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)

Fission in the landscape of heaviest elements: Some recent examples

International Nuclear Information System (INIS)

Khuyagbaatar, J.; Yakushev, A.; Düllmann, Ch.E.; Ackermann, D.; Andersson, L.-L.; Block, M.; Brand, H.; Even, J.; Forsberg, University; Hartmann, W.; Herzberg, R.-D.; Heßberger, F.P.; Hoffmann, J.; Hübner, A.; Jäger, E.; Jeppsson, J.; Kindler, B.; Kratz, J.V.; Krier, J.; Kurz, N.; Lommel, B.; Maiti, M.; Minami, S.; Rudolph, D.; Runke, J.; Sarmiento, L.G.; Schädel, M.; Schausten, B.; Steiner, J.; Heidenreich, T. Torres De; Uusitalo, J.; Wiehl, N.; Yakusheva, V.

2016-01-01

The fission process still remains a main factor that determines the stability of the atomic nucleus of heaviest elements. Fission half-lives vary over a wide range, 10"−"1"9−10"2"4 s. Present experimental techniques for the synthesis of the superheavy elements that usually measure α-decay chains are sensitive only in a limited range of half-lives, often 10"−"5−10"3 s. In the past years, measurement techniques for very short-lived and very long-lived nuclei were significantly improved at the gas-filled recoil separator TASCA at GSI Darmstadt. Recently, several experimental studies of fission-related phenomena have successfully been performed. In this paper, results on "2"5"4"−"2"5"6Rf and "2"6"6Lr are presented and corresponding factors for retarding the fission process are discussed.

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

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.

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

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.

''In situ'' investigations of the radioactive fissionable element infiltration and retention in different soils

International Nuclear Information System (INIS)

Oncescu, M.; Danis, A.; Sahagia, M.; Negrescu, C.; Bobe, M.; Balanescu, P.; Burcescu, M.; Tautu, N.

1980-01-01

''In situ'' investigations of the natural and forced infiltration and retention of the fissionable elements from a liquid residue in several natural compacted soils and compacted clays are presented. The velocities and flow rates for different stages of the residue infiltration are determined. The retention of the fissionable elements by variation of the fissionable element concentration with the distance from the place of the residue depot is investigated. (author)

The behaviour of spherical HTR fuel elements under accident conditions

Energy Technology Data Exchange (ETDEWEB)

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

1985-07-01

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

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)

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.)

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

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

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

What can we learn from the fission time of the super-heavy elements?

OpenAIRE

Boilley, D.; Marchix, A.; Wilgenbus, D.; Lallouet, Y.; Gimbert, F.; Abe, Y.

2007-01-01

International audience; Recent experiments performed at GANIL with a crystal blocking technique have shown direct evidences of long fission times in the Super-Heavy Elements (SHE) region. Aimed to localize the SHE island of stability, can these experiments give access to the fission barrier and then to the shell-correction energy? In this paper, we calculate the fission time of heavy elements by using a new code, KEWPIE2, devoted to the study of the SHE.We also investigate the effect of poten...

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

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.

Rapid separation of individual rare-earth elements from fission products

International Nuclear Information System (INIS)

Baker, J.D.; Gehrke, R.J.; Greenwood, R.C.; Meikrantz, D.H.

1980-01-01

A microprocessor-controlled radiochemical separation system has been developed to rapidly separate rare-earth elements from gross fission products. The system is composed of two high performance liquid chromatography columns coupled in series by a stream-splitting injection valve. The first column separates the rare-earth group by extraction chromatography using dihexyldiethylcarbamylmethylenephosphonate (DHDECMP) adsorbed on Vydac C 8 resin. The second column isolates the individual rare-earth elements by cation exchange using Aminex A-9 resin with α-hydroxyisobutyric acid (α-HIBA) as the eluent. With this system, fission-product rare-earth isotopes with half-lives as short as three minutes have been studied

Finite element simulation of fission gas release and swelling in UO2 fuel pellets

International Nuclear Information System (INIS)

Denis, Alicia C.

1999-01-01

A fission gas release model is presented, which solves the atomic diffusion problem with xenon and krypton elements tramps produced by uranium fission during UO 2 nuclear fuel irradiation. The model considers intra and intergranular precipitation bubbles, its re dissolution owing to highly energetic fission products impact, interconnection of intergranular bubbles and gas sweeping by grain border in movement because of grain growth. In the model, the existence of a thermal gradient in the fuel pellet is considered, as well as temporal variations of fission rate owing to changes in the operation lineal power. The diffusion equation is solved by the finite element method and results of gas release and swelling calculation owing to gas fission are compared with experimental data. (author)

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

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)

Fission product release from HTGR coated microparticles and fuel elements

International Nuclear Information System (INIS)

Gusev, A.A.; Deryugin, A.I.; Lyutikov, R.A.; Chernikov, A.S.

1991-01-01

The article presents the results of the investigation of fission products release from microparticles with UO 2 core and five-layer HII PyC- and SiC base protection layers of TRICO type as well as from spherical fuel elements based thereon. It is shown that relative release of short-lived xenon and crypton from microparticles does not exceed (2-3) 10 -7 . The release of gaseous fission products from fuel elements containing no damaged coated microparticles, is primarily determined by the contamination of matrix graphite with fuel. An analytical dependence is derived, the dependence described the relation between structural parameters of coated microparticles, irradiation conditions and fuel burnup at which depressurization of coated microparticles starts

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

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

Nuclear fission and the transuranium elements

International Nuclear Information System (INIS)

Seaborg, G.T.

1989-02-01

Many of the transuranium elements are produced and isolated in large quantities through the use of neutrons furnished by nuclear fission reactions: plutonium (atomic number 94) in ton quantities; neptunium (93), americium (95), and curium (96) in kilogram quantities; berkelium (97) in 100 milligram quantities; californium (98) in gram quantities; and einsteinium (99) in milligram quantities. Transuranium isotopes have found many practical applications---as nuclear fuel for the large-scale generation of electricity, as compact, long-lived power sources for use in space exploration, as means for diagnosis and treatment in the medical area, and as tools in numerous industrial processes. Of particular interest is the unusual chemistry and impact of these heaviest elements on the periodic table. This account will feature these aspects. 9 refs., 5 figs

Nuclear fission and the transuranium elements

Energy Technology Data Exchange (ETDEWEB)

Seaborg, G.T.

1989-02-01

Many of the transuranium elements are produced and isolated in large quantities through the use of neutrons furnished by nuclear fission reactions: plutonium (atomic number 94) in ton quantities; neptunium (93), americium (95), and curium (96) in kilogram quantities; berkelium (97) in 100 milligram quantities; californium (98) in gram quantities; and einsteinium (99) in milligram quantities. Transuranium isotopes have found many practical applications---as nuclear fuel for the large-scale generation of electricity, as compact, long-lived power sources for use in space exploration, as means for diagnosis and treatment in the medical area, and as tools in numerous industrial processes. Of particular interest is the unusual chemistry and impact of these heaviest elements on the periodic table. This account will feature these aspects. 9 refs., 5 figs.

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)

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)

The Metabolic Properties of the Fission Products and Actinide Elements

Energy Technology Data Exchange (ETDEWEB)

Hamilton M.D., J.G.

1948-03-01

An investigation of the assimilation, distribution, retention, an excretion of the fission products and actinide elements in the rat has been conducted at the Crocker Radiation Laboratory, University of California, Berkeley, California. These studies were initiated October 15, 1942, and are continuing at the present time. An extensive survey has been made of the metabolism of twenty-two different radio elements in the rat.

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

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

Fuel element tomography by gammametry

International Nuclear Information System (INIS)

Simonet, G.; Pineira, T.

1982-03-01

As from transversal gamma determinations of a cylindrical fuel element, the TOMOGAM program reconstitutes the distribution of fission products in a section. This direct, fast and non destructive method, makes it possible to have access to the behaviour of the fuel at any time: - the soluble fission products in the matrix represent the fuel itself and the distribution of the fissions, - the migrating elements inform on the temperature reached in accordance with the permitted powers, - the volatile nuclides build up in particular points where physical-chemical phenomena of fuel-cladding interaction are liable to corrode the latter. Hence, gamma spectrometry extends its possibilities of analysis relative to the performance of reactor elements [fr

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)

Use of ELOCA.Mk5 to calculate transient fission product release from CANDU fuel elements

International Nuclear Information System (INIS)

Walker, J.R.; de Vaal, J.W.; Arimescu, V.I.; McGrady, T.G.; Wong, C.

1992-04-01

A change in fuel element power output, or a change in heat transfer conditions, will result in an immediate change in the temperature distribution in a fuel element. The temperature distribution change will be accompanied by concomitant changes in fuel stress distribution that lead, in turn, to a release of fission products to the fuel-to-sheath gap. It is important to know the inventory of fission products in the fuel-to-sheath gap, because this inventory is a major component of the source term for many postulated reactor accidents. ELOCA.Mk5 is a FORTRAN-77 computer code that has been developed to estimate transient releases to the fuel-to-sheath gap in CANDU reactors. ELOCA.Mk5 is an integration of the FREEDOM fission product release model into the ELOCA fuel element thermo-mechanical code. The integration of FREEDOM into ELOCA allows ELOCA.Mk5 to model the feedback mechanisms between the fission product release and the thermo-mechanical response of the fuel element. This paper describes the physical model, gives details of the ELOCA.Mkt code, and describes the validation of the model. We demonstrate that the model gives good agreement with experimental results for both steady state and transient conditions

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

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.)

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)

Distribution of fission products in Peach Bottom HTGR fuel element E01-01

International Nuclear Information System (INIS)

Wichner, R.P.; Dyer, F.F.; Martin, W.J.; Fairchild, L.L.

1978-10-01

The fifth in a projected series of six postirradiation examinations of Peach Bottom High-Temperature Gas-Cooled Reactor driver fuel elements is described. The element analyzed received an equivalent of 897 full-power days of irradiation prior to the scheduled termination of Core 2 operation. The examination procedures emphasized the determination of fission product distributions in the graphite portions of the fuel element. Continuous axial scans indicated a 137 Cs inventory of 20.3 Ci in the graphite sleeve and 8.1 Ci in the spine at the time of element withdrawal from the core. In addition, the nuclides 134 Cs, /sup 110 m/Ag, 60 Co, and 154 Eu were found in the graphite portions of the fuel element in significant amounts. Radial distributions of these nuclides plus the beta-emitters 3 H, 14 C, and 90 Sr were obtained at four axial locations of the fueled region of the element sleeve and two axial locations of the element spine. The radial dissection was accomplished by use of a manipulator-operated lathe in a hot cell. In addition to fission product distributions, the appearance of the component parts of the element was recorded photographically, fuel compact and graphite dimensions were recorded at numerous locations, and metallographic examinations of the fuel were performed

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)

Computer modelling of water reactor fuel element performance and life time

International Nuclear Information System (INIS)

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

1983-01-01

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

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)

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

Fission energy of uranium isotopes and transuranium elements

International Nuclear Information System (INIS)

Nemirovskij, P.Eh.; Manevich, L.G.

1981-01-01

A comparison is made between the prompt fission energy, Esub(pr), calculated from the mass and binding energy spectrum and the Esub(pr) value obtained from the experimental data on the kinetic energy of fragments, the energy of prompt neutrons and prompt γ-quanta. Basing on the data on β-decay chains of fission fragments, the energies of neutrinos, γ-quanta and β-electrons are obtained, which permits to calculate the actual energy released during fission. The calculations are performed for thermal neutron-induced fission, fast-neutron induced fission and for fission after bombardment with 14 MeV neutrons. The available experimental data on the fission fragment kinetic energy, prompt γ-quanta energy and fission neutron energy are presented. The comparison of the Esub(pr) values obtained experimentally for the thermal-neutron-induced fission with the calculated Esub(pr) value shows that for 233 U, 239 Pu, 241 Pu the agreement is rather favourable. For 235 U the agreement is within the error limits. As to the Esub(pr) values for the fast-neutron-induced fission, the agreement between the calculated and experimental data for all nuclides is quite good

'Masurium' and the 'early transuranium elements' or how discovery of nuclear fission was not clearly seen

International Nuclear Information System (INIS)

Keller, C.

1988-01-01

Fifty years after the discovery of fission, the scientific community is aware that this type of nuclear reaction could have been discovered more than a decade earlier. Noddack, Tacke and Berg announced in 1925 the discovery of elements Z = 43 (masurium) and rhenium (Z = 75), the first one could be detected only in U-bearing minerals. A recent re-examination by P.H.M. von Assche of the published data clearly showed that the original claim for element Z = 43 of the authors in 1925 was correct and, therefore, they detected not only element Z = 43 but also the first fission product. Because this discovery of element Z = 43 could not be repeated by other authors as that time, the scientific credibility of Noddack-Tacke was very low in order to give credit to her proposal that the 'early' transuranium elements by Enrico Fermi might also be fragments of known (lighter) elements. Enrico Fermi in 1934 obtained these 'early' (and as we today know: wrong) transuranium isotopes by irradiation of uranium with neutrons. A 'wrong' periodic system in the thirties which placed Th, Pa and U as 6d-elements and not as 5f-actinides chemically helped to consider these fission products as transuranium elements Z = 93/94. In 1937/38 I. Curie and P. Savitch discovered an 'actinium-nuclide' with 3,5 h half-life which, however, had properties similar to lanthanium and not to actinium, as they stated. (orig.) [de

An optimization on strontium separation model for fission products (inactive trace elements) using artificial neural networks

International Nuclear Information System (INIS)

Moosavi, K.; Setayeshi, S.; Maragheh, M.Gh.; Ahmadi, S.J.; Kardan, M.R.; Banaem, L.M.

2009-01-01

In this study, an experimental design using artificial neural networks for an optimization on the strontium separation model for fission products (inactive trace elements) is investigated. The goal is to optimize the separation parameters to achieve maximum amount of strontium that is separated from the fission products. The result of the optimization method causes a proper purity of Strontium-89 that was separated from the fission products. It is also shown that ANN may be establish a method to optimize the separation model.

New approach to description of fusion-fission dynamics in super-heavy element formation

International Nuclear Information System (INIS)

Zagrebaev, V.I.

2002-01-01

A new mechanism of the fusion-fission process for a heavy nuclear system is proposed, which takes place in the (A 1 , A 2 ) space, where A 1 and A 2 are two nuclei, surrounded by a certain number of shared nucleons ΔA. The nuclei A 1 and A 2 gradually lose (or acquire) their individualities with increasing (or decreasing) a number of collectivized nucleons ΔA. The driving potential in the (A 1 , A 2 ) space is derived, which allows the calculation of both the probability of the compound nucleus formation and the mass distribution of fission and quasi-fission fragments in heavy ion fusion reactions. The cross sections of super-heavy element formation in the 'hot' and 'cold' fusion reactions have been calculated up to Z CN =118. (author)

Fission in R-processes Elements (FIRE) - Annual Report: Fiscal Year 2017

Energy Technology Data Exchange (ETDEWEB)

Schunck, Nicolas [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2017-10-18

The goal of the FIRE topical collaboration in nuclear theory is to determine the astrophysical conditions of the rapid neutron capture process (r-process), which is responsible for the formation of heavy elements. This will be achieved by including in r-process simulations the most advanced models of fission (spontaneous, neutron-induced, beta-delayed) that have been developed at LLNL and LANL. The collaboration is composed of LLNL (lead) and LANL for work on nuclear data (ground-state properties, fission, beta-decay), BNL for nuclear data management, and the university of Notre Dame and North Carolina State University for r-process simulations. Under DOE/NNSA agreement, both universities receive funds from the DOE Office of Science, while national laboratories receive funds directly from NA221.

Calculation of burnup and power dependence on fission gas released from PWR type reactor fuel element

International Nuclear Information System (INIS)

Edy-Sulistyono

1996-01-01

Burn up dependence of fission gas released and variation power analysis have been conducted using FEMXI-IV computer code program for Pressure Water Reactor Fuel During steady-state condition. The analysis result shows that the fission gas release is sensitive to the fuel temperature, the increasing of burn up and power in the fuel element under irradiation experiment

Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

International Nuclear Information System (INIS)

Verfondern, K.; Mueller, D.

1991-01-01

Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

Diffusion of Fission Product Elements in Compacted Bentonite

International Nuclear Information System (INIS)

Pratomo-Budiman-Sastrowardoyo; Dewi-Susilowati; Dadang-Suganda

2000-01-01

Study on diffusion of fission product in compacted bentonite has been conducted. The information about mobilities of these elements have been obtained from the studies resulted in many countries. It is presented that the diffusion coefficient was varied by the function of solution phase condition as well as the nature of bentonite. It is also showed that the diffusion coefficient decreased by the increasing of density, as well as the increasing of montmorillonite content in bentonite. The ratio of bentonite/silica-sand used, was related to the increasing of elements mobility. In many case variation of diffusion coefficient was related to the variation of pH, redox condition, and the presence of complex ant in solution phase. The lower diffusion coefficient could give the higher retardation factor, which is a favorable factor to retard the radionuclides release from a disposal facility to geosphere. (author)

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

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)

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.

Separation of transuranium elements and fission products from medium activity aqueous liquid wastes

International Nuclear Information System (INIS)

Gompper, K.; Kunze, S.; Eden, G.; Loesch, G.; Zemski, C.

1986-01-01

In the course of work performed between January 1981 and June 1985 on the separation of TRU elements and fission products three liquid alpha containing waste streams were treated: - medium level waste solutions, - waste solutions from the acid digestion of burnable alpha containing solid residues, - waste solutions from mixed oxide fuel element fabrication. The method of separation was initially developed and optimized with simulating substances. Subesequently it was tested with real waste solutions

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.)

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)

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

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.

Operational monitoring of the release behaviour of the AVR core

International Nuclear Information System (INIS)

Wawrzik, U.; Ivens, G.

1985-01-01

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

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

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

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)

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.

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

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

Two reports: (i) Correlation properties of delayed neutrons from fast neutron induced fission. (ii) Method and set-up for measurements of trace level content of heavy fissionable elements based on delayed neutron counting

International Nuclear Information System (INIS)

Piksaikin, V.M.; Isaev, S.G.; Goverdovski, A.A.; Pshakin, G.M.

1998-10-01

The document includes the following two reports: 'Correlation properties of delayed neutrons from fast neutron induced fission' and 'Method and set-up for measurements of trace level content of heavy fissionable elements based on delayed neutron counting. A separate abstract was prepared for each report

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.

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.)

Process for the extraction of fission products

International Nuclear Information System (INIS)

Anav, M.; Chesne, A.; Leseur, A.; Miquel, P.; Pascard, R.

1979-01-01

A process is described for the extraction of fission products contained in irradiated nuclear fuel elements which have been subject to a temperature of at least 1200 0 C during their irradiation prior to dissolving the fuel by the wet process. After mechanically treating the elements in order to decan and/or cut them they are brought into contact with water in order to pass the fission products into aqueous solution. The treated elements are then separated from the thus obtained aqueous solution. At least one of the fission products is then recovered from the aqueous solution. The fission products are iodine, cesium, rubidium and tritium

Fuel element structure - design, production and operational behaviour

International Nuclear Information System (INIS)

Pott, G.; Dietz, W.

1985-01-01

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

Evidence for bimodal fission in the heaviest elements

International Nuclear Information System (INIS)

Hulet, E.K.; Wild, J.F.; Lougheed, R.W.

1987-08-01

We have measured the mass and kinetic-energy partitioning in the spontaneous fission of five heavy nuclides: 258 Fm, 259 Md, 260 Md 258 No, and 260 [104]. Each was produced by heavy-ion reactions with either 248 Cm, 249 Bk, or 254 Es targets. Energies of correlated fragments from the isotopes with millisecond half lives, 258 No and 260 [104], were measured on-line by a special rotating-wheel instrument, while the others were determined off-line after mass separation. All fissioned with mass distributions that were symmetric. Total-kinetic-energy distributions peaked near either 200 or 235 MeV. Surprisingly, because only a single Gaussian energy distribution had been observed previously in actinide fission, these energy distributions were skewed upward or downward from the peak in each case, except for 260 [104], indicating a composite of two energy distributions. We were able to fit accurately two Gaussian curves to the gross energy distributions from the four remaining nuclides. From the multiple TKE distributions and the shapes of the mass distributions, we conclude that there is a low-energy fission component with liquid-drop characteristics which is admixed with a much higher-energy component due to closed fragment shells. We now have further evidence for this conclusion from measurements of the neutron multiplicity in the spontaneous fission of 260 Md. 25 refs., 9 figs

Fusion-fission dynamics and synthesis of the superheavy elements

International Nuclear Information System (INIS)

Abe, Yasuhisa

2003-01-01

Experiments of fusion-fission reactions clarify that the life time of nuclear fission is much longer than that expected from Bohr-Wheeler formula from the measurements of multiplicities of neutrons, gamma rays etc. emitted prior scission, and thereby appear to require a dynamical treatment of the process. Following the pioneering work by Kramers with the dissipation- fluctuation dynamics, the fissioning degree of freedom is described with the viewpoint of Brownian motion under incessant interactions with the heat bath particles, i.e., with nucleons in thermal equilibrium, in the present case. In the dynamical description the fission width is no more constant in time, but has a transient feature, as well as the reduction factor, the so-called Kramers factor. Both result in a longer life time, consistent with anomalous multiplicities measured. In the fusion process, Coulomb barriers play a crucial role in lighter heavy ion systems, but in very heavy systems it is known that there exists a hindrance in fusion. That is, the Coulomb barrier is not enough for determination of fusion probability, but an extra-energy above the barrier height is required for the system to fuse. This is understood by the properties of the Liquid Drop Model. After overcoming the Coulomb barrier, the ions touch with each other. But the united system, i.e., the pear-shaped configuration is located outside of the conditional saddle point or of the ridgeline. Therefore, in order to form the spherical compound nucleus, the system has to overcome one more barrier. Naturally, in such a situation, the kinetic energy carried in by the incident projectile has been more or less dissipated, i.e., the composite system is heated up. Thus, the shape evolution toward the spherical shape or toward the re-separation can be considered as a Brownian motion with the heat bath inside. The present author et al. have proposed the two-step model for fusion of massive heavy-ion systems where the fusion probability is

Calculated fission properties of the heaviest elements

International Nuclear Information System (INIS)

Moeller, P.; Nix, J.R.; Swiatecki, W.J.

