WorldWideScience

Sample records for fissionable material behaviour

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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)

  9. 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.

  10. 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

  11. 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.)

  12. 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.

  13. 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

  14. 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)

  15. 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)

  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. 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

  18. 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

  19. 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.

  20. 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.

  1. 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

  2. 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.)

  3. 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

  4. 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.

  5. 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

  6. 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.

  7. 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

  8. 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

  9. 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.

  10. 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)

  11. 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.

  12. 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)

  13. 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.

  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. 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.)

  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. 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)

  18. 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)

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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)

  5. 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)

  6. 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.

  7. 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

  8. 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)

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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)

  14. 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.

  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. 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)

  17. 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

  18. 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.

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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.

  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. 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

  7. 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

  8. 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)

  9. 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.)

  10. 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.)

  11. 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.)

  12. 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

  13. 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

  14. 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

  15. 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.

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  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. 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)

  3. 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

  4. 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.

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  7. 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

  8. 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.)

  9. 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)

  10. 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.

  11. 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

  12. 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...

  13. 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)

  14. 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

  15. 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.)

  16. 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

  17. 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

  18. 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.

  19. 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

  20. 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....

  1. GRASS-SST, Fission Products Gas Release and Fuel Swelling in Steady-State and Transients

    International Nuclear Information System (INIS)

    Zawadzki, S.

    2001-01-01

    1 - Description of program or function: GRASS-SST is a comprehensive, mechanistic model for the prediction of fission-gas behaviour in UO 2 -base fuels during steady-state and transient conditions. GRASS-SST treats fission-gas release and fuel swelling on an equal basis and simultaneously treats all major mechanisms that influence fission-gas behaviour. Models are included for intra- and inter-granular fission-gas bubble behaviour as well as a mechanistic description of the role of grain-edge inter-linked porosity on fission-gas release and swelling. GRASS-SST calculations include the effects of gas production from fissioning uranium atoms, bubble nucleation, a realistic equation of state for xenon, lattice bubble diffusivities based on experimental observations, bubble migration, bubble coalescence, re-solution, temperature and temperature gradients, inter-linked porosity, and fission-gas interaction with structural defects (dislocations and grain boundaries) on both the distribution of fission-gas within the fuel and on the amount of fission-gas released from the fuel. GRASS-SST includes the effects of the degree of nonequilibrium in the UO 2 lattice on fission-gas bubble mobility and bubble coalescence and also accounts for the observed formation of grain-surface channels. GRASS-SST also includes mechanistic models for grain-growth/grain boundary sweeping and for the behaviour of fission gas during liquefaction/dissolution and fuel melting conditions. 2 - Method of solution: A system of coupled equations for the evolution of the fission-gas bubble-size distributions in the lattice, on dislocations, on grain faces, and grain edges is derived based on the GRASS-SST models. Given a set of operating conditions, GRASS-SST calculates the bubble radii for the size classes of bubbles under consideration using a realistic equation of state for xenon as well as a generalised capillary relation. 3 - Restrictions on the complexity of the problem: Maxima of : 1 axial section

  2. 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)

  3. 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.

  4. 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)

  5. 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.)

  6. 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.)

  7. 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.

  8. 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

  9. Vaporization of low-volatile fission products under severe CANDU reactor accident conditions

    International Nuclear Information System (INIS)

    Lewis, B.J.; Corse, B.J.; Thompson, W.T.; Kaye, M.H.; Iglesias, F.C.; Elder, P.; Dickson, R.; Liu, Z.

    1997-01-01

    An analytical model has been developed to describe the release behaviour of low-volatile fission products from uranium dioxide fuel under severe reactor accident conditions. The effect of the oxygen potential on the chemical form and volatility of fission products is determined by Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix or fission product vaporization from the fuel surface. The effect of fuel volatilization (i.e., matrix stripping) on the release behaviour is also considered. The model has been compared to data from an out-of-pile annealing experiment performed in steam at the Chalk River Laboratories. (author)

  10. 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)

  11. 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

  12. 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)

  13. 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

  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. 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

  16. 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.

  17. 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

  18. 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.

  19. 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

  20. 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.

  1. 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.

  2. 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

  3. 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

  4. 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)

  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. 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

  7. 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)

  8. 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)

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

    International Nuclear Information System (INIS)

    Szuta, A.

    2003-01-01

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

  10. 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

  11. 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

  12. Fission cross section calculations for 209Bi target nucleus based on fission reaction models in high energy regions

    Directory of Open Access Journals (Sweden)

    Kaplan Abdullah

    2015-01-01

    Full Text Available Implementation of projects of new generation nuclear power plants requires the solving of material science and technological issues in developing of reactor materials. Melts of heavy metals (Pb, Bi and Pb-Bi due to their nuclear and thermophysical properties, are the candidate coolants for fast reactors and accelerator-driven systems (ADS. In this study, α, γ, p, n and 3He induced fission cross section calculations for 209Bi target nucleus at high-energy regions for (α,f, (γ,f, (p,f, (n,f and (3He,f reactions have been investigated using different fission reaction models. Mamdouh Table, Sierk, Rotating Liquid Drop and Fission Path models of theoretical fission barriers of TALYS 1.6 code have been used for the fission cross section calculations. The calculated results have been compared with the experimental data taken from the EXFOR database. TALYS 1.6 Sierk model calculations exhibit generally good agreement with the experimental measurements for all reactions used in this study.

  13. 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

  14. 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

  15. 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

  16. 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.

  17. Models for MOX fuel behaviour. A selective review

    International Nuclear Information System (INIS)

    Massih, Ali R.

    2006-01-01

    This report reviews the basic physical properties of light water reactor mixed-oxide (MOX) fuel comprising nuclear characteristics, thermal properties such as melting temperature, thermal conductivity, thermal expansion, and heat capacity, and compares these with properties of conventional UO 2 fuel. These properties are generally well understood for MOX fuel and are well described by appropriate models developed for engineering analysis. Moreover, certain modelling approaches of MOX fuel in-reactor behaviour, regarding densification, swelling, fission product gas release, helium release, fuel creep and grain growth, are evaluated and compared with the models for UO 2 . In MOX fuel the presence of plutonium rich agglomerates adds to the complexity of fuel behaviour on the micro scale. In addition, we survey the recent fuel performance experience and post irradiation examinations on several types of MOX fuel types. We discuss the data from these examinations, regarding densification, swelling, fission product gas release and the evolution of the microstructure during irradiation. The results of our review indicate that in general MOX fuel has a higher fission gas release and helium release than UO 2 fuel. Part of this increase is due to the higher operating temperatures of MOX fuel relative to UO 2 fuel due to the lower thermal conductivity of MOX material. But this effect by itself seems to be insufficient to make for the difference in the observed fission gas release of UO 2 vs. MOX fuel. Furthermore, the irradiation induced creep rate of MOX fuel is higher than that of UO 2 . This effect can reduce the pellet-clad interaction intensity in fuel rods. Finally, we suggest that certain physical based approaches discussed in the report are implemented in the fuel performance code to account for the behaviour of MOX fuel during irradiation

  18. Models for MOX fuel behaviour. A selective review

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-12-15

    This report reviews the basic physical properties of light water reactor mixed-oxide (MOX) fuel comprising nuclear characteristics, thermal properties such as melting temperature, thermal conductivity, thermal expansion, and heat capacity, and compares these with properties of conventional UO{sub 2} fuel. These properties are generally well understood for MOX fuel and are well described by appropriate models developed for engineering analysis. Moreover, certain modelling approaches of MOX fuel in-reactor behaviour, regarding densification, swelling, fission product gas release, helium release, fuel creep and grain growth, are evaluated and compared with the models for UO{sub 2}. In MOX fuel the presence of plutonium rich agglomerates adds to the complexity of fuel behaviour on the micro scale. In addition, we survey the recent fuel performance experience and post irradiation examinations on several types of MOX fuel types. We discuss the data from these examinations, regarding densification, swelling, fission product gas release and the evolution of the microstructure during irradiation. The results of our review indicate that in general MOX fuel has a higher fission gas release and helium release than UO{sub 2} fuel. Part of this increase is due to the higher operating temperatures of MOX fuel relative to UO{sub 2} fuel due to the lower thermal conductivity of MOX material. But this effect by itself seems to be insufficient to make for the difference in the observed fission gas release of UO{sub 2} vs. MOX fuel. Furthermore, the irradiation induced creep rate of MOX fuel is higher than that of UO{sub 2}. This effect can reduce the pellet-clad interaction intensity in fuel rods. Finally, we suggest that certain physical based approaches discussed in the report are implemented in the fuel performance code to account for the behaviour of MOX fuel during irradiation.

  19. 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

  20. 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

  1. 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 )

  2. 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)

  3. 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.

  4. 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)

  5. 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

  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. 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)

  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. Perspective on the fusion-fission energy concept

    International Nuclear Information System (INIS)

    Liikala, R.C.; Perry, R.T.; Teofilo, V.L.

    1978-01-01

    A concept which has potential for near-term application in the electric power sector of our energy economy is combining fusion and fission technology. The fusion-fission system, called a hybrid, is distinguished from its pure fusion counterpart by incorporation of fertile materials (uranium or thorium) in the blanket region of a fusion machine. The neutrons produced by the fusion process can be used to generate energy through fission events in the blanket or produce fuel for fission reactors through capture events in the fertile material. The performance requirements of the fusion component of hybrids is perceived as being less stringent than those for pure fusion electric power plants. The performance requirements for the fission component of hybrids is perceived as having been demonstrated or could be demonstrated with a modest investment of research and development funds. This paper presents our insights and observations of this concept in the context of why and where it might fit into the picture of meeting our future energy needs. A bibliography of hybrid research is given

  10. 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

  11. 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

  12. Correlation studies of neutron multiplicities in the 252Cf spontaneous fission

    International Nuclear Information System (INIS)

    Alkhazov, I.D.; Dmitriev, V.D.; Kovalenko, S.S.; Kuznetsov, A.V.; Malkin, L.Z.; Petrzhak, K.A.; Petrov, B.F.; Shpakov, V.I.

    1988-01-01

    Correlations between the numbers of neutrons emitted by the 252 Cf spontaneous fission fragments have been studied as a function of the fragment mass and total kinetic energy. Behaviour of the neutron number dispersions and covariances was studied for the region of symmetric fission. Parameters of the complementary fragment excitation energy distribution (mean values, dispersions, covariances) were determined. Various factors describing correlations between the complementary fragment excitation energies are considered

  13. 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

  14. 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

  15. 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.

  16. 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

  17. 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

  18. 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

  19. 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).

  20. 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.

  1. 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

  2. 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

  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. 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.

  5. Effectiveness in detecting fission fragments with ionization chambers

    International Nuclear Information System (INIS)

    Manrique Garcia, J.; Monne, G.

    1991-01-01

    Detection of fission fragments is important in nuclear measurements. When a high detection accuracy is required it is necessary to take in account the detection losses due to the absorption of fragments in the fissionable material. The losses corrections might change the final results in 2-3%. The traditional expression used in the calculation of the detection efficiency does not consider neither the density variation of the fissionable substance with its width, because it depends on the target material. That's why actually in many labs it is being searched new methods that allow to find the efficiency for each target. In this work a new method for determination of absorption efficiency is presented. The obtained results are analyzed

  6. Hybrid fission-fusion nuclear reactors

    International Nuclear Information System (INIS)

    Zucchetti, Massimo

    2011-01-01

    A fusion-fission hybrid could contribute to all components of nuclear power - fuel supply, electricity production, and waste management. The idea of the fusion-fission hybrid is many decades old. Several ideas, both new and revisited, have been investigated by hybrid proponents. These ideas appear to have attractive features, but they require various levels of advances in plasma science and fusion and nuclear technology. As a first step towards the development of hybrid reactors, fusion neutron sources can be considered as an option. Compact high-field tokamaks can be a candidate for being the neutron source in a fission-fusion hybrid, essentially due to their design characteristics, such as compact dimensions, high magnetic field, flexibility of operation. This study presents the development of a tokamak neutron source for a material testing facility using an Ignitor-based concept. The computed values show the potential of this neutron-rich device for fusion materials testing. Some full-power months of operation are sufficient to obtain relevant radiation damage values in terms of dpa. (Author)

  7. 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

  8. Nuclear fission and fission-product spectroscopy: 3. International workshop on nuclear fission and fission-product spectroscopy

    International Nuclear Information System (INIS)

    Goutte, Heloise; Fioni, Gabriele; Faust, Herbert; Goutte, Dominique

    2005-01-01

    The present book contains the proceedings of the third workshop in a series of workshops previously held in Seyssins in 1994 and 1998. The meeting was jointly organized by different divisions of CEA and two major international laboratories. In the opening address, Prof. B. Bigot, the French High Commissioner for Atomic Energy, outlined France's energy policy for the next few decades. He emphasized the continuing progress of nuclear fission in both technical and economic terms, allowing it to contribute to the energy needs of the planet even more in the future than it does today. Such progress implies a very strong link between fundamental and applied research based on experimental and theoretical approaches. The workshop gathered the different nuclear communities studying the fission process, including topics as the following: - nuclear fission experiments, - spectroscopy of neutron rich nuclei, - fission data evaluation, - theoretical aspects of nuclear fission, - and innovative nuclear systems and new facilities. The scientific program was suggested by an International Advisory Committee. About 100 scientists from 13 different countries attended the conference in the friendly working atmosphere of the Castle of Cadarache in the heart of the Provence. The proceedings of the workshop were divided into 11 sections addressing the following subject matters: 1. Cross sections and resonances (5 papers); 2. Fission at higher energies - I (5 papers); 3. Fission: mass and charge yields (4 papers); 4. Light particles and cluster emission (4 papers); 5. Spectroscopy of neutron rich nuclei (5 papers); 6. Resonances, barriers, and fission times (5 papers); 7. Fragment excitation and neutron emission (4 papers); 8. Mass and energy distributions (4 papers); 9. Needs for nuclear data and new facilities - I (4 papers); 10. Angular momenta and fission at higher Energies - II (3 papers); 11. New facilities - II (2 papers). A poster session of 8 presentations completed the workshop

  9. 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.

  10. 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.

  11. Release of fission products from irradiated aluminide fuel at high temperature

    International Nuclear Information System (INIS)

    Shibata, Toshikazu; Kanda, Keiji; Mishima, Kaichiro; Tamai, Tadaharu; Hayashi, Masatoshi; Snelgrove, James L.; Stahl, David; Matos, James E.; Travelli, Armando; Case, F. Neil; Posey, John C.

    1983-01-01

    Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel cladding material. The release of fission products from the fuel plate at temperature below 500 deg. C was found negligible. The first rapid release of fission products was observed with the occurrence of blistering at 561±1 deg. C on the plates. The next release at 585. C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 deg. C of U-Al x . The released material was mostly xenon, but small amounts of iodine and cesium were observed. (author)

  12. Release of fission products from irradiated aluminide fuel at high temperature

    International Nuclear Information System (INIS)

    Shibata, T.; Kanda, K.; Mishima, K.

    1982-01-01

    Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel-cladding material. The release of fission products from the fuel plate at temperature below 500 0 C was found negligible. The firist rapid release of fission products was observed with the occurrence of blistering at 561 +- 1 0 C on the plates. The next release at 585 0 C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 0 C of U-Al/sub x/. The released material was mostly xenon, but small amounts of iodine and cesium were observed

  13. BWR fuel clad behaviour following LOCA

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  14. 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

  15. 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)

  16. 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)

  17. Production and release of the fission gas in (Th U)O2 fuel rods

    International Nuclear Information System (INIS)

    Dias, Marcio S.

    1982-06-01

    The volume, composition and release of the fission gas products were caculated for (Th, U)O 2 fuel rods. The theorectical calculations were compared with experimental results available on the literature. In ThO 2 + 5% UO 2 fuel rods it will be produced approximated 5% more fission gas as compared to UO 2 fuel rods. The fission gas composition or Xe to Kr ratio has showed a decreasing fuel brunup dependence, in opposition to that of UO 2 . Under the same fuel rod operational conditions, the (Th, U)O 2 fission gas release will be smaller as compared to UO 2 . This behaviour of (Th, U)O 2 fuel comes from smallest gas atom difusivity and higher activation energies of the processes that increase the fission gas release. (Author) [pt

  18. Estimates of fission yields in nuclear criticality excursions

    International Nuclear Information System (INIS)

    Choi, J.S.; Thompson, J.W.; Reed, R.

    1995-06-01

    There is a need for computer simulation of hypothetical criticality excursions involving significant quantities of fissionable materials, especially in fissile aqueous system. The need arises due to the requirements for the emergency planning of facilities where the fissionable materials are handled, processed, or stored; and the regulatory requirements associated with facility operation or conversion. It is proposed here that a data base of fission yeilds for critical experiments and known accidents (both aqueous and solid) should be generated by using existing or new computer codes. The success in compiling this data base would provide useful source-terms for criticality excursions, realistic estimates of emergency-response boundary, as well as a replacement for the ''rule-of-thumb'' or ''bounding'' method. 10 refs

  19. 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

  20. 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

  1. 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

  2. Behaviour of irradiated uranium silicide fuel revisited

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  3. Neutron threshold activation detectors (TAD) for the detection of fissions

    Science.gov (United States)

    Gozani, Tsahi; Stevenson, John; King, Michael J.

    2011-10-01

    Prompt fission neutrons are one of the strongest signatures of the fission process. Depending on the fission inducing radiation, their average number ranges from 2.5 to 4 neutrons per fission. They are more energetic and abundant, by about 2 orders of magnitude, than the delayed neutrons (≈3 vs. ≈0.01) that are commonly used as indicators for the presence of fissionable materials. The detection of fission prompt neutrons, however, has to be done in the presence of extremely intense probing radiation that stimulated them. During irradiation, the fission stimulation radiation, X-rays or neutrons, overwhelms the neutron detectors and temporarily incapacitate them. Consequently, by the time the detectors recover from the source radiation, fission prompt neutrons are no longer emitted. In order to measure the prompt fission signatures under these circumstances, special measures are usually taken with the detectors such as heavy shielding with collimation, use of inefficient geometries, high pulse height bias and gamma-neutron separation via pulse-shape discrimination with an appropriate organic scintillator. These attempts to shield the detector from the flash of radiation result in a major loss of sensitivity. It can lead to a complete inability to detect the fission prompt neutrons. In order to overcome the blinding induced background from the source radiation, the detection of prompt fission neutrons needs to occur long after the fission event and after the detector has fully recovered from the source overload. A new approach to achieve this is to detect the delayed activation induced by the fission neutrons. The approach demonstrates a good sensitivity in adverse overload situations (gamma and neutron "flash") where fission prompt neutrons could normally not be detected. The new approach achieves the required temporal separation between the detection of prompt neutrons and the detector overload by the neutron activation of the detector material. The technique

  4. Neutron threshold activation detectors (TAD) for the detection of fissions

    International Nuclear Information System (INIS)

    Gozani, Tsahi; Stevenson, John; King, Michael J.

    2011-01-01

    Prompt fission neutrons are one of the strongest signatures of the fission process. Depending on the fission inducing radiation, their average number ranges from 2.5 to 4 neutrons per fission. They are more energetic and abundant, by about 2 orders of magnitude, than the delayed neutrons (∼3 vs. ∼0.01) that are commonly used as indicators for the presence of fissionable materials. The detection of fission prompt neutrons, however, has to be done in the presence of extremely intense probing radiation that stimulated them. During irradiation, the fission stimulation radiation, X-rays or neutrons, overwhelms the neutron detectors and temporarily incapacitate them. Consequently, by the time the detectors recover from the source radiation, fission prompt neutrons are no longer emitted. In order to measure the prompt fission signatures under these circumstances, special measures are usually taken with the detectors such as heavy shielding with collimation, use of inefficient geometries, high pulse height bias and gamma-neutron separation via pulse-shape discrimination with an appropriate organic scintillator. These attempts to shield the detector from the flash of radiation result in a major loss of sensitivity. It can lead to a complete inability to detect the fission prompt neutrons. In order to overcome the blinding induced background from the source radiation, the detection of prompt fission neutrons needs to occur long after the fission event and after the detector has fully recovered from the source overload. A new approach to achieve this is to detect the delayed activation induced by the fission neutrons. The approach demonstrates a good sensitivity in adverse overload situations (gamma and neutron 'flash') where fission prompt neutrons could normally not be detected. The new approach achieves the required temporal separation between the detection of prompt neutrons and the detector overload by the neutron activation of the detector material. The technique

  5. Neutron threshold activation detectors (TAD) for the detection of fissions

    Energy Technology Data Exchange (ETDEWEB)

    Gozani, Tsahi, E-mail: tgozani@rapiscansystems.com [Rapiscan Laboratories, Inc., 520 Almanor Ave., Sunnyvale, CA 94085 (United States); Stevenson, John; King, Michael J. [Rapiscan Laboratories, Inc., 520 Almanor Ave., Sunnyvale, CA 94085 (United States)

    2011-10-01

    Prompt fission neutrons are one of the strongest signatures of the fission process. Depending on the fission inducing radiation, their average number ranges from 2.5 to 4 neutrons per fission. They are more energetic and abundant, by about 2 orders of magnitude, than the delayed neutrons ({approx}3 vs. {approx}0.01) that are commonly used as indicators for the presence of fissionable materials. The detection of fission prompt neutrons, however, has to be done in the presence of extremely intense probing radiation that stimulated them. During irradiation, the fission stimulation radiation, X-rays or neutrons, overwhelms the neutron detectors and temporarily incapacitate them. Consequently, by the time the detectors recover from the source radiation, fission prompt neutrons are no longer emitted. In order to measure the prompt fission signatures under these circumstances, special measures are usually taken with the detectors such as heavy shielding with collimation, use of inefficient geometries, high pulse height bias and gamma-neutron separation via pulse-shape discrimination with an appropriate organic scintillator. These attempts to shield the detector from the flash of radiation result in a major loss of sensitivity. It can lead to a complete inability to detect the fission prompt neutrons. In order to overcome the blinding induced background from the source radiation, the detection of prompt fission neutrons needs to occur long after the fission event and after the detector has fully recovered from the source overload. A new approach to achieve this is to detect the delayed activation induced by the fission neutrons. The approach demonstrates a good sensitivity in adverse overload situations (gamma and neutron 'flash') where fission prompt neutrons could normally not be detected. The new approach achieves the required temporal separation between the detection of prompt neutrons and the detector overload by the neutron activation of the detector

  6. 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

  7. 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

  8. 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.

  9. 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

  10. Design considerations of fission and corrosion product in primary system of MONJU

    International Nuclear Information System (INIS)

    Yanagisawa, T.; Akagane, K.; Yamamoto, K.; Kawashima, K.

    1976-01-01

    General influence of fission and corrosion products in primary system on MONJU plant design is reviewed. Various research and development works are now in progress to decrease the generation rate, to remove the products more effectively and to develop the methods of evaluation the behaviour of radioactive products. The inventory and distribution of fission and corrosion products in the primary circuit of MONJU are given. The radiation levels on the primary components are estimated to be several roentgens per hour. (author)

  11. 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.

  12. 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.

  13. 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)

  14. 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

  15. 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

  16. Measurement of Fission Product Yields from Fast-Neutron Fission

    Science.gov (United States)

    Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Henderson, R.; Kenneally, J.; Macri, R.; McNabb, D.; Ryan, C.; Sheets, S.; Stoyer, M. A.; Tonchev, A. P.; Bhatia, C.; Bhike, M.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.

    2014-09-01

    One of the aims of the Stockpile Stewardship Program is a reduction of the uncertainties on fission data used for analyzing nuclear test data [1,2]. Fission products such as 147Nd are convenient for determining fission yields because of their relatively high yield per fission (about 2%) and long half-life (10.98 days). A scientific program for measuring fission product yields from 235U,238U and 239Pu targets as a function of bombarding neutron energy (0.1 to 15 MeV) is currently underway using monoenergetic neutron beams produced at the 10 MV Tandem Accelerator at TUNL. Dual-fission chambers are used to determine the rate of fission in targets during activation. Activated targets are counted in highly shielded HPGe detectors over a period of several weeks to identify decaying fission products. To date, data have been collected at neutron bombarding energies 4.6, 9.0, 14.5 and 14.8 MeV. Experimental methods and data reduction techniques are discussed, and some preliminary results are presented.

  17. Possible effects of oxidation on the transient release of fission gas from UO2

    International Nuclear Information System (INIS)

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

    1981-01-01

    The effect of varying the fuel composition from UO 2 to UOsub(2.3), on the transient behaviour of fission gas is simulated on the assumption that surface diffusion behaves in a similar manner to volume diffusion. The results may help in the understanding of fuel behaviour after pin failure in accident conditions in thermal reactor systems. (author)

  18. 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

  19. 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

  20. [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

  1. Attempt to enrich of a new spontaneous fissioning nuclide by evaporation of natural brine

    International Nuclear Information System (INIS)

    Adamek, A.; Zhuravleva, E.L.; Constantinescu, M.; Constantinescu, o.; Chuburkov, Yu.T.

    1983-01-01

    The enrichment of the new spontaneous fissioning nuclide discovered in the Cheleken brine, was made by evaporation. The purpose of this work was the comparison of behaviour of the new spontaneous fissioning nuclide with that of the known elements in the formation processes of the high concentration brines. Spontaneous fission of the nuclide was measured by means of the counters for multiple emission of neutrons. It is shown that the new spontaneous fissioning nuclide was enriched as well as other trace elements (Hg, Tl, Bi and Pb) in a solution remained after the evaporation of the initial solution. The conclusion is drawn that from the sea water brines could be obtained by evaporation which are enriched in trace elements with an enrichment degree higher than the natural brines

  2. 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

  3. 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

  4. Summarizing evaluation of the results of in-pile experiments for the transient fission gas release under accidental conditions of fast breeders

    International Nuclear Information System (INIS)

    Fischer, E.A.; Vaeth, L.

    1989-04-01

    The transient fission gas behaviour and the fission gas induced fuel motion were studied in in-pile experiments in different countries, under conditions typical for hypothetical accidents. This report summarizes first the different experiment series and the main results. Then, a comparative evaluation is given, which provides a basis for the choice of the fission gas parameters in the accident code SAS3D

  5. 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

  6. 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

  7. 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)

  8. 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

  9. 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

  10. 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

  11. 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)

  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. 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

  14. Challenges of structural materials for innovative nuclear systems in Europe

    International Nuclear Information System (INIS)

    Serrano, M.; Gomez-Briceno, D.

