. To examine the abiotic influences on the time of remaining in the material, the dampness of the materials was varied (0%, 5% and 10%). Moreover, the influence of attitude of materials was tested. The time of emigration from the material was noted for each individual. Emigration from all dry materials was completed 15 minutes at the latest after set up of the mites. The increase of dampness had an obvious effect on the time of remaining in the material. In this respect the material millet pelts showed the most favourable effect with 10% dampness. Increasing attitude of material the mobility of predatory mites will be influenced negatively above 75 cm. Up to 50 cm, mites have not a problem to move in the material and the time of remaining can be prolonged considerably.

  12. Fatigue life prediction of autofrettage tubes using actual material behaviour

    International Nuclear Information System (INIS)

    Jahed, Hamid; Farshi, Behrooz; Hosseini, Mohammad

    2006-01-01

    There is a profound Bauschinger effect in the behaviour of high-strength steels used in autofrettaged tubes. This has led to development of methods capable of considering experimentally obtained (actual) material behaviour in residual stress calculations. The extension of these methods to life calculations is presented here. To estimate the life of autofrettaged tubes with a longitudinal surface crack emanating from the bore more accurately, instead of using idealized models, the experimental loading-unloading stress-strain behaviour is employed. The resulting stresses are then used to calculate stress intensity factors by the weight function method as input to fatigue life determination. Fatigue lives obtained using the actual material behaviour are then compared with the results of frequently used ideal models including those considering Bauschinger effect factors and strain hardening in unloading. Using standard fatigue crack growth relationships, life of the vessel is then calculated based on recommended initial and final crack length. It is shown that the life gain due to autofrettage above 70% overstrain is considerable

  13. HEU and LEU MTR fuel elements as target materials for the production of fission molybdenum

    International Nuclear Information System (INIS)

    Sameh, A.A.; Bertram-Berg, A.

    1993-01-01

    The processing of irradiated MTR-fuels for the production of fission nuclides for nuclear medicine presents a significantly increasing task in the field of chemical separation technology of high activity levels. By far the most required product is MO-99, the mother nuclide of Tc-99m which is used in over 90% of the organ function tests in nuclear medicine. Because of the short half life of Mo-99 (66 h) the separation has to be carried out from shortly cooled neutron irradiated U-targets. The needed product purity, the extremely high radiation level, the presence of fission gases like xenon-133 and of volatile toxic isotopes such as iodine-131 and its compounds in kCi-scale require a sophisticated process technology

  14. Neutron Diffusion in a Space Lattice of Fissionable and Absorbing Materials

    Science.gov (United States)

    Feynman, R. P.; Welton, T. A.

    1946-08-27

    Methods are developed for estimating the effect on a critical assembly of fabricating it as a lattice rather than in the more simply interpreted homogeneous manner. An idealized case is discussed supposing an infinite medium in which fission, elastic scattering and absorption can occur, neutrons of only one velocity present, and the neutron m.f.p. independent of position and equal to unity with the unit of length used.

  15. Fuel elements and fuel element materials. Experimental facilities for fission products lift-off tests

    International Nuclear Information System (INIS)

    Blanchard, R.J.; Veyrat, J.F.

    1978-01-01

    One of the hypothetical accidents on the HTGR primary cooling circuits is the failure of a circuit resulting in a depressurization in the primary loops of the reactor. There is a risk of release of fission products in relation to the size of the failure. Experimental facilities for HTGR tests were developed: an in pile helium loop Comedie and an out of pile helium loop

  16. Study on fission blanket fuel cycling of a fusion-fission hybrid energy generation system

    International Nuclear Information System (INIS)

    Zhou, Z.; Yang, Y.; Xu, H.

    2011-01-01

    This paper presents a preliminary study on neutron physics characteristics of a light water cooled fission blanket for a new type subcritical fusion-fission hybrid reactor aiming at electric power generation with low technical limits of fission fuel. The major objective is to study the fission fuel cycling performance in the blanket, which may possess significant impacts on the feasibility of the new concept of fusion-fission hybrid reactor with a high energy gain (M) and tritium breeding ratio (TBR). The COUPLE2 code developed by the Institute of Nuclear and New Energy Technology of Tsinghua University is employed to simulate the neutronic behaviour in the blanket. COUPLE2 combines the particle transport code MCNPX with the fuel depletion code ORIGEN2. The code calculation results show that soft neutron spectrum can yield M > 20 while maintaining TBR >1.15 and the conversion ratio of fissile materials CR > 1 in a reasonably long refuelling cycle (>five years). The preliminary results also indicate that it is rather promising to design a high-performance light water cooled fission blanket of fusion-fission hybrid reactor for electric power generation by directly loading natural or depleted uranium if an ITER-scale tokamak fusion neutron source is achievable.

  17. Mechanical behaviour of composite materials made by resin film infusion

    Directory of Open Access Journals (Sweden)

    Casavola C.

    2010-06-01

    Full Text Available Innovative composite materials are frequently used in designing aerospace, naval and automotive components. In the typical structure of composites, multiple layers are stacked together with a particular sequence in order to give specific mechanical properties. Layers are organized with different angles, different sequences and different technological process to obtain a new and innovative material. From the standpoint of engineering designer it is useful to consider the single layer of composite as macroscopically homogeneous material. However, composites are non homogeneous bodies. Moreover, layers are not often perfectly bonded together and delamination often occurs. Other violations of lamination theory hypotheses, such as plane stress and thin material, are not unusual and in many cases the transverse shear flexibility and the thickness-normal stiffness should be considered. Therefore the real behaviour of composite materials is quite different from the predictions coming from the traditional lamination theory. Due to the increasing structural performance required to innovative composites, the knowledge of the mechanical properties for different loading cases is a fundamental source of concern. Experimental characterization of materials and structures in different environmental conditions is extremely important to understand the mechanical behaviour of these new materials. The purpose of the present work is to characterize a composite material developed for aerospace applications and produced by means of the resin film infusion process (RFI. Different tests have been carried out: tensile, open-hole and filled-hole tensile, compressive, openhole and filled-hole compressive. The experimental campaign has the aim to define mechanical characteristics of this RFI composite material in different conditions: environmental temperature, Hot/Wet and Cold.

  18. Release of fission products and post-pile creep behaviour of irradiated fuel rods stored under dry conditions

    International Nuclear Information System (INIS)

    Kaspar, G.; Peehs, M.; Bokelmann, R.; Jorde, D.; Schoenfeld, H.; Haas, W.; Bleier, A.; Rutsch, F.

    1985-06-01

    The release of moisture and fission products (Kr-85, H-3 and I-129) under dry storage conditions has been examined on six fuel rods which have become defective in the reactor. During the examinations, inert conditions prevailed and limited air inlet was allowed temporarily. The storage temperature was 400 0 C. The residual moisture content of the fuel rods was approx. 5 g. At the beginning of the test, the total moisture content and 0,05% (max.) of the fission gas inventory were released. Under inert conditions, fission gas was not released during a prolonged period of time. Under oxidizing conditions, however, fission gas was released in the course of UO 2 oxidation. Post-pile creep of Zircaloy cladding tubes was measured at temperatures between 350 and 395 0 C and interval gauge pressures between 69 and 110 bar. The creep curves indicate that the irradiated cladding tube specimens still bear internal residual stresses which contribute through their relaxation to the post-pile creep. (orig.) [de

  19. Material operating behaviour of ABB BWR control rods

    International Nuclear Information System (INIS)

    Rebensdorff, B.; Bart, G.

    2000-01-01

    The BWR control rods made by ABB use boron carbide (B 4 C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B 4 C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

  20. Swarm robotics and complex behaviour of continuum material

    Science.gov (United States)

    dell'Erba, Ramiro

    2018-05-01

    In swarm robotics, just as for an animal swarm in nature, one of the aims is to reach and maintain a desired configuration. One of the possibilities for the team, to reach this aim, is to see what its neighbours are doing. This approach generates a rules system governing the movement of the single robot just by reference to neighbour's motion. The same approach is used in position-based dynamics to simulate behaviour of complex continuum materials under deformation. Therefore, in some previous works, we have considered a two-dimensional lattice of particles and calculated its time evolution by using a rules system derived from our experience in swarm robotics. The new position of a particle, like the element of a swarm, is determined by the spatial position of the other particles. No dynamic is considered, but it can be thought as being hidden in the behaviour rules. This method has given good results in some simple situations reproducing the behaviour of deformable bodies under imposed strain. In this paper we try to stress our model to highlight its limits and how they can be improved. Some other, more complex, examples are computed and discussed. Shear test, different lattices, different fracture mechanisms and ASTM shape sample behaviour have been investigated by the software tool we have developed.

  1. Characterizing the tribological behaviour of fast breeder reactor materials

    International Nuclear Information System (INIS)

    Depierre, J.; Raffailhac, J.

    1984-04-01

    The object of these tests is to define the behaviour of material couples working in conditions as representative as possible of reactor operation. For this purpose a certain number of test installations have been developed to simulate the most typical cases of friction encountered: plane to plane geometry, rotational bearings, guiding bearings. Endurance tests have also been carried out on ball bearings and ballscrews samples. As said before, the test conditions attempt to reproduce as faithfully as possible the environment of the materials used in fast breeder reactors, particularly in: - using purified liquid sodium, and maintaining it isotherm, respectively at three temperature levels: 180, 400 and 550 0 C; - or using argon containing sodium aerosol particles. Some typical values of friction coefficients and rates of wear obtained during the tests with certain couples of materials are given here as examples. The aims which are currently guiding the direction of the tests are also briefly described

  2. On the fission track dating and annealing behaviour of accessory minerals of Eastern Ghats (Andhra Pradesh, India)

    International Nuclear Information System (INIS)

    Koul, S.L.

    1978-01-01

    Use of the etching of fission fragment damage tracks for an estimation of the uranium content of apatite and zircon crystals is described. The etching conditions have been studied for which visible tracks are developed. Fission track determined ages of 25 samples of apatite and zircon crystals from four widely separated regions of India; the Borra mines (Vishakapatanam), Kashipatnam (Vishakapatnam), the Khamam area (Andhra Pradesh) and the Kodrama mines (Bihar) have been determined. Mean ages for these regions are 456 +- 5, 389 +- 4, 486 +- 7 and 664 +- 7 million years respectively. It is concluded that the fission track ages of the minerals date the last metamorphic event of the Eastern Ghats, known as the Indian Ocean Cycle. Annealing studies confirm that radiation damaged fossil tracks can be erased in minerals under intense metamorphic episodes, thus resetting the geological clock. Extrapolation of the experimentally determined temperatures for annealing suggest that a temperature of 170 0 C in 10 6 years will erase all the tracks in the apatite mineral. The uranium concentration has been estimated to be approximately 10 -8 gm/gm in apatite and approximately 10 -6 gm/gm in zircon. (Auth.)

  3. Structural Behaviour of Strengthened Composite Materials. Experimental Studies

    Directory of Open Access Journals (Sweden)

    Vlad Munteanu

    2007-01-01

    Full Text Available Masonry represents one of the earliest structural materials used by mankind. A lot of the ancient building structures were made using masonry. A large number of these buildings have been stated historical monuments. Most commonly masonry elements which are able to cover large spans was masonry arches. The paper makes a detailed presentation on structural behaviour and failure mechanisms of a horizontally loaded masonry arch. The arch model was built at a 1 : 1 scale using solid bricks and M10Z mortar. It was firstly loaded with vertically acting dead loads and with horizontal load acting in its plane. In this loading hypothesis, a plastic hinge occurred leading to the failure of the arch and loss of load bearing capacity. In the next stage of the experimental program, the arch was strengthened using a composite material membrane at the upper face. The membrane consisted in a continuous, glass-fiber fabric and epoxy resin. After proper curing, the same loading hypothesis was used. The failure mechanisms changed and a larger horizontal loading level was noticed. Further on, the arch was rehabilitated using a different composite material layout, the membrane was applied both on upper and bottom faces as well as partially on the lateral faces of the arch. This new rehabilitation layout leads to a significant increase in the load bearing capacity of the arch. The failure mechanisms were changed causing a significantly better overall structural behaviour of the arch.

  4. Crack and fracture behaviour in tough ductile materials

    International Nuclear Information System (INIS)

    Venter, R.D.; Hoeppner, D.W.

    1985-10-01

    The report describes various approaches and developments pertaining to the understanding of crack and fracture behaviour in tough ductile materials. The fundamental elastic fracture mechanics concepts based on the concepts of energy, stress field, and displacement are introduced and their interrelationships demonstrated. The extension of these concepts to include elasto-plastic fracture mechanics considerations is reviewed in the context of the preferred options available for the development of appropriate design methodologies. The recommendations of the authors are directed towards the continued development of the J-integral concept. This energy-based concept, in its fundamental form, has a sound theoretical basis and as such offers the possibility of incorporating elasto-plastic fracture mechanics considerations in the crack and fracture behaviour of tough ductile materials. It must however be emphasized that the concise defintion of J becomes increasingly suspect as the crack length increases. J is not a material property, as is J IC , but emerges as a useful empirical parameter which is dependent upon the particular geometry and the loading imposed on the structure. It is proposed that 'lowest bound' J-resistance curves and the associated J-T curves be experimentally developed and employed in the design process. Improvements to these 'lowest bounds' can be developed through extensive analysis of the twin J-CTOA criteria and validation of this approach through near full scale tests

  5. 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)

  6. Fuel and fission product behaviour in early phases of a severe accident. Part II: Interpretation of the experimental results of the PHEBUS FPT2 test

    Energy Technology Data Exchange (ETDEWEB)

    Dubourg, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); 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); Ducher, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Gavillet, D. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); 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 bundle and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 mm and 900 mm) of the test section previously reported are interpreted in the present paper. Solid state interactions between fuel and cladding have been compared with the characteristics of interaction identified in the previous separate-effect tests. Corium resulting from the interaction between fuel and cladding was formed. The uranium concentration in the corium is compared to analytical tests and a scenario for the corium formation is proposed. The analysis showed that, despite the rather low fuel burn up, the conditions of temperature and oxygen potential reached during the starvation phase are able to give an early very significant release fraction of caesium. A significant part (but not all) of the molybdenum was segregated at grain boundaries and trapped in metallic inclusions from which they were totally removed in the final part of the experiment. During the steam starvation phase, the conditions of oxygen potential were favourable for the formation of simple Ba and BaO chemical forms but the temperature was too low to provoke their volatility. This is one important difference with out-of-pile experiments such as VERCORS for which only a combination of high temperature and low oxygen potential induced a significant barium release. Finally another significant difference with analytical out-of-pile experiments comes from the formation of foamy zones due to the fission gas presence in FPT2-type experiments which give an additional possibility for the formation of stable fission product compounds.

  7. Multiscale Simulation of Thermo-mechanical Processes in Irradiated Fission-reactor Materials

    International Nuclear Information System (INIS)

    Phillpot, Simon R.

    2012-01-01

    The work funded from this project has been published in six papers, with two more in draft form, with submission planned for the near future. The papers are: (1) Kinetically-Evolving Irradiation-Induced Point-Defect Clusters in UO 2 by Molecular-Dynamics Simulation; (2) Kinetically driven point-defect clustering in irradiated MgO by molecular-dynamics simulation; (3) Grain-Boundary Source/Sink Behavior for Point Defect: An Atomistic Simulation Study; (4) Energetics of intrinsic point defects in uranium dioxide from electronic structure calculations; (5) Thermodynamics of fission products in UO 2±x ; and (6) Atomistic study of grain boundary sink strength under prolonged electron irradiation. The other two pieces of work that are currently being written-up for publication are: (1) Effect of Pores and He Bubbles on the Thermal Transport Properties of UO2 by Molecular Dynamics Simulation; and (2) Segregation of Ruthenium to Edge Dislocations in Uranium Dioxide.

  8. Behaviour of core materials and fission product release in accident conditions in LWRs

    International Nuclear Information System (INIS)

    1993-06-01

    The meeting was convened to review the progress and to identify areas of concern, particularly the consequences of taking fuel to higher burnup, where further work would be valuable. Forty participants representing 14 countries attended the meeting. Twenty-four papers were presented and discussed during four technical sessions. Working Groups composed of the session chairmen and authors of papers prepared summaries of each session including conclusions and recommendations for future work. Refs, figs, tabs and plates

  9. Melting behaviour of raw materials and recycled stone wool waste

    DEFF Research Database (Denmark)

    Schultz-Falk, Vickie; Agersted, Karsten; Jensen, Peter Arendt

    2018-01-01

    Stone wool is a widely used material for building insulation, to provide thermal comfort along with fire stability and acoustic comfort for all types of buildings. Stone wool waste generated either during production or during renovation or demolition of buildings can be recycled back into the sto...... wool melt production. This study investigates and compares the thermal response and melting behaviour of a conventional stone wool charge and stone wool waste. The study combines differential scanning calorimetry (DSC), hot stage microscopy (HSM) and X-ray diffraction (XRD). DSC reveals...... that the conventional charge and stone wool waste have fundamentally different thermal responses, where the charge experiences gas release, phase transition and melting of the individual raw materials. The stone wool waste experiences glass transition, crystallization and finally melting. Both DSC and HSM measurements...

  10. Nuclear fission

    International Nuclear Information System (INIS)

    Kodama, T.

    1981-01-01

    The nuclear fission process is pedagogically reviewed from a macroscopic-microscopic point of view. The Droplet model is considered. The fission dynamics is discussed utilizing path integrals and semiclassical methods. (L.C.) [pt

  11. VESPA. Behaviour of long-lived fission and activation products in the nearfield of a nuclear waste repository and the possibilities of their retention

    Energy Technology Data Exchange (ETDEWEB)

    Bischofer, Barbara; Hagemann, Sven [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Koeln (Germany); Altmaier, Marcus [Karlsruher Institut fuer Technologie (KIT) (Germany); and others

    2016-06-15

    The present document is the final report of the Joint Research Project VESPA (Behaviour of Long-lived Fission and Activation Products in the Near Field of a Nuclear Waste Repository and the Possibilities of Their Retention), started in July 2010 with a duration of four years. The following four institutions were collaborative Partners in VESPA: - Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH - Institut fuer Energie- und Klimaforschung, IEK-6: Nukleare Entsorgung und Reaktorsicherheit, Forschungszentrum Juelich (FZJ) - Institut fuer Ressourcenoekologie (IRE), Helmholtz-Zentrum Dresden-Rossendorf (HZDR) - Institut fuer Nukleare Entsorgung (INE), Karlsruher Institut fuer Technologie (KIT) VESPA was funded by the German Federal Ministry of Economics and Energy (BMWi) under the contract numbers 02 E 10770 (GRS), 02 E 10780 (FZJ-IEF-6), 02 E 10790 (HZDR-IRE), 02 E 10800 (KIT-INE).

  12. A comparative study of fission gas behaviour in UO2 and MOX fuels using the meteor fuel performance code

    International Nuclear Information System (INIS)

    Struzik, C.; Garcia, Ph.; Noirot, L.

    2002-01-01

    The paper reviews some of the fission-gas-related differences observed between MOX MIMAS AUC fuels and homogeneous UO 2 fuels. Under steady-state conditions, the apparently higher fractional release in MOX fuels is interpreted with the METEOR fuel performance code as a consequence of their lower thermal conductivity and the higher linear heat rates to which MOX fuel rods are subjected. Although more fundamental diffusion properties are needed, the apparently greater swelling of MOX fuel rods at higher linear heat rates can be ascribed to enhanced diffusion properties. (authors)

  13. Isotopic Determination of Nuclear Materials Using Nuclear Fission Track Registration Technique and Thermal Ionization Mass Spectrometric Technique

    International Nuclear Information System (INIS)

    Jeon, Young Sin; Pyo, Hyeong Yeol; Park, Yong Joon; Song, Kyu Seok; Kim, Won Ho; Jee, Kwang Yong

    2007-05-01

    It is very important to develope the technology for the determination of isotopic ratios of hot particles( 234 U, 235 U, 236 U etc.) detected from swipe samples of various nuclear facilities. This technology is highly competitive internationally and has to be established independently as long as our government maintains atomic energy and treats nuclear materials. In this text, sample pretreatment procedure, gamma-ray counting, alpha or fission track techniques, isotopic analysis of U and Pu, background problems and detection limits for mass determination, and their application to the real swipe sample were described with detailed procedure. This technology would contribute to the Korean economy's high growth rate as well as to superiority of government's leading research and development programs if successfully established

  14. Imprints left by natural radioactivity in geological materials: uranium fission tracks and thermoluminescence applications in earth sciences

    International Nuclear Information System (INIS)

    Broquet, P.; Chambaudet, A.; Rebetez, M.; Charlet, J.M.

    1994-01-01

    In a rock, all minerals which contain uranium are host to a number of spontaneous fission phenomena forming a single damaged area called a ''latent track'', observations of which may lead to dating, uranium mapping and finding paleo-geo-thermometers (thermal history, used in oil exploration). The radioactive elements during the decay process release energy which is trapped as electrons into the physical or chemical defects of the crystalline lattice; this energy can be later released by heating the mineral (thermic stimulated luminescence); the thermoluminescence is characterized by a glow which spectrum constitutes a typical feature of the mineral, its crystallization conditions and the subsequent evolution of the material. Natural and induced glow curve may be produced. 6 figs., 52 refs

  15. Asymmetric fission and evaporation of C{sup r+}{sub 60} (r = 2-4) fullerene ions in ion-C{sub 60} collisions: III. Universal behaviour of fission

    Energy Technology Data Exchange (ETDEWEB)

    Bordenave-Montesquieu, D; Bordenave-Montesquieu, A; Rentenier, A; Moretto-Capelle, P [LCAR-IRSAMC, UMR 5589 Universite Paul Sabatier-CNRS, 118 rte de Narbonne, 31062 Toulouse Cedex (France)

    2005-04-14

    The behaviour of the asymmetrical fission (AF) scheme (correlated ion distributions) against the collision conditions is investigated using H{sup +}{sub x} (x = 1-3) and He{sup +} projectiles in the 1-130 keV collision energy range. The present work is an extension of our recent publications on this topic using 11 keV protons (Rentenier et al 2004 J. Phys. B: At. Mol. Opt. Phys. 37 2429 and 2455). The threshold for AF is observed at 2 keV proton energy corresponding to a maximum deposited energy equal to about 41 eV. The main result concerns the fragment distributions resulting from AF of C{sup r+}{sub 60} ions, and secondary dissociation of even-n C{sup +}{sub n} fragments, which are both found to remain independent of the projectile species and collision velocity. These findings indicate that they are insensitive to the internal energy distributions of the parent ions. In addition, a contribution of binary collisions between the projectile and individual carbon atoms of the C{sub 60} molecule to AF is identified in the C{sup +}{sub 1} production at the lowest collision velocities, the so-called impulsive fragmentation.

  16. Multiscale Simulation of Thermo-mechancial Processes in Irradiated Fission-reactor Materials.

    Energy Technology Data Exchange (ETDEWEB)

    Simon R. Phillpot

    2012-06-08

    The work funded from this project has been published in six papers, with two more in draft form, with submission planned for the near future. The papers are: (1) Kinetically-Evolving Irradiation-Induced Point-Defect Clusters in UO{sub 2} by Molecular-Dynamics Simulation; (2) Kinetically driven point-defect clustering in irradiated MgO by molecular-dynamics simulation; (3) Grain-Boundary Source/Sink Behavior for Point Defect: An Atomistic Simulation Study; (4) Energetics of intrinsic point defects in uranium dioxide from electronic structure calculations; (5) Thermodynamics of fission products in UO{sub 2{+-}x}; and (6) Atomistic study of grain boundary sink strength under prolonged electron irradiation. The other two pieces of work that are currently being written-up for publication are: (1) Effect of Pores and He Bubbles on the Thermal Transport Properties of UO2 by Molecular Dynamics Simulation; and (2) Segregation of Ruthenium to Edge Dislocations in Uranium Dioxide.

  17. Fuel behaviour and fission product release under realistic hydrogen conditions comparisons between HEVA 06 test results and Vulcain computations

    International Nuclear Information System (INIS)

    Dumas, J.M.; Lhiaubet, G.

    1989-07-01

    The HEVA 06 test was designed to simulate the conditions existing at the time when fission products are released from irradiated fuel under hydrogen conditions occurring in a PWR core at low pressure. The test conditions were defined from results provided by the core degradation module of the ESCADRE system (1): VULCAIN. This computer code has been recently used to analyse the early core degradation of a 900 MWe PWR in the AF accident sequence (as defined in WASH - 1400, USNRC - 1975). In this scenario, the core would begin to uncover about one day after scram with the system pressure at about 0.4 MPa. The fission product release starts 70 minutes after core dewatering. The F.P. are transferred to the core outlet in an increasingly hydrogen-rich steam atmosphere. The carrier gas is nearly pure hydrogen in the time period 100 - 130 minutes after core uncovering. A large release of F.P. is predicted in the upper part of the core when the steam starvation occurs. At that time, two thirds of the cladding have been oxidised on an average. Before each HEVA test a fuel sample with a burn-up of 36 GWd/tU is reirradiated in order to observe the release of short-lived fission products. A pre-oxidation was primarely conducted in the HEVA 06 test at a temperature of 1300 0 C and controlled to reach a 2/3 cladding oxidation state. Then the steam was progressively replaced by hydrogen and a heat-up rate of 1.5 0 C/s was induced to reach a temperature of 2100 0 C. The fuel was maintained at this temperature for half an hour in hydrogen. The volatile F.P. release kinetics were observed by on-line gamma spectrometry. Pre test calculations of F.P. release kinetics performed with the EMIS module based on the CORSOR models (3) are compared with the test results. Measured releases of cesium and iodine are really lower than those predicted. Axial and radial F.P. distributions in the fuel pellets are available from gamma tomography measurements performed after the test. Tellurium seems

  18. Development of a fission product transport module predicting the behavior of radiological materials during sever accidents in a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Hyung Seok; Rhee, Bo Wook; Kim, Dong Ha [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-09-15

    Korea Atomic Energy Research Institute is developing a fission product transport module for predicting the behavior of radioactive materials in the primary cooling system of a nuclear power plant as a separate module, which will be connected to a severe accident analysis code, Core Meltdown Progression Accident Simulation Software (COMPASS). This fission product transport (COMPASS-FP) module consists of a fission product release model, an aerosol generation model, and an aerosol transport model. In the fission product release model there are three submodels based on empirical correlations, and they are used to simulate the fission product gases release from the reactor core. In the aerosol generation model, the mass conservation law and Raoult's law are applied to the mixture of vapors and droplets of the fission products in a specified control volume to find the generation of the aerosol droplet. In the aerosol transport model, empirical correlations available from the open literature are used to simulate the aerosol removal processes owing to the gravitational settling, inertia impaction, diffusiophoresis, and thermophoresis. The COMPASS-FP module was validated against Aerosol Behavior Code Validation and Evaluation (ABCOVE-5) test performed by Hanford Engineering Development Laboratory for comparing the prediction and test data. The comparison results assuming a non-spherical aerosol shape for the suspended aerosol mass concentration showed a good agreement with an error range of about ±6%. It was found that the COMPASS-FP module produced the reasonable results of the fission product gases release, the aerosol generation, and the gravitational settling in the aerosol removal processes for ABCOVE-5. However, more validation for other aerosol removal models needs to be performed.

  19. The emergence of complex behaviours in molecular magnetic materials.

    Science.gov (United States)

    Goss, Karin; Gatteschi, Dante; Bogani, Lapo

    2014-09-14

    Molecular magnetism is considered an area where magnetic phenomena that are usually difficult to demonstrate can emerge with particular clarity. Over the years, however, less understandable systems have appeared in the literature of molecular magnetic materials, in some cases showing features that hint at the spontaneous emergence of global structures out of local interactions. This ingredient is typical of a wider class of problems, called complex behaviours, where the theory of complexity is currently being developed. In this perspective we wish to focus our attention on these systems and the underlying problematic that they highlight. We particularly highlight the emergence of the signatures of complexity in several molecular magnetic systems, which may provide unexplored opportunities for physical and chemical investigations.

  20. First Results for Fluid Dynamics, Neutronics and Fission Product Behaviour in HTR applying the HTR Code Package (HCP) Prototype

    International Nuclear Information System (INIS)

    Allelein, H.-J.; Kasselmann, S.; Xhonneux, A.; Lambertz, D.

    2014-01-01

    To simulate the different aspects of High Temperature Reactor (HTR) cores, a variety of specialized computer codes have been developed at Forschungszentrum Jülich (IEK-6) and Aachen University (LRST) in the last decades. In order to preserve knowledge, to overcome present limitations and to make these codes applicable to modern computer clusters, these individual programs are being integrated into a consistent code package. The so-called HTR code package (HCP) couples the related and recently applied physics models in a highly integrated manner and therefore allows to simulate phenomena with higher precision in space and time while at the same time applying state-of-the-art programming techniques and standards. This paper provides an overview of the status of the HCP and reports about first benchmark results for an HCP prototype which couples the fluid dynamics and time dependent neutronics code MGT-3D, the burn up code TNT and the fission product release code STACY. Due to the coupling of MGT-3D and TNT, a first step towards a new reactor operation and accident simulation code was made, where nuclide concentrations calculated by TNT are fed back into a new spectrum code of the HCP. Selected operation scenarios of the HTR-Module 200 concept plant and the HTTR were chosen to be simulated with the HCP prototype. The fission product release during normal operation conditions will be calculated with STACY based on a core status derived from SERPENT and MGT–3D. Comparisons will be shown against data generated by the legacy codes VSOP99/11, NAKURE and FRESCO-II. (author)

  1. Analytical methods for fissionable materials in the nuclear fuel cycle. Covering June 1974--June 1975

    International Nuclear Information System (INIS)

    Waterbury, G.R.

    1975-10-01

    Research progress is reported on method development for the dissolution of difficult-to-dissolve materials, the automated analysis of plutonium and uranium, the preparation of plutonium materials for the Safeguard Analytical Laboratory Evaluation (SALE) Program, and the analysis of HTGR fuel and SALE uranium materials. The previously developed Teflon-container, metal-shell apparatus was applied to the dissolution of various nuclear materials. Gas--solid reactions, mainly using chlorine at elevated temperatures, are promising for separating uranium from refractory compounds. An automated spectrophotometer designed for determining plutonium and uranium was tested successfully. Procedures were developed for this instrument to analyze uranium--plutonium mixtures and the effects of diverse ions upon the analysis of plutonium and uranium were further established. A versatile apparatus was assembled to develop electrotitrimetric methods that will serve as the basis for precise automated determinations of plutonium. Plutonium materials prepared for the Safeguard Analytical Laboratory Evaluation (SALE) Program were plutonium oxide, uranium--plutonium mixed oxide, and plutonium metal. Improvements were made in the methods used for determining uranium in HTGR fuel materials and SALE uranium materials. Plutonium metal samples were prepared, characterized, and distributed, and half-life measurements were in progress as part of an inter-ERDA-laboratory program to measure accurately the half-lives of long-lived plutonium isotopes

  2. Characterization of impact behaviour of armour plate materials

    Science.gov (United States)

    Bassim, M. N.; Bolduc, M.; Nazimuddin, G.; Delorme, J.; Polyzois, I.

    2012-08-01

    Three armour plate materials, including two steels, namely HHA and Mars 300, and an aluminium alloy 5083, were studied under impact loading to determine their behaviour and the mechanisms of deformation that lead to failure. The experimental testing was carried out using either a direct impact compression Split Hopkinson Bar or a torsion Hopkinson Bar. The impact properties and stress-strain cures were obtained as a function of the impact momentum in compression and the angle of twist in torsion. It was found that at the high strain rates developed in the specimen during the tests, the deformation occurs by the formation of adiabatic shear bands (ASBs) which may lead to the formation of cracks within the bands and the ultimate failure of the specimens. It was also found that below a certain impact momentum, the deformation is more uniform and no ASBs are formed. Also, ASBs are more likely to form in the BCC metals such as the two steels while diffuse ASBs associated with plastic flow are exhibited in the 5083 aluminum alloy. Microstructural techniques ranging from optical microscopy to atomic force microscopy (AFM) were used to study the topography of the ASBs. Also, modelling of the formation was performed. The results provide a comprehensive understanding of the role of ASBs in the failure of these materials.

  3. Characterization of impact behaviour of armour plate materials

    Directory of Open Access Journals (Sweden)

    Nazimuddin G.

    2012-08-01

    Full Text Available Three armour plate materials, including two steels, namely HHA and Mars 300, and an aluminium alloy 5083, were studied under impact loading to determine their behaviour and the mechanisms of deformation that lead to failure. The experimental testing was carried out using either a direct impact compression Split Hopkinson Bar or a torsion Hopkinson Bar. The impact properties and stress-strain cures were obtained as a function of the impact momentum in compression and the angle of twist in torsion. It was found that at the high strain rates developed in the specimen during the tests, the deformation occurs by the formation of adiabatic shear bands (ASBs which may lead to the formation of cracks within the bands and the ultimate failure of the specimens. It was also found that below a certain impact momentum, the deformation is more uniform and no ASBs are formed. Also, ASBs are more likely to form in the BCC metals such as the two steels while diffuse ASBs associated with plastic flow are exhibited in the 5083 aluminum alloy. Microstructural techniques ranging from optical microscopy to atomic force microscopy (AFM were used to study the topography of the ASBs. Also, modelling of the formation was performed. The results provide a comprehensive understanding of the role of ASBs in the failure of these materials.

  4. Automated inventory and material science scoping calculations under fission and fusion conditions

    Directory of Open Access Journals (Sweden)

    Mark R. Gilbert

    2017-09-01

    Full Text Available The FISPACT-II inventory simulation platform is a modern computational tool with advanced and unique capabilities. It is sufficiently flexible and efficient to make it an ideal basis around which to perform extensive simulation studies to scope a variety of responses of many materials (elements to several different neutron irradiation scenarios. This paper briefly presents the typical outputs from these scoping studies, which have been used to compile a suite of nuclear physics materials handbooks, providing a useful and vital resource for material selection and design studies. Several different global responses are extracted from these reports, allowing for comparisons between materials and between different irradiation conditions. A new graphical output format has been developed for the FISPACT-II platform to display these “global summaries”; results for different elements are shown in a periodic table layout, allowing side-by-side comparisons. Several examples of such plots are presented and discussed.

  5. Neutron data error estimate of criticality calculations for lattice in shielding containers with metal fissionable materials

    International Nuclear Information System (INIS)

    Vasil'ev, A.P.; Krepkij, A.S.; Lukin, A.V.; Mikhal'kova, A.G.; Orlov, A.I.; Perezhogin, V.D.; Samojlova, L.Yu.; Sokolov, Yu.A.; Terekhin, V.A.; Chernukhin, Yu.I.

    1991-01-01

    Critical mass experiments were performed using assemblies which simulated one-dimensional lattice consisting of shielding containers with metal fissile materials. Calculations of the criticality of the above assemblies were carried out using the KLAN program with the BAS neutron constants. Errors in the calculations of the criticality for one-, two-, and three-dimensional lattices are estimated. 3 refs.; 1 tab

  6. Accelerator-based approach experiments for remote identification of fissionable and other materials

    International Nuclear Information System (INIS)

    Chuvilo, I.V.; Danilov, M.M.; Katarzhnov, Yu.D.; Kushin, V.V.; Nedopekin, V.G.; Plotnikov, S.V.; Rogov, V.I.

    1998-01-01

    Recently there has been a great deal of interest in studying possible methods for remote non-destructive material composition testing, for example, for cargo identification at transportation, neutron logging etc., by means of nuclear detection (D.R. Brown, T. Gozani (1995)). Of current concern are the applications of pulsed fast neutron analysis in determining the composition of fissile objects (I.I. Zaliubovskiy et al. (1993)). In this paper the observed experimental results are discussed indicating the possibility of practical realization of the method for remote material identification. The approach is based on measuring gamma ray spectra from an object to be examined after its irradiation with short neutron pulses produced by an accelerator. The obtained time and energy gamma spectra are used for material inspection. The information is obtained by using time-of-flight (TOF) analysis between the accelerator pulse and the arrival of gamma rays in NaI detectors located far enough from an object to be examined. The method seems to be the most effective for fissile materials identification. (orig.)

  7. Development of radiation resistant structural materials utilizing fission research reactors in Japan (Role of research reactors)

    International Nuclear Information System (INIS)

    Shikama, T.; Tanigawa, H.; Nozawa, T.; Muroga, T.; Aoyama, T.; Kawamura, H.; Ishihara, M.; Ito, C.; Kaneda, S.; Mimura, S.

    2009-01-01

    Structural materials for next-generation nuclear power systems should have a good radiation resistance, where the expected accumulation dose will largely exceed 10 dpa. Among several candidate materials, materials of five categories, 1. Austenitic steels, including high nickel alloys, 2. Low activation ferritic martensitic steels, 3. ODS steels (austenitic and ferritic), 4. Vanadium based alloys, 5. Silicon carbide composites (SiC/SiCf). All have been most extensively studied in Japan, in collaboration among industries, national institutes such as Japan Atomic Energy Agency (JAEA), National Institute for Fusion Science (NIFS) and National Institute for Materials Science (NIMS), and universities. The high nickel base alloys were studied for their low swelling behaviors mainly by the NIMS and the austenitic steels are studied for their reliable engineering data base and their reliable performance in irradiation environments mainly by the JAEA, mainly for their application in the near-term projects such as the ITER and the Sodium Cooled Fast Reactors. The most extensive studies are now concentrated on the Low Activation Ferritic Marsensitic steels and ODS steels, for their application in a demonstration fusion reactor and prototype sodium cooled fast reactors. Fundamental studies on radiation effects are carried out, mainly utilizing Japan Materials Testing Rector (JMTR) with its flexible irradiation ability, up to a few dpa. For higher dpa irradiation, a fast test reactor, JOYO is utilized up to several 10s dpa. Some international collaborations such as Japan/USA and Japan/France are effective to utilize reactors abroad, such as High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory, and sodium cooled high flux fast reactors in France. Silicon carbide based composites are extensively studied by university groups led by Kyoto University and the JAEA. For their performance in heavy irradiation environments, the Japan/USA collaboration plays an important role

  8. Contributions to the R-curve behaviour of ceramic materials

    International Nuclear Information System (INIS)

    Fett, T.

    1994-12-01

    Several ceramic materials show an increase in crack growth resistance with increasing crack extension. Especially, in case of coarse-grained alumina this ''R-curve effect'' is caused by crack-face interactions in the wake of the advancing crack. Similar effects occur for whisker reinforced ceramics. Due to the crack-face interactions so-called ''bridging stresses'' are generated which transfer forces between the two crack surfaces. A second reason for an increase of crack-growth resistance are stress-induced phase transformations in zirconia ceramics with the tetragonal phase changing to the monoclinic phase. These transformations will affect the stress field in the surroundings of crack tips. The transformation generates a crack-tip transformation zone and, due to the stress balance, also residual stresses in the whole crack region which result in a residual stress intensity factor. This additional stress intensity factor is also a reason for the R-curve behaviour. In this report both effects are outlined in detail. (orig.) [de

  9. An experimental study of the behaviour of fission products following an accident on a swimming pool reactor

    International Nuclear Information System (INIS)

    Dadillon, J.

    1976-11-01

    In the estimation of nuclear risks connected with the running of a reactor an essential factor, sometimes neglected because insufficiently known, is the knowledge of the type, amount and behaviour of the contamination actually released inside the containment in the case of an accident. In the special case of swimming pool reactors the cooling fluid proves to be a very efficient barrier against contamination. Three experiments were carried out in the reactor CABRI, during which several fuel element plates were melted inside the core itself. (Author)

  10. Accidental behaviour of nuclear fuel in a warehousing site under air: investigation of the nuclear ceramic oxidation and of fission gas release; Comportement accidentel du combustible nucleaire dans un site d'entreposage sous air: Etude de l'oxydation de la ceramique nucleaire et du relachement des gaz de fission

    Energy Technology Data Exchange (ETDEWEB)

    Desgranges, L.

    2006-12-15

    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{sub 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{sub 2} to U{sub 3}O{sub 7}, and from U{sub 3}O{sub 7} to U{sub 3}O{sub 8}), and with further high activity laboratory experiments

  11. Qualification of SiC materials for fusion and fission reactors

    International Nuclear Information System (INIS)

    Ryazanov, Alexander

    2009-01-01

    Ceramic materials such as silicon carbide (SiC) and SiC/SiC composites are both considered, due to their high-temperature strength, pseudo-ductile fracture behavior and low-induced radioactivity, as candidate materials for fusion reactor (test blanket module for ITER) and high temperature gas-cooled reactors (HTGR). The radiation swelling and creep of SiC are very important physical phenomena that determine the radiation resistance of them in these reactors. Other important problem which exists especially in fusion reactor is an effect of accumulation of high concentrations of helium atoms in SiC (up to 15000-20000 at.ppm) due to (n,α) nuclear reaction on physical mechanical properties. An understanding of the physical mechanism of this phenomenon is very important for the investigations of helium atom effect on radiation swelling in SiC. In this report a compilation of non-irradiated and irradiated properties of SiC are provided and analyzed in terms of their application to fusion and high temperature gas cooled reactors. Special topic of this report is oriented on the micro structural changes in chemically vapor-deposited (CVD) high-purity beta-SiC during neutron and ion irradiations at elevated temperatures. The evolutions of various radiation induced defects including dislocation loops, network dislocations and cavities are presented here as a function of irradiation temperature and fluencies. These observations are discussed in relation with such irradiation phenomena in SiC as low temperature swelling and cavity swelling. One of the main difficulties in the radiation damage studies of SiC materials lies in the absence of theoretical models and interpretation of many physical mechanisms of radiation phenomena including the radiation swelling and creep. The point defects in ceramic materials are characterized by the charge states and they can have an effective charge. The internal effective electrical field is formed due to the accumulation of charged point

  12. The Role of Materialism on Social, Emotional and Behavioural Difficulties for British Adolescents

    Science.gov (United States)

    Maras, Pam; Moon, Amy; Gupta, Taveeshi; Gridley, Nicole

    2015-01-01

    The relationship between materialism and social-emotional behavioural difficulties (SEBDs) was assessed by comparing a sample of adolescents receiving in-school behavioural support with adolescents not receiving any support. All participants completed the Youth Materialism Scale and the Strengths and Difficulties Questionnaire. Binary logistic…

  13. A spallation-based irradiation test facility for fusion and future fission materials

    International Nuclear Information System (INIS)

    Samec, K.; Fusco, Y.; Kadi, Y.; Luis, R.; Romanets, Y.; Behzad, M.; Aleksan, R.; Bousson, S.