1986-09-01

A quantitative calculation is presented that shows where high-kinetic-energy symmetric fission occurs and why it is associated with a sudden and large decrease in fission half-lives. The study is based on calculations of potential-energy surfaces in the macroscopic-microscopic model and a semi-empirical model for the nuclear inertia. For the macroscopic part a Yukawa-plus-exponential model is used and for the microscopic part a folded-Yukawa single-particle potential is used. The three-quadratic-surface parameterization generates shapes for which the potential-energy surfaces are calculated. The use of this parameterization and the use of the finite-range macroscopic model allows for the study of two touching spheres and similar shapes. The results of the calculations in terms of potential-energy surfaces and fission half-lives are presented for heavy even nuclei. The surfaces are displayed in the form of contour diagrams as functions of two moments of the shape. 53 refs., 15 figs., 1 tab

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)

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

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

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

Release of radioactive fission products from BN-600 reactor untight fuel elements

International Nuclear Information System (INIS)

Osipov, S.L.; Tsikunov, A.G.; Lisitsin, E.C.

1996-01-01

The experimental data on the release of radioactive fission products from BN-600 reactor untight fuel elements are given in the report. Various groups of radionuclides: inert gases Xe, Kr, volatile Cs, J, non-volatile Nb, and La are considered. The results of calculation-experimental study of transfer and distribution of radionuclides in the reactor primary circuit, gas system and sodium coolant are considered. It is shown that some complex radioactivity transfer processes can be described by simple mathematical models. (author)

Study of a device for the direct measurement of the fission gas pressure inside an in-pile fuel element; Etude d'un dispositif pour la mesure directe de la pression des gaz de fission a l'interieur d'un element combustible en pile

Energy Technology Data Exchange (ETDEWEB)

Lavaud, B; Uschanoff, S [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The fission gas pressure inside a fuel element made of a refractory fuel constitutes an important limiting factor for the burn-up. Although it is possible to calculate approximately the volume of gas produced outside the fuel during its life-time; it is nevertheless very difficult to evaluate the pressure since the volume allowed to the fission gases, as well as their temperature are known only very approximately. This physical value, which is essential for the technologist, can only be known by direct in-pile measurement of the pressure. The report describes the equipment which has been developed for this test. (authors) [French] La pression des gaz de fission a l'interieur d'un element combustible a combustible refractaire constitue une des limitations importantes du taux de combustion. Si on peut approcher par calcul la determination du volume, des gaz degages hors du combustible au cours de sa vie, il est par contre tres difficile d'evaluer la pression car le volume alloue aux gaz de fission et leur temperature sont tres mal connus. Cette donnee essentielle pour le technologue ne peut etre atteinte que par une mesure directe en pile de la pression. Le rapport decrit l'appareillage qui a ete mis au point pour cet essai. (auteurs)

Stability of trans-fermium elements at high spin: Measuring the fission barrier of 254No

International Nuclear Information System (INIS)

Henning, Greg

2012-01-01

Super heavy nuclei provide opportunities to study nuclear structure near three simultaneous limits: in charge Z, spin I and excitation energy E*. These nuclei exist only because of a fission barrier, created by shell effects. It is therefore important to determine the fission barrier and its spin dependence B f (I), which gives information on the shell energy E(shell)(I). Theoretical calculations predict different fission barrier heights from B f (I = 0) = 6.8 MeV for a macro-microscopic model to 8.7 MeV for Density Functional Theory calculations using the Gogny or Skyrme interactions. Hence, a measurement of B f provides a test for theories.To investigate the fission barrier, an established method is to measure the rise of fission with excitation energy, characterized by the ratio of decay widths Γ(fission)/Γ(total), using transfer reactions. However, for heavy elements such as 254 No, there is no suitable target for a transfer reaction. We therefore rely on the complementary decay widths ratio Γ γ /Γ(fission) and its spin dependence, deduced from the entry distribution (I, E*).Measurements of the gamma-ray multiplicity and total energy for 254 No have been performed with beam energies of 219 and 223 MeV in the reaction 208 Pb( 48 Ca,2n) at ATLAS (Argonne Tandem Linac Accelerator System). The 254 No gamma rays were detected using the Gammasphere array as a calorimeter - as well as the usual high resolution γ-ray detector. Coincidences with evaporation residues at the Fragment Mass Analyzer focal plane separated 254 No gamma rays from those from fission fragments, which are ≥ 10 6 more intense. From this measurement, the entry distribution - i.e. the initial distribution of I and E* - is constructed. Each point (I,E*) of the entry distribution is a point where gamma decay wins over fission and, therefore, gives information on the fission barrier. The measured entry distributions show an increase in the maximum spin and excitation energy from 219 to 223 Me

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

Fission product release from defected nuclear reactor fuel elements

International Nuclear Information System (INIS)

Lewis, B.J.

1983-01-01

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

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

Fission fragment mass and angular distributions

Indian Academy of Sciences (India)

2015-07-22

Jul 22, 2015 ... Synthesis of heavy and superheavy elements is severely hindered by fission and fission-like processes. The probability of these fission-like, non-equilibrium processes strongly depends on the entrance channel parameters. This article attempts to summarize the recent experimental findings and classify the ...

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

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.

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

Radiation resistance of pyrocarbon-boned fuel and absorbing elements for HTGR

International Nuclear Information System (INIS)

Gurin, V.A.; Konotop, Yu.F.; Odejchuk, N.P.; Shirochenkov, S.D.; Yakovlev, V.K.; Aksenov, N.A.; Kuprienko, V.A.; Lebedev, I.G.; Samsonov, B.V.

1990-01-01

In choosing the reactor type, problems of nuclear and radiation safety are outstanding. The analysis of the design and experiments show that HTGR type reactors helium cooled satisfy all the safety requirements. It has been planned in the Soviet Union to construct two HTGR plants, VGR-50 and VG-400. Later it was decided to construct an experimental plant with a low power high temperature reactor (VGM). Spherical uranium-graphite fuel elements with coated fuel particles are supposed to be used in HTGR core. A unique technology for producing spherical pyrocarbon-bound fuel and absorbing elements of monolithic type has been developed. Extended tests were done to to investigate fuel elements behaviour: radiation resistance of coated fuel particles with different types of fuel; influence of the coated fuel particles design on gaseous fission products release; influence of non-sphericity on coated fuel particle performance; dependence of gaseous fission products release from fuel elements on the thickness of fuel-free cans; confining role of pyrocarbon as a factor capable of diminishing the rate of fission products release; radiation resistance of spherical fuel elements during burnup; radiation resistance of spherical absorbing elements to fast neutron fluence and boron burnup

Observation of new spontaneous fission activities from elements 100 to 105

International Nuclear Information System (INIS)

Somerville, L.P.

1982-03-01

Several new Spontaneous Fission (SF) activities have been found. No definite identification could be made for any of the new SF activities; however, half-lives and possible assignments to element-104 isotopes consistent with several cross bombardments include 257 Rf(3.8 s, 14% SF), 258 Rf(13 ms), 259 Rf(approx. 3 s, 8% SF), 260 Rf(approx. 20 ms), and 262 Rf(approx. 50 ms). The 80-ms SF activity claimed by the Dubna group for the discovery of element 104 ( 260 104) was not observed. A difficulty exists in the interpretation that 260 Rf is a approx. 20-ms SF activity: in order to be correct, for example, the SF activities with half-lives between 14 and 24 ms produced in the reactions 109- to 119-MeV 18 O + 248 Cm, 88- to 100-MeV 15 N + 249 Bk, and 96-MeV 18 O + 249 Cf must be other nuclides due to their large production cross sections, or the cross sections for production of 260 Rf must be enhanced by unknown mechanisms. Based on calculated total production cross sections a possible approx. 1% electron-capture branch in 258 Lr(4.5 s) to the SF emitter 258 No(1.2 ms) and an upper limit of 0.05% for SF branching in 254 No(55 s) were determined. Other measured half-lives from unknown nuclides produced in respective reactions include approx. 1.6 s ( 18 O + 248 CM), indications of a approx. 47-s SF activity (75-MeV 12 C + 249 Cf), and two or more SF activities with 3 s less than or equal to T/sub 1/2/ less than or equal to 60 s ( 18 O + 249 Bk). The most exciting conclusion of this work is that if the tentative assignments to even-even element 104 isotopes are correct, there would be a sudden change in the SF half-life systematics at element 104 which has been predicted theoretically and attributed to the disappearance of the second hump of the double-humped fission barrier

Irradiation of a CANDU UOsub(2) fuel element with twenty-three machined slits cut through the zircaloy sheath

International Nuclear Information System (INIS)

DaSilva, R.L.

1984-09-01

A CANDU fuel element was purposely defected, exposing a minimum UOsub(2) fuel stack area of 272 mmsup(2), by machining 23 longitudinal slits through the Zircaloy-4 sheathing. The element was then irradiated in the X-2 loop of the NRX reactor for a period of 14.64 effective full power days at a linear heat rating of 48 kW/m to investigate the relationship between fission product release and UOsub(2) oxidation behaviour in an element with minimal fuel-to-gap fission gas trapping. The fission product releases, as measured by on-line gamma-ray spectroscopy, revealed that the noble gases and radioiodines are both released from the UOsub(2) fuel matrix directly to the coolant via simple diffusion kinetics, and that their diffusivities in hyperstoichiometric UOsub(2) are approximately equal. The oxidation of UOsub(2) to the higher states UOsub(2+x), Usub(4)Osub(9) and Usub(3)Osub(8), was accompanied by substantial fuel swelling and sheath deformation preferentially located in the lower powered end of the element. The spalling and erosion behaviour of the fuel pellets was correlated to the rate of fuel oxidation

Pressure due to fission gases in a fuel element circulating in a reactor

International Nuclear Information System (INIS)

Fonteray, Jean

1965-01-01

This document states calculation hypotheses and methods used to assess pressures due to fission gases in a fuel element moving in a reactor channel in the reverse direction with respect to the cooling fluid. The calculation comprises the calculation of the temperature in the fuel rod, of the reduced diffusion coefficient, of the diffused gas fraction, of the pressure. The appendix describes the use of the SPM 076 software: input data, output results, computing time [fr

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Tight connection between fission gas discharge channels

International Nuclear Information System (INIS)

Jung, W.; Peehs, M.; Rau, P.; Krug, W.; Stechemesser, H.

1978-01-01

The invention is concerned with the tight connection between the fission gas discharge channel, leading away from the support plate of a gas-cooled reactor, and the top of the fuel element suspended from this support plate. The closure is designed to be gas-tight for the suspended as well as for the released fuel element. The tight connection has got an annular body resting on the core support plate in the mouth region of the fission gas discharge channel. This body is connected with the fission gas discharge channel in the fuel element top fitting via a gas-tight part and supported by a compression spring. Care is taken for sealing if the fuel element is removal. (RW) [de

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)

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

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)

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

International Nuclear Information System (INIS)

1973-01-01

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

Sommerfeld-Watson transformation for nuclear fission

International Nuclear Information System (INIS)

Alexandru, G.

1978-01-01

It is proved that the fission matrix element can be written like a Sommerfeld-Watson relation. This leads to a dispersion relation for the fission process in which the substraction term is uniquely determined. (author)

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.

Comparison of reconstructed radial pin total fission rates with experimental results in full scale BWR fuel elements

Energy Technology Data Exchange (ETDEWEB)

Giust, Flavio [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne, CH-1015 Lausanne (Switzerland); Nordostschweizerische Kraftwerke AG, Parkstrasse 23, CH-5401 Baden (Switzerland); Grimm, Peter; Jatuff, Fabian [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Chawla, Rakesh [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne, CH-1015 Lausanne (Switzerland)

2008-07-01

Total fission rate measurements have been performed on full size BWR fuel assemblies of type SVEA-96+ in the zero power reactor PROTEUS at the Paul Scherrer Institute. This work presents comparisons of reconstructed 2D pin fission rates in two configurations, I-1A and I-2A. Both configurations contain, in the central test zone, an array of 3x3 SVEA-96+ fuel elements moderated with light water at 20 deg. C. In configuration I-2A, an L-shaped hafnium control blade (half of a real cruciform blade) is inserted adjacent to the NW corner of the central fuel element. To minimize the impact of the surroundings, all measurements were done in fuel pins belonging to the central assembly. The 3x3 experimental configuration was modeled using the core monitoring and design tools that are applied at the Leibstadt Nuclear Power Plant (KKL). These are the 2D transport code HELIOS, used for the cross-section generation, and the 3D, 2-group nodal diffusion code PRESTO-2. The exterior is represented, in the axial and radial directions, by 2-group albedos calculated at the test zone boundary using a full-core 3D MCNPX model. The calculated-to-experimental (C/E) ratios of the total fission rates have a standard deviation of 1.3% in configuration I-1A (uncontrolled) and 3.2% in configuration I-2A (controlled). Sensitivity cases are analyzed to show the impact of certain parameters on the calculated fission rate distribution and reactivity. It is shown that the relative pin fission rate is only weakly dependent on these parameters. In cases without a control blade, the pin power reconstruction methodology delivers the same level of accuracy as 2D transport calculations. On the other hand, significant deviations, that are inherent to the use of reflected geometry in the lattice calculations, are observed in cases when the control blade is inserted. (authors)

Chemical behaviour of transuranic elements in the natural environment

International Nuclear Information System (INIS)

Kim, J.I.

1991-01-01

The chemical behaviour of transuranic elements in natural aquifer systems is governed by a variety of geochemical reactions, such as dissolution reaction (solubility), hydrolysis, complexation with inorganics or organics, redox reaction, colloid formation, geochemical interaction with surfaces of various minerals, coprecipitation, mineralisation etc. This paper reviews the present state of knowledge on some of these particular reactions. The emphasis is placed on how the individual reactions can be appraised for the long-term prediction of the geochemical behaviour of transuranic elements in the natural environment. Of the various reactions, the primary thermodynamic processes of dissolution of transuranic compounds in aquatic solution, complexation with important anions present in groundwater and colloid generation are discussed with notable examples. Various laser spectroscopy in use for the chemical speciation are mentioned briefly as for their spectroscopic capability as well as applicability. The present review discussion is primarily directed to a better understanding of the migration behaviour of transuranic elements in natural aquifer systems. (author) 100 refs

Behaviour of Lagrangian triangular mixed fluid finite elements

Indian Academy of Sciences (India)

The behaviour of mixed fluid finite elements, formulated based on the Lagrangian frame of reference, is investigated to understand the effects of locking due to incompressibility and irrotational constraints. For this purpose, both linear and quadratic mixed triangular fluid elements are formulated. It is found that there exists a ...

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.

Separation of short-lived fission products

International Nuclear Information System (INIS)

Tamai, Tadaharu; Ohyoshi, Emiko; Ohyoshi, Akira; Kiso, Yoshiyuki; Shinagawa, Mutsuaki.

1976-01-01

A rbief review is presented on the various methods of separation available for both gaseous and liquid states, for the separation of short-lived fission products formed by binary fission of neutron irradiated uranium. The means available for gaseous state are the hot atom reaction, the hydride method and on-line mass separation. For liquid state, use can be made of precipitation, ionic or atomic exchange, solvent extraction and paper electrophoresis. Particular reference is made to electrophoretic separation of ions produced by fission in aqueous solution of uranium. The principle of electrophoretic separation and the procedures for separating the element of interest from the other fission products are outlined, with reference made to the results obtained with the method by the present authors. The elements in question are alkalines, alkaline earths, rare earths, halogens, selenium and

Compilation of papers presented to the KTG conference on 'Advanced LWR fuel elements: Design, performance and reprocessing', 17-18 November 1988, Karlsruhe Nuclear Research Center

International Nuclear Information System (INIS)

Bahm, W.

1989-05-01

The two expert groups of the Nuclear Society (KTG), 'chemistry and waste disposal' and 'fuel elements' discussed interdisciplinary problems concerning the development and reprocessing of advanced fuel elements. The 10 lectures deal with waste disposal, mechanical layout, operating behaviour, operating experiences and new developments of fuel elements for water moderated reactors as well as operational experiences of the Karlsruhe reprocessing plant (WAK) with reprocessing of high burnup LWR and MOX fuel elements, the distribution of fission products, the condition of the fission products during dissolution and with the effects of the higher burnup of fuel elements on the PUREX process. (DG) [de

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

Fission of nuclei far from stability

International Nuclear Information System (INIS)

Schmidt, K.H.; Benlliure, J.; Junghans, A.R.

2000-11-01

The secondary-beam facility of GSI provided the technical equipment for a new kind of fission experiment. Fission properties of short-lived neutron-deficient nuclei have been investigated in inverse kinematics. The measured element distributions reveal new kinds of systematics on shell structure and even-odd effects and lead to an improved understanding of structure effects in nuclear fission. Prospects for further experimental studies are discussed. (orig.)

Actinide, Elemental, and Fission Product Measurements by ICPMS at the Savannah River Site

International Nuclear Information System (INIS)

Tovo, L.L.; Waller, P.R.; Clymire, J.; Jones, V.D.; Boyce, W.T.

1998-03-01

VG Elemental Inductively coupled plasma-mass spectrometer (ICPMS), PlasmaQuad 1 (PQ1) Model No. 4, installed in a radiohood, is used by the Savannah River Technology Center to provide non-routine mass measurements for environmental monitoring, waste tank characterization studies, isotope ratios for criticality determinations, and the measurement of elemental, fission product, and actinide mass distributions of the glass product from the Defense Waste Processing Facility (DWPF). Modifications to improve instrument reliability, sample preparation, and data handling, as well as modifications to the laboratory that permit measurements in a radioactive environment will be discussed. Based on our operating experience, two laboratory facilities are being prepared for additional instruments to operate in a radioactive environment. A separate instrument is being installed for non-radioactive measurements and method development

Superheavy-element fission tracks in iron meteorites, and reply by Bull, R.K

International Nuclear Information System (INIS)

Runcorn, S.K.

1980-01-01

Comment is made on the lack of superheavy element (SHE) fission tracks observed in silicates from the class IA Odessa iron meteorite by Bull (Nature; 282:393 (1979)). Two explanations are suggested. Firstly, a thermal history for Odessa can be constructed in which the meteorite reaches track retention temperatures only after a time corresponding to many half lives for the SHEs (taken to be approximately 100 Myr) has elapsed and secondly that the IA irons never took up many SHEs. These suggestions are discussed in a reply by Bull. (U.K.)

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

Separation of actinides and long-lived fission products from high-level radioactive wastes (a review)

International Nuclear Information System (INIS)

Kolarik, Z.

1991-11-01

The management of high-level radioactive wastes is facilitated, if long-lived and radiotoxic actinides and fission products are separated before the final disposal. Especially important is the separation of americium, curium, plutonium, neptunium, strontium, cesium and technetium. The separated nuclides can be deposited separately from the bulk of the high-level waste, but their transmutation to short-lived nuclides is a muchmore favourable option. This report reviews the chemistry of the separation of actinides and fission products from radioactive wastes. The composition, nature and conditioning of the wastes are described. The main attention is paid to the solvent extraction chemistry of the elements and to the application of solvent extraction in unit operations of potential partitioning processes. Also reviewed is the behaviour of the elements in the ion exchange chromatography, precipitation, electrolysis from aqueous solutions and melts, and the distribution between molten salts and metals. Flowsheets of selected partitioning processes are shown and general aspects of the waste partitioning are shortly discussed. (orig.) [de

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...

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

The role of the dinuclear system in the processes of nuclear fusion, quasi-fission, fission and cluster formation

International Nuclear Information System (INIS)

Volkov, V.V.

1999-01-01

The nuclear fusion, quasi-fission, fission and cluster formation in an excited nucleus are considered as the processes of the formation and evolution of the dinuclear system. This approach allows one to reveal new aspects of nuclear fusion, to show that quasi-fission plays an important role in nuclear reactions used to synthesise superheavy elements. A qualitative picture is given of the fission process of an excited nucleus and an important role of cluster formation in this process is shown

Fission-gas release in fuel performing to extended burnups in Ontario Hydro nuclear generating stations

International Nuclear Information System (INIS)

Floyd, M.R.; Novak, J.; Truant, P.T.

1992-06-01

The average discharge burnup of CANDU fuel is about 200 MWh/kgU. A significant number of 37-element bundles have achieved burnups in excess of 400 MWh/kgU. Some of these bundles have experienced failures related to their extended operation. To date, hot-cell examinations have been performed on fuel elements from nine 37-element bundles irradiated in Bruce NGS-A that have burnups in the range of 300-800 MWh/kgU. 1 Most of these have declining power histories from peak powers of up to 59 kW/m. Fission-gas releases of up to 26% have been observed and exhibit a strong dependence on fuel power. This obscures any dependence on burnup. The extent of fission-gas release at extended burnups was not predicted by low-burnup code extrapolations. This is attributed primarily to a reduction in fuel thermal conductivity which results in elevated operating temperatures. Reduced conductivity is due, at least in part, to the buildup of fission products in the fuel matrix. Some evidence of hyperstoichiometry exists, although this needs to be further investigated along with any possible relation to CANLUB graphite coating behaviour and sheath oxidation. Residual tensile sheath strains of up to 2% have been observed and can be correlated with fuel power/fission-gas release. SCC 2 -related defects have been observed in the sheath and endcaps of elements from bundles experiencing declining power histories to burnups in excess of 500 MWh/kgU. This indicates that the current recommended burnup limit of 450 MWh/kgU is justified. SCC-related defects have also been observed in ramped bundles having burnups < 450 MWh/kgU. Hence, additional guidelines are in place for power ramping extended-burnup fuel

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)

Characterisation and behaviour under irradiation of rare-earth doped powellite phases - Application to the long term behaviour of nuclear waste matrices

International Nuclear Information System (INIS)

Mendoza, C.