    2009-01-01

    New fusion and fission reactors for generation IV are envisaged to operate at conditions well above the actual ones for commercial fission reactors. This type of reactor combined a high operation temperature with a high neutron dose and an aggressive coolant, which imply new challenges for structural materials. One of the key issues to assure the safety and feasibility of these new nuclear systems is the selection of the structural materials, especially for in core components. Beside the differences between them, especially the amount of transmutation He in fusion reactors, similar structural materials have been selected. Some of the selected materials are well characterized at least at medium temperatures, as conventional ferritic/martensitic steels, but the qualification for higher temperatures is needed. For other materials, as ODS steels, there is a need for a complete characterization and qualification. In this paper a review of the operating conditions and selected structural materials for generation IV and fusion reactors within Europe is made. The needs for a complete characterization of these candidate materials are identified in terms of high temperature behaviour, radiation damage and coolant compatibility. (author)

  15. 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

  16. Preparation of a primary target for the production of fission products in a nuclear reactor

    International Nuclear Information System (INIS)

    Arino, H.; Cosolito, F.J.; George, K.D.; Thornton, A.K.

    1976-01-01

    A primary target for the production of fission products in a nuclear reactor, such as uranium or plutonium fission products, is comprised of an enclosed, cylindrical vessel, preferably comprised of stainless steel, having a thin, continuous, uniform layer of fissionable material, integrally bonded to its inner walls and a port permitting access to the interior of the vessel. A process is also provided for depositing uranium material on to the inner walls of the vessel. Upon irradiation of the target with neutrons from a nuclear reactor, radioactive fission products, such as molybdenum-99, are formed, and thereafter separated from the target by the introduction of an acidic solution through the port to dissolve the irradiated inner layer. The irradiation and dissolution are thus effected in the same vessel without the necessity of transferring the fissionable material and fission products to a separate chemical reactor. Subsequently, the desired isotopes are extracted and purified. Molybdenum-99 decays to technetium-99m which is a valuable medical diagnostic radioisotope. 3 claims, 1 drawing figure

  17. 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.

  18. 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.

  19. 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

  20. To follow or not? How animals in fusion-fission societies handle conflicting information during group decision-making.

    Science.gov (United States)

    Merkle, Jerod A; Sigaud, Marie; Fortin, Daniel

    2015-08-01

    When group members possess differing information about the environment, they may disagree on the best movement decision. Such conflicts result in group break-ups, and are therefore a fundamental driver of fusion-fission group dynamics. Yet, a paucity of empirical work hampers our understanding of how adaptive evolution has shaped plasticity in collective behaviours that promote and maintain fusion-fission dynamics. Using movement data from GPS-collared bison, we found that individuals constantly associated with other animals possessing different spatial knowledge, and both personal and conspecific information influenced an individual's patch choice decisions. During conflict situations, bison used group familiarity coupled with their knowledge of local foraging options and recently sampled resource quality when deciding to follow or leave a group - a tactic that led to energy-rewarding movements. Natural selection has shaped collective behaviours for coping with social conflicts and resource heterogeneity, which maintain fusion-fission dynamics and play an essential role in animal distribution. © 2015 John Wiley & Sons Ltd/CNRS.

  1. A general evaluation of the irradiation behaviour of dispersion fuels

    International Nuclear Information System (INIS)

    Hofman, G.L.

    1995-01-01

    The irradiation behaviour of aluminum-based dispersion fuels is evaluated with emphasis on metallurgical processes that control the dispersion behaviour. Phase transformations and microstructural changes resulting from fuel-matrix interactions and the effect of fissioning in fuel are discussed. (author)

  2. 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

  3. 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

  4. 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.

  5. 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...

  6. Singlet Fission in Rubrene Derivatives: Impact of Molecular Packing

    KAUST Repository

    Sutton, Christopher

    2017-03-13

    We examine the properties of six recently synthesized rubrene derivatives (with substitutions on the side phenyl rings) that show vastly different crystal structures. In order to understand how packing in the solid state affects the excited states and couplings relevant for singlet fission, the lowest excited singlet (S), triplet (T), multiexciton (TT), and charge-transfer (CT) states of the rubrene derivatives are compared to known singlet fission materials [tetracene, pentacene, 5,12-diphenyltetracene (DPT), and rubrene itself]. While a small difference of less than 0.2 eV is calculated for the S and TT energies, a range of 0.50 to 1.2 eV in the CT energies and nearly 3 orders of magnitude in the electronic couplings are computed for the rubrene derivatives in their crystalline packings, which strongly affects the role of the CT state in facilitating SF. To rationalize experimental observations of singlet fission occurring in amorphous phases of rubrene, DPT, and tetracene, we use molecular dynamics (MD) simulations to assess the impact of molecular packing and orientations and to gain a better understanding of the parameters that control singlet fission in amorphous films compared to crystalline packings. The MD simulations point to a crystalline-like packing for thin films of tetracene; on the other hand, DPT, rubrene, and the rubrene derivatives all show various degrees of disorder with a number of sites that have larger electronic couplings than in the crystal, which can facilitate singlet fission in such thin films. Our analysis underlines the potential of these materials as promising candidates for singlet fission and helps understand how various structural motifs affect the critical parameters that control the ability of a system to undergo singlet fission.

  7. Post-scission fission theory: Neutron emission in fission

    International Nuclear Information System (INIS)

    Madland, D.G.

    1997-01-01

    A survey of theoretical representations of two of the observables in neutron emission in fission is given, namely, the prompt fission neutron spectrum N (E) and the average prompt neutron multiplicity bar ν p . Early representations of the two observables are presented and their deficiencies are discussed. This is followed by summaries and examples of recent theoretical models for the calculation of these quantities. Emphasis is placed upon the predictability and accuracy of the recent models. In particular, the dependencies of N (E) and bar ν p upon the fissioning nucleus and its excitation energy are treated. Recent work in the calculation of the prompt fission neutron spectrum matrix N (E, E n ), where E n is the energy of the neutron inducing fission, is then discussed. Concluding remarks address the current status of our ability to calculate these observables with confidence, the direction of future theoretical efforts, and limitations to current (and future) approaches

  8. 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.)

  9. 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)

  10. UV photolysis of 4-iodo-, 4-bromo-, and 4-chlorophenol: Competition between C-Y (Y = halogen) and O-H bond fission

    Science.gov (United States)

    Sage, Alan G.; Oliver, Thomas A. A.; King, Graeme A.; Murdock, Daniel; Harvey, Jeremy N.; Ashfold, Michael N. R.

    2013-04-01

    The wavelength dependences of C-Y and O-H bond fission following ultraviolet photoexcitation of 4-halophenols (4-YPhOH) have been investigated using a combination of velocity map imaging, H Rydberg atom photofragment translational spectroscopy, and high level spin-orbit resolved electronic structure calculations, revealing a systematic evolution in fragmentation behaviour across the series Y = I, Br, Cl (and F). All undergo O-H bond fission following excitation at wavelengths λ ≲ 240 nm, on repulsive ((n/π)σ*) potential energy surfaces (PESs), yielding fast H atoms with mean kinetic energies ˜11 000 cm-1. For Y = I and Br, this process occurs in competition with prompt C-I and C-Br bond cleavage on another (n/π)σ* PES, but no Cl/Cl* products unambiguously attributable to one photon induced C-Cl bond fission are observed from 4-ClPhOH. Differences in fragmentation behaviour at longer excitation wavelengths are more marked. Prompt C-I bond fission is observed following excitation of 4-IPhOH at all λ ≤ 330 nm; the wavelength dependent trends in I/I* product branching ratio, kinetic energy release, and recoil anisotropy suggest that (with regard to C-I bond fission) 4-IPhOH behaves like a mildly perturbed iodobenzene. Br atoms are observed when exciting 4-BrPhOH at long wavelengths also, but their velocity distributions suggest that dissociation occurs after internal conversion to the ground state. O-H bond fission, by tunnelling (as in phenol), is observed only in the cases of 4-FPhOH and, more weakly, 4-ClPhOH. These observed differences in behaviour can be understood given due recognition of (i) the differences in the vertical excitation energies of the C-Y centred (n/π)σ* potentials across the series Y = I increased spin-orbit coupling in, particularly, 4-IPhOH. The present results provide (another) reminder of the risks inherent in extrapolating photochemical behaviour measured for one molecule at one wavelength to other (related) molecules and to

  11. 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.

  12. 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.)

  13. 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)

  14. Microscopic theory of singlet exciton fission. III. Crystalline pentacene

    International Nuclear Information System (INIS)

    Berkelbach, Timothy C.; Reichman, David R.; Hybertsen, Mark S.

    2014-01-01

    We extend our previous work on singlet exciton fission in isolated dimers to the case of crystalline materials, focusing on pentacene as a canonical and concrete example. We discuss the proper interpretation of the character of low-lying excited states of relevance to singlet fission. In particular, we consider a variety of metrics for measuring charge-transfer character, conclusively demonstrating significant charge-transfer character in the low-lying excited states. The impact of this electronic structure on the subsequent singlet fission dynamics is assessed by performing real-time master-equation calculations involving hundreds of quantum states. We make direct comparisons with experimental absorption spectra and singlet fission rates, finding good quantitative agreement in both cases, and we discuss the mechanistic distinctions that exist between small isolated aggregates and bulk systems

  15. Microscopic theory of singlet exciton fission. III. Crystalline pentacene

    Energy Technology Data Exchange (ETDEWEB)

    Berkelbach, Timothy C., E-mail: tcb2112@columbia.edu; Reichman, David R., E-mail: drr2103@columbia.edu [Department of Chemistry, Columbia University, 3000 Broadway, New York, New York 10027 (United States); Hybertsen, Mark S., E-mail: mhyberts@bnl.gov [Center for Functional Nanomaterials, Brookhaven National Laboratory, Upton, New York 11973-5000 (United States)

    2014-08-21

    We extend our previous work on singlet exciton fission in isolated dimers to the case of crystalline materials, focusing on pentacene as a canonical and concrete example. We discuss the proper interpretation of the character of low-lying excited states of relevance to singlet fission. In particular, we consider a variety of metrics for measuring charge-transfer character, conclusively demonstrating significant charge-transfer character in the low-lying excited states. The impact of this electronic structure on the subsequent singlet fission dynamics is assessed by performing real-time master-equation calculations involving hundreds of quantum states. We make direct comparisons with experimental absorption spectra and singlet fission rates, finding good quantitative agreement in both cases, and we discuss the mechanistic distinctions that exist between small isolated aggregates and bulk systems.

  16. Evaluations of fusion-fission (hybrid) concepts: market penetration analysis for fusion-fission hybrids. Part A

    International Nuclear Information System (INIS)

    Engel, R.L.; Deonigi, D.E.

    1976-01-01

    This report summarizes findings of the fusion-fission studies conducted for the Electric Power Research Institute by Battelle, Pacific Northwest Laboratories. This particular study focused on the evaluation of fissile material producing hybrids. Technical results of the evaluation of actinide burning are presented in a companion volume, Part B

  17. Diffusion in cladding materials

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  18. The role of grain boundary fission gases in high burn-up fuel under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Lemoine, F.; Papin, J.; Frizonnet, J.M.; Cazalis, B.; Rigat, H.

    2002-01-01

    In the frame of reactivity-initiated accidents (RIA) studies, the CABRI REP-Na programme is currently performed, focused on high burn-up UO 2 and MOX fuel behaviour. From 1993 to 1998, seven tests were performed with UO 2 fuel and three with MOX fuel. In all these tests, particular attention has been devoted to the role of fission gases in transient fuel behaviour and in clad loading mechanisms. From the analysis of experimental results, some basic phenomena were identified and a better understanding of the transient fission gas behaviour was obtained in relation to the fuel and clad thermo-mechanical evolution in RIA, but also to the initial state of the fuel before the transient. A high burn-up effect linked to the increasing part of grain boundary gases is clearly evidenced in the final gas release, which would also significantly contribute to the clad loading mechanisms. (authors)

  19. 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...

  20. 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)

  1. Fission-product yields for thermal-neutron fission of curium-243

    International Nuclear Information System (INIS)

    Breederland, D.G.

    1982-01-01

    Cumulative fission yields for 25 gamma rays emitted during the decay of 23 fission products produced by thermal-neutron fission of 243 Cm have been determined. Using Ge(Li) spectroscopy, 33 successive pulse-height spectra of gamma rays emitted from a 77-ng sample of 243 Cm over a period of approximately two and one-half months were analyzed. Reduction of these spectra resulted in the identification and matching of gamma-ray energies and half-lives to specific radionuclides. Using these results, 23 cumulative fission-product yields were calculated. Only those radionuclides having half-lives between 6 hours and 65 days were observed. Prior to this experiment, no fission-product yields had been recorded for 243 Cm

  2. Unit mechanisms of fission gas release: Current understanding and future needs

    Science.gov (United States)

    Tonks, Michael; Andersson, David; Devanathan, Ram; Dubourg, Roland; El-Azab, Anter; Freyss, Michel; Iglesias, Fernando; Kulacsy, Katalin; Pastore, Giovanni; Phillpot, Simon R.; Welland, Michael

    2018-06-01

    Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. This basic understanding of the fission gas behavior mechanisms has the potential to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.

  3. Neutron emission as a probe of fusion-fission and quasi-fission dynamics

    International Nuclear Information System (INIS)

    Hinde, D.J.

    1991-01-01

    Pre- and post scission neutron yeilds have been measured as a function of projectile mass, compound nucleus fissility, and fission mass-split and total kinetic energy (TKE) for 27 fusion-fission and quasi-fission reactions induced by beams of 16,18 O, 40 Ar and 64 Ni. A new method of interpretation of experimental pre-scission neutron multiplicities ν-pre and mean kinetic energies ε ν allows the extraction of fission time scales with much less uncertainty than previously, all fusion-fission results being consistent with a dynamical time scale of (35±15) x 10 -21 s for symmetric fission. All reactions show that ν-pre falls quite rapidly with increasing mass-asymmetry; evidence is presented that for fusion-fission reactions this is partly due to a reduction of the dynamical fission time scale with mass-asymmetry. For quasi-fission, the data indicate that the pre-scission multiplicity and mean neutron kinetic energy are very sensitive to the final mass-asymmetry, but that the time scale is virtually independent of mass-asymmetry. It is concluded that for fusion-fission there is no dependence of ν-pre on TKE, whilst for 64 Ni-induced quasi-fission reactions, a strong increase of ν-pre with decreasing TKE is observed, probably largely caused by neutron emission during the acceleration time of the fission fragments in these fast reactions. Interpretation of post-scission multiplicities in terms of fragment excitation energies leads to deduced time scales consistent with those determined from the pre-scission data. 54 refs., 17 tabs., 25 figs

  4. 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)

  5. 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

  6. 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

  7. Fission yield covariance generation and uncertainty propagation through fission pulse decay heat calculation

    International Nuclear Information System (INIS)

    Fiorito, L.; Diez, C.J.; Cabellos, O.; Stankovskiy, A.; Van den Eynde, G.; Labeau, P.E.

    2014-01-01

    Highlights: • Fission yield data and uncertainty comparison between major nuclear data libraries. • Fission yield covariance generation through Bayesian technique. • Study of the effect of fission yield correlations on decay heat calculations. • Covariance information contribute to reduce fission pulse decay heat uncertainty. - Abstract: Fission product yields are fundamental parameters in burnup/activation calculations and the impact of their uncertainties was widely studied in the past. Evaluations of these uncertainties were released, still without covariance data. Therefore, the nuclear community expressed the need of full fission yield covariance matrices to be able to produce inventory calculation results that take into account the complete uncertainty data. State-of-the-art fission yield data and methodologies for fission yield covariance generation were researched in this work. Covariance matrices were generated and compared to the original data stored in the library. Then, we focused on the effect of fission yield covariance information on fission pulse decay heat results for thermal fission of 235 U. Calculations were carried out using different libraries and codes (ACAB and ALEPH-2) after introducing the new covariance values. Results were compared with those obtained with the uncertainty data currently provided by the libraries. The uncertainty quantification was performed first with Monte Carlo sampling and then compared with linear perturbation. Indeed, correlations between fission yields strongly affect the uncertainty of decay heat. Eventually, a sensitivity analysis of fission product yields to fission pulse decay heat was performed in order to provide a full set of the most sensitive nuclides for such a calculation

  8. 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...

  9. The changing face of fission track dating: recent advances

    International Nuclear Information System (INIS)

    Green, P.F.; Duddy, I.R.; Gleadow, A.J.W.; Laslett, G.M.; Hegarty, K.A.; Lovering, J.F.

    1985-01-01

    Fission Track Dating has undergone something of a renaissance in recent years. What began as a rather unreliable method of geochronology has now become a rigorous and dependable means of measuring not only geological time but also paleotemperatures. Developments such as the zeta calibration technique, rigorous satistical data analysis, investigation of confined track lengths and detailed studies of annealing behaviour have all contributed to advances in the technique

  10. Computer program FPIP-REV calculates fission product inventory for U-235 fission

    Science.gov (United States)

    Brown, W. S.; Call, D. W.

    1967-01-01

    Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

  11. Nuclear fission and nuclear safeguards: Common technologies and challenges

    International Nuclear Information System (INIS)

    Keepin, G.R.

    1989-01-01

    Nuclear fission and nuclear safeguards have much in common, including the basic physical phenomena and technologies involved as well as the commitments and challenges posed by expanding nuclear programs in many countries around the world. The unique characteristics of the fission process -- such as prompt and delayed neutron and gamma ray emission -- not only provide the means of sustaining and controlling the fission chain reaction, but also provide unique ''signatures'' that are essential to quantitative measurement and effective safeguarding of key nuclear materials (notably 239 Pu and 235 U) against theft, loss, or diversion. In this paper, we trace briefly the historical emergence of safeguards as an essential component of the expansion of the nuclear enterprise worldwide. We then survey the major categories of passive and active nondestructive assay techniques that are currently in use or under development for rapid, accurate measurement and verification of safe-guarded nuclear materials in the many forms in which they occur throughout the nuclear fuel cycle. 23 refs., 14 figs

  12. Characteristics of diallyl phthalate resin as a fission track detector

    CERN Document Server

    Tsuruta, T

    1999-01-01

    Diallyl phthalate (DAP) resin plates were irradiated with fission fragments, and then etched in aqueous solution of KOH. Etched tracks were observed and counted by using an optical microscope. The detection efficiency of fission fragments was about 100% for both perpendicular and random incidence. DAP plates were insensitive to alpha particles and fast neutrons. These characteristics are suitable for detecting selected fission fragments, which coexist with alpha particles or fast neutrons. DAP plates are valuable for quantitative analysis of fissionable materials and neutron dosimetry. DAP and allyl diglycol carbonate (CR-39) were formed into copolymers in various ratios. The copolymers showed intermediate characteristics between DAP and CR-39. The fabrication of the copolymers made it possible to control the discrimination level for detection of heavy charged particles.

  13. Gas-phase transport of fission products

    International Nuclear Information System (INIS)

    Tang, I.N.; Munkelwitz, H.R.

    1982-01-01

    The paper presents the results of an experimental investigation to show the importance of nuclear aerosol formation as a mechanism for semi-volatile fission product transport under certain postulated HTGR accident conditions. Simulated fission product Sr and Ba as oxides are impregnated in H451 graphite and released at elevated temperatures into a dry helium flow. In the presence of graphite, the oxides are quantitatively reduced to metals, which subsequently vaporize at temperatures much lower than required for the oxides alone to vaporize in the absence of graphite. A substantial fraction of the released material is associated with particulate matter, which is collected on filters located downstream at ambient temperatures. Increasing carrier-gas flow rate greatly enhances the extent of particulate transport. The release and transport of simulated fission product Ag as metal are also investigated. Electron microscopic examinations of the collected Sr and Ag aerosols show large agglomerates composed of primary particles roughly 0.06 to 0.08 μm in diameter

  14. 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

  15. 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

  16. 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

  17. Methodology and application of the WIMS-D4M fission product data

    International Nuclear Information System (INIS)

    Mo, S.C.

    1995-01-01

    The WIMS-D4 code has been modified (WIMS-D4m) to generate burn-up dependent microscopic cross sections for use in full core depletion calculations. The calculation of neutron absorption by fission products can be obtained from a reduced fission-product-chain model that includes the 135 Xe and 149 Sm chains, and a lumped fission product to account for the absorption by fission products not explicitly treated. Burn-up calculations were performed for the ANS MEU core using WIMS and EPRI-CELL cross sections. The calculated eigenvalues and material loadings are in good agreements

  18. 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

  19. Studies of Fission-Induced Surface Damage in Actinides Using Ultracold Neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Broussard, Leah J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-03-05

    This report describes the results of the fission-induced actinide studies at LANL. Previously, there was no fission data at these energies though there were initial characterizations of UCN energy dependence and material thickness. The proof of principle was demonstrated and the initial characterizations of sputtered rates, angular and size distribution are underway.

  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. Fission fragment angular distributions and fission cross section validation

    International Nuclear Information System (INIS)

    Leong, Lou Sai

    2013-01-01

    The present knowledge of angular distributions of neutron-induced fission is limited to a maximal energy of 15 MeV, with large discrepancies around 14 MeV. Only 238 U and 232 Th have been investigated up to 100 MeV in a single experiment. The n-TOF Collaboration performed the fission cross section measurement of several actinides ( 232 Th, 235 U, 238 U, 234 U, 237 Np) at the n-TOF facility using an experimental set-up made of Parallel Plate Avalanche Counters (PPAC), extending the energy domain of the incident neutron above hundreds of MeV. The method based on the detection of the 2 fragments in coincidence allowed to clearly disentangle the fission reactions among other types of reactions occurring in the spallation domain. I will show the methods we used to reconstruct the full angular resolution by the tracking of fission fragments. Below 10 MeV our results are consistent with existing data. For example in the case of 232 Th, below 10 MeV the results show clearly the variation occurring at the first (1 MeV) and second (7 MeV) chance fission, corresponding to transition states of given J and K (total spin and its projection on the fission axis), and a much more accurate energy dependence at the 3. chance threshold (14 MeV) has been obtained. In the spallation domain, above 30 MeV we confirm the high anisotropy revealed in 232 Th by the single existing data set. I'll discuss the implications of this finding, related to the low anisotropy exhibited in proton-induced fission. I also explore the critical experiments which is valuable checks of nuclear data. The 237 Np neutron-induced fission cross section has recently been measured in a large energy range (from eV to GeV) at the n-TOF facility at CERN. When compared to previous measurements, the n-TOF fission cross section appears to be higher by 5-7 % beyond the fission threshold. To check the relevance of n-TOF data, we simulate a criticality experiment performed at Los Alamos with a 6 kg sphere of 237 Np. This

  2. Robust singlet fission in pentacene thin films with tuned charge transfer interactions.

    Science.gov (United States)

    Broch, K; Dieterle, J; Branchi, F; Hestand, N J; Olivier, Y; Tamura, H; Cruz, C; Nichols, V M; Hinderhofer, A; Beljonne, D; Spano, F C; Cerullo, G; Bardeen, C J; Schreiber, F

    2018-03-05

    Singlet fission, the spin-allowed photophysical process converting an excited singlet state into two triplet states, has attracted significant attention for device applications. Research so far has focused mainly on the understanding of singlet fission in pure materials, yet blends offer the promise of a controlled tuning of intermolecular interactions, impacting singlet fission efficiencies. Here we report a study of singlet fission in mixtures of pentacene with weakly interacting spacer molecules. Comparison of experimentally determined stationary optical properties and theoretical calculations indicates a reduction of charge-transfer interactions between pentacene molecules with increasing spacer molecule fraction. Theory predicts that the reduced interactions slow down singlet fission in these blends, but surprisingly we find that singlet fission occurs on a timescale comparable to that in pure crystalline pentacene. We explain the observed robustness of singlet fission in such mixed films by a mechanism of exciton diffusion to hot spots with closer intermolecular spacings.

  3. COCOSYS analysis for deposition of aerosols and fission products in PHEBUS FPT-2 containment

    International Nuclear Information System (INIS)

    Kontautas, A.; Babilas, E.; Urbonavičius, E.

    2012-01-01

    Highlights: ► Aerosol and fission product behaviour in containment is analyzed. ► Lumped-parameter code COCOSYS is used for the analysis. ► Detailed description of COCOSYS model nodalisation is presented. - Abstract: The issue of the source term of radioactive fission products release from the nuclear power plants to the environment is not resolved yet. Even though experiments are performed and many analyses are performed using different computer codes some questions remain unresolved. The analyses of aerosol transport and deposition processes in the containments of nuclear power plants are investigated for a long time and computer codes are more advanced than 20 years ago there is not developed generic methodology how to develop nodalisation for the lumped-parameter codes. The validation of the computer codes is also an issue. The PHEBUS FP experiments provide possibility for an extensive validation of the computer codes and assessment of different methods to develop nodalisation of the containment.This paper presents results of analysis of aerosol and fission product behaviour in PHEBUS FPT-2 test. It includes description of the PHEBUS containment, detailed description of nodalisation with the initial and boundary conditions used in the analysis and extensive comparison of calculated and measured results. Lumped-parameter code COCOSYS was used for the analysis. The calculated thermal-hydraulic results are in good agreement with measured, which ensures good basis for analysis of aerosol and fission product transport and deposition. The calculated airborn aerosol and fission product masses are in good agreement with measured as well. The aerosol deposition distribution shows that the calculated diffusive deposition on the external containment walls is lower than measured and that the diffusive deposition model implemented in COCOSYS code could not explain this result and further investigations are needed.

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  5. Fission-fragment attachment to aerosols and their transport through capillary tubes

    International Nuclear Information System (INIS)

    Novick, V.J.; Alvarez, J.L.; Greenwood, R.C.

    1981-01-01

    The transport of radioactive aerosols was studied using equipment, collectively called the Helium jet, that has been constructed to provide basic nuclear physics data on fission product nuclides. The transport of the fission products in the system depends on their attachment to aerosol particles. The system consists of 1) a tube furnace which generates aerosols by the sublimation or evaporation of source material, 2) a helium stream used to transport the aerosols, 3) a 25 m settling tube to eliminate the larger aerosols and smaller aerosols that would deposit in the capillary, 4) a Californium-252 self-fissioning source of fission product nuclides, and 5) a small capillary to carry the radioactive aerosols from the hot cell to the laboratory. Different source materials were aerosolized but NaCl is generally used because it yielded the highest transport efficiencies through the capillary. Particle size measurments were made with NaCl aerosols by using a cascade impactor, an optical light scattering device, and the capillary itself as a diffusion battery by performing radiation measurements and/or electrical conductivity measurements. Both radioactive and nonradioactive aerosols were measured in order to investigate the possibility of a preferential size range for fission product attachment. The measured size distributions were then used to calculate attachment coefficients and finally an attachment time

  6. 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)

  7. 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)

  8. Measurement of prompt fission gamma-ray spectra in fast neutron-induced fission

    International Nuclear Information System (INIS)

    Laborie, J.M.; Belier, G.; Taieb, J.