    2014-01-01

    The EU's FP7 TIARA program for developing accelerator-based facilities has recently demonstrated the unique capabilities of a compact and powerful spallation source for irradiating advanced nuclear materials. The spectrum and intensity of the neutron flux produced in the proposed facility fulfils the requirements of the proposed DEMO fusion reactor, ADS reactors and also Gen III / IV reactors. Test conditions can be modulated, covering temperature from 400 to 550 deg. C, liquid metal corrosion, cyclical or static stress up to 500 MPa and neutron/proton irradiation damage of up to 25 DPA per annum over a volume occupying one litre. The entire 'TMIF' facility fits inside a cube 2 metres on a side, and is dimensioned for an accelerator beam power of 100 kW, thus reducing costs and offering great versatility and flexibility. (authors)

  14. A spallation-based irradiation test facility for fusion and future fission materials

    CERN Document Server

    Samec, K; Kadi, Y; Luis, R; Romanets, Y; Behzad, M; Aleksan, R; Bousson, S

    2014-01-01

    The EU’s FP7 TIARA program for developing accelerator-based facilities has recently demonstrated the unique capabilities of a compact and powerful spallation source for irradiating advanced nuclear materials. The spectrum and intensity of the neutron flux produced in the proposed facility fulfils the requirements of the DEMO fusion reactor for ITER, ADS reactors and also Gen III / IV reactors. Test conditions can be modulated, covering temperature from 400 to 550°C, liquid metal corrosion, cyclical or static stress up to 500 MPa and neutron/proton irradiation damage of up to 25 DPA per annum. The entire “TMIF” facility fits inside a cube 2 metres on a side, and is dimensioned for an accelerator beam power of 100 kW, thus reducing costs and offering great versatility and flexibility.

  15. Method to separate fission noble gases from gaseous wastes of a reprocessing plant for nuclear fuel material

    International Nuclear Information System (INIS)

    Schnez, H.

    1977-01-01

    In order to avoid the high cost expenditure in the separation of fission noble gases from waste gas of the head end, the following economical method is suggested: The fission noble gases released in the solvent - after grinding and burn-up of the nuclear fuel elements and dissolving in HNO 3 - are purified in a known method and collected in an equalizing tank. From here, the fission noble gas quantity necessary as washing gas is recycled into the solvent, so that a part of the fission noble gas quantity flows in a circuit. The quantity of fission noble gas not required for the above is separated from the circuit, compressed and put into a storage container from where it can be put into gas flashs or be recycled in the gas circuit where necessary. Furthermore, the method involves that to separate krypton, the filtered fission noble gas is compressed, cooled and rectified, whereby the krypton mixture taken from the rectification column is stored under high pressure and the gas part containing xenon, occuring as liquid, is at least partly fed back to the solvent. (HPH) [de

  16. A small flat fission chamber

    International Nuclear Information System (INIS)

    Li Yijun; Wang Dalun; Chen Suhe

    1999-01-01

    With fission materials of depleted uranium, natural uranium, enriched uranium, 239 Pu, and 237 Np, the authors have designed and made a series of small flat fission chamber. The authors narrated the construction of the fission chamber and its technological process of manufacture, and furthermore, the authors have measured and discussed the follow correct factor, self-absorption, boundary effect, threshold loss factor, bottom scatter and or so

  17. Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

  18. A study of potential high band-gap photovoltaic materials for a two step photon intermediate technique in fission energy conversion. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Prelas, M.A.

    1996-01-24

    This report describes progress made to develop a high bandgap photovoltaic materials for direct conversion to electricity of excimer radiation produced by fission energy pumped laser. This report summarizes the major achievements in sections. The first section covers n-type diamond. The second section covers forced diffusion. The third section covers radiation effects. The fourth section covers progress in Schottky barrier and heterojunction photovoltaic cells. The fifth section covers cell and reactor development.

  19. Progress report on research and development work 1991 of the Institute of Genetics and Toxicology of Fissionable Materials, Karlsruhe Nuclear Research Center

    International Nuclear Information System (INIS)

    1991-03-01

    The present annual report describes the results of research work done by the Institute of Genetics and Toxicology of Fissionable Materials (IGT) in 1991. The following eight subjects were dealt with: genetic repair; genetic regulation; biological carcinogenesis; molecular genetics of eukaryontic genes; genetic mouse models for human illnesses; radiation toxicology of actinides; molecular and cellular environmental toxicology, and in vivo fractionation and speciation of actinides. (MG) [de

  20. Study of fission products (Cs, Ba, Mo, Ru) behaviour in irradiated and simulated nuclear fuels during severe accidents using X-ray absorption Spectroscopy, SIMS and EPMA

    International Nuclear Information System (INIS)

    Geiger, Ernesto

    2016-01-01

    The identification of Fission Products (FP) release mechanism from irradiated nuclear fuels during a severe accident is of main importance for the development of codes for the estimation of the source-term (nature and quantity of radionuclides released into the environment). among the many FP Ba, Cs, Mo and Ru present a particular interest, since they may interact with each other or other elements and thus affect their release. In the framework of this thesis, two work axes have been set up in order to identify, firstly, the chemical phases initially present before the accident and, secondly, their evolution during the accident itself. The experimental approach consisted in reproducing nuclear severe accidents conditions at laboratory scale using both irradiated fuels and model materials (natural UO_2 doped with 12 FP). The advantage of these latter is the possibility of using characterization methods such as X-ray absorption Spectroscopy which are not available for irradiated fuels. Three irradiated fuel samples have been studied, representative to an initial state (before the accident), to an intermediate stage (1773 K) and to an advanced stage (2873 K) of a nuclear severe accident. Regarding to model materials, many accident sequences have been carried out, from 573 to 1973 K. Experimental results have allowed to establish a new release mechanism, considering both reducing and oxidizing conditions during an accident. These results have also demonstrated the importance of model materials as a complement to irradiated nuclear fuels in the study of nuclear severe accidents. (author) [fr

  1. Ternary fission

    International Nuclear Information System (INIS)

    Wagemans, C.

    1991-01-01

    Since its discovery in 1946, light (charged) particle accompanied fission (ternary fission) has been extensively studied, for spontaneous as well as for induced fission reactions. The reason for this interest was twofold: the ternary particles being emitted in space and time close to the scission point were expected to supply information on the scission point configuration and the ternary fission process was an important source of helium, tritium, and hydrogen production in nuclear reactors, for which data were requested by the nuclear industry. Significant experimental progress has been realized with the advent of high-resolution detectors, powerful multiparameter data acquisition systems, and intense neutron and photon beams. As far as theory is concerned, the trajectory calculations (in which scission point parameters are deduced from the experimental observations) have been very much improved. An attempt was made to explain ternary particle emission in terms of a Plateau-Rayleigh hydrodynamical instability of a relatively long cylindrical neck or cylindrical nucleus. New results have also been obtained on the so-called open-quotes trueclose quotes ternary fission (fission in three about-equal fragments). The spontaneous emission of charged particles has also clearly been demonstrated in recent years. This chapter discusses the main characteristics of ternary fission, theoretical models, light particle emission probabilities, the dependence of the emission probabilities on experimental variables, light particle energy distributions, light particle angular distributions, correlations between light particle accompanied fission observables, open-quotes trueclose quotes ternary fission, and spontaneous emission of heavy ions. 143 refs., 18 figs., 8 tabs

  2. Uranium ores of Kazakhstan as the most technologic source of a fissionable material

    International Nuclear Information System (INIS)

    Berikbolov, B.R.

    1999-01-01

    Kazakhstan as is known has unique deposits of uranium. Its resources composed a third part of the world resources. The most important part of resources having a practical value, is related with depression in southern regions of the Republic. By now more than 15 deposits are discovered and partially explored. These deposits from three uranium provinces - Shu-Sarysu, Syr-Darya and Ili. The ores occur in friable water-bearing sandy horizons of Cretaceous and Paleogene age between waterproof agrillaceous sediments at depth from 100 up to 600 m. Ore bodies thickness changes from 5 to 10 m at uranian average-grade 0.03-0.1 %. Width of band shaped ore bodies changes from tens meters to the one kilometers and extent changes from one kilometer up to many tens kilometers. The important feature of deposits is their suitability for development by progressive in situ leaching (ISL) method. It was demonstrated, that uranium ores are comprehensive and, that is important, a lot of commercially important elements, containing in ores, gives in to extraction at development by the ISL method. The preliminary calculation of expenditures for the extraction of useful byproducts from ordinary sulphate solution have demonstrated rather high profitableness for rhenium, scandium, selenium, rare earth even at the very low contents in solution. It was pointed out, that whole technological chain applied now at industrial scale is oriented to mono-metallic uranium ores, therefore present technology of leaching and recovery of industrial solution does not allow ti extract all valuable components containing in ores. The development of new improved technological chain. beginning with a composition of leaching out reagent and up to applying of miscellaneous sorbing materials, can create new mineral-raw base of rare and dissipated elements and to lower considerably the price of uranium mining from sandstone deposits

  3. Numerical simulation of liquefaction behaviour of granular materials ...

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    cles using Discrete Element Method (DEM) is used to study the liquefaction behaviour of ... studies have focussed on the stress-strain relation- ... experimentation still remains quite problematic. ... distorting the periodic cell and changing its vol-.

  4. Micro plate fission chamber development

    International Nuclear Information System (INIS)

    Wang Mei; Wen Zhongwei; Lin Jufang; Jiang Li; Liu Rong; Wang Dalun

    2014-01-01

    To conduct the measurement of neutron flux and the fission rate distribution at several position in assemblies, the micro plate fission chamber was designed and fabricated. Since the requirement of smaller volume and less structure material was taken into consideration, it is convinient, commercial and practical to use fission chamber to measure neutron flux in specific condition. In this paper, the structure of fission chamber and process of fabrication were introduced and performance test result was presented. The detection efficiency is 91.7%. (authors)

  5. 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)

  6. The role of materialism on social, emotional and behavioural difficulties for British adolescents

    OpenAIRE

    Maras, Pam; Moon, Amy; Gupta, Taveeshi; Gridley, Nicole

    2015-01-01

    The relationship between materialism and social-emotional behavioural difficulties (SEBDs) was assessed by comparing a sample of adolescents receiving in-school behavioural support with adolescents not receiving any support. All participants completed the Youth Materialism Scale and the Strengths and Difficulties Questionnaire. Binary logistic regression indicated that adolescents who reported higher levels of materialism were more likely to be classified into a group considered ‘at-risk’ for...

  7. Potentials of fissioning plasmas

    International Nuclear Information System (INIS)

    Karlheinz, Thom.

    1979-01-01

    Successful experiments with the nuclear pumping of lasers have demonstrated that in gaseous medium the kinetic energy of fission fragments can be converted directly into non-equilibrium optical radiation. This confirms the concept that the fissioning medium in a gas-phase nuclear reactor shows an internal structure such as a plasma in nearly thermal equilibrium varying up to a state of extreme-non-equilibrium. The accompanying variations of temperatures, pressure and radiative spectrum suggest wide ranges of applications. For example, in the gas-phase fission reactor concept enriched uranium hexafluoride or an uranium plasma replaces conventional fuel elements and permits operation above the melting point of solid materials. This potential has been motivation for the US National Aeronautics and Space Administration (NASA) to conduct relevant research for high specific impulse propulsion in space. The need to separate the high temperature gaseous fuel from the surfaces of a containing vessel and to protect them against thermal radiation has led to the concept of an externally moderated reactor in which the fissioning gaseous material is suspended by fluid dynamic means and the flow of opaque buffer gas removes the power. The gaseous nuclear fuel can slowly be circulated through the reactor for continuous on-site reprocessing including the annihilation of transuranium actinides at fission when being fed back into the reactor. An equilibrium of the generation and destruction of such actinides at fission when being fed back into the reactor. An equilibrium of the generation and destruction of such actinides can thus be achieved. These characteristics and the unique radiative properties led to the expectation that the gas-phase fission reactor could feature improved safety, safeguarding and economy, in addition to new technologies such as processing, photochemistry and the transmission of power over large distances in space

  8. 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)

  9. Construction for fissionable material

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1978-01-01

    A nuclear reactor fuel assembly is designed to maintain its structural integrity during all phases of reactor operation. Spacer assemblies, containing a plurality of rectangular slotted plates intersecting and interlocking in egg-crate fashion, laterally maintain the fuel elements and guide tubes in a spaced array. Spacer assembly movement is restrained by collars mechanically fixed to guide tube sleeves at each spacer assembly location. (Auth.)

  10. Construction for fissionable material

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

    This invention is directed toward a nuclear fuel assembly guide tube arrangement which restrains spacer grid movement due to coolant flow and which offers secondary means for supporting a fuel assembly during handling and transfer operations. (Auth.)

  11. The evaluation of failure stress and released amount of fission product gas of power ramped rod by fuel behaviour analysis code 'FEMAXI-III'

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujita, Misao

    1984-01-01

    Pellet-Cladding Interaction(PCI) related in-pile failure of Zircaloy sheathed fuel rod is in general considered to be caused by combination of pellet-cladding mechanical interaction(PCMI) with fuel-cladding chemical interaction(FCCI). An understanding of a basic mechanism of PCI-related fuel failure is therefore necessary to get actual cladding hoop stress from mechanical interaction and released amounts of fission product(FP) gas of aggressive environmental agency from chemical interaction. This paper describes results of code analysis performed on fuel failure to cladding hoop stress and amounts of FP gas released under the condition associated with power ramping. Data from Halden(HBWR) and from Studsvik(R2) are used for code analysis. The fuel behaviour analysis code ''FEMAXI-III'' is used as an analytical tool. The followings are revealed from the study: (1) PCI-related fuel failure is dependent upon cladding hoop stress and released amounts of FP gas at power ramping. (2) Preliminary calculated threshold values of hoop stress and of released amounts of FP gas to PCI failure are respectively 330MPa, 10% under the Halden condition, 190MPa, 5% under the Inter ramp(BWR) condition, and 270MPa, 14% under the Over ramp(PWR) condition. The values of hoop stress calculated are almost in the similar range of those obtained from ex-reactor PCI simulated tests searched from references published. (3) The FEMAXI-III code verification is made in mechanical manner by using in-pile deformation data(diametral strain) obtained from power ramping test undertaken by JAERI. While, the code verification is made in thermal manner by using punctured FP gas data obtained from post irradiation examination performed on non-defected power ramped fuel rods. The calculations are resulted in good agreements to both, mechanical and thermal experimental data suggesting the validity of the code evaluation. (J.P.N.)

  12. Forced decontamination of fission products deposited on urban areas

    International Nuclear Information System (INIS)

    Warming, L.

    1984-12-01

    Long-lived fission products may be deposited in the environment following a serious reactor accident. Areas of special concern are cities where the collective dose might be high because of the population. An extensive literature list is presented here. Only a few of the references deal with the problem as a whole. Some references deal with non-radiaoctive materials but give us useful information about the behaviour of particles on outdoor surfaces. (author)

  13. Some aspects of the tribological behaviour of materials in sodium

    International Nuclear Information System (INIS)

    Campbell, C.S.; Lewis, M.W.J.

    1980-08-01

    The influence of boundary lubricating films formed by reaction of metallic surfaces with oxygen-containing sodium is discussed. In general, pre-existing surface metallic oxides are reduced in high-temperature low-oxygen sodium, and tribological behaviour is accordingly poor. Chromium-containing alloys, however, can react more readily with oxygen-containing sodium to form sodium chromite, NaCrO 2 , on the alloy surfaces. Such an oxide could plausibly account for significantly improved tribological behaviour at higher oxygen levels. Sodium chromite is only marginally stable at typical reactor outlet conditions and frictional behaviour of typical chromium-containing alloys has therefore been studied as a function of rig cold trap temperature for exposure temperatures ranging from 650 to 500 0 C in order to define the effective tribological boundary. The behaviour of aluminised surfaces has also been studied and results from sliding and fretting wear tests are discussed in the context of the role of a lubricating oxide, believed to be sodium aluminate (formed by reaction of aluminium and oxygen-containing sodium) which is considerably more stable than sodium chromite at reactor outlet temperatures. (author)

  14. Mechanical behaviour of dental composite filling materials using digital holography

    OpenAIRE

    Monteiro, J.M.; Lopes, H.; Vaz, M.A.P.; Campos, J.C. Reis

    2010-01-01

    One of the most common clinical problems in dentistry is tooth decay. Among the dental filling materials used to repair tooth structure that has been destroyed by decay are dental amalgam and composite materials based on acrylics. Dental amalgam has been used by dentists for the past 150 years as a dental restorative material due to its low cost, ease of application, strength, durability, and bacteriostatic effects. However its safety as a filling material has been questioned due to th...

  15. Comparative evaluation of solar, fission, fusion, and fossil energy resources. Part 2: Power from nuclear fission

    Science.gov (United States)

    Clement, J. D.

    1973-01-01

    Different types of nuclear fission reactors and fissionable materials are compared. Special emphasis is placed upon the environmental impact of such reactors. Graphs and charts comparing reactor facilities in the U. S. are presented.

  16. New results concerning the behaviour of fission gases in in-pile UO{sub 2} at high temperatures; Resultats nouveaux sur le comportement des gaz de fission a haute temperature dans l'UO{sub 2} en pile

    Energy Technology Data Exchange (ETDEWEB)

    Soulhier, R; Schurenkamper, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The authors consider in the first part the various phenomena giving rise to the production of fission gases towards the exterior of nuclear fuels. The following aspects are dealt with: diffusion, for which is considered the influence of the predecessors of the radioactive gases, the fission recoil, atom expulsion along the fission paths and the evaporation. In the second part the authors present the results obtained on UO{sub 2} samples subjected to irradiation at temperatures of between 150 and 2000 deg C: - At low temperatures the variation of the amount produced as a function of the half-life of the isotopes studied shows that recoil is hot the only cause of gas production. - Above 1800 deg C, a weight loss by evaporation has been observed and the influence of this phenomenon on gas liberation has been studied; thus the fraction of {sup 135}Xe liberated at 2000 deg C by processes other than evaporation is of the order of 10 per cent. - The influence of the various mechanism on the overall effect as a function of temperature is discussed. (authors) [French] Dans une premiere partie, les auteurs etudient les differents phenomenes pouvant donner lieu au degagement des gaz de fission hors d'un combustible. Sont traites successivement: la diffusion, pour laquelle on discute l'influence des predecesseurs des gaz radioactifs, le recul de fission, l'expulsion des atomes le long des trajets de fission et l'evaporation. Dans une deuxieme partie ils exposent les resultats obtenus sur des echantillons d'UO{sub 2} portes sous irradiation a des temperatures comprises entre 150 deg C et 2000 deg C: - A basse temperature la variation de la quantite degagee suivant la periode des isotopes etudies montre que le recul n'est pas la seule cause du degagement des gaz. - Au-dessus de 1800 deg C on a note une perte de poids par evaporation et on a evalue l'influence de ce phenomene sur la liberation des gaz: ainsi la fraction du {sup 135}Xe liberee a 2000 deg C par d'autres processus

  17. Comparison of predicted and measured fission product behaviour in the Fort St. Vrain HTGR during the first three cycles of operation

    International Nuclear Information System (INIS)

    Hanson, D.L.; Jovanovic, V.; Burnette, R.D.

    1985-01-01

    The 330 MW(e) Fort St. Vrain (M) High Temperature Gas-Cooled Reactor (HTGR) is fueled with (Th,U)C 2 /ThC 2 TRISO-coated fuel particles contained in prismatic graphite fuel elements. Fission product release from the reactor core has been monitored during the first three cycles of operation. In order to assess the validity of the design methods used to predict fission product source terms for HTGRs, fission product release from the reactor core has been predicted by the reference design methods and compared with reactor surveillance measurements and with the results of postirradiation examination (PIE) of spent FSV fuel elements. Overall, the predictive methods have been shown to be conservative: the predicted fission gas release at the end of Cycle 3 is about five times higher than observed. The dominant source of fission gas release is as-manufactured, heavy-metal contamination; in-service failure of the coated fuel particles appears to be negligible, which is consistent with the PIE of spent fuel elements removed during the first two refuelings. The predicted releases of fission metals are insignificant compared to the release and subsequent decay of their gaseous precursors, which is consistent with plateout probe measurements. (author)

  18. Oxidation and creep behaviour of dense silicon nitride materials with different compositions

    International Nuclear Information System (INIS)

    Ernstberger, U.

    1985-09-01

    The study was intended to yield information on the oxidation and creep behaviour of Si 3 N 4 materials of different composition and microstructure, and produced by different processes. The experiments carried out in a vast temperature and load range showed that the chemical grain boundary composition is the key parameter affecting the materials' high-temperature properties. Significant correlations could be established between oxidation and creep behaviour on the one hand, and between microstructure and the behaviour on the other. (orig./IHOE) [de

  19. Activation analysis and waste management for blanket materials of multi-functional experimental fusion–fission hybrid reactor (FDS-MFX)

    International Nuclear Information System (INIS)

    Jiang, Jieqiong; Yuan, Baoxin; Zou, Jun; Wu, Yican

    2014-01-01

    The preliminary studies of the activation analysis and waste management for blanket materials of the multi-functional experimental fusion–fission hybrid reactor, i.e. Multi-Functional eXperimental Fusion Driven Subcritical system named FDS-MFX, were performed. The neutron flux of the FDS-MFX blanket was calculated using VisualBUS code and Hybrid Evaluated Nuclear Data Library (HENDL) developed by FDS Team. Based on these calculated neutron fluxes, the activation properties of blanket materials were analyzed by the induced radioactivity, the decay heat and the contact dose rate for different regions of the FDS-MFX blanket. The safety and environment assessment of fusion power (SEAFP) strategy, which was developed in Europe, was applied to FDS-MFX blanket for the management of activated materials. Accordingly, the classification and management strategy of activated materials after different cooling time were proposed for FDS-MFX blanket

  20. Guidelines for Applying Cohesive Models to the Damage Behaviour of Engineering Materials and Structures

    CERN Document Server

    Schwalbe, Karl-Heinz; Cornec, Alfred

    2013-01-01

    This brief provides guidance for the application of cohesive models to determine damage and fracture in materials and structural components. This can be done for configurations with or without a pre-existing crack. Although the brief addresses structural behaviour, the methods described herein may also be applied to any deformation induced material damage and failure, e.g. those occurring during manufacturing processes. The methods described are applicable to the behaviour of ductile metallic materials and structural components made thereof. Hints are also given for applying the cohesive model to other materials.

  1. Burning behaviour of surgical materials; Brandverhalten von chirurgischen textilen Materialien

    Energy Technology Data Exchange (ETDEWEB)

    Nowak, W.; Weinberg, L.; Grossewinkelmann, A.; Berlien, H.P. [Klinikum Neukoelln (Germany). Klinik fuer Lasermedizin

    2004-07-01

    Besides other energy driven devices like electrocautery and endoscopic light sources, also medical laser has the risk to induce operating theatre fire. On the market several so called laser safe materials, based on different technical solutions, are available. Materials which are not at the same time resistant against penetration and perforation by the laser beam have the risk of secondary ignition and combustion of underlaying materials. This should be kept in mind, when using a laser. With a standard for testing we irradiated several typical surgical materials from different suppliers with an CO{sub 2}-laser, observing their perforation, ignition or combustion behavior. (orig.)

  2. Analysis of writing and erasing behaviours in phase change materials

    Energy Technology Data Exchange (ETDEWEB)

    Hyot, B. E-mail: bhyot@cea.fr; Poupinet, L.; Gehanno, V.; Desre, P.J

    2002-09-01

    An understanding of the process involved in writing and erasing of phase-change optical recording media is vital to the development of new, and the improvement of existing, products. The present work investigates both experimental and theoretical laser-induced fast structural transformations of GeSbTe thin films. Optical and microstructural changes are correlated using both a static tester and transmission electron microscopy. In the second part of this paper we try to elucidate the physics underlying the amorphous-to-crystalline phase transformation under short-pulse laser excitation. Both thermal and thermodynamical behaviours must be taken into account to illustrate real processes.

  3. Analysis of writing and erasing behaviours in phase change materials

    International Nuclear Information System (INIS)

    Hyot, B.; Poupinet, L.; Gehanno, V.; Desre, P.J.

    2002-01-01

    An understanding of the process involved in writing and erasing of phase-change optical recording media is vital to the development of new, and the improvement of existing, products. The present work investigates both experimental and theoretical laser-induced fast structural transformations of GeSbTe thin films. Optical and microstructural changes are correlated using both a static tester and transmission electron microscopy. In the second part of this paper we try to elucidate the physics underlying the amorphous-to-crystalline phase transformation under short-pulse laser excitation. Both thermal and thermodynamical behaviours must be taken into account to illustrate real processes

  4. Directive of The Minister of Economic Affairs and the State Secretary for Social Affairs and Public Health of 5 December 1969, Stcrt. 240 in implementation of Section 2 and other Sections of the Fissionable Materials, Ores and Radioactive Materials (Transport) Decree (Designation of Countries)

    International Nuclear Information System (INIS)

    1969-01-01

    This Directive designates the countries which are parties to the same international transport agreements as the Netherlands and which may therefore transport fissionable materials, ores and radioactive materials over Netherlands territory and territorial waters. (NEA) [fr

  5. Comparison of fission probabilities with emission of long range particles under the action of slow and fast neutrons on various materials; Probabilites comparees de fission avec emission de particules de long parcours pour divers materiaux sous l'action des neutrons lents et rapides

    Energy Technology Data Exchange (ETDEWEB)

    Netter, F; Faraggi, H; Garin-Bonnet, A; Julien, J; Corge, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Turkiewicz, J [Institut de Recherches Nucleaire de Varsovie (Poland)

    1958-07-01

    The authors describe relative cross-section measurements of fission of the isotopes of uranium and plutonium (more particularly {sup 235}U, {sup 238}U, {sup 239}Pu), with emission of long range particles, under the action of neutrons of various energies: thermal neutrons, pile neutrons, neutrons produced with the Van de Graaff accelerator by reaction of protons on tritium. The measurements are carried out: 1) with the aid of photographic plates, by submitting to the action of the neutrons a layer of fissile material coupled with an Ilford nuclear emulsion of 200 microns; a tin sheet laying between the plate and the layer stops the {alpha} particles and the fission fragments. By an appropriate development the tracks of the long range particles can be distinguished in the emulsion, from the tracks of the recoil protons resulting of fission neutrons, or of the last primary neutrons. For neutrons of energy under 1 MeV, the compared frequency of the tracks of long range particles and of the recoils caused by the fission neutrons gives a measurement of the fission cross-section with emission of long range particles relative to the product of the fission cross-section by the mean number of neutrons emitted by fission. For neutrons of higher energy, one measures only the frequency of the tracks of long range particles, comparatively with the flux of primary neutrons. Some precautions are taken to eliminate the action of thermal neutrons in the measurements with fast neutrons. 2) with the aid of a system of ionization chamber and proportional counter, the rate of coincidence between the impulsions caused by the long range particles and the impulsions provided by one of the fission fragments is measured comparatively with the counting rate of fission fragme (author) [French] Les auteurs decrivent des mesures relatives a la section efficace de fission des isotopes de l'uranium et du plutonium (notamment {sup 235}U, {sup 238}U, {sup 239}Pu) avec emission de particules de long

  6. Fission track analysis of Pu in small specimens of biological material: Technical progress report, August 1, 1987--July 31, 1988

    International Nuclear Information System (INIS)

    Wrenn, M.E.

    1988-01-01

    The objective of this research is to develop a highly specific and ultrasensitive method capable of detecting 100 aCi/liter of 239 Pu in human urine. The method using neutron induced fission track analysis is to be made free of interference from uranium, the only naturally occurring element with an isotope which fissions with thermal neutrons. A simplified flow diagram for the method is shown in Figure 1. Briefly 239 Pu is coprecipitated quantitatively from urine with rhodozonic acid. The precipitate containing the 239 Pu is dissolved in HCl and is sequentially passed through two ion exchange columns and reduced in volume. The element is then deposited in a circular area on a thick polycarbonate detector and a thinner detector is placed over the circular deposit. The plastic detectors are then irradiated to a high thermal neutron fluence in a research reactor. The detectors are etched in a caustic solution for controlled times and temperatures in order to develop the fission tracks. Images of tracks are formed both on the thin and thick plastic detectors. Total tracks in the thinner detector are measured with a locally developed spark counter and in the thick plastic are measured by counting with a microscope. The results will be made quantitative by constructing a calibration curve for 239 Pu. 3 refs., 9 figs., 3 tabs

  7. Thermal shock behaviour of mullite-cordierite refractory materials

    Czech Academy of Sciences Publication Activity Database

    Boccaccini, D. N.; Leonelli, C.; Romagnoli, M.; Pellacani, G. C.; Veronesi, P.; Dlouhý, Ivo; Boccaccini, A. R.

    2007-01-01

    Roč. 106, č. 3 (2007), s. 142-148 ISSN 1743-6753 R&D Projects: GA AV ČR IAA200410502 Institutional research plan: CEZ:AV0Z20410507 Keywords : refraktory materials * thermal shock * fracutre toughness Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass Impact factor: 1.074, year: 2007

  8. Ternary fission

    Indian Academy of Sciences (India)

    the energy minimization of all possible ternary breakups of a heavy radioactive nucleus. Further, within the TCM we have analysed the competition between different geometries as well as different positioning of the fragments. Also, an attempt was made to calculate the mass distribution of ternary fission process within the ...

  9. Corrosion behaviour of materials selected for FMIT lithium system

    Energy Technology Data Exchange (ETDEWEB)

    Bazinet, G.D.; Brehm, W.F.

    1983-09-01

    The corrosion behavior of selected materials in a liquid lithium environment was studied in support of system and component designs for the Fusion Materials Irradiation Test (FMIT) Facility. Testing conditions ranged from about 3700 to about6500 hours of exposure to flowing lithium at temperatures from 230/sup 0/ to 270/sup 0/C and static lithium at temperatures from 200/sup 0/ to 500/sup 0/C. Principal areas of investigation included lithium corrosion/erosion effects on FMIT lithium system baseline and candidate materials. Material coupons and full-size prototypic components were evaluated to determine corrosion rates, fatigue crack growth rates, structural compatibility, and component acceptability for the lithium system. Based on the results of these studies, concerns regarding system materials and component designs were satisfactorily resolved to support a 20-year design life requirement for the FMIT lithium system.

  10. Corrosion behaviour of materials selected for FMIT lithium system

    International Nuclear Information System (INIS)

    Bazinet, G.D.; Brehm, W.F.

    1983-01-01

    The corrosion behavior of selected materials in a liquid lithium environment was studied in support of system and component designs for the Fusion Materials Irradiation Test (FMIT) Facility. Testing conditions ranged from about 3700 to about6500 hours of exposure to flowing lithium at temperatures from 230 0 to 270 0 C and static lithium at temperatures from 200 0 to 500 0 C. Principal areas of investigation included lithium corrosion/erosion effects on FMIT lithium system baseline and candidate materials. Material coupons and full-size prototypic components were evaluated to determine corrosion rates, fatigue crack growth rates, structural compatibility, and component acceptability for the lithium system. Based on the results of these studies, concerns regarding system materials and component designs were satisfactorily resolved to support a 20-year design life requirement for the FMIT lithium system

  11. Comparison of nuclear irradiation parameters of fusion breeder materials in high flux fission test reactors and a fusion power demonstration reactor

    International Nuclear Information System (INIS)

    Fischer, U.; Herring, S.; Hogenbirk, A.; Leichtle, D.; Nagao, Y.; Pijlgroms, B.J.; Ying, A.

    2000-01-01

    Nuclear irradiation parameters relevant to displacement damage and burn-up of the breeder materials Li 2 O, Li 4 SiO 4 and Li 2 TiO 3 have been evaluated and compared for a fusion power demonstration reactor and the high flux fission test reactor (HFR), Petten, the advanced test reactor (ATR, INEL) and the Japanese material test reactor (JMTR, JAERI). Based on detailed nuclear reactor calculations with the MCNP Monte Carlo code and binary collision approximation (BCA) computer simulations of the displacement damage in the polyatomic lattices with MARLOWE, it has been investigated how well the considered HFRs can meet the requirements for a fusion power reactor relevant irradiation. It is shown that a breeder material irradiation in these fission test reactors is well suited in this regard when the neutron spectrum is well tailored and the 6 Li-enrichment is properly chosen. Requirements for the relevant nuclear irradiation parameters such as the displacement damage accumulation, the lithium burn-up and the damage production function W(T) can be met when taking into account these prerequisites. Irradiation times in the order of 2-3 full power years are necessary for the HFR to achieve the peak values of the considered fusion power Demo reactor blanket with regard to the burn-up and, at the same time, the dpa accumulation

  12. Description of scattering material behaviour and damage in inelastic materials; Beschreibung von streuendem Materialverhalten und von Schaedigung bei inelastischen Werkstoffen

    Energy Technology Data Exchange (ETDEWEB)

    Pensky, H.M.H.

    2000-07-01

    For realistic numerical simulations of the stress-strain behaviour of structures, models are necessary which describe elastic-inelastic and scattering material behaviour. The developed models simulate elastic, viscoplastic and anisotropic damage material phenomena. An approach is proposed for covering stochastic material beahviour by correspondingly distributed parameters of the deterministic material model. Numerical simulations of biaxial material tests and structural tests demonstrate the range of applicability. (orig.) [German] Die realitaetsnahe numerische Simulation des Spannungs-Verformungsverhaltens von Bauteilen erfordert Modelle zur Beschreibung inelastischen und streuenden Materialverhaltens. Die hier entwickelten Modelle beschreiben elastische, viskoplastische und anisotrope Schaedigungsphaenomene des Materialverhaltens. Desweiteren wird ein Konzept vorgestellt, mit dem streuendes Materialverhalten mit streuenden Materialparametersaetzen deterministischer Stoffmodelle beschreibbar ist. Numerische Simulationen von Werkstoff- und Bauteilversuchen veranschaulichen den Anwendungsbereich der Modelle. (orig.)

  13. Advanced BorobondTM Shields for Nuclear Materials Containment and BorobondTM Immobilization of Volatile Fission Products - Final CRADA Report

    International Nuclear Information System (INIS)

    Wagh, Arun S.

    2016-01-01

    Borobond is a company-proprietary material developed by the CRADA partner in collaboration with Argonne, and is based on Argonne's Ceramicrete technology. It is being used by DOE for nuclear materials safe storage, and Boron Products, LLC is the manufacturer and supplier of Borobond. The major objective of this project was to produce a more versatile composition of this material and find new applications. Major target applications were use for nuclear radiation shields, such as in dry storage casks; use in immobilization of most difficult waste streams, such as Hanford K-Basin waste; use for soluble and volatile fission products, such as Cs, Tc, Sr, and I; and use for corrosion and fire protection applications in nuclear facilities.

  14. Rheological behaviour of fibre-rich plant materials in fat-based food systems

    NARCIS (Netherlands)

    Bonarius, G.A.; Vieira, J.B.; Goot, van der A.J.; Bodnar, I.

    2014-01-01

    The potential use of fibre-rich materials as bulking agents to replace sucrose in chocolate confectionary products is investigated. Since the rheological behaviour of the molten chocolate mass is key in chocolate production, the rheology of fibre-rich materials in medium chain triglycerides (MCT) is

  15. Mechanical behaviour of new zirconia-hydroxyapatite ceramic materials

    Energy Technology Data Exchange (ETDEWEB)

    Delgado, J.A.; Morejon, L. [La Habana Univ. (Cuba). Centro de Biomateriales; Martinez, S. [Barcelona Univ. (Spain). Dept. Cristallografia, Mineralogia; Ginebra, M.P.; Carlsson, N.; Fernandez, E.; Planell, J.A. [Universidad Politecnica de Cataluna, Barcelona (Spain). CREB; Clavaguera-Mora, M.T.; Rodriguez-Viejo, J. [Universitat Autonoma de Barcelona (Spain). Dept. de Fisica

    2001-07-01

    In this work a new zirconia-hydroxyapatite ceramic material was obtained by uniaxial pressing and sintering in humid environment. The powder X-ray diffraction (XRD) patterns and infrared spectra (FT-IR) showed that the hydroxyapatite (HA) is the only calcium phosphate phase present. The fracture toughness for HA with 20 wt.% of magnesia partially stabilised zirconia (Mg-PSZ) was around 2.5 times higher than those obtained for HA pure, also the highest value of bending strength (160 MPa) was obtained for material reinforced with Mg-PSZ. For the MgPSZ-HA (20%) the fracture mechanism seems to be less transgranular. (orig.)

  16. Consultancy on the potential of fusion/fission sub-critical neutron systems for energy production and transmutation. Working material

    International Nuclear Information System (INIS)

    2005-01-01

    The Workshop on Sub-critical Neutron Production held at the University of Maryland and the Eisenhower Institute on 11-13 October 2004 brought together members of fusion, fission and accelerator technical communities to discuss issues of spent fuel, nonproliferation, reactor safety and the use of neutrons for sub-critical operation of nuclear reactors. The Workshop strongly recommended that the fusion community work closely with other technical communities to ensure that a wider range of technical solutions is available to solve the spent fuel problem and to utilize the current actinide inventories. Participants of the Workshop recommended that a follow-on Workshop, possibly under the aegis of the IAEA, should be held in the first half of the year 2005. The Consultancy Meeting is the response to this recommendation. The objectives of the Consultancy meeting were to hold discussions on the role of fusion/fission systems in sub-critical operations of nuclear reactors. The participants agreed that development of innovative (fourth generation) fission reactors, advanced fuel cycle options, and disposition of existing spent nuclear fuel inventories in various Member Sates can significantly benefit from including sub-critical systems, which are driven by external neutron sources. Spallation neutrons produced by accelerators have been accepted in the past as the means of driving sub-critical reactors. The accelerator community deserves credit in pioneering this novel approach to reactor design. Progress in the design and operation of fusion devices now offers additional innovative means, broadening the range of sub-critical operations of fission reactors. Participants felt that fusion should participate with accelerators in providing a range of technical options in reactor design. Participants discussed concrete steps to set up a small fusion/fission system to demonstrate actinide burning in the laboratory and what advice should be given to the Agency on its role in

  17. Ductile fracture behaviour of primary heat transport piping material ...

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    Abstract. Design of primary heat transport (PHT) piping of pressurised heavy water reactors (PHWR) has to ensure implementation of leak-before-break con- cepts. In order to be able to do so, the ductile fracture characteristics of PHT piping material have to be quantified. In this paper, the fracture resistance of SA333, Grade.

  18. Transitional behaviour of thickness effects in shipbuilding materials (MS plate)

    Science.gov (United States)

    Mahmud, S. M. Ikhtiar; Razib, Amirul Hasan; Rahman, Md. Rabab Raiyatur

    2017-12-01

    Majority of the crack propagation in ships and offshore structures are caused due to fatigue. Previously, it was known that fatigue strength of notched specimen is dependent on size, but recently it came to light that fatigue strength of some welded joints depends on the thickness. Much investigation is done on the fatigue growth of welded joints. Fatigue often results in fracture accidents, which starts from the sites of structural discontinuities because of the reason that they may induce local stress concentrations. Structural discontinuities include notches, holes, sharp corners, and weld defects. Weld defects include undercut, porosity, lack of fusion, slag inclusion, incomplete weld root penetration, and misalignments. In order to investigate the effects of plate thickness on fatigue strength, semi-elliptical side notches (U and V shaped) in plates are studied in the present research. First consider a simple problem of crack emanating from notches in plates where the solution of stress intensity factor is given by an empirical formula so that the thickness effect on fatigue strength can easily be investigated for a variety of geometrical parameters. The present study aims to investigate the transitional behaviour of thickness effect in plates on fatigue strength. In order to calculate the stress, finite element analysis is carried by using ANSYS.

  19. On the hydro-mechanical behaviour of MX80 bentonite-based materials

    Directory of Open Access Journals (Sweden)

    Yu-Jun Cui

    2017-06-01

    Full Text Available Bentonite-based materials have been considered in many countries as engineered barrier/backfilling materials in deep geological disposal of high-level radioactive waste. During the long period of waste storage, these materials will play an essential role in ensuring the integrity of the storage system that consists of the waste canisters, the engineered barrier/backfill, the retaining structures as well as the geological barrier. Thus, it is essential to well understand the hydro-mechanical behaviours of these bentonite-based materials. This review paper presents the recent advances of knowledge on MX80 bentonite-based materials, in terms of water retention properties, hydraulic behaviour and mechanical behaviour. Emphasis is put on the effect of technological voids and the role of the dry density of bentonite. The swelling anisotropy is also discussed based on the results from swelling tests with measurements of both axial and radial swelling pressures on a sand-bentonite mixture compacted at different densities. Microstructure observation was used to help the interpretation of macroscopic hydro-mechanical behaviour. Also, the evolution of soil microstructure thus the soil density over time is discussed based on the results from mock-up tests. This evolution is essential for understanding the long-term hydro-mechanical behaviour of the engineered barrier/backfill.

  20. The effect of the dislocation image force on the brittle behaviour of materials

    International Nuclear Information System (INIS)

    Lung, C.W.

    1986-06-01

    The dislocation image force due to the free surface of a finite width specimen makes the plastic zone at a crack tip larger. The effect of the dislocation image force on the fracture behaviour of materials with different geometrical shapes is discussed. It is found that the ratio V/A as an indication of the brittle behaviour of structural components is reasonable for elastic-plastic fracture. (author)

  1. Long-term behaviour of heat-resistant steels and high-temperature materials

    International Nuclear Information System (INIS)

    1987-01-01

    This book contains 10 lectures with the following subjects: On the effect of thermal pretreatment on the structure and creep behaviour of the alloy 800 H (V. Guttmann, J. Timm); Material properties of heat resistant ferritic and austenitic steels after cold forming (W. Bendick, H. Weber); Investigations for judging the working behaviour of components made of alloy 800 and alloy 617 under creep stress (H.J. Penkalla, F. Schubert); Creep behaviour of gas turbine materials in hot gas (K.H. Kloos et al.); Effect of small cold forming on the creep beahviour of gas turbine blades made of Nimonic 90 (K.H. Keienburg et al.); Investigations on creep fatigue alternating load strength of nickel alloys (G. Raule); Change of structure, creep fatigue behaviour and life of X20 Cr Mo V 12 1 (by G. Eggeler et al.); Investigations on thermal fatigue behaviour (K.H. Mayer et al.); Creep behaviour of similar welds of the steels 13 Cr Mo 4 4, 14 MoV 6 3, 10 Cr Mo 910 and GS-17 Cr Mo V 5 11 (K. Niel et al.); Determining the creep crack behaviour of heat resistant steels with samples of different geometry (K. Maile, R. Tscheuschner). (orig.,/MM) [de

  2. Material pre-conditioning effects on the creep behaviour of 316H stainless steel

    International Nuclear Information System (INIS)

    Mehmanparast, A.; Davies, C.M.; Dean, D.W.; Nikbin, K.