2010-09-01

This work deals with the behaviour under irradiation of a glass-ceramic made after heat treatment of a molybdenum rich R7/T7 type glass. Rare earth elements (Eu 3+ and Nd 3+ ) are used as surrogates of minor actinides and fission products as well as structural luminescent probes. We will focus on the behaviour of the crystalline phase which is a powellite type calcium molybdate that incorporated other elements including rare earth elements. In order to determine the crystalline-chemical properties of the powellite structure, Raman spectroscopy and photoluminescence analyses are led on natural powellite samples and synthetic ceramics with compositions from pure CaMoO 4 to Ca 0.76 Sr 0.1 Na 0.07 Eu 0.01 La 0.02 Nd 0.02 Pr 0.02 MoO 4 , a model composition of the crystalline phase of the glass-ceramic. The analyses of synthetic samples irradiated with He, Ar and Pb ions compared to the behaviour of a natural powellite sample that contains uranium indicate that powellite resist strongly to irradiation and never reach the amorphous state. (author)

Study of a device for the direct measurement of the fission gas pressure inside an in-pile fuel element

International Nuclear Information System (INIS)

Lavaud, B.; Uschanoff, S.

1964-01-01

The fission gas pressure inside a fuel element made of a refractory fuel constitutes an important limiting factor for the burn-up. Although it is possible to calculate approximately the volume of gas produced outside the fuel during its life-time; it is nevertheless very difficult to evaluate the pressure since the volume allowed to the fission gases, as well as their temperature are known only very approximately. This physical value, which is essential for the technologist, can only be known by direct in-pile measurement of the pressure. The report describes the equipment which has been developed for this test. (authors) [fr

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.

Cold valleys in fusion and fission

International Nuclear Information System (INIS)

Misicu, S.

2003-01-01

The cold fission configuration after the preformation of the fragments resembles a short-lived dinuclear or quasi-molecular system. The most conceivable scission configuration is given by two fission fragments in touching with the symmetry axes aligned (pole-pole orientation). This conclusion was based on the simple argument that this configuration offers the optimal tunneling time, i.e. the difference between the Coulomb barrier and the decay energy Q is minimal. Other orientations are apparently precluded in cold spontaneous fission and should be regarded as quasi-fission doorways in the synthesis of superheavy elements by cold fusion. (orig.)

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

CANDU fuel behaviour under LOCA conditions

International Nuclear Information System (INIS)

Kohn, E.

1989-07-01

This report summarizes the current understanding of CANDU fuel-element behaviour under loss-of-coolant (LOCA) accidents. It focuses on a key in-reactor verification experiment conducted at Idaho National Engineering Laboratory (INEL) and on three Canadian in-reactor tests. The in-reactor data, and the considerable body of supporting information developed from out-reactor tests, support the general conclusion that CANDU fuel behaviour during LOCA transients is well understood. Four elements of 37-element CANDU fuel-bundle design were tested under conditions typical of a large-break LOCA blowdown in a CANDU reactor. The purpose of the test was to confirm our current understanding of fuel behaviour under loss-of-coolant accident blowdown conditions. The test also provided data for comparison with predictions made with the steady-state and transient fuel-element performance codes ELESIM and ELOCA. Key components of typical LOCA transients were incorporated in the test: namely, a rapid depressurization rate of the hot coolant, a simultaneous power increase before decreasing to decay values (a power pulse), and prototype fuel element under pre-transient power and burnup conditions. The test was successfully completed in the Power Burst Facility (PBF) reactor at INEL under contract to Ontario Hydro and AECL. The three CANDU Owners Group LOCA tests performed at Chalk River Nuclear Laboratories measured both the thermal-mechanical response and fission-gas release resulting from exposure to a LOCA transient. Results from these three tests provided further confirmation that the behaviour of the fuel under LOCA conditions is understood

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.)

Process for separately recovering uranium, transuranium elements, and fission products of uranium from atomic reactor fuel

International Nuclear Information System (INIS)

Balal, A.L.; Metscher, K.; Muehlig, B.; Reichmuth, C.; Schwarz, B.; Zimen, K.E.

1976-01-01

Spent reactor fuel elements are dissolved in dilute nitric acid. After addition of acetic acid as a complexing agent, the nitric acid is partly decomposed and the mixture subjected to electrolysis while a carrier liquid, which may be dilute acetic acid or a dilute mixture of acetic acid and nitric acid is caused to flow in the electric field between the electrodes either against the direction of ion migration or transversely thereto. The ions of uranium, plutonium, and other transuranium elements, and of fission products accumulate in discrete portions of the electrolyte and are separately withdrawn as at least three fractions after one or more stages of electrolysis

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

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)

50 years with nuclear fission, April 25-28, 1989. A prelude to fission: France

International Nuclear Information System (INIS)

Savic, P.

1989-01-01

A personal account of the events leading to the 1938 discovery, by Irene Joliot-Curie and the author, of the unidentified element R 3.5h will be presented, as well as the experimental methods proving the R 3.5h had chemical properties similar to lanthanum which in turn led Hahn and Strassmann to the discovery of fission in 1939. The author points out that Irene Curie's intuition indicated the path towards the solution of the problem of uranium irradiated by neutrons, which made possible the discovery of fission by Hahn and Strassmann. Further applications of fission became possible by tremendous efforts and excellent contributions of scientists in the USA

Apparatus for measuring the release of fission gases and other fission products by degassing

Energy Technology Data Exchange (ETDEWEB)

Stradal, Karl Alfred

1970-10-15

In gas-cooled high-temperature reactors, the fuel is, in general, inserted in the fuel elements in the form of small particles, which are, for example, coated with pyrolytic carbon. The purpose of this coating is to keep the fission products separate from the coolant gas. The further development of these coated particles makes it necessary to check the retention capacity. One possible method of doing this is the degassing test after irradiation in the reactor. An apparatus is described below, which was developed and installed in order to measure to a higher degree of sensitivity and in serial measurements the release of fission gases and sparingly volatile fission products.

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)

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.

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

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 product elements which have been considered are Cs, I, Te, Sr and Ba but aspects of the behavior 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

Additional foraging elements reduce abnormal behaviour – fur-chewing and stereotypic behaviour – in farmed mink (Neovison vison)

DEFF Research Database (Denmark)

Malmkvist, Jens; Palme, Rupert; Svendsen, Pernille Maj

2013-01-01

We investigated whether provision of additional appetitive and consummatory elements of foraging reduces baseline stress and abnormal behaviour – in terms of fur-chewing and stereotypic behaviour – in farmed mink. We studied 200 juveniles (n = 100 females and 100 males) during the 5-month growth......; (iii) CONS, chunky feed (parts up to 42 mm), replacing conventional feed; (iv) BOTH, access to both biting ropes and chunky feed. In growing mink, biting ropes reduced fur-chewing (P = 0.044) and chunky feed reduced stereotypic behaviour (P = 0.038) and fur-chewing in female mink (P = 0.019). During......, stereotypic behaviour was reduced by provision of chunky feed, increasing the consummatory element in daily foraging. Fur-chewing was reduced upon access to either biting ropes or chunky feed in female mink throughout the study. Our findings support frustrated foraging, mainly consummatory, behind abnormal...

Studies on the chemistry of element 104 with the centrifuge system SISAK-3

Energy Technology Data Exchange (ETDEWEB)

Eberhardt, K.; Kratz, J.V.; Mendel, M.; Naehler, A.; Trautmann, N.; Wiehl, N. [Mainz Univ. (Germany); Alstadt, J.; Omtvedt, J.P. [Oslo Univ. (Norway); Malmbeck, R.; Skarnemark, G.; Wierczinski, B. [Chalmers Univ. of Technology, Goeteborg (Sweden); Eichler, B.; Gaeggeler, H.W.; Jost, D.T.; Tuerler, A. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1997-09-01

The centrifuge system SISAK-3 combined with a detector for the measurement of {alpha}-particles and spontaneous fission (SF) fragments in a flowing organic liquid has been applied to study the chemical behaviour of element 104 produced in the reaction {sup 248}Cm({sup 18}O,5n){sup 261}104. (author) 2 figs., 2 refs.

(d,p)-transfer induced fission of heavy radioactive beams

CERN Document Server

Veselsky, Martin

2012-01-01

(d,p)-transfer induced fission is proposed as a tool to study low energy fission of exotic heavy nuclei. 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 offers a possibility for systematic study the low energy fission of heavy exotic nuclei at the ISOLDE.

Boundary element method for modelling creep behaviour

International Nuclear Information System (INIS)

Zarina Masood; Shah Nor Basri; Abdel Majid Hamouda; Prithvi Raj Arora

2002-01-01

A two dimensional initial strain direct boundary element method is proposed to numerically model the creep behaviour. The boundary of the body is discretized into quadratic element and the domain into quadratic quadrilaterals. The variables are also assumed to have a quadratic variation over the elements. The boundary integral equation is solved for each boundary node and assembled into a matrix. This matrix is solved by Gauss elimination with partial pivoting to obtain the variables on the boundary and in the interior. Due to the time-dependent nature of creep, the solution has to be derived over increments of time. Automatic time incrementation technique and backward Euler method for updating the variables are implemented to assure stability and accuracy of results. A flowchart of the solution strategy is also presented. (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-11-01

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

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)

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)

Characterization of aerosols containing fissionable elements using solid-state track recorders

International Nuclear Information System (INIS)

Roberts, J.H.; Kafalenos, V.P.; Yule, T.J.

1976-01-01

An aerosol of U 3 O 3 highly enriched in 235 U was generated with a nebulizer from a suspension of U 3 O 8 powder in distilled water. The aerosol was collected on a membrane filter. Polycarbonate plastic, placed in good contact with the filter, was used to record fission tracks when the package was exposed to known fluences of slow neutrons. Fission-track stars associated with individual particles of U 3 O 8 were observed in the plastic. The fission-track distributions were converted to a particle size distribution for the aerosol. For a log normal distribution the geometric mean and standard deviation can be determined with better than 5% accuracy. This method can be applied to plutonium and other transuranic aerosols. (orig.) [de

Heavy neutron-deficient radioactive beams: fission studies and fragment distributions

Energy Technology Data Exchange (ETDEWEB)

Schmidt, K.H.; Benlliure, J.; Heinz, A.; Voss, B. [Gesellschaft fuer Schwerionenforschung mbH, Darmstadt (Germany); Boeckstiegel, C.; Grewe, A.; Steinhaeuser, S.; Clerc, H.G.; Jong, M. de; Junghans, A.R.; Mueller, J. [Technische Hochschule Darmstadt (Germany). Inst. fuer Kernphysik; Pfuetzner, M. [Warsaw Univ. (Poland). Inst. of Experimental Physics

1998-02-01

The secondary-beam facility of GSI Darmstadt was used to study the fission process of short-lived radioactive nuclei. Relativistic secondary projectiles were produced by fragmentation of a 1 A GeV {sup 238}U primary beam and identified in nuclear charge and mass number. Their production cross sections were determined, and the fission competition in the statistical deexcitation was deduced for long isotopical chains. New results on the enhancement of the nuclear level density in spherical and deformed nuclei due to collective rotational and vibrational excitations were obtained. Using these reaction products as secondary beams, the dipole giant resonance was excited by electromagnetic interactions in a secondary lead target, and fission from excitation energies around 11 MeV was induced. The fission fragments were identified in nuclear charge, and their velocity vectors were determined. Elemental yields and total kinetic energies have been determined for a number of neutron-deficient actinides and preactinides which were not accessible with conventional techniques. The characteristics of multimodal fission of nuclei around {sup 226}Th were systematically investigated and related to the influence of shell effects on the potential energy and on the level density between fission barrier and scission. A systematic view on the large number of elemental yields measured gave rise to a new interpretation of the enhanced production of even elements in nuclear fission and allowed for a new understanding of pair breaking in large-scale collective motion. (orig.)

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

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.

Measurements of the Fast Fission Factor ({epsilon}) in UO{sub 2}-Elements

Energy Technology Data Exchange (ETDEWEB)

Nylund, O

1961-03-15

These experiments were undertaken in order to check the {epsilon} - calculations for the R3/Adam reactor project and to obtain a more rigorous basis for calculations on fuel elements similar to the R3/Adam clusters. Different number of rods (19, 7, 1) and coolants (air, D{sub 2}O, H{sub 2}O) have been investigated. The measurements were performed in the central channel of the R1 research reactor making it possible also to study the influence due to adjacent R1 fuel rods. The method used (an activation technique) gives directly the relationship R between the fission rates in {sup 238}U and {sup 235}U . Knowing this quantity it is possible to compute {epsilon}. However, since {epsilon} is dependent on how it is defined, it is more convenient to use R in a comparison with calculated values. Calculations for some of the cases studied experimentally have been made according to a new heterogeneous method. The agreement between the calculated and measured R-values is relatively good, but there is a tendency for the latter to be somewhat higher, especially for the 19-rod element.

An Evaluation of a Fission Product Inventory for CANDU Fuels

International Nuclear Information System (INIS)

Jung, Jong Yeob; Park, Joo Hwan

2007-01-01

Fission products are released by two processes when a single channel accident occurs. One is a 'prompt release' and the other is a 'delayed release'. Prompt release assumes that the gap inventory of the fuel elements is released by a fuel element failure at the time of an accident. Delayed release assumes that the inventories within the grain or at the grain boundary are released after a accident due to a diffusion through grains, an oxidation of the fuel and an interaction between the fuel and the Zircaloy sheath. Therefore, the calculation of a fission product inventory and its distribution in a fuel during a normal operating is the starting point for the assessment of a fission product release for single channel accidents. In this report, the fission product inventories and their distributions within s fuel under a normal operating condition are evaluated for three types of CANDU fuels such as the 37 element fuel, CANFLEX-NU and CANFLEX-RU fuel bundles in the 'limiting channel'. To accomplish the above mentioned purposes, the basic power histories for each type of CANDU fuel were produced and the fission product inventories were calculated by using the ELESTRES code

Irradiation tests of THTR fuel elements in the DRAGON reactor (irradiation experiment DR-K3)

International Nuclear Information System (INIS)

Burck, W.; Duwe, R.; Groos, E.; Mueller, H.

1977-03-01

Within the scope of the program 'Development of Spherical Fuel Elements for HTR', similar fuel elements (f.e.) have been irradiated in the DRAGON reactor. The f.e. were fabricated by NUKEM and were to be tested under HTR conditions to scrutinize their employability in the THTR. The fuel was in the form of coated particles moulded into A3 matrix. The kernels of the particles were made of mixed oxide of uranium and thorium with an U 235 enrichment of 90%. One aim of the post irradiation examination was the investigation of irradiation induced changes of mechanical properties (dimensional stability and elastic behaviour) and of the corrosion behaviour which were compared with the properties determined with unirradiated f.e. The measurement of the fission gas release in annealing tests and ceramografic examinations exhibited no damage of the coated particles. The measured concentration distribution of fission metals led to conclusions about their release. All results showed, that neither the coated particles nor the integral fuel spheres experienced any significant changes that could impair their utilization in the THTR. (orig./UA) [de

Two specialized delayed-neutron detector designs for assays of fissionable elements in water and sediment samples

International Nuclear Information System (INIS)

Balestrini, S.J.; Balagna, J.P.; Menlove, H.O.

1976-01-01

Two specialized neutron-sensitive detectors are described which are employed for rapid assays of fissionable elements by sensing for delayed neutrons emitted by samples after they have been irradiated in a nuclear reactor. The more sensitive of the two detectors, designed to assay for uranium in water samples, is 40% efficient; the other, designed for sediment sample assays, is 27% efficient. These detectors are also designed to operate under water as an inexpensive shielding against neutron leakage from the reactor and neutrons from cosmic rays. (Auth.)

Leaching of actinides and fission products from ILW embedded in cement and bitumen, and their mobility in natural salt rock

International Nuclear Information System (INIS)

Flambard, A.R.; Keiling, C.; Fusban, H.U.; Marx, G.

1986-01-01

Real and simulated ILW embedded in cement and bitumen has been subjected to leaching through various binary brines. The resulting solutions containing americium, plutonium, cesium, antimony, ruthenium, cobalt, and strontium have been led through columns packed with the natural Na3γ salt rock from the Gorleben salt dome, in order to determine the mobility characteristics of these elements in the near-field range of a projected waste repository in the Gorleben salt dome, specifically for the case of water intrusion. Leaching data and experimental results are explained and discussed, special attention being given to the impact of the pH-value of the systems studied, and to the formation of carrier (or 'pseudo') colloids during radionuclide release. The paper also gives data obtained on the mobility of transuranium elements and fission products, together with information on differences in behaviour of the actinides and the fission products (ruthenium in particular). (orig.) [de

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

International Nuclear Information System (INIS)

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

1979-09-01

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

PIN99W, Modelling of VVER and PWR Fuel Rod Thermomechanical Behaviour

International Nuclear Information System (INIS)

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

2000-01-01

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

Fission properties of very heavy actinides

International Nuclear Information System (INIS)

Hoffman, D.C.

1979-01-01

The existing data on neutron-emission, kinetic-energy and mass distributions, and half-lives for spontaneous fission of the heavy actinides are reviewed. A comparison of the data for the Fm isotopes with heavier and lighter nuclides suggests that the properties of the heavy Fm isotopes may be unique and can qualitatively be explained on the basis of fragment shell effects, i.e., symmetric fission results in two fragments with configurations close to the doubly magic 132 Sn nucleus. The effect of excitation energy and the use of systematics and theoretical predictions of fission properties and half-lives in the identification of new heavy element isotopes is discussed. 54 references

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.

Fission-product release during accidents

International Nuclear Information System (INIS)

Hunt, C.E.L.; Cox, D.S.

1991-09-01

One of the aims when managing a reactor accident is to minimize the release of radioactive fission products. Release is dependent not only on the temperature, but also on the partial pressure of oxygen. Strongly oxidizing atmospheres, such as those that occurred during the Chernobyl accident, released semi-volatile elements like ruthenium, which has volatile oxides. At low temperatures, UO 2 oxidization to U 3 O 8 can result in extensive breakup of the fuel, resulting in the release of non-volatile fission products as aerosols. Under less oxidizing conditions, when hydrogen accumulates from the zirconium-water reaction, the resulting low oxygen partial pressure can significantly reduce these reactions. At TMI-2, only the noble gases and volatile fission products were released in significant quantities. A knowledge of the effect of atmosphere as well as temperature on the release of fission products from damaged reactor cores is therefore a useful, if not necessary, component of information required for accident management

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.

A warning on fission resonance intergrals: Caveat utor

International Nuclear Information System (INIS)

Holden, N.E.

1988-01-01

A common error is made in defining the resonance integral in most tabulations and handbooks. Although it has a minor effect on the capture resonance integral and on the fission resonance integral for the fissile nuclides, it leads to gross errors in the fission resonance integral for the fertile nuclides. The errors in the fission resonance integral for fertile nuclides of the elements from thorium through curium in the ENDF/B-V library will be presented. Let the user beware

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)

ELOCA: fuel element behaviour during high temperature transients

International Nuclear Information System (INIS)

Sills, H.E.

1979-03-01

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

Measurement of fission yields far from the center of isotopic distributions in the thermal neutron fission of 235U

International Nuclear Information System (INIS)

Shmid, M.

1979-08-01

The main purpose of this work was to measure independent yields, in the thermal neutron fission of 235 U, of fission products which lie far from the centers of the isotopic and isobaric yield distributions. These measurements were used to test the predictions of semi-empirical systematics of fission yields and theoretical fission models. Delay times were measured as a function of temperature in the range 1200-2000degC. The very low delay times achieved in the present work permitted expanding the measurable region to the isotopes 147 , 148 Cs and 99 Rb which are of special interest in the present work. The delay times of Sr and Ba isotopes achieved were more than two orders of magnitude lower than values reported in the literature and thus short-lived isotopes of these elements could be separated for the first time by mass spectrometry. The half-lives of 147 Ba, 148 Ba, 149 La and 149 Ce were measured for the first time. The isotopic distributions of fission yields were measured for the elements Rb, Sr, Cs and Ba in the thermal neutron fission of 235 U, those of 99 Rb, 147 Cs and 148 Cs having been measured for the first time. A comparison of the experimental yields with the predictions of the currently accepted semi-empirical systematics of fission yields, which is the odd-even effect systematics, shows that the systematics succeeds in accounting for the strong odd-even proton effect and the weaker odd-even neutron effect and also in predicting the shape of the distributions in the central region. It is shown that prompt neutron emission broadens the distribution only slightly in the wing of heavy isotopes and more significantly in the wing of light isotopes. But the effect of prompt neutron emission cannot explain the large discrepancies existing between the predictions of fission models and the experimentally measured fission yield in the wings of the isotopic distributions. (B.G.)

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

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

Thermodynamic analysis of volatile organometallic fission products

International Nuclear Information System (INIS)

Auxier II, J.D.; Hall, H.L.; Cressy, Derek

2016-01-01

The ability to perform rapid separations in a post nuclear weapon detonation scenario is an important aspect of national security. In the past, separations of fission products have been performed using solvent extraction, precipitation, etc. The focus of this work is to explore the feasibility of using thermochromatography, a technique largely employed in superheavy element chemistry, to expedite the separation of fission products from fuel components. A series of fission product complexes were synthesized and the thermodynamic parameters were measured using TGA/DSC methods. Once measured, these parameters were used to predict their retention times using thermochromatography. (author)

Final report on fabrication and study of SYNROC containing radioactive waste elements

International Nuclear Information System (INIS)

Reeve, K.D.; Levins, D.M.; Seatonberry, B.W.; Ryan, R.K.; Hart, K.P.; Stevens, G.T.

1987-01-01

Two facilities for the fabrication and testing of Synroc samples containing separate additions of the transuranic actinides americium, plutonium, curium and neptunium, a fission product solution, and two radioisotopes of caesium and strontium were designed, built and operated by the AAEC at the Lucas Heights Research Laboratories. Twenty-one 75 g batches of radioactive Synroc were made and representative samples were characterised by alpha track etching, scanning electron microscopy and aqueous leach testing, mostly at 70 deg C. Where comparisons were possible, radioactive fission products behaved as expected from non-radioactive tests. The leaching behaviour of the actinides was complex but as a group they were the least leachable of all the elements studied

Multi-channel probes to understand fission dynamics

Energy Technology Data Exchange (ETDEWEB)

Mosby, Shea Morgan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-04-15

Explaining the origin of the elements is a major outstanding question in nuclear astrophysics. Observed elemental abundance distribution shows strong nuclear physics effects. In conclusion, neutron-induced reactions are important for nuclear astrophysics and applied fields in nuclear energy and security. LANSCE has a program to address many of these questions directly with neutron beams on (near-)stable nuclei. Increasing demand for correlated data to test details of fission models poses additional challenges. Possibilities exist to extend existing experimental efforts to radioactive beam facilities. Kinematic focusing from using inverse kinematics has potential to circumvent some challenges associated with measuring correlations between fission output channels.