    2012-01-01

    Knowledge of prompt fission gamma-ray emission has been of major interest in reactor physics for a few years. Since very few experimental spectra were ever published until now, new measurements would be also valuable to improve our understanding of the fission process. An experimental method is currently being developed to measure the prompt fission gamma-ray spectrum from some tens keV up to 10 MeV at least. The mean multiplicity and total energy could be deduced. In this method, the gamma-rays are measured with a bismuth germanate (BGO) detector which has the advantage to present a high P/T ratio and a high efficiency compared to other gamma-ray detectors. The prompt fission neutrons are rejected by the time of flight technique between the BGO detector and a fission trigger given by a fission chamber or a scintillating active target. Energy and efficiency calibration of the BGO detector were carried out up to 10.76 MeV by means of the Al-27(p, gamma) reaction. First prompt fission gamma-ray spectrum measurements performed for the spontaneous fission of Cf-252 and for 1.7 and 15.6 MeV neutron-induced fission of U-238 at the CEA, DAM, DIF Van de Graaff accelerator, will be presented. (authors)

  9. Fission product release from UO2 during irradiation. Diffusion data and their application to reactor fuel pins

    International Nuclear Information System (INIS)

    Findlay, J.R.; Johnson, F.A.; Turnbull, J.A.; Friskney, C.A.

    1980-01-01

    Release of fission product species from UO 2 , and to a limited extent from (U, Pu)0 2 was studied using small scale in-reactor experiments in which these interacting variables may be separated, as far as is possible, and their influences assessed. Experiments were at fuel ratings appropriate to water reactor fuel elements and both single crystal and poly-crystalline specimens were used. They employed highly enriched uranium such that the relative number of fissions occurring in plutonium formed by neutron capture was small. The surface to volume ratio (S/V) of the specimens was well defined thus reducing the uncertainties in the derivation of diffusion coefficients. These experiments demonstrate many of the important characteristics of fission product behaviour in UO 2 during irradiation. The samples used for these experiments were small being always less than 1g with a fissile content usually between 2 and 5mg. Polycrystalline materials were taken from batches of production fuel prepared by conventional pressing and sintering techniques. The enriched single crystals were grown from a melt of sodium and potassium chloride doped with UO 2 powder 20% 235 U content. The irradiations were performed in the DIDO reactor at Harwell. The neutron flux at the specimen was 4x10 16 neutrons m -2 s -1 providing a heat rating within the samples of 34.5 MW/teU

  10. Measurements of fission cross-sections and of neutron production rates

    International Nuclear Information System (INIS)

    Billaud, P.; Clair, C.; Gaudin, M.; Genin, R.; Joly, R.; Leroy, J.L.; Michaudon, A.; Ouvry, J.; Signarbieux, C.; Vendryes, G.

    1958-01-01

    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 10 B layers, was used as the neutron flux monitor. b) Measurement of the fission cross section of 235 U. We intend to measure the variation of the neutron induced fission cross section of 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 235 U (20 deposits of 25 cm diameter). The relative fission data were corrected for the neutron spectrum measured with a set of BF 3 proportional counters. c) Mean number ν of neutrons emitted in neutron induced fission. We measured the value of ν for several fissile isotopes in the case of fission induced by 14 MeV neutrons. The 14 MeV neutrons were produced by D (t, n) α reaction by means of a 300 kV Cockcroft Walton generator. (author) [fr

  11. 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

  12. 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.)

  13. Metallic fission product releases from HTR-spherical fuel elements

    International Nuclear Information System (INIS)

    Helmbold, M.; Amian, W.; Stoever, D.; Hecker, R.

    1978-01-01

    Fission product releases from fuel determines to a large extent the feasibility of a special reactor concept. Basic data describing the diffusion behaviour from coated particle fuel are presented concerning isotopes Cs 137 , Sr 90 and Agsup(110m). Taking into account these data for typical 3000MWth plants release calculations are performed. Sensitive release parameters could be defined and the results show low release figures for all the considered reactor concepts. (author)

  14. 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

  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. 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

  17. A method to evaluate fission gas release during irradiation testing of spherical fuel - HTR2008-58184

    International Nuclear Information System (INIS)

    Van Der Merwet, H.; Venter, J.

    2008-01-01

    The evaluation of fission gas release from spherical fuel during irradiation testing is critical to understand expected fuel performance under real reactor conditions. Online measurements of Krypton and Xenon fission products explain coated particle performance and contributions from graphitic matrix materials used in fuel manufacture and irradiation rig materials. Methods that are being developed to accurately evaluate fission gas release are described here together with examples of evaluations performed on irradiation tests HFR-K5, -K6 and EU1bis. (authors)

  18. UV photolysis of 4-iodo-, 4-bromo-, and 4-chlorophenol: Competition between C–Y (Y = halogen) and O–H bond fission

    International Nuclear Information System (INIS)

    Sage, Alan G.; Oliver, Thomas A. A.; King, Graeme A.; Murdock, Daniel; Harvey, Jeremy N.; Ashfold, Michael N. R.

    2013-01-01

    The wavelength dependences of C–Y and O–H bond fission following ultraviolet photoexcitation of 4-halophenols (4-YPhOH) have been investigated using a combination of velocity map imaging, H Rydberg atom photofragment translational spectroscopy, and high level spin-orbit resolved electronic structure calculations, revealing a systematic evolution in fragmentation behaviour across the series Y = I, Br, Cl (and F). All undergo O–H bond fission following excitation at wavelengths λ≲ 240 nm, on repulsive ((n/π)σ*) potential energy surfaces (PESs), yielding fast H atoms with mean kinetic energies ∼11 000 cm −1 . For Y = I and Br, this process occurs in competition with prompt C–I and C–Br bond cleavage on another (n/π)σ* PES, but no Cl/Cl* products unambiguously attributable to one photon induced C–Cl bond fission are observed from 4-ClPhOH. Differences in fragmentation behaviour at longer excitation wavelengths are more marked. Prompt C–I bond fission is observed following excitation of 4-IPhOH at all λ≤ 330 nm; the wavelength dependent trends in I/I* product branching ratio, kinetic energy release, and recoil anisotropy suggest that (with regard to C–I bond fission) 4-IPhOH behaves like a mildly perturbed iodobenzene. Br atoms are observed when exciting 4-BrPhOH at long wavelengths also, but their velocity distributions suggest that dissociation occurs after internal conversion to the ground state. O–H bond fission, by tunnelling (as in phenol), is observed only in the cases of 4-FPhOH and, more weakly, 4-ClPhOH. These observed differences in behaviour can be understood given due recognition of (i) the differences in the vertical excitation energies of the C–Y centred (n/π)σ* potentials across the series Y = I < Br < Cl and the concomitant reduction in C–Y bond strength, cf. that of the rival O–H bond, and (ii) the much increased spin-orbit coupling in, particularly, 4-IPhOH. The present results provide (another) reminder of the

  19. UV photolysis of 4-iodo-, 4-bromo-, and 4-chlorophenol: competition between C-Y (Y = halogen) and O-H bond fission.

    Science.gov (United States)

    Sage, Alan G; Oliver, Thomas A A; King, Graeme A; Murdock, Daniel; Harvey, Jeremy N; Ashfold, Michael N R

    2013-04-28

    The wavelength dependences of C-Y and O-H bond fission following ultraviolet photoexcitation of 4-halophenols (4-YPhOH) have been investigated using a combination of velocity map imaging, H Rydberg atom photofragment translational spectroscopy, and high level spin-orbit resolved electronic structure calculations, revealing a systematic evolution in fragmentation behaviour across the series Y = I, Br, Cl (and F). All undergo O-H bond fission following excitation at wavelengths λ ≲ 240 nm, on repulsive ((n∕π)σ∗) potential energy surfaces (PESs), yielding fast H atoms with mean kinetic energies ∼11,000 cm(-1). For Y = I and Br, this process occurs in competition with prompt C-I and C-Br bond cleavage on another (n∕π)σ∗ PES, but no Cl∕Cl∗ products unambiguously attributable to one photon induced C-Cl bond fission are observed from 4-ClPhOH. Differences in fragmentation behaviour at longer excitation wavelengths are more marked. Prompt C-I bond fission is observed following excitation of 4-IPhOH at all λ ≤ 330 nm; the wavelength dependent trends in I∕I∗ product branching ratio, kinetic energy release, and recoil anisotropy suggest that (with regard to C-I bond fission) 4-IPhOH behaves like a mildly perturbed iodobenzene. Br atoms are observed when exciting 4-BrPhOH at long wavelengths also, but their velocity distributions suggest that dissociation occurs after internal conversion to the ground state. O-H bond fission, by tunnelling (as in phenol), is observed only in the cases of 4-FPhOH and, more weakly, 4-ClPhOH. These observed differences in behaviour can be understood given due recognition of (i) the differences in the vertical excitation energies of the C-Y centred (n∕π)σ∗ potentials across the series Y = I bond strength, cf. that of the rival O-H bond, and (ii) the much increased spin-orbit coupling in, particularly, 4-IPhOH. The present results provide (another) reminder of the risks inherent in extrapolating photochemical

  20. Development and manufacturing of special fission chambers for in-core measurement requirements in nuclear reactors

    International Nuclear Information System (INIS)

    Geslot, B.; Berhouet, F.; Oriol, L.; Breaud, S.; Jammes, C.; Filliatre, P.; Villard, J. F.

    2009-01-01

    The Dosimetry Command control and Instrumentation Laboratory (LDCI) at CEA/Cadarache is specialized in the development, design and manufacturing of miniature fission chambers (from 8 mm down to 1.5 mm in diameter). The LDCI fission chambers workshop specificity is its capacity to manufacture and distribute special fission chambers with fissile deposits other than U 235 (typically Pu 242 , Np 237 , U 238 , Th 232 ). We are also able to define the characteristics of the detector for any in-core measurement requirements: sensor geometry, fissile deposit material and mass, filling gas composition and pressure, operating mode (pulse, current or Campbelling) with associated cable and electronics. The fission chamber design relies on numerical simulation and modeling tools developed by the LDCI. One of our present activities in fission chamber applications is to develop a fast neutron flux instrumentation using Campbelling mode dedicated to measurements in material testing reactors. (authors)

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. Effect of fission yield libraries on the irradiated fuel composition in Monte Carlo depletion calculations

    International Nuclear Information System (INIS)

    Mitenkova, E.; Novikov, N.

    2014-01-01

    Improving the prediction of radiation parameters and reliability of fuel behaviour under different irradiation modes is particularly relevant for new fuel compositions, including recycled nuclear fuel. For fast reactors there is a strong dependence of nuclide accumulations on the nuclear data libraries. The effect of fission yield libraries on irradiated fuel is studied in MONTEBURNS-MCNP5-ORIGEN2 calculations of sodium fast reactors. Fission yield libraries are generated for sodium fast reactors with MOX fuel, using ENDF/B-VII.0, JEFF3.1, original library FY-Koldobsky, and GEFY 3.3 as sources. The transport libraries are generated from ENDF/B-VII.0 and JEFF-3.1. Analysis of irradiated MOX fuel using different fission yield libraries demonstrates the considerable spread in concentrations of fission products. The discrepancies in concentrations of inert gases being ∼25%, up to 5 times for stable and long-life nuclides, and up to 10 orders of magnitude for short-lived nuclides. (authors)

  6. Prompt fission neutron spectra of n + 235U above the (n, nf) fission threshold

    International Nuclear Information System (INIS)

    Shu Nengchuan; Chen Yongjing; Liu Tingjin; Jia Min

    2015-01-01

    Calculations of prompt fission neutron spectra (PFNS) from the 235 U(n, f) reaction were performed with a semi-empirical method for En = 7.0 and 14.7 MeV neutron energies. The total PFNS were obtained as a superposition of (n, xnf) pre-fission neutron spectra and post-fission spectra of neutrons which were evaporated from fission fragments, and these two kinds of spectra were taken as an expression of the evaporation spectrum. The contributions of (n, xnf) fission neutron spectra on the calculated PFNS were discussed. The results show that emission of one or two neutrons in the (n, nf) or (n, 2nf) reactions influences the PFNS shape, and the neutron spectra of the (n, xnf) fission-channel are soft compared with the neutron spectra of the (n, f) fission channel. In addition, analysis of the multiple-chance fission component showed that second-chance fission dominates the PFNS with an incident neutron energy of 14.7 MeV whereas first-chance fission dominates the 7 MeV case. (authors)

  7. 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)

  8. Freedom: a transient fission-product release model for radioactive and stable species

    International Nuclear Information System (INIS)

    Macdonald, L.D.; Lewis, B.J.; Iglesias, F.C.

    1989-05-01

    A microstructure-dependent fission-gas release and swelling model (FREEDOM) has been developed for UO 2 fuel. The model describes the transient release behaviour for both the radioactive and stable fission-product species. The model can be applied over the full range of operating conditions, as well as for accident conditions that result in high fuel temperatures. The model accounts for lattice diffusion and grain-boundary sweeping of fusion products to the grain boundaries, where the fission gases accumulate in grain-face bubbles as a result of vacancy diffusion. Release of fission-gas to the free void of the fuel element occurs through the interlinkage of bubbles and cracks on the grain boundaries. This treatment also accounts for radioactive chain decay and neutron-induced transmutation effects. These phenomena are described by mass balance equations which are numerically solved using a moving-boundary, finite-element method with mesh refinement. The effects of grain-face bubbles on fuel swelling and fuel thermal conductivity are included in the ELESIM fuel performance code. FREEDOM has an accuracy of better than 1% when assessed against an analytic solution for diffusional release. The code is being evaluated against a fuel performance database for stable gas release, and against sweep-gas and in-cell fission-product release experiments at Chalk River for active species

  9. 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.

  10. Cumulative fission yield of Ce-148 produced by thermal-neutron fission of U-235

    International Nuclear Information System (INIS)

    Hasan, A.A.

    1984-12-01

    Cumulative fission yield of 148 cesium isotopes and some other fission products produced by thermal-neutron fission of 235 uranium is determined by Germanium/Lithium spectroscopic methods. The measuremets were done at Tsing-Hua open pool reactor using 3 to 4 mg of 93.15% enriched 235 uranium samples. Gamma rays are assigned to the responsible fission products by matching gamma rays energies and half lives. Fission rate is calculated by fission track method. Cumulative fission yields of 148 cesium, 90 krypton, 130 iodine, 144 lanthanum, 89 krypton, 136 xenon, 137 xenon and 140 cesium are calculated. This values are compared with previously predicted values and showed good agreement. 21 Ref

  11. Unit mechanisms of fission gas release: Current understanding and future needs

    Energy Technology Data Exchange (ETDEWEB)

    Tonks, Michael; Andersson, David; Devanathan, Ram; Dubourg, Roland; El-Azab, Anter; Freyss, Michel; Iglesias, Fernando; Kulacsy, Katalin; Pastore, Giovanni; Phillpot, Simon R.; Welland, Michael

    2018-06-01

    Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel properties and, once the gas is released into the gap between the fuel and cladding, lowering gap thermal conductivity and increasing gap pressure. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are being applied to provide unprecedented understanding of the unit mechanisms that define the fission product behavior. In this article, existing research on the basic mechanisms behind the various stages of fission gas release during normal reactor operation are summarized and critical areas where experimental and simulation work is needed are identified. This basic understanding of the fission gas behavior mechanisms has the potential to revolutionize our ability to predict fission product behavior during reactor operation and to design fuels that have improved fission product retention. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.

  12. 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

  13. Fission cross-section calculations and the multi-modal fission model

    International Nuclear Information System (INIS)

    Hambsch, F.J.

    2004-01-01

    New, self consistent, neutron-induced reaction cross section calculations for 235,238 U, 237 Np have been performed. The statistical model code STATIS was improved to take into account the multimodality of the fission process. The three most dominant fission modes, the two asymmetric standards I (S1) and standard II (S2) modes and the symmetric superlong (SL) mode have been taken into account. De-convoluted fission cross sections for those modes for 235,238 U(n,f) and 237 Np(n,f) based on experimental branching ratios, were calculated for the first time up to the second chance fission threshold. For 235 U(n,f), the calculations being made up to 28 MeV incident neutron energy, higher fission chances have been considered. This implied the need for additional calculations for the neighbouring isotopes. As a side product also mass yield distributions could be calculated at energies hitherto not accessible by experiment. Experimental validation of the predictions is being envisaged

  14. Analysis of 137Cs in fission based neutron dosimetry

    International Nuclear Information System (INIS)

    Peltonen, T.

    1995-11-01

    137 Cs analysis is based on dissolving an irradiated fission dosimeter and chemically separating the cesium from the rest of the fission material. The samples consisted of uranium and neptunium in the form of metal or oxide. The uranium samples were dissolved in nitric acid and the neptunium samples in a mixture of nitric acid and chloric acid with addition of hydrogen peroxide. Cs was precipitated into a mixture of ammonium molyndophoshate and cellulose powder. A preparate for measurement was made from the precipitate and covered with polyethen plastic. Since other fission products than cesium were precipitated as well from the more recently irradiated samples, the activity measurements could not be carried out with a NaI(Tl) cavity crystal, but had to be made with a less efficient but more selective germanium semiconductor crystal. The method is well suited for 137 Cs determination, especially for older dosimeters where the more short-lived fission products have decayed. (orig.) (6 refs., 7 figs., 7 tabs.)

  15. Polyphase diffusion of fission products in graphite

    International Nuclear Information System (INIS)

    Dannert, V.

    1989-05-01

    The report attempts to give an introduction into the subject of fission product transport in nuclear graphite and results in an extended proposal of a transport-model. Beginning with a rough description of the graphite in question, an idea about the physical transport-phenomena in graphite is developed. Some of the basic experimental methods, especially techniques of porosimetry, determination of sorption-isotherms and of course several transport-experiments, are briefly described and their results are discussed. Some of the most frequent transport models are introduced and assessed with the criteria emphasized in this report. An extended model is proposed including the following main ideas: The transport of the fission-products is regarded as a two-phase-diffusion process through the open pores of the graphite. The two phases are: surface-diffusion and gas-diffusion. A time-dependent coupling of the two diffusion-phases by sorption-isotherms and a concentration-dependence of the surface diffusion coefficient, also related to the physical behaviour of the sorption-isotherms, are the basic properties of the proposed model. (orig./HP) [de

  16. Burn-up calculation of fusion-fission hybrid reactor using thorium cycle

    International Nuclear Information System (INIS)

    Shido, S.; Matsunaka, M.; Kondo, K.; Murata, I.; Yamamoto, Y.

    2006-01-01

    A burn-up calculation system has been developed to estimate performance of blanket in a fusion-fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In this system, neutron flux is calculated by MCNP4B and then burn-up calculation is performed by ORIGEN2. The cross-section library for ORIGEN2 is made from the calculated neutron flux and evaluated nuclear data. The 3-dimensional ITER model was used as a base fusion reactor. The nuclear fuel (reprocessed plutonium as the fission materials mixed with thorium as the fertile materials), transmutation materials (minor actinides and long-lived fission products) and tritium breeder were loaded into the blanket. Performances of gas-cooled and water-cooled blankets were compared with each other. As a result, the proposed reactor can meet the requirement for TBP and power density. As far as nuclear waste incineration is concerned, the gas-cooled blanket has advantages. On the other hand, the water cooled-blanket is suited to energy production. (author)

  17. Regulatory aspects of fusion power-lessons from fission plants

    International Nuclear Information System (INIS)

    Natalizio, A.; Brunnader, H.; Sood, S.K.

    1993-01-01

    Experience from fission reactors has shown the regulatory process for licensing a nuclear facility to be legalistic, lengthy, unpredictable, and costly. This experience also indicates that much of the regulatory debate is focused on safety margins, that is, the smaller the safety margins the bigger the regulatory debate and the greater the amount of proof required to satisfy the regulatory. Such experience suggests that caution and prudence guide the development of a regulatory regime for fusion reactors. Fusion has intrinsic safety and environmental advantages over fission, which should alleviate significantly, or even eliminate, the regulatory problems associated with fission. The absence of a criticality concern and the absence of fission products preclude a Chernobyl type accident from occurring in a fusion reactor. Although in a fusion reactor there are large inventories of radioactive products that can be mobilized, the total quantity is orders of magnitude smaller than in fission power reactors. The bulk of the radioactivity in a fusion reactor is either activation products in steel structures, or tritium fuel supplies safely stored in the form of a metal tritide in storage beds. The quantity of tritium that can be mobilized under accident conditions is much less than ten million curies. This compares very favorably with a fission product inventory greater than ten billion curies in a fission power reactor. Furthermore, in a fission reactor, all of the reactivity is contained in a steel vessel that is pressurized to about 150 atmospheres, whereas in a fusion reactor, the inventory of radioactive material is dispersed in different areas of the plant, such that it is improbable that a single event could give rise to the release of the entire inventory to the environment. With such significant intrinsic safety advantages there is no a priori need to make fusion requirements/regulations more demanding and more stringent than fission

  18. Development and manufacturing of special fission chambers for in-core measurement requirements in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Geslot, B.; Berhouet, F.; Oriol, L.; Breaud, S.; Jammes, C.; Filliatre, P.; Villard, J. F. [CEA, DEN, Dosimetry Command Control and Instrumentation Laboratory, F-13109 Saint-Paul-lez-Durance (France)

    2009-07-01

    The Dosimetry Command control and Instrumentation Laboratory (LDCI) at CEA/Cadarache is specialized in the development, design and manufacturing of miniature fission chambers (from 8 mm down to 1.5 mm in diameter). The LDCI fission chambers workshop specificity is its capacity to manufacture and distribute special fission chambers with fissile deposits other than U{sup 235} (typically Pu{sup 242}, Np{sup 237}, U{sup 238}, Th{sup 232}). We are also able to define the characteristics of the detector for any in-core measurement requirements: sensor geometry, fissile deposit material and mass, filling gas composition and pressure, operating mode (pulse, current or Campbelling) with associated cable and electronics. The fission chamber design relies on numerical simulation and modeling tools developed by the LDCI. One of our present activities in fission chamber applications is to develop a fast neutron flux instrumentation using Campbelling mode dedicated to measurements in material testing reactors. (authors)

  19. 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.

  20. 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.

  1. 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

  2. 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

  3. 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.)

  4. 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).

  5. Study on the effect factor of the absolute fission rates measured by depleted uranium fission chamber

    International Nuclear Information System (INIS)

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

    2003-01-01

    The absolute fission rates was measured by the depleted uranium fission chamber. The efficiency of the fission fragments recorded in the fission chamber was analyzed. The factor influencing absolute fission rates was studied in the experiment, including the disturbing effect between detectors and the effect of the structural of the fission chamber, etc

  6. Convective-diffusive transport of fission products in the gap of a failed fuel element

    International Nuclear Information System (INIS)

    Lian, Z.W.; Carlucci, L.N.; Arimescu, V.I.

    1995-03-01

    A model is presented to describe the transport behaviour of gaseous fission products along the axial fuel-to-sheathe gap of a failed fuel element to the coolant system. The model is applicable to an element having failed under normal operating conditions or loss-of coolant-accident conditions. Because of the large differences in operating parameters, the transport characteristics of gaseous fission products in a failed element under these two operating conditions are significantly different. However, in both cases the transport process can be described by convection-diffusion caused by the continuous release of fission products from the fuel to the gap. Under normal operating conditions, the bulk-flow velocity is found to be negligible, due to the low release rate of fission products from fuel. The process can be well approximated by the diffusion of fission products in a stagnant gas-steam mixture. The effect of convection on the fission product transport, however, becomes significant under loss-of-coolant-accident conditions, where the release rates of fission products from fuel can be several orders of magnitude higher that that under normal operating conditions. The convection of the mixture in the gap not only contributes an additional flux to the gas-mixture transport, but also increases the gradient of fission products concentration across the opening, and therefore increases the diffusion flux to the coolant. As a result of the bulk flow, the transport of fission products along the gap is accelerated and the hold-up of short-lived isotopes in the gap is significantly reduced. Steam ingress through the opening into the gap is obstructed by the bulk flow, resulting in low steam concentrations in the gap under loss-of-coolant-accident conditions. (author). 6 refs., 8 figs

  7. Investigation of short-living fission products from the spontaneous fission of Cf-252

    International Nuclear Information System (INIS)

    Klonk, H.

    1976-01-01

    In this paper, a method of separating and measuring fission products of Cf-252 is presented. The measurement was achieved by means of γ-spectrometry and thus provides a quantitative analysis with a good separation of the fission products with respect to both atomic number Z and mass number A. The separation of the fission products from the fission source was achieved by means of solid traps. An automatic changing apparatus made it possible to keep irradiation and measuring times short, so even very short-lived fission products could be registered. The quantitative evaluation of primary fission products was made possible by correction according to Bateman equations. With that, the yields of single nuclides and the dispersion of charge can be determined. (orig./WL) [de

  8. 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

  9. Pulsed fission/fusion hybrid engines

    International Nuclear Information System (INIS)

    Hudson, G.C.

    1979-01-01

    Research into high-thrust, high-specific impulse rocket engines using energy from nuclear reactions which has been conducted at this organization will be discussed. The engines are all conceptual in nature, yet are within the realization of conventional or near-term technology. The engine concepts under study at Foundation, Inc. are designed to obviate or minimize these negative effects of the ORION scheme. By using non-chemical triggers to initiate a non-breakeven fusion reaction at the core of a target composed of both fission and fusion fuel, it should be possible to employ the fusion neutrons thus produced to begin a fission reaction in U-235 or Pu-239. Since the density of the target can be increased by as much as a factor of 250 through compression of the pellet, the amount of fission material necessary to produce a critical mass can be greatly reduced. (This also means that the amount of fission products produced for a giventhrust level is also reduced from the ORION levels.) Coupling this eeffect to the large number of 14 MeV fusion neutrons produced early in the compression process and subsequently to the heating of some additional fusion fuel surrounding the critical mass leads to the very efficient burnup of the target. This insures both high yield from the target as well as low cost per MJ energy released. Finally, the use of such small pellets allows the scale of the energy released to be tailored to a level usable in rocket engines of a few tens of tons thrust level. (orig.) [de

  10. 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.

  11. New ceramics for nuclear industry. Case of fission and fusion reactors

    International Nuclear Information System (INIS)

    Yvars, M.