    2013-01-01

    Material pre-conditioning by, for example, pre-strain through component bending and welding is known to alter the creep deformation and creep crack growth (CCG) behaviour of 316H stainless steel. Experimental test data on the creep deformation and crack growth behaviour of 316H weldment compact tension specimens at 550 °C, where the starter defect was introduced into the heat affected zone (HAZ), have been compared to those of obtained from similar specimens manufactured from parent material, which had been subjected to 8% compressive plastic pre-strain at room temperature. Similar degrees of accelerated cracking behaviour compared to parent material, for given values of C*, were exhibited in both 316H HAZ and pre-compressed parent materials. This acceleration has been attributed to the influence of material hardening effects and the reduction of creep ductility in the pre-conditioned materials. These results are discussed in terms of the potential for using material pre-conditioning to assist in predicting the long term cracking behaviour of high temperature 316H stainless steel plant components from shorter term laboratory CCG tests

  3. Modeling the behaviour of shape memory materials under large deformations

    Science.gov (United States)

    Rogovoy, A. A.; Stolbova, O. S.

    2017-06-01

    In this study, the models describing the behavior of shape memory alloys, ferromagnetic materials and polymers have been constructed, using a formalized approach to develop the constitutive equations for complex media under large deformations. The kinematic and constitutive equations, satisfying the principles of thermodynamics and objectivity, have been derived. The application of the Galerkin procedure to the systems of equations of solid mechanics allowed us to obtain the Lagrange variational equation and variational formulation of the magnetostatics problems. These relations have been tested in the context of the problems of finite deformation in shape memory alloys and ferromagnetic materials during forward and reverse martensitic transformations and in shape memory polymers during forward and reverse relaxation transitions from a highly elastic to a glassy state.

  4. Characterization of baking behaviour of carbonaceous materials by dilatation investigations

    Energy Technology Data Exchange (ETDEWEB)

    Born, M.; Seichter, A.; Starke, S.

    1990-01-01

    An increase in volume can be observed in carbonaceous materials during baking which is assumed to be the reason for strains and crack formation. It occurs most pronouncedly within a temperature range from 100 to 200{degree}C. The causes of such phenomena in products pressed at different temperatures are analyzed by means of a gas pressure model and a relaxation model. The factors influencing dilatation are subject to thermal analysis. 15 refs., 13 figs.

  5. Modelling isothermal fission gas release

    International Nuclear Information System (INIS)

    Uffelen, P. van

    2002-01-01

    The present paper presents a new fission gas release model consisting of two coupled modules. The first module treats the behaviour of the fission gas atoms in spherical grains with a distribution of grain sizes. This module considers single atom diffusion, trapping and fission induced re-solution of gas atoms associated with intragranular bubbles, and re-solution from the grain boundary into a few layers adjacent to the grain face. The second module considers the transport of the fission gas atoms along the grain boundaries. Four mechanisms are incorporated: diffusion controlled precipitation of gas atoms into bubbles, grain boundary bubble sweeping, re-solution of gas atoms into the adjacent grains and gas flow through open porosity when grain boundary bubbles are interconnected. The interconnection of the intergranular bubbles is affected both by the fraction of the grain face occupied by the cavities and by the balance between the bubble internal pressure and the hydrostatic pressure surrounding the bubbles. The model is under validation. In a first step, some numerical routines have been tested by means of analytic solutions. In a second step, the fission gas release model has been coupled with the FTEMP2 code of the Halden Reactor Project for the temperature distribution in the pellets. A parametric study of some steady-state irradiations and one power ramp have been simulated successfully. In particular, the Halden threshold for fission gas release and two simplified FUMEX cases have been computed and are summarised. (author)

  6. Fatigue behaviour of metallic materials; Ermuedungsverhalten metallischer Werkstoffe

    Energy Technology Data Exchange (ETDEWEB)

    Christ, H.J. [ed.

    1998-12-31

    The 16 contributions selected for this book, each from experts in their fields, are intended to give a broad survey of the phenomenon and mechanisms of fatigue in metallic materials, addressing important aspects and showing the cross-disciplinarity of scientific research required to obtain a complete picture. Emphasis has been placed on the matter being discussed in a way that is easy to digest as well as complete in information, which was possible only by deliberate restriction to the essential knowledge available today, leaving aside what recent scientific research may have revealed, or whatever interesting specific aspects there may be. The known mechanisms of fatigue and their effects in metallic materials as well as the conclusions to be drawn from the engineering angle with regard to the applicability of the materials and systems design are the points of main interest of the book, which offers readers to develop a sound, general understanding of the processes involved and a feeling for the effects induced in the materiuals by cyclic stress. (orig./CB) [Deutsch] In diesem 16 Fachbeitraege enthaltenden Buch wird versucht, einen ueberschau- und erfassbaren Ueberblick ueber die Ermuedung metallischer Werkstoffe unter Beruecksichtigung der wichtigen Teilaspekte und Wissenschaftsgebiete darzustellen. Die Betonung wird bewusst auf Verstaendlichkeit und Uebersichtlichkeit gelegt, was nur durch Einschraenkung der Breite der Behandlung und durch Verzicht auf neueste wissenschaftliche Details moeglich ist. Im Vordergrund stehen die bei der Ermuedung ablaufenden werkstoffkundlichen Vorgaenge und die sich daraus ergebenden Konsequenzen fuer den Werkstoffeinsatz und die -auslegung. Primaer soll ein solides Grundverstaendnis fuer die moeglichen Prozesse vermittelt werden, aus dem sich ein Gefuehl fuer die Vorgaenge im Werkstoff bei zyklischer Beanspruchung entwickeln kann. (orig.)

  7. Influence of the recycled material percentage on the rheological behaviour of HDPE for injection moulding process.

    Science.gov (United States)

    Javierre, C; Clavería, I; Ponz, L; Aísa, J; Fernández, A

    2007-01-01

    The amount of polymer material wasted during thermoplastic injection moulding is very high. It comes from both the feed system of the part, and parts necessary to set up the mould, as well as the scrap generated along the process due to quality problems. The residues are managed through polymer recycling that allows reuse of the materials in the manufacturing injection process. Recycling mills convert the parts into small pieces that are used as feed material for injection, by mixing the recycled feedstock in different percentages with raw material. This mixture of both raw and recycled material modifies material properties according to the percentage of recycled material introduced. Some of the properties affected by this modification are those related to rheologic behaviour, which strongly conditions the future injection moulding process. This paper analyzes the rheologic behaviour of material with different percentages of recycled material by means of a capillary rheometer, and evaluates the influence of the corresponding viscosity curves obtained on the injection moulding process, where small variations of parameters related to rheological behaviour, such as pressure or clamping force, can be critical to the viability and cost of the parts manufactured by injection moulding.

  8. Corrosion behaviour of construction materials for high temperature steam electrolysers

    DEFF Research Database (Denmark)

    Nikiforov, Aleksey; Petrushina, Irina; Christensen, Erik

    2011-01-01

    temperature proton exchange membrane (PEM) steam electrolysers. Steady-state voltammetry was used in combination with scanning electron microscopy and energy-dispersive X-ray spectroscopy to evaluate the stability of the mentioned materials. It was found that stainless steels were the least resistant...... to corrosion under strong anodic polarisation. Among alloys, Ni-based showed the highest corrosion resistance in the simulated PEM electrolyser medium. In particular, Inconel 625 was the most promising among the tested corrosion-resistant alloys for the anodic compartment in high temperature steam electrolysis...

  9. Mesoscopic approach to modeling elastic-plastic polycrystalline material behaviour

    International Nuclear Information System (INIS)

    Kovac, M.; Cizelj, L.

    2001-01-01

    Extreme loadings during severe accident conditions might cause failure or rupture of the pressure boundary of a reactor coolant system. Reliable estimation of the extreme deformations can be crucial to determine the consequences of such an accident. One of important drawbacks of classical continuum mechanics is idealization of inhomogenous microstructure of materials. This paper discusses the mesoscopic approach to modeling the elastic-plastic behavior of a polycrystalline material. The main idea is to divide the continuum (e.g., polycrystalline aggregate) into a set of sub-continua (grains). The overall properties of the polycrystalline aggregate are therefore determined by the number of grains in the aggregate and properties of randomly shaped and oriented grains. The random grain structure is modeled with Voronoi tessellation and random orientations of crystal lattices are assumed. The elastic behavior of monocrystal grains is assumed to be anisotropic. Crystal plasticity is used to describe plastic response of monocrystal grains. Finite element method is used to obtain numerical solutions of strain and stress fields. The analysis is limited to two-dimensional models.(author)

  10. Corrosion behaviour of construction materials for high temperature water electrolysers

    Energy Technology Data Exchange (ETDEWEB)

    Nikiforov, A.; Petruchina, I.; Christensen, E.; Bjerrum, N.J.; Tomas-Garcya, A.L. [Technical Univ. of Denmark, Lyngby (Denmark). Dept. of Chemistry, Materials Science Group

    2010-07-01

    This presentation reported on a study in which the feasibility of using different corrosion resistant stainless steels as a possible metallic bipolar plate and construction material was evaluated in terms of corrosion resistance under conditions corresponding to the conditions in high temperature proton exchange membrane (PEM) water electrolysers (HTPEMWE). PEM water electrolysis technology has been touted as an effective alternative to more conventional alkaline water electrolysis. Although the energy efficiency of this technology can be increased considerably at temperatures above 100 degrees C, this increases the demands to all the used materials with respect to corrosion stability and thermal stability. In this study, Ni-based alloys as well as titanium and tantalum samples were exposed to anodic polarization in 85 per cent phosphoric acid electrolyte solution. Tests were performed at 80 and 120 degrees C to determine the dependence of corrosion speed and working temperature. Platinum and gold plates were also tested for a comparative evaluation. Steady-state voltammetry was used along with scanning electron microscopy and energy-dispersive X-ray spectroscopy. Titanium showed the poorest corrosion resistance, while Ni-based alloys showed the highest corrosion resistance, with Inconel R 625 being the most promising alloy for the bipolar plate of an HTPEMWE. 3 refs., 1 tab., 2 figs.

  11. Selfwelding, friction and wear behaviour of special materials in sodium under corroding conditions

    International Nuclear Information System (INIS)

    Borgstedt, H.U.; Mattes, K.; Wild, E.

    1975-11-01

    Control rod guides and fuel element duct load pads have to be fabricated from materials exhibiting optimum slide behaviour. Galling or self-welding under static conditions should not be tolerated. Given bearing clearances have to be maintained constant and loop contamination, caused by wear particles, have to be prevented. Since high friction between contacting pads may impose severe limitations on core compaction, for the duct load pads a maximum friction coefficient of 0.5 is acceptable. The effect of sodium corrosion should not impair the friction and wear behaviour of the materials applied. This report covers the work performed to optain appropriate mechanical design data. (orig.) [de

  12. Who is reducing their material consumption and why? A cross-cultural analysis of dematerialization behaviours.

    Science.gov (United States)

    Whitmarsh, Lorraine; Capstick, Stuart; Nash, Nicholas

    2017-06-13

    The environmental and economic imperatives to dematerialize economies, or 'do more with less', have been established for some years. Yet, to date, little is known about the personal drivers associated with dematerializing. This paper explores the prevalence and profile of those who are taking action to reduce consumption in different cultural contexts (UK and Brazil) and considers influences on dematerialization behaviours. We find that exemplar behaviours (avoiding buying new things and avoiding packaging) are far less common than archetypal environmental behaviours (e.g. recycling), but also that cultural context is important (Brazilians are more likely to reduce their material consumption than people in the UK). We also find that the two dematerialization behaviours are associated with different pro-environmental actions (more radical action versus green consumption, respectively); and have distinct, but overlapping, psychological (e.g. identity) and socio-demographic (e.g. education) predictors. Comparing a more traditional value-identity model of pro-environmental behaviour with a motivation-based (self-determination) model, we find that the latter explains somewhat more variance than the former. However, overall, little variance is explained, suggesting that additional factors at the personal and structural levels are important for determining these consumption behaviours. We conclude by outlining policy implications and avenues for further research.This article is part of the themed issue 'Material demand reduction'. © 2017 The Author(s).

  13. Thermoresponsive behaviour of AM2O8 materials

    International Nuclear Information System (INIS)

    Allen, Simon

    2003-01-01

    This thesis investigates the synthesis and structural characterisation of AM 2 O 8 phases, many of which show negative thermal expansion (NTE); relevant literature is reviewed in Chapter One. Chapter Two describes the synthesis, structure solution, and mechanistic role of a new family of low-temperature (LT) orthorhombic AM 2 O 8 polymorphs (A IV = Zr, Hf; M VI = Mo, W). These materials are key intermediates in the preparation of cubic AM 2 O 8 phases from AM 2 O 7 (OH) 2 (H 2 O) 2 . The structure of LT-AM 2 O 8 has been elucidated by combined laboratory X-ray and neutron powder diffraction. Variable temperature X-ray diffraction (VTXRD) studies have shown LT-AMo 2 O 8 phases exhibit anisotropic NTE. LT-ZrMo 2 O 8 has been shown to undergo spontaneous rehydration, allowing preparation of ZrMo 2 O 7 (OD) 2 (D 2 O) 2 and assignment of D 2 O/OD positions within the structure by neutron diffraction. Using this result, a reversible topotactic dehydration pathway from AM 2 O 7 (OH) 2 (H 2 O) 2 to LT-AM 2 O 8 is proposed. Chapter Three investigates the order-disorder phase transition with concurrent oxygen mobility in cubic AM 2 O 8 materials; studies include comprehensive VT neutron diffraction of cubic ZrMo 2 O 8 to reveal a static to dynamic transition at 215 K, and novel quench-anneal/quench-warm variable temperature/time diffraction experiments on ZrWMoO 8 which lead to an activation energy of 40 kJmol -1 for oxygen migration. In Chapter Four 17 O-labelled cubic ZrW 2 O 8 has been prepared to understand the oxygen migration process by VT MAS NMR. In situ hydrothermal studies of cubic ZrMo 2 O 8 using synchrotron radiation have shown direct hydration to ZrMo 2 O 7 (OH) 2 (H 2 O) 2 . In Chapter Five VTXRD of trigonal α-AMo 2 O 8 phases reveals a previously unknown second-order phase transition at 487 K (A = Zr) or 463 K (A = Hf) from P3-bar 1c to P3-bar m1. Rigid-body Rietveld refinements have shown this is due to alignment of apical Mo-O groups with the c axis in the

  14. Enhanced Generic Phase-field Model of Irradiation Materials: Fission Gas Bubble Growth Kinetics in Polycrystalline UO2

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert O.; Gao, Fei; Sun, Xin

    2012-05-30

    Experiments show that inter-granular and intra-granular gas bubbles have different growth kinetics which results in heterogeneous gas bubble microstructures in irradiated nuclear fuels. A science-based model predicting the heterogeneous microstructure evolution kinetics is desired, which enables one to study the effect of thermodynamic and kinetic properties of the system on gas bubble microstructure evolution kinetics and morphology, improve the understanding of the formation mechanisms of heterogeneous gas bubble microstructure, and provide the microstructure to macroscale approaches to study their impact on thermo-mechanical properties such as thermo-conductivity, gas release, volume swelling, and cracking. In our previous report 'Mesoscale Benchmark Demonstration, Problem 1: Mesoscale Simulations of Intra-granular Fission Gas Bubbles in UO2 under Post-irradiation Thermal Annealing', we developed a phase-field model to simulate the intra-granular gas bubble evolution in a single crystal during post-irradiation thermal annealing. In this work, we enhanced the model by incorporating thermodynamic and kinetic properties at grain boundaries, which can be obtained from atomistic simulations, to simulate fission gas bubble growth kinetics in polycrystalline UO2 fuels. The model takes into account of gas atom and vacancy diffusion, vacancy trapping and emission at defects, gas atom absorption and resolution at gas bubbles, internal pressure in gas bubbles, elastic interaction between defects and gas bubbles, and the difference of thermodynamic and kinetic properties in matrix and grain boundaries. We applied the model to simulate gas atom segregation at grain boundaries and the effect of interfacial energy and gas mobility on gas bubble morphology and growth kinetics in a bi-crystal UO2 during post-irradiation thermal annealing. The preliminary results demonstrate that the model can produce the equilibrium thermodynamic properties and the morphology of gas

  15. Consultancy to review and finalize the IAEA publication 'Compendium on the use of fusion/fission hybrids for the utilization and transmutation of actinides and long-lived fission products'. Working material

    International Nuclear Information System (INIS)

    2004-01-01

    In addition to the traditional fission reactor research, fusion R and D activities are becoming of interest also to nuclear fission power development. There is renewed interest in utilizing fusion neutrons, Heavy Liquid Metals, and molten salts for innovative systems (energy production and transmutation). Indeed, for nuclear power development to become sustainable as a long-term energy option, innovative fuel cycle and reactor technologies will have to be developed to solve the problems of resource utilization and long-lived radioactive waste management. In this context Member States clearly expressed the need for comparative assessments of various transmutation reactors. Both the fusion and fission communities are currently investigating the potential of innovative reactor and fuel cycle strategies that include a fusion/fission system. The attention is mainly focused on substantiating the potential advantages of such systems: utilization and transmutation of actinides and long-lived fission products, intrinsic safety features, enhanced proliferation resistance, and fuel breeding capabilities. An important aspect of the ongoing activities is the comparison with the accelerator driven subcritical system (spallation neutron source), which is the other main option for producing excess neutrons. Apart from comparative assessments, knowledge preservation is another subject of interest to the Member States: the goal, applied to fusion/fission systems, is to review the status of, and to produce a 'compendium' of past and present achievements in this area

  16. Ideological Fission

    DEFF Research Database (Denmark)

    Christiansen, Steen Ledet

    ; it is a materialisation of an ideological fission which attempts to excise certain ideological constructions, yet paradoxically casting them in a form that is recognizable and familiar. The monstrous metonomy which is used shows us glimpses of a horrid being, intended to vilify the attack on New York City. However......, it is a being which is reminiscent of earlier monsters - from Godzilla to The Blob. It is evident that the Cloverfield monster is a paradoxical construction which attempts to articulate fear and loathing about terrorism, but ends up trapped in an ideological dead-end maze, unable to do anything other than...

  17. Book of abstracts of the joint EC-IAEA topical meeting on development of new structural materials for advanced fission and fusion reactor systems

    International Nuclear Information System (INIS)

    2009-01-01

    Materials performance and reliability are key issues for the safety and competitiveness of future nuclear installations: Generation IV nuclear systems for increased sustainability, advanced systems for non-electrical uses of nuclear energy, partitioning and transmutation systems, as well as thermo-nuclear fusion systems. These systems will have to feature high thermal efficiency and optimized utilization of fuel combined with minimized nuclear waste. For the sustainability of the nuclear option, there is a renewed interest worldwide in new reactor systems, closed fuel cycle research and technology development, and nuclear process heat applications. This requires the development and qualification of new high temperature structural materials with improved radiation and corrosion resistance. To achieve the challenging materials performance parameters, focused research and targeted testing of new candidate materials are necessary. Recent developments regarding new classes of materials with improved microstructural features, such as fibre-reinforced ceramic composite materials, oxide dispersion strengthened steels or advanced ferritic-martensitic steels are promising since they combine good radiation resistance and corrosion properties with high-temperature strength and toughness. In view of a successful and timely implementation of design parameters, in particular for primary circuits, new structural materials have to be qualified during the next decade. To this end an international R and D effort is being undertaken. Recent progress in materials science, supported by computer modelling and advanced materials characterisation techniques, has the potential to accelerate the process of new structural materials development. The scope of the meeting is information exchange and cross-fertilisation of various disciplines, including an overview of recent status of world-wide R and D activities. A comprehensive review of the designs of fission as well as fusion reactor systems

  18. Materials with complex behaviour II properties, non-classical materials and new technologies

    CERN Document Server

    Oechsner, Andreas

    2012-01-01

    This book reviews developments and trends in advanced materials and their properties; modeling and simulation of non-classical materials and new technologies for joining materials. Offers tools for characterizing and predicting properties and behavior.

  19. Mica fission detectors

    International Nuclear Information System (INIS)

    Wong, C.; Anderson, J.D.; Hansen, L.; Lehn, A.V.; Williamson, M.A.

    1977-01-01

    The present development status of the mica fission detectors is summarized. It is concluded that the techniques have been refined and developed to a state such that the mica fission counters are a reliable and reproducible detector for fission events

  20. Fission gas behaviour and interdiffusion layer growth in in-pile and out-of-pile irradiated U-Mo/Al nuclear fuels

    International Nuclear Information System (INIS)

    Zweifel, Tobias

    2014-01-01

    Worldwide, research and test reactors are to convert their fuel from highly towards lower enriched uranium, among them the FRM II. One prospective fuel is an alloy of uranium and molybdenum (abbr. U-Mo). Test irradiations showed an insufficient irradiation behavior of this new fuel due to the growth of an interdiffusion layer (abbr. IDL) between the U-Mo fuel and the surrounding Al matrix. Furthermore, this layer accumulates fission gases. In this work, heavy ion irradiated U-Mo/Al layer systems were studied and compared to in-reactor irradiated fuel to study the fission gas dynamics. It is demonstrated that the gas behavior is identical for both in-reactor and out-of-reactor approaches.

  1. Behaviour of cementitious materials: sulfates and temperature actions

    International Nuclear Information System (INIS)

    Barbarulo, Remi

    2002-09-01

    The research work presented in this Ph.D. thesis is related to the nuclear waste underground repository concept. Concrete could be used in such a repository, and would be subjected to variations of temperature in presence of sulfate, a situation that could induce expansion of concrete. The research was lead in three parts: an experimental study of the possibility of an internal sulfate attack on mortars; an experimental study and modeling of the chemical equilibriums of the CaO-SiO 2 -Al 2 O 3 -SO 3 -H 2 O system; and a modeling of the mechanisms of internal and external sulfate attacks, and the effect of temperature. The results show that mortars can develop expansions after a steam-cure during hydration, but also when a long steam-cure is applied to one-year-old mortars, which is a new point. Ettringite precipitation can be considered as responsible for these expansions. The experimental study of the CaO-SiO 2 -Al 2 O 3 -SO 3 -H 2 O system clarified the role of Calcium Silicate Hydrates (C-S-H) on chemical equilibriums of cementitious materials. Sulfate sorption on C-S-H has been studied in detail. The quantity of sulfate bound to the C-S-H mainly depends on the sulfate concentration in solution, on the Ca/Si ratio of the C-S-H and is not significantly influenced by temperature. Aluminium inclusion in the C-S-H seems to be a significant phenomenon. Temperature increases the calcium sulfo-aluminate solubilities and thus increases sulfates concentration in solution. A modeling of the chemical system is proposed. Simulations of external sulfate attack (15 mmol/L of Na 2 SO 4 ) predict ettringite precipitation at 20 and 85±C. Simulation of internal sulfate attack was performed at a local scale (a hydrated cement grain). An initial inhomogeneity can lead, after a thermal curing at 85±C, to ettringite precipitation in zones originally free from ettringite. This new-formed ettringite could be the origin of the expansions. (author) [fr

  2. Thermal stresses in hexagonal materials - heat treatment influence on their mechanical behaviour

    International Nuclear Information System (INIS)

    Gloaguen, D.; Freour, S.; Guillen, R.; Royer, J.; Francois, M.

    2004-01-01

    Internal stresses due to anisotropic thermal and plastic properties were investigated in rolled zirconium and titanium. The thermal stresses induced by a cooling process were predicted using a self-consistent model and compared with experimental results obtained by X-ray diffraction. The study of the elastoplastic response during uniaxial loading was performed along the rolling and the transverse direction of the sheet, considering the influence of the texture and the thermal stresses on the mechanical behaviour. An approach in order to determine the thermal behaviour of phases embedded in two-phase materials is also presented. For zirconium, the residual stresses due to thermal anisotropy are rather important (equivalent to 35% of the yield stress) and consequently they play an important role on the elastoplastic transition contrary to titanium. The study of two-phase material shows the influence and the interaction of the second phase on the thermal behaviour in the studied phase

  3. Effect of Anisotropy on the Resilient Behaviour of a Granular Material in Low Traffic Pavement.

    Science.gov (United States)

    Jing, Peng; Nowamooz, Hossein; Chazallon, Cyrille

    2017-12-03

    Granular materials are often used in pavement structures. The influence of anisotropy on the mechanical behaviour of granular materials is very important. The coupled effects of water content and fine content usually lead to more complex anisotropic behaviour. With a repeated load triaxial test (RLTT), it is possible to measure the anisotropic deformation behaviour of granular materials. This article initially presents an experimental study of the resilient repeated load response of a compacted clayey natural sand with three fine contents and different water contents. Based on anisotropic behaviour, the non-linear resilient model (Boyce model) is improved by the radial anisotropy coefficient γ ₃ instead of the axial anisotropy coefficient γ ₁. The results from both approaches ( γ ₁ and γ ₃) are compared with the measured volumetric and deviatoric responses. These results confirm the capacity of the improved model to capture the general trend of the experiments. Finally, finite element calculations are performed with CAST3M in order to validate the improvement of the modified Boyce model (from γ ₁ to γ ₃). The modelling results indicate that the modified Boyce model with γ ₃ is more widely available in different water contents and different fine contents for this granular material. Besides, based on the results, the coupled effects of water content and fine content on the deflection of the structures can also be observed.

  4. Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

    International Nuclear Information System (INIS)

    Poirier, J.; Dupouy, J.M.

    In this conference are presented papers dealing with swelling of metals and alloys, (and specially ferritic steels), structural evolution and stability under irradiation, modifications of mechanical properties, consequences on the behaviour of fuel elements and the optimization of materials selection, and irradiation creep [fr

  5. Fundamental principles of the cyclic behaviour and the fatigue damage for metallic materials

    International Nuclear Information System (INIS)

    Vogt, J.B.

    2001-01-01

    The aim of this paper is a pedagogic presentation of the basic concepts concerning the cyclic behaviour and the fatigue damage of metallic materials in order to offer a better understand of mechanisms. The following aspects are taking into account: the fatigue fracture, the cyclic accommodation, the dislocations structures, the surface and bulk cracks and the influence of the medium. (A.L.B.)

  6. Experimental studies of the crack behaviour during elastoplastic deformations of materials

    International Nuclear Information System (INIS)

    Hollstein, R.

    1982-01-01

    In C-, SEN- and WOL X-samples of the materials StE 460 (Ni-V), 22NiCr37, and 30CrNiMo8 a transition from linear elasticity to elastoplastic behaviour is observed with increasing temperature. Before crack propagation can be observed, a stretching zone at the crack tip is formed, which depends on the material and the stress conditions. (DG) [de

  7. Hot cell works and related irradiation tests in fission reactor for development of new materials for nuclear application

    International Nuclear Information System (INIS)

    Shikama, Tatsuo

    1999-01-01

    Present status of research works in Oarai Branch, Institute for Materials Research, Tohoku University, utilizing Japan Materials Testing Reactor and related hot cells will be described.Topics are mainly related with nuclear materials studies, excluding fissile materials, which is mainly aiming for development of materials for advanced nuclear systems such as a nuclear fusion reactor. Conflict between traditional and routined procedures and new demands will be described and future perspective is discussed. (author)

  8. Chemical aspects of fission product transport in the primary circuit of a light water reactor

    International Nuclear Information System (INIS)

    Bowsher, B.R.; Dickinson, S.; Nichols, A.L.; Ogden, J.S.; Potter, P.E.

    1985-01-01

    The transport and fission products in the primary circuit of a light water reactor are of fundamental importance in assessing the consequences of severe accidents. Recent experimental studies have concentrated upon the behaviour of simulant fission product species such as caesium iodide, caesium hydroxide and tellurium, in terms of their vapour deposition characteristics onto metals representative of primary circuit materials. An induction furnace has been used to generate high-density/structural materials aerosols for subsequent analysis, and similar equipment has been incorporated into a glove-box to study lightly-irradiated UO/sub 2/ clad in Zircaloy. Analytical techniques are being developed to assist in the identification of fission product chemical species released from the fuel at temperatures from 1000 to 2500 0 C. Matrix isolation-infrared spectroscopy has been used to identify species in the vapour phase, and specific data using this technique are reported

  9. Hook tool manufacture in New Caledonian crows: behavioural variation and the influence of raw materials.

    Science.gov (United States)

    Klump, Barbara C; Sugasawa, Shoko; St Clair, James J H; Rutz, Christian

    2015-11-18

    New Caledonian crows use a range of foraging tools, and are the only non-human species known to craft hooks. Based on a small number of observations, their manufacture of hooked stick tools has previously been described as a complex, multi-stage process. Tool behaviour is shaped by genetic predispositions, individual and social learning, and/or ecological influences, but disentangling the relative contributions of these factors remains a major research challenge. The properties of raw materials are an obvious, but largely overlooked, source of variation in tool-manufacture behaviour. We conducted experiments with wild-caught New Caledonian crows, to assess variation in their hooked stick tool making, and to investigate how raw-material properties affect the manufacture process. In Experiment 1, we showed that New Caledonian crows' manufacture of hooked stick tools can be much more variable than previously thought (85 tools by 18 subjects), and can involve two newly-discovered behaviours: 'pulling' for detaching stems and bending of the tool shaft. Crows' tool manufactures varied significantly: in the number of different action types employed; in the time spent processing the hook and bending the tool shaft; and in the structure of processing sequences. In Experiment 2, we examined the interaction of crows with raw materials of different properties, using a novel paradigm that enabled us to determine subjects' rank-ordered preferences (42 tools by 7 subjects). Plant properties influenced: the order in which crows selected stems; whether a hooked tool was manufactured; the time required to release a basic tool; and, possibly, the release technique, the number of behavioural actions, and aspects of processing behaviour. Results from Experiment 2 suggested that at least part of the natural behavioural variation observed in Experiment 1 is due to the effect of raw-material properties. Our discovery of novel manufacture behaviours indicates a plausible scenario for the

  10. Attachment of gaseous fission products to aerosols

    International Nuclear Information System (INIS)

    Skyrme, G.

    1985-01-01

    Accidents may occur in which the integrity of fuel cladding is breached and volatile fission products are released to the containment atmosphere. In order to assess the magnitude of the subsequent radiological hazard it is necessary to know the transport behaviour of such fission products. It is frequently assumed that the fission products remain in the gaseous phase. There is a possibility, however, that they may attach themselves to particles and hence substantially modify their transport properties. This paper provides a theoretical assessment of the conditions under which gaseous fission products may be attached to aerosol particles. Specific topics discussed are: the mass transfer of a gaseous fission product to an isolated aerosol particle in an infinite medium; the rate at which the concentration of fission products in the gas phase diminishes within a container as a result of deposition on a population of particles; and the distribution of deposited fission product between different particle sizes in a log-normal distribution. It is shown that, for a given mass, small particles are more efficient for fission product attachment, and that only small concentrations of such particles may be necessary to achieve rapid attachment. Conditions under which gaseous fission products are not attached to particles are also considered, viz, the competing processes of deposition onto the containment walls and onto aerosol particles, and the possibility of the removal of aerosols from the containment by various deposition processes, or agglomeration, before attachment takes place. (author)

  11. Chemical immobilization of fission products reactive with nuclear reactor components

    International Nuclear Information System (INIS)

    Grossman, L.N.; Kaznoff, A.I.; Clukey, H.V.

    1975-01-01

    This invention teaches a method of immobilizing deleterious fission products produced in nuclear fuel materials during nuclear fission chain reactions through the use of additives. The additives are disposed with the nuclear fuel materials in controlled quantities to form new compositions preventing attack of reactor components, especially nuclear fuel cld, by the deleterious fission products. (Patent Office Record)

  12. Experimental creep behaviour determination of cladding tube materials under multi-axial loadings

    International Nuclear Information System (INIS)

    Grosjean, Catherine; Poquillon, Dominique; Salabura, Jean-Claude; Cloue, Jean-Marc

    2009-01-01

    Cladding tubes are structural parts of nuclear plants, submitted to complex thermomechanical loadings. Thus, it is necessary to know and predict their behaviour to preserve their integrity and to enhance their lifetime. Therefore, a new experimental device has been developed to control the load path under multi-axial load conditions. The apparatus is designed to determine the thermomechanical behaviour of zirconium alloys used for cladding tubes. First results are presented. Creep tests with different biaxial loadings were performed. Results are analysed in terms of thermal expansion and of creep strain. The anisotropy of the material is revealed and iso-creep strain curves are given.

  13. Influence of the temperature on materials electric behaviour: Understanding and students’ learning difficulties

    Directory of Open Access Journals (Sweden)

    Antonio García Carmona

    2006-03-01

    Full Text Available In this article, we defend that in the teaching/learning of the electricity, its contents must be associa ted with contents concerning the structure and behaviour of the matter. Thus, it is possible to understand some electricity topics as the influence of the temperature on electric behaviour of materials. In this sense, we propose a conceptual framework for its teaching, coherent with the Spanish Physics and Chemistry curriculum of Secondary Education. Likewise, we show the results of a research carried out with 60 pupils (age 14-15, about theirs understanding levels and theirs learning difficulties regarding considered topic.

  14. Fission theory and actinide fission data

    Energy Technology Data Exchange (ETDEWEB)

    Michaudon, A.

    1975-06-01

    The understanding of the fission process has made great progress recently, as a result of the calculation of fission barriers, using the Strutinsky prescription. Double-humped shapes were obtained for nuclei in the actinide region. Such shapes could explain, in a coherent manner, many different phenomena: fission isomers, structure in near-threshold fission cross sections, intermediate structure in subthreshold fission cross sections and anisotropy in the emission of the fission fragments. A brief review of fission barrier calculations and relevant experimental data is presented. Calculations of fission cross sections, using double-humped barrier shapes and fission channel properties, as obtained from the data discussed previously, are given for some U and Pu isotopes. The fission channel theory of A. Bohr has greatly influenced the study of low-energy fission. However, recent investigation of the yields of prompt neutrons and γ rays emitted in the resonances of {sup 235}U and {sup 239}Pu, together with the spin determination for many resonances of these two nuclei cannot be explained purely in terms of the Bohr theory. Variation in the prompt neutron and γ-ray yields from resonance to resonance does not seem to be due to such fission channels, as was thought previously, but to the effect of the (n,γf) reaction. The number of prompt fission neutrons and the kinetic energy of the fission fragments are affected by the energy balance and damping or viscosity effects in the last stage of the fission process, from saddle point to scission. These effects are discussed for some nuclei, especially for {sup 240}Pu.

  15. Nanocrystalline SiC and Ti3SiC2 Alloys for Reactor Materials: Diffusion of Fission Product Surrogates

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jiang, Weilin [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-11-01

    MAX phases, such as titanium silicon carbide (Ti3SiC2), have a unique combination of both metallic and ceramic properties, which make them attractive for potential nuclear applications. Ti3SiC2 has been suggested in the literature as a possible fuel cladding material. Prior to the application, it is necessary to investigate diffusivities of fission products in the ternary compound at elevated temperatures. This study attempts to obtain relevant data and make an initial assessment for Ti3SiC2. Ion implantation was used to introduce fission product surrogates (Ag and Cs) and a noble metal (Au) in Ti3SiC2, SiC, and a dual-phase nanocomposite of Ti3SiC2/SiC synthesized at PNNL. Thermal annealing and in-situ Rutherford backscattering spectrometry (RBS) were employed to study the diffusivity of the various implanted species in the materials. In-situ RBS study of Ti3SiC2 implanted with Au ions at various temperatures was also performed. The experimental results indicate that the implanted Ag in SiC is immobile up to the highest temperature (1273 K) applied in this study; in contrast, significant out-diffusion of both Ag and Au in MAX phase Ti3SiC2 occurs during ion implantation at 873 K. Cs in Ti3SiC2 is found to diffuse during post-irradiation annealing at 973 K, and noticeable Cs release from the sample is observed. This study may suggest caution in using Ti3SiC2 as a fuel cladding material for advanced nuclear reactors operating at very high temperatures. Further studies of the related materials are recommended.

  16. Nuclear fission and reactions

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    The nuclear fission research programs are designed to elucidate basic features of the fission process. Specifically, (1) factors determining how nucleons of a fissioning nucleus are distributed between two fission fragments, (2) factors determining kinetic energy and excitation energies of fragments, and (3) factors controlling fission lifetimes. To these ends, fission studies are reported for several heavy elements and include investigations of spontaneous and neutron-induced fission, heavy ion reactions, and high energy proton reactions. The status of theoretical research is also discussed. (U.S.)

  17. A Comparative Study of the Behaviour of Five Dense Glass Materials Under Shock Loading Conditions

    Science.gov (United States)

    Radford, Darren D.; Proud, William G.; Field, John E.

    2001-06-01

    Previous work at the Cavendish Laboratory on the properties of glasses under shock loading has demonstrated that the material response is highly dependent upon the composition of the glass. The shock response of glass materials with an open structure, such as borosilicate, exhibits a ramping behaviour in the longitudinal stress histories due to structural collapse. Glass materials with a “filled” microstructure, as in the case of Type-D, Extra Dense Flint (DEDF) do not exhibit a ramping behaviour and behave in a manner similar to polycrystalline ceramics [1]. The current investigation compares the behaviour of five such glasses (SF15, DEDF, LACA, SF57 and DEDF-927210) under shock loading conditions. It is observed that slight changes in material composition can have a large affect on the inelastic behaviour. Principal Hugoniot and shear strength data are presented for all of the materials for pressures ranging from 2 to 14 GPa. Evidence of the so-called failure-front [2] is presented via lateral stress histories measured using manganin stress gauges and confirmed with high-speed photography. 1. Bourne, N.K., Millett, J.C.F., and Field, J.E., “On the strength of shocked glasses” Proc. R. Soc. Lond. A 455 (1999) 1275-1282 2. Brar, N.S., “Failure Waves in Glass and Ceramics Under Shock Compression”, in "Shock Compression of Condensed Matter 1999", ed. M.D. Furnish, L.C. Chhabildas, and R.S. Hixson, American Institute of Physics, Woodbury, New York, (1999) 601-606

  18. Actual point about fission products vitrification

    International Nuclear Information System (INIS)

    Bonniaud, R.

    1982-05-01

    The main characteristics concerning the continuous vitrification process for the confinement of fission product solutions operated at AVM are summarized. The general principle of a vitrification plant is described. The AVM plant efficiency as also its conception of consumable parts interchangeability are satisfying. The evolution of the process and its application developped in two ways: a more spaced installation conception and the improvement of the weak points remarked at AVM, as also the capacity of output. Two industrial units are designed at La Hague. The future evolution of the process aims at manufacturing glass at higher temperatures about 1400 degrees Celsius. Some problems remain to be resolved for the using of ceramic melters associated with a calcination unit. The studies provide for a satisfying behaviour for the material to long-term. The risks of damage by crystallisation, leaching and effects of alpha emission are analysed [fr

  19. Measurements of fission yields

    International Nuclear Information System (INIS)

    Denschlag, H.O.

    2000-01-01

    After some historical introductory remarks on the discovery of nuclear fission and early fission yield determinations, the present status of knowledge on fission yields is briefly reviewed. Practical and fundamental reasons motivating the pursuit of fission yield measurements in the coming century are pointed out. Recent results and novel techniques are described that promise to provide new interesting insights into the fission process during the next century. (author)

  20. Radiochemical studies on fission

    Energy Technology Data Exchange (ETDEWEB)

    None

    1973-07-01

    Research progress is reported on nuclear chemistry; topics considered include: recoil range and kinetic energy distribution in the thermal neutron ftssion of /sup 245/Cm; mass distribution and recoil range measurements in the reactor neutron-induced fission of /sup 232/U; fission yields in the thermal neutron fission of /sup 241/PU highly asymmetric binary fission of uranium induced by reactor neutrons; and nuclear charge distribution in low energy fission. ( DHM)

  1. Analytical methods for fissionable material determinations in the nuclear fuel cycle. Progress report, October 1, 1978-September 30, 1979

    International Nuclear Information System (INIS)

    Waterbury, G.R.

    1980-03-01

    Work continues on the development of dissolution techniques for difficult-to-dissolve nuclear materials, the development of methods and automated instruments for plutonium, uranium, and thorium determinations, and the preparation of plutonium materials for the Safeguards Analytical Laboratory Evaluation (SALE) program and distribution by the National Bureau of Standards (NBS) as standard reference materials (SRMs). We are measuring the loner plutonium isotope half-lives, evaluating the isotope correlation techniques and the chemistry involved in the mass-spectrometric ion-bead techniques, and analyzing the SALE uranium materials. Completed subtasks include evaluations of various Teflon materials to recommend those acceptable for the dissolution apparatus developed at LASL, investigations of laser-enhanced dissolution of refractory materials, determinations of diverse ion effects on the microgram-sensitive method for determining uranium, fabrication of the first automated controlled-potential coulometric analyzer for determining plutonium, preparation of a 244 Pu material for distribution by NBS as a SRM, and determination of the half-life of 239 Pu. Work has been started on a spectrophotometric method for determining microgram quantities of plutonium, a microcomplexometric titration method for determining uranium, the use of new reagents for separations of plutonium, the preparation and packaging of a new lot of high-purity plutonium metal for distribution by NBS as a plutonium chemical SRM, and determination of half-lives of other plutonium isotopes

  2. Analytical methods for fissionable material determinations in the nuclear fuel cycle. Progress report, October 1, 1978-September 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    Waterbury, G.R. (comp.)

    1980-03-01

    Work continues on the development of dissolution techniques for difficult-to-dissolve nuclear materials, the development of methods and automated instruments for plutonium, uranium, and thorium determinations, and the preparation of plutonium materials for the Safeguards Analytical Laboratory Evaluation (SALE) program and distribution by the National Bureau of Standards (NBS) as standard reference materials (SRMs). We are measuring the loner plutonium isotope half-lives, evaluating the isotope correlation techniques and the chemistry involved in the mass-spectrometric ion-bead techniques, and analyzing the SALE uranium materials. Completed subtasks include evaluations of various Teflon materials to recommend those acceptable for the dissolution apparatus developed at LASL, investigations of laser-enhanced dissolution of refractory materials, determinations of diverse ion effects on the microgram-sensitive method for determining uranium, fabrication of the first automated controlled-potential coulometric analyzer for determining plutonium, preparation of a /sup 244/Pu material for distribution by NBS as a SRM, and determination of the half-life of /sup 239/Pu. Work has been started on a spectrophotometric method for determining microgram quantities of plutonium, a microcomplexometric titration method for determining uranium, the use of new reagents for separations of plutonium, the preparation and packaging of a new lot of high-purity plutonium metal for distribution by NBS as a plutonium chemical SRM, and determination of half-lives of other plutonium isotopes.