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)

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)

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

Experimental research of fuel element reliability

International Nuclear Information System (INIS)

Cech, B.; Novak, J.; Chamrad, B.

1980-01-01

The rate and extent of the damage of the can integrity for fission products is the basic criterion of reliability. The extent of damage is measurable by the fission product leakage into the reactor coolant circuit. An analysis is made of the causes of the fuel element can damage and a model is proposed for testing fuel element reliability. Special experiments should be carried out to assess partial processes, such as heat transfer and fuel element surface temperature, fission gas liberation and pressure changes inside the element, corrosion weakening of the can wall, can deformation as a result of mechanical interactions. The irradiation probe for reliability testing of fuel elements is described. (M.S.)

Hyperfission - a new mode of nuclear fission

International Nuclear Information System (INIS)

Ion, D.B.; Ivascu, M.; Ion-Mihai, R.

1988-02-01

In this paper the nuclear hyperfission as a new mode of fission, possible for heavy elements with Z > 92, is investigated. The Q-systematics, hyperfissibility parameters, hyperfission barrier as well as the essential hindrance factors are presented. The hyperfission hindrance factor relative to that of fission is found to be in the interval 1.0x10 -17 - 3.4x10 -16 for the parent nuclei with Z = 92-108. (orig.)

Major and trace elements regulation in natural granitic waters: Application to deep radioactive waste disposals

International Nuclear Information System (INIS)

Michard, G.; Negrel, G.; Toulhoat, P.; Beaucaire, C.; Ouzounian, G.

1991-01-01

In order to forecast the evolution of deep groundwaters in the environment of a radioactive waste disposal, one must be able to understand the behaviour of major and trace elements in natural systems. From granitic geothermal and groundwater systems the authors establish that major elements are controlled by mineral precipitation. Regulation levels depend both on equilibration temperature and mobile anion concentration (mainly C1). From empirical laws, the regulation levels with temperature of some trace elements (alkaline and most divalent) elements can be estimated, although a precise explanation for the regulation mechanism is not yet available. They demonstrate that some transition metals are controlled by sulphide precipitation; that uranium is controlled by uraninite solubility; that trivalent and tetravalent metals are present in association with colloidal particles. Maximum regulation levels can be estimated. Such studies can also be useful to forecast the concentration levels of many elements related to nuclear wastes, mainly fission products, uranium, thorium and by analogy artificial actinide elements, as the behaviour of corresponding natural elements can be evaluated

Finite element simulation of the micromagnetic behaviour of nanoelements

International Nuclear Information System (INIS)

Ridley, P.H.W.

2000-06-01

Over recent years the investigation into the magnetic behaviour of nanostructured permalloy has become more advanced due to improvements in numerical micromagnetic methods on the theoretical side and high accuracy electron-beam lithography methods experimentally. The interest in such structures of magnetic material is increasing mainly due to the possible potential use in future high-density magnetic storage media applications. When the material is discretized into a nanoelement structure at the sub micron level theoretical micromagnetic techniques may be employed in order to investigate the magnetization behaviour. This thesis describes a theoretical study of the hysteresis and domain behaviour in thin film permalloy nanoelements. To carry out our investigations we have developed a dynamical micromagnetic model based on the use of the finite element method. The results presented in this thesis begin with a test of the performance of our model. We then proceed with an investigation into the effect of size, elongation and geometry on the transition states for single nanoelements. The investigation is then extended to look at the magnetization behaviour of arrays of interacting nanoelements in relation to their separation and material properties. The reversal mechanism of the arrays is very sensitive to the degree of disorder. In the case of an aligned uniaxial anisotropy a highly symmetric cooperative switching mechanism is observed. A large anisotropy has the effect of stabilizing states during the reversal process leading to distinctive switching. A random anisotropy breaks this high symmetry sufficiently to reduce the cooperative switching leading to a relatively random reversal of individual elements. The theoretical predictions are compared with experimental observations. (author)

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

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

Increasing of MERARG experimental performances: on-line fission gas release measurement by mass spectrometry

Energy Technology Data Exchange (ETDEWEB)

Pontillon, Y.; Capdevila, H.; Clement, S. [CEA, DEN, DEC, SA3C, LAMIR, F-13108 Saint Paul lez Durance, (France); Guigues, E.; Janulyte, A.; Zerega, Y.; Andre, J. [Aix-Marseille Universite, LISA EA 4672, 13397 MARSEILLE cedex 20, (France)

2015-07-01

The MERARG device - implemented at the LECASTAR Hot Laboratory, at the CEA Cadarache - allows characterizing nuclear fuels with respect to the behaviour of fission gases during thermal transients representative of normal or off normal operating nuclear power plant conditions. The fuel is heated in order to extract a part or the total gas inventory it contains. Fission Gas Release (FGR) is actually recorded by mean of both on-line gamma spectrometry station and micro gas chromatography. These two devices monitor the quantity and kinetics of fission gas release rate. They only address {sup 85}Kr radioactive isotope and the elemental quantification of Kr, Xe and He (with a relatively low detection limit in the latter case, typically 5-10 ppm). In order to better estimate the basic mechanisms that promote fission gas release from irradiated nuclear fuels, the CEA fuel study department decided to improve its experimental facility by modifying MERARG to extend the studies of gamma emitter fission gases to all gases (including Helium) with a complete isotopic distribution capability. To match these specifications, a Residual Gas Analyser (RGA) has been chosen as mass spectrometer. This paper presents a review of the main aspects of the qualification/calibration phase of the RGA type analyser. In particular, results recorded over three mass ranges 1-10 u, 80-90 u and 120-140 u in the two classical modes of MERARG, i.e. on-line and off-line measurements are discussed. Results obtained from a standard gas bottle show that the quantitative analysis at a few ppm levels can be achieved for all isotopes of Kr and Xe, as well as masses 2 and 4 u. (authors)

Natural fission reactors - the Oklo phenomenon

International Nuclear Information System (INIS)

Anon.

1989-01-01

Overview describes the discovery of the site, location of the reactors and site geology and discusses the permanence of fission products, nuclear reaction control mechanisms and trace concentrations of elements that act as poisons. (Author)

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

Immobilization of fission products arising from pyrometallurgical reprocessing in chloride media

Science.gov (United States)

Leturcq, G.; Grandjean, A.; Rigaud, D.; Perouty, P.; Charlot, M.

2005-12-01

Spent nuclear fuel reprocessing to recover energy-producing elements such as uranium or plutonium can be performed by a pyrochemical process. In such method, the actinides and fission products are extracted by electrodeposition in a molten chloride medium. These processes generate chlorinated alkali salt flows contaminated by fission products, mainly Cs, Ba, Sr and rare earth elements constituting high-level waste. Two possible alternatives are investigated for managing this wasteform; a protocol is described for dechlorinating the fission products to allow vitrification, and mineral phases capable of immobilizing chlorides are listed to allow specification of a dedicated ceramic matrix suitable for containment of these chlorinated waste streams. The results of tests to synthesize chlorosilicate phases are also discussed.

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

Systematics of neutron-induced fission yields

International Nuclear Information System (INIS)

Blachot, J.; Brissot, R.

1983-10-01

The main characteristics of the mass and charge distributions for thermal neutron induced fission of actinides are reviewed. We show that these distributions can be reasonably reproduced with only 24 data as input. We use a representation where the element yields together with the most probable mass Ap(Z) play the dominant role. The ability of this model to calculate mass yields for the fission of not yet measured actinides is also shown. The influence of the excitation energy of the fissile system on charge and mass distribution is also discussed

New fission valley for 258Fm and nuclei beyond

International Nuclear Information System (INIS)

Moeller, P.; Nix, J.R.; Swiatecki, W.J.

1986-01-01

Experimental results on the fission properties of nuclei close to 264 Fm show sudden and large changes with a change of only one or two neutrons or protons. The nucleus 258 Fm, for instance, undergoes symmetric fission with a half-life of about 0.4 ms and a kinetic energy peaked at about 235 MeV whereas 256 Fm undergoes asymmetric fission with a half-life of about 3 h and a kinetic energy peaked at about 200 MeV. Qualitatively, these sudden changes hve been postulated to be due to the emergence of fragment shells in symmetric fission products close to 132 Sn. A quantitative calculation that shows where high-kinetic-energy symmetric fission occurs and why it is associated with a sudden and large decrease in fission half-lives. The study is based on calculations of potential-energy surfaces in the macroscopic-microscopic model and a semi-empirical model for the nuclear inertia. The implications of the new fission valley on the stability of the heaviest elements is discussed. 33 refs., 12 figs

A discrete element model for the investigation of the geometrically nonlinear behaviour of solids

Science.gov (United States)

Ockelmann, Felix; Dinkler, Dieter

2018-07-01

A three-dimensional discrete element model for elastic solids with large deformations is presented. Therefore, an discontinuum approach is made for solids. The properties of elastic material are transferred analytically into the parameters of a discrete element model. A new and improved octahedron gap-filled face-centred cubic close packing of spheres is split into unit cells, to determine the parameters of the discrete element model. The symmetrical unit cells allow a model with equal shear components in each contact plane and fully isotropic behaviour for Poisson's ratio above 0. To validate and show the broad field of applications of the new model, the pin-pin Euler elastica is presented and investigated. The thin and sensitive structure tends to undergo large deformations and rotations with a highly geometrically nonlinear behaviour. This behaviour of the elastica can be modelled and is compared to reference solutions. Afterwards, an improved more realistic simulation of the elastica is presented which softens secondary buckling phenomena. The model is capable of simulating solids with small strains but large deformations and a strongly geometrically nonlinear behaviour, taking the shear stiffness of the material into account correctly.

The fission yeast CENP-B protein Abp1 prevents pervasive transcription of repetitive DNA elements.

Science.gov (United States)

Daulny, Anne; Mejía-Ramírez, Eva; Reina, Oscar; Rosado-Lugo, Jesus; Aguilar-Arnal, Lorena; Auer, Herbert; Zaratiegui, Mikel; Azorin, Fernando

2016-10-01

It is well established that eukaryotic genomes are pervasively transcribed producing cryptic unstable transcripts (CUTs). However, the mechanisms regulating pervasive transcription are not well understood. Here, we report that the fission yeast CENP-B homolog Abp1 plays an important role in preventing pervasive transcription. We show that loss of abp1 results in the accumulation of CUTs, which are targeted for degradation by the exosome pathway. These CUTs originate from different types of genomic features, but the highest increase corresponds to Tf2 retrotransposons and rDNA repeats, where they map along the entire elements. In the absence of abp1, increased RNAPII-Ser5P occupancy is observed throughout the Tf2 coding region and, unexpectedly, RNAPII-Ser5P is enriched at rDNA repeats. Loss of abp1 also results in Tf2 derepression and increased nucleolus size. Altogether these results suggest that Abp1 prevents pervasive RNAPII transcription of repetitive DNA elements (i.e., Tf2 and rDNA repeats) from internal cryptic sites. Copyright © 2016 Elsevier B.V. All rights reserved.

A contribution to the search for the element 114 in nature

International Nuclear Information System (INIS)

McMinn, J.

1975-09-01

Starting from a lead ore having an age of 1.3 x 10 9 years, an already chemically pre-enriched material up to a factor of about 2 to 3 x 10 8 was available for the search of the element 114 based on the properties predicted for eka lead. A further physical enrichment was undertaken using the Juelich high performance electromagnetic isotope separator. The exposed target material (synthetic quartz glass plates)in the mass range 278 to 290 in the mass separator was irradiated with thermal neutrons and the fission tracks were counted under an optical microscope. The background due to fissionable impurities in the quartz ( -13 atoms per SiO 2 molecule) and that due to the use of the mass separator were studied in preliminary experiments. The possible escape of the presumably volatile element 114 from the fission track detectors during irradiation as a result of nuclear γ-heating was also investigated in simulation experiments. The diffusion behaviour of carrier-free 203 Hg implanted into quartz plates in the mass separator, as determined by a radiometric analysis, showed that the expected losses are negligible below 200 0 C. The counting results from the fission track detectors gave a maximum value of 5.1 x 10 -22 barns for the product of the fission cross section sigmasub(f) and the number of eka-lead atoms per lead carrier atom. In order to be able to make an estimate of the absolute concentration limits, the fission cross sections of eka-lead for thermal neutrons were approximated using an extrapolation method. Assuming these approximate fission cross sections, the absolute concentration limits of 8 x 10sup(-24), 3 x 10 -22 and 1 x 10 -19 share of eka-lead atoms per carrier atom were obtained for the isotopes 295, 297 and 298 (sigma = 70 barns, 2 barns and 4 gbarns). (orig./LH) [de

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

Flibe blanket concept for transmuting transuranic elements and long lived fission products

International Nuclear Information System (INIS)

Gohar, Y.

2000-01-01

A Molten salt (Flibe) fusion blanket concept has been developed to solve the disposition problems of the spent nuclear fuel and the transuranic elements. This blanket concept can achieve the top rated solution, the complete elimination of the transuranic elements and the long-lived fission products. Small driven fusion devices with low neutron wall loading and low neutron fluence can perform this function. A 344-MW integrated fusion power from D-T plasmas for thirty years with an availability factor of 0.75 can dispose of 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. In addition, the utilization of this blanket concept eliminates the need for a geological repository site, which is a major advantage. This application provides an excellent opportunity to develop and to enhance the public acceptance of the fusion energy for the future. The energy from the transmutation process is utilized to produce revenue. Flibe, lithium-lead eutectic, and liquid lead are possible candidates. The liquid blankets have several features, which are suited for W application. It can operate at constant thermal power without interruption for refueling by adjusting the concentration of the transuranic elements and lithium-6. These liquids operate at low-pressure, which reduces the primary stresses in the structure material. Development and fabrication costs of solid transuranic materials are eliminated. Burnup limit of the transuranic elements due to radiation effects is eliminated. Heat is generated within the liquid, which simplifies the heat removal process without producing thermal stresses. These blanket concepts have large negative temperature coefficient with respect to the blanket reactivity, which enhances the safety performance. These liquids are chemically and thermally stable under irradiation conditions, which minimize the radioactive waste volume. The operational record of the Molten Salt Breeder Reactor with Flibe was very successful

Isotopic composition of primary xenon and the fission of Pu-244

Energy Technology Data Exchange (ETDEWEB)

Levskii, L K

1983-05-01

The hypothesis that the origin of xenon on earth is due to the fission of uranium and/or transuranium elements is examined. The isotopic composition of primary xenon on earth is calculated using a model (Levskii, 1980) of the isotopic composition of rare gases which is based on the hypothesis of the heterogeneity of the isotopic composition of the elements of the solar system. The isotopic composition of fission-produced xenon in the atmosphere and solid earth is determined to correspond to the abundance of xenon isotopes as a result of the spontaneous fission of Pu-244 (half-life of 8.2 x 10 to the 7th years). The amount of fission-produced xenon in the atmosphere is shown to amount to about 30 percent (Xe-136). Under certain conditions, the degree of the degassing of the solid earth for xenon is 25 percent, which corresponds to a ratio of Kr-84/Xe-130 45 for the earth as a whole.

Behaviour of radioactive elements in the vegetation of uranium mine

International Nuclear Information System (INIS)

Shitaka, Yukinori

2001-03-01

To improve the predictability of radium transfer from soil to plant, the behaviour was compared with other alkaline earth elements. I analyzed the radium transfer ratio as the factor of solubility from soil to water, absorbability by plant from soil, and the allocation to each organ for three grass fields in Ningyo-Toge environmental engineering center. The results indicated that the behaviour of radium is correlated with that of other alkaline earth elements. Moreover negative correlation was found between the absorbability of each element and the concentration of Mg and/or Ca in soluble fraction. From the results, it was suggested that the absorption of radium should be influenced by the regulation of the absorption of these nutrients. Secondly, to clarify the soil- and plant-specific feature of the radium transfer, the radium transfer was investigated for twelve plant species from four vegetation. There was 60 times of difference in radium concentration among plant species. It was observed that there is strong correlation between the transfer factor of radium and that of calcium. The differences in radium concentration of plant were explained by the soil-specific difference in the concentration ratio of radium and the plant-specific difference in calcium concentration. (author)

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.

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

International Nuclear Information System (INIS)

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

1975-01-01

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

A Model to Reproduce the Response of the Gaseous Fission Product Monitor (GFPM) in a CANDU{sup R} 6 Reactor (An Estimate of Tramp Uranium Mass in a Candu Core)

Energy Technology Data Exchange (ETDEWEB)

Mostofian, Sara; Boss, Charles [AECL Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga Ontario L5K 1B2 (Canada)

2008-07-01

In a Canada Deuterium Uranium (Candu) reactor, the fuel bundles produce gaseous and volatile fission products that are contained within the fuel matrix and the welded zircaloy sheath. Sometimes a fuel sheath can develop a defect and release the fission products into the circulating coolant. To detect fuel defects, a Gaseous Fission Product Monitoring (GFPM) system is provided in Candu reactors. The (GFPM) is a gamma ray spectrometer that measures fission products in the coolant and alerts the operator to the presence of defected fuel through an increase in measured fission product concentration. A background fission product concentration in the coolant also arises from tramp uranium. The sources of the tramp uranium are small quantities of uranium contamination on the surfaces of fuel bundles and traces of uranium on the pressure tubes, arising from the rare defected fuel element that released uranium into the core. This paper presents a dynamic model that reproduces the behaviour of a GFPM in a Candu 6 plant. The model predicts the fission product concentrations in the coolant from the chronic concentration of tramp uranium on the inner surface of the pressure tubes (PT) and the surface of the fuel bundles (FB) taking into account the on-power refuelling system. (authors)

Fission product release profiles from spherical HTR fuel elements at accident temperatures

International Nuclear Information System (INIS)

Schenk, W.; Pitzer, D.; Nabielek, H.

1986-10-01

A total of 22 fuel elements with modern TRISO particles has been tested in the temperature range 1500-2500 0 C. Additionally, release profiles of iodine and other isotopes have been obtained with seven UO 2 samples at 1400-1800 0 C. For heating times up to 100 hours at the maximum temperature, the following results are pertinent to HTR accident conditions: Ag 110 m is the only fission products to be released at 1200-1600 0 C by diffusion through intact SiC, but it is of low significance in accident assessments; cesium, iodine, strontium, and noble gas releases up to 1600 0 C are solely due to various forms of contamination; at 1700-1800 0 C, corrosion induced SiC defects cause the release of Cs, Sr, I/Xe/Kr; above 2000 0 C, thermal decomposition of the silicon carbide layer sets in while pyrocarbons still remain intact. Around 1600 0 C, the accident specific contribution of cesium, strontium, iodine, and noble gases is negligible. (orig./HP) [de

Fission in intermediate energy heavy ion reactions

International Nuclear Information System (INIS)

Wilhelmy, J.B.; Begemann-Blaich, M.; Blaich, T.; Boissevain, J.; Fowler, M.M.; Gavron, A.; Jacak, B.V.; Lysaght, P.S.; Britt, H.C.; Fields, D.J.; Hansen, L.F.; Lanier, R.G.; Massoletti, D.J.; Namboodiri, M.M.; Remington, B.A.; Sangster, T.C.; Struble, G.L.; Webb, M.L.; Chan, Y.D.; Dacai, A.; Harmon, A.; Leyba, J.; Pouliot, J.; Stokstad, R.G.; Hansen, O.; Levine, M.J.; Thorn, C.E.; Trautmann, W.; Dichter, B.; Kaufman, S.; Videbaek, F.; Fraenkel, Z.; Mamane, G.; Cebra, D.; Westfall, G.D.

1989-01-01

A systematic study of reaction mechanisms at intermediate energies (50-100 MeV/A) has been performed at the Lawrence Berkeley Laboratory's BeValac using medium weight projectiles on medium and heavy element targets. A gas and plastic phoswich detector system was employed which gave large geometric coverage and a wide dynamic response. The particles identified with the gas detectors could be characterized into three components - intermediate mass fragments (IMF), fission fragments (FF) and heavy residues (HR). Major observed features are: The reaction yields are similar in the 50 to 100 MeV/A range, central collisions have high multiplicty of IMF's with broad angular correlations consistent with a large participant region, effects of final state Coulomb interactions are observed and give information on the size and temporal behavior of the source, true fission yields are dependent on target fissility and correlated with relatively peripheral collisions. Analysis of fission and evaporation yields implies limiting conditions for which fission decay remains a viable deexcitation channel. (orig.)

Fission in intermediate energy heavy ion reactions

International Nuclear Information System (INIS)

Wilhelmy, J.B.; Begemann-Blaich, M.; Blaich, T.

1989-01-01

A systematic study of reaction mechanisms at intermediate energies (50--100 MeV/A) has been performed at the Lawrence Berkeley Laboratory's BeValac using medium weight projectiles on medium and heavy element targets. A gas and plastic phoswich detector system was employed which gave large geometric coverage and a wide dynamic response. The particles identified with the gas detectors could be characterized into three components - intermediate mass fragments (IMF), fission fragments (FF) and heavy residues (HR). Major observed features are: the reaction yields are similar in the 50 to 100 MeV/A range, central collisions have high multiplicity of IMF's with broad angular correlations consistent with a large participant region, effects of final state Coulomb interactions are observed and give information on the size and temporal behavior of the source, true fission yields are dependent on target fissility and correlated with relatively peripheral collisions. Analysis of fission and evaporation yields implies limiting conditions for which fission decay remains a viable deexcitation channel. 7 figs

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

Valence electron structure of cast iron and graphltization behaviour criterion of elements

Institute of Scientific and Technical Information of China (English)

刘志林; 李志林; 孙振国; 杨晓平; 陈敏

1995-01-01

The valence electron structure of common alloy elements in phases of cast iron is calculated- The relationship between the electron structure of alloy elements and equilibrium, non-equilibrium solidification and graphitization is revealed by defining the bond energy of the strongest bond in a phase as structure formation factor S. A criterion of graphitization behaviour of elements is advanced with the critical value of the structure formation factor of graphite and the n of the strongest covalent bond in cementite. It is found that this theory conforms to practice very well when the criterion is applied to the common alloy elements.