    1979-10-01

    The ceramics used in the nuclear field are described as is their behaviour under radiation. 1) Power reactors - nuclear fission. Ceramics enter into the fabrication of nuclear fuels: oxides, carbides, uranium or plutonium nitrides or oxy-nitrides. Silicon carbide SiC is used for preparing the fuels of helium cooled high temperature reactors. Its use is foreseen in the design of gas high temperature gas thermal exchangers, as is silicon nitride (Si 3 N 4 ). In the materials for safety or control rods, the intense neutron flows induce nuclear reactions which increase the temperature of the neutron absorbing material. Boron carbide B 4 C, rare earth oxides Ln 2 O 3 , or B 4 C-Cu or B 4 C-Al cermets are employed. Burnable poison materials are formed of Al 2 O 3 -B 4 C or Al 2 O 3 -Ln 2 O 3 cermets. The moderators of thermal neutron reactors are in high purety polycrystalline graphite. For the thermal insulation of reactor vessels and jackets, honeycomb ceramics are used as well as ceramic fibres on an increasing scale (kaolin, alumina and other fibres). 2) fusion reactors (Tokomak). These require refractory materials with a low atomic number. Carbon fibres, boron carbide, some borons (Al B 12 ), silicon nitrides and oxy-nitrides and high density alumina are the substances considered [fr

  12. Subthreshold neutron interrogator for detection of radioactive materials

    Science.gov (United States)

    Evans, Michael L.; Menlove, Howard O.; Baker, Michael P.

    1980-01-01

    A device for detecting fissionable material such as uranium in low concentrations by interrogating with photoneutrons at energy levels below 500 keV, and typically about 26 keV. Induced fast neutrons having energies above 500 keV by the interrogated fissionable material are detected by a liquid scintillator or recoil proportional counter which is sensitive to the induced fast neutrons. Since the induced fast neutrons are proportional to the concentration of fissionable material, detection of induced fast neutrons indicate concentration of the fissionable material.

  13. 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.

  14. 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

  15. 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)

  16. The separation of plutonium from uranium and fission products on zirconium phosphate columns

    Energy Technology Data Exchange (ETDEWEB)

    Gal, I; Ruvarac, A [Institute of Nuclear Sciences Boris Kidric, Laboratorija za visoku aktivnost, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    In recent years special attention has been given to the ion-exchange properties of zirconium phosphate and similar compounds in aqueous solutions. These inorganic cation exchangers are stable in oxidizing media and at elevated temperatures. Their resistance to ionizing radiation makes them particularly suitable for work with radioactive solutions. On account of this we considered ir worthwhile to investigate the separation of plutonium from uranium and fission products on zirconium phosphate columns. We were interested in nitric and solutions containing macro-amounts of uranium (a few grams per litre), and micro-amounts of plutonium and long-lived fission products. To obtain a better insight into the ion-exchange behaviour of the different ionic species towards zirconium phosphate, we first determined the dependence of the distribution coefficients of uranium, plutonium and fission product cations on the aqueous nitric acid concentration. Then, taking the distribution data as a guide, we separated plutonium on small glass columns filled with zirconium phosphate and calculated the decontamination factors (author)

  17. 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.

  18. Sequential Detection of Fission Processes for Harbor Defense

    Energy Technology Data Exchange (ETDEWEB)

    Candy, J V; Walston, S E; Chambers, D H

    2015-02-12

    With the large increase in terrorist activities throughout the world, the timely and accurate detection of special nuclear material (SNM) has become an extremely high priority for many countries concerned with national security. The detection of radionuclide contraband based on their γ-ray emissions has been attacked vigorously with some interesting and feasible results; however, the fission process of SNM has not received as much attention due to its inherent complexity and required predictive nature. In this paper, on-line, sequential Bayesian detection and estimation (parameter) techniques to rapidly and reliably detect unknown fissioning sources with high statistical confidence are developed.

  19. 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.

  20. 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

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

    International Nuclear Information System (INIS)

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

    1992-05-01

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

  2. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1984-09-01

    This is the tenth issue of a report series on Fission Product Data, which informs us about all the activities in this field, which are planned, ongoing, or have recently been completed. The types of activities included are measurements, compilations and evaluations of: fission product yields (neutron induced and spontaneous fission), neutron reaction cross sections of fission products, data related to the radioactive decay of fission products, delayed neutron data of fission products, lumped fission product data (decay heat, absorption, etc.). There is also a section with recent references relative to fission product nuclear data

  3. Composite Materials under Extreme Radiation and Temperature Environments of the Next Generation Nuclear Reactors

    International Nuclear Information System (INIS)

    Simos, N.

    2011-01-01

    In the nuclear energy renaissance, driven by fission reactor concepts utilizing very high temperatures and fast neutron spectra, materials with enhanced performance that exceeds are expected to play a central role. With the operating temperatures of the Generation III reactors bringing the classical reactor materials close to their performance limits there is an urgent need to develop and qualify new alloys and composites. Efforts have been focused on the intricate relations and the high demands placed on materials at the anticipated extreme states within the next generation fusion and fission reactors which combine high radiation fluxes, elevated temperatures and aggressive environments. While nuclear reactors have been in operation for several decades, the structural materials associated with the next generation options need to endure much higher temperatures (1200 C), higher neutron doses (tens of displacements per atom, dpa), and extremely corrosive environments, which are beyond the experience on materials accumulated to-date. The most important consideration is the performance and reliability of structural materials for both in-core and out-of-core functions. While there exists a great body of nuclear materials research and operating experience/performance from fission reactors where epithermal and thermal neutrons interact with materials and alter their physio-mechanical properties, a process that is well understood by now, there are no operating or even experimental facilities that will facilitate the extreme conditions of flux and temperature anticipated and thus provide insights into the behaviour of these well understood materials. Materials, however, still need to be developed and their interaction and damage potential or lifetime to be quantified for the next generation nuclear energy. Based on material development advances, composites, and in particular ceramic composites, seem to inherently possess properties suitable for key functions within the

  4. Effects of fissioning nuclei distributions on fragment mass distributions for high energy fission

    Directory of Open Access Journals (Sweden)

    Rossi P C R

    2012-02-01

    Full Text Available We study the effects of fissioning nuclei mass- and energy-distributions on the formation of fragments for fission induced by high energy probes. A Monte Carlo code called CRISP was used for obtaining mass distributions and spectra of the fissioning nuclei for reactions induced by 660 MeV protons on 241Am and on 239Np, by 500 MeV protons on 208Pb, and by Bremsstrahlung photons with end-point energies at 50 MeV and 3500 MeV on 238U. The results show that even at high excitation energies, asymmetric fission may still contribute significantly to the fission cross section of actinide nuclei, while it is the dominante mode in the case of lead. However, more precise data for high energy fission on actinide are necessary in order to allow definite conclusions.

  5. 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)

  6. 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

  7. Angular momenta of fission fragments in the {alpha}-accompanied fission of {sup 252}Cf

    Energy Technology Data Exchange (ETDEWEB)

    Jandel, M.; Kliman, J.; Krupa, L.; Morhac, M. [Slovak Academy of Sciences, Department of Nuclear Physics, Bratislava (Slovakia); Joint Institute for Nuclear Research, Flerov Laboratory for Nuclear Reactions, Dubna (Russian Federation); Hamilton, J.H.; Kormicki, J.; Ramayya, A.V.; Hwang, J.K.; Luo, Y.X.; Fong, D.; Gore, P. [Vanderbilt University, Department of Physics, Nashville, TN (United States); Ter-Akopian, G.M.; Oganessian, Yu.Ts.; Rodin, A.M.; Fomichev, A.S.; Popeko, G.S. [Joint Institute for Nuclear Research, Flerov Laboratory for Nuclear Reactions, Dubna (Russian Federation); Daniel, A.V. [Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Rasmussen, J.O.; Macchiavelli, A.O.; Stoyer, M.A. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Donangelo, R.; Cole, J.D.

    2005-06-01

    For the first time, average angular momenta of the ternary fission fragments {sup 100,102}Zr, {sup 106}Mo, {sup 144,146}Ba and {sup 138,140,142}Xe from the {alpha}-accompanied fission of {sup 252}Cf were obtained from relative intensities of prompt {gamma}-ray transitions with the use of the statistical model calculation. Average values of the angular momenta were compared with the corresponding values for the same fission fragments from the binary fission of {sup 252}Cf. Results indicate the presence of a decreasing trend in the average values of angular momenta induced in ternary fission fragments compared to the same binary fission fragments. On the average, the total angular momentum extracted for ternary fission fragments is {proportional_to}1.4{Dirac_h} lower than in binary fission. Consequently, results indicate that the mechanism of the ternary {alpha}-particles emission may directly effect an induction of angular momenta of fission fragments, and possible scenarios of such mechanisms are discussed. Further, the dependence of the angular momenta of {sup 106}Mo and {sup 140}Xe on the number of emitted neutrons from correlated pairs of primary fragments was obtained also showing a decreasing dependence of average angular momenta with increasing number of emitted neutrons. Consequences are briefly discussed. (orig.)

  8. 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

  9. 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.

  10. Specific fission J-window and angular momentum dependence of the fission barrier

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Hiroshi; Saito, Tadashi; Takahashi, Naruto; Yokoyama, Akihiko [Osaka Univ., Suita (Japan); Shinohara, Atsushi

    1997-04-01

    A method to determine a unique J-window in the fission process was devised and the fissioning nuclide associated with thus extracted J-window was identified for each of the heavy-ion reaction systems. Obtained fission barriers at the resulting J-window were compared with the calculated values by the rotating finite range model (RFRM). The deduced barriers for individual nuclides were compared with the RFRM barriers to reproduce more or less the angular momentum dependence the RFRM prediction. The deduced systematic behavior of the fission barrier indicates no even-odd and shell corrections are necessary. The nuclear dissipation effect based on Kramer`s model revealed substantial reduction of the statistically deduced barrier heights and brought a fairly large scattering from the RFRM J-dependence. However, introduction of the temperature-dependent friction coefficient ({gamma} = 2 for T {>=} 1.0 MeV and 0.5 for T < 1.0 MeV) was found to bring about satisfactory agreement with both RFRM fission barriers and the pre-fission neutron multiplicity systematics. (author). 81 refs.

  11. Fission product chemistry in severe nuclear reactor accidents, specialists' meeting at JRC-Ispra, 15-17 January 1990

    International Nuclear Information System (INIS)

    Nichols, A.L.

    1990-05-01

    A specialists' 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). (author)

  12. Design of ITER neutron monitor using micro fission chambers

    International Nuclear Information System (INIS)

    Nishitani, Takeo; Ebisawa, Katsuyuki; Ando, Toshiro; Kasai, Satoshi; Johnson, L.C.; Walker, C.

    1998-08-01

    We are designing micro fission chambers, which are pencil size gas counters with fissile material inside, to be installed in the vacuum vessel as neutron flux monitors for ITER. We found that the 238 U micro fission chambers are not suitable because the detection efficiency will increase up to 50% in the ITER life time by breading 239 Pu. We propose to install 235 U micro fission chambers on the front side of the back plate in the gap between adjacent blanket modules and behind the blankets at 10 poloidal locations. One chamber will be installed in the divertor cassette just under the dome. Employing both pulse counting mode and Campbelling mode in the electronics, we can accomplish the ITER requirement of 10 7 dynamic range with 1 ms temporal resolution, and eliminate the effect of gamma-rays. We demonstrate by neutron Monte Carlo calculation with three-dimensional modeling that we avoid those detection efficiency changes by installing micro fission chambers at several poloidal locations inside the vacuum vessel. (author)

  13. 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

  14. Neutron-induced fission cross sections

    International Nuclear Information System (INIS)

    Weigmann, H.

    1991-01-01

    In the history of fission research, neutron-induced fission has always played the most important role. The practical importance of neutron-induced fission rests upon the fact that additional neutrons are produced in the fission process, and thus a chain reaction becomes possible. The practical applications of neutron-induced fission will not be discussed in this chapter, but only the physical properties of one of its characteristics, namely (n,f) cross sections. The most important early summaries on the subject are the monograph edited by Michaudon which also deals with the practical applications, the earlier review article on fission by Michaudon, and the review by Bjornholm and Lynn, in which neutron-induced fission receives major attention. This chapter will attempt to go an intermediate way between the very detailed theoretical treatment in the latter review and the cited monograph which emphasizes the applied aspects and the techniques of fission cross-section measurements. The more recent investigations in the field will be included. Section II will survey the properties of cross sections for neutron-induced fission and also address some special aspects of the experimental methods applied in their measurement. Section Ill will deal with the formal theory of neutron-induced nuclear reactions for the resolved resonance region and the region of statistical nuclear reactions. In Section IV, the fission width, or fission transmission coefficient, will be discussed in detail. Section V will deal with the broader structures due to incompletely damped vibrational resonances, and in particular will address the special case of thorium and neighboring isotopes. Finally, Section VI will briefly discuss parity violation effects in neutron-induced fission. 74 refs., 14 figs., 3 tabs

  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. 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)

  17. Correlation of errors in the Monte Carlo fission source and the fission matrix fundamental-mode eigenvector

    International Nuclear Information System (INIS)

    Dufek, Jan; Holst, Gustaf

    2016-01-01

    Highlights: • Errors in the fission matrix eigenvector and fission source are correlated. • The error correlations depend on coarseness of the spatial mesh. • The error correlations are negligible when the mesh is very fine. - Abstract: Previous studies raised a question about the level of a possible correlation of errors in the cumulative Monte Carlo fission source and the fundamental-mode eigenvector of the fission matrix. A number of new methods tally the fission matrix during the actual Monte Carlo criticality calculation, and use its fundamental-mode eigenvector for various tasks. The methods assume the fission matrix eigenvector is a better representation of the fission source distribution than the actual Monte Carlo fission source, although the fission matrix and its eigenvectors do contain statistical and other errors. A recent study showed that the eigenvector could be used for an unbiased estimation of errors in the cumulative fission source if the errors in the eigenvector and the cumulative fission source were not correlated. Here we present new numerical study results that answer the question about the level of the possible error correlation. The results may be of importance to all methods that use the fission matrix. New numerical tests show that the error correlation is present at a level which strongly depends on properties of the spatial mesh used for tallying the fission matrix. The error correlation is relatively strong when the mesh is coarse, while the correlation weakens as the mesh gets finer. We suggest that the coarseness of the mesh is measured in terms of the value of the largest element in the tallied fission matrix as that way accounts for the mesh as well as system properties. In our test simulations, we observe only negligible error correlations when the value of the largest element in the fission matrix is about 0.1. Relatively strong error correlations appear when the value of the largest element in the fission matrix raises

  18. Radiochemistry and the Study of Fission

    Energy Technology Data Exchange (ETDEWEB)

    Rundberg, Robert S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-11-14

    These are slides from a lecture given at UC Berkeley. Radiochemistry has been used to study fission since its 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. The following topics are covered: In the beginning: the discovery of fission; forensics using fission products: what can be learned from fission products, definitions of R-values and Q-values, fission bases, K-factors and fission chambers, limitations; the neutron energy dependence of the mass yield distribution (the two mode fission hypothesis); the influence of nuclear structure on the mass yield distribution. In summary: Radiochemistry has been used to study fission since its discovery. Radiochemical measurement of fission product yields have provided the highest precision data for developing fission models and for nuclear forensics. The two-mode fission hypothesis provides a description of the neutron energy dependence of the mass yield curve. However, data is still rather sparse and more work is needed near second and third chance fission. Radiochemical measurements have provided evidence for the importance of nuclear states in the compound nucleus in predicting the mass yield curve in the resonance region.

  19. Radiochemistry and the Study of Fission

    International Nuclear Information System (INIS)

    Rundberg, Robert S.

    2016-01-01

    These are slides from a lecture given at UC Berkeley. Radiochemistry has been used to study fission since its 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. The following topics are covered: In the beginning: the discovery of fission; forensics using fission products: what can be learned from fission products, definitions of R-values and Q-values, fission bases, K-factors and fission chambers, limitations; the neutron energy dependence of the mass yield distribution (the two mode fission hypothesis); the influence of nuclear structure on the mass yield distribution. In summary: Radiochemistry has been used to study fission since its discovery. Radiochemical measurement of fission product yields have provided the highest precision data for developing fission models and for nuclear forensics. The two-mode fission hypothesis provides a description of the neutron energy dependence of the mass yield curve. However, data is still rather sparse and more work is needed near second and third chance fission. Radiochemical measurements have provided evidence for the importance of nuclear states in the compound nucleus in predicting the mass yield curve in the resonance region.

  20. The Growth of Sea cucumber Stichopus herrmanni After Transverse Induced Fission in Two and Three Fission Plane

    Directory of Open Access Journals (Sweden)

    Retno Hartati

    2016-06-01

    Full Text Available Transverse induced fission proven could be done in Teripang Tril, Stichopus herrmanni. This present works aimed to analyze wound recovery, regeneration period and growth of Teripang Trill  after asexual reproduction by fission using two and three fission plane. Observations were made every day until the sea cucumber body separated into two or more (depending on treatment and reared for 16 weeks.  The results showed that there are differences in wound recovery, regeneration period and growth of S. herrmanni depend on their different fission plane. The wound recovery and regeneration period (days of anterior, middle and posterior individu S. herrmanni resulted from two and three fission plane were varied but the two fission plane the anterior individu recover for longer period than posterior part and  the wound recover process in both end for thee fission plane was same. Average growth of anterior and posterior fragment were longer for two fission plane than three fission plane.  The middle fragment (M1 and M2 both fission plane was able to grow but very low.  It showed that three fission plane gave very slow growth in every fragment of the body. Keywords: growth, post-fission, fission plane, Stichopus herrmanni

  1. 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

  2. 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

  3. Effect of fission dynamics on the spectra and multiplicities of prompt fission neutrons

    International Nuclear Information System (INIS)

    Nix, J.R.; Madland, D.G.; Sierk, A.J.

    1985-01-01

    With the goal of examining their effect on the spectra and multiplicities of the prompt neutrons emitted in fission, we discuss recent advances in a unified macroscopic-microscopic description of large-amplitude collective nuclear dynamics. The conversion of collective energy into single-particle excitation energy is calculated for a new surface-plus-window dissipation mechanism. By solving the Hamilton equations of motion for initial conditions appropriate to fission, we obtain the average fission-fragment translational kinetic energy and excitation energy. The spectra and multiplicities of the emitted neutrons, which depend critically upon the average excitation energy, are then calculated on the basis of standard nuclear evaporation theory, taking into account the average motion of the fission fragments, the distribution of fission-fragment residual nuclear temperature, the energy dependence of the cross section for the inverse process of compound-nucleus formation, and the possibility of multiple-chance fission. Some illustrative comparisons of our calculations with experimental data are shown

  4. 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

  5. NEET Micro-Pocket Fission Detector. Final Project report

    Energy Technology Data Exchange (ETDEWEB)

    Unruh, T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rempe, Joy [Idaho National Lab. (INL), Idaho Falls, ID (United States); McGregor, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ugorowski, Philip [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reichenberger, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ito, Takashi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Villard, J. -F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Alternative Energies and Atomic Energy Commission, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), is funded by the Nuclear Energy Enabling Technologies (NEET) program to develop and test Micro-Pocket Fission Detectors (MPFDs), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package. When deployed, these sensors will significantly advance flux detection capabilities for irradiation tests in US Material Test Reactors (MTRs). Ultimately, evaluations may lead to a more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, and high performance reactors, allowing several Department of Energy Office of Nuclear Energy (DOE-NE) programs to obtain higher accuracy/higher resolution data from irradiation tests of candidate new fuels and materials. Specifically, deployment of MPFDs will address several challenges faced in irradiations performed at MTRs: Current fission chamber technologies do not offer the ability to measure fast flux, thermal flux and temperature within a single compact probe; MPFDs offer this option. MPFD construction is very different than current fission chamber construction; the use of high temperature materials allow MPFDs to be specifically tailored to survive harsh conditions encountered in-core of high performance MTRs. The higher accuracy, high fidelity data available from the compact MPFD will significantly enhance efforts to validate new high-fidelity reactor physics codes and new multi-scale, multi-physics codes. MPFDs can be built with variable sensitivities to survive the lifetime of an experiment or fuel assembly in some MTRs, allowing for more efficient and cost effective power monitoring. The small size of the MPFDs allows multiple sensors to be deployed, offering the potential to

  6. 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.)

  7. 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

  8. 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

  9. 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.

  10. 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.

  11. Detection of gaseous fission products in water - a method of monitoring fuel sheathing failures

    Energy Technology Data Exchange (ETDEWEB)

    Tunnicliffe, P. R.; Whittier, A. C.

    1959-05-15

    The gaseous activities stripped from samples of effluent coolant from the NRU fuel elements tested in the central thimble of the NRX reactor (NRU loop) and from the NRX main effluent have been investigated. The activities obtained from the NRU loop can be attributed to gaseous fission products only. Design data have been obtained for a 'Gaseous Fission Product Monitor' to be installed for use with the NRU reactor. It is expected that this monitor will have high sensitivity to activity indicative of an incipient fuel element sheath failure. No qualitative determination of the various gaseous activities obtained from the NRX effluent has been made. A strong component of 25 {+-}1 seconds half-life is not consistent with O-19. Limited information concerning sheath failures in NRX was obtained. Of six failures observed in parallel with the installed delayed neutron monitors, three of these gave pre-warnings and in each case the gaseous fission product monitor showed a substantially greater sensitivity. An experiment in which small samples of uranium, inserted into the NRX reactor, could be exposed at will to a stream of water showed the behaviour of the two types of monitors to be similar. However, a number of signals were detected only by the gaseous fission product monitor. These can be attributed to its sensitivity to relatively long lived fission products. (author)

  12. 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

  13. Fuel pins irradiation: experimental devices and analytical behaviour

    International Nuclear Information System (INIS)

    Lemaignan, C.

    1996-01-01

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

  14. Experimental verification of the fission chamber gamma signal suppression by the Campbelling mode

    Energy Technology Data Exchange (ETDEWEB)

    Vermeeren, L.; Weber, M. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Oriol, L.; Breaud, S.; Filliatre, P.; Geslot, B.; Jammes, C. [CEA, Centre de Cadarache, F-13109 Saint-Paul-lez-Durance (France); Normand, S.; Lescop, B. [CEA, Centre de Saclay, F-91191 Gif sur Yvette Cedex (France)

    2009-07-01

    For the on-line monitoring of high fast neutron fluxes in the presence of a strong thermal neutron component, SCK-CEN and CEA are jointly developing a Fast Neutron Detector System, based on {sup 242}Pu fission chambers as sensors and including dedicated electronics and data processing systems. Irradiation tests in the BR2 reactor of {sup 242}Pu fission chambers operating in current mode showed that in typical MTR (Materials Test Reactors) conditions the fission chamber currents are dominated by the gamma contribution. In order to reduce the gamma contribution to the signal, it was proposed to use the fission chambers in Campbelling mode. An irradiation experiment in the BR2 reactor with a {sup 242}Pu and a {sup 235}U fission chamber, both equipped with a suitable cable for measurements in Campbelling mode, proved the effectiveness of the suppression of the gamma-induced signal component by the Campbelling mode: gamma contribution reduction factors of 26 for the {sup 235}U fission chamber and more than 80 for the {sup 242}Pu fission chamber were obtained. The experimental data also prove that photofission contributions are negligibly small. Consequently, in typical MTR conditions the gamma contribution to the fission chamber Campbelling signal can be neglected. (authors)

  15. 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.

  16. Fission product vapour - aerosol interactions in the containment: simulant fuel studies

    International Nuclear Information System (INIS)

    Beard, A.M.; Benson, C.G.; Bowsher, B.R.

    1988-12-01

    Experiments have been conducted in the Falcon facility to study the interaction of fission product vapours released from simulant fuel samples with control rod aerosols. The aerosols generated from both the control rod and fuel sample were chemically distinct and had different deposition characteristics. Extensive interaction was observed between the fission product vapours and the control rod aerosol. The two dominant mechanisms were condensation of the vapours onto the aerosol, and chemical reactions between the two components; sorption phenomena were believed to be only of secondary importance. The interaction of fission product vapours and reactor materials aerosols could have a major impact on the transport characteristics of the radioactive emission from a degrading core. (author)

  17. IFPE/IFA-432, Fission Gas Release, Mechanical Interaction BWR Fuel Rods, Halden

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    1996-01-01

    Description: It contains data from experiments that have been performed at the IFE/OECD Halden Reactor Project, available for use in fuel performance studies. It covers experiments on thermal performance, fission product release, clad properties and pellet clad mechanical interaction. It includes also experimental data relevant to high burn-up behaviour. IFA-432: Measurements of fuel temperature response, fission gas release and mechanical interaction on BWR-type fuel rods up to high burn-ups. The assembly featured several variations in rod design parameters, including fuel type, fuel/cladding gap size, fill gas composition (He and Xe) and fuel stability. It contained 6 BWR-type fuel rods with fuel centre thermocouples at two horizontal planes, rods were also equipped with pressure transducers and cladding extensometers. Only data from 6 rods are compiled here

  18. Determination of the fission products yields, lanthanide and yttrium, in the fission of 238U with neutrons of fission spectra

    International Nuclear Information System (INIS)

    Nicoli, I.G.

    1981-06-01

    A radiochemical investigation is performed to measure the cumulative fission product yields of several lantanides and yttrium nuclides in the 238 U by fission neutron spectra. Natural and depleted uranium are irradiated under the same experimental conditions in order to find a way to subtract the contribution of the 235 U fission. 235 U percentage in the natural uranium was 3.5 times higher than in the depleted uranium. Uranium oxides samples are irradiated inside the core of the Argonaut Reactor, at the Instituto de Engenharia Nuclear, and the lantanides and yttrium are chemically separated. The fission products gamma activities were detected, counted and analysed in a system constituted by a high resolution Ge(Li) detector, 4096 multichannel analyser and a PDP-11 computer. Cumulative yields for fission products with half-lives between 1 to 33 hours are measured: 93 Y, 141 La, 142 La, 143 Ce and 149 Nd. The chain total yields are calculated. The cumulative fission yields measured for 93 Y, 141 La, 142 La, 143 Ce and 149 Nd are 4,49%, 4,54%, 4,95%, 4,16% and 1,37% respectively and they are in good agreement with the values found in the literature. (Author) [pt

  19. Gas dynamics models for an oscillating gaseous core fission reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.; Dam, H. van; Hoogenboom, J.E. (Interuniversitair Reactor Inst., Delft (Netherlands))

    1991-01-01

    Two one-dimensional models are developed for the investigation of the gas dynamical behaviour of the fuel gas in a cylindrical gaseous core fission reactor. By numerical and analytical calculations, it is shown that, for the case where a direct energy extraction mechanism (such as magneto-hydrodynamics (MHD)) is not present, increasing density oscillations occur in the gas. Also an estimate is made of the attainable direct energy conversion efficiency, for the case where a direct energy extraction mechanism is present. (author).