  3. The effect of low-concentration inorganic materials on the behaviour of supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Imre, A.R., E-mail: imre@aeki.kfki.h [KFKI Atomic Energy Research Institute, POB 49, Budapest (Hungary); Hazi, G.; Horvath, A.; Maraczy, Cs. [KFKI Atomic Energy Research Institute, POB 49, Budapest (Hungary); Mazur, V.; Artemenko, S. [Odessa State Academy of Refrigeration, 1/3 Dvoryanslaya Str., 65026, Odessa (Ukraine)

    2011-01-15

    Research highlights: Small amount of inorganic materials (like corrosion products) can be dissolved in the supercritical water. Pseudo-critical temperature and other properties will be changed. Thermal and hydraulic behaviours of the SCW with small amount of contaminants differ in great extent from the behaviour of pure SCW. - Abstract: Supercritical water is a promising working fluid in the new Generation IV nuclear power plants. Due to the presence of the pseudo-critical line, the thermo-hydraulics (thermal and flow properties) and the physical chemistry of the supercritical water differ significantly from the pressurized hot water used in pressurized water reactors. In this study we would like to analyse the effect of small amount of inorganic material on the thermo-hydraulics of the supercritical water cooled nuclear reactors and other, non-nuclear supercritical water loops.

  4. Behaviour of LWR core materials under accident conditions. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1996-12-01

    At the invitation of the Government of the Russian Federation, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA convened a Technical Committee Meeting on Behaviour of LWR Core Materials Under Accident Conditions from 9 to 13 October 1995 in Dimitrovgrad to analyze and evaluate the behaviour of LWR core materials under accident conditions with special emphasis on severe accidents. In-vessel severe accidents phenomena were considered in detail, but specialized thermal hydraulic aspects as well as ex-vessel phenomena were outside the scope of the meeting. Forty participants representing eight countries attended the meeting. Twenty-three papers were presented and discussed during five sessions. Refs, figs, tabs

  5. A review of mechanical and tribological behaviour of polymer composite materials

    Science.gov (United States)

    Prabhakar, K.; Debnath, S.; Ganesan, R.; Palanikumar, K.

    2018-04-01

    Composite materials are finding increased applications in many industrial applications. A nano-composite is a matrix to which nanosized particles have been incorporated to drastically improve the mechanical performance of the original material. The structural components produced using nano-composites will exhibit a high strength-to-weight ratio. The properties of nano-composites have caused researchers and industries to consider using this material in several fields. Polymer nanocomposites consists of a polymer material having nano-particles or nano-fillers dispersed in the polymer matrix which may be of different shapes with at least one of the dimensions less than 100nm. In this paper, comprehensive review of polymer nanocomposites was done majorly in three different areas. First, mechanical behaviour of polymer nanocomposites which focuses on the mechanical property evaluation such as tensile strength, impact strength and modulus of elasticity based on the different combination of filler materials and nanoparticle inclusion. Second, wear behavior of Polymer composite materials with respect to different impingement angles and variation of filler composition using different processing techniques. Third, tribological (Friction and Wear) behaviour of nanocomposites using various combination of nanoparticle inclusion and time. Finally, it summarized the challenges and prospects of polymer nanocomposites.

  6. Stb 342 - Decree of 4 June 1987 amending the Decree on the transport of fissionable materials, ores and radioactive substances

    International Nuclear Information System (INIS)

    1987-01-01

    The 1969 transport Decree governs all modes of transport of fissile and radioactive materials as well as ores in and to and from the Netherlands. The 1987 Decree amends it, in particular, for modernization purposes. (NEA) [fr

  7. Material, behavioural, cultural and psychosocial factors in the explanation of socioeconomic inequalities in oral health.

    Science.gov (United States)

    Duijster, Denise; Oude Groeniger, Joost; van der Heijden, Geert J M G; van Lenthe, Frank J

    2017-12-19

    This study aimed to assess the contribution of material, behavioural, cultural and psychosocial factors in the explanation of socioeconomic inequalities (education and income) in oral health of Dutch adults. Cross-sectional data from participants (25-75 years of age) of the fifth wave of the GLOBE cohort were used (n = 2812). Questionnaires were used to obtain data on material factors (e.g. financial difficulties), behavioural factors (e.g. smoking), cultural factors (e.g. cultural activities) and psychosocial factors (e.g. psychological distress). Oral health outcomes were self-reported number of teeth and self-rated oral health (SROH). Mediation analysis, using multivariable negative binomial regression and logistic regression, was performed. Education level and income showed a graded positive relationship with both oral health outcomes. Adding material, behavioural, cultural and psychosocial factors substantially reduced the rate ratio for the number of teeth of the lowest education group from 0.79 (95% confidence interval (CI): 0.75-0.83) to 0.92 (95% CI: 0.87-0.97) and of the lowest income group from 0.80 (95% CI: 0.73-0.88) to 1.04 (95% CI: 0.96-1.14). Inclusion of all factors also substantially reduced the odds ratio for poor SROH of the lowest education group from 1.61 (95% CI: 1.28-2.03) to 1.12 (95% CI: 0.85-1.48) and of the lowest income groups from 3.18 (95% CI: 2.13-4.74) to 1.48 (95% CI: 0.90-2.45). In general, behavioural factors contributed most to the explanation of socioeconomic inequalities in adult oral health, followed by material factors. The contribution of cultural and psychosocial factors was relatively moderate. © The Author 2017. Published by Oxford University Press on behalf of the European Public Health Association.

  8. Validation of a New Elastoplastic Constitutive Model Dedicated to the Cyclic Behaviour of Brittle Rock Materials

    Science.gov (United States)

    Cerfontaine, B.; Charlier, R.; Collin, F.; Taiebat, M.

    2017-10-01

    Old mines or caverns may be used as reservoirs for fuel/gas storage or in the context of large-scale energy storage. In the first case, oil or gas is stored on annual basis. In the second case pressure due to water or compressed air varies on a daily basis or even faster. In both cases a cyclic loading on the cavern's/mine's walls must be considered for the design. The complexity of rockwork geometries or coupling with water flow requires finite element modelling and then a suitable constitutive law for the rock behaviour modelling. This paper presents and validates the formulation of a new constitutive law able to represent the inherently cyclic behaviour of rocks at low confinement. The main features of the behaviour evidenced by experiments in the literature depict a progressive degradation and strain of the material with the number of cycles. A constitutive law based on a boundary surface concept is developed. It represents the brittle failure of the material as well as its progressive degradation. Kinematic hardening of the yield surface allows the modelling of cycles. Isotropic softening on the cohesion variable leads to the progressive degradation of the rock strength. A limit surface is introduced and has a lower opening than the bounding surface. This surface describes the peak strength of the material and allows the modelling of a brittle behaviour. In addition a fatigue limit is introduced such that no cohesion degradation occurs if the stress state lies inside this surface. The model is validated against three different rock materials and types of experiments. Parameters of the constitutive laws are calibrated against uniaxial tests on Lorano marble, triaxial test on a sandstone and damage-controlled test on Lac du Bonnet granite. The model is shown to reproduce correctly experimental results, especially the evolution of strain with number of cycles.

  9. Analytical methods for fissionable materials in the nuclear fuel cycle. Progress report, July 1, 1975--September 30, 1976

    International Nuclear Information System (INIS)

    Waterbury, G.R.

    1976-12-01

    Progress continued on development of dissolution techniques for difficult-to-dissolve nuclear materials, development of methods and automated instruments for determinations of plutonium and uranium, preparation of plutonium-containing materials for the Safeguards Analytical Laboratory Evaluation (SALE) program, analysis of SALE uranium materials, and measurement of plutonium isotope half-lives. Gas-solid reactions at elevated temperatures using reactive gases such as chlorine continue to show promise for separating uranium from refractory materials. An extensive study of nonaqueous solvents for the dissolution of refractory materials is in progress. An extraction-separation procedure, highly specific for microgram amounts of uranium, has been developed, and its adaptation to the Los Alamos Scientific Laboratory (LASL) automated spectrophotometer is being evaluated. Development of an electrometric analysis method for plutonium is nearing completion, and design of an automated instrument using the method has been started. Batches of plutonium oxide and mixed uranium--plutonium, intended for issue as Secondary Reference and Calibration Test Materials, are being recharacterized for assay and isotopic contents. The half-life of 239 Pu has been determined by isotope-dilution mass-spectrometric measurement of 235 U grow-in as a function of time

  10. Analytical methods for fissionable materials in the nuclear fuel cycle. Progress report, July 1, 1975--September 30, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Waterbury, G.R. (comp.)

    1976-12-01

    Progress continued on development of dissolution techniques for difficult-to-dissolve nuclear materials, development of methods and automated instruments for determinations of plutonium and uranium, preparation of plutonium-containing materials for the Safeguards Analytical Laboratory Evaluation (SALE) program, analysis of SALE uranium materials, and measurement of plutonium isotope half-lives. Gas-solid reactions at elevated temperatures using reactive gases such as chlorine continue to show promise for separating uranium from refractory materials. An extensive study of nonaqueous solvents for the dissolution of refractory materials is in progress. An extraction-separation procedure, highly specific for microgram amounts of uranium, has been developed, and its adaptation to the Los Alamos Scientific Laboratory (LASL) automated spectrophotometer is being evaluated. Development of an electrometric analysis method for plutonium is nearing completion, and design of an automated instrument using the method has been started. Batches of plutonium oxide and mixed uranium--plutonium, intended for issue as Secondary Reference and Calibration Test Materials, are being recharacterized for assay and isotopic contents. The half-life of /sup 239/Pu has been determined by isotope-dilution mass-spectrometric measurement of /sup 235/U grow-in as a function of time.

  11. Theoretical study of laser feedback interferometry for dynamical material's behaviour studies

    International Nuclear Information System (INIS)

    Le-Barbier, Laura

    2017-01-01

    The purpose of this thesis is to study the feasibility of optical feedback interferometry (OFI) for measuring velocities for dynamical material's behaviour studies. Dynamical material's behaviour studies permit to analyse the shocked material when subjects to shocks (laser shocks, isentropic compression, projectiles, etc.). In these conditions, we seek to measure velocities up to 10 km/s. The OFI technique is regularly used as an embedded system to measure slow velocities in various fields. However, very few studies have been performed for determining velocities measurement limits for this system. As a matter of fact, the optical feedback induces nonlinear effects into the laser's cavity: it disrupts the laser's emitted optical power. Depending on the optical feedback strength, the laser can show chaotic behaviour, then it is no longer possible to get the information for the target's velocity or displacement regarding the signal. In this study, we have been developing mathematical models and performing a wide range of numerical simulations to study the performances and the limits of the OFI technique. We have been also studying the influence of the targets reflectivity, the length and the modulation frequency of the external cavity. (author) [fr

  12. Analytical methods for fissionable material determinations in the nuclear fuel cycle. Progress report, October 1, 1976--September 30, 1977

    International Nuclear Information System (INIS)

    Waterbury, G.R.

    1978-01-01

    Development of dissolution techniques for difficult-to-dissolve nuclear materials, development of methods and automated instruments for plutonium and uranium determinations, preparation of plutonium-containing materials for the Safeguards Analytical Laboratory Evaluation (SALE) program, analysis of SALE uranium materials, preparation of certified reference material plutonium metal, measurement of longer plutonium isotope half-lives, and study of ion exchange behavior of elements in various media continued. Gas-solid reaction of carbonyl chloride with uranium-bearing materials at elevated temperature is superior to reaction with chlorine for uranium volatilization and separation. Neither reaction with a variety of nonaqueous solvents nor reaction with molten selenium oxide provides practical dissolution of refractory materials characteristic of nuclear fuel cycle materials. The LASL automated spectrophotometer has been used to determine 0.1-mg amounts without instrumental or procedural changes. A microgram-sensitive spectrophotometric method for uranium has been developed, and the automated spectrophotometer is being modified to its use. A controlled-potential coulometric method has been developed for selective determination of plutonium. An automated analyzer to use this method is being built. Uranium-plutonium mixed oxide powder, for SALE samples, has not remained stable during storage, but high-density pellets have. In a DOE interlaboratory program, the half-life of 239 Pu has been measured, experiments on 241 Pu half-life measurement are in progress, and 240 Pu half-life measurement is planned. Ion exchange distributions for over 50 elements have been measured to determine cation exchange in nitric acid and anion exchange in both hydrobromic and hydriodic acids

  13. Analytical methods for fissionable material determinations in the nuclear fuel cycle. Progress report, October 1, 1976--September 30, 1977

    Energy Technology Data Exchange (ETDEWEB)

    Waterbury, G.R. (comp.)

    1978-01-01

    Development of dissolution techniques for difficult-to-dissolve nuclear materials, development of methods and automated instruments for plutonium and uranium determinations, preparation of plutonium-containing materials for the Safeguards Analytical Laboratory Evaluation (SALE) program, analysis of SALE uranium materials, preparation of certified reference material plutonium metal, measurement of longer plutonium isotope half-lives, and study of ion exchange behavior of elements in various media continued. Gas-solid reaction of carbonyl chloride with uranium-bearing materials at elevated temperature is superior to reaction with chlorine for uranium volatilization and separation. Neither reaction with a variety of nonaqueous solvents nor reaction with molten selenium oxide provides practical dissolution of refractory materials characteristic of nuclear fuel cycle materials. The LASL automated spectrophotometer has been used to determine 0.1-mg amounts without instrumental or procedural changes. A microgram-sensitive spectrophotometric method for uranium has been developed, and the automated spectrophotometer is being modified to its use. A controlled-potential coulometric method has been developed for selective determination of plutonium. An automated analyzer to use this method is being built. Uranium-plutonium mixed oxide powder, for SALE samples, has not remained stable during storage, but high-density pellets have. In a DOE interlaboratory program, the half-life of /sup 239/Pu has been measured, experiments on /sup 241/Pu half-life measurement are in progress, and /sup 240/Pu half-life measurement is planned. Ion exchange distributions for over 50 elements have been measured to determine cation exchange in nitric acid and anion exchange in both hydrobromic and hydriodic acids.

  14. Study by electronic structure calculations of the radiation damage in the UO2 nuclear fuel: behaviour of the point defects and fission gases

    International Nuclear Information System (INIS)

    Vathonne, Emerson

    2014-01-01

    Uranium dioxide (UO 2 ) is worldwide the most widely used fuel in nuclear plants in the world and in particular in pressurized water reactors (PWR). In-pile the fission of uranium nuclei creates fission products and point defects in the fuel. The understanding of the evolution of these radiation damages requires a multi-scale modelling approach of the nuclear fuel, from the scale of the pellet to the atomic scale. We used an electronic structure calculation method based on the density functional theory (DFT) to model radiation damage in UO 2 at the atomic scale. A Hubbard-type Coulomb interaction term is added to the standard DFT formalism to take into account the strong correlations of the 5f electrons in UO 2 . This method is used to study point defects with various charge states and the incorporation and diffusion of krypton in uranium dioxide. This study allowed us to obtain essential data for higher scale models but also to interpret experimental results. In parallel of this study, three ways to improve the state of the art of electronic structure calculations of UO 2 have been explored: the consideration of the spin-orbit coupling neglected in current point defect calculations, the application of functionals allowing one to take into account the non-local interactions such as van der Waals interactions important for rare gases and the use of the Dynamical Mean Field Theory combined to the DFT method in order to take into account the dynamical effects in the 5f electron correlations. (author) [fr

  15. Analytical methods for fissionable material determinations in the nuclear fuel cycle. Progress report, October 1, 1977--September 30, 1978

    International Nuclear Information System (INIS)

    Waterbury, G.R.

    1979-01-01

    Work has continued on the development of dissolution techniques for difficult-to-dissolve nuclear materials, development of methods and automated instruments for plutonium and uranium determinations, preparation of plutonium-containing materials for the Safeguards Analytical Laboratory Evaluation (SALE) program, preparation of plutonium materials for distribution by the National Bureau of Standards (NBS) as standard reference materials (SRMs), measurement of longer plutonium isotope half-lives, and analysis of SALE uranium materials. New tasks include the development of methods and automated instruments for the determination of thorium and uranium, and an evaluation of the ion-exchange-bead technique for the mass spectrometric measurement of uranium and plutonium isotope distributions. Completed tasks include the measurements of ion exchange distributions of over 50 elements on cation exchange resins from nitric acid media and anion exchange resins from hydrobromic and hydriodic acid media. Using a newly developed procedure, the LASL automated spectrophotometer was modified to determine microgram levels of uranium and to determine milligram levels of uranium and plutonium. Construction of an automated controlled-potential analyzer for the determination of plutonium is nearing completion. Apparatus and procedures for the separation and complexometric titration of thorium and uranium are being developed

  16. BEHAVIOUR OF BACKFILL MATERIALS FOR ELECTRICAL GROUNDING SYSTEMS UNDER HIGH VOLTAGE CONDITIONS

    Directory of Open Access Journals (Sweden)

    S. C. LIM

    2015-06-01

    Full Text Available Backfill materials like Bentonite and cement are effective in lowering grounding resistance of electrodes for a considerable period. During lightning, switching impulses and earth fault occurrences in medium and high voltage networks, the grounding system needs to handle extremely high currents either for a short duration or prolonged period respectively. This paper investigates the behaviour of bentonite, cement and sand under impulse and alternating high voltage (50Hz conditions. Fulguritic-formation was observed in all materials under alternating high voltage. The findings reveal that performance of grounding systems under high voltage conditions may significantly change from the outcomes anticipated at design stage.

  17. Analysis of the elastic behaviour of nonclassical nonlinear mesoscopic materials in quasi-static experiments

    International Nuclear Information System (INIS)

    Ruffino, E.; Scalerandi, M.

    2000-01-01

    As discovered by recent quasi-static and dynamic resonance experiments, the classical nonlinear theory fails in describing the hysteretic behaviour of nonlinear mesoscopic materials like rocks, concrete, etc. The paper applies the local interaction simulation approach (LISA) for studying such kind of nonclassical nonlinearity. To this purpose, in the LISA treatment of ultrasonic wave propagation has been included a phenomenological model, based on the PM space approach, of the local mesoscopic features of rocks and other materials with localized damages. A quantitative comparison of simulation and experimental results in quasi-static experiments is also presented

  18. Fission products collecting devices

    International Nuclear Information System (INIS)

    Matsumoto, Hiroshi

    1979-01-01

    Purpose: To enable fission products trap with no contamination to coolants and cover gas by the provision of a fission products trap above the upper part of a nuclear power plant. Constitution: Upon fuel failures in a reactor core, nuclear fission products leak into coolants and move along the flow of the coolants to the coolants above the reactor core. The fission products are collected in a trap container and guided along a pipeline into fission products detector. The fission products detector monitors the concentration of the fission products and opens the downstream valve of the detector when a predetermined concentration of the fission products is detected to introduce the fission products into a waste gas processing device and release them through the exhaust pipe. (Seki, T.)

  19. Energy released in fission

    International Nuclear Information System (INIS)

    James, M.F.

    1969-05-01

    The effective energy released in and following the fission of U-235, Pu-239 and Pu-241 by thermal neutrons, and of U-238 by fission spectrum neutrons, is discussed. The recommended values are: U-235 ... 192.9 ± 0.5 MeV/fission; U-238 ... 193.9 ± 0.8 MeV/fission; Pu-239 ... 198.5 ± 0.8 MeV/fission; Pu-241 ... 200.3 ± 0.8 MeV/fission. These values include all contributions except from antineutrinos and very long-lived fission products. The detailed contributions are discussed, and inconsistencies in the experimental data are pointed out. In Appendix A, the contribution to the total useful energy release in a reactor from reactions other than fission are discussed briefly, and in Appendix B there is a discussion of the variations in effective energy from fission with incident neutron energy. (author)

  20. Apparatus for storing and processing fissionable substances

    International Nuclear Information System (INIS)

    Dubovsky, B.G.; Bogatyrev, V.K.; Vladykov, G.M.; Sviridenko, V.Y.

    1974-01-01

    An apparatus is described for storing and processing fissionable substances in which there is provided a protective shield in the form of a layer of neutron absorbing material located in direct proximity to a vessel with a fissionable substance contained therein. The layer of neutron retarding material according to the present invention has alternating projections and depressions facing the layer of neutron-absorbing material. (author)

  1. Muon induced fission and fission track dating of minerals

    International Nuclear Information System (INIS)

    Marques, A.

    1988-01-01

    The effects of muon induced fission on geological dating of samples by the fission track method are evaluated for the case of muscovite minerals. It is found a small but significant effect, greater for the longer ages. Since calculations are developped under the hypothesis of constant atmosphere and primary cosmic ray flux it is suggested that any discrepancy found in ages of very old material that cannot be accounted for by well known environmental influences, be taken as an indication of variation on either the atmospheric stopping power or the intensity of cosmic radiation along the ages. (author) [pt

  2. Corrosion behaviour of container materials for geological disposal of high level radioactive waste

    International Nuclear Information System (INIS)

    Accary, A.

    1985-01-01

    The disposal of high level radioactive waste in geological formations, based on the multibarrier concept, may include the use of a container as one of the engineered barriers. In this report the requirements imposed on this container and the possible degradation processes are reviewed. Further on an overview is given of the research being carried out by various research centres in the European Community on the assessment of the corrosion behaviour of candidate container materials. The results obtained on a number of materials under various testing conditions are summarized and evaluated. As a result, three promising materials have been selected for a detailed joint testing programme. It concerns two highly corrosion resistant alloys, resp. Ti-Pd (0.2 Pd%) and Hastelloy C4 and one consumable material namely a low carbon steel. Finally the possibilities of modelling the corrosion phenomena are discussed

  3. Stb No. 404 - Decree of 12 July 1983 amending the fissionable materials, ores and radioactive substances (Transport) Decree

    International Nuclear Information System (INIS)

    1983-01-01

    For the Netherlands, international carriage by air of radioactive materials is governed by the regulations of the internationl Air Transport Association (IATA) which are partly based on the IAEA's recommendations in this respect. These were revised in 1973, and the present Decree amends the Transport Decree of 1969 to align it with the 1973 revision followed by IATA. (NEA) [fr

  4. Advances on fission chamber modelling

    International Nuclear Information System (INIS)

    Filliatre, Philippe; Jammes, Christian; Geslot, Benoit; Veenhof, Rob

    2013-06-01

    In-vessel, online neutron flux measurements are routinely performed in mock-up and material testing reactors by fission chambers. Those measurements have a wide range of applications, including characterization of experimental conditions, reactor monitoring and safety. Depending on the application, detectors may experience a wide range of constraints, of several magnitudes, in term of neutron flux, gamma-ray flux, temperature. Hence, designing a specific fission chamber and measuring chain for a given application is a demanding task. It can be achieved by a combination of experimental feedback and simulating tools, the latter being based on a comprehensive understanding of the underlying physics. A computation route that simulates fission chambers, named CHESTER, is presented. The retrieved quantities of interest are the neutron-induced charge spectrum, the electronic and ionic pulses, the mean current and variance, the power spectrum. It relies on the GARFIELD suite, originally developed for drift chambers, and makes use of the MAGBOLTZ code to assess the drift parameters of electrons within the filling gas, and the SRIM code to evaluate the stopping range of fission products. The effect of the gamma flux is also estimated. Computations made with several fission chambers exemplify the possibilities of the route. A good qualitative agreement is obtained when comparing the results with the experimental data available to date. In a near future, a comprehensive experimental programme will be undertaken to qualify the route using the known neutron sources, mock-up reactors and wide choice of fission chambers, with a stress on the predictiveness of the Campbelling mode. Depending on the results, a refinement of the modelling and an effort on the accuracy of input data are also to be considered. CHESTER will then make it possible to predict the overall sensitivity of a chamber, and to optimize the design for a given application. Another benefit will be to increase the

  5. Nuclear fission with inertial confinement

    CERN Document Server

    Koshkarev, D G

    2002-01-01

    The possibility of initiating the explosive fission reaction in a small quantity of fissile material through the heavy ions beam from the powerful accelerator-driver, developed for realization of the thermonuclear synthesis in the deuterium-tritium cylindrical targets with the direct ignition, is considered. The consequences of applying this method in the nuclear engineering are discussed

  6. Prehistory effects on the VHCF behaviour of engineering metallic materials with different strengthening mechanisms

    International Nuclear Information System (INIS)

    Zimmermann, M; Stoecker, C; Mueller-Bollenhagen, C; Christ, H-J

    2010-01-01

    Engineering materials often undergo a plastic deformation during manufacturing, hence the effect of a predeformation on the subsequent fatigue behaviour has to be considered. The effect of a prestrain on the microstructure is strongly influenced by the strengthening mechanism. Different mechanisms are relevant in the materials applied in this study: a solid-solution hardened and a precipitation-hardened nickel-base alloy and a martensite-forming metastable austenitic steel. Prehistory effects become very important, when fatigue failure at very high number of cycles (N > 10 7 ) is considered, since damage mechanisms occur different to those observed in the range of conventional fatigue limit. With the global strain amplitude being well below the static elastic limit, only inhomogeneously distributed local plastic deformation takes place in the very high cycle fatigue (VHCF) region. The dislocation motion during cyclic loading thus depends on the effective flow stress, which is defined by the global cyclic stress-strain relation and the local stress distribution as a consequence of the interaction between dislocations and precipitates, grain boundaries, martensite phases and micro-notches. As a consequence, no significant prehistory effect was observed for the VHCF behaviour of the solid-solution hardening alloy, while the precipitation-hardening alloy shows a perceptible prehistory dependence. In the case of the austenitic steel, strain-hardening and the volume fraction of the deformation-induced martensite dominate the fatigue behaviour.

  7. Evaluation the homogenisation behaviour of Sm-Fe-Nb materials by Moessbauer spectroscopy

    International Nuclear Information System (INIS)

    Sinan, S. A.; Muryaed, Y.; Alhweg, F. A.

    2004-01-01

    The microstructure of cast and annealed Sm-Fe-Nb materials were investigated by Moessbauer spectroscopy. The aim of the present work is to study the effect of Nb additions upon the microstructure of Sm 2 Fe 17 material and evaluation the homogenisation behaviour of different Sm-Fe-Nb materials. The niobium free cast material consisting of the Sm 2 Fe 17 phase and significant amounts of the free iron (α -Fe). Therefore, the homogenisation process is necessary to eliminate the free iron and produce a single Sm 2 Fe 17 phase material. This process takes long annealing time, up to seven days. The Sm 9 .5 Fe 8 7.5 Nb 3 alloy contains the lowest amount of α-Fe among, the Sm-Fe-Nb materials. Thus the homogenisation step was carried out with treatment time (12 hours) smaller than the reported annealing time of Nb-free material (Sm 2 Fe 17 ). Therefore, the addition of at 3% Nb reduces the manufacturing cost of the Sm 2 Fe 17 and makes this based material for permanent magnets, more industrially desirable, due to elimination the free iron with lowest treatment time. Also it was found that the existence of the paramagnetic NbFe 2 phase becomes higher after the homogenisation process, which can be explained due to the diffusion of Nb from Sm 2 Fe 17 phase to paramagnetic NbFe 2 phase, during the annealing process. (authors)

  8. The effects of the finest grains on the mechanical behaviours of nanocrystalline materials

    International Nuclear Information System (INIS)

    Hu Lingling; Huo Ruxiao; Zhou Jianqiu; Wang Ying; Zhang Shu

    2012-01-01

    This article proposes a new constitutive model to account for effects of the finest grains, with sizes ranging from 2 to 4 nm, on the mechanical behaviours of nanocrystalline (NC) materials. In this model, the normal nanograins (ranging from 20 to 100 nm) were treated as though they were composed of a grain interior (GI) and a grain boundary (GB) affected zone (GBAZ). The finest grains were considered to be part of the GBAZ, denoted as super triple junctions (STJs). For the initial plastic deformation stage of the NC materials, a phenomenological constitutive equation was suggested to predict the deformation behaviours of the GBAZ. The formation of GB dislocation (GBD) pileups provides dramatic strain hardening in deformed NC materials and thereby enhances their ductility. Then, the constitutive equations to describe the plastic deformation of the GI and the GBAZ lattice region were established. In this stage, the GBAZ are already saturated with GBD pileups, and GI deformation is the dominant mechanism. Finally, the mechanical model for the NC materials with the finest grains was built using the self-consistent method, and an overall moderate “work hardening,” sustained over a long range of plastic strain, was predicted. The effects of TJs/STJs on the deformation mechanism were quantitatively analysed. The analysis demonstrated that the existence of the finest grains will simultaneously lead to good strength and good ductility.

  9. Drying and moisture resorption behaviour of various electrode materials and separators for lithium-ion batteries

    Science.gov (United States)

    Stich, Michael; Pandey, Nisrit; Bund, Andreas

    2017-10-01

    The drying behaviour and water uptake of a variety of commonly used electrode materials (graphite, LiFePO4, LiMn2O4, LiCoO2, Li(NiCoMn)O2) and separators (polyolefin, glass fibre) for lithium-ion batteries (LIBs) are investigated. The drying experiments are carried out using a coulometric Karl Fischer titrator in combination with a vaporiser. This setup leads to a highly sensitive and precise method to quantify water amounts in the microgram range in solid materials. Thereby the mass specific drying behaviour at RT and 120 °C is determined as well as the water resorption of the investigated materials in conditioned air atmosphere (T: 25 °C, RH: 40%). By extracting characteristic water detection rate curves for the investigated materials, a method is developed to predict the water detection beyond the runtime of the experiment. The results help optimising drying procedures of LIB components and thus can save time and costs. It is also shown, that water contaminations in graphite/LiFePO4 coin cells with a LiPF6 based electrolyte lead to a faster capacity fade during cycling and a significant change of the cell impedance.

  10. Effects of loose housing and the provision of alternative nesting material on peri-partum sow behaviour and piglet survival

    NARCIS (Netherlands)

    Bolhuis, J.E.; Raats-van den Boogaard, A.M.E.; Hoofs, A.I.J.; Soede, N.M.

    2018-01-01

    Sows are strongly motivated to perform nestbuilding behaviour before parturition. This behaviour is often restricted in commercial systems due to confinement of the sow and lack of suitable nesting material to be used on slatted floors. This study aimed to investigate effects of loose vs. crated

  11. Decoring Behaviour of Chosen Moulding Materials with Alkali Silicate Based Inorganic Binders

    Directory of Open Access Journals (Sweden)

    Conev M.

    2017-06-01

    Full Text Available This paper contains basic information about new processes for cores for cylinder heads production with alkali silicate based inorganic binders. Inorganic binders are coming back to the foreground due to their ecologically friendly nature and new technologies for cores production and new binder systems were developed. Basically these binder systems are modified alkali silicates and therefore they carry some well-known unfavourable properties with their usage. To compensate these disadvantages, the binder systems are working with additives which are most often in powder form and are added in the moulding material. This paper deals with decoring behaviour of different moulding sands as well as the influence of chosen additives on knock-out properties in laboratory terms. For this purpose, specific methods of specimen production are described. Developed methods are then used to compare decoring behaviour of chosen sands and binder systems.

  12. Leaching behaviour of municipal solid waste incineration bottom ash: From granular material to monolithic concrete.

    Science.gov (United States)

    Sorlini, Sabrina; Collivignarelli, Maria Cristina; Abbà, Alessandro

    2017-09-01

    The aim of this work was to assess the leaching behaviour of the bottom ash derived from municipal solid waste incineration (MSWI) used in concrete production. In particular, the release of pollutants was evaluated by the application of different leaching tests, both on granular materials and monolithic samples (concrete mixtures cast with bottom ash). The results confirmed that, according to Italian regulations, unwashed bottom ashes present critical issues for the use as alternative aggregates in the construction sector due to the excessive release of pollutants; instead, the leachate from washed bottom ashes was similar to natural aggregates. The concentration of pollutants in the leachate from concrete mixtures was lower than regulation limits for reuse. The crushing process significantly influenced the release of pollutants: this behaviour was due both to the increase in surface area and the release of contaminants from cement. Moreover, the increase in contact time (up to 64 days) involved more heavy metals to be released.

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

    International Nuclear Information System (INIS)

    Mogensen, M.; Walker, C.T.

    1999-01-01

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

  14. Mechanical Behaviour of Conventional Materials at Experimental Conditions of Deep Drawing Technological Process

    Science.gov (United States)

    Nikolov, N.; Pashkouleva, D.; Kavardzhikov, V.

    2012-09-01

    The paper deals with experimental investigations on the mechanical behaviour of body-centred-cubic (BCC) and face-centred-cubic (FCC)-conventionally structured sheet metalic-metalic materials under stress-strain conditions of a deep drawing process determined by a coefficient close to the limiting one for Steel 08 and punch diameter of 50 mm. The mechanical characteristics of the investigated materials are identified by one-dimensional tension tests. The materials' responses, as results of identical loading conditions, are described by the change of blank sizes and characteristics of the forming processes. The chosen deformation path ensures obtaining a qualitative steel piece and leads to failures of aluminium and brass blanks. The reported results could be useful for investigations and predictions of the mechanical responses of such type metallic structures applying microscopic instrumented observations and numerical simulations.

  15. Chemical Separation of Fission Products in Uranium Metal Ingots from Electrolytic Reduction Process

    International Nuclear Information System (INIS)

    Lee, Chang-Heon; Kim, Min-Jae; Choi, Kwang-Soon; Jee, Kwang-Yong; Kim, Won-Ho

    2006-01-01

    Chemical characterization of various process materials is required for the optimization of the electrolytic reduction process in which uranium dioxide, a matrix of spent PWR fuels, is electrolytically reduced to uranium metal in a medium of LiCl-Li 2 O molten at 650 .deg. C. In the uranium metal ingots of interest in this study, residual process materials and corrosion products as well as fission products are involved to some extent, which further adds difficulties to the determination of trace fission products. Besides it, direct inductively coupled plasma atomic emission spectrometric (ICP-AES) analysis of uranium bearing materials such as the uranium metal ingots is not possible because a severe spectral interference is found in the intensely complex atomic emission spectra of uranium. Thus an adequate separation procedure for the fission products should be employed prior to their determinations. In present study ion exchange and extraction chromatographic methods were adopted for selective separation of the fission products from residual process materials, corrosion products and uranium matrix. The sorption behaviour of anion and tri-nbutylphosphate (TBP) extraction chromatographic resins for the metals in acidic solutions simulated for the uranium metal ingot solutions was investigated. Then the validity of the separation procedure for its reliability and applicability was evaluated by measuring recoveries of the metals added

  16. Fission Research at IRMM

    Directory of Open Access Journals (Sweden)

    Al-Adili A.

    2010-03-01

    Full Text Available Fission Research at JRC-IRMM has a longstanding tradition. The present paper is discussing recent investigations of fission fragment properties of 238 U(n,f, 234 U(n,f, prompt neutron emission in fission of 252 Cf(SF as well as the prompt fission neutron spectrum of 235 U(n,f and is presenting the most important results.

  17. Overview of research by the fission group in Trombay

    Indian Academy of Sciences (India)

    In the late eighties, heavy-ion beams from the pelletron-based medium energy heavy- ... (9) Ternary fission/light charged particle (LCP) accompanied fission .... There is a clear deviation in the behaviour of neck emission of α-particles at high.

  18. Tensile behaviour of geopolymer-based materials under medium and high strain rates

    Science.gov (United States)

    Menna, Costantino; Asprone, Domenico; Forni, Daniele; Roviello, Giuseppina; Ricciotti, Laura; Ferone, Claudio; Bozza, Anna; Prota, Andrea; Cadoni, Ezio

    2015-09-01

    Geopolymers are a promising class of inorganic materials typically obtained from an alluminosilicate source and an alkaline solution, and characterized by an amorphous 3-D framework structure. These materials are particularly attractive for the construction industry due to mechanical and environmental advantages they exhibit compared to conventional systems. Indeed, geopolymer-based concretes represent a challenge for the large scale uses of such a binder material and many research studies currently focus on this topic. However, the behaviour of geopolymers under high dynamic loads is rarely investigated, even though it is of a fundamental concern for the integrity/vulnerability assessment under extreme dynamic events. The present study aims to investigate the effect of high dynamic loading conditions on the tensile behaviour of different geopolymer formulations. The dynamic tests were performed under different strain rates by using a Hydro-pneumatic machine and a modified Hopkinson bar at the DynaMat laboratory of the University of Applied Sciences of Southern Switzerland. The results are processed in terms of stress-strain relationships and strength dynamic increase factor at different strain-rate levels. The dynamic increase factor was also compared with CEB recommendations. The experimental outcomes can be used to assess the constitutive laws of geopolymers under dynamic load conditions and implemented into analytical models.

  19. Equilibrium fission model calculations

    International Nuclear Information System (INIS)

    Beckerman, M.; Blann, M.

    1976-01-01

    In order to aid in understanding the systematics of heavy ion fission and fission-like reactions in terms of the target-projectile system, bombarding energy and angular momentum, fission widths are calculated using an angular momentum dependent extension of the Bohr-Wheeler theory and particle emission widths using angular momentum coupling

  20. Study of the sorption of nuclear fuel fission products using non-ion exchange polymeric and inorganic materials

    International Nuclear Information System (INIS)

    Manorik, P.

    1997-01-01

    New tetrazamacrocycle and crown-ether poly-styrene and divinylbenzenepolystyrene derivatives (P-TAM and P-CE respectively) have been synthesized and studied as chemisorbents for Cs, Sr (P-CE) and Ru, Co (P-TAM) radioisotopes. It has been found that the tetraazamacro cycle modifier concentration in such material is about 5 x 10 -4 M/g material. Model solutions, containing Ru and other platinoid salts, and also Co salts, were used in experiments at concentrations of 5 x 10 -3 - 5 x 10 -6 M/L. It was shown that P-TAM quantitatively removes Ru and other platinoids from the water solution at pH = 0-1 during 1-2 days and practically all modifying groups participate in the sorption process. It was established that in alkaline solutions this sorbent also adsorbs Co, Cr and other 3d-metal ions. P-TAM also demonstrates a very high platinoid sorption selectivity (close to 100%) from solutions containing, for example, platinoids in the presence of a large excess of 3d-metal ions. 137 Cs sorption by some types of P-CE and IN-CE from Ringer-Locka model waste solution was also studied and it was found that K d (Bq/g) strongly depends not only on the crown-ether ring size but also on the form in which crown-ethers exist on the surface, for example when the modifier was in the form of the complex such as P-CE-L(L=K 3 Fe(CN) 6 or L=K 4 Fe(CN) 6 ) there was an increase in K' d value (Bq/mole of CE). The results of a study of 85 Sr removal by P-CE and IN-CE show that the chemisorption capacity strongly depends not only on the ''hole'' size of crown-ethers but also on the nature of encapsulated complexes, including ferrocyanides of different types. 32 refs, 5 figs, 5 tabs

  1. Neutron multicounter detector for investigation of content and spatial distribution of fission materials in large volume samples

    International Nuclear Information System (INIS)

    Swiderska-Kowalczyk, M.; Starosta, W.; Zoltowski, T.

    1998-01-01

    The experimental device is a neutron coincidence well counter. It can be applied for passive assay of fissile - especially for plutonium bearing - materials. It consist of a set of 3 He tubes placed inside a polyethylene moderator; outputs from the tubes, first processed by preamplifier/amplifier/discriminator circuits, are then analysed using neutron correlator connected with a PC, and correlation techniques implemented in software. Such a neutron counter allows for determination of plutonium mass ( 240 Pu effective mass) in nonmultiplying samples having fairly big volume (up to 0.14 m 3 ). For determination of neutron sources distribution inside the sample, the heuristic methods based on hierarchical cluster analysis are applied. As an input parameters, amplitudes and phases of two-dimensional Fourier transformation of the count profiles matrices for known point sources distributions and for the examined samples, are taken. Such matrices are collected by means of sample scanning by detection head. During clustering process, counts profiles for unknown samples fitted into dendrograms using the 'proximity' criterion of the examined sample profile to standard samples profiles. Distribution of neutron sources in an examined sample is then evaluated on the basis of comparison with standard sources distributions. (author)

  2. Decree of 4 September 1969, Stb. 403, concerning the implementation of Sections 16, 17, 19, paragraph 1 and 21 of the Nuclear Energy Act (Nuclear Installations, Fissionable Materials and Ores)

    International Nuclear Information System (INIS)

    1969-01-01

    This Decree lays down the licensing system for fissionable materials and ores except during transport or storage incidental to transport. It also provides for a procedure for objections by third parties against the granting of a licence. Such licences are granted jointly by the Minister for Economic Affairs and the Minister for Social Affairs and Public Health, where necessary in agreement with the other Ministers concerned. (NEA) [fr

  3. Strength behaviour of sintered steel from the view of design-relevant material data

    International Nuclear Information System (INIS)

    Sonsino, C.M.; Esper, F.J.; Leuze, G.

    1982-01-01

    A reliable design of sintered components and an aimed material's selection requires the knowledge of designrelevant material data as Cyclic stress-strain-curves, crack propagation and fracture toughness properties as well as statistically evaluated S-N-curves, because conventional material data as tensile strength, monotonic yield strength, elongation, area reduction and impact strength can lead to a false estimation of the material's fatigue behaviour. For this reason the powder metallurgical industry began to determine design-relevant material data on the example of the porous Fe-Cu-C- and Fe-Cu-Ni-alloys. The fatigue tests with notched specimen and different modes of loading show that porous sintered parts having mechanical notches are less sensitive to external notches than wrought steel, because crack-propagation is delayed by pores. The possibility to manufacture cyclic hardening alloys, their relative notch-insensitivity and with wrought steel comparable scatter of fatigue properties show the importance of sintered alloys as alternative materials. (orig.) [de

  4. Aerosols and fission product transport

    International Nuclear Information System (INIS)

    Megaw, W.J.

    1987-12-01

    A survey is presented of current knowledge of the possible role of aerosols in the consequences of in- and out-of-core LOCAs and of end fitting failures in CANDU reactors. An extensive literature search has been made of research on the behaviour of aerosols in possible accidents in water moderated and cooled reactors and the results of various studies compared. It is recommended that further work should be undertaken on the formation of aerosols during these possible accidents and to study their subsequent behaviour. It is also recommended that the fission products behaviour computer code FISSCON II should be re-examined to determine whether it reflects the advances incorporated in other codes developed for light water reactors which have been extensively compared. 47 refs

  5. Target conception for the Munich fission fragment accelerator

    CERN Document Server

    Maier, H J; Gross, M L; Grossmann, R; Kester, O; Thirolf, P

    1999-01-01

    For the new high-flux reactor FRM II, the fission fragment accelerator MAFF is under design. MAFF will supply intense mass-separated radioactive ion beams of very neutron-rich nuclei with energies around the Coulomb barrier. A central part of this accelerator is the ion source with the fission target, which is operated at a neutron flux of 1.5x10 sup 1 sup 4 cm sup - sup 2 s sup - sup 1. The target consists of typically 1 g of sup 2 sup 3 sup 5 U dispersed in a cylindrical graphite matrix, which is encapsulated in a Re container. To enable diffusion and extraction of the fission products, the target has to be maintained at a temperature of up to 2400 deg. C during operation. It has to stand this temperature for at least one reactor cycle of 1250 h. Comprehensive tests are required to study the long-term behaviour of the involved materials at these conditions prior to operation in the reactor. The present paper gives details of the target conception and the projected tests.