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)

Compound Nucleus Reactions in LENR, Analogy to Uranium Fission

Science.gov (United States)

Hora, Heinrich; Miley, George; Philberth, Karl

2008-03-01

The discovery of nuclear fission by Hahn and Strassmann was based on a very rare microanalytical result that could not initially indicate the very complicated details of this most important process. A similarity is discussed for the low energy nuclear reactions (LENRs) with analogies to the yield structure found in measurements of uranium fission. The LENR product distribution measured earlier in a reproducible way in experiments with thin film electrodes and a high density deuteron concentration in palladium has several striking similarities with the uranium fission fragment yield curve.ootnotetextG.H. Miley and J.A. Patterson, J. New Energy 1, 11 (1996); G.H. Miley et al, Proc ICCF6, p. 629 (1997).This comparison is specifically focussed to the Maruhn-Greiner local maximum of the distribution within the large-scale minimum when the fission nuclei are excited. Implications for uranium fission are discussed in comparison with LENR relative to the identification of fission a hypothetical compound nuclear reaction via a element ^306X126 with double magic numbers.

Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

International Nuclear Information System (INIS)

Weissfloch, R.

The fuel elements of High-Temperature Reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are present. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can serve as an interim state on the way to a complete theory. (U.S.)

Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Weissfloch, R

1973-07-15

The fuel elements of high-temperature reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons, the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are presented. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons, a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can be held as an interim state on the way to a complete theory.

Plutonium and surrogate fission products in a composite ceramic waste form

International Nuclear Information System (INIS)

Esh, D. W.; Frank, S. M.; Goff, K. M.; Johnson, S. G.; Moschetti, T. L.; O'Holleran, T.

1999-01-01

Argonne National Laboratory is developing a ceramic waste form to immobilize salt containing fission products and transuranic elements. Preliminary results have been presented for ceramic waste forms containing surrogate fission products such as cesium and the lanthanides. In this work results from scanning electron microscopy/energy dispersive spectroscopy and x-ray diffraction are presented in greater detail for ceramic waste forms containing surrogate fission products. Additionally, results for waste forms containing plutonium and surrogate fission products are presented. Most of the surrogate fission products appear to be silicates or aluminosilicates whereas the plutonium is usually found in an oxide form. There is also evidence for the presence of plutonium within the sodalite phase although the chemical speciation of the plutonium is not known

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

A generic biokinetic model for predicting the behaviour of the lanthanide elements in the human body

International Nuclear Information System (INIS)

Taylor, D.M.; Leggett, R.W.

2003-01-01

Information on the biokinetics of the 15 elements of the lanthanide series, 57 La to 71 Lu, is too sparse to permit individual development of meaningful biokinetic models to describe the behaviour of each of the elements in humans. The lanthanides show a regular gradation in chemical properties across the series, and animal studies indicate that this is reflected in regular differences in their deposition in tissues such as the liver and skeleton. These regular differences in chemical and biological behaviour have been utilised to construct a generic lanthanide biokinetic model and to define element-specific parameters for each element in the series. This report describes the use of the available biokinetic data for humans and animals to derive the parameters for each of the elements. (author)

On the safety of conceptual fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Okrent, D.; Badham, V.; Caspi, S.; Chan, C.K.; Ferrell, W.J.; Frederking, T.H.K.; Grzesik, J.; Lee, J.Y.; McKone, T.E.; Pomraning, G.C.; Ullman, A.Z.; Ting, T.D.; Kim, Y.I.

1979-01-01

A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (c) induced radioactivity, and (d) tritium. As a result of these studies, it appears that highly reliable and even redundent decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered. (Auth.)

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)

Mass distributions in monoenergetic-neutron-induced fission of 232Th

International Nuclear Information System (INIS)

Glendenin, L.E.; Gindler, J.E.; Ahmad, I.; Henderson, D.J.; Meadows, J.W.

1980-01-01

Fission product yields for 38 masses were determined for the fission of 232 Th with essentially monoenergetic neutrons of 2.0, 3.0, 4.0, 5.9, 6.4, 6.9, 7.6, and 8.0 MeV. Fission product activities were measured by Ge(Li) γ-ray spectrometry of irradiated 232 Th foils and by chemical separation of the fission product elements followed by β counting. The mass yield data for 232 Th(n,f ) show a sensitive increase of fission yields in the near-symmetric mass region (valley) with increasing incident neutron energy E/sub n/ and a pronounced dip in yield at the onset of second-chance fission just above the neutron binding energy (at approx. 6 MeV) where the excitation energy is lowered by competition with neutron evaporation prior to fission. The effect of second-chance fission is also seen in the yields of asymmetric peak products. A distinct third peak is observed at symmetry in the valley of the mass distribution, and enhanced yields are observed in the asymmetric peaks at masses associated with even Z (proton pairing effect). The fission yeilds of 232 Th(n,f ) are compared with those of 238 U(n,f ) and 232 Th

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Fission distribution measurements of Atucha's fuel pellets with solid state track detectors

International Nuclear Information System (INIS)

Ricabarra, M.D. Bovisio de; Waisman, Dina.

1979-08-01

Distribution of fissions in a UO 2 rod has been measured by means of solid state detectors. Mica muscovite and Makrofol-N detectors were used in the experiment. The merits of mica muscovite relative to the Makrofol-N for the detection of fission fragments have been verified. However both fission track detectors closely agree (0,5%) in the final fission distribution of the UO 2 rod. Sensitivity of the detectors shows to be linear in the range between 50.000and 360.000 fission tracks per square centimeter. Due to the high spatial resolution this method is better than any other technique. Determination were made in UO 2 pellets similar to the fuel element of the Atucha reactor. The average fission rate in the rod has been measured within 0,8% error, and provides an accurate determination for the distribution of fissions in the rod wich is needed for the determination of energy liberated per fission in the natural uranium rod.(author) [es

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

The solubility of solid fission products in carbides and nitrides of uranium and plutonium. Part I: literature review on experimental results

International Nuclear Information System (INIS)

Benedict, U.

1977-01-01

This review compiles the available data on the solubility of the most important non-volatile fission products in the carbides, nitrides, and carbonitrides of uranium and plutonium. It includes some elements which are not fission products, but belong to a group of the Periodic Table which contains one or more fission products elements

Source term and behavioural parameters for a postulated HIFAR loss-of-coolant accident

International Nuclear Information System (INIS)

May, F.G.

1987-01-01

The fraction of the fission product inventory which might be released into the atmosphere of the HIFAR reactor containment building (RCB) during a postulated loss-of-coolant accident (LOCA) has been evaluated as a function of time, for each classification of airborne radioactivity. This appraisal will be used as the source term for a computer program, which uses realistic attenuation of the fission product aerosol in a single compartment model with a defined leakrate to predict possible radioactive releases into the environment in a hypothetical bounding case reactor accident which is rather more severe in all major aspects than any single LOCA. Also given are the parameters governing the attenuation of the aerosol and vapours in the atmosphere of the RCB so that their behaviour may be accurately modelled. The source terms for several other types of accident involving the meltdown of fuel elements have also been considered but in less detail than the LOCA case. In some of the cases, the fission products are released directly to atmosphere, so there is no attenuation of the release by deposition within the RCB

Infiltration behaviour of elemental mercury DNAPL in fully and partially water saturated porous media

Science.gov (United States)

D'Aniello, Andrea; Hartog, Niels; Sweijen, Thomas; Pianese, Domenico

2018-02-01

Mercury is a contaminant of global concern due to its harmful effects on human health and for the detrimental consequences of its release in the environment. Sources of liquid elemental mercury are usually anthropogenic, such as chlor-alkali plants. To date insight into the infiltration behaviour of liquid elemental mercury in the subsurface is lacking, although this is critical for assessing both characterization and remediation approaches for mercury DNAPL contaminated sites. Therefore, in this study the infiltration behaviour of elemental mercury in fully and partially water saturated systems was investigated using column experiments. The properties affecting the constitutive relations governing the infiltration behaviour of liquid Hg0, and PCE for comparison, were determined using Pc(S) experiments with different granular porous media (glass beads and sands) for different two- and three-phase configurations. Results showed that, in water saturated porous media, elemental mercury, as PCE, acted as a non-wetting fluid. The required entry head for elemental mercury was higher (from about 5 to 7 times). However, due to the almost tenfold higher density of mercury, the required NAPL entry heads of 6.19 cm and 12.51 cm for mercury to infiltrate were 37.5% to 20.7% lower than for PCE for the same porous media. Although Leverett scaling was able to reproduce the natural tendency of Hg0 to be more prone than PCE to infiltrate in water saturated porous media, it considerably underestimated Hg0 infiltration capacity in comparison with the experimental results. In the partially water saturated system, in contrast with PCE, elemental mercury also acted as a nonwetting fluid, therefore having to overcome an entry head to infiltrate. The required Hg0 entry heads (10.45 and 15.74 cm) were considerably higher (68.9% and 25.8%) than for the water saturated porous systems. Furthermore, in the partially water saturated systems, experiments showed that elemental mercury displaced

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

Microprobe study of fission product behavior in high-burnup HTR fuels

International Nuclear Information System (INIS)

Kleykamp, H.

Electron microprobe analysis of irradiated coated particles with high burnup (greater than 50 percent fima) gives detailed information on the chemical state and the transport behavior of the fission products in UO 2 and UC 2 kernels and in the coatings. In oxide fuel kernels, metallic inclusions and ceramic precipitations are observed. The solubility behavior of the fission products in the fuel matrix has been investigated. Fission product inclusions could not be detected in carbide fuel kernels; post irradiation annealed UC 2 kernels, however, give information on the element combinations of some fission product phases. Corresponding to the chemical state in the kernel, Cs, Sr, Ba, Pd, Te and the rare earths are released easily and diffuse through the entire pyrocarbon coating. These fission products can be retained by a silicon carbide layer. The initial stage of a corrosive attack of the SiC coating by the fission products is evidenced

Interaction of noble-metal fission products with pyrolytic silicon carbide

International Nuclear Information System (INIS)

Lauf, R.J.; Braski, D.N.

1982-01-01

Fuel particles for the High-Temperature Gas-Cooled Reactor (HTGR) contain layers of pyrolytic carbon and silicon carbide, which act as a miniature pressure vessel and form the primary fission product barrier. Of the many fission products formed during irradiation, the noble metals are of particular interest because they interact significantly with the SiC layer and their concentrations are somewhat higher in the low-enriched uranium fuels currently under consideration. To study fission product-SiC interactions, particles of UO 2 or UC 2 are doped with fission product elements before coating and are then held in a thermal gradient up to several thousand hours. Examination of the SiC coatings by TEM-AEM after annealing shows that silver behaves differently from the palladium group

Fission barriers of superheavy nuclei

International Nuclear Information System (INIS)

Burvenich, T.

2001-01-01

Full text: Self consistent microscopic mean-field models are powerful tools for the description of nuclear structure phenomena in the region of known elements, where they have reached a good quality. Especially the Skyrme-Hartree-Fock (SHF) method and the Relativistic Mean-Field (RMF) model will be considered in the discussion of the properties of these models. When it comes to extrapolation to the region of superheavy elements, though there is agreement concerning the global trends, these model exhibit significant differences in their predictions concerning fission barrier heights and structures. (Author)

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

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)

Study of 235U very asymmetric thermal fission

International Nuclear Information System (INIS)

Sida, J.L.

1989-12-01

The fission fragment separator Lohengrin of the Institut Laue-Langevin in Grenoble was used to determine the yields of the very asymmetric light fission products (A=84-69) as a function of A, Z, and the kinetic energy E. The proton pairing effect causes fine structures in the mass distribution, in the mean nuclear charge anti Z and its variance σ z , and in the mean kinetic energies of the elements. The neutron pairing effect in the production yields is found for the first time of the same order of magnitude than the proton pairing effect. In the mass region investigated both are the largest observed in fission of 235 U. A decrease in the mean kinetic energy for the isotopes of Ni and Cu was observed. It points to a large deformation at scission. Our results support the view that very asymmetric low-energy fission is a weakly dissipative process. The highly deformed transient system breaks by a slow necking-in process [fr

Investigation of the fission yields of the fast neutron-induced fission of {sup 233}U; Mesure de la distribution en masse et en charge des produits de la fission rapide de l'{sup 233}U

Energy Technology Data Exchange (ETDEWEB)

Galy, J

1999-09-01

As a stars, a survey of the different methods of investigations of the fission product yields and the experimental data status have been studied, showing advantages and shortcomings for the different approaches. An overview of the existing models for the fission product distributions has been as well intended. The main part of this thesis was the measurement of the independent yields of the fast neutron-induced fission of{sup 233}U, never investigated before this work. The experiment has been carried out using the mass separator OSIRIS (Isotope Separator On-Line). Its integrated ion-source and its specific properties required an analysis of the delay-parameter and ionisation efficiency for each chemical species. On the other hand, this technique allows measurement of independent yields and cumulative yields for elements from Cu to Ba, covering most of the fission yield distribution. Thus, we measured about 180 independent yields from Zn (Z=30) to Sr (Z=38) in the mass range A=74-99 and from Pd (Z=46) to Ba (Z=56) in the mass range A=113-147, including many isomeric states. An additional experiment using direct {gamma}-spectroscopy of aggregates of fission products was used to determine more than 50 cumulative yields of element with half-life from 15 min to a several days. All experimental data have been compared to estimates from a semi-empirical model, to calculated values and to evaluated values from the European library JEF 2.2. Furthermore, a study of both thermal and fast neutron-induced fission of {sup 233}U measured at Studsvik, the comparison of the OSIRIS and LOHENGRIN facilities and the trends in new data for the Reactors Physics have been discussed. (author)

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)

Nuclear fission: a review of experimental advances and phenomenology

Science.gov (United States)

Andreyev, A. N.; Nishio, K.; Schmidt, K.-H.

2018-01-01

In the last two decades, through technological, experimental and theoretical advances, the situation in experimental fission studies has changed dramatically. With the use of advanced production and detection techniques both much more detailed and precise information can now be obtained for the traditional regions of fission research and, crucially, new regions of nuclei have become routinely accessible for fission studies. This work first of all reviews the recent developments in experimental fission techniques, in particular the resurgence of transfer-induced fission reactions with light and heavy ions, the emerging use of inverse-kinematic approaches, both at Coulomb and relativistic energies, and of fission studies with radioactive beams. The emphasis on the fission-fragment mass and charge distributions will be made in this work, though some of the other fission observables, such as prompt neutron and γ-ray emission will also be reviewed. A particular attention will be given to the low-energy fission in the so far scarcely explored nuclei in the very neutron-deficient lead region. They recently became the focus for several complementary experimental studies, such as β-delayed fission with radioactive beams at ISOLDE(CERN), Coulex-induced fission of relativistic secondary beams at FRS(GSI), and several prompt fusion–fission studies. The synergy of these approaches allows a unique insight in the new region of asymmetric fission around {\\hspace{0pt}}180 Hg, recently discovered at ISOLDE. Recent extensive theoretical efforts in this region will also be outlined. The unprecedented high-quality data for fission fragments, completely identified in Z and A, by means of reactions in inverse kinematics at FRS(GSI) and VAMOS(GANIL) will be also reviewed. These experiments explored an extended range of mercury-to-californium elements, spanning from the neutron-deficient to neutron-rich nuclides, and covering both asymmetric, symmetric and transitional fission regions

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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.

Remarks on the fission barriers of super-heavy nuclei

Energy Technology Data Exchange (ETDEWEB)

Hofmann, S. [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Goethe-Universitaet Frankfurt, Institut fuer Physik, Frankfurt (Germany); Heinz, S.; Mann, R.; Maurer, J.; Muenzenberg, G.; Barth, W.; Dahl, L.; Kindler, B.; Kojouharov, I.; Lang, R.; Lommel, B.; Runke, J.; Scheidenberger, C.; Tinschert, K. [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Antalic, S. [Comenius University, Department of Nuclear Physics and Biophysics, Bratislava (Slovakia); Eberhardt, K.; Thoerle-Pospiech, P.; Trautmann, N. [Johannes Gutenberg-Universitaet Mainz, Mainz (Germany); Grzywacz, R. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); University of Tennessee, Knoxville, TN (United States); Hamilton, J.H. [Vanderbilt University, Department of Physics and Astronomy, Nashville, TN (United States); Henderson, R.A.; Kenneally, J.M.; Moody, K.J.; Shaughnessy, D.A.; Stoyer, M.A. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Miernik, K. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); University of Warsaw, Warsaw (Poland); Miller, D. [University of Tennessee, Knoxville, TN (United States); Morita, K. [RIKEN Nishina Center for Accelerator-Based Science, Wako, Saitama (Japan); Nishio, K. [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Popeko, A.G.; Yeremin, A.V. [Joint Institute for Nuclear Research, Dubna (Russian Federation); Roberto, J.B.; Rykaczewski, K.P. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Uusitalo, J. [University of Jyvaeskylae, Department of Physics, Jyvaeskylae (Finland)

2016-04-15

Shell-correction energies of super-heavy nuclei are approximated by using Q{sub α} values of measured decay chains. Five decay chains were analyzed, which start at the isotopes {sup 285}Fl, {sup 294}118, {sup 291}Lv, {sup 292}Lv and {sup 293}Lv. The data are compared with predictions of macroscopic-microscopic models. Fission barriers are estimated that can be used to eliminate uncertainties in partial fission half-lives and in calculations of evaporation-residue cross-sections. In that calculations, fission probability of the compound nucleus is a major factor contributing to the total cross-section. The data also provide constraints on the cross-sections of capture and quasi-fission in the entrance channel of the fusion reaction. Arguments are presented that fusion reactions for synthesis of isotopes of elements 118 and 120 may have higher cross-sections than assumed so far. (orig.)

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

Mechanical behaviour of synthetic surgical meshes: finite element simulation of the herniated abdominal wall.

Science.gov (United States)

Hernández-Gascón, B; Peña, E; Melero, H; Pascual, G; Doblaré, M; Ginebra, M P; Bellón, J M; Calvo, B

2011-11-01

The material properties of meshes used in hernia surgery contribute to the overall mechanical behaviour of the repaired abdominal wall. The mechanical response of a surgical mesh has to be defined since the haphazard orientation of an anisotropic mesh can lead to inconsistent surgical outcomes. This study was designed to characterize the mechanical behaviour of three surgical meshes (Surgipro®, Optilene® and Infinit®) and to describe a mechanical constitutive law that accurately reproduces the experimental results. Finally, through finite element simulation, the behaviour of the abdominal wall was modelled before and after surgical mesh implant. Uniaxial loading of mesh samples in two perpendicular directions revealed the isotropic response of Surgipro® and the anisotropic behaviour of Optilene® and Infinit®. A phenomenological constitutive law was used to reproduce the measured experimental curves. To analyze the mechanical effect of the meshes once implanted in the abdomen, finite element simulation of the healthy and partially herniated repaired rabbit abdominal wall served to reproduce wall behaviour before and after mesh implant. In all cases, maximal displacements were lower and maximal principal stresses higher in the implanted abdomen than the intact wall model. Despite the fact that no mesh showed a behaviour that perfectly matched that of abdominal muscle, the Infinit® mesh was able to best comply with the biomechanics of the abdominal wall. Copyright © 2011 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

Baby fission chambers; Etude de chambres a fission miniatures

Energy Technology Data Exchange (ETDEWEB)

Guery, U; Tachon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

The present report is intended, on the one band, as a study of the main types of fission chambers produced to date, and on the other, to deal more generally with this type of detector. Originally, it was with a view to the charting of neutron scatter in 'Proserpine' that the authors undertook the study of these chambers. During the course of the task, it was considered worth tbe trouble of developing its scope to include a more general application: neutron scatter measurement of various energy neutrons within a reduced volume with slight local disturbance. (author) [French] Le present rapport se propose, d'une part, d'exposer les principales realisations de chambres a fission, d'autre part de faire une mise au point a caractere plus general sur ces detecteurs. Au depart, c'est surtout en vue des mesures de densite neutronique dans 'Proserpine' que les auteurs ont etudie ces chambres; au cours de la mise au point, il a paru interessant de developper leur etude pour des applications plus generales: mesures de densites de neutrons de differentes energies dans un element de volume tres reduit et avec faible perturbation locale. (auteur)

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)

Application of mercury cathode electrolysis to fission-product separation

International Nuclear Information System (INIS)

Besson, A.; Prigent, Y.; Van-Kote, F.

1969-01-01

A method involving controlled potential mercury cathode electrolysis has been developed to separate fission products. It allows the radiochemical determination of Ag, Cd, Pd, Rh, Ru, Sn, Te, Sb and Mo from solutions of fission products highly concentrated in mineral salts. The general procedure consists in three main steps: electrolytic amalgam generation, destruction of amalgams and ultimate purification of elements by other means. Electrolytic operations last about five hours. Chemical yields lie between 10 per cent and 70 per cent. (authors) [fr

Investigation of fission products release and structural changes of WWER spent fuel in inert and oxidizing environment

International Nuclear Information System (INIS)

Kungurtsev, I.A.; Smirnov, V.P.; Kuzmin, I.V.; Lebeduk, I.V.; Pimonov, Y.I.; Sohcilin, G.I.; Stupina, L.N.; Chesanov, V.V.; Shtuckert, Y.A.; Zvir, E.A.

1996-01-01

At the Research Institute of Atomic Reactors in-cell experiments were carried out which were aimed at investigation of WWER spent fuel behaviour under accident conditions. Gaseous and volatile fission products release and the influence of gaseous swelling, fuel interaction with the cladding and oxidation on it have been investigated. At the present time, series of experiments in inert and air environments have been finished and the tests in steam environment have been carrying out. In all series the samples in the form of fuel pellets fragments and fuel elements pieces were used. This report presents some results of annealing tests and investigations of the sample microstructure after annealing. (author). 4 refs, 15 figs, 2 tabs

[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

Independent yields of Rb and Cs isotopes from thermal-neutron induced fission of 235U

International Nuclear Information System (INIS)

Balestrini, S.J.; Decker, R.; Wollnik, H.; Wuensch, K.D.; Jung, G.; Koglin, E.; Siegert, G.