  20. 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

  1. Natural fission reactors from Gabon. Contribution to the study of the conditions of stability of a natural radioactive wastes storage site (2 Ga)

    International Nuclear Information System (INIS)

    Pourcelot, L.

    1997-01-01

    The natural fission reactors of Oklo consists of a core of uraninite (60%) with fission products, embedded in a pure clay matrix. Thus, the aim of geological, mineral, and geochemical studies of the Oklo Reactors is to assess the behaviour of fission products in an artificial waste depository. Previous studies have shown that Reactor Zone 10, located in the Oklo mine, represents an example for an exceptional confinement of fission products since 2 Ga. In reactor Zone 9, located in Oklo open pit, migrations are more important. Reactor ZOne 13 was influenced by a thermal event due to a doleritic intrusion, located some twenty meters far away, one Ga years after fission reaction operations. In this study,we characterized temperature and redox conditions of fluids by using stable isotopes of uraninites and clays. Moreover mineralogical and chemical characteristics were defined. (author)

  2. 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)

  3. 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.

  4. Determination of the fission barrier height in fission of heavy radioactive beams induced by the (d,p)-transfer

    CERN Multimedia

    A theoretical framework is described, allowing to determine the fission barrier height using the observed cross sections of fission induced by the (d,p)-transfer with accuracy, which is not achievable in another type of low-energy fission of neutron-deficient nuclei, the $\\beta$-delayed fission. The primary goal is to directly determine the fission barrier height of proton-rich fissile nuclei, preferably using the radio-active beams of isotopes of odd elements, and thus confirm or exclude the low values of fission barrier heights, typically extracted using statistical calculations in the compound nucleus reactions at higher excitation energies. Calculated fission cross sections in transfer reactions of the radioactive beams show sufficient sensitivity to fission barrier height. In the probable case that fission rates will be high enough, mass asymmetry of fission fragments can be determined. Results will be relevant for nuclear astrophysics and for production of super-heavy nuclei. Transfer induced fission of...

  5. Fusion-Fission like studies from medium heavy to light compound systems

    International Nuclear Information System (INIS)

    Heusch, B.

    1991-01-01

    It has been shown that for systems as light as A CN = 47 up to systems just above the Businaro Gallone point in the mass region of 100 to 110 the probability for a system to deexcite by the fission channel, is not negligible. As predicted, the asymmetrical separation becomes dominant when the A CN mass is decreasing but the symmetrical mode remains measurable. The ambiguities in the measured outgoing fragment distributions arise from the competition with IMF emissions as well as dynamical fission processes which depend strongly on the studied system. Fully relaxed DIC has also be used to interpret the results. I tried to show that precise checks on the behavior of two neighbouring systems as well as search for entrance channel effect and/or energy dependence bring evidence enough that the deexcitation of the compound nucleus can account for the symmetric and asymmetric fission channels as well as IMF emissions. This is strongly supported by different recent calculations all done in this frame. These all conclusions indicate also that the RLDM fails in the data interpretation. The strength of the fission channel depends strongly on the possibilities a system has to deexcite. For very light systems especially the number of open channels available determines directly the flux repartition between direct or compound processes and therefore very large differences in the general behaviour of two neighbouring systems can be observed. 15 figs

  6. Fission fragment distributions within dynamical approach

    Energy Technology Data Exchange (ETDEWEB)

    Mazurek, K. [Institute of Nuclear, Physics Polish Academy of Sciences, Krakow (Poland); Nadtochy, P.N. [Omsk State Technical University, Omsk (Russian Federation); Ryabov, E.G.; Adeev, G.D. [Omsk State University, Physics Department, Omsk (Russian Federation)

    2017-04-15

    The review covers recent developments and achievements in the dynamical description of fission process at high excitation energy. It is shown that the dynamical approach based on multidimensional Langevin equations combined with the statistical description of nuclear decay by particles evaporation is capable of fairly well describing the formation of fission fragment mass-energy, charge, and angular distributions of fission fragments in coincidence with the pre- and post-scission particle emission. The final yields of fission and evaporation residues channels products could be obtained. The detailed description of fission dynamics allows studying different stages of fission process, indicating the most important ingredients governing fission process and studying in detail such fundamental nuclear properties as nuclear viscosity and fission timescale. The tasks and perspectives of multidimensional dynamical approach are also discussed. (orig.)

  7. Fission fragment yields from heavy-ion-induced reactions measured with a fragment separator

    Science.gov (United States)

    Tarasov, O. B.; Delaune, O.; Farget, F.; Morrissey, D. J.; Amthor, A. M.; Bastin, B.; Bazin, D.; Blank, B.; Cacéres, L.; Chbihi, A.; Fernández-Dominguez, B.; Grévy, S.; Kamalou, O.; Lukyanov, S. M.; Mittig, W.; Pereira, J.; Perrot, L.; Saint-Laurent, M.-G.; Savajols, H.; Sherrill, B. M.; Stodel, C.; Thomas, J. C.; Villari, A. C.

    2018-04-01

    The systematic study of fission fragment yields under different initial conditions has provided valuable experimental data for benchmarking models of fission product yields. Nuclear reactions using inverse kinematics coupled to the use of a high-resolution spectrometer with good fragment identification are shown here to be a powerful tool to measure the inclusive isotopic yields of fission fragments. In-flight fusion-fission was used in this work to produce secondary beams of neutron-rich isotopes in the collisions of a 238U beam at 24 MeV/u with 9Be and 12C targets at GANIL using the LISE3 fragment separator. Unique identification of the A, Z, and atomic charge state, q, of fission products was attained with the Δ E- TKE-B ρ- ToF measurement technique. Mass, and atomic number distributions are reported for the two reactions. The results show the importance of different reaction mechanisms in the two cases. The optimal target material for higher yields of neutron-rich high- Z isotopes produced in fusion-fission reactions as a function of projectile energy is discussed.

  8. Tip model of cold fission

    International Nuclear Information System (INIS)

    Goennenwein, F.; Boersig, B.

    1991-01-01

    Cold fission is defined to be the limiting case of nuclear fission where virtually all of the available energy is converted into the total kinetic energy of the fragments. The fragments have, therefore, to be born in or at least close to their respective ground states. Starting from the viewpoint that cold fission corresponds to most compact scission configurations, energy constraints have been exploited to calculate minimum tip distances between the two nascent fragments in binary fission. Crucial input parameters to this tip model of cold fission are the ground-state deformations of fragment nuclei. It is shown that the minimum tip distances being compatible with energy conservation vary strongly with both the mass and charge fragmentation of the fission prone nucleus. The tip distances refer to nuclei with equivalent sharp surfaces. In keeping with the size of the surface width of leptodermous nuclei, only configurations where the tip distances are smaller than a few fm may be considered as valid scission configurations. From a comparison with experimental data on cold fission this critical tip distance appears to be 3.0 fm for the model parameters chosen. Whenever the model calculation yields tip distances being smaller than the critical value, a necessary condition for attaining cold fission is considered to be fulfilled. It is shown that this criterion allows to understand in fair agreement with experiment which mass fragmentations are susceptible to lead to cold fission and which fragment-charge divisions are the most favored in each isobaric mass chain. Being based merely on energy arguments, the model cannot aim at predicting fragment yields in cold fission. However, the tip model proposed appears well suited to delineate the phase space where cold fission phenomena may come into sight. (orig.)

  9. Axisymmetric Magnetic Mirror Fusion-Fission Hybrid

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R. W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Martovetsky, N. N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Molvik, A. W. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Ryutov, D. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Simonen, T. C. [Univ. of California, Berkeley, CA (United States)

    2011-05-13

    The achieved performance of the gas dynamic trap version of magnetic mirrors and today’s technology we believe are sufficient with modest further efforts for a neutron source for material testing (Q=Pfusion/Pinput~0.1). The performance needed for commercial power production requires considerable further advances to achieve the necessary high Q>>10. An early application of the mirror, requiring intermediate performance and intermediate values of Q~1 are the hybrid applications. The Axisymmetric Mirror has a number of attractive features as a driver for a fusion-fission hybrid system: geometrical simplicity, inherently steady-state operation, and the presence of the natural divertors in the form of end tanks. This level of physics performance has the virtue of low risk and only modest R&D needed and its simplicity promises economy advantages. Operation at Q~1 allows for relatively low electron temperatures, in the range of 4 keV, for the DT injection energy ~ 80 keV. A simple mirror with the plasma diameter of 1 m and mirror-to-mirror length of 35 m is discussed. Simple circular superconducting coils are based on today’s technology. The positive ion neutral beams are similar to existing units but designed for steady state. A brief qualitative discussion of three groups of physics issues is presented: axial heat loss, MHD stability in the axisymmetric geometry, microstability of sloshing ions. Burning fission reactor wastes by fissioning actinides (transuranics: Pu, Np, Am, Cm, .. or just minor actinides: Np, Am, Cm, …) in the hybrid will multiply fusion’s energy by a factor of ~10 or more and diminish the Q needed to less than 1 to overcome the cost of recirculating power for good economics. The economic value of destroying actinides by fissioning is rather low based on either the cost of long-term storage or even deep geologic disposal so most of the revenues of hybrids will come from electrical power. Hybrids that obtain revenues from

  10. 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

  11. Insights into nuclear structure and the fission process from spontaneous fission

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, J.H.; Butler-Moore, K.; Ramayya, A.V. [Vanderbilt Univ., Nashville, TN (United States)] [and others

    1993-12-31

    The {gamma}-rays emitted following spontaneous and induced fission are rich sources of information about the structure of neutron-rich nuclei and about the fission process itself. The study of spontaneous fissioning isotopes with large Ge detector arrays are providing a wealth of such information as seen, for example, in recent reports. In this paper we present some of our most recent results on nuclear structure studies and conclusions on the fission process itself. In our work, we have employed in spontaneous fission, a triple gamma coincidence study for the first time and a high resolution, X-ray detector-{gamma}-coincidence study. These data provide powerful ways of separating the gamma rays which belong to a particular nucleus. The triple coincidence technique was used to uniquely identify the levels in {sup 136}Te and higher spin states in its N=84 isotones, {sup 138}Xe and {sup 140}Ba{sup 171}. Some other examples of the level structures observed in the low and high mass partners are presented, including a detailed analysis of the backbending of the moment of inertia in {sup 112,114,116}Pd. Finally, we present the first examples of how our analysis allows one to extract a detailed picture of the dependence of the angular momentum on the mass and atomic numbers of the fission fragments and of the long-sought neutron multiplicity distribution from zero-n to ten-n as a function of the charge and mass asymmetry.

  12. Dual-fission chamber and neutron beam characterization for fission product yield measurements using monoenergetic neutrons

    Science.gov (United States)

    Bhatia, C.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.; Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rundberg, R. S.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Macri, R.; Ryan, C.; Sheets, S. A.; Stoyer, M. A.; Tonchev, A. P.

    2014-09-01

    A program has been initiated to measure the energy dependence of selected high-yield fission products used in the analysis of nuclear test data. We present out initial work of neutron activation using a dual-fission chamber with quasi-monoenergetic neutrons and gamma-counting method. Quasi-monoenergetic neutrons of energies from 0.5 to 15 MeV using the TUNL 10 MV FM tandem to provide high-precision and self-consistent measurements of fission product yields (FPY). The final FPY results will be coupled with theoretical analysis to provide a more fundamental understanding of the fission process. To accomplish this goal, we have developed and tested a set of dual-fission ionization chambers to provide an accurate determination of the number of fissions occurring in a thick target located in the middle plane of the chamber assembly. Details of the fission chamber and its performance are presented along with neutron beam production and characterization. Also presented are studies on the background issues associated with room-return and off-energy neutron production. We show that the off-energy neutron contribution can be significant, but correctable, while room-return neutron background levels contribute less than <1% to the fission signal.

  13. 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.)

  14. Reexamination of fission fragment angular distributions and the fission process: Formalism

    International Nuclear Information System (INIS)

    Bond, P.D.

    1985-01-01

    The theory of fission fragment angular distributions is examined and the universally used expression is found to be valid only under restrictive assumptions. A more general angular distribution formula is derived and applied to recent data of high spin systems. At the same time it is shown that the strong anisotropies observed from such systems can be understood without changing the essential basis of standard fission theory. The effects of reaction mechanisms other than complete fusion on fission fragment angular distributions are discussed and possible angular distribution signatures of noncompound nucleus formation are mentioned

  15. 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

  16. Investigation of the heavy nuclei fission with anomalously high values of the fission fragments total kinetic energy

    Science.gov (United States)

    Khryachkov, Vitaly; Goverdovskii, Andrei; Ketlerov, Vladimir; Mitrofanov, Vecheslav; Sergachev, Alexei

    2018-03-01

    Binary fission of 232Th and 238U induced by fast neutrons were under intent investigation in the IPPE during recent years. These measurements were performed with a twin ionization chamber with Frisch grids. Signals from the detector were digitized for further processing with a specially developed software. It results in information of kinetic energies, masses, directions and Bragg curves of registered fission fragments. Total statistics of a few million fission events were collected during each experiment. It was discovered that for several combinations of fission fragment masses their total kinetic energy was very close to total free energy of the fissioning system. The probability of such fission events for the fast neutron induced fission was found to be much higher than for spontaneous fission of 252Cf and thermal neutron induced fission of 235U. For experiments with 238U target the energy of incident neutrons were 5 MeV and 6.5 MeV. Close analysis of dependence of fission fragment distribution on compound nucleus excitation energy gave us some explanation of the phenomenon. It could be a process in highly excited compound nucleus which leads the fissioning system from the scission point into the fusion valley with high probability.

  17. The use of recoil for the separation of uranium fission products; Utilisation du recul pour la separation des produits de fission de l'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Henry, R; Herczec, C [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    The recoil distance of fission fragments in U{sub 3}O{sub 8} is about 8 microns. By using highly diluted suspensions of uranium oxide particles having dimension much smaller than this figure (mean diameter 0,5 micron), we were able to study the re-adsorption of fission products on uranium oxide. Separation results have been studied as a function of the nature of the irradiation medium (solid or liquid) and the separation medium, of particle size and of concentration of particles in the dispersing medium. Decay curves can be used to discriminate between {sup 239}Np and mixed fission products. Most of the {sup 239}Np is found in the U{sub 3}O{sub 8} particles. The location of fission products in solid dispersing media has been determined, fission products being found always inside the dispersing medium particles. The results obtained can be applied to the rapid separation of short-lived fission products from a uranium-free starting material. (author) [French] Le parcours de recul des fragments de fission est en moyenne de 8 microns dans l'U{sub 3}O{sub 8}. En prenant des suspensions d'oxyde d'uranium dont les particules, tres diluees, ont des dimensions nettement inferieures a cette valeur (diametre moyen 0,5 micron), on a pu etudier directement la readsorption des produits de fission sur l'oxyde d'uranium. Les resultats de separation ont ete etudies en fonction de la nature du milieu d'irradiation (solide ou liquide) et du milieu de separation, de la taille des particules d'oxyde et de leur concentration dans le milieu dispersant. Les courbes de decroissance permettent de determiner la perturbation apportee dans les mesures par le {sup 239}Np qui reste en majorite dans les grains d'U{sub 3}O{sub 8}. On a determine enfin l'emplacement des produits de fission dans le cas des melanges solides; ils se trouvent toujours a l'interieur des grains du milieu recepteur. Les resultats obtenus permettent d'envisager la separation rapide de produits de fission a periode courte a

  18. Optimization of Surrounding Reflector Material for Hyb-WT

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Hong, Song Hee; Kim, Myung Hyun

    2013-01-01

    The choice of reflector material is crucial for fusion and hybrid reactors as it was for the fission reactors. Multiple reflector materials were studied for pure fusion blanket design. The purpose of reflector in fusion blanket is to enhance the tritium breeding ratio (TBR). In fusion fission hybrid blanket the roll of reflector is slightly changed as it include the fission core and the performance of fission core also needs to be optimized and evaluated with the choice of reflector material, along with the enhancement of TBR. The performance parameters of Hyb-WT are significantly influenced by the choice of reflector material. TiC is best for TRU transmutation, TBR and reduced the neutron wall loading and graphite is best for FP transmutation. Strategy of multi reflector materials gives the best TRU and FP transmutation performance and also enhanced TBR with reduced neutron wall loading and it is a better choice for Hyb-WT reflector. The neutron flux is primarily dominated by the fission neutrons

  19. Chemistry studies in support of Phebus-FP: multicomponent aerosol behaviour. V. 1

    International Nuclear Information System (INIS)

    1992-01-01

    A major programme of integral tests has been completed in the Falcon facility at Winfrith. This two-year programme was designed to study the transport of fission product vapours and aerosols under severe accident conditions. Simulant or trace-irradiated fuel samples were heated up to 2000 deg K together with bulk-reactor materials, and the transport of the released fission products was followed through a pathway that simulated to some degree the upper plenum, hot-leg structures and the containment. Various analytical techniques were used to provide information on the chemical species and physical forms of the release. Mass spectrometry was developed to sample the high-temperature vapours at ambient pressure and determine their chemical forms, on-line gamma-ray spectroscopy was used to distinguish between gas-borne and deposited radionuclides, and speciation samplers (Maypacks) were used to follow the chemical behaviour of iodine within the containment vessel as a function of time. The experiments were designed and analysed with the assistance of modelling specialists to maximize the benefit of this work for the assessment and development of primary circuit and containment codes

  20. Fifty years with nuclear fission

    International Nuclear Information System (INIS)

    Behrens, J.W.; Carlson, A.D.

    1989-01-01

    The news of the discovery of nuclear fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fifieth anniversary of its discovery by holding a topical meeting entitled, ''Fifty Years with Nuclear Fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent development in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicated a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two fully days of sessions (April 27 and 28) at the main site of the NIST in Gaithersburg, Maryland. The wide range of topics covered in this Volume 1 by this topical meeting included plenary invited, and contributed sessions entitled: Preclude to the First Chain Reaction -- 1932 to 1942; Early Fission Research -- Nuclear Structure and Spontaneous Fission; 50 Years of Fission, Science, and Technology; Nuclear Reactors, Secure Energy for the Future; Reactors 1; Fission Science 1; Safeguards and Space Applications; Fission Data; Nuclear Fission -- Its Various Aspects; Theory and Experiments in Support of Theory; Reactors and Safeguards; and General Research, Instrumentation, and By-Product. The individual papers have been cataloged separately

  1. Fission meter and neutron detection using poisson distribution comparison

    Science.gov (United States)

    Rowland, Mark S; Snyderman, Neal J

    2014-11-18

    A neutron detector system and method 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. Comparison of the observed neutron count distribution with a Poisson distribution is performed to distinguish fissile material from non-fissile material.

  2. 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)

  3. Behaviour of short-lived iodines in operating UO2 fuel elements

    International Nuclear Information System (INIS)

    Lipsett, J.J.; Hastings, I.J.; Hunt, C.E.L.

    1984-11-01

    Sweep gas experiments have been done to determine the behaviour of short-lived fission products within operating UO 2 fuel elements at linear powers of 45, 54, and 60 KW/m, and to burnups of 70, 80, and 50 MWh/kgU respectively. Although radioiodine transport was not observed directly during normal operation, equilibrium gap inventories for I-131 were deduced from the shutdown decay behaviour of the fission gases. These inventories were a strong function of fuel power and ranged from 10 GBq (0.27 Ci) to 100 GBq (2.7 Ci) over the range tested. We conclude that the iodine inventory was adsorbed onto the fuel and/or sheath surfaces with a volatile fraction of less than 10 -2 and a charcoal-filter-penetrating fraction of less than 2x10 -4

  4. 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

  5. Energy production using fission fragment rockets

    International Nuclear Information System (INIS)

    Chapline, G.; Matsuda, Y.

    1991-08-01

    Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: Approximately twice as efficient if one can directly convert the fission fragment energy into electricity; by reducing the buildup of a fission fragment inventory in the reactor one could avoid a Chernobyl type disaster; and collecting the fission fragments outside the reactor could simplify the waste disposal problem. 6 refs., 4 figs., 2 tabs

  6. 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.

  7. 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.

  8. 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

  9. HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

    Energy Technology Data Exchange (ETDEWEB)

    Sell, D. A.; Baily, C. E.; Malewitz, T. J.; Medvedev, P. G.; Porter, D. L.; Hilton, B. A.

    2016-09-01

    A system has been developed to perform measurements on irradiated, sodium bonded-metallic fuel elements to determine the amount of fission gas retained in the fuel material after release of the gas to the element plenum. During irradiation of metallic fuel elements, most of the fission gas developed is released from the fuel and captured in the gas plenums of the fuel elements. A significant amount of fission gas, however, remains captured in closed porosities which develop in the fuel during irradiation. Additionally, some gas is trapped in open porosity but sealed off from the plenum by frozen bond sodium after the element has cooled in the hot cell. The Retained fission Gas (RFG) system has been designed, tested and implemented to capture and measure the quantity of retained fission gas in characterized cut pieces of sodium bonded metallic fuel. Fuel pieces are loaded into the apparatus along with a prescribed amount of iron powder, which is used to create a relatively low melting, eutectic composition as the iron diffuses into the fuel. The apparatus is sealed, evacuated, and then heated to temperatures in excess of the eutectic melting point. Retained fission gas release is monitored by pressure transducers during the heating phase, thus monitoring for release of fission gas as first the bond sodium melts and then the fuel. A separate hot cell system is used to sample the gas in the apparatus and also characterize the volume of the apparatus thus permitting the calculation of the total fission gas release from the fuel element samples along with analysis of the gas composition.

  10. 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.

  11. 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.

  12. 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.

  13. Physics and chemistry of fission

    International Nuclear Information System (INIS)

    1979-01-01

    Full text: In the pleasant and hospitable atmosphere of the Kernforschungsanlage Juelich in the Federal Republic of Germany, the IAEA symposium on the Physics and Chemistry of Fission took place. Almost 200 scientists attended, 154 abstracts were submitted, and 57 papers presented, but more important than the numbers was the quality of the contributions and the progress reported at the symposium. The neutron was discovered almost 50 years ago; 40 years ago the idea of nuclear fission was born. Since then, a number of laboratories have worked hard to explain the phenomenon of fission One would expect that by now scientists would know exactly what happens in a nucleus before and during the process of fission, particularly as there are hundreds of power and research reactors in operation, and fission of uranium isotopes is the basis of their functioning. At first glance, fission seems a simple process: a neutron hits and penetrates the uranium nucleus which becomes excited, i.e. has a surplus of energy. One way to get rid of this energy is for the nucleus to split into two parts; additional products of this process are energy and more neutrons. Nature, however, seems to dislike such straightforward explanations. In the case of fission, scientists have observed a number of phenomena which disagree with a simple model. Sometimes, a nucleus will split into two parts without being 'attacked' by a neutron; this spontaneous fission opens up a new line of fission research and several contributions at the symposium reported on sophisticated experiments designed to unravel some of its specific details. Sometimes, a fissioning nucleus will emit another particle: ternary fission has become a powerful tool for studying the properties of nuclei during the fission process. For the scientist, it is fascinating to observe how the nucleus behaves during fission. They invent models which are supposed to reproduce the most probable course of events leading to fission. In one of these

  14. Theoretical Description of the Fission Process

    International Nuclear Information System (INIS)

    Nazarewicz, Witold

    2009-01-01

    Advanced theoretical methods and high-performance computers may finally unlock the secrets of nuclear fission, a fundamental nuclear decay that is of great relevance to society. In this work, we studied the phenomenon of spontaneous fission using the symmetry-unrestricted nuclear density functional theory (DFT). Our results show that many observed properties of fissioning nuclei can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. From the calculated collective potential and collective mass, we estimated spontaneous fission half-lives, and good agreement with experimental data was found. We also predicted a new phenomenon of trimodal spontaneous fission for some transfermium isotopes. Our calculations demonstrate that fission barriers of excited superheavy nuclei vary rapidly with particle number, pointing to the importance of shell effects even at large excitation energies. The results are consistent with recent experiments where superheavy elements were created by bombarding an actinide target with 48-calcium; yet even at high excitation energies, sizable fission barriers remained. Not only does this reveal clues about the conditions for creating new elements, it also provides a wider context for understanding other types of fission. Understanding of the fission process is crucial for many areas of science and technology. Fission governs existence of many transuranium elements, including the predicted long-lived superheavy species. In nuclear astrophysics, fission influences the formation of heavy elements on the final stages of the r-process in a very high neutron density environment. Fission applications are numerous. Improved understanding of the fission process will enable scientists to enhance the safety and reliability of the nation's nuclear stockpile and nuclear reactors. The deployment of a fleet of safe and efficient advanced reactors, which will also minimize radiotoxic

  15. Experimental approach to fission process of actinides

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Hiroshi [Osaka Univ., Toyonaka (Japan). Faculty of Science

    1997-07-01

    From experimental views, it seems likely that the mechanism of nuclear fission process remains unsolved even after the Bohr and Weeler`s study in 1939. Especially, it is marked in respect of mass distribution in unsymmetric nuclear fission. The energy dependency of mass distribution can be explained with an assumption of 2-mode nuclear fission. Further, it was demonstrated that the symmetrical fission components and the unsymmetrical ones have different saddle and fission points. Thus, the presence of the 2-mode fission mechanism was confirmed. Here, transition in the nuclear fission mechanism and its cause were investigated here. As the cause of such transition, plausible four causes; a contribution of multiple-chance fission, disappearance of shell effects, beginning of fission following collective excitation due to GDR and nuclear phase transition were examined in the condition of excitation energy of 14.0 MeV. And it was suggested that the transition in the nuclear fission concerned might be related to phase transition. In addition, the mechanism of nuclear fission at a low energy and multi-mode hypothesis were examined by determination of the energy for thermal neutron fission ({sup 233,235}U and {sup 239}Pu) and spontaneous nuclear fission ({sup 252}Cf). (M.N.)

  16. 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)

  17. Dynamic of fission and quasi-fission revealed by pre-scission neutron evaporation

    International Nuclear Information System (INIS)

    Hinde, D.J.