  6. The Microwave Noise Behaviour Of Dual Material Gate Silicon On Insulator

    Science.gov (United States)

    Jafar, N.; Soin, N.

    2009-06-01

    This work presents the noise behaviour due to the applied Dual Material Gate (DMG) on the 75 nm n-channel Silicon On Insulator (SOI) device operating in the fully depletion mode, particularly for microwave circuit design. Influences of DMG properties namely the gate length ratio (L1:L2) and gate material workfunction difference (ΔΦM) as well as structural and operational parameters which are silicon thickness (TSi) and threshold voltage (VTH) setting variation on the noise performance were carried out on simulation basis using ATLAS 2D. Results show better noise performance in DMG as compare to the standard gate structure of FD-SOI devices. Higher VTH for DMG design is recommended for minimized noise figure in line with the advantage of inverse VTH roll-off characteristics for short channel effects suppression.

  7. Tribological and mechanical behaviours of rattan-fibre-reinforced friction materials under dry sliding conditions

    Science.gov (United States)

    Ma, Yunhai; Wu, Siyang; Tong, Jin; Zhao, Xiaolou; Zhuang, Jian; Liu, Yucheng; Qi, Hongyan

    2018-03-01

    This work was mainly aimed to study the physical, mechanical and tribological behaviours of the friction materials reinforced by different contents of rattan fibre. These friction materials were fabricated by a compression moulder and tested using a constant speed tester at different friction temperatures. It was found that the friction coefficients of the friction materials added with rattan fibre were relatively stable and no obvious fade was observed in comparison with specimen F-0 (containing 0 wt.% rattan fibres). The fade ratio of specimen F-5 (containing 5 wt.% rattan fibres) was 10.3% and its recovery ratio was 92.4%, indicating the excellent performances of fade resistance and recovery. And the specimen F-5 exhibited the lowest wear rate (0.541 × 10‑7 cm3(N · m)‑1 at 350 °C) among all tested specimens. The worn surface morphologies of the friction materials showed that the appropriate addition of rattan fibres effectively reduced abrasive wear and adhesion wear. The specimen F-5 had a smooth worn surface (Sa = 1.885 μm) with the superior fibre-matrix interfacial adhesion and a lot of secondary contact plateaus, which indicated the highest wear resistance property. The rattan-fibre-reinforced friction materials could be widely applied to automotive friction brake field according to their economic, environmental and social benefits.

  8. Peltier heat measurements at a junction between materials exhibiting Fermi gas and Fermi liquid behaviour

    International Nuclear Information System (INIS)

    Kuznetsov, V L; Kuznetsova, L A; Rowe, D M

    2003-01-01

    The feasibility of improving the conversion efficiency of a thermoelectric converter by employing interfaces between materials exhibiting Fermi gas (FG) and Fermi liquid (FL) behaviour has been studied. Thermocouples consisting of a semiconductor and a strongly correlated material have been fabricated and the Peltier heat measured over the temperature range 15 deg 330 K. A number of materials possessing different types of strong electron correlation have been synthesized including the heavy fermion compound YbAl 3 , manganite La 0.7 Ca 0.3 MnO 3 and high-T c superconductor YBa 2 Cu 3 O 7δ . n- and p-Bi 2 Te 3 -based solid solutions as well as n-Bi 0.85 Sb 0.15 solid solution have also been synthesized and used as materials exhibiting FG properties. Experimental measurements of the Peltier heat were compared to the results of calculations based on preliminary measured thermoelectric properties of materials and electrical contact resistance at the interfaces. The potential of employing FG/FL interfaces in thermoelectric energy conversion is discussed

  9. Research activities at JAERI on core material behaviour under severe accident conditions

    International Nuclear Information System (INIS)

    Uetsuka, H.; Katanashi, S.; Ishijima, K.

    1996-01-01

    At the Japan Atomic Energy Research Institute (JAERI), experimental studies on physical phenomena under the condition of a severe accident have been conducted. This paper presents the progress of the experimental studies on fuel and core materials behaviour such as the thermal shock fracture of fuel cladding due to quenching, the chemical interaction of core materials at high temperatures and the examination of TMI-2 debris. The mechanical behaviour of fuel rod with heavily embrittled cladding tube due to the thermal shock during delayed reflooding have been investigated at the Nuclear Safety Research Reactor (NSSR) of JAERI. A test fuel rod was heated in steam atmosphere by both electric and nuclear heating using the NSSR, then the rod was quenched by reflooding at the test section. Melting of core component materials having relatively low melting points and their eutectic reaction with other materials significantly influence on the degradation and melt down of fuel bundles during severe accidents. Therefore basic information on the reaction of core materials is necessary to understand and analyze the progress of core melting and relocation. Chemical interactions have been widely investigated at high temperatures for various binary systems of core component materials including absorber materials such as Zircaloy/Inconel, Zircaloy/stainless steel, Zircaloy/(Ag-In-Cd), stainless steel B 4 C and Zircaloy/B 4 C. It was found that the reaction generally obeyed a parabolic rate law and the reaction rate was determined for each reaction system. Many debris samples obtained from the degraded core of TMI-2 were transported to JAERI for numerous examinations and analyses. The microstructural examination revealed that the most part of debris was ceramic and it was not homogeneous in a microscopic sense. The thermal diffusivity data was also obtained for the temperature range up to about 1800K. The data from the large scale integral experiments were also obtained through the

  10. Coupling between mechanical behaviour and drying of cementing materials: experimental study on mortars

    International Nuclear Information System (INIS)

    Yurtdas, I.

    2003-10-01

    The aim of this work is to understand the desiccation effects on the mechanical behaviour of cement materials. Two mortars of ratio E/C=0.5 and 0.8 have been tested. All the tests have been implemented after a six months maturing in water. The experimental study has been carried out as follows: 1)tests characterizing the differed behaviour and the transport properties have been carried out 2)tests characterizing the short term multiaxial mechanical behaviour have been carried out. The desiccation shrinkage in terms of the weight loss presents three characteristic phases. The permeability measurement on the mortar 05 shows that the permeability of the specimens dried and crept is greater than those of the specimens dried before being crept, and the permeability of the specimens submitted to a desiccation creep and then dried is sensibly the same as the last one in spite of a very important differed deformation. The influence of the desiccation on the uniaxial and deviatoric compressions resistance depends of the binding agent: for a cement paste of good quality (E/C=0.5), the resistances increase with the desiccation because of the capillary depression and of the hydric gradients. For a cement paste of low quality (E/C=0.8), there is a competitive effect between the increase of the microcracks induced and the specimen rigidification; the microcracking becomes then the parameter controlling the rupture process. The elasto-plastic behaviour becomes a damageable elasto-plastic behaviour during desiccation which induces, as the decrease of the E/C ratio, a translation of the elastic limit surfaces and ruptures towards higher stresses. In parallel, the elastic properties and the incompressibility modulus are damaged and the volume deformations increase after the drying. At last, the decrease of the Young modulus and the passage to the third shrinkage phase in terms of the weight loss coincide. This can be attributed to the induced microcracking: this decrease of the

  11. The Investigation of Knitted Materials Bonded Seams Behaviour upon Cyclical Fatigue Loading

    Directory of Open Access Journals (Sweden)

    Gita BUSILIENĖ

    2017-08-01

    Full Text Available In this research uniaxial tension behaviour of PES knitted materials with bonded seams is analysed. The objects of the investigation were two types of knitted materials, having the same fibre composition (93 % PES, 7 % EL, but different in knitting pattern, i. e. plain single jersey and rib 1 × 1. Bonded overlap seams were formed by changing the orientation of knitted materials strips, i. e. parallel/parallel, parallel/bias, parallel/perpendicular, bias/bias and bias/perpendicular. The strips of each knitted material were joined by two types of thermoplastic polyurethane (PU films different in thickness (75 mm and 150 mm. Mechanical characteristics of bonded seams were defined in longitudinal direction. During uniaxial tension such parameters as maximal force Fmax (N and maximal elongation ɛmax (% were recorded from typical tension diagrams. The changes of tested specimens strength and deformation were compared before and after cyclical fatigue tension the conditions of which were 50 cycles up to tension force F equal 24.5 N. The results have shown that changes before and after cyclical fatigue tension are mostly determined by the structure of knitted materials, the orientation of knitted materials strips in bonded seam, but not effected by thermoplastic polyurethane film. These results are opposite compared to the results of biaxial tension of the same type of specimens, which have shown that changes before and after cyclical fatigue punching are mostly determined by the type of thermoplastic film, but not effected by the orientation of knitted materials strips in bonded seams. DOI: http://dx.doi.org/10.5755/j01.ms.23.2.16065

  12. Magnetic hysteresis at the domain scale of a multi-scale material model for magneto-elastic behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Vanoost, D., E-mail: dries.vanoost@kuleuven-kulak.be [KU Leuven Technology Campus Ostend, ReMI Research Group, Oostende B-8400 (Belgium); KU Leuven Kulak, Wave Propagation and Signal Processing Research Group, Kortrijk B-8500 (Belgium); Steentjes, S. [Institute of Electrical Machines, RWTH Aachen University, Aachen D-52062 (Germany); Peuteman, J. [KU Leuven Technology Campus Ostend, ReMI Research Group, Oostende B-8400 (Belgium); KU Leuven, Department of Electrical Engineering, Electrical Energy and Computer Architecture, Heverlee B-3001 (Belgium); Gielen, G. [KU Leuven, Department of Electrical Engineering, Microelectronics and Sensors, Heverlee B-3001 (Belgium); De Gersem, H. [KU Leuven Kulak, Wave Propagation and Signal Processing Research Group, Kortrijk B-8500 (Belgium); TU Darmstadt, Institut für Theorie Elektromagnetischer Felder, Darmstadt D-64289 (Germany); Pissoort, D. [KU Leuven Technology Campus Ostend, ReMI Research Group, Oostende B-8400 (Belgium); KU Leuven, Department of Electrical Engineering, Microelectronics and Sensors, Heverlee B-3001 (Belgium); Hameyer, K. [Institute of Electrical Machines, RWTH Aachen University, Aachen D-52062 (Germany)

    2016-09-15

    This paper proposes a multi-scale energy-based material model for poly-crystalline materials. Describing the behaviour of poly-crystalline materials at three spatial scales of dominating physical mechanisms allows accounting for the heterogeneity and multi-axiality of the material behaviour. The three spatial scales are the poly-crystalline, grain and domain scale. Together with appropriate scale transitions rules and models for local magnetic behaviour at each scale, the model is able to describe the magneto-elastic behaviour (magnetostriction and hysteresis) at the macroscale, although the data input is merely based on a set of physical constants. Introducing a new energy density function that describes the demagnetisation field, the anhysteretic multi-scale energy-based material model is extended to the hysteretic case. The hysteresis behaviour is included at the domain scale according to the micro-magnetic domain theory while preserving a valid description for the magneto-elastic coupling. The model is verified using existing measurement data for different mechanical stress levels. - Highlights: • A ferromagnetic hysteretic energy-based multi-scale material model is proposed. • The hysteresis is obtained by new proposed hysteresis energy density function. • Avoids tedious parameter identification.

  13. Characterization of a facility for the measurement of fission fragment transport effects: experimental determination of the fission rates for fissile and fissionable isotopes

    International Nuclear Information System (INIS)

    Benetti, P.; Raselli, G.L.; Tigliole, A. Borio di; Cagnazzo, M.; Cesana, A.; Mongelli, S.; Terrani, M.

    2002-01-01

    The transfer facility of the LENA laboratory allows the direct neutron irradiation of fissionable material in the D channel of the TRIGA reactor. A test measurement carried out with a ionization chamber and a 239 Pu sample shows the possibility to use this tool for the study of the transport effects of the fission fragment emerging from thin layers of fissile materials. (author)

  14. Fission level densities

    International Nuclear Information System (INIS)

    Maslov, V.M.

    1998-01-01

    Fission level densities (or fissioning nucleus level densities at fission saddle deformations) are required for statistical model calculations of actinide fission cross sections. Back-shifted Fermi-Gas Model, Constant Temperature Model and Generalized Superfluid Model (GSM) are widely used for the description of level densities at stable deformations. These models provide approximately identical level density description at excitations close to the neutron binding energy. It is at low excitation energies that they are discrepant, while this energy region is crucial for fission cross section calculations. A drawback of back-shifted Fermi gas model and traditional constant temperature model approaches is that it is difficult to include in a consistent way pair correlations, collective effects and shell effects. Pair, shell and collective properties of nucleus do not reduce just to the renormalization of level density parameter a, but influence the energy dependence of level densities. These effects turn out to be important because they seem to depend upon deformation of either equilibrium or saddle-point. These effects are easily introduced within GSM approach. Fission barriers are another key ingredients involved in the fission cross section calculations. Fission level density and barrier parameters are strongly interdependent. This is the reason for including fission barrier parameters along with the fission level densities in the Starter File. The recommended file is maslov.dat - fission barrier parameters. Recent version of actinide fission barrier data obtained in Obninsk (obninsk.dat) should only be considered as a guide for selection of initial parameters. These data are included in the Starter File, together with the fission barrier parameters recommended by CNDC (beijing.dat), for completeness. (author)

  15. Fast fission phenomena

    International Nuclear Information System (INIS)

    Gregoire, Christian.

    1982-03-01

    Experimental studies of fast fission phenomena are presented. The paper is divided into three parts. In the first part, problems associated with fast fission processes are examined in terms of interaction potentials and a dynamic model is presented in which highly elastic collisions, the formation of compound nuclei and fast fission appear naturally. In the second part, a description is given of the experimental methods employed, the observations made and the preliminary interpretation of measurements suggesting the occurence of fast fission processes. In the third part, our dynamic model is incorporated in a general theory of the dissipative processes studied. This theory enables fluctuations associated with collective variables to be calculated. It is applied to highly inelastic collisions, to fast fission and to the fission dynamics of compound nuclei (for which a schematic representation is given). It is with these calculations that the main results of the second part can be interpreted [fr

  16. A new method for the experimental study of fatigue behaviour of thermoplastic materials

    Directory of Open Access Journals (Sweden)

    M. Sanità

    2008-10-01

    Full Text Available Nowadays most industrial realities undergo a strong push to improve cost-effectiveness, productivity and quality of manufactured products. In particular we focussed our attention in the area of design of plastic structural components, including both optimization of existing structures and design of new ones. In this case, but the following considerations have a more general value, these needs could be translated into demanding requirements of cost-effectiveness, weight reduction, reduced time-to-market with guarantee reliability. From a material perspective this means demanding mechanical performances, attention to safety margins and need of a better control of key design parameters. To obtain these results, we need to develop a new approach and effective tools in the design of plastic materials and components aimed at tailoring part behaviour to endurance and performance requirements.The target of the project is to find effective tools for predicting life endurance and damage evolution of plastic materials and components under mechanical/thermal service loading, in order to support the development of new material formulations and the design and optimization of structural components. In a particular way, we focussed our work in the characterization and modellization of materials durability and damage mechanisms.One of the main problems related to materials durability is due to fatigue failure. Fatigue process is a progressive weakening of a component with increasing time under load such that loads to be supported satisfactorily for short duration produce failure after long durations [1, 2, 3]. Fatigue failure should not be thought only as the breaking of the specimen into two separated pieces, but as a progressive material damage accumulation [2]. Material damage during fatigue loading manifests as progressive reduction of stiffness and as creep [5].As standard fatigue testing are expensive in terms of money and time, it is essential to develop

  17. Multi-scale modeling of the thermo-hydro- mechanical behaviour of heterogeneous materials. Application to cement-based materials under severe loads

    International Nuclear Information System (INIS)

    Grondin, Frederic Alain

    2005-01-01

    The work of modeling presented here relates to the study of the thermo-hydro- mechanical behaviour of porous materials based on hydraulic binder such as concrete, High Performance Concrete or more generally cement-based materials. This work is based on the exploitation of the Digital Concrete model, of the finite element code Symphonie developed in the Scientific and Technical Centre for Building (CSTB), in coupling with the homogenization methods to obtain macroscopic behaviour laws drawn from the Micro-Macro relations. Scales of investigation, macroscopic and microscopic, has been exploited by simulation in order to allow the comprehension fine of the behaviour of cement-based materials according to thermal, hydrous and mechanical loads. It appears necessary to take into account various scales of modeling. In order to study the behaviour of the structure, we are brought to reduce the scale of investigation to study the material more particularly. The research tasks presented suggest a new approach for the identification of the multi-physic behaviour of materials by simulation. In complement of the purely experimental approach, based on observations on the sample with measurements of the apparent parameters on the macroscopic scale, this new approach allows to obtain the fine analysis of elementary mechanisms in acting within the material. These elementary mechanisms are at the origin of the evolution of the macroscopic parameters measured in experimental tests. In this work, coefficients of the thermo-hydro-mechanical behaviour law of porous materials and the equivalent hydraulic conductivity were obtained by a multi-scales approach. Applications has been carried out on the study of the damaged behaviour of cement-based materials, in the objective to determine the elasticity tensor and the permeability tensor of a High Performance Concrete at high temperatures under a mechanical load. Also, the study of the strain evolution of cement-based materials at low

  18. 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

  19. Flexural Behaviour of Precast Aerated Concrete Panel (PACP with Added Fibrous Material: An Overview

    Directory of Open Access Journals (Sweden)

    Abdul Rahim Noor Hazlin

    2017-01-01

    Full Text Available The usage of precast aerated concrete panel as an IBS system has become the main alternative to conventional construction system. The usage of this panel system contributes to a sustainable and environmental friendly construction. This paper presents an overview of the precast aerated concrete panel with added fibrous material (PACP. PACP is fabricated from aerated foamed concrete with added Polypropylene fibers (PP. The influence of PP on the mechanical properties of PACP are studied and reviewed from previous research. The structural behaviour of precast concrete panel subjected to flexure load is also reviewed. It is found that PP has significant affects on the concrete mixture’s compressive stregth, tensile strength and flexural strength. It is also found that PP manage to control the crack propagation in the concrete panel.

  20. Fatigue crack growth behaviour of carbon steel piping material subjected to single overload/under-load

    International Nuclear Information System (INIS)

    Arora, Punit; Tripathi, R.; Singh, P.K.; Bhasin, V.; Vijayan, P.K.

    2016-01-01

    The objective of the present study is to understand the Fatigue Crack Growth Rate (FCGR) behaviour after single over-load/ under-load event on carbon steel piping material. The tests have been carried out on standard Compact Tension (CT) specimens. The effect of different crack length to width ratio (a/W) of specimen and overload/under-load ratios on FCGR have been studied. The studies have shown significant reduction in FCG rate after overload event. The strain field has been measured using Digital Image Correlation (DIC) technique ahead of the crack tip to quantify the plastic zone size due to overload and constant amplitude load. In addition, plastic zone calculations have also been carried out using 3D finite element analyses for the prediction of post overload FCGR/ life. The predicted FCGR are in agreement with experimentally determined FCGR. (author)

  1. Unravelling the High-Pressure Behaviour of Dye-Zeolite L Hybrid Materials

    Directory of Open Access Journals (Sweden)

    Lara Gigli

    2018-02-01

    Full Text Available Self-assembly of chromophores nanoconfined in porous materials such as zeolite L has led to technologically relevant host-guest systems exploited in solar energy harvesting, photonics, nanodiagnostics and information technology. The response of these hybrid materials to compression, which would be crucial to enhance their application range, has never been explored to date. By a joint high-pressure in situ synchrotron X-ray powder diffraction and ab initio molecular dynamics approach, herein we unravel the high-pressure behaviour of hybrid composites of zeolite L with fluorenone dye. High-pressure experiments were performed up to 6 GPa using non-penetrating pressure transmitting media to study the effect of dye loading on the structural properties of the materials under compression. Computational modelling provided molecular-level insight on the response to compression of the confined dye assemblies, evidencing a pressure-induced strengthening of the interaction between the fluorenone carbonyl group and zeolite L potassium cations. Our results reveal an impressive stability of the fluorenone-zeolite L composites at GPa pressures. The remarkable resilience of the supramolecular organization of dye molecules hyperconfined in zeolite L channels may open the way to the realization of optical devices able to maintain their functionality under extreme conditions.

  2. Effect of material variation on the biomechanical behaviour of orthodontic fixed appliances: a finite element analysis.

    Science.gov (United States)

    Papageorgiou, Spyridon N; Keilig, Ludger; Hasan, Istabrak; Jäger, Andreas; Bourauel, Christoph

    2016-06-01

    Biomechanical analysis of orthodontic tooth movement is complex, as many different tissues and appliance components are involved. The aim of this finite element study was to assess the relative effect of material alteration of the various components of the orthodontic appliance on the biomechanical behaviour of tooth movement. A three-dimensional finite element solid model was constructed. The model consisted of a canine, a first, and a second premolar, including the surrounding tooth-supporting structures and fixed appliances. The materials of the orthodontic appliances were alternated between: (1) composite resin or resin-modified glass ionomer cement for the adhesive, (2) steel, titanium, ceramic, or plastic for the bracket, and (3) β-titanium or steel for the wire. After vertical activation of the first premolar by 0.5mm in occlusal direction, stress and strain calculations were performed at the periodontal ligament and the orthodontic appliance. The finite element analysis indicated that strains developed at the periodontal ligament were mainly influenced by the orthodontic wire (up to +63 per cent), followed by the bracket (up to +44 per cent) and the adhesive (up to +4 per cent). As far as developed stresses at the orthodontic appliance are concerned, wire material had the greatest influence (up to +155 per cent), followed by bracket material (up to +148 per cent) and adhesive material (up to +8 per cent). The results of this in silico study need to be validated by in vivo studies before they can be extrapolated to clinical practice. According to the results of this finite element study, all components of the orthodontic fixed appliance, including wire, bracket, and adhesive, seem to influence, to some extent, the biomechanics of tooth movement. © The Author 2015. Published by Oxford University Press on behalf of the European Orthodontic Society. All rights reserved. For permissions, please email: journals.permissions@oup.com.

  3. Fission product yields

    International Nuclear Information System (INIS)

    Valenta, V.; Hep, J.

    1978-01-01

    Data are summed up necessary for determining the yields of individual fission products from different fissionable nuclides. Fractional independent yields, cumulative and isobaric yields are presented here for the thermal fission of 235 U, 239 Pu, 241 Pu and for fast fission (approximately 1 MeV) of 235 U, 238 U, 239 Pu, 241 Pu; these values are included into the 5th version of the YIELDS library, supplementing the BIBFP library. A comparison is made of experimental data and possible improvements of calculational methods are suggested. (author)

  4. Fission neutron multiplicity calculations

    International Nuclear Information System (INIS)

    Maerten, H.; Ruben, A.; Seeliger, D.

    1991-01-01

    A model for calculating neutron multiplicities in nuclear fission is presented. It is based on the solution of the energy partition problem as function of mass asymmetry within a phenomenological approach including temperature-dependent microscopic energies. Nuclear structure effects on fragment de-excitation, which influence neutron multiplicities, are discussed. Temperature effects on microscopic energy play an important role in induced fission reactions. Calculated results are presented for various fission reactions induced by neutrons. Data cover the incident energy range 0-20 MeV, i.e. multiple chance fission is considered. (author). 28 refs, 13 figs

  5. Intermediate energy nuclear fission

    International Nuclear Information System (INIS)

    Hylten, G.

    1982-01-01

    Nuclear fission has been investigated with the double-kinetic-energy method using silicon surface barrier detectors. Fragment energy correlation measurements have been made for U, Th and Bi with bremsstrahlung of 600 MeV maximum energy. Distributions of kinetic energy as a function of fragment mass are presented. The results are compared with earlier photofission data and in the case of bismuth, with calculations based on the liquid drop model. The binary fission process in U, Yb, Tb, Ce, La, Sb, Ag and Y induced by 600 MeV protons has been investigated yielding fission cross sections, fragment kinetic energies, angular correlations and mass distributions. Fission-spallation competition calculations are used to deduce values of macroscopic fission barrier heights and nuclear level density parameter values at deformations corresponding to the saddle point shapes. We find macroscopic fission barriers lower than those predicted by macroscopic theories. No indication is found of the Businaro Gallone limit expected to occur somewhere in the mass range A = 100 to A = 140. For Ce and La asymmetric mass distributions similar to those in the actinide region are found. A method is described for the analysis of angular correlations between complementary fission products. The description is mainly concerned with fission induced by medium-energy protons but is applicable also to other projectiles and energies. It is shown that the momentum and excitation energy distributions of cascade residuals leading to fission can be extracted. (Author)

  6. Fission rate measurements in fuel plate type assembly reactor cores

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1988-01-01

    The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

  7. Long Term Behaviour of Cementitious Materials in the Korean Repository Environment

    International Nuclear Information System (INIS)

    Park, J.-W.; Kim, C.-L.

    2013-01-01

    The safe management of radioactive waste is a national task required for sustainable generation of nuclear power and for energy self-reliance in Korea. After the selection of the final candidate site for low- and intermediate-level waste (LILW) disposal in Korea, a construction and operation license was issued for the Wolsong LILW Disposal Center (WLDC) for the first stage of disposal. Underground silo type disposal has been determined for the initial phase. The engineered barrier system of the disposal silo consists of waste packages, disposal containers, backfills, and a concrete lining. Main objective of our study in this IAEA-CRP is to investigate closure concepts and cementitious backfill materials for the closure of silos. For this purpose, characterisation of cementitious materials, development of silo closure concept, and evaluation of long-term behaviour of cementitious materials, including concrete degradation in repository environment, have been carried out. The overall implementation plan for the CRP comprises performance testing for the physic-chemical properties of cementitious materials, degradation modelling of concrete structures, comparisons of performance for silo closure options, radionuclide transport modelling (considering concrete degradation in repository conditions), and the implementation of an input parameter database and quality assurance for safety/performance assessment. In particular, the concrete degradation modelling study has been focused on the corrosion of reinforcement steel induced by chloride attack, which was of primary concern in the safety assessment of the WLDC. A series of electrochemical experiments were conducted to investigate the effect of dissolved oxygen, pH, and Cl on the corrosion rate of reinforcing steel in a concrete structure saturated with groundwater. Laboratory-scale experiments and a thermodynamic modelling were performed to understand the porosity change of cement pastes, which were prepared using

  8. Creep behaviour and microstructure of the ferritic material No. 1-6770 under irradiation

    International Nuclear Information System (INIS)

    Herschbach, K.; Ehrlich, K.; Materna, E.

    Creep behaviour under irradiation of the ferritic steel-DIN-1-6770 is quite different of austenitic steel behaviour, in particular temperature sensitivity is important and response to stress is non linear. The microstructure stays unchanged

  9. Critical masses of miniexplosion in fission-fusion hybrid systems

    Energy Technology Data Exchange (ETDEWEB)

    Kaliski, S [Polska Akademia Nauk, Warsaw. Inst. Podstawowych Problemow Techniki

    1976-01-01

    The critical mass of the fissionable material subjected to the explosive compression and the action of the neutron stream originating from the process of D-T fusion in the spherical cavity was estimated. High energy recovery from the fissionable material was obtained and the energy of the laser pulse was minimized.

  10. A nonlinear dynamical system approach for the yielding behaviour of a viscoplastic material.

    Science.gov (United States)

    Burghelea, Teodor; Moyers-Gonzalez, Miguel; Sainudiin, Raazesh

    2017-03-08

    A nonlinear dynamical system model that approximates a microscopic Gibbs field model for the yielding of a viscoplastic material subjected to varying external stresses recently reported in R. Sainudiin, M. Moyers-Gonzalez and T. Burghelea, Soft Matter, 2015, 11(27), 5531-5545 is presented. The predictions of the model are in fair agreement with microscopic simulations and are in very good agreement with the micro-structural semi-empirical model reported in A. M. V. Putz and T. I. Burghelea, Rheol. Acta, 2009, 48, 673-689. With only two internal parameters, the nonlinear dynamical system model captures several key features of the solid-fluid transition observed in experiments: the effect of the interactions between microscopic constituents on the yield point, the abruptness of solid-fluid transition and the emergence of a hysteresis of the micro-structural states upon increasing/decreasing external forces. The scaling behaviour of the magnitude of the hysteresis with the degree of the steadiness of the flow is consistent with previous experimental observations. Finally, the practical usefulness of the approach is demonstrated by fitting a rheological data set measured with an elasto-viscoplastic material.

  11. The nuclear fission

    International Nuclear Information System (INIS)

    Fiorentino, J.

    1983-01-01

    The nuclear fission process considering initially the formation of compound nucleus and finishing with radioactive decay of fission products is studied. The process is divided in three parts which consist of the events associated to the nucleus of intermediate transitional state, the scission configuration, and the phenomenum of post scission. (M.C.K.) [pt

  12. Fission gas detection system

    International Nuclear Information System (INIS)

    Colburn, R.P.

    1984-01-01

    A device for collecting fission gas released by failed fuel rods which device uses a filter adapted to pass coolant but to block passage of fission gas bubbles due to the surface tension of the bubbles. The coolant may be liquid metal. (author)

  13. Muon-induced fission

    International Nuclear Information System (INIS)

    Polikanov, S.

    1980-01-01

    A review of recent experimental results on negative-muon-induced fission, both of 238 U and 232 Th, is given. Some conclusions drawn by the author are concerned with muonic atoms of fission fragments and muonic atoms of the shape isomer of 238 U. (author)

  14. Relativistic Coulomb Fission

    Science.gov (United States)

    Norbury, John W.

    1992-01-01

    Nuclear fission reactions induced by the electromagnetic field of relativistic nuclei are studied for energies relevant to present and future relativistic heavy ion accelerators. Cross sections are calculated for U-238 and Pu-239 fission induced by C-12, Si-28, Au-197, and U-238 projectiles. It is found that some of the cross sections can exceed 10 b.

  15. 3-D analysis of fatigue crack behaviour in a shot peened steam turbine blade material

    Energy Technology Data Exchange (ETDEWEB)

    He, B.Y., E-mail: Binyan.he@soton.ac.uk [Engineering Materials, Faculty of Engineering and the Environment, University of Southampton, Southampton SO17 1BJ (United Kingdom); Katsamenis, O.L. [muVIS X-ray Imaging Centre, Faculty of Engineering and the Environment, University of Southampton, Southampton SO17 1BJ (United Kingdom); Mellor, B.G.; Reed, P.A.S. [Engineering Materials, Faculty of Engineering and the Environment, University of Southampton, Southampton SO17 1BJ (United Kingdom)

    2015-08-26

    Serial mechanical sectioning and high resolution X-ray tomography have been used to study the three-dimensional morphology of small fatigue cracks growing in a 12 Cr tempered martensitic steam turbine blade material. A range of surface conditions has been studied, namely polished and shot peened (with varying levels of intensity). In the polished (unpeened) condition, inclusions (alumina and manganese sulphide) played an important role in initiating and controlling early fatigue crack behaviour. When fatigue cracks initiated from an alumina stringer, the crack morphology was normally dominated by single stringers, which were always in the centre of the fatigue crack, indicating its primary role in initiation. Manganese sulphide inclusion groups however seemed to dominate and affect the crack path along both the surface and depth crack growth directions. The more intensely shot peened condition did not however evidence inclusion or stringer affected fatigue crack initiation or growth behaviour; sub-surface crack coalescence being clearly observed by both serial sectioning and computed tomography (CT) imaging techniques at a depth of about 150–180 μm. These sub-surface crack coalescences can be linked to both the extent of the compressive residual stress as well as the depth of the plastic deformation arising from the intense shot peening process. Shot peening appears to provide a different defect population that initiates fatigue cracks and competes with the underlying metallurgical defect populations. The most beneficial shot peening process would in this case appear to “deactivate” the original metallurgical defect population and substitute a known defect distribution from the shot peening process from which fatigue cracks grow rather slowly in the strain hardened surface layer which also contains compressive residual stresses. A benefit to fatigue life in bending, even under Low Cycle Fatigue (LCF) conditions, has been observed in these tests if a

  16. Melt-Flow Behaviours of Thermoplastic Materials under Fire Conditions: Recent Experimental Studies and Some Theoretical Approaches

    Directory of Open Access Journals (Sweden)

    Paul Joseph

    2015-12-01

    Full Text Available Polymeric materials often exhibit complex combustion behaviours encompassing several stages and involving solid phase, gas phase and interphase. A wide range of qualitative, semi-quantitative and quantitative testing techniques are currently available, both at the laboratory scale and for commercial purposes, for evaluating the decomposition and combustion behaviours of polymeric materials. They include, but are not limited to, techniques such as: thermo-gravimetric analysis (TGA, oxygen bomb calorimetry, limiting oxygen index measurements (LOI, Underwriters Laboratory 94 (UL-94 tests, cone calorimetry, etc. However, none of the above mentioned techniques are capable of quantitatively deciphering the underpinning physiochemical processes leading to the melt flow behaviour of thermoplastics. Melt-flow of polymeric materials can constitute a serious secondary hazard in fire scenarios, for example, if they are present as component parts of a ceiling in an enclosure. In recent years, more quantitative attempts to measure the mass loss and melt-drip behaviour of some commercially important chain- and step-growth polymers have been accomplished. The present article focuses, primarily, on the experimental and some theoretical aspects of melt-flow behaviours of thermoplastics under heat/fire conditions.

  17. Melt-Flow Behaviours of Thermoplastic Materials under Fire Conditions: Recent Experimental Studies and Some Theoretical Approaches.

    Science.gov (United States)

    Joseph, Paul; Tretsiakova-McNally, Svetlana

    2015-12-15

    Polymeric materials often exhibit complex combustion behaviours encompassing several stages and involving solid phase, gas phase and interphase. A wide range of qualitative, semi-quantitative and quantitative testing techniques are currently available, both at the laboratory scale and for commercial purposes, for evaluating the decomposition and combustion behaviours of polymeric materials. They include, but are not limited to, techniques such as: thermo-gravimetric analysis (TGA), oxygen bomb calorimetry, limiting oxygen index measurements (LOI), Underwriters Laboratory 94 (UL-94) tests, cone calorimetry, etc. However, none of the above mentioned techniques are capable of quantitatively deciphering the underpinning physiochemical processes leading to the melt flow behaviour of thermoplastics. Melt-flow of polymeric materials can constitute a serious secondary hazard in fire scenarios, for example, if they are present as component parts of a ceiling in an enclosure. In recent years, more quantitative attempts to measure the mass loss and melt-drip behaviour of some commercially important chain- and step-growth polymers have been accomplished. The present article focuses, primarily, on the experimental and some theoretical aspects of melt-flow behaviours of thermoplastics under heat/fire conditions.

  18. Corrosion behaviour of container materials for geological disposal of high-level waste. Joint annual progress report 1983

    International Nuclear Information System (INIS)

    1985-01-01

    Within the framework of the Community R and D programme on management and storage of radioactive waste (shared-cost action), a research activity is aiming at the assessment of corrosion behaviour of potential container materials for geological disposal of vitrified high-level wastes. In this report, the results obtained during the year 1983 are described. Research performed at the Studiecentrum voor Kernenergie/Centre d'Etudes de l'Energie Nucleaire (SCK/CEN) at Mol (B), concerns the corrosion behaviour in clay environments. The behaviour in salt is tested by the Kernforschungszentrum (KfK) at Karlsruhe (D). Corrosion behaviour in granitic environments is being examined by the Commissariat a l'Energie Atomique (CEA) at Fontenay-aux-Roses (F) and the Atomic Energy Research Establishment (AERE) at Harwell (UK); the first is concentrating on corrosion-resistant materials and the latter on corrosion-allowance materials. Finally, the Centre National de la Recherche Scientifique (CNRS) at Vitry (F) is examining the formation and behaviour of passive layers on the metal alloys in the various environments

  19. Study of hypernuclei fission

    International Nuclear Information System (INIS)

    Malek, F.

    1990-01-01

    This work is about PS177 experience made on LEAR machine at CERN in 1988. The annihilation reaction of anti protons on a target of Bismuth or Uranium is studied. Lambda particles are produced by this reaction, in the nucleus in 2% of cases 7.1 10 -3 hypernuclei by stopped antiproton in the target are produced. The prompt hypernucleus fission probability of uranium is 75% and that of Bismuth 10%. The mass distribution of fission fragments is symmetrical ((≡ the excitation energy of the nucleus is very high). If the nucleus hasn't fissioned, the non-mesonic lambda decay, gives it an energy of 100 MeV, what allows to fission later. This fission is delayed because the hypernucleus lifetime is 1.3 +0.25 -0.21 10 -10 sec for Bismuth [fr

  20. The nuclear fission process

    International Nuclear Information System (INIS)

    Wagemans, C.

    1991-01-01

    Fifty years after its discovery, the nuclear fission phenomenon is of recurring interest. When its fundamental physics aspects are considered, fission is viewed in a very positive way, which is reflected in the great interest generated by the meetings and large conferences organized for the 50th anniversary of its discovery. From a purely scientific and practical point of view, a new book devoted to the (low energy) nuclear fission phenomenon was highly desirable considering the tremendous amount of new results obtained since the publication of the book Nuclear Fission by Vandenbosch and Huizenga in 1973 (Academic Press). These new results could be obtained thanks to the growth of technology, which enabled the construction of powerful new neutron sources, particle and heavy ion accelerators, and very performant data-acquisition and computer systems. The re-invention of the ionization chamber, the development of large fission fragment spectrometers and sophisticated multiparameter devices, and the production of exotic isotopes also contributed significantly to an improved understanding of nuclear fission. This book is written at a level to introduce graduate students to the exciting subject of nuclear fission. The very complete list of references following each chapter also makes the book very useful for scientists, especially nuclear physicists. The book has 12 chapters covering the fission barrier and the various processes leading to fission as well as the characteristics of the various fission reaction products. In order to guarantee adequate treatment of the very specialized research fields covered, several distinguished scientists actively involved in some of these fields were invited to contribute their expertise as authors or co-authors of the different chapters

  1. Dynamics of process at the final stage of nuclear fission

    International Nuclear Information System (INIS)

    Koljari, I.G.; Mavlitov, N.D.

    2005-01-01

    Numerous experimental data show, that the final stage of nuclear fission near to a scission point plays an essential role at formation of characteristics of fission products. At the description of a final stage of fission there is a number of problems: Definition of the form of the nuclear near the scission point and definition forms of a fission fragments; The account of dynamic processes in compound nuclear directly before of fission. The condition of the quasistatic al adiabatic process - dS/dt=0 - is applied in a point of transition from the uniform compound nuclei to several forms for definition of generalized coordinates and speeds. Calculation of dependence of post neutrons from nuclear mass of fission fragments for reactions is α+ 83 Bi 209 → 85 At 213 (E lab = 45 MeV); α+ 92 U 242 → 94 Pu 242 (E lab = 45 MeV); 8 O 18 + 79 Au 197 → 97 Fr 215 (E lab = 159 MeV). System of equations, which describes behaviour of system in a point of nuclear fission-transition from the uniform form to system of a two (and, probably more) fission fragments is given. The system of the equations allows in a fission point to define the generalized coordinates, and the generalized speeds for each of the generalized coordinates of collective deformation variables

  2. Influence of pH and oxygen content of buffer solutions on the corrosion behaviour of metallic materials

    International Nuclear Information System (INIS)

    Wiedemann, K.H.

    1977-05-01

    The application of solutions to the decontamination of materials in nuclear installations is based on the condition that their corrosion behaviour is clearly understood. Since electrochemical corrosion is due to cathodic and anodic partial reactions which are influenced in different ways by the pH of the solution and the oxygen content it is suggested that the results of electrochemical experiments with buffer solutions be used as a model for predicting the corrosion behaviour of materials in other solutions. In the tests described here potentio-kinetic current-potential-curves have been traced and galvanic corrosion tests have been made. The results obtained in ascorbic acid, potassium hydrogen phthalate, ammonium citrate and acetate, sodium and potassium tartrate, ammonium hydrogen phosphate, sodium carbonate, hexamethylene tetramin, ethylene diamine enable - on the basis of summarized current-potential-curves - the metals studied to be classified in four groups characterized by clear differences concerning the influence of pH on the corrosion behaviour. (Auth.)

  3. Fission 2009 4. International Workshop on Nuclear Fission and Fission Product Spectroscopy - Compilation of slides

    International Nuclear Information System (INIS)

    2009-01-01

    This conference is dedicated to the last achievements in experimental and theoretical aspects of the nuclear fission process. The topics include: mass, charge and energy distribution, dynamical aspect of the fission process, nuclear data evaluation, quasi-fission and fission lifetime in super heavy elements, fission fragment spectroscopy, cross-section and fission barrier, and neutron and gamma emission. This document gathers the program of the conference and the slides of the presentations

  4. Behaviour of the radionuclides in the environment

    International Nuclear Information System (INIS)

    Lauria, Dejanira da Costa

    2007-01-01

    This chapter approaches the behaviour of radionuclides in the environment, the isotopes of the natural radioactive series, some aspects of the isotope behaviour on the environment and fission and activation radioactive isotopes

  5. Protective Behaviour of Citizens to Transport Accidents Involving Hazardous Materials: A Discrete Choice Experiment Applied to Populated Areas nearby Waterways.