1979-01-01

The relative yields of Rb and Cs isotopes from thermal-neutron fission of 235 U have been redetermined using the mass separator OSTIS, on-line at a neutron guide of the High-Flux Beam Reactor at the Institut Laue-Langevin, Grenoble, France. The separator ion source was a hot oven containing 235 U in a graphite matrix. The neutron beam was pulsed. Alkali fission products diffused out of the graphite and were ionized, thus producing a stepwise increase in the analyzed ion beam proportional to the independent fission yield. The ion beam and the fissions in the source were monitored simultaneously. The diffusion of Rb and Cs from the source was exponential in time with half-lives ranging from 2.8 to 18 sec, depending upon the element and source temperature. The independent fission yields of Rb and Cs are normalized by equating their element yields to each other and to a value computed from the charge distributions observed with the recoil separator LOHENGRIN and well established mass yields. Fractional independent yields are deduced from the independent fission yields, and these compare very well with the EOZ model described by Wahl

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.

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

Packing Guidelines for Optimizing Singlet Fission Matrix Elements in Noncovalent Dimers

Czech Academy of Sciences Publication Activity Database

Buchanan, E. A.; Michl, Josef

2017-01-01

Roč. 139, č. 44 (2017), s. 15572-15575 ISSN 0002-7863 R&D Projects: GA ČR GA15-19143S Institutional support: RVO:61388963 Keywords : exciton fission * solar cells * thin film Subject RIV: CF - Physical ; Theoretical Chemistry OBOR OECD: Physical chemistry Impact factor: 13.858, year: 2016

Examination of irradiated fuel elements using gamma scanning technique

International Nuclear Information System (INIS)

Ichim, O.; Mincu, M.; Man, I.; Stanica, M.

2016-01-01

The purpose of this paper is to validate the gamma scanning technique used to calculate the activity of gamma fission products from CANDU/TRIGA irradiated fuel elements. After a short presentation of the equipments used and their characteristics, the paper describes the calibration technique for the devices and how computed tomography reconstruction is done. Following the previously mentioned steps is possible to obtain the axial and radial profiles and the computed tomography reconstruction for calibration sources and for the irradiated fuel elements. The results are used to validate the gamma scanning techniques as a non-destructive examination method. The gamma scanning techniques will be used to: identify the fission products in the irradiated CANDU/TRIGA fuel elements, construct the axial and radial distributions of fission products, get the distribution in cross section through computed tomography reconstruction, and determine the nuclei number and the fission products activity of the irradiated CANDU/TRIGA fuel elements. (authors)

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

Physics and chemistry of fission

International Nuclear Information System (INIS)

1979-01-01

models, the nucleus is imagined to be a very small drop of liquid; the theorist then devises many schemes that lead to its splitting into two smaller drops. It is surprising how many detailed features theorists can predict, based on such simple models, and how well these predictions have been confirmed by experiments. The symposium summarized the progress in this field, and indicated how many intricate details can be introduced into a simple liquid drop model to give better agreement with experimental results. Step by step, a picture is emerging and being continuously improved, coming closer and closer to the truth. However, the liquid drop model has several competitors of which the most prominent is the shell model. This model assumes that the constituents of the nucleus are sorted into well defined energy levels, or shells; the distribution of protons and neutrons in these shells, and their movements from one shell to another, can provide an explanation for many experimentally observed facts New theoretical results drawn from these models were reported at the symposium, together with some efforts to combine different theoretical concepts, thereby trying to create a unified picture of nuclear fission. The nucleus is too small for the scientist to be able to observe directly what is happening inside it. There is really only one way for an experimenter to study a process in a nucleus, he must measure the energies of the particles coming out of it. In the case of fission, these 'messengers' can be the fission fragments, i.e. nuclei of elements that are being created from the splitting atom, but they can also be the neutrons or gamma rays which emerge during and after the fission process. Several sessions in the symposium were devoted to reports of such studies. With the help of sophisticated electronics systems, or complicated radiochemical methods, experimenters have measured the energies of fission fragments, neutrons, gamma and X-rays, and other particles emerging from

Directions for nuclear research in the transplutonium elements

International Nuclear Information System (INIS)

Wilhelmy, J.B.; Chasman, R.R.; Friedman, A.M.; Ahmad, I.

1983-01-01

The study of the heavy nuclides has played a vital role in our understanding of the alpha decay process, nuclear fission, nuclear binding energies and the limits of nuclear stability. This study has led to the understanding of novel shape degrees of freedom, such as the very large quadrupole deformations associated with the fission isomer process, and the very recently discovered octupole deformation. The existence of these unique phenomena in the heavy element region is not accidental. Fission isomerism is due to the delicate balance between nuclear forces holding the nucleus together and Coulomb forces causing nuclear fission. Octupole deformation arises from the increasing strength of matrix elements with increasing oscillator shell. Both illustrate the unique features of the heavy element region. Fission studies have given us information about large collective aspects in nuclei and the importance that nuclear structural effects can play in altering these macro properties. A new class of atomic studies has become possible with the availability of heavy elements. With these isotopes, we are now able to produce electric fields of such magnitude that it becomes possible to spontaneously create positron-electron pairs in the vacuum. We have organized this presentation into three major sections: nuclear structure, fission studies and atomic studies of supercritical systems. In each we will try to emphasize the new directions which can benefit from the continued availability of isotopes supplied by the Trans-plutonium Production Program. 117 references

Nuclear structure in cold rearrangement processes in fission and fusion

Energy Technology Data Exchange (ETDEWEB)

Armbruster, P.

1998-11-01

In fission and fusion of heavy nuclei large numbers of nucleons are rearranged at a scale of excitation energy very small compared to the binding energy of the nuclei. The energies involved are less than 40 MeV at nuclear temperatures below 1.5 MeV. The shapes of the configurations in the rearrangement of a binary system into a monosystem in fusion, or vice versa in fission, change their elongations by as much as 8 fm, the radius of the monosystem. The dynamics of the reactions macroscopically described by a potential energy surface, inertia parameters, dissipation, and a collision energy is strongly modified by the nuclear structure of the participating nuclei. Experiments showing nuclear structure effects in fusion and fission of the heaviest nuclei are reviewed. The reaction kinematics and the multitude of isotopes involved are investigated by detector techniques and by recoil spectrometers. The advancement of the latter allows to find very small reaction branches in the range of 10{sup -5} to 10{sup -10}. The experiments reveal nuclear structure effects in all stages of the rearrangement processes. These are discussed pointing to analogies in fusion and fission on the microscopic scale, notwithstanding that both processes macroscopically are irreversible. Heavy clusters, as 132Sn, 208Pb, nuclei with closed shell configurations N=82,126, Z=50,82 survive in large parts of the nuclear rearrangement. They determine the asymmetry in the mass distribution of low energy fission, and they allow to synthesise superheavy elements, until now up to element 112. Experiments on the cold rearrangement in fission and fusion are presented. Here, in the range of excitation energies below 12 MeV the phenomena are observed most convincingly. (orig.)

Recovery of trans-plutonium elements; Recuperation des elements transplutoniens

Energy Technology Data Exchange (ETDEWEB)

Espie, J Y; Poncet, B; Simon, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1970-07-01

The object of this work is to study the recovery of americium and curium from the fission-product solution obtained from the processing of irradiated fuel elements made of natural metallic uranium alloyed with aluminium, iron and silicon; these elements have been subjected to an average irradiation of 4000 MW days/ton in a gas-graphite type reactor having a thermal power of 3.7 MW/ton of uranium. The process used consists of 3 extraction cycles and one americium-curium separation: - 1) extraction cycle in 40 per cent TBP: extraction of actinides and lanthanides; elimination of fission products; - 2) extraction cycle in 8 per cent D2EHPA: decontamination from the fission products, decontamination of actinides from lanthanides; - 3) extraction cycle in 40 per cent TBP: separation of the complexing agent and concentration of the actinides; - 4) americium-curium separation by precipitation. (authors) [French] Cette etude a pour objet, la recuperation de l'americium et du curium de la solution de produits de fission provenant du traitement de combustibles irradies a base d'uranium naturel metallique allie a l'aluminium, le fer, et le silicium, et ayant subi une irradiation moyenne de 4000 MWj/t dans une pile du type graphite-gaz, dont la puissance thermique est de 3.7 MW/t d'uranium. Le procede utilise comprend 3 cycles d'extraction et une separation americium-curium: - 1. cycle d'extraction dans le TBP a 40 pour cent: extraction des actinides et des lanthanides, elimination des produits de fission; - 2. cycle d'extraction dans le D2EHPA a 8 pour cent: decontamination en produits de fission, decontamination des actinides en lanthanides; - 3. cycle d'extraction dans le TBP a 40 pour cent: separation du complexant et concentration des actinides; - 4. separation americium-curium par precipitation. (auteurs)

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)

Fission times studies of the Z=124 superheavy nucleus by X-ray fluorescence

International Nuclear Information System (INIS)

Airiau, Maud

2016-01-01

Since the 1960's nuclear structure model have predicted the existence of an island of stability of superheavy elements. It should be located around the next magic numbers expected at N=172 or 184 and between Z=114 and 126 depending on the model. Very high fission barrier of a few MeV are predicted to be generated by microscopic effects for those nuclei for which large fission times distributions extended to very high fission times are induced. Fission time measurements of the superheavy element Z=124 have been made by us using the X-ray fluorescence technique, a method based on the filling of inner-shell electronic vacancies created during the collision leading to the formation of the compound nucleus. The aim of this experiment was to detect in coincidence both fission fragments and characteristic X-rays from the Z=124, created by the reaction 238 U+ 70,76 Ge. The main difficulty was to identify those X-rays due to the fact that gamma-rays from fission fragments were emitted in the same energy range, which affected our photon multiplicities for any fragment selection. This new difficulty brings an important limitation to the study of some particular superheavy elements by the X-ray fluorescence method. K X-rays spectra have been simulated using MCDF (Multi-Configuration-Dirac-Fock) and then compared to the experimental ones in order to get a maximal K X-ray multiplicity compatible with our data. The extracted results were about 6-7% for 76 Ge and from 12 to 14% for 70 Ge. Those values remain compatible with the experimental signature of long lifetime component observed for the same system but using a blocking technique in single crystals. (author) [fr

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

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)

The analysis of effect of gaseous fission products on service-ability of various types of fuel elements used in research reactors

International Nuclear Information System (INIS)

Vatulin, A.A.; Kulakov, G.V.; Kosaurov, A.A.; Dobrikova, I.V.; Morozov, A.A.

2005-01-01

The analysis of the work to license U-Mo dispersion fuel that has been carried out by program RERTR shows that at high burnups the U-Mo particles interacts with an Al matrix to form porosity in the central areas of the fuel meat. This results leads to change the loading of cladding due to the formation of long-length defects in the fuel meat. As a result, the pressure of gases leads to a larger shape change (pillowing) of fuels and in some instances to their failure. The paper presents the results of modeling pillowing process proceeding in Russia's research reactor fuels of different types (tubular and rod fuels) that is effected by the accumulation and pressure of fission gas in meat porosity. The problem has been resolved in the viscous-plasticity statement using the program MARC for finite element analysis. The behavior of fission gas (FG) in open porosity has been analyzed to reveal the influence of the fuel design on their resistance to the pressure of FG. (author)

Fundamental aspects of nuclear reactor fuel elements

International Nuclear Information System (INIS)

Olander, D.R.

1976-01-01

The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO 2 , fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO 2 , radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies

FRESCO-II: A computer program for analysis of fission product release from spherical HTGR-fuel elements in irradiation and annealing experiments

International Nuclear Information System (INIS)

Krohn, H.; Finken, R.

1983-06-01

The modular computer code FRESCO has been developed to describe the mechanism of fission product release from a HTGR-Core under accident conditions. By changing some program modules it has been extended to take into account the transport phenomena (i.e. recoil) too, which only occur under reactor operating conditions and during the irradiation experiments. For this report, the release of cesium and strontium from three HTGR-fuel elements has been evaluated and compared with the experimental data. The results show that the measured release can be described by the considered models. (orig.) [de

Microscopic predictions of fission yields based on the time dependent GCM formalism

International Nuclear Information System (INIS)

Regnier, D.; Dubray, N.; Verriere, M.; Schunck, N.

2016-01-01

Accurate knowledge of fission fragment yields is an essential ingredient of numerous applications ranging from the formation of elements in the r-process to fuel cycle optimization in nuclear energy. The need for a predictive theory applicable where no data is available, together with the variety of potential applications, is an incentive to develop a fully microscopic approach to fission dynamics. One of the most promising theoretical frameworks is the time-dependent generator coordinate method (TDGCM) applied under the Gaussian overlap approximation (GOA). Previous studies reported promising results by numerically solving the TDGCM+GOA equation with a finite difference technique. However, the computational cost of this method makes it difficult to properly control numerical errors. In addition, it prevents one from performing calculations with more than two collective variables. To overcome these limitations, we have developed the new code FELIX-1.0 that solves the TDGCM+GOA equation based on the Galerkin finite element method. In this article, we briefly illustrate the capabilities of the solver FELIX-1.0, in particular its validation for n+ 239 Pu low energy induced fission. FELIX-1.0 gives full control on the numerical precision of fission product yields in neutron-induced fission, and its scalability also enables series of dynamical calculations on several potential energy surfaces. Preliminary results suggest an important sensitivity of our two-dimensional approach to the input potential energy surface

ESOL facility for the generation and radiochemical separation of short half-life fission products

International Nuclear Information System (INIS)

Gehrke, R.J.; Meikrantz, D.H.; Baker, J.D.; Anderl, R.A.; Novick, V.J.; Greenwood, R.C.

1988-01-01

A facility has been developed at the Idaho National Engineering Laboratory (INEL) for the generation and rapid radiochemical separation of short half-life mixed fission products. This facility, referred to as the Idaho Elemental Separation On Line (ESOL), consists of electro-plated sources of spontaneously fissioning 252 Cf with a helium jet transport arrangement to continuously deliver short half-life, mixed fission products to the radiochemistry laboratory for rapid, computer controlled, radiochemical separations. 18 refs., 13 figs

Microscopic predictions of fission yields based on the time dependent GCM formalism

Science.gov (United States)

Regnier, D.; Dubray, N.; Schunck, N.; Verrière, M.

2016-03-01

Accurate knowledge of fission fragment yields is an essential ingredient of numerous applications ranging from the formation of elements in the r-process to fuel cycle optimization in nuclear energy. The need for a predictive theory applicable where no data is available, together with the variety of potential applications, is an incentive to develop a fully microscopic approach to fission dynamics. One of the most promising theoretical frameworks is the time-dependent generator coordinate method (TDGCM) applied under the Gaussian overlap approximation (GOA). Previous studies reported promising results by numerically solving the TDGCM+GOA equation with a finite difference technique. However, the computational cost of this method makes it difficult to properly control numerical errors. In addition, it prevents one from performing calculations with more than two collective variables. To overcome these limitations, we developed the new code FELIX-1.0 that solves the TDGCM+GOA equation based on the Galerkin finite element method. In this article, we briefly illustrate the capabilities of the solver FELIX-1.0, in particular its validation for n+239Pu low energy induced fission. This work is the result of a collaboration between CEA,DAM,DIF and LLNL on nuclear fission theory.

A model of fission gas behavior during steady-state operation

International Nuclear Information System (INIS)

Villalobos, A.

1981-01-01

A model of fission gas behavior during the steady-state operation of a nuclear reactor that uses uranium dioxide as fuel is developed. The basic physical phenomena encountered in analyzing the disposition of fission gas have been retained, but in a simplified form for ease of calculation. The analysis code, includes treatment of intragranular, grain face, and grain edge gas and release to the open spaces. The code is utilized to obtain comparisons with experimental data and to perform fuel behavior sensitivity studies. The results obtained in the sensitivity studies indicate the importance of including grain face and grain edge bubbles treatments in modeling fission gas. It is found that representation of release in different sections of the fuel pin is possible in a simple way by assuming evenly spaced bubbles on the edge, and that grain edge bubble interlinkage is a necessary condition for release to the open spaces. It is also indicated by the sensitivity studies that fission gas swelling is mainly due to grain edge bubbles. Grain face bubbles, although large in size, are few in number and contribute little to swelling. Intragranular swelling is intermediate between these two values. The resulting code can be used in predicting fuel element performance, that is necessary in nuclear fuel design, safety analysis, and interpretation of experimental data on fuel element behavior

Independent yields of Rb and Cs isotopes from thermal-neutron induced fission of /sup 235/U

Energy Technology Data Exchange (ETDEWEB)

Balestrini, S.J.; Decker, R.; Wollnik, H.; Wuensch, K.D.; Jung, G.; Koglin, E.; Siegert, G.

1979-12-01

The relative yields of Rb and Cs isotopes from thermal-neutron fission of /sup 235/U have been redetermined using the mass separator OSTIS, on-line at a neutron guide of the High-Flux Beam Reactor at the Institut Laue-Langevin, Grenoble, France. The separator ion source was a hot oven containing /sup 235/U in a graphite matrix. The neutron beam was pulsed. Alkali fission products diffused out of the graphite and were ionized, thus producing a stepwise increase in the analyzed ion beam proportional to the independent fission yield. The ion beam and the fissions in the source were monitored simultaneously. The diffusion of Rb and Cs from the source was exponential in time with half-lives ranging from 2.8 to 18 sec, depending upon the element and source temperature. The independent fission yields of Rb and Cs are normalized by equating their element yields to each other and to a value computed from the charge distributions observed with the recoil separator LOHENGRIN and well established mass yields. Fractional independent yields are deduced from the independent fission yields, and these compare very well with the EOZ model described by Wahl.

Assessments of sheath strain and fission gas release data from 20 years of power reactor fuel irradiations

International Nuclear Information System (INIS)

Purdy, P.L.; Manzer, A.M.; Hu, R.H.; Gibb, R.A.; Kohn, E.

1997-01-01

Over the past 20 years, many fuel elements or bundles discharged from Canadian CANDU power reactors have been examined in the AECL hot cells. The post-irradiation examination (PIE) database covers a wide range of operating conditions, from which fuel performance characteristics can be assessed. In the present analysis, a PIE database was compiled representing elements from a total of 129 fuel bundles, of which 26% (34 bundles) were confirmed to have one or more defective elements. This comprehensive database was assessed in terms of measured sheath strain and fission gas release (FGR) for intact elements, in an attempt to identify any changes in these parameters over the history of CANDU reactor operation. Results from this assessment indicate that, for the data that are typical of normal CANDU operating conditions, tensile sheath strain and FGR have remained within 0.5% and 8%, respectively. Those data beyond these ranges are from fuel operated under abnormal conditions, not representative of normal operation, and thus do not indicate a trend toward unexpected fuel behaviour. The distributions of the PIE measurements indicate that maximum expected sheath strains and FGR for normally operated fuel are 0.7% and 13%, respectively. (author)

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.

Baby fission chambers; Etude de chambres a fission miniatures

Energy Technology Data Exchange (ETDEWEB)

Guery, U.; Tachon, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

The present report is intended, on the one band, as a study of the main types of fission chambers produced to date, and on the other, to deal more generally with this type of detector. Originally, it was with a view to the charting of neutron scatter in 'Proserpine' that the authors undertook the study of these chambers. During the course of the task, it was considered worth tbe trouble of developing its scope to include a more general application: neutron scatter measurement of various energy neutrons within a reduced volume with slight local disturbance. (author) [French] Le present rapport se propose, d'une part, d'exposer les principales realisations de chambres a fission, d'autre part de faire une mise au point a caractere plus general sur ces detecteurs. Au depart, c'est surtout en vue des mesures de densite neutronique dans 'Proserpine' que les auteurs ont etudie ces chambres; au cours de la mise au point, il a paru interessant de developper leur etude pour des applications plus generales: mesures de densites de neutrons de differentes energies dans un element de volume tres reduit et avec faible perturbation locale. (auteur)

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.

Simulation of pellet-cladding thermomechanical interaction and fission gas release

International Nuclear Information System (INIS)

Denis, A.; Soba, A.

2001-01-01

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

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)

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

Irradiation of mixed UO2-PuO2 oxide samples for fast neutron reactor fuel elements

International Nuclear Information System (INIS)

Mikailoff, H.; Mustelier, J.; Bloch, J.; Conte, M.; Hayet, L.; Lauthier, J.C.; Leclere, J.

1968-01-01

Thermal flux irradiation testings of small mixed oxide pellets UPuO 2 fuel elements were performed in support of the fuel reference design for the Phenix fast reactor. The effects of different parameters (stoichiometry, pellet density, pellet clad gap). on the behaviour of the oxide (temperature distribution, microstructural changes, fission gas release) were investigated in various irradiation conditions. In particular, the effect of fuel density decrease and power rate increase on thermal performances were determined on short term irradiations of porous fuels. (authors) [fr

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

International Nuclear Information System (INIS)

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

1985-04-01

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

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.

Effect of microbial processes on transuranium elements behaviour in soil, plants and animal organism

International Nuclear Information System (INIS)

Uajldung, R.Eh.; Garlend, T.P.

1985-01-01

Results of preliminary studies discussed in the present paper bring about the supposition that concentration and chemical from of an element in a plant play an essential role in variation of its availability for animals consuming plants. That is why any assessment of long-term behaviour of transuranium elements in terrestrialenvironment should be based on determination of factors affecting solubility and forms of soluble compounds in soil. These factors include concentration and chemical form of the element migrating to soil; effect of the properties of soil on element distribution between solid and liquid phases; effect soil processes on kinetics of sorption reactions, concentration of transuranium elements, forms of soluble and non-soluble chemical compounds

The discovery of uranium fission

International Nuclear Information System (INIS)

Brix, P.

1990-01-01

Uranium was discovered 200 years ago. Its radioactive character was first demonstrated in 1896 and two years later radium was extracted from uranium minerals. In 1911 studies with alpha rays from radioactive decay led to the unexpected discovery of the atomic nucleus. Exposure of beryllium to alpha rays yielded neutrons, first detected in 1932. Starting in 1934, neutron irradiation of uranium produced radioactive substances erroneously attributed to transuranium elements but with confusing properties. Painstaking experiments by chemists left no doubt on 17 December 1938 that barium was produced by these irradiations: the neutrons had split some uranium nuclei. The physics of the fission process was understood two weeks later; after a few months, neutron multiplication was found to be probable. This review deals with the eminent scientists involved, their successes, errors and disappointments, and the unexpected insights which occurred on the paths and detours of scientific research. It is, therefore, instructive also to discuss how fission was not discovered. The momentous discovery must be considered inevitable; the great tragedy was that Germany started World War II just at the time when the possibility of nuclear chain reactions and bombs became known. The consequences and anxieties that remain after 50 years of nuclear fission demand that mankind act with reason and conscience to maintain peace. (author)

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

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.)