    1991-06-01

    The dependence of pre-scission neutron multiplicities (ν-pre) on the mass-split and total kinetic energy (TKE) in fusion-fission and quasi-fission has been measured for a wide range of projectile-target combinations. the data indicate that the fusion-fission time scale is shorter for asymmetric splits than for symmetric splits, whilst there is no dependence on TKE. For quasi-fission reactions induced using 64 Ni projectiles, ν-pre falls rapidly with increasing TKE, indicating that these neutrons are emitted near to or after scission. A new interpretation of both neutron multiplicities and mean energies (the neutron clock-thermometer) allows the extraction of time scales with much less uncertainty than previously, and also gives information about the deformation from which the neutrons are emitted. 15 refs., 13 figs

  18. Theory of nuclear fission: a review

    International Nuclear Information System (INIS)

    Mosel, U.

    1976-01-01

    General properties of nuclear fission are reviewed and related to our present knowledge of fission theory. For this purpose the basic reasons for the shape of the fission barriers are discussed and their consequences compared with experimental results on barrier shapes and structures. Special emphasis is put on the asymmetry of the fission barriers and mass-distributions and its relation to the shells of the nascent fragment shells. Finally the problem of calculating fission cross sections is discussed

  19. Transmutation of fission products and actinide waste at Hanford

    Energy Technology Data Exchange (ETDEWEB)

    Daemen, L.L.; Pitcher, E.J.; Russell, G.J. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    The authors studied the neutronics of an ATW system for the transmutation of the fission products ({sup 99}Tc in particular) and the type of actinide waste stored in several tanks at Hanford. The heart of the system is a highly-efficient neutron production target. It is surrounded by a blanket containing a moderator/reflector material, as well as the products to be transmuted. The fission products are injected into the blanket in the form of an aqueous solution in heavy water, whereas an aqueous actinides slurry is circulated in the outer part of the blanket. For the sake of definiteness, the authors focussed on {sup 99}Tc (the most difficult fission product to transmute), and {sup 239}Pu, {sup 237}Np, and {sup 241}Am. Because of the low thermal neutron absorption cross-section of {sup 99}Tc, considerable care and effort must be devoted to the design of a very efficient neutron source.

  20. Gamma-ray multiplicity distribution in ternary fission of {sup 252}Cf

    Energy Technology Data Exchange (ETDEWEB)

    Jandel, M [Department of Nuclear Physics, Slovak Academy of Sciences, Dubravska cesta 9, Bratislava (Slovakia); Kliman, J [Department of Nuclear Physics, Slovak Academy of Sciences, Dubravska cesta 9, Bratislava (Slovakia); Krupa, L [Department of Nuclear Physics, Slovak Academy of Sciences, Dubravska cesta 9, Bratislava (Slovakia); Morhac, M [Department of Nuclear Physics, Slovak Academy of Sciences, Dubravska cesta 9, Bratislava (Slovakia); Hamilton, J H [Department of Physics, Vanderbilt University, Nashville, TN (United States); Kormicki, J [Department of Physics, Vanderbilt University, Nashville, TN (United States); Ramayya, A V [Department of Physics, Vanderbilt University, Nashville, TN (United States); Hwang, J K [Department of Physics, Vanderbilt University, Nashville, TN (United States); Luo, Y X [Department of Physics, Vanderbilt University, Nashville, TN (United States); Fong, D [Department of Physics, Vanderbilt University, Nashville, TN (United States); Gore, P [Department of Physics, Vanderbilt University, Nashville, TN (United States); Akopian, G M Ter; Oganessian, Yu Ts; Rodin, A M; Fomichev, A S; Popeko, G S; Daniel, A V [Flerov Laboratory for Nuclear Reactions, Joint Institute for Nuclear Research, Dubna (Russian Federation); Rasmussen, J O; Macchiavelli, A O [Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Stoyer, M A [Lawrence Livermore National Laboratory, Livermore, CA (United States); Donangelo, R [Instituto de Fisica, Universidade Federal do Rio de Janeiro, 21945-970 Rio de Janeiro (Brazil); Cole, J D [Idaho National Engineering and Environmental Laboratory, Idaho Falls, ID (United States)

    2002-12-01

    From multiparameter data obtained at Lawrence Berkeley National Laboratory, the integral characteristics of the prompt {gamma}-ray emission were extracted for tripartition of {sup 252}Cf with He, Be and C being the third light charged particle. We used multifold {gamma}-ray coincidence spectra for the determination of {gamma}-ray multiplicities assuming a Gaussian distribution for {gamma}-ray multiplicity. The multiplicity distribution characteristics, i.e. mean multiplicity and its dispersion were obtained by minimizing with respect to the calculated values of probabilities of multifold {gamma}-ray coincidences using a combinatoric method. Comparison with the known experimental data from binary fission was made. Further, we investigated dependencies of the mean {gamma}-ray multiplicity on the kinetic energy of the light charged particle. The mean {gamma}-ray multiplicity for He ternary fission is found to increase rapidly with increasing kinetic energy of He in the region less than 11 MeV and then decrease slowly with increasing kinetic energy of He. The anomalous behaviour of {gamma}-ray emission is discussed. The mean {gamma}-ray multiplicity was determined for the first time for Be and C ternary fission. For Be, the {gamma}-ray multiplicity as a function of kinetic energy was obtained as well.

  1. Design of In-vessel neutron monitor using micro fission chambers for ITER

    International Nuclear Information System (INIS)

    Nishitani, Takeo; Kasai, Satoshi

    2001-10-01

    A neutron monitor using micro fission chambers to be installed inside the vacuum vessel has been designed for compact ITER (ITER-FEAT). We investigated the responses of the micro fission chambers to find the suitable position of micro fission chambers by a neutron Monte Carlo calculation using MCNP version 4b code. It was found that the averaged output of the micro fission chambers behind blankets at upper outboard and lower outboard is insensitive to the changes in the plasma position and the neutron source profile. A set of 235 U micro fission chamber and ''blank'' detector which is a fissile material free detector to identify noise issues such as from γ-rays are installed behind blankets. Employing both pulse counting mode and Campbelling mode in the electronics, the ITER requirement of 10 7 dynamic range with 1 ms temporal resolution can be accomplished. The in-situ calibration has been simulated by MCNP calculation, where a point source of 14 MeV neutrons is moving on the plasma axis. It was found that the direct calibration is possible by using a neutron generator with an intensity of 10 11 n/s. The micro fission chamber system can meet the required 10% accuracy for a fusion power monitor. (author)

  2. 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

  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. Contribution to the study of the diffusion of fission products in uranium; Contribution a l'etude de la diffusion des produits de fission dans l'uranium

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-10-15

    In this work we have developed a simple method for determining the diffusion constants and the solid-state solubilities of a metal which is volatile and only slightly soluble in second metal. This method has been applied to the behaviour of certain fission products in {gamma} uranium: strontium, barium, lanthanum, samarium and cerium. This work has made it possible to show the effect of the atomic radius of the solute on the diffusion constants. (author) [French] Dans ce travail nous avons mis au point une methode simple permettant de determiner les constantes de diffusion ainsi que les solubilites a l'etat solide d'un metal volatil et peu soluble dans un autre. Cette methode a ete appliquee au comportement de certains produits de fission dans l'uranium {gamma}: strontium, baryum, lanthane, samarium et cerium. Cette etude a permis de mettre en evidence le role du rayon atomique du solute sur les constantes de diffusion. (auteur)

  5. Future research program on prompt γ-ray emission in nuclear fission

    Science.gov (United States)

    Oberstedt, S.; Billnert, R.; Hambsch, F.-J.; Lebois, M.; Oberstedt, A.; Wilson, J. N.

    2015-12-01

    In recent years the measurement of prompt fission γ-ray spectra (PFGS) has gained renewed interest, after about forty years since the first comprehensive studies of the reactions 235U(n th , f), 239Pu(n th ,f) and 252Cf(sf). The renaissance was initiated by requests for new values especially for γ-ray multiplicity and average total energy release per fission in neutron-induced fission of 235U and 239Pu. Both isotopes are considered the most important ones with respect to the modeling of innovative cores required for the Generation-IV reactors, the majority working with fast neutrons. During the last 5 years we have conducted a systematic study of spectral data for thermal-neutron-induced fission on 235U and 241Pu as well as for the spontaneous fission of 252Cf with unprecedented accuracy. From the new data we conclude that those reactions do not considerably contribute to the observed heat excess and suspect other reactions playing a significant role. Possible contributions may originate from fast-neutron-induced reactions on 238U, which is largely present in the fuel, or from γ-induced fission from neutron capture in the construction material. A first experiment campaign on prompt γ-ray emission from fast-neutron-induced fission on 235,238U was successfully performed in order to test our assumptions. In the following we attempt to summarize, what has been done in the field to date, and to motivate future measurement campaigns exploiting dedicated neutron and photon beams as well as upcoming highly efficient detector assemblies.

  6. Absolute nuclear material assay

    Science.gov (United States)

    Prasad, Manoj K [Pleasanton, CA; Snyderman, Neal J [Berkeley, CA; Rowland, Mark S [Alamo, CA

    2010-07-13

    A method of absolute nuclear material assay of an unknown source comprising counting neutrons from the unknown source and providing an absolute nuclear material assay utilizing a model to optimally compare to the measured count distributions. In one embodiment, the step of providing an absolute nuclear material assay comprises utilizing a random sampling of analytically computed fission chain distributions to generate a continuous time-evolving sequence of event-counts by spreading the fission chain distribution in time.

  7. PolyPole-1: An accurate numerical algorithm for intra-granular fission gas release

    International Nuclear Information System (INIS)

    Pizzocri, D.; Rabiti, C.; Luzzi, L.; Barani, T.; Van Uffelen, P.; Pastore, G.

    2016-01-01

    The transport of fission gas from within the fuel grains to the grain boundaries (intra-granular fission gas release) is a fundamental controlling mechanism of fission gas release and gaseous swelling in nuclear fuel. Hence, accurate numerical solution of the corresponding mathematical problem needs to be included in fission gas behaviour models used in fuel performance codes. Under the assumption of equilibrium between trapping and resolution, the process can be described mathematically by a single diffusion equation for the gas atom concentration in a grain. In this paper, we propose a new numerical algorithm (PolyPole-1) to efficiently solve the fission gas diffusion equation in time-varying conditions. The PolyPole-1 algorithm is based on the analytic modal solution of the diffusion equation for constant conditions, combined with polynomial corrective terms that embody the information on the deviation from constant conditions. The new algorithm is verified by comparing the results to a finite difference solution over a large number of randomly generated operation histories. Furthermore, comparison to state-of-the-art algorithms used in fuel performance codes demonstrates that the accuracy of PolyPole-1 is superior to other algorithms, with similar computational effort. Finally, the concept of PolyPole-1 may be extended to the solution of the general problem of intra-granular fission gas diffusion during non-equilibrium trapping and resolution, which will be the subject of future work. - Highlights: • A new numerical algorithm (PolyPole-1) for intra-granular fission gas release in time-varying conditions is developed. • The concept combines the modal analytic solution for constant conditions and a polynomial correction. • PolyPole-1 is extensively verified and compared to other state-of-the-art algorithms. • PolyPole-1 exhibits a superior accuracy and a similar computational time relative to other algorithms. • The PolyPole-1 algorithm can be

  8. PolyPole-1: An accurate numerical algorithm for intra-granular fission gas release

    Energy Technology Data Exchange (ETDEWEB)

    Pizzocri, D. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division, Via La Masa 34, 20156 Milano (Italy); Rabiti, C. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Luzzi, L.; Barani, T. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division, Via La Masa 34, 20156 Milano (Italy); Van Uffelen, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Pastore, G., E-mail: giovanni.pastore@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States)

    2016-09-15

    The transport of fission gas from within the fuel grains to the grain boundaries (intra-granular fission gas release) is a fundamental controlling mechanism of fission gas release and gaseous swelling in nuclear fuel. Hence, accurate numerical solution of the corresponding mathematical problem needs to be included in fission gas behaviour models used in fuel performance codes. Under the assumption of equilibrium between trapping and resolution, the process can be described mathematically by a single diffusion equation for the gas atom concentration in a grain. In this paper, we propose a new numerical algorithm (PolyPole-1) to efficiently solve the fission gas diffusion equation in time-varying conditions. The PolyPole-1 algorithm is based on the analytic modal solution of the diffusion equation for constant conditions, combined with polynomial corrective terms that embody the information on the deviation from constant conditions. The new algorithm is verified by comparing the results to a finite difference solution over a large number of randomly generated operation histories. Furthermore, comparison to state-of-the-art algorithms used in fuel performance codes demonstrates that the accuracy of PolyPole-1 is superior to other algorithms, with similar computational effort. Finally, the concept of PolyPole-1 may be extended to the solution of the general problem of intra-granular fission gas diffusion during non-equilibrium trapping and resolution, which will be the subject of future work. - Highlights: • A new numerical algorithm (PolyPole-1) for intra-granular fission gas release in time-varying conditions is developed. • The concept combines the modal analytic solution for constant conditions and a polynomial correction. • PolyPole-1 is extensively verified and compared to other state-of-the-art algorithms. • PolyPole-1 exhibits a superior accuracy and a similar computational time relative to other algorithms. • The PolyPole-1 algorithm can be

  9. 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.

  10. Some aspects of the research and development programmes on the behaviour of containments during severe accidents

    International Nuclear Information System (INIS)

    Dufresne, J.

    1989-01-01

    The R and D programmes relating to the behaviour of containments during severe accidents cover several domains: .leaktightness of the containment: this programme concerns the mechanical resistance of the concretes and the cracking criteria, on the one hand, and the leak rate through the porosities or cracks, on the other; . gaseous releases inside the containment. In addition to the releases of steam and fission products from the primary circuit, the gaseous H 2 0 and C0 2 releases from the concrete must also be studied: firstly during the corium-concrete interaction, and secondly during the heating of the internal surface of the containment which can be raised to a high temperature on contact with the atmosphere, for example during hydrogen combustion; . the release of fission products during the corium-concrete interactions; . the behaviour of the fission products inside the containment, particularly as regards iodine

  11. Mechanisms of fission neutron emission

    International Nuclear Information System (INIS)

    Maerten, H.

    1991-01-01

    The time evolution in fission is the starting point for discussing not only the main mechanism of fission neutron emission, the evaporation from fully accelerated fragments, but also possible secondary ones connected with dynamical features of nuclear fission. ''Asymptotic'' conditions as relevant for describing the particle release from highly excited, rapidly moving fragments are defined. Corresponding statistical model approaches to fission neutron emission, based on the adequate consideration of the intricate fragment occurrence probability, reproduce most of the experimental data. The remarkable influence of fission modes on neutron observables is analyzed in the framework of a macroscopic-microscopic scission point model consistent with energy conservation. Finally, chances and deficiencies for solving the mechanism puzzle are summarized. (author). 87 refs, 21 figs

  12. 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.

  13. Development of a “Fission-proxy” Method for the Measurement of 14-MeV Neutron Fission Yields at CAMS

    Energy Technology Data Exchange (ETDEWEB)

    Gharibyan, Narek [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-10-25

    Relative fission yield measurements were made for 50 fission products from 25.6±0.5 MeV alpha-induced fission of Th-232. Quantitative comparison of these experimentally measured fission yields with the evaluated fission yields from 14-MeV neutron-induced fission of U-235 demonstrates the feasibility of the proposed fission-proxy method. This new technique, based on the Bohr-independence hypothesis, permits the measurement of fission yields from an alternate reaction pathway (Th-232 + 25.6 MeV α → U-236* vs. U-235 + 14-MeV n → U-236*) given that the fission process associated with the same compound nucleus is independent of its formation. Other suitable systems that can potentially be investigated in this manner include (but are not limited to) Pu-239 and U-237.

  14. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1981-06-01

    This is the seventh issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields (neutron induced and spontaneous fission); neutron reaction cross sections of fission products; data related to the radioactive decay of fission products; delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The sixth issue of this series has been published in June 1980 as INDC(NDS)-113/G+P. The present issue includes contributions which were received by NDS between 1 August 1980 and 25 May 1981

  15. 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.

  16. Nuclear fission and neutron-induced fission cross-sections

    CERN Document Server

    James, G D; Michaudon, A; Michaudon, A; Cierjacks, S W; Chrien, R E

    2013-01-01

    Nuclear Fission and Neutron-Induced Fission Cross-Sections is the first volume in a series on Neutron Physics and Nuclear Data in Science and Technology. This volume serves the purpose of providing a thorough description of the many facets of neutron physics in different fields of nuclear applications. This book also attempts to bridge the communication gap between experts involved in the experimental and theoretical studies of nuclear properties and those involved in the technological applications of nuclear data. This publication will be invaluable to those interested in studying nuclear fis

  17. Degraded core accidents: review of aerosol behaviour in the containment of a PWR

    International Nuclear Information System (INIS)

    Nichols, A.L.; Walker, B.C.

    1981-09-01

    Low probability-high consequence accidents have become an important issue in reactor safety studies. Such accidents would involve damage to the core and the subsequent release of radioactive fission products into the environment. Aerosols play a major role in the transport and removal of these fission products in the reactor building containment. The aerosol mechanisms, computer modelling codes and experimental studies used to predict aerosol behaviour in the containment of a PWR are reviewed. There are significant uncertainties in the aerosol source terms and specific recommendations have been made for further studies, particularly with respect to code development and high density aerosol-fission product transport within closed systems. (author)

  18. 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.

  19. Magnetically insulated fission electric cells for direct energy conversion

    International Nuclear Information System (INIS)

    Slutz, S.A.; Seidel, D.B.; Lipinski, R.J.; Rochau, G.E.; Brown, L.C.

    2003-01-01

    The principles of fission electric cells are reviewed. A detailed Monte Carlo model of the efficiency of a fission electric cell is presented and a theory of magnetically insulated fission electric cells (MIFECs) is developed. It is shown that the low operating voltages observed in previous MIFEC experiments were due to nonoptimal magnetic field profiles. Improved magnetic field profiles are presented. It is further shown that the large electric field present in a MIFEC limits the structure of the cathode and can lead to a displacement instability of the cathode toward the anode. This instability places constraints on the number of cells that can be strung together without some external cathode support. The large electric field stress also leads to electrical surface breakdown of the cathode. It is shown that this leads to the formation of a virtual cathode resulting in geometry constraints for spherical cells. Finally it is shown that the requirements of magnetic insulation and high efficiency leads to very low average density of the fissile material. Thus a reactor using fission electric cells for efficient direct energy conversion will be large and require a very large number of cells. This could be mitigated somewhat by the use of exotic fuels

  20. JEFF-3T. Decay data and fission yield libraries

    International Nuclear Information System (INIS)

    Bersillon, O.; Blachot, J.; Dean, C.J.; Mills, R.W.; Nichols, A.L.; Nouri, A.

    2002-01-01

    Comprehensive decay-data and fission-yield libraries provide important input to a wide range of nuclear physics codes for nuclear applications. A new initiative has begun under the auspices of the NEA/OECD to generate improved data sets that will constitute the JEFF-3 libraries in ENDF-6 format, primarily for nuclear power, fuel reprocessing and waste management needs. Various sources of decay data have been accessed in order to assemble these files: NUBASE, ENSDF, UKPADD-6 and UKHEDD-2. Efforts have also focused on the evaluation of decay data for a number of important short-lived fission products, so that artificial adjustments to some of the relevant decay data and fission yields are not required to accommodate a previous lack of such data. Fission yields were adopted from UK evaluations recently undertaken to create the UKFY3 library. Decay-data files for 3 755 nuclides have been prepared, including sets of data for the stable nuclides (i.e. mass, natural abundance, spin and parity). Problems in the assignment of ENDF material numbers were addressed, while format and consistency tests were made using CHECKR and FIZCON, respectively. The assembly processes are discussed and reviewed, and the contents of the JEFF-3T starter libraries are described. (author)

  1. 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

  2. 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

  3. 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

  4. 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

  5. A comparison of single knock-on and complete bubble destruction models of the fission induced re-solution of gas atoms from bubbles

    International Nuclear Information System (INIS)

    Wood, M.H.

    1978-03-01

    In previous theoretical studies of the behaviour of the fission gases in nuclear fuel, the Nelson single knock-on model of the fission induced re-solution of gas atoms from fission gas bubbles has been employed. In the present investigation, predictions from this model are compared with those from a complete bubble destruction model of the re-solution process. The main conclusions of the study are that the complete bubble destruction model predicts more gas release after a particular irradiation time than the single knock-on model, for the same choice of the model parameters, and that parameter sets chosen to give the same gas release predict significantly different bubble size distribution functions. (author)

  6. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, G.

    1976-05-01

    The purpose of this series is to inform scientists working on Fission Product Nuclear Data, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. This report consists of reproductions of essentially unaltered original contributions which the authors have sent to IAEA/NDS. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields; neutron cross-section data of fission products; data related to β-, γ-decay of fission products; delayed neutron data; and fission product decay-heat

  7. 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

  8. Fusion-Fission Hybrid for Fissile Fuel Production without Processing

    Energy Technology Data Exchange (ETDEWEB)

    Fratoni, M; Moir, R W; Kramer, K J; Latkowski, J F; Meier, W R; Powers, J J

    2012-01-02

    Two scenarios are typically envisioned for thorium fuel cycles: 'open' cycles based on irradiation of {sup 232}Th and fission of {sup 233}U in situ without reprocessing or 'closed' cycles based on irradiation of {sup 232}Th followed by reprocessing, and recycling of {sup 233}U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE1) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in

  9. Ejection of Uranium Atoms from UO{sub 2} by Fission Fragments

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Goesta

    1964-02-15

    The numbers of uranium atoms ejected from the surface of sintered plates of UO{sub 2} by fission fragments have been measured over the fission density range 5x10{sup 15} to 7x10{sup 16} fissions/cm{sup 3}. The number of uranium atoms ejected per escaping fragment was about 9. The measurements were performed by irradiating the plates in vacuum and collecting a fraction of the uranium atoms ejected on catcher foils. The amount collected was determined by fission counting. Saturation of the amount collected, as reported by Rogers and Adam, was not observed. The numbers of uranium atoms ejected as knock-ons under the same experimental conditions have been calculated. The reasonably close agreement between the experimental and theoretical values indicates that, under the prevailing experimental conditions, mainly knock-ons are ejected. Other ejection mechanisms, e. g. evaporation of material in thermal spikes, are probably insignificant; this is in contrast to the usual interpretation of the ejection process. The mean range in UO{sub 2}, of fission products of mass number 140 was found to be 7.37 {+-} 0. 05 mg/cm{sup 2} by direct gamma spectrometric, determination of the fraction of {sup 140}La escaping from the surface of the plates.

  10. Physics and technology of nuclear materials

    CERN Document Server

    Ursu, Ioan

    2015-01-01

    Physics and Technology of Nuclear Materials presents basic information regarding the structure, properties, processing methods, and response to irradiation of the key materials that fission and fusion nuclear reactors have to rely upon. Organized into 12 chapters, this book begins with selectively several fundamentals of nuclear physics. Subsequent chapters focus on the nuclear materials science; nuclear fuel; structural materials; moderator materials employed to """"slow down"""" fission neutrons; and neutron highly absorbent materials that serve in reactor's power control. Other chapters exp

  11. Control of a laser inertial confinement fusion-fission power plant

    Science.gov (United States)

    Moses, Edward I.; Latkowski, Jeffery F.; Kramer, Kevin J.

    2015-10-27

    A laser inertial-confinement fusion-fission energy power plant is described. The fusion-fission hybrid system uses inertial confinement fusion to produce neutrons from a fusion reaction of deuterium and tritium. The fusion neutrons drive a sub-critical blanket of fissile or fertile fuel. A coolant circulated through the fuel extracts heat from the fuel that is used to generate electricity. The inertial confinement fusion reaction can be implemented using central hot spot or fast ignition fusion, and direct or indirect drive. The fusion neutrons result in ultra-deep burn-up of the fuel in the fission blanket, thus enabling the burning of nuclear waste. Fuels include depleted uranium, natural uranium, enriched uranium, spent nuclear fuel, thorium, and weapons grade plutonium. LIFE engines can meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the highly undesirable stockpiles of depleted uranium, spent nuclear fuel and excess weapons materials.

  12. 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

  13. Fission signal detection using helium-4 gas fast neutron scintillation detectors

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, J. M., E-mail: lewisj@ufl.edu; Kelley, R. P.; Jordan, K. A. [Nuclear Engineering Program, University of Florida, Gainesville, Florida 32611 (United States); Murer, D. [Arktis Radiation Detectors Ltd., 8045 Zurich (Switzerland)

    2014-07-07

    We demonstrate the unambiguous detection of the fission neutron signal produced in natural uranium during active neutron interrogation using a deuterium-deuterium fusion neutron generator and a high pressure {sup 4}He gas fast neutron scintillation detector. The energy deposition by individual neutrons is quantified, and energy discrimination is used to differentiate the induced fission neutrons from the mono-energetic interrogation neutrons. The detector can discriminate between different incident neutron energies using pulse height discrimination of the slow scintillation component of the elastic scattering interaction between a neutron and the {sup 4}He atom. Energy histograms resulting from this data show the buildup of a detected fission neutron signal at higher energies. The detector is shown here to detect a unique fission neutron signal from a natural uranium sample during active interrogation with a (d, d) neutron generator. This signal path has a direct application to the detection of shielded nuclear material in cargo and air containers. It allows for continuous interrogation and detection while greatly minimizing the potential for false alarms.

  14. A stochastic approach to fission

    International Nuclear Information System (INIS)

    Boilley, D.; Suraud, E.; Abe, Yasuhisa

    1992-01-01

    A microscopically derived Langevin equation is applied to thermally induced nuclear fission. An important memory effect is pointed out. A strong friction coefficient, calculated from microscopic quantities, tends to decrease the stationary limit of the fission rate and to increase the transient time. Fission was described as a diffusion over a barrier of a collective variable, and a Langevin Equation (LE) was used to study the phenomenon. A study of the stationary flow over the saddle point with a Fokker-Planck Equation (FPE), equivalent to the LE was used to give formula for the stationary fission rate (or reaction rate for the chemistry applications). More recently, a complete study of the fission process was performed numerically with both FPE and LE. A long transient time, that could allow more pre-scission neutrons to evaporate, was pointed out. The derivation of this new LE is recalled, followed by the description of the memory dependence and by the effect of a large friction coefficient on the fission rate. (author) 6 refs., 3 figs

  15. Fission product release from SLOWPOKE-2 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

    1994-12-31

    Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

  16. Fission Fragment Angular Distributions in the $^{234}$U(n,f) and $^{236}$U(n,f) reactions

    CERN Multimedia

    We propose to measure the fission fragment angular distribution (FFAD) of the $^{234}$U(n,f) and $^{236}$U (n,f) reactions with the PPAC detection setup used in previous n_TOF-14 experiment. This experiment would take advantage of the high resolution of the n_TOF facility to investigate the FFAD behaviour in the pronounced vibrational resonances that have been observed between 0.1 and 2 MeV for the thorium cycle isotopes. In addition, the angular distribution of these isotopes will be measured for the first time beyond 14 MeV. Furthermore, the experiment will also provide the fission cross section with reduced statistical uncertainty, extending the $^{236}$U(n,f) data up to 1 GeV

  17. Adequate Measuring Technology and System of Fission Gas release Behavior from Voloxidation Process

    International Nuclear Information System (INIS)

    Park, Geun Il; Park, J. J.; Jung, I. H.; Shin, J. M.; Yang, M. S.; Song, K. C.