    Directory of Open Access Journals (Sweden)

    Esther W de Bekker-Grob

    Full Text Available To improve the information for and preparation of citizens at risk to hazardous material transport accidents, a first important step is to determine how different characteristics of hazardous material transport accidents will influence citizens' protective behaviour. However, quantitative studies investigating citizens' protective behaviour in case of hazardous material transport accidents are scarce.A discrete choice experiment was conducted among subjects (19-64 years living in the direct vicinity of a large waterway. Scenarios were described by three transport accident characteristics: odour perception, smoke/vapour perception, and the proportion of people in the environment that were leaving at their own discretion. Subjects were asked to consider each scenario as realistic and to choose the alternative that was most appealing to them: staying, seeking shelter, or escaping. A panel error component model was used to quantify how different transport accident characteristics influenced subjects' protective behaviour.The response was 44% (881/1,994. The predicted probability that a subject would stay ranged from 1% in case of a severe looking accident till 62% in case of a mild looking accident. All three transport accident characteristics proved to influence protective behaviour. Particularly a perception of strong ammonia or mercaptan odours and visible smoke/vapour close to citizens had the strongest positive influence on escaping. In general, 'escaping' was more preferred than 'seeking shelter', although stated preference heterogeneity among subjects for these protective behaviour options was substantial. Males were less willing to seek shelter than females, whereas elderly people were more willing to escape than younger people.Various characteristics of transport accident involving hazardous materials influence subjects' protective behaviour. The preference heterogeneity shows that information needs to be targeted differently depending on

  6. Non-fossil reduction materials in the silicon process - properties and behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Myrhaug, Edin Henrik

    2003-07-01

    The purpose of this work has been to clarify the effect of using biocarbon as a reduction material in the silicon process. It was decided to compare the biocarbon with fossil carbon and find possible differences both on process performance and eventually on product quality. The elements in the raw materials added to the silicon process goes into three different products: silicon metal, silica dust and into open air. Based on analysis of raw materials and of produced silicon metal and microsilica extensive material balances have been established. One important result from these are the distribution factors that indicate how much of the trace elements that goes into each medium. Another result is that the boiling point of an element or a compound gives a good indication of were it ends. A high boiling point indicates that the element ends up in the silicon metal, while a low boiling point indicates that the element goes with off-gas into air. With an intermediate boiling point, the element goes into the silica dust. The SiO-reactivity of the reduction materials are commonly acknowledged to affect strongly the productivity and consumption figures of the silicon process. Based on data from thermogravimetric experiments with chemical reaction between carbonaceous spheres and SiO-gas, kinetic parameters have been estimated from the shrinking core model for some selected reduction materials of various sizes and spanning a wide range of SiO-reactivity figures. This model describes the degree of conversion versus time for a single sphere where the chemical reaction progresses in a topochemical manner from the outer surface of the solid towards the centre forming a porous product layer around an unreacted shrinking core. This behaviour is for the selected reduction materials to a large extent supported by an investigation of cross section pictures of fully and 50% converted spheres obtained with a microprobe. The estimated kinetic parameters obtained from the

  7. Internat. conference about the radiation behaviour of metallic canning and structure materials for fast breeders in Ajaccio (Korsika)

    International Nuclear Information System (INIS)

    Anderko, K.; Ehrlich, K.

    1979-01-01

    The program includes 48 plenary reports as well as 22 contributions in the form of a poster view and has the following structure: - swelling of ferritic steel - structural instability under radiation - theory of swelling - experiments about the swelling of austenitic steels - mechanical properties after radiation - fuel element behaviour and material optimization - radiation creeping. Additional to the items respecting the conference titel some material problems of the fusion reactor were discussed. (orig./RW) [de

  8. Investigation of delayed fission gas release

    International Nuclear Information System (INIS)

    Cayet, Nicolas

    1996-05-01

    The study of the fission gas release process in the high burnup rig IFA-562 has revealed a particular fuel behaviour: a delay in the fission gas release process. It appeared that an important release of gas was measured by the pressure transducers once the power had decreased, whereas, during steady-state operation, the pressure did not increase very much. After examinations, the gap size has been concluded to be the main parameter involving this delay. However the burnup could have been a potential factor, its role is mainly to close the gap by swelling. The observations of low burnup rods have shown the same delayed fission gas release, the gap being small by design and closed essentially by thermal expansion. The study of the kinetics has demonstrated the time-independency of the phenomenon. Thus the proposed mechanism driving this delayed fission gas release would involve three consecutives stages. During steady-state, the gas is released into the interlinkage network of grain boundary bubbles and cracks. Due to the closed gap, the gas is trapped in some void volumes, unable to escape the pellet. During power reduction, the gap and some old/new cracks open, immediately providing a path for the gas to the pressure transducers and explaining this delay in the fission gas release. (author)

  9. Corrosion behaviour of boiler tube materials during combustion of fuels containing Zn and Pb

    Energy Technology Data Exchange (ETDEWEB)

    Bankiewicz, D.

    2012-11-01

    Many power plants burning challenging fuels such as waste-derived fuels experience failures of the superheaters and/or increased waterwall corrosion due to aggressive fuel components already at low temperatures. To minimize corrosion problems in waste-fired boilers, the steam temperature is currently kept at a relatively low level which drastically limits power production efficiency. The elements found in deposits of waste and waste-derived fuels burning boilers that are most frequently associated with high-temperature corrosion are: Cl, S, and there are also indications of Br; alkali metals, mainly K and Na, and heavy metals such as Pb and Zn. The low steam pressure and temperature in waste-fired boilers also influence the temperature of the waterwall steel which is nowadays kept in the range of 300 deg C - 400 deg C. Alkali chloride (KCl, NaCl) induced high-temperature corrosion has not been reported to be particularly relevant at such low material temperatures, but the presence of Zn and Pb compounds in the deposits have been found to induce corrosion already in the 300 deg C - 400 deg C temperature range. Upon combustion, Zn and Pb may react with Cl and S to form chlorides and sulphates in the flue gases. These specific heavy metal compounds are of special concern due to the formation of low melting salt mixtures. These low melting, gaseous or solid compounds are entrained in the flue gases and may stick or condense on colder surfaces of furnace walls and superheaters when passing the convective parts of the boiler, thereby forming an aggressive deposit. A deposit rich in heavy metal (Zn, Pb) chlorides and sulphates increases the risk for corrosion which can be additionally enhanced by the presence of a molten phase. The objective of this study was to obtain better insight into high-temperature corrosion induced by Zn and Pb and to estimate the behaviour and resistance of some boiler superheater and waterwall materials in environments rich in those heavy metals

  10. NUMERICAL MODELLING OF THE SOIL BEHAVIOUR BY USING NEWLY DEVELOPED ADVANCED MATERIAL MODEL

    Directory of Open Access Journals (Sweden)

    Jan Veselý

    2017-02-01

    Full Text Available This paper describes a theoretical background, implementation and validation of the newly developed Jardine plastic hardening-softening model (JPHS model, which can be used for numerical modelling of the soils behaviour. Although the JPHS model is based on the elasto-plastic theory, like the Mohr-Coulomb model that is widely used in geotechnics, it contains some improvements, which removes the main disadvantages of the MC model. The presented model is coupled with an isotopically hardening and softening law, non-linear elastic stress-strain law, non-associated elasto-plastic material description and a cap yield surface. The validation of the model is done by comparing the numerical results with real measured data from the laboratory tests and by testing of the model on the real project of the tunnel excavation. The 3D numerical analysis is performed and the comparison between the JPHS, Mohr-Coulomb, Modified Cam-Clay, Hardening small strain model and monitoring in-situ data is done.

  11. 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)

  12. Digital signal processing for velocity measurements in dynamical material's behaviour studies

    International Nuclear Information System (INIS)

    Devlaminck, Julien; Luc, Jerome; Chanal, Pierre-Yves

    2014-01-01

    In this work, we describe different configurations of optical fiber interferometers (types Michelson and Mach-Zehnder) used to measure velocities during dynamical material's behaviour studies. We detail the algorithms of processing developed and optimized to improve the performance of these interferometers especially in terms of time and frequency resolutions. Three methods of analysis of interferometric signals were studied. For Michelson interferometers, the time-frequency analysis of signals by Short-Time Fourier Transform (STFT) is compared to a time-frequency analysis by Continuous Wavelet Transform (CWT). The results have shown that the CWT was more suitable than the STFT for signals with low signal-to-noise, and low velocity and high acceleration areas. For Mach- Zehnder interferometers, the measurement is carried out by analyzing the phase shift between three interferometric signals (Triature processing). These three methods of digital signal processing were evaluated, their measurement uncertainties estimated, and their restrictions or operational limitations specified from experimental results performed on a pulsed power machine. (authors)

  13. Contribution to the study of cementitious and clayey materials behaviour in the context of deep geological disposal: transport aspect, durability and thermo-hydro-mechanical behaviour

    International Nuclear Information System (INIS)

    Galle, C.

    2011-07-01

    Deep geological formation disposal is the reference solution in France for the management of medium and high activities radioactive waste. In this context, to demonstrate the feasibility of such a disposal, it is necessary to evaluate the long-term performances and the behaviour of the materials engaged in the elaboration of engineered barrier systems (EBS) and waste package elements. The studies mentioned and synthesized in this HDR thesis focused mainly on the convective transport of gas (under pressure gradient) in cementitious matrices, by coupling microstructure aspect (porosity/pores sizes distribution) and hydric environment (water saturation). Works on physico-chemical durability allowed the description of the chemical degradation of cement-based materials in extreme conditions using ammonium nitrate, to increase the materials damaging processes in order to identify functional margins. In relationship with the interim storage management phase, studies related to the behaviour and characterization of concrete submitted to high temperatures (up to 400 C) were also described. Finally, results concerning the gas (H 2 ) overpressure resistance of engineered barriers made of compacted clays were summarized. (author)

  14. Fission gas release at high burn-up: beyond the standard diffusion model

    International Nuclear Information System (INIS)

    Landskron, H.; Sontheimer, F.; Billaux, M.R.

    2002-01-01

    At high burn-up standard diffusion models describing the release of fission gases from nuclear fuel must be extended to describe the experimental loss of xenon observed in the fuel matrix of the rim zone. Marked improvements of the prediction of integral fission gas release of fuel rods as well as of radial fission gas profiles in fuel pellets are achieved by using a saturation concept to describe fission gas behaviour not only in the pellet rim but also as an additional fission gas path in the whole pellet. (author)

  15. The VERDI fission fragment spectrometer

    Directory of Open Access Journals (Sweden)

    Frégeau M.O.

    2013-12-01

    Full Text Available The VERDI time-of-flight spectrometer is dedicated to measurements of fission product yields and of prompt neutron emission data. Pre-neutron fission-fragment masses will be determined by the double time-of-flight (TOF technique. For this purpose an excellent time resolution is required. The time of flight of the fragments will be measured by electrostatic mirrors located near the target and the time signal coming from silicon detectors located at 50 cm on both sides of the target. This configuration, where the stop detector will provide us simultaneously with the kinetic energy of the fragment and timing information, significantly limits energy straggling in comparison to legacy experimental setup where a thin foil was usually used as a stop detector. In order to improve timing resolution, neutron transmutation doped silicon will be used. The high resistivity homogeneity of this material should significantly improve resolution in comparison to standard silicon detectors. Post-neutron fission fragment masses are obtained form the time-of-flight and the energy signal in the silicon detector. As an intermediary step a diamond detector will also be used as start detector located very close to the target. Previous tests have shown that poly-crystalline chemical vapour deposition (pCVD diamonds provides a coincidence time resolution of 150 ps not allowing complete separation between very low-energy fission fragments, alpha particles and noise. New results from using artificial single-crystal diamonds (sCVD show similar time resolution as from pCVD diamonds but also sufficiently good energy resolution.

  16. The shock behaviour of a SiO2-Li2O transparent glass-ceramic armour material

    International Nuclear Information System (INIS)

    Pickup, I.M.; Millett, J.C.F.; Bourne, N.K.

    2004-01-01

    The dynamic behaviour of a transparent glass-ceramic material, Transarm, developed by Alstom UK for the UK MoD has been studied. Plate impact experiments have been used to measure the materials Hugoniot characteristics and failure behaviour. Longitudinal stresses have been measured using embedded and back surface mounted Manganin gauges. Above a threshold stress of ca. 4 GPa, the longitudinal stress histories exhibit a significant secondary rise, prior to attaining their Hugoniot stress. Lateral stresses were also measured by embedding Manganin gauges in longitudinal cuts. Significant secondary rises in stress were observed when the applied longitudinal stress exceeded the 4 GPa threshold, indicating the presence of a failure front. The dynamic shear strength of the glass has been measured using the longitudinal and lateral data. Even though significant strength drops have been measured before and behind the failure front, the material has a high post-failure strength compared to non- crystalline glasses

  17. The Shock Behaviour of a SiO2-Li2O Transparent Glass-Ceramic Armour Material

    Science.gov (United States)

    Pickup, I. M.; Millett, J. C. F.; Bourne, N. K.

    2004-07-01

    The dynamic behaviour of a transparent glass-ceramic material, Transarm, developed by Alstom UK for the UK MoD has been studied. Plate impact experiments have been used to measure the materials Hugoniot characteristics and failure behaviour. Longitudinal stresses have been measured using embedded and back surface mounted Manganin gauges. Above a threshold stress of ca. 4 GPa, the longitudinal stress histories exhibit a significant secondary rise, prior to attaining their Hugoniot stress. Lateral stresses were also measured by embedding Manganin gauges in longitudinal cuts. Significant secondary rises in stress were observed when the applied longitudinal stress exceeded the 4 GPa threshold, indicating the presence of a failure front. The dynamic shear strength of the glass has been measured using the longitudinal and lateral data. Even though significant strength drops have been measured before and behind the failure front, the material has a high post-failure strength compared to non- crystalline glasses.

  18. On Identification of Critical Material Attributes for Compression Behaviour of Pharmaceutical Diluent Powders

    Directory of Open Access Journals (Sweden)

    Jianyi Zhang

    2017-07-01

    anticipated that the expansion was induced by elastic recovery to a limited extent, while the shrinkage was primarily due to the solidification during storage. It was also found that, for all powders considered, the powder compressibility and the elastic recovery depended significantly on the particle breakage tendency: a decrease in the particle breakage tendency led to a slight decrease in the powder compressibility and a significant drop in immediate elastic recovery. This implies that the particle breakage tendency is a critical material attribute in controlling the compression behaviour of pharmaceutical powders.

  19. Behaviour of neutron moderator materials at high temperatures in CASTOR registered -casks: qualification and assessment

    International Nuclear Information System (INIS)

    Krietsch, T.; Wolff, D.; Knopp, U.; Brocke, H.D.

    2004-01-01

    The Federal Institute for Materials Research and Testing (BAM) is the responsible German authority for the assessment of mechanical and thermal designs of transport and storage casks for radioactive materials. BAM checks up the proofs of the applicants in their safety reports and assesses the conformity to the Regulations for the Safe Transport of Radioactive Material. One applicant is the Gesellschaft fuer Nuklear-Behaelter mbH (GNB) with a new generation of transport and storage casks of CASTOR registered -design. GNB typically uses ultra high molecular weight Polyethylene (UHMW-PE) for the moderation of free neutrons. Rods made of UHMW-PE are positioned in axial bore holes in the wall of the cask and plates of UHMW-PE are in free spaces between primary and secondary lid and between the bottom of the cask and an outer plate (Figure 1). Because of the heat generated by the radioactive inventory and because of a strained spring at the bottom of every bore hole, UHMW-PE is subjected to permanent thermal and mechanical loads as well as loads from gamma and neutron radiation. UHMW-PE has been used under routine- and normal conditions of transport for maximum temperatures up to 130 C. For new generations of CASTOR registered -design maximum temperatures will be increased up to 160 C. That means a permanent use of UHMW-PE at temperatures within and above the melting region of the crystallites. In this paper, some results of special investigations for the proofs of usability of UHMW-PE at temperatures up to 160 C under real conditions of transport and storage in CASTOR registered -casks are given. For that, investigations on temperature dependent expansion behaviour under laboratory conditions as well as in large scale experiments, especially in the case of multiple heating and cooling, were done. Besides, geometrical creep strength for long-term loading by temperatures and pressures with regard to the chemical and physical stability properties of UHMW-PE above the

  20. Social inequalities in self-rated health in Ukraine in 2007: the role of psychosocial, material and behavioural factors.

    Science.gov (United States)

    Platts, Loretta G; Gerry, Christopher J

    2017-04-01

    Despite Ukraine's large population, few studies have examined social inequalities in health. This study describes Ukrainian educational inequalities in self-rated health and assesses how far psychosocial, material and behavioural factors account for the education gradient in health. Data were analyzed from the 2007 wave of the Ukrainian Longitudinal Monitoring Survey. Education was categorized as: lower secondary or less, upper secondary and tertiary. In logistic regressions of 5451 complete cases, stratified by gender, declaring less than average health was regressed on education, before and after adjusting for psychosocial, material and behavioural factors. In analyses adjusted for socio-demographic characteristics, compared with those educated up to lower secondary level, tertiary education was associated with lower risk of less than average health for both men and women. Including material factors (income quintiles, housing assets, labour market status) reduced the association between education and health by 55-64% in men and 35-47% in women. Inclusion of health behaviours (physical activity, smoking, alcohol consumption and body mass index) reduced the associations by 27-30% in men and 19-27% in women; in most cases including psychosocial factors (marital status, living alone, trust in family and friends) did not reduce the size of the associations. Including all potential explanatory factors reduced the associations by 68-84% in men and 43-60% in women. The education gradient in self-rated health in Ukraine was partly accounted for by material and behavioural factors. In addition to health behaviours, policymakers should consider upstream determinants of health inequalities, such as joblessness and poverty. © The Author 2016. Published by Oxford University Press on behalf of the European Public Health Association. All rights reserved.

  1. Fission in a Plasma

    Energy Technology Data Exchange (ETDEWEB)

    Younes, W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-10-26

    A three-year theory project was undertaken to study the fission process in extreme astrophysical environments, such as the crust of neutron stars. In the first part of the project, the effect of electron screening on the fission process was explored using a microscopic approach. For the first time, these calculations were carried out to the breaking point of the nucleus. In the second part of the project, the population of the fissioning nucleus was calculated within the same microscopic framework. These types of calculations are extremely computer-intensive and have seldom been applied to heavy deformed nuclei, such as fissioning actinides. The results, tools and methodologies produced in this work will be of interest to both the basic-science and nuclear-data communities.

  2. Thermomechanical behaviour of two heterogeneous tungsten materials via 2D and 3D image-based FEM

    International Nuclear Information System (INIS)

    Zivelonghi, Alessandro

    2011-01-01

    An advanced numerical procedure based on imaging of the material microstructure (Image- Based Finite Element Method or Image-Based FEM) was extended and applied to model the thermomechanical behaviour of novel materials for fusion applications. Two tungsten based heterogeneous materials with different random morphologies have been chosen as challenging case studies: (1) a two-phase mixed ductile-brittle W/CuCr1Zr composite and (2) vacuum plasma-sprayed tungsten (VPS-W 75 vol.%), a porous coating system with complex dual-scale microstructure. Both materials are designed for the future fusion reactor DEMO: W/CuCr1Zr as main constituent of a layered functionally graded joint between plasma-facing armor and heat sink whereas VPS-W for covering the first wall of the reactor vessel in direct contact with the plasma. The primary focus of this work was to investigate the mesoscopic material behaviour and the linkage to the macroscopic response in modeling failure and heat-transfer. Particular care was taken in validating and integrating simulation findings with experimental inputs. The solution of the local thermomechanical behaviour directly on the real material microstructure enabled meaningful insights into the complex failure mechanism of both materials. For W/CuCr1Zr full macroscopic stress-strain curves including the softening and failure part could be simulated and compared with experimental ones at different temperatures, finding an overall good agreement. The comparison of simulated and experimental macroscopic behaviour of plastic deformation and rupture also showed the possibility to indirectly estimate micro- and mesoscale material parameters. Both heat conduction and elastic behaviour of VPS-W have been extensively investigated. New capabilities of the Image-Based FEM could be shown: decomposition of the heat transfer reduction as due to the individual morphological phases and back-fitting of the reduced stiffness at interlamellar boundaries. The

  3. Microscopic Theory of Fission

    International Nuclear Information System (INIS)

    Younes, W; Gogny, D

    2008-01-01

    In recent years, the microscopic method has been applied to the notoriously difficult problem of nuclear fission with unprecedented success. In this paper, we discuss some of the achievements and promise of the microscopic method, as embodied in the Hartree-Fock method using the Gogny finite-range effective interaction, and beyond-mean-field extensions to the theory. The nascent program to describe induced fission observables using this approach at the Lawrence Livermore National Laboratory is presented

  4. Fusion-fission dynamics

    International Nuclear Information System (INIS)

    Blocki, J.; Planeta, R.; Brzychczyk, J.; Grotowski, K.

    1992-01-01

    Classical dynamical calculations of the heavy ion induced fission processes have been performed for the reactions 40 Ar+ 141 Pr, 20 Ne+ 165 Ho and 12 C+ 175 Lu leading to the iridium like nucleus. As a result prescission lifetimes were obtained and compared with the experimental values. The comparison between the calculated and experimental lifetimes indicates that the one-body dissipation picture is much more relevant in describing the fusion-fission dynamics than the two-body one. (orig.)

  5. Behaviour of Danish weaner and grower pigs is affected by the type and quantity of enrichment material provided

    DEFF Research Database (Denmark)

    Hakansson, Franziska; Lund, Vibe Pedersen; Kirchner, Marlene

    Inappropriate behaviour is known to reduce the welfare of pigs and therefore, determining factors influencing the quality of pig behaviour in commercial systems is of importance. As part of a larger project, this study investigated the effect of selected management parameters on different aspects...... and w/g pigs, were performed at each farm. Additionally, space allowance (WQ), tail biting (WQ), percentage of nursing sows, breed, weaning-age, type and amount of rooting material were collected. From the single measurements, WQ-criteria scores and the corresponding principle score for ‘Appropriate...... Behaviour’ were calculated according to the latest published version of WQ. Th e relation between selected management factors and the aggregated behaviour scores was tested with the help of Pearson correlations (*/ ** = significance at 0.05/ 0.01 level). The results of this study indicate an effect...

  6. Adsorption of fission products on mediterranean mud; Adsorption des produits de fission sur des vases de mediterranee

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, P; Gailledreau, C [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    Partition coefficients of some fission products have been measured in sea water on mud taken from the bottom of the Mediterranean sea. A discussion follows on the behaviour of these radioisotopes. (author) [French] On a mesure les coefficients de partage dans l'eau de mer de quelques produits de fission a longue periode sur des echantillons de vase preleves en Mediterranee. Les valeurs trouvees sont elevees. Le comportement de ces isotopes radioactifs est discutee. (auteur)

  7. Chemical factors affecting fission product transport in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

    1984-10-01

    This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly

  8. Measurements of Short-Lived Fission Isomers

    Science.gov (United States)

    Finch, Sean; Bhike, Megha; Howell, Calvin; Krishichayan, Fnu; Tornow, Werner

    2016-09-01

    Fission yields of the short lived isomers 134mTe (T1 / 2 = 162 ns) and 136mXe (T1 / 2 = 2 . 95 μs) were measured for 235U and 238U. The isomers were detected by the γ rays associated with the decay of the isomeric states using high-purity germanium detectors. Fission was induced using both monoenergetic γ rays and neutrons. At TUNL's High-Intensity Gamma-ray Source (HI γS), γ rays of 9 and 11 MeV were produced . Monoenergetic 8 MeV neutrons were produced at TUNL's tandem accelerator laboratory. Both beams were pulsed to allow for precise time-gated spectroscopy of both prompt and delayed γ rays following fission. This technique offers a non-destructive probe of special nuclear materials that is sensitive to the isotopic identity of the fissile material.

  9. Hydro-mechanical behaviour of crushed COx argillite used as backfilling material in HLW repository

    International Nuclear Information System (INIS)

    Tang Chaosheng; Shi Bin; Cui Yujun; Anh-Minh Tang

    2010-01-01

    At present, the crushed Callovo-Oxfordian (COx) argillite powder is proposed as an alternative backfilling material in France, which will be constructed in the engineering barrier of high-level radioactive waste (HLW) repository. In this investigation, the compression behavior of two crushed COx argillite powders (coarser one and finer one) was studied by running l-D compression tests with several loading-unloading cycles. After the final dry density 2.0 g/cm 3 was reached, the specimen was flooding with distilled water and the evolution of axial stress was studied during saturation process. The effects of initial axial stress level and grain size distribution (GSD) on hydro-mechanical behaviour of compacted specimen were analyzed. The results show that the compression curves are significantly influenced by the GSD of the soils. To obtain the same degree of compaction, the axial stress applied to finer soil is much higher than that of coarser soil. In addition, the compression index of the finer soil is bigger than that of coarser soil. The swelling index at initial water content increases with the dry density and seems to be independent of the GSD. During saturation, the initial lower axial stress causes obvious swelling behavior for both the coarser and finer powder samples and the corresponding axial stress increase gradually. At initial higher axial stress condition, monotone collapse behavior is observed for the coarser powder samples. Whereas the axial stress decrease firstly, then increase and finally decrease again for the finer powder samples. After saturation, the equilibrium axial stresses of finer powder samples are higher than that of coarser powder samples. (authors)

  10. Crystal plasticity-based modeling for predicting anisotropic behaviour and formability of metallic materials

    International Nuclear Information System (INIS)

    Pham, Son; Jeong, Youngung; Creuziger, Adam; Iadicola, Mark; Foecke, Tim; Rollett, Anthony

    2016-01-01

    Metallic materials often exhibit anisotropic behaviour under complex load paths because of changes in microstructure, e.g., dislocations and crystallographic texture. In this study, we present the development of constitutive model based on dislocations, point defects and texture in order to predict anisotropic response under complex load paths. In detail, dislocation/solute atom interactions were considered to account for strain aging and static recovery. A hardening matrix based on the interaction of dislocations was built to represent the cross-hardening of different slip systems. Clear differentiation between forward and backward slip directions of dislocations was made to describe back stresses during path changes. In addition, we included dynamic recovery in order to better account for large plastic deformation. The model is validated against experimental data for AA5754-O with path changes, e.g., Figure 1 [1] Another effort is to include microstructure in forming predictions with a minimal increase in computational time. This effort enables comprehensive investigations of the influence of texture-induced anisotropy on formability [2]. Application of these improvements to predict forming limits of various BCC textures, such as γ, ρ, α, η and ϵ fibers and a random (R) texture. These simulations demonstrate that the crystallographic texture has significant (both positive and negative) effects on the forming limit diagrams (Figure 2). For example, the y fiber texture, that is often sought through thermo-mechanical processing due to high r-value, had the highest forming limit in the balanced biaxial strain path but the lowest forming limit under the plane strain path among textures under consideration. (paper)

  11. Do material, psychosocial and behavioural factors mediate the relationship between disability acquisition and mental health? A sequential causal mediation analysis.

    Science.gov (United States)

    Aitken, Zoe; Simpson, Julie Anne; Gurrin, Lyle; Bentley, Rebecca; Kavanagh, Anne Marie

    2018-01-29

    There is evidence of a causal relationship between disability acquisition and poor mental health; however, the mechanism by which disability affects mental health is poorly understood. This gap in understanding limits the development of effective interventions to improve the mental health of people with disabilities. We used four waves of data from the Household, Income and Labour Dynamics in Australia Survey (2011-14) to compare self-reported mental health between individuals who acquired any disability (n=387) and those who remained disability-free (n=7936). We tested three possible pathways from disability acquisition to mental health, examining the effect of material, psychosocial and behavioural mediators. The effect was partitioned into natural direct and indirect effects through the mediators using a sequential causal mediation analysis approach. Multiple imputation using chained equations was used to assess the impact of missing data. Disability acquisition was estimated to cause a five-point decline in mental health [estimated mean difference: -5.3, 95% confidence interval (CI) -6.8, -3.7]. The indirect effect through material factors was estimated to be a 1.7-point difference (-1.7, 95% CI -2.8, -0.6), explaining 32% of the total effect, with a negligible proportion of the effect explained by the addition of psychosocial characteristics (material and psychosocial: -1.7, 95% CI -3.0, -0.5) and a further 5% by behavioural factors (material-psychosocial-behavioural: -2.0, 95% CI -3.4, -0.6). The finding that the effect of disability acquisition on mental health operates predominantly through material rather than psychosocial and behavioural factors has important implications. The results highlight the need for better social protection, including income support, employment and education opportunities, and affordable housing for people who acquire a disability. © The Author(s) 2018; all rights reserved. Published by Oxford University Press on behalf of the

  12. Electric nets and sticky materials for analysing oviposition behaviour of gravid malaria vectors

    Directory of Open Access Journals (Sweden)

    Dugassa Sisay

    2012-11-01

    Full Text Available Abstract Background Little is known about how malaria mosquitoes locate oviposition sites in nature. Such knowledge is important to help devise monitoring and control measures that could be used to target gravid females. This study set out to develop a suite of tools that can be used to study the attraction of gravid Anopheles gambiae s.s. towards visual or olfactory cues associated with aquatic habitats. Methods Firstly, the study developed and assessed methods for using electrocuting nets to analyse the orientation of gravid females towards an aquatic habitat. Electric nets (1m high × 0.5m wide were powered by a 12V battery via a spark box. High and low energy settings were compared for mosquito electrocution and a collection device developed to retain electrocuted mosquitoes when falling to the ground. Secondly, a range of sticky materials and a detergent were tested to quantify if and where gravid females land to lay their eggs, by treating the edge of the ponds and the water surface. A randomized complete block design was used for all experiments with 200 mosquitoes released each day. Experiments were conducted in screened semi-field systems using insectary-reared An. gambiae s.s. Data were analysed by generalized estimating equations. Results An electric net operated at the highest spark box energy of a 400 volt direct current made the net spark, creating a crackling sound, a burst of light and a burning smell. This setting caught 64% less mosquitoes than a net powered by reduced voltage output that could neither be heard nor seen (odds ratio (OR 0.46; 95% confidence interval (CI 0.40-0.53, p Conclusion A square of four e-nets with yellow sticky boards as a collection device can be used for quantifying the numbers of mosquitoes approaching a small oviposition site. Shiny sticky surfaces attract gravid females possibly because they are visually mistaken as aquatic habitats. These materials might be developed further as gravid traps

  13. Detection of fission fragments by secondary emission; Detection des fragments de fission par emission secondaire

    Energy Technology Data Exchange (ETDEWEB)

    Audias, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    This fission fragment detecting apparatus is based on the principle that fragments traversing a thin foil will cause emission of secondary electrons. These electrons are then accelerated (10 kV) and directly detected by means of a plastic scintillator and associated photomultiplier. Some of the advantages of such a detector are, its rapidity, its discriminating power between alpha particles and fission fragments, its small energy loss in detecting the fragments and the relatively great amount of fissionable material which it can contain. This paper is subdivided as follows: a) theoretical considerations b) constructional details of apparatus and some experimental details and c) a study of the secondary emission effect itself. (author) [French] Le detecteur de fragments de fission que nous avons realise est base sur le principe de l'emission secondaire produite par les fragments de fission traversant une feuille mince: les electrons secondaires emis sont acceleres a des tensions telles (de l'ordre de 10 kV), qu'ils soient directement detectables par un scintillateur plastique associe a un photomultiplicateur. L'interet d'un tel detecteur reside: dans sa rapidite, sa tres bonne discrimination alpha, fission, la possibilite de detecter les fragments de fission avec une perte d'energie pouvant rester relativement faible, et la possibilite d'introduire des quantites de matiere fissile plus importantes que dans les autres types de detecteurs. Ce travail comporte: -) un apercu bibliographique de la theorie du phenomene, -) realisation et mise au point du detecteur avec etude experimentale de quelques parametres intervenant dans l'emission secondaire, -) etude de l'emission secondaire (sur la face d'emergence des fragments de fission) en fonction de l'energie du fragment et en fonction de l'epaisseur de matiere traversee avant emission secondaire, et -) une etude comparative de l'emission secondaire sur la face d'incidence et sur la face d'emergence des fragments de

  14. A cluster dynamics study of fission gases in uranium dioxide

    International Nuclear Information System (INIS)

    Skorek, Richard

    2013-01-01

    During in-pile irradiation of nuclear fuels a lot of rare gases are produced, mainly xenon and krypton. The behaviour of these highly insoluble fission gases may lead to an additional load of the cladding, which may have detrimental safety consequences. For these reasons, fission gas behaviour (diffusion and clustering) has been extensively studied for years.In this work, we present an application of Cluster Dynamics to address the behaviour of fission gases in UO_2 which simultaneously describes changes in rare gas atom and point defect concentrations in addition to the bubble size distribution. This technique, applied to Kr implanted and annealed samples, yields a precise interpretation of the release curves and helps justifying the estimation of the Kr diffusion coefficient, which is a data very difficult to obtain due to the insolubility of the gas. (author) [fr

  15. Joint EC-IAEA topical meeting on development of new structural materials for advanced fission and fusion reactor systems. PowerPoint presentations

    International Nuclear Information System (INIS)

    2009-01-01

    The key topics of the meeting are the following: Radiation damage phenomena and modelling of material properties under irradiation; On-going challenges in radiation materials science; Key material parameters and operational conditions of selected reactor designs; Microstructures and mechanical properties of nuclear structural materials; Pathways to development of new structural materials; Qualification of new structural materials; Advanced microstructure probing methods; Special emphasis is given to the application of nuclear techniques in the development and qualification of new structural materials.

  16. Who is reducing their material consumption and why? A cross-cultural analysis of dematerialisation behaviours

    OpenAIRE

    Whitmarsh, Lorraine; Capstick, Stuart; Nash, Nicholas

    2017-01-01

    The environmental and economic imperatives to dematerialise economies, or ‘do more with less’, have\\ud been established for some years. Yet, to date little is known about the personal drivers associated with\\ud dematerialising. This paper explores the prevalence and profile of those who are taking action to reduce\\ud consumption in different cultural contexts (UK and Brazil) and considers influences on dematerialisation\\ud behaviours. We find exemplar behaviours (avoiding buying new things an...

  17. A physical detail relevant to the Savic-Kasanin theory of behaviour of materials under high pressure

    International Nuclear Information System (INIS)

    Celebonovic, V.

    1982-01-01

    P. Savic and R. Kasanin have proposed a theory of behaviour of materials under high pressure (Savic, 1981). Their theory can be applied to the explanation of the internal structures of planets and stars. The author proposes, a simple method for the calculation of the internal temperatures of the terrestrial planets. All the parameters needed for the application of the method can be obtained from the SK theory. (Auth.)

  18. Volatilization and reaction of fission products in flowing steam

    International Nuclear Information System (INIS)

    Johnson, I.; Steidl, D.V.; Johnson, C.E.

    1985-01-01

    The principal risk to the public from nuclear power plants derives from the highly radioactive atoms (fission products) generated as energy is produced in the nuclear fuel. The revolatilization of fission products from reactor system surfaces due to self-heating by radioactive decay has become a complicating factor in the source-term redefinition effort. It has had a major impact on calculations of fission product distributions in accident safety analyses. The focus of this research effort was to investigate the volatilization and transport of fission products and control rod materials in a flowing gaseous steam-hydrogen mixture. Fission product and control rod materials in various combinations were studied including CsI, CsOH, TeO 2 , SrO, Ag, In, Cd and Mn. The vaporization behavior of the deposits were characterized with respect to vaporization rates, chemical species and downstream transport behavior

  19. Simulation of Fission Product Liftoff Behavior During Depressurization Transients

    International Nuclear Information System (INIS)

    Tak, Nam-il; Yoon, Churl; Lee, Sung Nam

    2016-01-01

    As one of crucial technologies for the NHDD project, the development of the GAMMA-FP code is on-going. The GAMMA-FP code is targeted for fission product transport analysis under accident conditions. A well-known experiment named COMEDIE considered two important phenomena, i.e., fission product plateout and liftoff, for fission product transport within the primary circuit of a prismatic high temperature gas cooled reactor. The accumulated fission products on the structural material via the plateout can be liftoff during a blowdown phase after a pipe break accident. Since the fission product liftoff can increase a radioactivity risk, it is important to predict the amount of fission product liftoff during depressurization accidents. In this work, a model for fission product liftoff is implemented into the GAMMA-FP code and the GAMMA-FP code with the implemented model is validated using the COMEDIE blowdown test data. The results of GAMMA-FP show that the GAMMA-FP code can reliably simulate a pressure transient during blowdown phase after a pipe break accident. In addition, a reasonable amount of fission product liftoff was predicted by the GAMMA-FP code. The maximum difference between the measured and predicted liftoff fraction was less than a factor of 10. More in-depth study is required to increase the accuracy of prediction for a fission product liftoff

  20. Fission Detection Using the Associated Particle Technique

    International Nuclear Information System (INIS)

    R.P. Keegan; J.P. Hurley; J.R. Tinsley; R. Trainham; S.C. Wilde

    2008-01-01

    A beam of tagged 14 MeV neutrons from the deuterium-tritium (DT) reaction is used to induce fission in a target composed of depleted uranium. The generator yield is 10 7 neutrons/second radiated into a 4 x 4 in. NaI detectors are used for gamma-ray detection. The fission process is known to produce multiple gamma-rays and neutrons. Triple coincidences (α-γ-γ) are measured as a function of neutron flight time up to 90 ns after fission, where the α-particle arises from the DT reaction. A sudden increase in the triple coincidence rate at the location of the material is used to localize and detect fission in the interrogated target. Comparisons are made with experiment runs where lead, tungsten, and iron were used as target materials. The triple coincidence response profile from depleted uranium is noted to be different to those observed from the other target materials. The response from interrogation targets composed of fissile material is anticipated to be even more unique than that observed from depleted uranium

  1. Corrosion behaviour of container materials for the disposal of high-level waste forms in rock salt formations

    International Nuclear Information System (INIS)

    Smailos, E.; Schwarzkopf, W.; Koester, R.

    1987-05-01

    Extensive laboratory-scale experiments to evaluate the long-term corrosion behaviour of selected materials in brines and first in situ experiments were performed. In the laboratory experiments the materials Ti 99.8-Pd, Hastelloy C4 and hot-rolled low carbon steel as well cast steel, spheroidal cast iron, Si-cast iron and the Ni-Resists type D2 and D4 were investigated. The investigated parameters were: temperature, gamma-radiation and different compositions of salt brines. (orig./PW) [de

  2. SIAM CM 09 - The SIAM method for applying cohesive models to the damage behaviour of engineering materials and structures

    International Nuclear Information System (INIS)

    Scheider, Ingo; Cornec, Alfred; Schwalbe, Karl-Heinz

    2009-01-01

    This document provides guidance on the determination of damage and fracture of ductile metallic materials and structures made thereof, based mainly on experience obtained at GKSS. The method used for the fracture prediction is the cohesive model, in which material separation is represented by interface elements and their constitutive behaviour, the so-called traction-separation law, in the framework of finite elements. Several traction-separation laws are discussed, some of which are already implemented in commercial finite element codes and therefore easy applicable. Methods are described for the determination of the cohesive parameters, using a hybrid experimental/numerical approach. (orig.)

  3. Research into material behaviour of the polymeric samples obtained after 3D-printing and subjected to compression test

    Science.gov (United States)

    Petrov, Mikhail A.; Kosatchyov, Nikolay V.; Petrov, Pavel A.

    2016-10-01

    The paper represents the results of the study concerning the investigation of the influence of the filling grade (material density) on the force characteristic during the uniaxial compression test of the cylindrical polymer probes produced by additive technology based on FDM. The authors have shown that increasing of the filling grate follows to the increase of the deformation forces. However, the dependency is not a linear function and characterized by soft-elastic model of material behaviour, which is typical for polymers partly crystallized structure.

  4. SIAM CM 09 - The SIAM method for applying cohesive models to the damage behaviour of engineering materials and structures

    Energy Technology Data Exchange (ETDEWEB)

    Scheider, Ingo; Cornec, Alfred [GKSS-Forschungszentrum Geesthacht GmbH (Germany). Inst. fuer Materialforschung; Schwalbe, Karl-Heinz

    2009-12-19

    This document provides guidance on the determination of damage and fracture of ductile metallic materials and structures made thereof, based mainly on experience obtained at GKSS. The method used for the fracture prediction is the cohesive model, in which material separation is represented by interface elements and their constitutive behaviour, the so-called traction-separation law, in the framework of finite elements. Several traction-separation laws are discussed, some of which are already implemented in commercial finite element codes and therefore easy applicable. Methods are described for the determination of the cohesive parameters, using a hybrid experimental/numerical approach. (orig.)

  5. Crystallization behaviour and thermal stability of two aluminium-based metallic glass powder materials

    Energy Technology Data Exchange (ETDEWEB)

    Li, X.P.; Yan, M. [University of Queensland, School of Mechanical and Mining Engineering, ARC Centre of Excellence for Design in Light Metals, Brisbane, QLD 4072 (Australia); Yang, B.J. [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Wang, J.Q., E-mail: jqwang@imr.ac.cn [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Schaffer, G.B. [University of Queensland, School of Mechanical and Mining Engineering, ARC Centre of Excellence for Design in Light Metals, Brisbane, QLD 4072 (Australia); Qian, M., E-mail: ma.qian@uq.edu.au [University of Queensland, School of Mechanical and Mining Engineering, ARC Centre of Excellence for Design in Light Metals, Brisbane, QLD 4072 (Australia)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer The crystallization paths and products of Al{sub 86}Ni{sub 7}Y{sub 4.5}Co{sub 1}La{sub 1.5} powder have been identified. Black-Right-Pointing-Pointer The thermal stability of Al{sub 86}Ni{sub 7}Y{sub 4.5}Co{sub 1}La{sub 1.5} powder has been assessed. Black-Right-Pointing-Pointer The Al{sub 86}Ni{sub 7}Y{sub 4.5}Co{sub 1}La{sub 1.5} powder shows a wide processing window of 75 K. Black-Right-Pointing-Pointer The powder has the potential to be consolidated into thick BMG components based on the findings. Black-Right-Pointing-Pointer The Al{sub 85}Ni{sub 5}Y{sub 6}Co{sub 2}Fe{sub 2} powder shows similar characteristics but inferior thermal stability. - Abstract: The crystallization behaviour and thermal stability of two Al-based metallic glass powder materials, Al{sub 85}Ni{sub 5}Y{sub 6}Co{sub 2}Fe{sub 2} and Al{sub 86}Ni{sub 6}Y{sub 4.5}Co{sub 2}La{sub 1.5}, have been investigated using differential scanning calorimetry (DSC), X-ray diffraction (XRD) and electron microscopy. Both alloy powders show a distinct three-stage crystallization process with a similar gap of {approx}75 K between the onset crystallization temperature (T{sub x}) and the second crystallization temperature. Crystallization occurs by the precipitation and growth of fcc-Al, without intermetallic formation. The apparent activation energy for each stage of crystallization was determined from DSC analyses and the phases resulting from each crystallization stage were identified by XRD and electron microscopy. The critical cooling rate for each alloy powder was calculated from the DSC data. These results are necessary to inform the consolidation of amorphous powder particles of Al{sub 85}Ni{sub 5}Y{sub 6}Co{sub 2}Fe{sub 2} or Al{sub 86}Ni{sub 6}Y{sub 4.5}Co{sub 2}La{sub 1.5} into thick (>1 mm) metallic glass components.

  6. Methods to Collect, Compile, and Analyze Observed Short-lived Fission Product Gamma Data

    Energy Technology Data Exchange (ETDEWEB)

    Finn, Erin C.; Metz, Lori A.; Payne, Rosara F.; Friese, Judah I.; Greenwood, Lawrence R.; Kephart, Jeremy D.; Pierson, Bruce D.; Ellis, Tere A.

    2011-09-29

    A unique set of fission product gamma spectra was collected at short times (4 minutes to 1 week) on various fissionable materials. Gamma spectra were collected from the neutron-induced fission of uranium, neptunium, and plutonium isotopes at thermal, epithermal, fission spectrum, and 14-MeV neutron energies. This report describes the experimental methods used to produce and collect the gamma data, defines the experimental parameters for each method, and demonstrates the consistency of the measurements.