Fundamental aspects of nuclear reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Olander, D.R.

1976-01-01

The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO/sub 2/, fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO/sub 2/, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies. (DG)

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)

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

The calculation - experimental investigations of the HTGR fuel element construction

International Nuclear Information System (INIS)

Eremeev, V.S.; Kolesov, V.S.; Chernikov, A.S.

1985-01-01

One of the most important problems in the HTGR development is the creation of the fuel element gas-tight for the fission products. This problem is being solved by using fuel elements of dispersion type representing an ensemble of coated fuel particles dispersed in the graphite matrix. Gas-tightness of such fuel elements is reached at the expense of deposing a protective coating on the fuel particles. It is composed of some layers serving as diffusion barriers for fission products. It is apparent that the rate of fission products diffusion from coated fuel particles is determined by the strength and temperature of the protective coating

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)

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

International Nuclear Information System (INIS)

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

1985-09-01

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

Process for ultimate storage of radioactive fission products

International Nuclear Information System (INIS)

Baukal, W.; Gruenthaler, K.H.; Neumann, K.

1980-01-01

In order to exclude cracking in the cooling phase during sealing of radioactive oxidic fission products in glass melts, metallic filling elements - e.g. wires, tissues - are proposed to be incorporated in the mould before the glass melt is poured in. Especially nickel alloys with corrosion proof surface layers, e.g. titanium nitride, silicon carbide, silicon nitride, aluminium oxide, suit best. These elements reduce thermal stresses and effect high thermal conductance towards the mould wall. (UWI) [de

RIBD-IRT, Isotope Buildup and Isotope Decay from Fission Source

International Nuclear Information System (INIS)

1990-01-01

1 - Description of problem or function: RIBD-IRT calculates isotopic concentrations resulting from two fission sources with normal down- chain decay by beta emission and isomeric transfers and inter-chain coupling resulting from (n,gamma) reactions. Calculations can be made to follow an irradiation history through an unlimited number of step changes of unrestricted duration and variability including shutdown periods, restarts at different power levels and/or any other level changes. In addition, the program permits to track and modify the concentration of individual elements as they decay with time following reactor shutdown. Tracking individual elements enables one to estimate time-dependent source terms for a hypothetical LOCA based on known or postulated fission product release mechanisms. 2 - Method of solution: RIBD-IRT is a grid processor. It organizes the various members described by the fission product library data into a grid with the various linkages established from chain branching data, yield data, and neutron capture cross sections with their branching ratios. Radioactive decay includes not only the simple member-to-member cascade but also the more complex forms where branching may be partially or completely skip one or two intervening members

Systematics of criticality properties of actinide nuclides and its bearing on the long lived fission waste problem

International Nuclear Information System (INIS)

Srinivasan, M.; Rao, K.S.; Garg, S.B.; Iyengar, P.K.

1989-01-01

This paper reports on a systematic analysis of the criticality parameters of over twenty fissile and fertile isotopes of eight transthorium actinide elements that has been carried out by us. It is observed that K ∞ increases and critical mass decreases monotonically with the fissility parameter (Z 2 /A) of the nuclides. This implies that each and every isotope of transuranic elements such as Np, Am, Cm etc. which are produced as by-products during reactor operation is a more valuable nuclear fuel than the corresponding fissile/fissible isotopes of plutonium. This finding has a profound bearing on the long lived fission waste problem and supports the view that the byproduct actinide elements should be separated from the high level waste stream and recycled back into fission reactors, thereby eliminating one of the commonly voiced concerns regarding the acceptability of nuclear fission power

Transactions of 2. international seminars on the mathematical/mechanical modelling of reactor fuel elements

International Nuclear Information System (INIS)

Lassmann, K.

1991-01-01

Fuel element modelling is a wide field of activity that spans decades of research and code development for different reactor systems and very different situations such as normal operation, off-normal situations and severe accidents. Modern computer technology helps to take the full advantage of detailed model development performed over the past for daily design analyses, safety analyses, conception of new experiments and investigation of an improved nuclear fuel utilization and fuel element performance. The basic development of the concepts of fuel element modelling can be considered as finished. The future trends are the development of refined models based on a deeper understanding of the physical and mechanical basis. Areas of interest are transient phenomena especially the fission product behaviour, burnup-enhanced phenomena, PCI and fuel reliability, severe core damage and chemical aspects. The seminar presentations reflect this variety

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)

Fission-product burnup chain model for research reactor application

Energy Technology Data Exchange (ETDEWEB)

Kim, Jung Do; Gil, Choong Sup; Lee, Jong Tai [Korea Atomic Energy Research Inst., Daeduk (Republic of Korea)

1990-12-01

A new fission-product burnup chain model was developed for use in research reactor analysis capable of predicting the burnup-dependent reactivity with high precision over a wide range of burnup. The new model consists of 63 nuclides treated explicitly and one fissile-independent pseudo-element. The effective absorption cross sections for the preudo-element and the preudo-element yield of actinide nuclides were evaluated in the this report. The model is capable of predicting the high burnup behavior of low-enriched uranium-fueled research reactors.(Author).

Thorium content of a mineral ore from Morro do Ferro by fission track technique

International Nuclear Information System (INIS)

Oliveira, C.A.N. de.

1980-10-01

The feasibility to determine thorium concentrations by fission track technique in samples of mineral ore has been demonstrated. The literature registers only the application of the fission track technique to mineral ore in the case where the fissionable element is uranium. The technique was applied to determine the thorium concentration of an ore sample from Morro do Ferro, taking advantage of the high thorium to uranium ratio in that mineral. The sample analysed presented a thorium concentration of 2467 +- 400 mg Th/Kg ore. The so called wet method was adopted by using the Bayer made Makrofol KG 10μm thick, as the detector foil, immersed in the thorium solution. The technique is also useful to determine thorium concentrations in environmental samples because of the following aspects: high sensitivity; fast chemical separation of interfering elements; low cost; and operational simplicity. (Author) [pt

Study of the desintegration of short-life fission products. Application to the mass distribution in the fission of 238U and 233U induced by 14MeV neutrons

International Nuclear Information System (INIS)

Cavallini, Pierre.

1975-01-01

Nuclear spectrometry of short-life fission products was investigated, together with direct applications to the study of mass and charge distribution in fission reactions. It is shown that, by choosing judiciously the target in which the fission product is created and owing to the differences in stabilities and evaporation temperatures of the compounds obtained, it is possible to separate some elements. For example, niobium was separated by heating after irradiation of a mixture of UC and RuCl 3 , and sublimation in a tube with temperature gradient. It was thus possible to study the 99 Nb isotope. Other classical chemical separation processes were used for yttrium and strontium. The half-lifes beta and gamma spectra, decay schemes of 93 Sr, 94 Y and 95 Y were studied. It was shown how to obtain mass distribution in fission using a nondestructive gamma analysis method. As an application, results obtained in the fission of 233 U and 238 U at 14 MeV are given [fr

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

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

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

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)

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

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

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

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.

Postirradiation examinations of the BG-9 element

International Nuclear Information System (INIS)

Strain, R.V.; Renfro, C.W.; Neimark, L.A.

1976-10-01

Postirradiation examinations were performed on the GB-9 element irradiated in the Oak Ridge Reactor. This vented fuel element was irradiated to a peak burnup of 54,000 MWd/MT at a peak power rating of 14.8 kW/ft and a maximum outside surface temperature of 685 +- 15 0 C. The maximum diametral increase of the element was 0.2% ΔD/D. Volatile fission products migrated to the ends of the fuel column but had not been transported beyond the blanket region of the element. Fission-product attack of the cladding was found to a depth of 4.4 mils adjacent to the mixed-oxide fuel. The maximum fission-product attack (approximately 6.2 mils) was found adjacent to slightly hyperstoichiometric UO 2 pellets at the upper end of the fuel column. The more severe attack in this region has been attributed to the higher oxygen potential adjacent to the hyperstoichiometric UO 2 fuel. A stress-rupture test was performed on a section of the element. The cladding ruptured after 30 min at 700 0 C and approximately 21,500-psi loop stress, resulting in a diametral strain of 1.6%

Experimental evaluation of the structural behaviour of adobe masonry structural elements

OpenAIRE

Varum, H.; Costa, A.; Pereira, H.; Almeida, J.; Rodrigues, H.; Silveira, D.

2007-01-01

Rehabilitation and strengthening of existing adobe masonry constructions have been neglected during the last decades. In Aveiro, Portugal, many adobe buildings present an important level of structural damage and, in many cases, are even near to ruin, having the majority a high vulnerability to seismic actions. To face the lack of information concerning the mechanical properties and structural behaviour of adobe elements, it was developed an experimental campaign. The composition and mechanica...

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

Mechanical properties and structure of Zircaloy attached by UO2+x and fission products

International Nuclear Information System (INIS)

Holub, F.

1987-08-01

The aim of this project was to determine the combined long-term effect of simulated fission products and hyperstoichiometric uranium dioxide on the mechanical properties and structure of Zircaloy. Three groups of fission product elements or compounds were defined: The rare earth oxides CeO 2 , La 2 O 3 , Nd 2 O 3 , Y 2 O 3 ; The metals No, Ru, Ag; The low melting elements Te, Sb and Cd. Each of these groups of fission products was mixed with UO 2+x in proportion related for burnups of 5, 10 and 30%. The simulated fuel mixtures were filled into tubular Zircaloy casings, plugged and welded. These specimens were annealed at 350, 500 and 700 deg. C up to 17,500 hours. The test results indicate different kinds of action of the simulated fuel constituents. Mixtures of rare earth oxides and UO 2+x embrittle Zircaloy drastically at higher temperatures. There exists a mutual intensifying effect of rare earth oxides and UO 2+x . UO 2+x and (Mo + Ru + Ag) and their mixtures act very similar on Zircaloy. The low melting fission products (Te + Sb + Cd) influence the ductility of Zircaloy in an advantageous manner, compared to pure UO 2+x fuel. The layer of zirconium tellurides seems to protect the Zircaloy metal against the embrittling attack of oxygen from UO 2+x . The most important events of tensile tests at 400 deg. C are the high values of the elongation of specimens which are brittled at room temperature. It should guarantee the integrity of fuel elements, which have been attacked chemically by fission products at temperatures of 400 deg. C and higher

On behaviour of fuel elements subject to combined cyclic thermomechanical loads

International Nuclear Information System (INIS)

Hsu, T.R.

1980-01-01

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

Ceramic Hosts for Fission Products Immobilization

Energy Technology Data Exchange (ETDEWEB)

Peter C Kong

2010-07-01

Natural spinel, perovskite and zirconolite rank among the most leach resistant of mineral forms. They also have a strong affinity for a large number of other elements and including actinides. Specimens of natural perovskite and zirconolite were radioisotope dated and found to have survived at least 2 billion years of natural process while still remain their loading of uranium and thorium . Developers of the Synroc waste form recognized and exploited the capability of these minerals to securely immobilize TRU elements in high-level waste . However, the Synroc process requires a relatively uniform input and hot pressing equipment to produce the waste form. It is desirable to develop alternative approaches to fabricate these durable waste forms to immobilize the radioactive elements. One approach is using a high temperature process to synthesize these mineral host phases to incorporate the fission products in their crystalline structures. These mineral assemblages with immobilized fission products are then isolated in a durable high temperature glass for periods measured on a geologic time scale. This is a long term research concept and will begin with the laboratory synthesis of the pure spinel (MgAl2O4), perovskite (CaTiO3) and zirconolite (CaZrTi2O7) from their constituent oxides. High temperature furnace and/or thermal plasma will be used for the synthesis of these ceramic host phases. Nonradioactive strontium oxide will be doped into these ceramic phases to investigate the development of substitutional phases such as Mg1-xSrxAl2O4, Ca1-xSrxTiO3 and Ca1-xSrxZrTi2O7. X-ray diffraction will be used to establish the crystalline structures of the pure ceramic hosts and the substitution phases. Scanning electron microscopy and energy dispersive X-ray analysis (SEM-EDX) will be performed for product morphology and fission product surrogates distribution in the crystalline hosts. The range of strontium doping is planned to reach the full substitution of the divalent

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

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.

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)

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)

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)

Vented nuclear fuel element

International Nuclear Information System (INIS)

Oguma, M.; Hirose, Y.

1976-01-01

A description is given of a vented nuclear fuel element having a plenum for accumulation of fission product gases and plug means for delaying the release of the fission product gases from the plenum, the plug means comprising a first porous body wettable with a liquid metal and a second porous body non-wettable with the liquid metal, the first porous body being impregnated with the liquid metal and in contact with the liquid metal

The HTR modular power reactor system. Qualification of fuel elements and materials

International Nuclear Information System (INIS)

Heidenreich, U.; Breitling, H.; Nieder, R.; Ohly, W.; Mittenkuehler, A.; Ragoss, H.; Seehafer, H.J.; Wirtz, K.; Serafin, N.

1989-01-01

For further development of the HTR modular power reactor system (HTR-M-KW), the project activities for 'Qualification of fuel elements and materials' reported here cover the work for specifying the qualifications to be met by metallic and ceramic materials, taking into account the design-based requirements and the engineered safety requirements. The fission product retention data determined for the HTR modular reactor fuel elements could be better confirmed by evaluation of the experiments, and have been verified by various calculation methods for different operating conditions. The qualification of components was verified by strength analyses including a benchmark calculation for specified normal operation and emergencies; the results show a convenient behaviour of the components and their materials. In addition, a fuel element burnup measuring system was designed that applies Cs-137 gamma spectroscopy; its feasibility was checked by appropriate analyses, and qualification work is in progress. The installation of a prototype measurement system is the task for project No. 03 IAT 211. (orig.) [de

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)

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.

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)

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

Nuclear fuel element

International Nuclear Information System (INIS)

1974-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding, and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

Nuclear fuel element

International Nuclear Information System (INIS)

Thompson, J.R.; Rowland, T.C.

1976-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting, fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

Development of computer models for fuel element behaviour in water reactors

International Nuclear Information System (INIS)

Gittus, J.H.

1987-03-01

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

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

Investigations concerning the source term for the emission of fission products and transuranic elements from the highly radioactive waste in the temperature region between 200 and 11000C

International Nuclear Information System (INIS)

Baumgartner, F.; Krebs, K.; Merte, B.

1984-04-01

Starting point of these investigations was the research report 290 of the former Institute for Reactor Safety, Cologne. The largely incorrect assumptions made in the chemical statement led to high radioactive doses not only for the closer, but also for the further surroundings of a reprocessing plant in case that the cooling system of the fuel element storage or the storage tank for the highly radioactive waste (HAW) should become inoperative for an extended period of time. However, as a result of this temperature region which we used in our experiments, the investigations concerning the source term of the HAW between 1100 and 1200 0 C became interesting again, because a glass melt of about 1100 0 C is used during the PAMELA-process for the vitrification of the HAW. The aqueous HAW-solution is thereby poured onto a hot glass melt. The solution thereby dries up, the HAW-salt is calcined and finally sinks into the glass melt. From today's point of view an investigation concerning the source term of the fission products and transuranic elements from the HAW at 1100 0 C therefore provides information regarding the fission product- and transuranic element volatilization within the melting furnace during the vitrification process. The results may now therefore also be considered a contribution for the layout of the off-gas purification system during the PAMELA-process

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

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

Fission of intermediate mass nuclei by bremsstrahlung photons in the energy range 0.8-1.8 GeV

International Nuclear Information System (INIS)

Lima, D.A. de.

1983-01-01

The fission of intermediate mass nuclei in the Al-Ta internal induced by bremsstrahlung photons of maximum energies between 0,8 to 1,8 GeV is studied. Thin targets of Nd and Sm and dense targets of Al,Ti,Co,Zr,Nb,Ag,In and Ta are utilized, and all the aspects related with the fission fragment absorption by the targets themselves are considered. The samples are exposed in th 2,5 GeV Electron Synchrotron at Bonn University. Muscovite mica, CR-39 and makrofol are used as fission fragments detectors. Fission cross sections and nuclear fissionabilities of the studied elements are estimated. (L.C.) [pt

Detection and location of leaking TRIGA fuel elements

International Nuclear Information System (INIS)

Bouchey, G.D.; Gage, S.J.

1970-01-01

Several TRIGA facilities have experienced difficulty resulting from cladding failures of aluminum clad TRIGA fuel elements. Recently, at the University of Texas at Austin reactor facility, fission product releases were observed during 250 kW operation and were attributed to a leaking fuel element. A rather extensive testing program has been undertaken to locate the faulty element. The used sniffer device is described, which provides a quick, easily constructed, and extremely sensitive means of locating leaking fuel elements. The difficulty at The University of Texas was compounded by extremely low levels and the sporadic nature of the releases. However, in the more typical situation, in which a faulty element consistently releases relatively large quantities of fission gas, such a device should locate the leak with little difficulty

Models and methods for predicting the release of fission products during hypothetical accidents in HTGRs

International Nuclear Information System (INIS)

Bailly, H.W.

1988-01-01

The paper deals with experiments, computational models and methods used to describe the fission product transport (diffusion and particle failure) in the fuel elements of a pebble-bed high-temperature module reactor (HTGR Module) during hypothetical accidents. The codes which describe the diffusion of fission products in the fuel elements are e.g. GETTER and FRESCO. PANAMA, IA/KWU failure function and the so called GOODIN models describe the particle failure. All these models may be used in the risk analysis. The experimental results obtained at the Nuclear Research Center Julich, Germany are discussed and compared with the model calculations for these experiments

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)

Infiltration behaviour of elemental mercury DNAPL in fully and partially water saturated porous media

NARCIS (Netherlands)

D'Aniello, Andrea; Hartog, Niels; Sweijen, Thomas; Pianese, Domenico

Mercury is a contaminant of global concern due to its harmful effects on human health and for the detrimental consequences of its release in the environment. Sources of liquid elemental mercury are usually anthropogenic, such as chlor-alkali plants. To date insight into the infiltration behaviour of

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)

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)

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

International Nuclear Information System (INIS)

Hastings, I.J.; McCracken, D.

1984-09-01

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

Crystal Plasticity Finite Element Analysis of Loading-Unloading Behaviour in Magnesium Alloy Sheet

International Nuclear Information System (INIS)

Hama, Takayuki; Fujimoto, Hitoshi; Takuda, Hirohiko

2010-01-01

Magnesium alloy sheets exhibit strong inelastic response during unloading. In this study crystal plasticity finite element analysis of loading-unloading behaviour during uniaxial tension in a rolled magnesium alloy sheet was carried out, and the mechanism of this inelastic response was examined in detail in terms of macroscopic and mesoscopic deformations. The unloading behaviour obtained by the simulation was in good agreement with the experiment in terms of variation with stress of instantaneous tangent modulus during unloading. Variations of activities of each family of slip systems during the deformation showed that the activation of basal slip systems is the largest during unloading, and the slip direction during unloading is opposite from during loading. These results indicated that one of the factors of the inelastic behaviour during unloading is the fact that the basal slip systems are easily activated during unloading because of their low strengths.

Separation and utilization of fission products considering economic aspects

International Nuclear Information System (INIS)

Beer, M.; Gorski, B.; Hennrich, M.; Pfrepper, G.; Richter, M.

1982-01-01

The quantity of usable fission products which will be obtained by nuclear fission till the year 2000 is estimated on the basis of prognostics for the development of nuclear energy in the world considering especially the development in the U.S.S.R. and the CMEA. The possibilities of utilization of cesium as gamma-ray source are discussed, and the present fields of application of palladium and the development of its price on the world market are shown. The fields of application of technetium, which wasn't available as artificial element in a greater quantity till now, have to be developed. The economic estimations base on data of a project for the separation of fission products in connection with a reprocessing plant, which was developed in the U.S.A. in 1978. The data show, that it is possible to produce the platinum metals and cesium with profit, the same can be expected for technetium. (author)

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

Asymmetric fission of 47V induced by the 23Na+24Mg reaction

International Nuclear Information System (INIS)

Beck, C.; Djerroud, B.; Haas, F.; Freeman, R.M.; Hachem, A.; Heusch, B.; Morsad, A.; Vuillet-A-Cilles, M.; Sanders, S.J.

1993-01-01

The properties of fully energy-damped processes (deep-inelastic orbiting, fusion-evaporation, and fusion-fission processes) have been investigated in the nearly mass-symmetric entrance-channel 23 Na + 24 Mg reaction leading to the 47 V compound nucleus. By comparison with previous data for the mass-asymmetric 35 Cl + 12 C reaction forming the same compound system at the same excitation energy, no entrance-channel effects are observed in either the evaporation residue or the fusion-fission yields. This is in contrast to the situation with the 28 Si + 12 C and 24 Mg + 16 O reactions where an orbiting process is evident. The asymmetrical elemental distributions of the fusion-fission fragments of the massA=47 system are well described by a transition-state model that accounts for the spin and mass-asymmetry dependence of the fission saddle point

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

Thermodynamic parameters and transport coefficients of the U-C-F gas mixture in the steady flow gaseous core fission reactor

International Nuclear Information System (INIS)

Berg, M.S. van den.

1995-01-01

Thermodynamic parameters and transport coefficients have been calculated for a multicomponent reacting U-C-F gas mixture in the steady flow gaseous core fission reactor. Element abundances are consistent with thermodynamic equilibrium between the gas mixture and a cooled solid graphite wall at 2500 K. Results are presented for various pressures, a fluorine potential of 5.6 and temperatures between 2500 and 7000 K. As a result of dissociation processes of uranium and carbon fluoride compounds, ''effective'' values of thermodynamic parameters and transport coefficients show anomalous behaviour with respect to so-called ''frozen'' values. The chemical reaction energy of the U-C-F gas mixture has been calculated as the driving-force behind the process of fuel redistribution to attain criticality conditions inside a functioning reactor. (author)

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

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.

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

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.

Fission gas retention in irradiated metallic fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Gruber, E.; Kramer, J.M.