    2006-09-01

    Based on the published literature and an understanding of available hot cell technologies, more accurate measuring methods for each volatile fission product released from voloxidation process were reviewed and selected. The conceptual design of an apparatus for measuring volatile and/or semi-volatile fission products released from spent fuel was prepared. It was identified that on-line measurement techniques can be applied for gamma-emitting fission products, and off-line measurement such as chemical/or neutron activation analysis can applied for analyzing beta-emitting fission gases. Collection methods using appropriate material or solutions were selected to measure the release fraction of beta-emitting gaseous fission products at IMEF M6 hot cell. Especially, the on-line gamma-ray counting system for monitoring of 85Kr and the off-line measuring system of 14C was established. On-line measuring system for obtaining removal ratios of the semi-volatile fission products, mainly gamma-emitting fission products such as Cs, Ru etc., was also developed at IMEF M6 hot cell which was based on by measuring fuel inventory before and after the voloxidation test through gamma measuring technique. The development of this measurement system may enable basic information to be obtained to support design of the off-gas treatment system for the voloxidation process at INL, USA

  18. Role of fission-reactor-testing capabilities in the development of fusion technology

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Schmunk, R.E.; Takata, M.L.; Watts, K.D.

    1981-01-01

    Testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for testing when fusion reactors become available. Fission testing is capable of filling many gaps in fusion reactor design information, and thus should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, we investigated radiation damage to magnet insulators. This work is now continuing with the use of an improved test capsule. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, we explored a conceptual design for a fission-based Integrated Test Facility (ITF), which can accommodate entire First Wall/Blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

  19. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1983-08-01

    This is the ninth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); Neutron reaction cross sections of fission products; Data related to the radioactive decay of fission products; Delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The eighth issue of this series has been published in July 1982 as INDC(NDS)-130. The present issue includes contributions which were received by NDS between 1 August 1982 and 25 June 1983

  20. On the possibility of unified description of fission and collision of nuclei

    International Nuclear Information System (INIS)

    Shigin, V.A.

    1978-01-01

    A system of two touching nuclei is comparatively stable against disintegration and fusion and unstable against chaotic nucleon exchange between the nuclei which can essentially change the size of the nuclei. These features of touching nuclei were used to suggest a fission mechanism involving production of a light nucleus (α particle) at the periphery of the compound nucleus, growth of the light nucleus caused by exchange of nucleons and subsequent disintegration into two nuclei. In this paper the potential relief of such a fission is presented. The life times of the system against disintegration, fusion, exchange by a nucleon and change of the nuclear sizes are estimated. It is shown that the behaviour of colliding nuclei after nearing is determined by the same potential relief and exchange by nucleons. Formation of the compound nucleus is explained due to transition of nucleons from one nucleus into another

  1. Stochastic simulation of destruction processes in self-irradiated materials

    Directory of Open Access Journals (Sweden)

    T. Patsahan

    2017-09-01

    Full Text Available Self-irradiation damages resulting from fission processes are common phenomena observed in nuclear fuel containing (NFC materials. Numerous α-decays lead to local structure transformations in NFC materials. The damages appearing due to the impacts of heavy nuclear recoils in the subsurface layer can cause detachments of material particles. Such a behaviour is similar to sputtering processes observed during a bombardment of the material surface by a flux of energetic particles. However, in the NFC material, the impacts are initiated from the bulk. In this work we propose a two-dimensional mesoscopic model to perform a stochastic simulation of the destruction processes occurring in a subsurface region of NFC material. We describe the erosion of the material surface, the evolution of its roughness and predict the detachment of the material particles. Size distributions of the emitted particles are obtained in this study. The simulation results of the model are in a qualitative agreement with the size histogram of particles produced from the material containing lava-like fuel formed during the Chernobyl nuclear power plant disaster.

  2. Properties of Fission-Product decay heat from Minor-Actinide fissioning systems

    International Nuclear Information System (INIS)

    Oyamatsu, Kazuhiro; Mori, Hideki

    2000-01-01

    The aggregate Fission-Product (FP) decay heat after a pulse fission is examined for Minor Actinide (MA) fissiles 237 Np, 241 Am, 243 Am, 242 Cm and 244 Cm. We find that the MA decay heat is comparable but smaller than that of 235 U except for cooling times at about 10 8 s (approx. = 3 y). At these cooling times, either the β or γ component of the FP decay heat for these MA's is substantially larger than the one for 235 U. This difference is found to originate from the cumulative fission yield of 106 Ru (T 1/2 = 3.2x10 7 s). This nuclide is the parent of 106 Rh (T 1/2 = 29.8 s) which is the dominant source of the decay heat at 10 8 s (approx. = 3 y). The fission yield is nearly an increasing function of the fissile mass number so that the FP decay heat is the largest for 244 Cm among the MA's at the cooling time. (author)

  3. 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

  4. Thermal safety analysis for pebble bed blanket fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie

    1998-01-01

    Pebble bed blanket hybrid reactor may have more advantages than slab element blanket hybrid reactor in nuclear fuel production and nuclear safety. The thermo-hydraulic calculations of the blanket in the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor developed in China are carried out using the Code THERMIX and auxiliary code. In the calculations different fuel pebble material and steady state, depressurization and total loss of flow accident conditions are included. The results demonstrate that the conceptual design of the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor with dump tank is feasible and safe enough only if the suitable fuel pebble material is selected and the suitable control system and protection system are established. Some recommendations for due conceptual design are also presented

  5. Systematic evaluation of prompt neutron spectra in fission

    International Nuclear Information System (INIS)

    Osawa, Takaaki

    1995-01-01

    To create the nuclear data fail JEND-32, the prompt fission neutron spectra X(E) of 233 U, 235 U, 238 U and 239 Pu were reevaluated and some improvement were added to the calculation models. We tried to extend the calculation method of fission spectra of nuclides with poor measurement data in consideration of increasing the importance of nuclear data of minor actinoids. We improved and extended the following five points. (1) On JENDL-3.1, the fission spectra of principal fissible materials had been calculated by the Modland-Nix model which the neutron emissions of fragments were calculated under the approximation of the constant inverse process cross section. In the paper, the spectra were calculated by the use of the inverse process cross section depend on the energy obtained by the calculation of the optical model. The result showed the increase of low energy components and the softening effect of spectra (2) On JENDL-3.1, the all fission processes were assumed to undergo (n,f) reaction. In the paper, they were calculated by the multi-chance fission such as (n, n'f), (n, 2nf) and (n, 3nf) etc. Softening of the spectra (En > 6 MeV) was obtained by this method. (3) The level density parameter (LDP) has been assumed as a = A/C in either case of light fragment (LF) and heavy fragment (HF) in the original Madland-Nix model. But we used LDP based on the Ignatyuk model under consideration of the shell effects of nuclear fragments, hence the neutron spectra of heavy fragments were hardening. (4) Nuclear temperature of both fragments had been assumed to be the same at original model, but now R T = Tm/TmH was derived to calculate them. The ratio of middle/both side components of spectra was changed. (5) Unknown neutron fission spectra of minor actinide were able to the assumed on the basis of Moriyama-Ohnishi model. (S.Y.)

  6. Monte carlo sampling of fission multiplicity.

    Energy Technology Data Exchange (ETDEWEB)

    Hendricks, J. S. (John S.)

    2004-01-01

    Two new methods have been developed for fission multiplicity modeling in Monte Carlo calculations. The traditional method of sampling neutron multiplicity from fission is to sample the number of neutrons above or below the average. For example, if there are 2.7 neutrons per fission, three would be chosen 70% of the time and two would be chosen 30% of the time. For many applications, particularly {sup 3}He coincidence counting, a better estimate of the true number of neutrons per fission is required. Generally, this number is estimated by sampling a Gaussian distribution about the average. However, because the tail of the Gaussian distribution is negative and negative neutrons cannot be produced, a slight positive bias can be found in the average value. For criticality calculations, the result of rejecting the negative neutrons is an increase in k{sub eff} of 0.1% in some cases. For spontaneous fission, where the average number of neutrons emitted from fission is low, the error also can be unacceptably large. If the Gaussian width approaches the average number of fissions, 10% too many fission neutrons are produced by not treating the negative Gaussian tail adequately. The first method to treat the Gaussian tail is to determine a correction offset, which then is subtracted from all sampled values of the number of neutrons produced. This offset depends on the average value for any given fission at any energy and must be computed efficiently at each fission from the non-integrable error function. The second method is to determine a corrected zero point so that all neutrons sampled between zero and the corrected zero point are killed to compensate for the negative Gaussian tail bias. Again, the zero point must be computed efficiently at each fission. Both methods give excellent results with a negligible computing time penalty. It is now possible to include the full effects of fission multiplicity without the negative Gaussian tail bias.

  7. Yields of fission products produced by thermal-neutron fission of 229Th

    International Nuclear Information System (INIS)

    Dickens, J.K.; McConnell, J.W.

    1983-01-01

    Absolute yields have been determined for 47 gamma rays emitted in the decay of 37 fission products representing 25 mass chains created during thermal-neutron fission of 229 Th. Using a Ge(Li) detector, spectra were obtained of gamma rays emitted between 15 min and 0.4 yr after very short irradiations by thermal neutrons of a 15-μg sample of 229 Th. On the basis of measured gamma-ray yields and known nuclear data, yields for cumulative production of 37 fission products were deduced. The absolute overall normalization uncertainty is 235 U, we postulate a simple functional dependence sigma = sigma(Z/sub p/), and using this dependence obtain values of Z/sub p/(A) for 15 mass chains created during fission of 229 Th. Values of Z/sub p/(A) were estimated for other mass chains based upon results of a recent study of Z/sub p/(A). Charge distributions determined using the deduced mass distribution and the deduced sets of Z/sub p/(A) and sigma(Z/sub p/) are in very good agreement with recent measurements, exhibiting a pronounced even-odd effect in elemental yields. These results may be used to predict unmeasured yields for 229 Th fission

  8. Systematics of Fission-Product Yields

    International Nuclear Information System (INIS)

    Wahl, A.C.

    2002-01-01

    Empirical equations representing systematics of fission-product yields have been derived from experimental data. The systematics give some insight into nuclear-structure effects on yields, and the equations allow estimation of yields from fission of any nuclide with atomic number Z F = 90 thru 98, mass number A F = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru ∼200 MeV--the ranges of these quantities for the fissioning nuclei investigated. Calculations can be made with the computer program CYFP. Estimates of uncertainties in the yield estimates are given by equations, also in CYFP, and range from ∼ 15% for the highest yield values to several orders of magnitude for very small yield values. A summation method is used to calculate weighted average parameter values for fast-neutron (∼ fission spectrum) induced fission reactions

  9. Systematics of Fission-Product Yields

    Energy Technology Data Exchange (ETDEWEB)

    A.C. Wahl

    2002-05-01

    Empirical equations representing systematics of fission-product yields have been derived from experimental data. The systematics give some insight into nuclear-structure effects on yields, and the equations allow estimation of yields from fission of any nuclide with atomic number Z{sub F} = 90 thru 98, mass number A{sub F} = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru {approx}200 MeV--the ranges of these quantities for the fissioning nuclei investigated. Calculations can be made with the computer program CYFP. Estimates of uncertainties in the yield estimates are given by equations, also in CYFP, and range from {approx} 15% for the highest yield values to several orders of magnitude for very small yield values. A summation method is used to calculate weighted average parameter values for fast-neutron ({approx} fission spectrum) induced fission reactions.

  10. Future research program on prompt γ-ray emission in nuclear fission

    Energy Technology Data Exchange (ETDEWEB)

    Oberstedt, S.; Hambsch, F.J. [Joint Research Centre IRMM, European Commission, Geel (Belgium); Billnert, R. [Joint Research Centre IRMM, European Commission, Geel (Belgium); Chalmers Tekniska Hoegskola, Fundamental Fysik, Goeteborg (Sweden); Lebois, M.; Wilson, J.N. [Institut de Physique Nucleaire Orsay, Orsay (France); Oberstedt, A. [Chalmers Tekniska Hoegskola, Fundamental Fysik, Goeteborg (Sweden); Ossolution Consulting, Oerebro (Sweden)

    2015-12-15

    In recent years the measurement of prompt fission γ-ray spectra (PFGS) has gained renewed interest, after about forty years since the first comprehensive studies of the reactions {sup 235}U(n{sub th}, f), {sup 239}Pu(n{sub th},f) and {sup 252}Cf(sf). The renaissance was initiated by requests for new values especially for γ-ray multiplicity and average total energy release per fission in neutron-induced fission of {sup 235}U and {sup 239}Pu. Both isotopes are considered the most important ones with respect to the modeling of innovative cores required for the Generation-IV reactors, the majority working with fast neutrons. During the last 5 years we have conducted a systematic study of spectral data for thermal-neutron-induced fission on {sup 235}U and {sup 241}Pu as well as for the spontaneous fission of {sup 252}Cf with unprecedented accuracy. From the new data we conclude that those reactions do not considerably contribute to the observed heat excess and suspect other reactions playing a significant role. Possible contributions may originate from fast-neutron-induced reactions on {sup 238}U, which is largely present in the fuel, or from γ-induced fission from neutron capture in the construction material. A first experiment campaign on prompt γ-ray emission from fast-neutron-induced fission on {sup 235,238}U was successfully performed in order to test our assumptions. In the following we attempt to summarize, what has been done in the field to date, and to motivate future measurement campaigns exploiting dedicated neutron and photon beams as well as upcoming highly efficient detector assemblies. (orig.)

  11. Asymmetry in ternary fission induced by polarized neutrons and fission mechanism

    International Nuclear Information System (INIS)

    Bunakov, V.E.; Gennenvajn, F.; Dzhessinger, P.; Mutterer, M.; Petrov, G.A.

    2003-01-01

    The results of measuring the P-odd, P-even (right-left) and T-odd asymmetries of the charged particles emission in the double and ternary fission, induced by the polarized neutrons, are considered. It is shown, what kind of information on the mechanism of the ternary nuclear fission may be obtained from the theoretical analysis of these data [ru

  12. 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

  13. Summary of the materials breakout group

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F. [Pacific Northwest Laboratories, Richland, WA (United States); Sommer, W. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    This breakout group discussed the following topics and reached a number of recommendations summarized in this report. The topics of discussion included (1) a comparison of expected materials response at spallation sources to experience with fission reactors; (2) synergistic effects from high-energy particle radiation, corrosion, and cyclic stresses; (3) estimates, based on experience with materials developed for fission and fusion reactors, for -materials testing needs; - facilities for materials development, both available and needed; - candidate materials; - time frame for material qualification; and - expected program costs.

  14. Thorium-uranium fission radiography

    Science.gov (United States)

    Haines, E. L.; Weiss, J. R.; Burnett, D. S.; Woolum, D. S.

    1976-01-01

    Results are described for studies designed to develop routine methods for in-situ measurement of the abundance of Th and U on a microscale in heterogeneous samples, especially rocks, using the secondary high-energy neutron flux developed when the 650 MeV proton beam of an accelerator is stopped in a 42 x 42 cm diam Cu cylinder. Irradiations were performed at three different locations in a rabbit tube in the beam stop area, and thick metal foils of Bi, Th, and natural U as well as polished silicate glasses of known U and Th contents were used as targets and were placed in contact with mica which served as a fission track detector. In many cases both bare and Cd-covered detectors were exposed. The exposed mica samples were etched in 48% HF and the fission tracks counted by conventional transmitted light microscopy. Relative fission cross sections are examined, along with absolute Th track production rates, interaction tracks, and a comparison of measured and calculated fission rates. The practicality of fast neutron radiography revealed by experiments to data is discussed primarily for Th/U measurements, and mixtures of other fissionable nuclei are briefly considered.

  15. Transient fission gas release from UO2 fuel for high temperature and high burnup

    International Nuclear Information System (INIS)

    Szuta, M.

    2001-01-01

    In the present paper it is assumed that the fission gas release kinetics from an irradiated UO 2 fuel for high temperature is determined by the kinetics of grain growth. A well founded assumption that Vitanza curve describes the change of uranium dioxide re-crystallization temperature and the experimental results referring to the limiting grain size presented in the literature are used to modify the grain growth model. Algorithms of fission gas release due to re-crystallization of uranium dioxide grains are worked out. The defect trap model of fission gas behaviour described in the earlier papers is supplemented with the algorithms. Calculations of fission gas release in function of time, temperature, burn-up and initial grain sizes are obtained. Computation of transient fission gas release in the paper is limited to the case where steady state of irradiation to accumulate a desired burn-up is performed below the temperature of re-crystallization then the subsequent step temperature increase follows. There are considered two kinds of step temperature increase for different burn-up: the final temperature of the step increase is below and above the re-crystallization temperature. Calculations show that bursts of fission gas are predicted in both kinds. The release rate of gas liberated for the final temperature above the re-crystallization temperature is much higher than for final temperature below the re-crystallization temperature. The time required for the burst to subside is longer due to grain growth than due to diffusion of bubbles and knock-out release. The theoretical results explain qualitatively the experimental data but some of them need to be verified since this sort of experimental data are not found in the available literature. (author)

  16. 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.

  17. Fission-product source terms

    International Nuclear Information System (INIS)

    Lorenz, R.A.

    1981-01-01

    This presentation consists of a review of fission-product source terms for light water reactor (LWR) fuel. A source term is the quantity of fission products released under specified conditions that can be used to calculate the consequences of the release. The source term usually defines release from breached fuel-rod cladding but could also describe release from the primary coolant system, the reactor containment shell, or the site boundary. The source term would be different for each locality, and the chemical and physical forms of the fission products could also differ

  18. Nuclear fission studies: from LOHENGRIN to FIPPS

    International Nuclear Information System (INIS)

    Chebboubi, Abdelaziz

    2015-01-01

    Nuclear fission consists in splitting a nucleus, in general an actinide, into smaller nuclei. Despite nuclear fission was discovered in 1939 by Hahn and Strassman, fission models cannot predict the fission observables with an acceptable accuracy for nuclear fuel cycle studies for instance. Improvement of fission models is an important issue for the knowledge of the process itself and for the applications. To reduce uncertainties of the nuclear data used in a nuclear reactor simulation, a validation of the models hypothesis is mandatory. In this work, two features of the nuclear fission were investigated in order to test the resistance of the theories. One aspect is the study of the symmetric fission fragments through the measurement of their yield and kinetic energy distribution. The other aspect is the study of the fission fragment angular momentum.Two techniques are available to assess the angular momentum of a fission fragment. The first one is to look at the properties of the prompt gamma. The new spectrometer FIPPS (Fission Product Prompt gamma-ray Spectrometer), is currently under development at the ILL and will combine a fission filter with a large array of gamma and neutron detectors in order to respond to these issues. The first part of this work is dedicated to the study of the properties of a Gas Filled Magnet (GFM) which is the type of fission filter considered for the FIPPS project.The second part of this work deals with the measurement of isomeric yields and evaluations of the angular momentum distribution of fission fragments. The study of the spherical nucleus 132 Sn shed the light on the current limits of fission models. Finally, the last part of this work is about the measurement of the yields and kinetic energy distributions of symmetric fission fragments. Since models predict the existence of fission modes, the symmetry region is a suitable choice to investigate this kind of prediction. In parallel with all these studies, an emphasis on the

  19. Generalized Energy-Dependent Q Values for Fission

    Energy Technology Data Exchange (ETDEWEB)

    Vogt, R

    2010-03-31

    We extend Madland's parameterization of the energy release in fission to obtain the dependence of the fission Q value for major and minor actinides on the incident neutron energies in the range 0 {le} E{sub n} {le} 20 MeV. Our parameterization is based on the actinide evaluations recommended for the ENDF/B-VII.1 release. This paper describes the calculation of energydependent fission Q values based on the calculation of the prompt energy release in fission by Madland. This calculation was adopted for use in the LLNL ENDL database and then generalized to obtain the prompt fission energy release for all actinides. Here the calculation is further generalized to the total energy release in fission. There are several stages in a fission event, depending on the time scale. Neutrons and gammas may be emitted at any time during the fission event.While our discussion here is focussed on compound nucleus creation by an incident neutron, similar parameterizations could be obtained for incident gammas or spontaneous fission.

  20. New fission-neutron-spectrum representation for ENDF

    International Nuclear Information System (INIS)

    Madland, D.G.

    1982-04-01

    A new representation of the prompt fission neutron spectrum is proposed for use in the Evaluated Nuclear Data File (ENDF). The proposal is made because a new theory exists by which the spectrum can be accurately predicted as a function of the fissioning nucleus and its excitation energy. Thus, prompt fission neutron spectra can be calculated for cases where no measurements exist or where measurements are not possible. The mathematical formalism necessary for application of the new theory within ENDF is presented and discussed for neutron-induced fission and spontaneous fission. In the case of neutron-induced fission, expressions are given for the first-chance, second-chance, third-chance, and fourth-chance fission components of the spectrum together with that for the total spectrum. An ENDF format is proposed for the new fission spectrum representation, and an example of the use of the format is given

  1. Code Development on Fission Product Behavior under Severe Accident-Validation of Aerosol Sedimentation

    International Nuclear Information System (INIS)

    Ha, Kwang Soon; Kim, Sung Il; Jang, Jin Sung; Kim, Dong Ha

    2016-01-01

    The gas and aerosol phases of the radioactive materials move through the reactor coolant systems and containments as loaded on the carrier gas or liquid, such as steam or water. Most radioactive materials might escape in the form of aerosols from a nuclear power plant during a severe reactor accident, and it is very important to predict the behavior of these radioactive aerosols in the reactor cooling system and in the containment building under severe accident conditions. Aerosols are designated as very small solid particles or liquid droplets suspended in a gas phase. The suspended solid or liquid particles typically have a range of sizes of 0.01 m to 20 m. Aerosol concentrations in reactor accident analyses are typically less than 100 g/m3 and usually less than 1 g/m3. When there are continuing sources of aerosol to the gas phase or when there are complicated processes involving engineered safety features, much more complicated size distributions develop. It is not uncommon for aerosols in reactor containments to have bimodal size distributions for at least some significant periods of time early during an accident. Salient features of aerosol physics under reactor accident conditions that will affect the nature of the aerosols are (1) the formation of aerosol particles, (2) growth of aerosol particles, (3) shape of aerosol particles. At KAERI, a fission product module has been developed to predict the behaviors of the radioactive materials in the reactor coolant system under severe accident conditions. The fission product module consists of an estimation of the initial inventories, species release from the core, aerosol generation, gas transport, and aerosol transport. The final outcomes of the fission product module designate the radioactive gas and aerosol distribution in the reactor coolant system. The aerosol sedimentation models in the fission product module were validated using ABCOVE and LACE experiments. There were some discrepancies on the predicted

  2. Code Development on Fission Product Behavior under Severe Accident-Validation of Aerosol Sedimentation

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang Soon; Kim, Sung Il; Jang, Jin Sung; Kim, Dong Ha [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The gas and aerosol phases of the radioactive materials move through the reactor coolant systems and containments as loaded on the carrier gas or liquid, such as steam or water. Most radioactive materials might escape in the form of aerosols from a nuclear power plant during a severe reactor accident, and it is very important to predict the behavior of these radioactive aerosols in the reactor cooling system and in the containment building under severe accident conditions. Aerosols are designated as very small solid particles or liquid droplets suspended in a gas phase. The suspended solid or liquid particles typically have a range of sizes of 0.01 m to 20 m. Aerosol concentrations in reactor accident analyses are typically less than 100 g/m3 and usually less than 1 g/m3. When there are continuing sources of aerosol to the gas phase or when there are complicated processes involving engineered safety features, much more complicated size distributions develop. It is not uncommon for aerosols in reactor containments to have bimodal size distributions for at least some significant periods of time early during an accident. Salient features of aerosol physics under reactor accident conditions that will affect the nature of the aerosols are (1) the formation of aerosol particles, (2) growth of aerosol particles, (3) shape of aerosol particles. At KAERI, a fission product module has been developed to predict the behaviors of the radioactive materials in the reactor coolant system under severe accident conditions. The fission product module consists of an estimation of the initial inventories, species release from the core, aerosol generation, gas transport, and aerosol transport. The final outcomes of the fission product module designate the radioactive gas and aerosol distribution in the reactor coolant system. The aerosol sedimentation models in the fission product module were validated using ABCOVE and LACE experiments. There were some discrepancies on the predicted

  3. Device for measuring fission product density

    International Nuclear Information System (INIS)

    Kaneda, Mitsunori.

    1980-01-01

    Purpose: To determine the fission product density of xenon or the like and enable measurement of real time of fission product density in a reactor by calculating the disintegration and annihilation of the fission product on the basis of neutron detected output. Constitution: The neutron flux in a reactor is detected by a detector, and applied to first and second density calculators. Second fission product density signal of xenon or the like outputted from first device is again inputted to the device to form an annihilation signal due to disintegration to determine the present density of the second fission product of xenon or the like corresponding to the decrease of the neutron due to the poison of xeron or the like. Similarly, second device determines the first fission product density of iodine or the like. (Sekiya, K.)

  4. 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

  5. Chemical activity of noble gases Kr and Xe and its impact on fission gas accumulation in the irradiated UO2 fuel

    International Nuclear Information System (INIS)

    Szuta, M.