  7. Nuclear fission and neutron-induced fission cross-sections

    Energy Technology Data Exchange (ETDEWEB)

    James, G.D.; Lynn, J.E.; Michaudon, A.; Rowlands, J.; de Saussure, G.

    1981-01-01

    A general presentation of current knowledge of the fission process is given with emphasis on the low energy fission of actinide nuclei and neutron induced fission. The need for and the required accuracy of fission cross section data in nuclear energy programs are discussed. A summary is given of the steps involved in fission cross section measurement and the range of available techniques. Methods of fission detection are described with emphasis on energy dependent changed and detector efficiency. Examples of cross section measurements are given and data reduction is discussed. The calculation of fission cross sections is discussed and relevant nuclear theory including the formation and decay of compound nuclei and energy level density is introduced. A description of a practical computation of fission cross sections is given.

  8. Chemistry of actinides and fission products

    International Nuclear Information System (INIS)

    Pruett, D.J.; Sherrow, S.A.; Toth, L.M.

    1988-01-01

    This task is concerned primarily with the fundamental chemistry of the actinide and fission product elements. Special efforts are made to develop research programs in collaboration with researchers at universities and in industry who have need of national laboratory facilities. Specific areas currently under investigation include: (1) spectroscopy and photochemistry of actinides in low-temperature matrices; (2) small-angle scattering studies of hydrous actinide and fission product polymers in aqueous and nonaqueous solvents; (3) kinetic and thermodynamic studies of complexation reactions in aqueous and nonaqueous solutions; and (4) the development of inorganic ion exchange materials for actinide and lanthanide separations. Recent results from work in these areas are summarized here

  9. The evaluation for reference fission yield of 238U fission

    International Nuclear Information System (INIS)

    Liang Qichang; Liu Tingjin

    1998-01-01

    In the fission yield data evaluation and measurement, the reference yield is very important, good or poor recommended or measurement values depend upon the reference data to a great extent. According to the CRP's requirement, the evaluation of reference fission yields have been and will be carried out in CNDC, as a part of the whole work (contract No.9504/R 0 /Regular Budget Fund), the evaluation for 29 reference fission yields of 15 product nuclides from 238 U fission have been completed

  10. Fission fragment angular momentum

    International Nuclear Information System (INIS)

    Frenne, D. De

    1991-01-01

    Most of the energy released in fission is converted into translational kinetic energy of the fragments. The remaining excitation energy will be distributed among neutrons and gammas. An important parameter characterizing the scission configuration is the primary angular momentum of the nascent fragments. Neutron emission is not expected to decrease the spin of the fragments by more than one unit of angular momentum and is as such of less importance in the determination of the initial fragment spins. Gamma emission is a suitable tool in studying initial fragment spins because the emission time, number, energy, and multipolarity of the gammas strongly depend on the value of the primary angular momentum. The main conclusions of experiments on gamma emission were that the initial angular momentum of the fragments is large compared to the ground state spin and oriented perpendicular to the fission axis. Most of the recent information concerning initial fragment spin distributions comes from the measurement of isomeric ratios for isomeric pairs produced in fission. Although in nearly every mass chain isomers are known, only a small number are suitable for initial fission fragment spin studies. Yield and half-life considerations strongly limit the number of candidates. This has the advantage that the behavior of a specific isomeric pair can be investigated for a number of fissioning systems at different excitation energies of the fragments and fissioning nuclei. Because most of the recent information on primary angular momenta comes from measurements of isomeric ratios, the global deexcitation process of the fragments and the calculation of the initial fragment spin distribution from measured isomeric ratios are discussed here. The most important results on primary angular momentum determinations are reviewed and some theoretical approaches are given. 45 refs., 7 figs., 2 tabs

  11. Lunar surface fission power supplies: Radiation issues

    International Nuclear Information System (INIS)

    Houts, M.G.; Lee, S.K.

    1994-01-01

    A lunar space fission power supply shield that uses a combination of lunar regolith and materials brought from earth may be optimal for early lunar outposts and bases. This type of shield can be designed such that the fission power supply does not have to be moved from its landing configuration, minimizing handling and required equipment on the lunar surface. Mechanisms for removing heat from the lunar regolith are built into the shield, and can be tested on earth. Regolith activation is greatly reduced compared with a shield that uses only regolith, and it is possible to keep the thermal conditions of the fission power supply close to these seen in free space. For a well designed shield, the additional mass required to be brought fro earth should be less than 1000 kg. Detailed radiation transport calculations confirm the feasibility of such a shield

  12. Lunar surface fission power supplies: Radiation issues

    International Nuclear Information System (INIS)

    Houts, M.G.; Lee, S.K.

    1994-01-01

    A lunar space fission power supply shield that uses a combination of lunar regolith and materials brought from earth may be optimal for early lunar outposts and bases. This type of shield can be designed such that the fission power supply does not have to be moved from its landing configuration, minimizing handling and required equipment on the lunar surface. Mechanisms for removing heat from the lunar regolith are built into the shield, and can be tested on earth. Regolith activation is greatly reduced compared with a shield that uses only regolith, and it is possible to keep the thermal conditions of the fission power supply close to those seen in free space. For a well designed shield, the additional mass required to be brought from earth should be less than 1,000 kg. Detailed radiation transport calculations confirm the feasibility of such a shield

  13. Laser Shock Processing of Metallic Materials: Coupling of Laser-Plasma Interaction and Material Behaviour Models for the Assessment of Key Process Issues

    International Nuclear Information System (INIS)

    Ocana, J. L.; Morales, M.; Molpeceres, C.; Porro, J. A.

    2010-01-01

    Profiting by the increasing availability of laser sources delivering intensities above 109 W/cm 2 with pulse energies in the range of several Joules and pulse widths in the range of nanoseconds, laser shock processing (LSP) is consolidating as an effective technology for the improvement of surface mechanical and corrosion resistance properties of metals. The main advantage of the laser shock processing technique consists on its capability of inducing a relatively deep compression residual stresses field into metallic alloy pieces allowing an improved mechanical behaviour, explicitly, the life improvement of the treated specimens against wear, crack growth and stress corrosion cracking. Although significant work from the experimental side has been contributed to explore the optimum conditions of application of the treatments and to assess their ultimate capability to provide enhanced mechanical behaviour to work-pieces of typical materials, only limited attempts have been developed in the way of full comprehension and predictive assessment of the characteristic physical processes and material transformations with a specific consideration of real material properties. In the present paper, a review on the physical issues dominating the development of LSP processes from a high intensity laser-matter interaction point of view is presented along with the theoretical and computational methods developed by the authors for their predictive assessment and practical results at laboratory scale on the application of the technique to different materials.

  14. Characterising the thermoforming behaviour of glass fibre textile reinforced thermoplastic composite materials

    Science.gov (United States)

    Kuhtz, M.; Maron, B.; Hornig, A.; Müller, M.; Langkamp, A.; Gude, M.

    2018-05-01

    Textile reinforced thermoplastic composites are predestined for highly automated medium- and high-volume production processes. The presented work focusses on experimental studies of different types of glass fibre reinforced polypropylene (GF-PP) semi-finished thermoplastic textiles to characterise the forming behaviour. The main deformation modes fabric shear, tension, thought-thickness compression and bending are investigated with special emphasis on the impact of the textile structure, the deformation temperature and rate dependency. The understanding of the fundamental forming behaviour is required to allow FEM based assessment and improvement of thermoforming process chains.

  15. Fission product detection

    International Nuclear Information System (INIS)

    Liatard, E.; Akrouf, S.; Bruandet, J.F

    1987-01-01

    The response of photovoltaic cells to heavy ions and fission products have been tested on beam. Their main advantages are their extremely low price, their low sensitivity to energetic light ions with respect to fission products, and the possibility to cut and fit them together to any shape without dead zone. The time output signals of a charge sensitive preamplifier connected to these cells allows fast coincidences. A resolution of 12ns (F.W.H.M.) have been measured between two cells [fr

  16. Low energy nuclear fission

    International Nuclear Information System (INIS)

    Nifenecker, H.

    1982-02-01

    In these lectures we present the liquid drop model of fission and compare some of its prediction with experiment. The liquid drop analogy allows to define in a rather simple and intuitive way a number of useful concepts and possible observables. We then discuss, using the example of the oscillator model, the generality of shell effects. We show how a synthesis of the liquid drop model and of the shell model can be made using the Strutinsky shell averaging procedure. Some experimental data related to the existence of shape isomers are presented and discussed. We conclude by discussing some aspects, both experimental and theoretical, of fission dynamics

  17. Fission of heavy hypernuclei

    International Nuclear Information System (INIS)

    Nifenecker, H.

    1993-01-01

    The results on delayed and prompt fission of heavy hypernuclei obtained by the LEAR PS177 collaboration are recalled and discussed. It is shown that the hypernuclei life-times can be explained in term of a weak strangeness violating lambda-nucleon interaction with a cross section close to 6.0 10 -15 barns. The lambda attachment function is shown to be sensitive to the scission configuration, just before fission, and to the neck dynamics. This function provides a new way to study the nuclear scission process. (author)

  18. Fission gas measuring technology

    International Nuclear Information System (INIS)

    Lee, Hyung Kwon; Kim, Eun Ka; Hwang, Yong Hwa; Lee, Eun Pyo; Chun, Yong Bum; Seo, Ki Seog; Park, Dea Gyu; Chu, Yong Sun; Ahn, Sang Bok.

    1998-02-01

    Safety and economy of nuclear plant are greatly affected by the integrity of nuclear fuels during irradiation reactor core. A series of post-irradiation examination (PIE) including non-destructive and destructive test is to be conducted to evaluate and characterize the nuclear performance. In this report, a principle of the examination equipment to measure and analyse fission gases existing nuclear fuels were described and features of the component and device consisting the fission gas measuring equipment are investigated. (author). 4 refs., 2 tabs., 6 figs

  19. Fission gas measuring technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyung Kwon; Kim, Eun Ka; Hwang, Yong Hwa; Lee, Eun Pyo; Chun, Yong Bum; Seo, Ki Seog; Park, Dea Gyu; Chu, Yong Sun; Ahn, Sang Bok

    1998-02-01

    Safety and economy of nuclear plant are greatly affected by the integrity of nuclear fuels during irradiation reactor core. A series of post-irradiation examination (PIE) including non-destructive and destructive test is to be conducted to evaluate and characterize the nuclear performance. In this report, a principle of the examination equipment to measure and analyse fission gases existing nuclear fuels were described and features of the component and device consisting the fission gas measuring equipment are investigated. (author). 4 refs., 2 tabs., 6 figs.

  20. Fission modelling with FIFRELIN

    International Nuclear Information System (INIS)

    Litaize, Olivier; Serot, Olivier; Berge, Leonie

    2015-01-01

    The nuclear fission process gives rise to the formation of fission fragments and emission of particles (n,γ, e - ). The particle emission from fragments can be prompt and delayed. We present here the methods used in the FIFRELIN code, which simulates the prompt component of the de-excitation process. The methods are based on phenomenological models associated with macroscopic and/or microscopic ingredients. Input data can be provided by experiment as well as by theory. The fission fragment de-excitation can be performed within Weisskopf (uncoupled neutron and gamma emission) or a Hauser-Feshbach (coupled neutron/gamma emission) statistical theory. We usually consider five free parameters that cannot be provided by theory or experiments in order to describe the initial distributions required by the code. In a first step this set of parameters is chosen to reproduce a very limited set of target observables. In a second step we can increase the statistics to predict all other fission observables such as prompt neutron, gamma and conversion electron spectra but also their distributions as a function of any kind of parameters such as, for instance, the neutron, gamma and electron number distributions, the average prompt neutron multiplicity as a function of fission fragment mass, charge or kinetic energy, and so on. Several results related to different fissioning systems are presented in this work. The goal in the next decade will be i) to replace some macroscopic ingredients or phenomenological models by microscopic calculations when available and reliable, ii) to be a support for experimentalists in the design of detection systems or in the prediction of necessary beam time or count rates with associated statistics when measuring fragments and emitted particle in coincidence iii) extend the model to be able to run a calculation when no experimental input data are available, iv) account for multiple chance fission and gamma emission before fission, v) account for the

  1. Low energy nuclear fission

    International Nuclear Information System (INIS)

    Nifenecker, H.

    1980-08-01

    In these lectures the liquid drop model of fission is presented and some of its predictions compared with experiment. The liquid drop analogy allows to define in a rather simple and intuitive way a number of useful concepts and possible observables. It is shown how a synthesis of the liquid drop model and of the shell model can be made using the Strutinsky shell averaging procedure. Some experimental data related to the existence of shape isomers are presented and discussed. We conclude by discussing some aspects, both experimental and theoretical, of fission dynamics

  2. Behavior of solid fission products in irradiated fuel

    International Nuclear Information System (INIS)

    Song, Ung Sup; Jung, Yang Hong; Kim, Hee Moon; Yoo, Byun Gok; Kim, Do Sik; Choo, Yong Sun; Hong, Kwon Pyo

    2004-01-01

    Many fission products are generated by fission events in UO 2 fuel under irradiation in nuclear reactor. Concentration of each fission product is changed by conditions of neutron energy spectrum, fissile material, critical thermal power, irradiation period and cooling time. Volatile materials such as Cs and I, the fission products, degrade nuclear fuel rod by the decrease of thermal conductivity in pellet and the stress corrosion cracking in cladding. Metal fission products (white inclusion) make pellet be swelled and decrease volume of pellet by densification. It seems that metal fission products are filled in the pore in pellet and placed between UO 2 lattices as interstitial. In addition, metal oxide state may change structural lattice volume. Considering behavior of fission products mentioned above, concentration of them is important. Fission products could be classified as bellows; solid solution in matrix : Sr, Zr, Nb, Y, La, Ce, Pr, Nd, Pm, Sm - metal precipitates : Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sb, Te - oxide precipitates : Ba, Zr, Nb, Mo, (Rb, Cs, Te) - volatile and gases : Kr, Xe, Br, I, (Rb, Cs, Te)

  3. Fission Product Library and Resource

    Energy Technology Data Exchange (ETDEWEB)

    Burke, J. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Padgett, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-09-29

    Fission product yields can be extracted from an irradiated sample by performing gamma ray spectroscopy on the whole sample post irradiation. There are several pitfalls to avoid when trying to determine a specific isotope's fission product yield.

  4. Spontaneous fission of superheavy nuclei

    Indian Academy of Sciences (India)

    The fission-like configurations are used for the total deformation energy calculations. A ... oscillator potential for the two fission fragment regions reads as ... Beyond this limit, the contribution of more remote levels is negligible. Once the density ...

  5. Mirror fusion--fission hybrids

    International Nuclear Information System (INIS)

    Lee, J.D.

    1978-01-01

    The fusion-fission concept and the mirror fusion-fission hybrid program are outlined. Magnetic mirror fusion drivers and blankets for hybrid reactors are discussed. Results of system analyses are presented and a reference design is described

  6. Separation Of Uranium From Fission Products Zr And Ru With 30% TBP (Tri Butyl Phosphate) Dodecane In Nitric Acid Medium As An Extract Material

    International Nuclear Information System (INIS)

    Herdady, R. Didiek; Masduki, Busron; Sigit

    2000-01-01

    Separation of uranium from fission products Zr and Ru in batch process with Tbp 30% - dodecane in nitric acid medium has been investigated. The extraction was carried out on various acidity of 1,006 M, 1.990 M, 2,980 M, 4,006 M, and 5,006 M, and uranium concentration in feed of 100.30 g/l; 149.96 g/l, 250.30 g/l and 300.7 g/l. The results showed that equilibrium of extraction was achieved at 25 minutes, enhancement factor of ruthenium increased and of zirconium decreased Utilization of grand concentration of uranium in feed caused decreasing of distribution coefficient, zirconium and ruthenium. The better contribution of experiments was obtained at the acidity of 2 M and uranium concentration in feed of 149.9 g/l with the decontamination factor of zirconium, FD zr-u was 1,65 and of ruthenium, FD ru-u was 1,52

  7. Modelling of the physical behaviour of water saturated clay barriers. Laboratory tests, material models and finite element application

    International Nuclear Information System (INIS)

    Boergesson, L.; Johannesson, L.E.; Sanden, T.; Hernelind, J.

    1995-09-01

    This report deals with laboratory testing and modelling of the thermo-hydro-mechanical (THM) properties of water saturated bentonite based buffer materials. A number of different laboratory tests have been performed and the results are accounted for. These test results have lead to a tentative material model, consisting of several sub-models, which is described in the report. The tentative model has partly been adapted to the material models available in the finite element code ABAQUS and partly been implemented and incorporated in the code. The model that can be used for ABAQUS calculations agrees with the tentative model with a few exceptions. The model has been used in a number of verification calculations, simulating different laboratory tests, and the results have been compared with actual measurements. These calculations show that the model generally can be used for THM calculations of the behaviour of water saturated buffer materials, but also that there is still a lack of some understanding. It is concluded that the available model is relevant for the required predictions of the THM behaviour but that a further improvement of the model is desirable

  8. Process for treating fission waste

    International Nuclear Information System (INIS)

    Rohrmann, C.A.; Wick, O.J.

    1983-01-01

    A method is described for the treatment of fission waste. A glass forming agent, a metal oxide, and a reducing agent are mixed with the fission waste and the mixture is heated. After melting, the mixture separates into a glass phase and a metal phase. The glass phase may be used to safely store the fission waste, while the metal phase contains noble metals recovered from the fission waste

  9. Improved Fission Neutron Data Base for Active Interrogation of Actinides

    Energy Technology Data Exchange (ETDEWEB)

    Pozzi, Sara; Czirr, J. Bart; Haight, Robert; Kovash, Michael; Tsvetkov, Pavel

    2013-11-06

    This project will develop an innovative neutron detection system for active interrogation measurements. Many active interrogation methods to detect fissionable material are based on the detection of neutrons from fission induced by fast neutrons or high-energy gamma rays. The energy spectrum of the fission neutrons provides data to identify the fissionable isotopes and materials such as shielding between the fissionable material and the detector. The proposed path for the project is as follows. First, the team will develop new neutron detection systems and algorithms by Monte Carlo simulations and bench-top experiments. Next, They will characterize and calibrate detection systems both with monoenergetic and white neutron sources. Finally, high-fidelity measurements of neutron emission from fissions induced by fast neutrons will be performed. Several existing fission chambers containing U-235, Pu-239, U-238, or Th-232 will be used to measure the neutron-induced fission neutron emission spectra. The challenge for making confident measurements is the detection of neutrons in the energy ranges of 0.01 – 1 MeV and above 8 MeV, regions where the basic data on the neutron energy spectrum emitted from fission is least well known. In addition, improvements in the specificity of neutron detectors are required throughout the complete energy range: they must be able to clearly distinguish neutrons from other radiations, in particular gamma rays and cosmic rays. The team believes that all of these challenges can be addressed successfully with emerging technologies under development by this collaboration. In particular, the collaboration will address the area of fission neutron emission spectra for isotopes of interest in the advanced fuel cycle initiative (AFCI).

  10. Fission product release from fuel of water-cooled reactors

    International Nuclear Information System (INIS)

    Strupczewski, A.; Marks, P.; Klisinska, M.

    1997-01-01

    The report contains a review of theoretical models and experimental works of gaseous and volatile fission products from uranium dioxide fuel. The experimental results of activity release at low burnup and the model of fission gas behaviour at initial stage of fuel operational cycle are presented. Empirical models as well as measured results of transient fission products release rate in the temperature up to UO 2 melting point, with consideration of their chemical reactions with fuel and cladding, are collected. The theoretical and experimental data were used for calculations of gaseous and volatile fission products release, especially iodine and caesium, to the gas volume of WWER-1000 and WWER-440 type fuel rods at low and high burnup and their further release from defected rods at the assumed loss-of-coolant accident. (author)

  11. 50 years of nuclear fission

    International Nuclear Information System (INIS)

    Hilscher, D.

    1989-01-01

    The article tells the story of the discovery of nuclear fission in Berlin 50 years ago by Otto Hahn and Fritz Strassmann in cooperation with Lise Meitner. 50 years later nuclear fission is still a subject of research. Some question remain unanswered. Selected new research results are used to discuss the dynamics of the collective movement of the elementary nuclear fission process. (orig.) [de

  12. Fission dynamics of hot nuclei

    Indian Academy of Sciences (India)

    2014-04-05

    Apr 5, 2014 ... across the fission barrier is very small or in other words, the fission barrier is much ... of this shape evolution, the gross features of the fissioning nucleus can be described ..... [7] Y Abe, C Gregoire and H Delagrange, J. Phys.

  13. Status of fission yield measurements

    International Nuclear Information System (INIS)

    Maeck, W.J.

    1979-01-01

    Fission yield measurement and yield compilation activities in the major laboratories of the world are reviewed. In addition to a general review of the effort of each laboratory, a brief summary of yield measurement activities by fissioning nuclide is presented. A new fast reactor fission yield measurement program being conducted in the US is described

  14. The discovery of fission

    International Nuclear Information System (INIS)

    McKay, H.A.C.

    1978-01-01

    In this article by the retired head of the Separation Processes Group of the Chemistry Division, Atomic Energy Research Establishment, Harwell, U.K., the author recalls what he terms 'an exciting drama, the unravelling of the nature of the atomic nucleus' in the years before the Second World War, including the discovery of fission. 12 references. (author)

  15. Thermodynamics of nuclear materials

    International Nuclear Information System (INIS)

    1979-01-01

    Full text: The science of chemical thermodynamics has substantially contributed to the understanding of the many problems encountered in nuclear and reactor technology. These problems include reaction of materials with their surroundings and chemical and physical changes of fuels. Modern reactor technology, by its very nature, has offered new fields of investigations for the scientists and engineers concerned with the design of nuclear fuel elements. Moreover, thermodynamics has been vital in predicting the behaviour of new materials for fission as well as fusion reactors. In this regard, the Symposium was organized to provide a mechanism for review and discussion of recent thermodynamic investigations of nuclear materials. The Symposium was held in the Juelich Nuclear Research Centre, at the invitation of the Government of the Federal Republic of Germany. The International Atomic Energy Agency has given much attention to the thermodynamics of nuclear materials, as is evidenced by its sponsorship of four international symposia in 1962, 1965, 1967, and 1974. The first three meetings were primarily concerned with the fundamental thermodynamics of nuclear materials; as with the 1974 meeting, this last Symposium was primarily aimed at the thermodynamic behaviour of nuclear materials in actual practice, i.e., applied thermodynamics. Many advances have been made since the 1974 meeting, both in fundamental and applied thermodynamics of nuclear materials, and this meeting provided opportunities for an exchange of new information on this topic. The Symposium dealt in part with the thermodynamic analysis of nuclear materials under conditions of high temperatures and a severe radiation environment. Several sessions were devoted to the thermodynamic studies of nuclear fuels and fission and fusion reactor materials under adverse conditions. These papers and ensuing discussions provided a better understanding of the chemical behaviour of fuels and materials under these

  16. A simple treatment of fission gas for normal and accident conditions

    International Nuclear Information System (INIS)

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

    1980-01-01

    A set of simple modules have been developed to describe fission gas release and swelling in oxide nuclear fuels for use in fuel behaviour codes. The methods used are simplifications of earlier more detailed work and contain several important developments that allow for improved accuracy over earlier simple treatments and the description of the fission gas bubble population with little penalty in computer time or storage. The three modules are: (i) intragranular fission gas behaviour during normal operation, which treats gas bubble nucleation, growth and destruction by fission fragments and the diffusion of gas to the grain boundaries by single gas atom diffusion, (ii) intragranular fission gas behaviour during rapid transients which treats the migration and coalescence of gas bubbles, the sweeping up of fission gas atoms by bubbles and the drift of gas bubbles to the grain boundary under the driving force of the temperature gradient, and (iii) intergranular fission gas behaviour, which treats the growth and interaction of face and edge bubbles on the grain boundary, their interlinkage and gas release. All these models allow for transient behaviour and are compared with experimental observations of both macroscopic swelling and gas release (and retention) and microscopic observations of bubble sizes and concentrations. (author)

  17. Elastocapillary Instability in Mitochondrial Fission

    Science.gov (United States)

    Gonzalez-Rodriguez, David; Sart, Sébastien; Babataheri, Avin; Tareste, David; Barakat, Abdul I.; Clanet, Christophe; Husson, Julien

    2015-08-01

    Mitochondria are dynamic cell organelles that constantly undergo fission and fusion events. These dynamical processes, which tightly regulate mitochondrial morphology, are essential for cell physiology. Here we propose an elastocapillary mechanical instability as a mechanism for mitochondrial fission. We experimentally induce mitochondrial fission by rupturing the cell's plasma membrane. We present a stability analysis that successfully explains the observed fission wavelength and the role of mitochondrial morphology in the occurrence of fission events. Our results show that the laws of fluid mechanics can describe mitochondrial morphology and dynamics.

  18. A threshold for dissipative fission

    International Nuclear Information System (INIS)

    Thoennessen, M.; Bertsch, G.F.

    1993-01-01

    The empirical domain of validity of statistical theory is examined as applied to fission data on pre-fission data on pre-fission neutron, charged particle, and γ-ray multiplicities. Systematics are found of the threshold excitation energy for the appearance of nonstatistical fission. From the data on systems with not too high fissility, the relevant phenomenological parameter is the ratio of the threshold temperature T thresh to the (temperature-dependent) fission barrier height E Bar (T). The statistical model reproduces the data for T thresh /E Bar (T) thresh /E Bar (T) independent of mass and fissility of the systems

  19. Performance and first results of fission product release and transport provided by the VERDON facility

    Energy Technology Data Exchange (ETDEWEB)

    Gallais-During, A., E-mail: annelise.gallais-during@cea.fr [CEA, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Bonnin, J.; Malgouyres, P.-P. [CEA, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Morin, S. [IRSN, F-13108 Saint-Paul-lez-Durance (France); Bernard, S.; Gleizes, B.; Pontillon, Y.; Hanus, E.; Ducros, G. [CEA, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-01

    Highlights: • A new facility to perform experimental LWR severe accidents sequences is evaluated. • In the furnace a fuel sample is heated up to 2600 °C under a controlled gas atmosphere. • Innovative thermal gradient tubes are used to study fission product transport. • The new VERDON facility shows an excellent consistency with results from VERCORS. • Fission product re-vapourization results confirm the correct functioning of the gradient tubes. - Abstract: One of the most important areas of research concerning a hypothetical severe accident in a light water reactor (LWR) is determining the source term, i.e. quantifying the nature, release kinetics and global released fraction of the fission products (FPs) and other radioactive materials. In line with the former VERCORS programme to improve source term estimates, the new VERDON laboratory has recently been implemented at the CEA Cadarache Centre in the LECA-STAR facility. The present paper deals with the evaluation of the experimental equipment of this new VERDON laboratory (furnace, release and transport loops) and demonstrates its capability to perform experimental sequences representative of LWR severe accidents and to supply the databases necessary for source term assessments and FP behaviour modelling.

  20. Fission and r-process nucleosynthesis in neutron star mergers

    International Nuclear Information System (INIS)

    Giuliani, Samuel Andrea

    2018-01-01

    rates are used in r-process calculations for matter dynamically ejected in neutron star mergers and we compare our results with those obtained from a more conventional set of reaction rates. We find that all the models predict the onset of fission above the shell closure N=184 and Z=100 due to the sudden decrease in fission barriers. However, the amount of material accumulated at N=184 turns out to be very sensitive to the height of the fission barriers and the shell gap. Finally, we have also explored the impact of recent advances in fission calculations on the theoretical estimation of spontaneous fission lifetimes. We find that performing dynamical approaches based on the minimization of the integral action with nontraditional collective degrees of freedom has a strong impact in the fission barriers and the spontaneous fission lifetimes. The possible consequences of this new approach for the calculation of neutron induced fission rates has to be addressed.

  1. Numerical experiments in finite element analysis of thermoelastoplastic behaviour of materials. Further developments of the PLASTEF code

    International Nuclear Information System (INIS)

    Basombrio, F.G.; Sarmiento, G.S.

    1980-01-01

    In a previous paper the finite element code PLASTEF for the numerical simulation of thermoelastoplastic behaviour of materials was presented in its general outline. This code employs an initial stress incremental procedure for given histories of loads and temperature. It has been formulated for medium sized computers. The present work is an extension of the previous paper to consider additional aspects of the variable temperature case. Non-trivial tests of this type of situation are described. Finally, details are given of some concrete applications to the prediction of thermoelastoplastic collapse of nuclear fuel element cladding. (author)

  2. Problems to be solved about inelastic behaviour of materials and inelastic analysis of structures at elevated temperature

    International Nuclear Information System (INIS)

    Ledermann, P.; Escatha, Y. d'.

    1981-01-01

    At elevated temperature, ASME CODE CASE N 47 demands, in its design and analysis part to demonstrate that none of eight damages, related to the monotonic and cyclic inelastic behaviour of the material and structure, will happen during the whole life of the reactor. However this demonstration, for strain limits and creep fatigue failure, using a purely elastic analysis as in the ASME CODE Section III, is usually impossible. Inelastic analysis is then necessary. We review some of the research work (theorical and experimental) which is being done to qualify methods for an inelastic analysis of structures at elevated temperature [fr

  3. Fusion-fission hybrid studies in the United States

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Berwald, D.H.; Cheng, E.T.; Delene, J.G.; Jassby, D.L.

    1986-01-01

    Systems and conceptual design studies have been carried out on the following three hybrid types: (1) The fission-suppressed hybrid, which maximizes fissile material produced (Pu or 233 U) per unit of total nuclear power by suppressing the fission process and multiplying neutrons by (n,2n) reactions in materials like beryllium. (2) The fast-fission hybrid, which maximizes fissile material produced per unit of fusion power by maximizing fission of 238 U (Pu is produced) in which twice the fissile atoms per unit of fusion power (but only a third per unit of nuclear power) are made. (3) The power hybrid, which amplifies power in the blanket for power production but does not produce fuel to sell. All three types must sell electrical power to be economical

  4. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    International Nuclear Information System (INIS)

    1985-01-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity

  5. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity.

  6. Barium 139 as Fission Indicator

    Energy Technology Data Exchange (ETDEWEB)

    Broda, E.

    1943-07-01

    This report is based on a measurement performed at the Cavendish Laboratory (Cambridge) by E. Broda in December 1943 where a technique has been worked out for measuring the fission density in a uranium containing medium in relative units by determining the amount of a suitable fission product formed. Generally a given fission product will be formed in natural uranium by slow neutron fission of U235 or by fast neutron fission of either U235 or U238. It is intended to translate the relative units into absolute units by comparison of the Ba yield with the indication of UF6 fission chamber in the same medium. This has to be done separately for fast and slow neutron fission as the yields may be different. Another application of the technique developed is the measurement of thermal neutron density in an uraniferous medium without using a detector subject to variations of sensitivity according to the properties of the medium. (nowak)

  7. Numerical study of the influence of material parameters on the mechanical behaviour of a rehabilitated edentulous mandible.

    Science.gov (United States)

    Favot, Louis-Marc; Berry-Kromer, Valérie; Haboussi, Mohamed; Thiebaud, Frédéric; Ben Zineb, Tarak

    2014-03-01

    The study dealt with full dental prosthetic reconstruction on four implants. The aim was to analyse the influence of material parameters on the mechanical behaviour of the restored mandible compared to the natural mandible. A finite element model of an edentulous mandible with prosthetic rehabilitation was established. Four materials were investigated for the framework of the prosthesis (zirconia, titanium, gold and nickel-titanium (NiTi)), as well as three cortical bone thicknesses. Various muscles were employed to simulate the main stages of mastication. Three distinct phases of mastication were modelled: maximum intercuspation, incisal clench and unilateral molar clench. The zirconia framework demonstrated the highest stresses and NiTi the weakest. The highest stresses in the framework were obtained during maximum intercuspation. The highest stresses at the bone-implant interface were recorded on the working axial implant during unilateral molar clench and on tilted implants during maximum intercuspation. The influence of the framework's material stiffness on the stresses at the bone-implant interface was insignificant for axial implants (except the right implant during unilateral molar clench) and slightly more significant for tilted implants. Mandibular flexion decreased with an increase of the cortical bone thickness and the stiffness of the prosthetic framework's material. Among all materials, NiTi allowed a better preservation of the mandibular flexure, during all the mastication stages. Compared to stiffer materials, NiTi also permitted physiological mechanical conditions at the bone/implant interface, in almost all mastication stages. Copyright © 2013 Elsevier Ltd. All rights reserved.

  8. An incremental analysis of a deep drawing steel’s material behaviour undergoing the predeformation using drawbeads

    Science.gov (United States)

    Schmid, H.; Suttner, S.; Merklein, M.

    2017-09-01

    Nowadays lightweight design in metal forming processes leads to complex deep drawing geometries, which can cause multiple damages. Therefore, drawbeads are one way to regulate and control material flow during the forming process. Not only in research, but also in industrial practice, it could be determined that material is work hardened passing drawbead geometries. It particularly means when material is pre-deformed with tensile and alternating bending loads. This incident also gives the opportunity to utilize it in a reasonable way if examined properly. To investigate these findings, a process oriented and comprehensive analysis of the material behaviour during these forming operations is needed. In this paper, sheet metal strips are linearly drawn through a drawbead and stopped after passing the drawbead. Within this forming operation, the material undergoes non-linear straining before reaching the in-plane position again. Here, the process will be stopped to investigate a permanent strengthening local along the sheet thickness. Therefore, microhardness measurements are realized before and after passing the drawbead. Because of its common use and its wide known material data, a deep drawing steel DC will be used for these studies. Additionally, the strategy is applied to advanced high strength steel.

  9. Contained fissionly vaporized imploded fission explosive breeder reactor

    International Nuclear Information System (INIS)

    Marwick, E.F.

    1978-01-01

    Disclosed is a nuclear reactor system which produces useful thermal power and breeds fissile isotopes wherein large spherical complex slugs containing fissile and fertile isotopes as well as vaporizing and tamping materials are exploded seriatim in a large containing chamber having walls protected from the effects of the explosion by about two thousand tons of slurry of fissile and fertile isotopes in molten alkali metal. The slug which is slightly sub-critical prior to its entry into the centroid portion of the chamber, then becomes slightly more than prompt-critical because of the near proximity of neutron-reflecting atoms and of fissioning atoms within the slurry. The slurry is heated by explosion of the slugs and serves as a working fluid for extraction of heat energy from the reactor. Explosive debris is precipitated from the slurry and used for the fabrication of new slugs

  10. Biomechanical testing of materials in avian nests provides insight into nest construction behaviour

    OpenAIRE

    Biddle, Lucia E.; Deeming, D. Charles; Goodman, Adrian M.

    2015-01-01

    Animals that use materials to build nest structures have long since fascinated biologists and engineers alike. Avian nests are generally composed of collected materials brought together into a cup-like structure in which the bird sits to incubate eggs and, in many cases, it is where chicks are reared. Hence, the materials in a nest can be presumed to be loaded in compression, but relatively few studies have investigated the mechanical role of the nest elements and their position w...

  11. Fusion-fission dynamics

    International Nuclear Information System (INIS)

    Blocki, J.; Planeta, R.; Brzychczyk, J.; Grotowski, K.

    1991-04-01

    Classical dynamical calculations of the heavy ion induced fission process for the reactions 40 Ar+ 141 Pr, 20 Ne+ 165 Ho and 12 C+ 175 Lu leading to the iridium like nucleus have been performed. As a result prescission lifetimes were obtained and compared with the experimental values. The agreement between the calculated and experimental lifetimes indicates that the one-body dissipation picture is much more relevant in describing the fusion-fission dynamics than the two-body one. Somewhat bigger calculated times than the experimental ones in case of the C+Lu reaction at 16 MeV/nucleon may be a signal on the energy range applicability of the one-body dissipation model. (author)

  12. Building materials. VOC emissions, diffusion behaviour and implications from their use

    International Nuclear Information System (INIS)

    Katsoyiannis, Athanasios; Leva, Paolo; Barrero-Moreno, Josefa; Kotzias, Dimitrios

    2012-01-01

    Five cement- and five lime-based building materials were examined in an environmental chamber for their emissions of Volatile Organic Compounds (VOCs). Typical VOCs were below detection limits, whereas not routinely analysed VOCs, like neopentyl glycol (NPG), dominated the cement-based products emissions, where, after 72 h, it was found to occur, in levels as high as 1400 μg m −3 , accounting for up to 93% of total VOCs. The concentrations of NPG were not considerably changed between the 24 and 72 h of sampling. The permeability of building materials was assessed through experiments with a dual environmental chamber; it was shown that building materials facilitate the diffusion of chemicals through their pores, reaching equilibrium relatively fast (6 h). - Highlights: ► Neopentyl glycol is reported in emissions from building materials for the first time. ► Neopentyl glycol dominates the VOC emissions from cement-based building materials. ► A dual chamber was developed to control diffusion through building materials. ► Building materials facilitate diffusion of indoor air pollutants through their pores. - Neopentyl glycol was detected in high concentrations in emissions from building materials.

  13. Comparison of the dynamic behaviour of brain tissue and two model materials

    NARCIS (Netherlands)

    Brands, D.W.A.; Bovendeerd, P.H.M.; Peters, G.W.M.; Wismans, J.S.H.M.; Paas, M.H.J.W.; Bree, van J.L.M.J.; Brands, D.W.A.

    1999-01-01

    Linear viscoelastic material parameters of porcine brain tissue and two brain substitute/ materials for use in mechanical head models (edible bone gelatin and dielectric silicone gel) were determined in small deformation, oscillatory shear experiments. Frequencies to 1000 Hertz could be obtained

  14. Steady State Crack Propagation in Layered Material Systems Displaying Visco-plastic Behaviour

    DEFF Research Database (Denmark)

    Nielsen, Kim Lau

    2012-01-01

    The steady state fracture toughness of elastic visco-plastic materials is studied numerically, using both a conventional and a higher order model. Focus is on the combined effect of strain hardening, strain gradient hardening and strain rate hardening on cracking in layered material systems...

  15. Report on the studies on the corrosion behaviour of the constructional materials for the gate cooling system

    International Nuclear Information System (INIS)

    Elayathu, N.S.D.; Balachandra, J.

    1974-01-01

    The gate cooling system of the Trombay R-5 reactor, now under construction, is proposed to be a laminated gate designed to operate with 50 % KBO 2 solution within the temperature limits 30 deg C and 50 deg C. With a view to find suitable constructional materials for the gate, the corrosion behaviour of stainless steel 304 L(ASTM 240-69), lead (ASTM B-29), aluminium (as Boral), neoprene, perspex and carbon steel (ASTM A 302 grade B) has been investigated in 50 % KBO 2 solution at 45 deg C. After definite periods of exposure, their coupons were examined metallographically at different magnifications to assess the nature and extent of sub-surface attack. The results show that out of the materials studied, carbon steel, lead and aluminium are more liable to corrosion in the borate solution and hence their use should be avoided. (M.G.B.)

  16. Geology behind nuclear fission technology

    International Nuclear Information System (INIS)

    Dhana Raju, R.

    2005-01-01

    Geology appears to have played an important role of a precursor to Nuclear Fission Technology (NFT), in the latter's both birth from the nucleus of an atom of and most important application as nuclear power extracted from Uranium (U), present in its minerals. NFT critically depends upon the availability of its basic raw material, viz., nuclear fuel as U and/ or Th, extracted from U-Th minerals of specific rock types in the earth's crust. Research and Development of the Nuclear Fuel Cycle (NFC) depends heavily on 'Geology'. In this paper, a brief review of the major branches of geology and their contributions during different stages of NFC, in the Indian scenario, is presented so as to demonstrate the important role played by 'Geology' behind the development of NFT, in general, and NFC, in particular. (author)

  17. The fission track method

    International Nuclear Information System (INIS)

    Hansen, K.

    1990-01-01

    During the last decade fission track (FT) analysis has evolved as an important tool in exploration for hydrocarbon resources. Most important is this method's ability to yield information about temperatures at different times (history), and thus relate oil generation and time independently of other maturity parameters. The purpose of this paper is to introduce the basics of the method and give an example from the author's studies. (AB) (14 refs.)

  18. Characterisation of material behaviour in high temperature aqueous environments by means of electrochemical techniques

    International Nuclear Information System (INIS)

    Bojinov, M.; Laitinen, T.; Maekelae, K.; Sirkiae, P.; Beverskog, B.

    1998-01-01

    Electrochemical measurements in solutions simulating power plant coolants are complicated by the low conductivity of the water, especially in the case of boiling water reactor (BWR) environments. To be able to obtain useful information also in BWR conditions, electrochemical techniques based on a thin-layer electrode arrangement are introduced. This arrangement makes it possible to perform voltammetric and electrochemical impedance measurements in high-temperature water with a room temperature conductivity (κ) as low as 0.1 μScm -1 . A combination of these results with those obtained by means of measuring the resistance of the surface film using the contact electric resistance (CER) technique facilitates versatile characterisation of oxide film behaviour. Examples are given on impedance and CER measurements of the oxide films formed on AISI 316 stainless steel in high temperature high purity (κ -1 ) water and on OX18H10T stainless steel in VVER water. Correlations between temperature, hydrogen and oxygen content of the solution and the oxide behaviour are discussed. (author)

  19. Explaining the impact of poverty on old-age frailty in Europe: material, psychosocial and behavioural factors.

    Science.gov (United States)

    Stolz, Erwin; Mayerl, Hannes; Waxenegger, Anja; Freidl, Wolfgang

    2017-12-01

    Previous research found poverty to be associated with adverse health outcomes among older adults but the factors that translate low economic resources into poor physical health are not well understood. The goal of this analysis was to assess the impact of material, psychosocial, and behavioural factors as well as education in explaining the poverty-health link. In total, 28 360 observations from 11 390 community-dwelling respondents (65+) in the Survey of Health, Ageing and Retirement in Europe (2004-13, 10 countries) were analysed. Multilevel growth curve models were used to assess the impact of combined income and asset poverty risk on old-age frailty (frailty index) and associated pathway variables. In total, 61.8% of the variation of poverty risk on frailty level was explained by direct and indirect effects. Results stress the role of material and particularly psychosocial factors such as perceived control and social isolation, whereas the role of health behaviour was negligible. We suggest to strengthen social policy and public health efforts in order to fight poverty and its deleterious health effects from early age on as well as to broaden the scope of interventions with regard to psychosocial factors. © The Author 2017. Published by Oxford University Press on behalf of the European Public Health Association. All rights reserved.

  20. Analysis and description of the long-term creep behaviour of high-temperature gas turbine materials

    International Nuclear Information System (INIS)

    Bartsch, H.