    2006-01-01

    It is generally accepted that most of the insoluble inert gas atoms Xe and Kr produced during fissioning are retained in the fuel irradiated at a temperature lower than the threshold. Experimental data imply that we can assume that after irradiation exposure in excess of 10 18 fissions/cm 3 the single gas atom diffusion can be disregarded in description of fission gas behaviour. It is assumed that the vicinity of the fission fragment trajectory is the place of intensive irradiation induced chemical interaction of the fission gas products with UO 2 . Significant part of fission gas product is thus expected to be chemically bound in the matrix of UO 2 . Experiments with mixture of noble gases, coupled with theoretical calculations, provide strong evidence for direct bonds between Ar, Kr, or Xe atoms and the U atom of the CUO molecule. Because of its positive charge, the UO 2 2+ ion, which is isoelectronic with CUO, should form even stronger bonds with noble gas atoms, which could lead to a growing number of complexes that contain direct noble gas - to - actinide bonds. Considering the huge amount of gas immobilised in the UO 2 fuel the solution process and in consequence the re-solution process of rare gases is to be replaced by the chemical bonding process. This explains the fission gas accumulation in the irradiated UO 2 fuel. (author)

  6. 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).

  7. Fission cross section and fission fragment angular distribution for oriented nucleus fission by intermediate energy neutrons (epsilon < or approximately 1 Mev)

    International Nuclear Information System (INIS)

    Barabanov, A.L.; Grechukhin, D.P.

    1985-01-01

    General analysis is conducted, and formulae for fission cross section and angular distribution of fission fragments of oriented nuclei by fast neutrons are presented. Geometrical coefficients making up the formulae permitting to carry out calculations for target nuclei with spins I=3/2, 5/2, 7/2 at interaction energies epsilon < or approximately 1 MeV are tabulated. Results of demonstrative calculation of fission fragment angular distribution of oriented sup(235)U nuclei by 0.1 <= epsilon <= 1.0 MeV neutrons reveal that angular distribution weakly depends on the set of permeability factors of neutron waves applied in the calculations

  8. Contribution to the study of nuclear fission

    International Nuclear Information System (INIS)

    Serot, O.

    2009-09-01

    The author proposes an overview of his research activity during the past fifteen years and more particularly that dealing with nuclear fission. The first part reports works on nucleus physics at the scission via the investigation of ternary fission (experimental procedure, influence of fission modes, influence of resonance spin, influence of excitation energy of the fissioning nucleus, emission probabilities, energy spectra of ternary alphas and tritons, emission mechanism). The second part reports measurements and assessments of neutron-induced fission cross sections. The third part reports the investigation of some properties of fission products (efficiencies, branching ratios of the main delayed neutron precursors)

  9. Fission - track age of the Marjalahti Pallasite

    International Nuclear Information System (INIS)

    Bondar, Yu.V.; Perelygin, V.P.

    2006-01-01

    Full text: Investigation of fossil charged-particle tracks in various mineral phases of extraterrestrial samples is a powerful method for research the early stages of the solar system. Over geological time, meteorites crystals have accumulated a record of tracks produced by heavily charged energetic particles from both internal (spontaneous fission of 238U and some other extinct isotopes) and external sources (galactic cosmic rays with Z>20). The fortunate fact that meteorite grains can accumulate latent and very long-lived tracks since soon after the end of nucleosynthesis in the solar nebula enables one to decode their radiation history and to detect any thermal events in the meteorite cosmic history by revealing these tracks through suitable etching procedures. Only a few minerals in meteorites (mainly phosphates) contain small amount of uranium; the fact that 238 U undergoes fission with fission-decay constant λ f ∼ 8.2x10 -17 yr -1 allows one to use this isotope as a chronometer. By measuring the U concentration in the crystals (by reactor irradiation) and the density of the spontaneous-fission tracks it is relatively easy to calculate the 'fission-track age' if 238 U is the main source of fission tracks. However the fission-track dating of extraterrestrial samples compared with the terrestrial ones has some peculiar features due to presence of a number of other potential track sources except the spontaneous fission of 238 U, such as the spontaneous fission of presently extinct 244 Pu, heavy nuclei of cosmic rays and induced fission by cosmic ray primaries. Only tracks from the spontaneous fission of U and Pu are suitable for fission-track dating. The competing effects of these fissioning elements, whose half-lives differ by a factor of ∼50, form a basis for a fission-track chronology for samples older than ∼ 4.0 Gyr. Over small intervals in time (∼ few x10 8 yr ) the track density from spontaneous fission of 238 U is nearly constant. However, the

  10. An optimized symbiotic fusion and molten-salt fission reactor system

    International Nuclear Information System (INIS)

    Blinkin, V.L.; Novikov, V.M.

    A symbiotic fusion-fission reactor system which breeds nuclear fuel is discussed. In the blanket of the controlled thermonuclear reactor (CTR) uranium-233 is generated from thorium, which circulates in the form of ThF 4 mixed with molten sodium and beryllium fluorides. The molten-salt fission reactor (MSR) burns up the uranium-233 and generates tritium for the fusion reactor from lithium, which circulates in the form of LiF mixed with BeF 2 and 233 UF 4 through the MSR core. With a CTR-MSR thermal power ratio of 1:11 the system can produce electrical energy and breed fuel with a doubling time of 4-5 years. The system has the following special features: (1) Fuel reprocessing is much simpler and cheaper than for contemporary fission reactors; reprocessing consists simply in continuous removal of 233 U from the salt circulating in the CTR blanket by the fluorination method and removal of xenon from the MSR fuel salt by gas scavenging; the MSR fuel salt is periodically exchanged for fresh salt and the 233 U is then removed from it; (2) Tritium is produced in the fission reactor, which is a much simpler system than the fusion reactor; (3) The CTR blanket is almost ''clean''; no tritium is produced in it and fission fragment activity does not exceed the activity induced in the structural materials; (4) Almost all the thorium introduced into the CTR blanket can be used for producing 233 U

  11. R and D in Ciemat Nuclear Fission Department

    International Nuclear Information System (INIS)

    Diaz, J. L.; Diaz Arocas, P.; Gomez Briceno, D.; Gonzalez de la Huebra Gordo, A.; Gonzalez Romero, E.; Herranz Puebla, L. E.; Sola Farre, R.

    2000-01-01

    The technologically developed countries count on nuclear fission as a durable energy resource to produce electricity, facing the future by establishing research programmes to enhance the safety and extend the lifetime of the current power plants and to achieve the adequate management of radioactive waste. At the same time, the progress in the development of a new generation of reactors based in innovative safety concepts. The Nuclear fission Department has the ultimate objective of providing technical support to the Spanish nuclear reactors through applied research and development focused on improving the safety and performance of the operating power plants, and cooperating in the activities related to radioactive waste. In this context, the Departament has been organised in four R and D project covering the areas of Safety, Materials, Radioactive. (Author)

  12. 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

  13. 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

  14. Laser microsampling method for determination of retained fission gas in irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Graczyk, D.G.; Bandyopadhyay, G.; Gehl, S.M.; Hughes, J.P.; Goodspeed, H.T.

    1979-10-01

    A small ruby laser adapted to fire through a microscope is used to release fission gases from specific sites on a plane surface of an irradiated fuel specimen. Interaction of the focused laser pulse with the specimen surface results in a conical crater from which sampled material has been vaporized; the crater is surrounded by a heat-affected zone in which intergranular fracture and grain separation allow release of grain-boundary gases. Procedures for measuring the amount of krypton-85 released by laser heating and the volume of material from which the release occurred are presented. The data obtained may be used to obtain local krypton fission-gas concentrations and the intragranular/intergranular distribution

  15. The Radiological and Thermal Characteristics of Fission Waste from a Deep-Burn Fusion-Fission Hybrid (LIFE) and Implications for Repository Performance

    International Nuclear Information System (INIS)

    Shaw, H.F.; Blink, J.; Farmer, J.; Latkowski, J.; Kramer, K.

    2009-01-01

    We are studying the use of a Laser Inertial-confinement Fusion Engine (LIFE) to drive a hybrid fusion-fission system that can generate electrical power and/or burn nuclear waste. The system uses the neutrons from laser driven ICF to produce tritium and to drive nuclear reactions in a subcritical fission blanket. The fusion neutron source obviates the need for a self-sustaining chain reaction in the fission blanket. Either fissile or fertile could be used as fission fuel, thus eliminating the need for isotopic enrichment. The 'driven' system potentially allows very high levels of burnup to be reached, extracting a large fraction of the available energy in the fission fuel without the need for reprocessing. In this note, we discuss the radionuclide inventory of a depleted uranium (DU) fuel burned to greater than 95% FIMA (Fissions per Initial heavy Metal Atom), the implications for thermal management of the resulting waste, and the implications of this waste for meeting the dose standards for releases from a geological repository for high-level waste. The fission waste discussed here would be that produced by a LIFE hybrid with a 500-MW fusion source. The fusion neutrons are multiplied and moderated by a sequence of concentric shells of materials before encountering the fission fuel, and fission in this region is largely due to thermal neutrons. The fission blanket consists of 40 metric tons (MT) of DU, assumed to be in the form of TRISO-like UOC fuel particles embedded in 2-cm-diameter graphite pebbles. (It is recognized that TRISO-based fuel may not reach the high burnup of the fertile fuel considered here, and other fuel options are being investigated. We postulate the existence of a fuel that can reach >95% FIMA so that the waste disposal implications of high burnup can be assessed.) The engine and plant design considered here would receive one load of fission fuel and produce ∼2 GWt of power (fusion + fission) over its 50- to 70-year lifetime. Neutron and

  16. 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)

  17. Fission Product Yield Study of 235U, 238U and 239Pu Using Dual-Fission Ionization Chambers

    Science.gov (United States)

    Bhatia, C.; Fallin, B.; Howell, C.; Tornow, W.; Gooden, M.; Kelley, J.; Arnold, C.; Bond, E.; Bredeweg, T.; Fowler, M.; Moody, W.; Rundberg, R.; Rusev, G.; Vieira, D.; Wilhelmy, J.; Becker, J.; Macri, R.; Ryan, C.; Sheets, S.; Stoyer, M.; Tonchev, A.

    2014-05-01

    To resolve long-standing differences between LANL and LLNL regarding the correct fission basis for analysis of nuclear test data [M.B. Chadwick et al., Nucl. Data Sheets 111, 2891 (2010); H. Selby et al., Nucl. Data Sheets 111, 2891 (2010)], a collaboration between TUNL/LANL/LLNL has been established to perform high-precision measurements of neutron induced fission product yields. The main goal is to make a definitive statement about the energy dependence of the fission yields to an accuracy better than 2-3% between 1 and 15 MeV, where experimental data are very scarce. At TUNL, we have completed the design, fabrication and testing of three dual-fission chambers dedicated to 235U, 238U, and 239Pu. The dual-fission chambers were used to make measurements of the fission product activity relative to the total fission rate, as well as for high-precision absolute fission yield measurements. The activation method was employed, utilizing the mono-energetic neutron beams available at TUNL. Neutrons of 4.6, 9.0, and 14.5 MeV were produced via the 2H(d,n)3He reaction, and for neutrons at 14.8 MeV, the 3H(d,n)4He reaction was used. After activation, the induced γ-ray activity of the fission products was measured for two months using high-resolution HPGe detectors in a low-background environment. Results for the yield of seven fission fragments of 235U, 238U, and 239Pu and a comparison to available data at other energies are reported. For the first time results are available for neutron energies between 2 and 14 MeV.

  18. Fifty years with nuclear fission

    International Nuclear Information System (INIS)

    Behrens, J.W.; Carlson, A.D.

    1989-01-01

    The news of the discovery of nucler fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fiftieth anniversary of its discovery by holding a topical meeting entitled, ''Fifty years with nuclear fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent developments in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicating a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two full days of sessions (April 27 and 28) at the main sites of the NIST in Gaithersburg, Maryland. The wide range of topics covered by Volume 2 of this topical meeting included plenary invited, and contributed sessions entitled, Nuclear fission -- a prospective; reactors II; fission science II; medical and industrial applications by by-products; reactors and safeguards; general research, instrumentation, and by-products; and fission data, astrophysics, and space applications. The individual papers have been cataloged separately

  19. 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)

  20. Shell effects in fission and quasi-fission of heavy and superheavy nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Itkis, M.G. E-mail: itkis@flnr.jinr.ru; Aeystoe, J.; Beghini, S.; Bogachev, A.A.; Corradi, L.; Dorvaux, O.; Gadea, A.; Giardina, G.; Hanappe, F.; Itkis, I.M.; Jandel, M.; Kliman, J.; Khlebnikov, S.V.; Kniajeva, G.N.; Kondratiev, N.A.; Kozulin, E.M.; Krupa, L.; Latina, A.; Materna, T.; Montagnoli, G.; Oganessian, Yu.Ts.; Pokrovsky, I.V.; Prokhorova, E.V.; Rowley, N.; Rubchenya, V.A.; Rusanov, A.Ya.; Sagaidak, R.N.; Scarlassara, F.; Stefanini, A.M.; Stuttge, L.; Szilner, S.; Trotta, M.; Trzaska, W.H.; Vakhtin, D.N.; Vinodkumar, A.M.; Voskressenski, V.M.; Zagrebaev, V.I

    2004-04-05

    Results of the experiments aimed at the study of fission and quasi-fission processes in the reactions {sup 12}C+{sup 204}Pb, {sup 48}Ca+{sup 144,154}Sm, {sup 168}Er, {sup 208}Pb, {sup 244}Pu, {sup 248}Cm; {sup 58}Fe+{sup 208}Pb, {sup 244}Pu, {sup 248}Cm, and {sup 64}Ni+{sup 186}W, {sup 242}Pu are presented in the work. The choice of the above-mentioned reactions was inspired by recent experiments on the production of the isotopes {sup 283}112, {sup 289}114 and {sup 283}116 at Dubna [1],[2] using the same reactions. The {sup 58}Fe and {sup 64}Ni projectiles were chosen since the corresponding projectile-target combinations lead to the synthesis of even heavier elements. The experiments were carried out at the U-400 accelerator of the Flerov Laboratory of Nuclear Reactions (JINR, Russia), the XTU Tandem accelerator of the National Laboratory of Legnaro (LNL, Italy) and the Accelerator of the Laboratory of University of Jyvaskyla (JYFL, Finland) using the time-of-flight spectrometer of fission fragments CORSET[3] and the neutron multi-detector DEMON[4],[5]. The role of shell effects and the influence of the entrance channel on the mechanism of the compound nucleus fusion-fission and the competitive process of quasi-fission are discussed.

  1. 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 deposition of 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 behavior of simulant fission product species such as cesium iodide, cesium hydroxide and tellurium, in terms of their vapor 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 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 vapor phase, and specific data using this technique are reported

  2. Independent fission yields of Rb and Cs from thermal-neutron-induced fission of 239Pu

    International Nuclear Information System (INIS)

    Balestrini, S.J.; Forman, L.

    1975-01-01

    The relative independent fission yields of Rb and Cs from thermal-neutron-induced fission of 239 Pu have been measured on line using a mass spectrograph and thermalized neutrons from a burst reactor. Independent yields were derived by normalizing the measurements to products of chain yields and fractional independent yields, estimating the latter from measured cumulative yields of Kr and Xe. Comparing the independent yields with those from 238 U fission, the 239 Pu results show shifts in isotopic yield distribution toward lower mass for both Rb and Cs and also toward the production of more Cs and less Rb when 239 Pu is fissioned

  3. Research activities in fission chamber modeling in support of the nuclear energy industry

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, C.; Filliatre, P.; Geslot, B.; Oriol, L.; Berhouet, F.; Villard, J. F. [Commissariat a l' Energie Atomique, DEN/SPEX/LDCI, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France); Vermeeren, L. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2009-07-01

    Fission chambers are widely used in the nuclear industry. As an example, they play a major role in the control of any fission reactor and are thus regarded as a key component for ensuring their safety. They are also employed in the material testing reactors for monitoring irradiations. We have recently started a research program, the objective of which is to improve the performance of those neutron detectors in terms of lifetime, calibration and online diagnosis. In this paper, we present several studies carried out in order to model the signal delivered by a fission chamber. First, the simulation of the deposit evolution allowed us to select the most appropriate fissile material for a given spectrum and fluence. Second, we studied the impact of the bias voltage and filling gas characteristics on the charge collection time. Finally, the simulation of a pulse signal prior to amplification showed how it is important to have a satisfactory knowledge of the energy for creating ion pairs to accurately assess the signal in current or Campbelling mode. (authors)

  4. Research activities in fission chamber modeling in support of the nuclear energy industry

    International Nuclear Information System (INIS)

    Jammes, C.; Filliatre, P.; Geslot, B.; Oriol, L.; Berhouet, F.; Villard, J. F.; Vermeeren, L.

    2009-01-01

    Fission chambers are widely used in the nuclear industry. As an example, they play a major role in the control of any fission reactor and are thus regarded as a key component for ensuring their safety. They are also employed in the material testing reactors for monitoring irradiations. We have recently started a research program, the objective of which is to improve the performance of those neutron detectors in terms of lifetime, calibration and online diagnosis. In this paper, we present several studies carried out in order to model the signal delivered by a fission chamber. First, the simulation of the deposit evolution allowed us to select the most appropriate fissile material for a given spectrum and fluence. Second, we studied the impact of the bias voltage and filling gas characteristics on the charge collection time. Finally, the simulation of a pulse signal prior to amplification showed how it is important to have a satisfactory knowledge of the energy for creating ion pairs to accurately assess the signal in current or Campbelling mode. (authors)

  5. A position sensitive parallel plate avalanche fission detector for use in particle induced fission coincidence measurements

    NARCIS (Netherlands)

    Plicht, J. van der

    1980-01-01

    A parallel plate avalanche detector developed for the detection of fission fragments in particle induced fission reactions is described. The active area is 6 × 10 cm2; it is position sensitive in one dimension with a resolution of 2.5 mm. The detector can withstand a count rate of 25000 fission

  6. Physics of neutron emission in fission

    International Nuclear Information System (INIS)

    Lemmel, H.D.

    1989-06-01

    The document contains the proceedings of the IAEA Consultants' Meeting on the Physics of Neutron Emission in Fission, Mito City (Japan), 24-27 May 1988. Included are the conclusions and recommendations reached at the meeting and the papers presented by the meeting participants. These papers cover the following topics: Energy dependence of the number of fission neutrons ν-bar (3 papers), multiplicity distribution of fission neutrons (3 papers), competition between neutron and γ-ray emission (4 papers), the fission neutron yield in resonances (2 papers) and the energy spectrum of fission neutrons in experiment (9 papers), theory (4 papers) and evaluation (1 paper). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  7. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, G.

    1975-01-01

    This is the first issue of a report series on Fission Product Nuclear Data (FPND), published every six months by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). Its purpose is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields; neutron cross-section data of fission products; data related to β-, γ-decay of fission products; delayed neutron data; and fission product decay-heat. The present issue includes contributions which were received by NDS before 1 November 1975

  8. 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

  9. Fission product range effects on HEU fissile gas monitoring for UF6 gas

    International Nuclear Information System (INIS)

    Munro, J.K. Jr.; Valentine, T.E.; Perez, R.B.

    1997-01-01

    The amount of 235 U in UF 6 flowing in a pipe can be monitored by counting gamma rays emitted from fission fragments carried along by the flowing gas. Neutron sources are mounted in an annular sleeve that is filled with moderator material and surrounds the pipe. This provides a source of thermal neutrons to produce the fission fragments. Those fragments that remain in the gas stream following fission are carried past a gamma detector. A typical fragment will be quite unstable, giving up energy as it decays to a more stable isotope with a significant amount of this energy being emitted in the form of gamma rays. A given fragment can emit several gamma rays over its lifetime. The gamma ray emission activity level of a distribution of fission fragments decreases with time. The monitoring system software uses models of these processes to interpret the gamma radiation counting data measured by the gamma detectors

  10. Active-interrogation measurements of fast neutrons from induced fission in low-enriched uranium

    International Nuclear Information System (INIS)

    Dolan, J.L.; Marcath, M.J.; Flaska, M.; Pozzi, S.A.; Chichester, D.L.; Tomanin, A.; Peerani, P.

    2014-01-01

    A detection system was designed with MCNPX-PoliMi to measure induced-fission neutrons from U-235 and U-238 using active interrogation. Measurements were then performed with this system at the Joint Research Centre in Ispra, Italy on low-enriched uranium samples. Liquid scintillators measured induced fission neutrons to characterize the samples in terms of their uranium mass and enrichment. Results are presented to investigate and support the use of organic liquid scintillators with active interrogation techniques to characterize uranium containing materials. -- Highlights: • We studied low-enriched uranium using active-interrogation experiments including a deuterium–tritium neutron generator and an americium–lithium isotopic neutron source. • Liquid scintillators measured induced-fission neutrons from the active-interrogation methods. • Fast-neutron (DT) and thermal-neutron (Am–Li) interrogation resulted in the measurement of trends in uranium mass and 235 U enrichment respectively. • MCNPX-PoliMi, the Monte Carlo transport code, simulated the measured induced-fission neutron trends in the liquid scintillators

  11. Active-interrogation measurements of fast neutrons from induced fission in low-enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    Dolan, J.L., E-mail: jldolan@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Marcath, M.J.; Flaska, M.; Pozzi, S.A. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Chichester, D.L. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Tomanin, A.; Peerani, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Ispra (Italy)

    2014-02-21

    A detection system was designed with MCNPX-PoliMi to measure induced-fission neutrons from U-235 and U-238 using active interrogation. Measurements were then performed with this system at the Joint Research Centre in Ispra, Italy on low-enriched uranium samples. Liquid scintillators measured induced fission neutrons to characterize the samples in terms of their uranium mass and enrichment. Results are presented to investigate and support the use of organic liquid scintillators with active interrogation techniques to characterize uranium containing materials. -- Highlights: • We studied low-enriched uranium using active-interrogation experiments including a deuterium–tritium neutron generator and an americium–lithium isotopic neutron source. • Liquid scintillators measured induced-fission neutrons from the active-interrogation methods. • Fast-neutron (DT) and thermal-neutron (Am–Li) interrogation resulted in the measurement of trends in uranium mass and {sup 235}U enrichment respectively. • MCNPX-PoliMi, the Monte Carlo transport code, simulated the measured induced-fission neutron trends in the liquid scintillators.

  12. Fission dynamics with systems of intermediate fissility

    Indian Academy of Sciences (India)

    results concerning nuclear dissipation and fission time-scale obtained from several of these studies. In particular ... alent to the assumption that fission is delayed, namely, that the fission probability is not .... parameters to be adjusted on the experimental data. ..... (b) Time distribution of all fission events for the 132Ce nucleus.

  13. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1982-07-01

    This is the eighth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. Therefore, the IAEA cannot be held responsible for the information contained nor for any consequences resulting from the use of this information. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); Neutron reaction cross sections of fission products; Data related to the radioactive decay of fission products; Delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The seventh issue of this series has been published in July 1981 as INDC(NDS)-116. The present issue includes contributions which were received by NDS between 1 August 1981 and 15 June 1982

  14. First-Principles Quantum Dynamics of Singlet Fission: Coherent versus Thermally Activated Mechanisms Governed by Molecular π Stacking

    Science.gov (United States)

    Tamura, Hiroyuki; Huix-Rotllant, Miquel; Burghardt, Irene; Olivier, Yoann; Beljonne, David

    2015-09-01

    Singlet excitons in π -stacked molecular crystals can split into two triplet excitons in a process called singlet fission that opens a route to carrier multiplication in photovoltaics. To resolve controversies about the mechanism of singlet fission, we have developed a first principles nonadiabatic quantum dynamical model that reveals the critical role of molecular stacking symmetry and provides a unified picture of coherent versus thermally activated singlet fission mechanisms in different acenes. The slip-stacked equilibrium packing structure of pentacene derivatives is found to enhance ultrafast singlet fission mediated by a coherent superexchange mechanism via higher-lying charge transfer states. By contrast, the electronic couplings for singlet fission strictly vanish at the C2 h symmetric equilibrium π stacking of rubrene. In this case, singlet fission is driven by excitations of symmetry-breaking intermolecular vibrations, rationalizing the experimentally observed temperature dependence. Design rules for optimal singlet fission materials therefore need to account for the interplay of molecular π -stacking symmetry and phonon-induced coherent or thermally activated mechanisms.

  15. 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

  16. 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.

  17. Overview on international experimental programmes on power ramping and fission gas release

    International Nuclear Information System (INIS)

    Knaab, H.; Lang, P.M.; Mogard, H.

    1983-01-01

    During the last years a number of internationally sponsored experimental programmes were initiated to study the LWR fuel behaviour during ramping and fission gas release at higher burnup levels. Common interest and the limited availability of experimental facilities and appropriate test fuel rods have led to valuable cooperation of many organizations throughout the nuclear community. These programmes are performed by the experimental staff from research centers with their experimental facilities. Fuel vendors and several utilities contribute by supply and irradiation of test fuel rods. The aim of this paper is to provide a synopsis of the following programmes: Studsvik Projects: Interramp, Overramp, Superramp, Demoramp I and II; Petten, High Burnup PWR Ramp Test Programme; Mol, Tribulation Programme; BNWL, High Burnup Effects Programme; Risoe Fission Gas Project; Related tasks within the OECD Halden Reactor Project. The objectives of the programme, their work scope and main results will be summarized on the basis of presently available information. An outlook to future proposed programmes will be given. (author)

  18. Overview on international experimental programmes on power ramping and fission gas release

    International Nuclear Information System (INIS)

    Knaab, H.; Lang, P.M.

    1985-01-01

    During the last few years a number of internationally sponsored experimental programmes have been initiated to study LWR fuel behaviour during ramping and fission gas release at high burnup levels. Common interest and the limited availability of experimental facilities and appropriate test fuel rods have led to valuable cooperation between many organizations throughout the nuclear community. These programmes are carried out by experienced staff from research centres using the centres' experimental facilities. Fuel vendors and several utilities contribute by supplying and irradiating the test fuel rods. The aim of this paper is to provide a synopsis of the following programmes: (a) Studsvik Projects: Interramp, Demoramp I and II, Overramp, Superramp; (b) Petten, High Burnup PWR Ramp Test Programme; (c) Mol, Tribulation Programme; (d) BNWL, High Burnup Effects Programme; (e) Riso Fission Gas Project; and (f) related tasks within the OECD Halden Reactor Project. The objectives of the programmes, their scope and the main results will be summarized. An overview of proposed future programmes will be given. (author)

  19. Some applications of fission-based testing capabilities in the development of fusion technology

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.; Schmunk, R.E.; Takata, M.L.; Watts, K.D.

    1981-10-01

    The testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for fusion testing when they do become available. Fission testing is capable of filling many gaps in fusion reactor design information, and should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, work was performed on the irradiation of magnet insulators. This work is continuing with an improved test environment. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, a conceptual design is presented for a fission-based Integrated Test Facility (ITF), which can accommodate entire wall/blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

  20. 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