    1985-01-01

    On a series of standard high-temperature gas turbine materials, creep tests were accomplished with the aim to obtain improved data on the long-term creep behaviour. The tests were carried out in the range of the main application temperatures of the materials and in the range of low stresses and elongations similar to operation conditions. They lasted about 5000 to 16000 h at maximum. At all important temperatures additional annealing tests lasting up to about 10000 h were carried out for the determination of a material-induced structure contraction. Thermal tension tests were effected for the description of elastoplastic short-time behaviour. As typical selection of materials the nickel investment casting alloys IN-738 LC, IN-939 and Udimet 500 for industrial turbine blades, IN-100 for aviation turbine blades and IN-713 C for integrally cast wheels of exhaust gas turbochargers were investigated, and also the nickel forge alloy Inconel 718 for industrial and aviation turbine disks and Nimonic 101 for industrial turbine blades and finally the cobalt alloy FSC 414 for guide blades and heat accumulation segments of industrial gas turbines. The creep tests were started on long-period individual creep testing machines with high strain measuring accuracy and economically continued on long-period multispecimen creep testing machines with long duration of test. The test results of this mixed test method were first subjected to a conventional evaluation in logarithmic time yield and creep diagrams which besides creep strength curves provided creep stress limit curves down to 0.2% residual strain. (orig./MM) [de

  1. [Fission product yields of 60 fissioning reactions]. Final report

    International Nuclear Information System (INIS)

    Rider, B.F.

    1995-01-01

    In keeping with the statement of work, I have examined the fission product yields of 60 fissioning reactions. In co-authorship with the UTR (University Technical Representative) Talmadge R. England ''Evaluation and Compilation of Fission Product Yields 1993,'' LA-UR-94-3106(ENDF-349) October, (1994) was published. This is an evaluated set of fission product Yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set

  2. Modeling the Behaviour of an Advanced Material Based Smart Landing Gear System for Aerospace Vehicles

    International Nuclear Information System (INIS)

    Varughese, Byji; Dayananda, G. N.; Rao, M. Subba

    2008-01-01

    The last two decades have seen a substantial rise in the use of advanced materials such as polymer composites for aerospace structural applications. In more recent years there has been a concerted effort to integrate materials, which mimic biological functions (referred to as smart materials) with polymeric composites. Prominent among smart materials are shape memory alloys, which possess both actuating and sensory functions that can be realized simultaneously. The proper characterization and modeling of advanced and smart materials holds the key to the design and development of efficient smart devices/systems. This paper focuses on the material characterization; modeling and validation of the model in relation to the development of a Shape Memory Alloy (SMA) based smart landing gear (with high energy dissipation features) for a semi rigid radio controlled airship (RC-blimp). The Super Elastic (SE) SMA element is configured in such a way that it is forced into a tensile mode of high elastic deformation. The smart landing gear comprises of a landing beam, an arch and a super elastic Nickel-Titanium (Ni-Ti) SMA element. The landing gear is primarily made of polymer carbon composites, which possess high specific stiffness and high specific strength compared to conventional materials, and are therefore ideally suited for the design and development of an efficient skid landing gear system with good energy dissipation characteristics. The development of the smart landing gear in relation to a conventional metal landing gear design is also dealt with

  3. Fission-track studies of uranium distribution in geological samples

    International Nuclear Information System (INIS)

    Brynard, H.J.

    1983-01-01

    The standard method of studying uranium distribution in geological material by registration of fission tracks from the thermal neutron-induced fission of 235 U has been adapted for utilisation in the SAFARI-1 reactor at Pelindaba. The theory of fission-track registration as well as practical problems are discussed. The method has been applied to study uranium distribution in a variety of rock types and the results are discussed in this paper. The method is very sensitive and uranium present in quantities far below the detection limit of the microprobe have been detected

  4. Determination of 233U, 235U, 238U and 239Pu fission yields induced by fission and 14.7 MeV neutrons

    International Nuclear Information System (INIS)

    Laurec, Jean; Adam, Albert; Bruyne, Thierry de.

    1981-12-01

    The 233 U, 235 U, 238 U, 239 Pu fission yields have been determined by a radiochemical method. A target and a fission chamber made of same fissible material are irradied together. The total fission number is measured from the fission chamber. The fission product activities are directly measured on the target using calibrated Ge-Li detectors. The fissible material masses are determined by alpha and mass spectrometries. The irradiations were made on the critical assemblies PROSPERO and CALIBAN and on the 14 MeV neutron generator of C.E. VALDUC. 3 to 5% fission yield errors are got for the most measured nuclides: 95 Zr, 97 Zr, 99 Mo, 103 Ru, 131 I, 132 Te, 140 Ba, 141 Ce, 143 Ce, 144 Ce, 147 Nd [fr

  5. Fission neutron output measurements at LANSCE

    International Nuclear Information System (INIS)

    Nelson, Ronald Owen; Haight, Robert C.; Devlin, Matthew J.; Fotiadis, Nikolaos; Laptev, Alexander; O'Donnell, John M.; Taddeucci, Terry N.; Tovesson, Fredrik; Ullmann, J.L.; Wender, Stephen A.; Bredeweg, T.A.; Jandel, M.; Vieira, D.J.; Wu, Ching-Yen; Becker, J.A.; Stoyer, M.A.; Henderson, R.; Sutton, M.; Belier, Gilbert; Chatillon, A.; Granier, Thierry; Laurent, Benoit; Taieb, Julien

    2010-01-01

    Accurate data for both physical properties and fission properties of materials are necessary to properly model dynamic fissioning systems. To address the need for accurate data on fission neutron energy spectra, especially at outgoing neutron energies below about 200 keV and at energies above 8 MeV, ongoing work at LANSCE involving collaborators from LANL, LLNL and CEA Bruyeres-le-Chatel is extending the energy range, efficiency and accuracy beyond previous measurements. Initial work in the outgoing neutron energy range from 1 to 7 MeV is consistent with current evaluations and provides a foundation for extended measurements. As part of these efforts, a new fission fragment detector that reduces backgrounds and improves timing has been designed fabricated and tested, and new neutron detectors are being assessed for optimal characteristics. Simulations of experimental designs are in progress to ensure that accuracy goals are met. Results of these measurements will be incorporated into evaluations and data libraries as they become available.

  6. Capsule shell material impacts the in vitro disintegration and dissolution behaviour of a green tea extract

    OpenAIRE

    Glube, Natalie; Moos, Lea von; Duchateau, Guus

    2013-01-01

    Purpose In vitro disintegration and dissolution are routine methods used to assess the performance and quality of oral dosage forms. The purpose of the current work was to determine the potential for interaction between capsule shell material and a green tea extract and the impact it can have on the release. Methods A green tea extract was formulated into simple powder-in-capsule formulations of which the capsule shell material was either of gelatin or HPMC origin. The disintegration times we...

  7. Behaviour of contact layer material between cermet fuel element core and can

    International Nuclear Information System (INIS)

    Gavrilin, S.S.; Permyakov, L.N.; Simakov, G.A.; Chernikov, A.S.

    1996-01-01

    The structural state of the contact layer between the shell of the Zr1Nb alloy and cermet fuel element core containing up to 70% of uranium dioxides is experimental studied. The silumin alloy was used as contact material. The results of studies on interaction zones, formed on the Zr1Nb - silumin boundary after fuel elements manufacture and also under temperature conditions, modeling the maximum design and hypothetical accidents accompanied by the contact material melting, are presented [ru

  8. Tritium chemistry in fission and fusion reactors

    International Nuclear Information System (INIS)

    Roth, E.; Masson, M.; Briec, M.

    1986-09-01

    We are interested in the behaviour of tritium inside the solids where it is generated both in the case of fission nuclear reactor fuel elements, and in that of blankets of future fusion reactor. In the first case it is desirable to be able to predict whether tritium will be found in the hulls or in the uranium oxide, and under what chemical form, in order to take appropriate steps for it's removal in reprocessing plants. In fusion reactors breeding large amounts of tritium and burning it in the plasma should be accomplished in as short a cycle as possible in order to limit inventories that are associated with huge activities. Mastering the chemistry of every step is therefore essential. Amounts generated are not of the same order of magnitude in the two cases studied. Ternary fissions produce about 66 10 13 Bq (18 000 Ci) per year of tritium in a 1000 MWe fission generator, i.e., about 1.8 10 10 Bq (0.5 Ci) per day per ton of fuel

  9. Coulomb fission and transfer fission at heavy ion collisions

    International Nuclear Information System (INIS)

    Himmele, G.

    1981-01-01

    In the present thesis the first direct evidence of nuclear fission after inelastic scattering of heavy ions (sup(183,184)W, 152 Sm → 238 U; 184 W → 232 Th; 184 W, 232 Th → 248 Cm) is reported. Experiments which were performed at the UNILAC of the Gesellschaft fuer Schwerionenforschung in Darmstadt show the observed heavy ion induced fission possesses significant properties of the Coulomb fission. The observed dependence of the fission probability for inelastic scattering on the projectile charge proves that the nuclear fission is mediated by the electromagnetic interaction between heavy ions. This result suggests moreover a multiple Coulomb-excitation preceding the fission. Model calculations give a first indication, that the Coulomb fission proceeds mainly from the higher β phonons. In the irradiation with 184 W the fission probability of 232 Th is for all incident energies about 40% smaller that at 238 U. The target dependence of the Coulomb fission however doesn't allow, to give quantitative statements about the position and B(E2)-values of higher lying β phonons. (orig./HSI) [de

  10. Influence of Parameters of Core Bingham Material on Critical Behaviour of Three-Layered Annular Plate

    Directory of Open Access Journals (Sweden)

    Pawlus Dorota

    2017-12-01

    Full Text Available The paper presents the dynamic response of annular three-layered plate subjected to loads variable in time. The plate is loaded in the plane of outer layers. The plate core has the electrorheological properties expressed by the Bingham body model. The dynamic stability loss of plate with elastic core is determined by the critical state parameters, particularly by the critical stresses. Numerous numerical observations show the influence of the values of viscosity constant and critical shear stresses, being the Bingham body parameters, on the supercritical viscous fluid plate behaviour. The problem has been solved analytically and numerically using the orthogonalization method and finite difference method. The solution includes both axisymmetric and asymmetric plate dynamic modes.

  11. The gem anvil cell: high-pressure behaviour of diamond and related materials

    International Nuclear Information System (INIS)

    Xu Jian; Mao Hokwang; Hemley, Russell J

    2002-01-01

    The moissanite anvil cell has been used to study the high-pressure behaviour of diamond. The first-order Raman shift of diamond shows a strong dependence on hydrostaticity, with very different pressure dependences observed under hydrostatic and non-hydrostatic conditions. The shift of the second-order Raman band under hydrostatic pressures was determined for the first time. Sapphire has almost no peaks above 1000 cm -1 in the Raman spectrum and no absorption in the ultraviolet range; it is therefore especially useful for studies in those spectral regions. A sapphire anvil cell was used in a study of graphite up to 24 GPa. A phase transition was found near 18 GPa, consistent with previous reports, and no peaks characteristic of diamond in the 1330 cm -1 range were found, indicating that the phase is not diamond

  12. The gem anvil cell: high-pressure behaviour of diamond and related materials

    CERN Document Server

    Xu Jian; Hemley, R J

    2002-01-01

    The moissanite anvil cell has been used to study the high-pressure behaviour of diamond. The first-order Raman shift of diamond shows a strong dependence on hydrostaticity, with very different pressure dependences observed under hydrostatic and non-hydrostatic conditions. The shift of the second-order Raman band under hydrostatic pressures was determined for the first time. Sapphire has almost no peaks above 1000 cm sup - sup 1 in the Raman spectrum and no absorption in the ultraviolet range; it is therefore especially useful for studies in those spectral regions. A sapphire anvil cell was used in a study of graphite up to 24 GPa. A phase transition was found near 18 GPa, consistent with previous reports, and no peaks characteristic of diamond in the 1330 cm sup - sup 1 range were found, indicating that the phase is not diamond.

  13. Evaluation of the suitability of tin slag in cementitious materials: Mechanical properties and Leaching behaviour

    Science.gov (United States)

    Rustandi, Andi; Wafa' Nawawi, Fuad; Pratesa, Yudha; Cahyadi, Agung

    2018-01-01

    Tin slag, a by-product of tin production has been used in cementitious application. The present investigation focuses on the suitability of tin slag as primary component in cement and as component that substitute some amount of Portland Cement. The tin slags studied were taken from Bangka, Indonesia. The main contents of the tin slag are SiO2, Al2O3, and Fe2O3 according to the XRF investigation. The aim of this article was to study the mechanical behaviour (compressive strength), microstructure and leaching behaviour of tin slag blended cement. This study used air-cooled tin slag that had been passed through 400# sieve to replace Portland Cement with ratio 0, 10, 20, 30, 40 by weight. Cement pastes and tin slag blended cement pastes were prepared by using water/cement ratio (W/C) of 0.40 by weight and hydrated for various curing ages of 3, 7, 14 days The microstructure of the raw tin slag was investigated using Scanning Electron Microscope (SEM). The phase composition of each cement paste was investigated using X-ray Diffraction (XRD). The aim of the leachability test was to investigate the environmental impacts of tin slag blended cement product in the range 4-8 pH by using static pH-dependent leaching test. The result show that the increase of the tin slag content decreasing the mortar compressive strength at early ages. The use of tin slag in cement provide economic benefits for all related industries.

  14. Transport of fission products in matrix and graphite

    International Nuclear Information System (INIS)

    Hoinkis, E.

    1983-06-01

    In the past years new experimental methods were applied to or developed for the investigation of fission product transport in graphitic materials and to characterization of the materials. Models for fission product transport and computer codes for the calculation of core release rates were improved. Many data became available from analysis of concentration profiles in HTR-fuel elements. New work on the effect on diffusion of graphite corrosion, fast neutron flux and fluence, heat treatment, chemical interactions and helium pressure was reported on recently or was in progress in several laboratories. It seemed to be the right time to discuss the status of transport of metallic fission products in general, and in particular the relationship between structural and transport properties. Following a suggestion a Colloquium was organized at the HMI Berlin. Interdisciplinary discussions were stimulated by only inviting a limited number of participants who work in different fields of graphite and fission product transport research. (orig./RW)

  15. Influence of the cosmic-ray induced fission tracks on the fission track of extraterrestric minerals via the 238U spontaneous fission

    International Nuclear Information System (INIS)

    Damm, G.; Thiel, K.

    1977-01-01

    The age determined by counting fission tracks of lunar and meteorite materials is obviously falsified by additional fission track parts not to be accounted for by the spontaneous fission of uranium 238. For this p and n induced fissions of U, Th and other hreavy elements through the cosmic radiation come into consideration. In order to determine the possible part of such interference factors, a simulation experiment at the proton synchrocycloton (CERN, Geneva) has been carried out and independently of this, the production rates for the p and n induced U, Th, Bi, Pb and Au in the surface-near regolith layers of the moon were calculated. It could be seen that the irradiation age as well as the spacial distribution of the heavy metals in the samples to be dated must be considered. (RB) [de

  16. Dynamic behaviour of interphases and its implication on high-energy-density cathode materials in lithium-ion batteries

    Science.gov (United States)

    Li, Wangda; Dolocan, Andrei; Oh, Pilgun; Celio, Hugo; Park, Suhyeon; Cho, Jaephil; Manthiram, Arumugam

    2017-01-01

    Undesired electrode–electrolyte interactions prevent the use of many high-energy-density cathode materials in practical lithium-ion batteries. Efforts to address their limited service life have predominantly focused on the active electrode materials and electrolytes. Here an advanced three-dimensional chemical and imaging analysis on a model material, the nickel-rich layered lithium transition-metal oxide, reveals the dynamic behaviour of cathode interphases driven by conductive carbon additives (carbon black) in a common nonaqueous electrolyte. Region-of-interest sensitive secondary-ion mass spectrometry shows that a cathode-electrolyte interphase, initially formed on carbon black with no electrochemical bias applied, readily passivates the cathode particles through mutual exchange of surface species. By tuning the interphase thickness, we demonstrate its robustness in suppressing the deterioration of the electrode/electrolyte interface during high-voltage cell operation. Our results provide insights on the formation and evolution of cathode interphases, facilitating development of in situ surface protection on high-energy-density cathode materials in lithium-based batteries. PMID:28443608

  17. Corrosion behaviour of container materials for the disposal of high-level wastes in rock salt formations

    International Nuclear Information System (INIS)

    Smailos, E.; Schwarzkopf, W.; Koester, R.

    1986-01-01

    In 1983-84 extensive laboratory-scale experiments (immersion tests) to evaluate the long-term corrosion behaviour of selected materials in salt brines and first in situ experiments were performed. In the laboratory experiments the materials Ti 99.8-Pd, Hastelloy C4 and hot-rolled low carbon steel (reference materials in the joint European corrosion programme) as well as cast steel, spheoroidal cast iron, Si-cast iron and the Ni-Resists type D2 and D4 were investigated. The investigated parameters were: temperature (90 0 C; 170 0 C, 200 0 C), gamma-radiation (10 5 rad/h) and different compositions of salt brines. The results obtained show that, in addition to Ti 99.8-Pd, also Hastelloy C4 and unalloyed steels are in principle suitable for being used for long-term stable HLW-containers if the gamma dose rate is reduced by suitable shielding. Furthermore, the susceptibility of Hastelloy C4 to crevice corrosion must be taken into account. Further studies will be necessary to provide final evidence of the suitability of the materials examined. These will mainly involve clarification of questions related to hydrogen embrittlement (Ti 99.8-Pd, unalloyed steels) and to the influence of pressure and saline impurities (e.g. antiJ, antiBr) on corrosion

  18. Measurements of fission cross-sections and of neutron production rates; Mesures de sections efficaces de fission et du nombre de neutrons prompts emis par fission

    Energy Technology Data Exchange (ETDEWEB)

    Billaud, P; Clair, C; Gaudin, M; Genin, R; Joly, R; Leroy, J L; Michaudon, A; Ouvry, J; Signarbieux, C; Vendryes, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    a) Measurements of neutron induced fission cross-sections in the low energy region. The variation of the fission cross sections of several fissile isotopes has been measured and analysed, for neutron energies below 0,025 eV. The monochromator was a crystal spectrometer used in conjunction with a mechanical velocity selector removing higher order Bragg reflections. The fissile material was laid down on the plates of a fission chamber by painting technic. An ionization chamber, having its plates coated with thin {sup 10}B layers, was used as the neutron flux monitor. b) Measurement of the fission cross section of {sup 235}U. We intend to measure the variation of the neutron induced fission cross section of {sup 235}U over the neutron energy range from 1 keV by the time of flight method. The neutron source is the uranium target of a pulsed 28 MeV electron linear accelerator. The detector is a large fission chamber, with parallel plates, containing about 10 g of {sup 235}U (20 deposits of 25 cm diameter). The relative fission data were corrected for the neutron spectrum measured with a set of BF{sub 3} proportional counters. c) Mean number {nu} of neutrons emitted in neutron induced fission. We measured the value of {nu} for several fissile isotopes in the case of fission induced by 14 MeV neutrons. The 14 MeV neutrons were produced by D (t, n) {alpha} reaction by means of a 300 kV Cockcroft Walton generator. (author)Fren. [French] a) Mesures de sectionficaces de fission a basse energie. Nous avons mesure et analyse la variation de la section efficace de fission de divers isotopes fissiles pour des neutrons d'energie inferieure a 0,025 eV. Le monochromateur est constitue par un spectrometre a cristal auquel est associe un selecteur mecanique destine a eliminer les diffractions de Bragg d'ordre superieur au premier. Le materiau fissile est contenu dans une chambre a fission sous forme de depots realises par peinture; une chambre d'ionisation a depots minces de B{sub 10

  19. Theoretical and experimental study of high strain, high strain rate materials viscoplastic behaviour. Application to Mars 190 steel and tantalum

    International Nuclear Information System (INIS)

    Juanicotena, A.

    1998-01-01

    This work enters in the general framework of the study and modelling of metallic materials viscoplastic behaviour in the area of high strain and high strain rate, from 10 4 to 10 5 s -1 . We define a methodology allowing to describe the behaviour of armor steel Mars 190 and tantalum in the initial area. In a first time, the study of visco-plasticity physical mechanisms shows the necessity to take into account some fundamental processes of the plastic deformation. Then, the examination of various constitutive relations allows to select the Preston-Tonks-Wallace model, that notably reproduce the physical phenomenon of the flow stress saturation. In a second part, a mechanical characterization integrating loading direction, strain rate and temperature effects is conducted on the two materials. Moreover, these experimental results allow to calculate associated constants to Preston-Tonks-Wallace, Zerilli-Armstrong and Johnson-Cook models for each material. In a third time, in order to evaluate and to validate these constitutive laws, we conceive and develop an experimental device open to reach the area of study: the expanding spherical shell test. It concerns to impose a free radial expanding to a thin spherical shell by means a shock wave generated by an explosive. By the radial expanding velocity measure, we can determine stress, strain rate and strain applied on the spherical shell at each time. In a four and last part, we evaluate constitutive models out of their optimization area's. This validation is undertaken by comparisons 'experimental results/calculations' with the help of global experiences like expanding spherical shell test and Taylor test. (author)

  20. Methylene blue adsorption by algal biomass based materials: biosorbents characterization and process behaviour.

    Science.gov (United States)

    Vilar, Vítor J P; Botelho, Cidália M S; Boaventura, Rui A R

    2007-08-17

    Dead algal biomass is a natural material that serves as a basis for developing a new family of sorbent materials potentially suitable for many industrial applications. In this work an algal industrial waste from agar extraction process, algae Gelidium and a composite material obtained by immobilization of the algal waste with polyacrylonitrile (PAN) were physical characterized and used as biosorbents for dyes removal using methylene blue as model. The apparent and real densities and the porosity of biosorbents particles were determined by mercury porosimetry and helium picnometry. The methylene blue adsorption in the liquid phase was the method chosen to calculate the specific surface area of biosorbent particles as it seems to reproduce better the surface area accessible to metal ions in the biosorption process than the N2 adsorption-desorption dry method. The porous texture of the biosorbents particles was also studied. Equilibrium isotherms are well described by the Langmuir equation, giving maximum uptake capacities of 171, 104 and 74 mg g(-1), respectively for algae, algal waste and composite material. Kinetic experiments at different initial methylene blue concentrations were performed to evaluate the equilibrium time and the importance of the driving force to overcome mass transfer resistances. The pseudo-first-order and pseudo-second-order kinetic models adequately describe the kinetic data. The biosorbents used in this work proved to be promising materials for removing methylene blue from aqueous solutions.

  1. Fission barriers of light nuclei

    International Nuclear Information System (INIS)

    Grotowski, K.; Planeta, R.; Blann, M.; Komoto, T.

    1989-01-01

    Experimental fission excitation functions for compound nuclei /sup 52/Fe, /sup 49/Cr, /sup 46/V, and /sup 44/Ti formed in heavy-ion reactions are analyzed in the Hauser-Feshbach/Bohr-Wheeler formalism using fission barriers based on the rotating liquid drop model of Cohen et al. and on the rotating finite range model of Sierk. We conclude that the rotating finite range approach gives better reproduction of experimental fission yields, consistent with results found for heavier systems

  2. Singlet exciton fission in polycrystalline pentacene: from photophysics toward devices.

    Science.gov (United States)

    Wilson, Mark W B; Rao, Akshay; Ehrler, Bruno; Friend, Richard H

    2013-06-18

    Singlet exciton fission is the process in conjugated organic molecules bywhich a photogenerated singlet exciton couples to a nearby chromophore in the ground state, creating a pair of triplet excitons. Researchers first reported this phenomenon in the 1960s, an event that sparked further studies in the following decade. These investigations used fluorescence spectroscopy to establish that exciton fission occurred in single crystals of several acenes. However, research interest has been recently rekindled by the possibility that singlet fission could be used as a carrier multiplication technique to enhance the efficiency of photovoltaic cells. The most successful architecture to-date involves sensitizing a red-absorbing photoactive layer with a blue-absorbing material that undergoes fission, thereby generating additional photocurrent from higher-energy photons. The quest for improved solar cells has spurred a drive to better understand the fission process, which has received timely aid from modern techniques for time-resolved spectroscopy, quantum chemistry, and small-molecule device fabrication. However, the consensus interpretation of the initial studies using ultrafast transient absorption spectroscopy was that exciton fission was suppressed in polycrystalline thin films of pentacene, a material that would be otherwise expected to be an ideal model system, as well as a viable candidate for fission-sensitized photovoltaic devices. In this Account, we review the results of our recent transient absorption and device-based studies of polycrystalline pentacene. We address the controversy surrounding the assignment of spectroscopic features in transient absorption data, and illustrate how a consistent interpretation is possible. This work underpins our conclusion that singlet fission in pentacene is extraordinarily rapid (∼80 fs) and is thus the dominant decay channel for the photoexcited singlet exciton. Further, we discuss our demonstration that triplet excitons

  3. Building materials. VOC emissions, diffusion behaviour and implications from their use.

    Science.gov (United States)

    Katsoyiannis, Athanasios; Leva, Paolo; Barrero-Moreno, Josefa; Kotzias, Dimitrios

    2012-10-01

    Five cement- and five lime-based building materials were examined in an environmental chamber for their emissions of Volatile Organic Compounds (VOCs). Typical VOCs were below detection limits, whereas not routinely analysed VOCs, like neopentyl glycol (NPG), dominated the cement-based products emissions, where, after 72 h, it was found to occur, in levels as high as 1400 μg m(-3), accounting for up to 93% of total VOCs. The concentrations of NPG were not considerably changed between the 24 and 72 h of sampling. The permeability of building materials was assessed through experiments with a dual environmental chamber; it was shown that building materials facilitate the diffusion of chemicals through their pores, reaching equilibrium relatively fast (6 h). Copyright © 2012 Elsevier Ltd. All rights reserved.

  4. Fatigue behaviour of coke drum materials under thermal-mechanical cyclic loading

    Directory of Open Access Journals (Sweden)

    Jie Chen

    2014-01-01

    Full Text Available Coke drums are vertical pressure vessels used in the delayed coking process in petroleum refineries. Significant temperature variation during the delayed coking process causes damage in coke drums in the form of bulging and cracking. There were some studies on the fatigue life estimation for the coke drums, but most of them were based on strain-fatigue life curves at constant temperatures, which do not consider simultaneous cyclic temperature and mechanical loading conditions. In this study, a fatigue testing system is successfully developed to allow performing thermal-mechanical fatigue (TMF test similar to the coke drum loading condition. Two commonly used base and one clad materials of coke drums are then experimentally investigated. In addition, a comparative study between isothermal and TMF lives of these materials is conducted. The experimental findings lead to better understanding of the damage mechanisms occurring in coke drums and more accurate prediction of fatigue life of coke drum materials.

  5. Prediction of material creep behaviour for strain based life assessment applications

    Energy Technology Data Exchange (ETDEWEB)

    Rantala, J.H.; Hurst, R.C. [EC JRC IAM, Petten (Netherlands); Bregani, F. [ENEL, Milan (Italy)

    1998-12-31

    In this work the idea of using constant load uniaxial creep test results instead of constant stress results for developing a CDM creep model for the P92 material is demonstrated. Due to limited availability of creep test results this work is based on incomplete test data and a general stress rupture line. In spite of these limitations a material creep model was developed for use in a FE analysis. Using P91 material as an example, a method is proposed to account for differences in strain evolution as a function of stress which normally manifests itself as lower strain values at low stresses in a normalised time-strain plot. This allows the CDM model to be used both in FE analysis and in strain-based life assessment engineering calculations. (orig.) 3 refs.

  6. Prediction of material creep behaviour for strain based life assessment applications

    Energy Technology Data Exchange (ETDEWEB)

    Rantala, J H; Hurst, R C [EC JRC IAM, Petten (Netherlands); Bregani, F [ENEL, Milan (Italy)

    1999-12-31

    In this work the idea of using constant load uniaxial creep test results instead of constant stress results for developing a CDM creep model for the P92 material is demonstrated. Due to limited availability of creep test results this work is based on incomplete test data and a general stress rupture line. In spite of these limitations a material creep model was developed for use in a FE analysis. Using P91 material as an example, a method is proposed to account for differences in strain evolution as a function of stress which normally manifests itself as lower strain values at low stresses in a normalised time-strain plot. This allows the CDM model to be used both in FE analysis and in strain-based life assessment engineering calculations. (orig.) 3 refs.

  7. Influence of material and solution composition on the extrusion/erosion behaviour of compacted bentonite

    International Nuclear Information System (INIS)

    Schatz, Timothy; Martikainen, Jari; Koskinen, Kari

    2010-01-01

    Document available in extended abstract form only. In principle, in a KBS-3 type repository, the volume of a deposition hole is fixed and the bentonite buffer mass accordingly balanced to lead to the development of a suitable swelling pressure upon saturation. However, fractures intersecting the deposition holes give rise to the possibility that volume constrained conditions do not universally exist. Such fractures may provide pathways for the continued, localised, free swelling of bentonite buffer material. Loss of mass from the deposition hole by extrusion into intersecting fractures may compromise the long-term safety and performance of the buffer component of the engineered barrier system. Furthermore, the continued hydration and expansion of extruded bentonite in these fracture environments could lead to the separation of colloid-sized (or larger) particles by diffusion or shear which may have to be accounted for in possible radionuclide migration scenarios. Geochemical conditions, with respect to both solution and material composition, are considered to play important roles regarding the fracture extrusion/erosion of bentonite buffer material. For example, calcium-montmorillonite exhibits limited free swelling relative to sodium-montmorillonite and the colloidal and rheological properties of montmorillonite dispersions are sensitive to the presence of electrolytes. Insofar as both the buffer material composition (due to ion exchange) and groundwater composition (dilution resulting from infiltration of glacial melt water) are expected to evolve with time, so too might the potential for fracture extrusion/erosion of buffer material vary over time. The hydraulic characteristics of the intersecting fracture are expected to influence the extrusion/erosion process as well. To evaluate the effect of material and solution composition on the potential for extrusion of buffer mass into intersecting fractures, a series of batch experiments were performed. In these

  8. Chemical concentration of a new natural spontaneously fissionable nuclide from solutions with low salt background

    International Nuclear Information System (INIS)

    Korotkin, Yu.S.; Ter-Akop'yan, G.M.; Popeko, A.G.; Drobina, T.P.; Zhuravleva, E.L.

    1982-01-01

    The results of experiments on further concentration of a new natural spontaneously fissionable nuclide, the concentrates of which form the Cheleken geothermal brines have been obtained, are presented. The conclusions are drown about the chemical nature of a new spontaneously fissionable nuclide. It is a chalcophile element which copreipitates with sulphides of copper, lead, arsenic and mercury from weakly acid solutions. The behaviour of the new nuclide in sulphide systems in many respects is similar to the behaviour of polonium, astatine and probably of bismuth. The most probable stable valence of the new nuclide varies from +1 up to +3. The data available on the chemical behaviour of the new nuclide as well as the analysis over contamination by spontaneously fissionable isotopes permit to state that the new natural spontaneously fissionable nuclide does not relate to the known isotopes

  9. Microbial ecology of Rum Jungle, III. Leaching behaviour of sulphidic waste material under controlled conditions

    International Nuclear Information System (INIS)

    Babij, T.; Goodman, A.; Khalid, A.M.; Ralph, B.J.

    1981-12-01

    The discharge, into river systems, of acid and heavy metals generated by leaching of sulphidic waste materials at the abandoned opencut uranium mine at Rum Jungle, Northern Territory, is causing continuing pollution of the surrounding environment. The maximum effects of acid and microorganisms on samples from the overburden dump material, under defined and controlled environmental conditions, were assessed using reactor systems. These samples came from the overburden dump resulting from the mining of White's orebody. Similarly, the stability of tailings material under conditions of flooding and increasing acidity was determined. At ph 2.5, metals in White's dump material were solubilised by acid attack only, whereas at pH 3.5, bacterial activity (principally that of Thiobacillus ferrooxidans) generated acidity and contributed significantly to metal release. Under microaerophilic conditions Thiobacillus ferrooxidans continued to effect metal release from the ore, but did not produce further acidity. If White's overburden is returned to the acidic, flooded opencuts, complete solubilisation of the material will occur. The exclusion of oxygen from the dump will not necessarily stop bacterially catalysed leaching processes. Under highly aerated and agitated flooded conditions the tailings material was not active, except for copper release of about 2 g kg -1 ore at pH 4.0. The only deleterious element released by increasing acidity was copper, which was 100 per cent solubilised at pH 2.5. Uranium was always lss than 3 μg kg -1 ore, and lead was detected only at pH 2.5. Indigenous leaching bacteria did not develop

  10. Heterogeneous catalytic materials solid state chemistry, surface chemistry and catalytic behaviour

    CERN Document Server

    Busca, Guido

    2014-01-01

    Heterogeneous Catalytic Materials discusses experimental methods and the latest developments in three areas of research: heterogeneous catalysis; surface chemistry; and the chemistry of catalysts. Catalytic materials are those solids that allow the chemical reaction to occur efficiently and cost-effectively. This book provides you with all necessary information to synthesize, characterize, and relate the properties of a catalyst to its behavior, enabling you to select the appropriate catalyst for the process and reactor system. Oxides (used both as catalysts and as supports for cata

  11. Nuclear Forensics and Radiochemistry: Fission

    Energy Technology Data Exchange (ETDEWEB)

    Rundberg, Robert S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-11-07

    Radiochemistry has been used to study fission since it’ discovery. Radiochemical methods are used to determine cumulative mass yields. These measurements have led to the two-mode fission hypothesis to model the neutron energy dependence of fission product yields. Fission product yields can be used for the nuclear forensics of nuclear explosions. The mass yield curve depends on both the fuel and the neutron spectrum of a device. Recent studies have shown that the nuclear structure of the compound nucleus can affect the mass yield distribution.

  12. Fission yield measurements at IGISOL

    Science.gov (United States)

    Lantz, M.; Al-Adili, A.; Gorelov, D.; Jokinen, A.; Kolhinen, V. S.; Mattera, A.; Moore, I.; Penttilä, H.; Pomp, S.; Prokofiev, A. V.; Rakopoulos, V.; Rinta-Antila, S.; Simutkin, V.; Solders, A.

    2016-06-01

    The fission product yields are an important characteristic of the fission process. In fundamental physics, knowledge of the yield distributions is needed to better understand the fission process. For nuclear energy applications good knowledge of neutroninduced fission-product yields is important for the safe and efficient operation of nuclear power plants. With the Ion Guide Isotope Separator On-Line (IGISOL) technique, products of nuclear reactions are stopped in a buffer gas and then extracted and separated by mass. Thanks to the high resolving power of the JYFLTRAP Penning trap, at University of Jyväskylä, fission products can be isobarically separated, making it possible to measure relative independent fission yields. In some cases it is even possible to resolve isomeric states from the ground state, permitting measurements of isomeric yield ratios. So far the reactions U(p,f) and Th(p,f) have been studied using the IGISOL-JYFLTRAP facility. Recently, a neutron converter target has been developed utilizing the Be(p,xn) reaction. We here present the IGISOL-technique for fission yield measurements and some of the results from the measurements on proton induced fission. We also present the development of the neutron converter target, the characterization of the neutron field and the first tests with neutron-induced fission.

  13. Fission yield measurements at IGISOL

    Directory of Open Access Journals (Sweden)

    Lantz M.

    2016-01-01

    Full Text Available The fission product yields are an important characteristic of the fission process. In fundamental physics, knowledge of the yield distributions is needed to better understand the fission process. For nuclear energy applications good knowledge of neutroninduced fission-product yields is important for the safe and efficient operation of nuclear power plants. With the Ion Guide Isotope Separator On-Line (IGISOL technique, products of nuclear reactions are stopped in a buffer gas and then extracted and separated by mass. Thanks to the high resolving power of the JYFLTRAP Penning trap, at University of Jyväskylä, fission products can be isobarically separated, making it possible to measure relative independent fission yields. In some cases it is even possible to resolve isomeric states from the ground state, permitting measurements of isomeric yield ratios. So far the reactions U(p,f and Th(p,f have been studied using the IGISOL-JYFLTRAP facility. Recently, a neutron converter target has been developed utilizing the Be(p,xn reaction. We here present the IGISOL-technique for fission yield measurements and some of the results from the measurements on proton induced fission. We also present the development of the neutron converter target, the characterization of the neutron field and the first tests with neutron-induced fission.

  14. Fusion-fission type collisions

    International Nuclear Information System (INIS)

    Oeschler, H.

    1980-01-01

    Three examples of fusion-fission type collisions on medium-mass nuclei are investigated whether the fragment properties are consistent with fission from equilibrated compound nuclei. Only in a very narrow band of angular momenta the data fulfill the necessary criteria for this process. Continuous evolutions of this mechnism into fusion fission and into a deep-inelastic process and particle emission prior to fusion have been observed. Based on the widths of the fragment-mass distributions of a great variety of data, a further criterion for the compound-nucleus-fission process is tentatively proposed. (orig.)

  15. Structure, properties and wear behaviour of multilayer coatings consisting of metallic and covalent hard materials, prepared by magnetron sputtering

    International Nuclear Information System (INIS)

    Schier, V.

    1995-12-01

    Novel multilayer coatings with metallic and covalent layer materials were prepared by magnetron sputtering and characterised concerning structure, properties and application behaviour. At first single layer coatings were deposited for the determination of the material properties. To evaluate relations between structure and properties of the multilayer coatings, different multilayer concepts were realised: - coatings consisting of at most 7 layers of metallic hard materials, - 100-layer coatings consisting of metallic and covalent hard materials, - TiN-TiC multilayer coatings with different numbers of layers (between 10 and 1000), - 150-layer coatings, based on TiN-TiC multilayers, with thin ( 4 C, AlN, SiC, a:C, Si 3 N 4 , SiAlON). X-rays and electron microscopic analysis indicate in spite of nonstoichiometric compositions single phase crystalline structures for nonreactively and reactively sputtered metastable single layer Ti(B,C)-, Ti(B,N)- and Ti(B,C,N)-coatings. These single layer coatings show excellent mechanical properties (e.g. hardness values up to 6000 HV0,05), caused by lattice stresses as well as by atomic bonding conditions similar to those in c:BN and B 4 C. The good tribological properties shown in pin-on-disk-tests can be attributed to the very high hardness of the coatings. The coatings consisting of at most 7 layers of metallic hard materials show good results mainly for the cutting of steel Ck45, due to the improved mechanical properties (e.g. hardness, toughness) of the multilayers compared to the single layer coatings. This improvement is caused by inserting the hard layer materials and the coherent reinforcement of the coatings. (orig.)

  16. Towards a mechanistic understanding of transient fission gas release

    International Nuclear Information System (INIS)

    Matthews, J.R.; Small, G.J.

    1989-01-01

    Recent experimental results on transient fission gas release from oxide fuels are briefly reviewed. These together with associated microstructural observations are compared with the main models for fission gas behaviour. Single gas atom diffusion, bubble migration, heterogeneous percolation and grain boundary sweeping are examined as possible release mechanisms. The role of gas trapping in bubbles and re-solution by irradiation and thermal processes are included in the comparison. As much of the data, and the main range of interest for light water reactor fuels, is for release during mild transients in fuel with a burn-up below 4%, the role of gas retention on grain boundaries is very important and in some cases dominant. The grain boundaries are found to respond very differently to various gas arrival rates and to local temperature conditions. This can lead to early interlinkage and release in some cases, but retention with accompanying large swelling in others. The role of fission products and the local oxygen content of the fuel are found to be important. The effective fuel stoichiometry is likely to change significantly during transients with substantial effects on the transport processes controlling fission gas behaviour. The results of the evaluation of the models are summarized in mechanism maps for intragranular and grain boundary behaviour. (author). 36 refs, 8 figs

  17. Fission product chemistry in severe nuclear reactor accidents

    International Nuclear Information System (INIS)

    Nichols, A.L.

    1990-09-01

    A specialist's meeting was held at JRC-Ispra from 15 to 17 January 1990 to review the current understanding of fission-product chemistry during severe accidents in light water reactors. Discussions focussed on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. Recommendations for future chemistry work were made covering the following areas: (a) fuel degradation and fission-product release, (b) transport and attenuation processes in the reactor coolant system, (c) containment chemistry (iodine behaviour and core-concrete interactions)

  18. The chemistry of fission products for accident analysis

    International Nuclear Information System (INIS)

    Potter, P.E.

    1985-01-01

    Current knowledge concerning the chemical state of the fission product elements during the development of accidents in water reactor systems is reviewed in this paper. The fission products elements which have been considered are Cs, I, Te, Sr and Ba but aspects of the behaviour of Mo, Ru and the lanthanides are also discussed. Some features of the reactions of the various species of these elements with other components of the reactor systems are described. The importance of having an adequate knowledge of thermodynamic data and phase equilibria of relatively simple systems in order to interpret experimental observations on complex multi-component systems is stressed

  19. The Phebus Fission Product and Source Term International Programmes

    International Nuclear Information System (INIS)

    Clement, B.; Zeyen, R.

    2005-01-01

    The international Phebus FP programme, initiated in 1988 is one of the major research programmes on light water reactors severe accidents. After a short description of the facility and of the test matrix, the main outcomes and results of the first four integral tests are provided and analysed. Several results were unexpected and some are of importance for safety analyses, particularly concerning fuel degradation, cladding oxidation, chemical form of some fission products, especially iodine, effect of control rod materials on degradation and chemistry, iodine behaviour in the containment. Prediction capabilities of calculation tools have largely been improved as a result of this research effort. However, significant uncertainties remain for a number of phenomena, requiring detailed physical analysis and implementation of improved models in codes, sustained by a number of separate-effect experiments. This is the subject of the new Source Term programme for a better understanding of the phenomenology on important safety issues, in accordance with priorities defined in the EURSAFE project of the 5 th European framework programme aiming at reducing the uncertainties on Source Term analyses. It covers iodine chemistry, impact of boron carbide control rods degradation and oxidation, air ingress situations and fission product release from fuel. Regarding the interpretation of Phebus, an international co-operation has been established since over ten years, particularly helpful for the improvement and common understanding of severe accident phenomena. Few months ago, the Phebus community was happy to welcome representatives of a large number of organisations from the following new European countries: the Czech republic, Hungary, Lithuania, Slovakia, Slovenia and also from Bulgaria and Romania. (author)

  20. The Influence of Material Properties on the Behaviour of Rayleigh Edge Waves in Thin Orthotropic Media

    Czech Academy of Sciences Publication Activity Database

    Červ, Jan

    2008-01-01

    Roč. 2, č. 5 (2008), s. 762-772 ISSN 1970-8734 R&D Projects: GA AV ČR(CZ) IAA200760611 Institutional research plan: CEZ:AV0Z20760514 Keywords : rayleigh edge waves * elastic orthotropic material * plane state of stress Subject RIV: BI - Acoustics