1987-01-01

Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5 wt. % Fs) are presented. (The symbol 'Fs' designates fissium, a 'pseudo-element' which, in reality, is an alloy whose composition is representative of fission products that remain in reprocessed fuel). The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by collecting the gases released from short segments of EBR-II driver fuel that were melted in a gas-tight furnace. Comparison of the calculations with the measurements shows quantitative agreement in both the magnitude and the axial variation of the retained gas content. (orig.)

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.

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

International Nuclear Information System (INIS)

Languille, A.

1984-02-01

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

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).

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

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

Stability of superheavy elements around Z=120 probed by the study of their fission time obtained by the crystal blocking method; Stabilite d'elements superlourds au voisinage de Z=120 testee par l'etude de leurs temps de fission deduits par la methode du blocage cristallin

Energy Technology Data Exchange (ETDEWEB)

Laget, M

2007-10-15

While the existence of an island of stability beyond Z=110 is theoretically acquired, the location of this island ranges from Z=114 to Z=126 depending on models. In this work, the stability of super-heavy nuclei is probed through the study of their fission time. The chosen experimental method, the crystal blocking method, is sensitive to the presence of possible long time components in the fission time distribution which indicates a fission mechanism occurring after the formation of a compound nucleus. The blocking dips were therefore constituted for the various products of the reaction U{sup 238} + Ni (6.6 MeV/A) {yields} 120, the experimental set-up allowing us to clearly identify and select the reaction mechanisms. The comparison of the blocking dip constituted for quasi-elastic scattering events with the one obtained for the fission fragments of a Z=120, combined with the study of kinematical properties of these fission fragments, give evidences of the existence of very long fission times (> 10{sup -18} s) only compatible with a fusion-fission mechanism implying a non vanishing fission barrier height for Z=120. The second part outlines microscopic calculations of fission barrier heights, carried out in the framework of the finite temperature of the Hartree-Fock-Bogoliubov (HFB) theory. Because of the progressive vanishing of the pairing correlation with T, which happens differently at the ground state and at the top of the barrier, B{sub f} first grows until T {approx_equal} 0.8 MeV before dropping with T owing to shell-effects damping with temperature. (author)

Emergent Behaviour

NARCIS (Netherlands)

Blom, H.A.P.; Everdij, M.H.C.; Bouarfa, S.; Cook, A; Rivas, D

2016-01-01

In complexity science a property or behaviour of a system is called emergent if it is not a property or behaviour of the constituting elements of the system, though results from the interactions between its constituting elements. In the socio-technical air transportation system these interactions

Neutron rich clusters and the dynamics of fission and fusion

International Nuclear Information System (INIS)

Armbruster, P.

1988-07-01

In this lecture I want to discuss experimental evidence for the appearance of cluster aspects in the dynamics of large rearrangement processes, as fusion and fission. Clusters in the sense as used in my lecture are the strongly bound doubly magic nuclei as 20 Ca 28 48 , 28 Ni 50 78 , 132 50 Sn 82 , and 208 82 Pb 126 and the superheavy nucleus 298 114 184 . Two of these nuclei, 78 Ni and 298 114 have not yet been identified. I discuss first the experimental findings from heavy element production. Then I cover the stability of cluster aspects to intrinsic excitation energy in fusion and fission. (orig./HSI)

Spontaneous-fission half-lives for even nuclei with Z> or =92

International Nuclear Information System (INIS)

Randrup, J.; Larsson, S.E.; Moller, P.; Nilsson, S.G.; Pomorski, K.; Sobiczewski, A.

1976-01-01

The spontaneous-fission process for doubly even nuclei with Z> or =92 is studied in a semiempirical WKB framework. One-dimensional fission barrier potentials are established from theoretical deformation-energy surfaces based on the droplet model and the modified-oscillator model. The effects of axial asymmetry as well as reflection asymmetry have been taken into account. Macroscopic (irrotational flow) inertial-mass functions and, alternatively, microscopic (cranking model) inertial mass parameters have been employed for the calculation of the fission half-lives. With one over-all normalization parameter it is possible to fit the experimental half-lives to within a factor of 20 on the average. The resulting effective inertial-mass functions are used to estimate the stability of the transactinide elements. Only minor differences with previous estimates for the r process and superheavy nuclei are encountered

Advanced system for separation of rare-earth fission products

International Nuclear Information System (INIS)

Baker, J.D.; Gehrke, R.J.; Greenwood, R.C.; Meikrantz, D.H.

1982-01-01

A microprocessor-controlled radiochemical separation system has been further advanced to separate individual rare-earth elements from mixed fission products in times of a few minutes. The system was composed of an automated chemistry system fed by two approximately 300 μg 252 Cf sources coupled directly by a He-jet to transport the fission products. Chemical separations were performed using two high performance liquid chromatography columns coupled in series. The first column separated the rare-earth group by extraction chromatography using dihexyldiethylcarbamoylmethylphosphonate (DHDECMP) adsorbed on Vydac C 8 resin. The second column isolated the individual rare-earth elements by cation exchange chromatography using Aminex A-9 resin with α-hydroxyisobutyric acid (α-HIBA) as the eluent. Significant results, which have been obtained to date with this advanced system, are the identification of several new neutron-rich rare-earth isotopes including 155 Pm (T=48+-4 s) and 163 Gd (T=68+-3 s). In addition, a half-life of 41+-4 s is reported for 160 Eu. (author)

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

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)

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)

Steady-state fission gas behavior in uranium-plutonium-zirconium metal fuel elements

International Nuclear Information System (INIS)

Steele, W.G.; Wazzan, A.R.; Okrent, D.

1989-01-01

An analysis of fission gas release and induced swelling in steady state irradiated U-Pu-Zr metal fuels is developed and computer coded. The code is used to simulate, with fair success, some gas release and induced swelling data obtained under the IFR program. It is determined that fuel microstructural changes resulting from zirconium migration, anisotropic swelling, and thermal variations are major factors affecting swelling and gas release behavior. (orig.)

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

International Nuclear Information System (INIS)

Denis, Alicia C.; Soba, Alejandro

2000-01-01

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

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

Stabilization effect of fission source in coupled Monte Carlo simulations

Energy Technology Data Exchange (ETDEWEB)

Olsen, Borge; Dufek, Jan [Div. of Nuclear Reactor Technology, KTH Royal Institute of Technology, AlbaNova University Center, Stockholm (Sweden)

2017-08-15

A fission source can act as a stabilization element in coupled Monte Carlo simulations. We have observed this while studying numerical instabilities in nonlinear steady-state simulations performed by a Monte Carlo criticality solver that is coupled to a xenon feedback solver via fixed-point iteration. While fixed-point iteration is known to be numerically unstable for some problems, resulting in large spatial oscillations of the neutron flux distribution, we show that it is possible to stabilize it by reducing the number of Monte Carlo criticality cycles simulated within each iteration step. While global convergence is ensured, development of any possible numerical instability is prevented by not allowing the fission source to converge fully within a single iteration step, which is achieved by setting a small number of criticality cycles per iteration step. Moreover, under these conditions, the fission source may converge even faster than in criticality calculations with no feedback, as we demonstrate in our numerical test simulations.

Stabilization effect of fission source in coupled Monte Carlo simulations

Directory of Open Access Journals (Sweden)

Börge Olsen

2017-08-01

Full Text Available A fission source can act as a stabilization element in coupled Monte Carlo simulations. We have observed this while studying numerical instabilities in nonlinear steady-state simulations performed by a Monte Carlo criticality solver that is coupled to a xenon feedback solver via fixed-point iteration. While fixed-point iteration is known to be numerically unstable for some problems, resulting in large spatial oscillations of the neutron flux distribution, we show that it is possible to stabilize it by reducing the number of Monte Carlo criticality cycles simulated within each iteration step. While global convergence is ensured, development of any possible numerical instability is prevented by not allowing the fission source to converge fully within a single iteration step, which is achieved by setting a small number of criticality cycles per iteration step. Moreover, under these conditions, the fission source may converge even faster than in criticality calculations with no feedback, as we demonstrate in our numerical test simulations.

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.

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

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.)

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

International Nuclear Information System (INIS)

Dadillon, J.

1976-11-01

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

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.

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.

Fission product concentration evolution in sodium pool following a fuel subassembly failure in an LMFBR

International Nuclear Information System (INIS)

Natesan, K.; Velusamy, K.; Selvaraj, P.; Kasinathan, N.; Chellapandi, P.; Chetal, S.; Bhoje, S.

2003-01-01

During a fuel element failure in a liquid metal cooled fast breeder reactor, the fission products originating from the failed pins mix into the sodium pool. Delayed Neutron Detectors (DND) are provided in the sodium pool to detect such failures by way of detection of delayed neutrons emitted by the fission products. The transient evolution of fission product concentration is governed by the sodium flow distribution in the pool. Transient hydraulic analysis has been carried out using the CFD code PHOENICS to estimate fission product concentration evolution in hot pool. k- ε turbulence model and zero laminar diffusivity for the fission product concentration have been considered in the analysis. Times at which the failures of various fuel subassemblies (SA) are detected by the DND are obtained. It has been found that in order to effectively detect the failure of every fuel SA, a minimum of 8 DND in hot pool are essential

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)

Study of distribution coefficient of some radionuclides in liquid-solid systems

International Nuclear Information System (INIS)

Ghafar, M.; Abdul-Hadi, A.; Al-Hassanieh, O.

2001-10-01

The distribution of the elements U, Th, Am, Eu, Cs, Sr and Ra in a solid-aqueous system. Natural phosphate in contact with groundwater, was investigated usingγ and α spectroscopy. The effect of various factors such as phase contact time, Ph, particle size of the solid phase, and the concentration of a element Ca, were studied. The obtained results show that, more than 98% of the actinide elements and europium are adsorbed by the solid phase under all conditions. The fission products Cs and Sr have different behaviours, depending on experiment conditions. The behaviour of Ra is closer to the actinides than to the fission products. There are small differences between the behaviour of the actinide elements, which can be interpreted by migration mechanism from the aqueous to the solid phase i.e. adsorption or precipitation. (author)

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

Observed mass distribution of spontaneous fission fragments from samples of lime - an SSNTD study

CERN Document Server

Paul, D; Ghose, D; Sastri, R C

1999-01-01

SSNTD is one of the most commonly used detectors in the studies involving nuclear phenomena. The ease of registration of the presence of alpha particles and fission fragments has made it particularly suitable in studies where stable long exposures are needed to extract reliable information. Studies on the presence of alpha emitting nuclides in the environment assume importance since they are found to be carcinogenic. Lime samples from Silchar in Assam of Eastern India have shown the presence of spontaneous fission fragments besides alphas. In the present study we look at the ratio of the average mass distribution of these fission fragments, that gives us an indication of the presence of the traces of transuranic elements.

Finite element simulation of the welding process and structural behaviour of welded components

International Nuclear Information System (INIS)

Locci, J.M.; Rouvray, A. de; Barbe, B.; Poirier, J.

1977-01-01

In the field of inelastic analysis of nuclear metal structures, the computation of residual stresses in welds, and their effects on the strength of welded components is of major importance. This paper presents an experimentally checked finite element simulation with the general nonlinear program PAM NEP-D, of the electron beam welding of two thick hemispherical shells, and the behaviour of the welded sphere under various additional thermomechanical sollicitations. (Auth.)

On fission product retention in the core of the low powered high temperature reactor under accident conditions

International Nuclear Information System (INIS)

Bastek, H.

1984-01-01

In the core of the high temperature reactor the fuel element and the coated particles contained herein provide the safest enclosure for fission products. The complex process of fission product transport out of the particle kernel, through the particle coating and within the fuel element graphite is described in a simplified form by the Fick's diffusion. The effective diffusion coefficient is used for calculation. Starting from the existing ideas of fission product transport five burn-up and temperature-dependent diffusion coefficients for Cesium in (Th,U)O 2 -kernels are derived in this study. The results have been gained from several fuel element radiation experiments in recent years, which showed extreme variation in regard to burn-up, temperature cycle, neutron flux and operation time. Cs-137 release measurements from single particle kernels were present from all the experiments. Furthermore, annealing tests of AVR-fuel elements were analyzed. Heat-temperatur and heating-time, the fuel element burn-up in the AVR-reactor, as well as the measured Cs-137 inventory of the fuel elements before and after annealing, are included in the investigation as essential parameters. With the aid of the derived diffusion coeffizients and already present data sets the Cs-137 release of fuel elements into a small reactor core is investigated under unrestricted core heat-up. While the released Cs-137 is derived mainly from defective particles at accident temperatures up to 1600 0 C, the main part diffuses through the particle coating at higher accident temperatures. (orig./HP) [de

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.

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.

Kr and Xe isotopes from spontaneous fission of 248Cm and 250Cf

International Nuclear Information System (INIS)

Srinivasan, B.

1980-01-01

Relative yields of Kr and Xe isotopes from the spontaneous fission of 248 Cm and 250 Cf have been determined mass spectrometrically. The yields are as follows: 83 Kr/ 84 Kr/ 85 Kr/ 86 Kr = 0.223/0.458/0.596/ identical 1.00 and 0.306/0.582/0.793/ identical 1.00; 131 Xe/ 134 Xe/ 136 Xe = 0.486/0.819/1.075 identical 1.00 and 0.343/0.506/0.851/ identical 1.00 from 248 Cm and 250 Cf, respectively. The Xe yields from 248 Cm agree with an earlier determination by Leich et al. Neither of these yield patterns matches that of fissiogenic Kr and Xe in carbonaceous chondrites and hence 248 Cm and 250 Cf are ruled out as progenitors of the meteoritic Kr and Xe. In general, none of the spontaneously fissioning nuclides of actinide elements can be identified as a possible progenitor. Even the mixtures of actinides, including a combination of 248 Cm and 250 Cm, are unsuitable. The origin of anomalous Kr and Xe in carbonaceous chondrites must then be traced either to the spontaneous fission of a superheavy element or to peculiarities in specific nucleosynthetic reactions. (orig.)

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

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

FINITE ELEMENT ANALYSIS OF THE BEHAVIOUR OF REINFORCED CONCRETE COLUMNS CONFINED BY OVERLAPPING HOOPS SUBJECTED TO RAPID CONCENTRIC LOADING

Directory of Open Access Journals (Sweden)

Xiang Zeng

2017-12-01

Full Text Available The strain rate sensitivity of concrete material was discovered approximately one hundred years ago, and it has a marked effect on the behaviour of concrete members subjected to dynamic loadings such as strong earthquake and impact loading. Because of the great importance of the confined reinforced concrete (RC columns in RC structures, the dynamic behaviour of the columns induced by the strain rate effect has been studied, but only few experiments and analyses have been conducted. To investigate the behaviour of overlapping hoop-confined square reinforced normal-strength concrete columns, considering the strain rate effect at a strain rate of 10-5/sec to 10-1/sec induced by earthquake excitation, an explicit dynamic finite element analysis (FEA model was developed in ABAQUS to predict the behaviour of confined RC columns subjected to the rapid concentric loading. A locally modified stress-strain relation of confined concrete with the strain rate sensitivity of the concrete material and the confining effect of overlapping hoops were proposed to complete the simulation of the dynamic behaviour of concrete with the concrete plastic-constitutive model in ABAQUS. The finite element predictions are consistent with the existing test results. Based on the FEA model, a parametric investigation was conducted to capture more information about the behaviour of confined RC columns under varying loading rates.

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

Quantitative analysis of fission products by {gamma} spectrography; Analyse quantitative des produits de fission par spectrographie {gamma}

Energy Technology Data Exchange (ETDEWEB)

Malet, G

1962-07-01

The activity of the fission products present in treated solutions of irradiated fuels is given as a function of the time of cooling and of the irradiation time. The variation of the ratio ({sup 144}Ce + {sup 144}Pr activity/{sup 137}Cs activity) as a function of these same parameters is also given. From these results a method is deduced giving the 'age' of the solution analyzed. By {gamma}-scintillation spectrography it was possible to estimate the following elements individually: {sup 141}Ce, {sup 144}Ce + {sup 144}Pr, {sup 103}Ru, {sup 106}Ru + {sup 106}Rh, {sup 137}Cs, {sup 95}Zr + {sup 95}Nb. Yield curves are given for the case of a single emitter. Of the various existing methods, that of the least squares was used for the quantitative analysis of the afore-mentioned fission products. The accuracy attained varies from 3 to 10%. (author) [French] L'activite des produits de fission presents dans les solutions de traitement de combustibles irradies est donnee en fonction du temps de refroidissement et du temps d'irradiation. On etudie de plus la variation du rapport Activite du {sup 144}Ce + {sup 144}Pr /Activite du {sup 137}Cs en fonction de ces memes parametres. De ces resultats, on deduit une methode donnant l'age de la solution analysee. La spectrographie {gamma} a scintillation a permis le dosage individuel des produits suivants: {sup 141}Ce, {sup 144}Ce + {sup 144}Pr, {sup 103}Ru, {sup 106}Ru + {sup 106}Rh, {sup 137}Cs, {sup 95}Zr + {sup 95}Nb. Des courbes de rendement sont donnees dans le cas d'un emetteur unique. Des differentes methodes existantes, la methode des moindres carres a ete employee pour l'analyse quantitative des produits de fission precites. La precision obtenue varie entre 3 et 10 pour cent. (auteur)

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

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.)

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

Quantitative analysis of fission products by γ spectrography

International Nuclear Information System (INIS)

Malet, G.

1962-01-01

The activity of the fission products present in treated solutions of irradiated fuels is given as a function of the time of cooling and of the irradiation time. The variation of the ratio ( 144 Ce + 144 Pr activity)/ 137 Cs activity) as a function of these same parameters is also given. From these results a method is deduced giving the 'age' of the solution analyzed. By γ-scintillation spectrography it was possible to estimate the following elements individually: 141 Ce, 144 Ce + 144 Pr, 103 Ru, 106 Ru + 106 Rh, 137 Cs, 95 Zr + 95 Nb. Yield curves are given for the case of a single emitter. Of the various existing methods, that of the least squares was used for the quantitative analysis of the afore-mentioned fission products. The accuracy attained varies from 3 to 10%. (author) [fr

Geochemical properties and nuclear chemical characteristics of Oklo natural fission reactors

Energy Technology Data Exchange (ETDEWEB)

Hidaka, Hiroshi [Hiroshima Univ., Higashi-Hiroshima (Japan). Faculty of Science

1997-07-01

There are six uranium deposits in the Gabonese Republic in the cnetral Africa. `Fission reactor zone`, the fission chain reactions generated about 200 billion years ago, was existed in a part of them. CEA begun geochemical researches of Oklo deposits etc. in 1991. The geochemical and nuclear chemical properties of Oklo were reviewed from the results of researches. Oklo deposits is consisted of main five sedimentary faces such as sandstone (FA), Black Shale formation (FB), mudstone (FC), tuff (FD) and volcaniclastic sandstone (FE) from the bottom on the base rock of granite in the Precambrian era. Uranium is enriched in the upper part of FA layer and the under part of FB layer. {sup 235}U/{sup 238}U, U content, fission proportion, duration time, neutron fluence, temperature, restitution factor of {sup 235}U and epithermal index ({gamma}) were investigated and compared. The geochemical properties of Oklo are as followed: large enrich of uranium, the abundance ratio of {sup 235}U as same as that of enriched uranium, interaction of natural water and small rear earth elements. These factors made casually Oklo fission reactor. (S.Y.)

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.

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.

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

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

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)

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

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.

The transport behaviour of elemental mercury DNAPL in saturated porous media: Analysis of field observations and two-phase flow modelling

NARCIS (Netherlands)

Sweijen, T.; Hartog, Niels; Marsman, A.; Keijzer, T.J.S.

2014-01-01

Mercury is a contaminant of global concern. The use of elemental mercury in various (former) industrial processes, such as chlorine production at chlor-alkali plants, is known to have resulted in soil and groundwater contaminations worldwide. However, the subsurface transport behaviour of elemental

European institute for transuranium elements. Annual report 1987

International Nuclear Information System (INIS)

1988-01-01

The Report covers work done in the field of operation limits of nuclear fuels, transient behaviour of oxide fuels and fission product release under severe fuel damage conditions, actinide cycle safety, actinide research, safeguards reactor safety waste management and radiation protection

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

The Munich accelerator for fission fragments MAFF

International Nuclear Information System (INIS)

Habs, D.; Gross, M.; Assmann, W.; Ames, F.; Bongers, H.; Emhofer, S.; Heinz, S.; Henry, S.; Kester, O.; Neumayr, J.; Ospald, F.; Reiter, P.; Sieber, T.; Szerypo, J.; Thirolf, P.G.; Varentsov, V.; Wilfart, T.; Faestermann, T.; Kruecken, R.; Maier-Komor, P.

2003-01-01

The Munich Accelerator for Fission Fragments MAFF has been designed for the new Munich research reactor FRM-II. It will deliver several intense beams (∼3x10 11 s -1 ) of very neutron-rich fission fragments with a final energy of 30 keV (low-energy beam) or energies between 3.7 and 5.9 MeV·A (high-energy beam). Such beams are of interest for the creation of super-heavy elements by fusion reactions, nuclear spectroscopy of exotic nuclei, but they also have a potential for applications, e.g. in medicine. Presently the Munich research reactor FRM-II is ready for operation, but authorities delay the final permission to turn the reactor critical probably till the end of 2002. Only after this final permission the financing of the major parts of MAFF can start. On the other hand all major components have been designed and special components have been tested in separate setups

Detector for gaseous nuclear fission products

International Nuclear Information System (INIS)

Kobayashi, Yoshihiro; Kubo, Katsumi.

1979-01-01

Purpose: To facilitate the fabrication of a precipitator type detector, as well as improve the reliability. Constitution: Gas to be measured flown in an anode is stored in a gas processing system. By applying a voltage between the anode and the cathode, if positively charged Rb or Cs which is the daughter products of gaseous fission products are present in the gas to be measured, the daughter products are successively deposited electrostatically to the cathode. The daughter products issue beta-rays and gamma-rays to ionize the argon gas at the anode, whereby ionizing current flows between both of the electrodes. Pulses are generated from the ionizing current, and presence or absence, as well as the amount of the gaseous fission products are determined by the value recorded for the number of the pulses to thereby detect failures in the nuclear fuel elements. After the completion of the detection, the inside of the anode is evacuated and the cathode is heated to evaporate and discharge the daughter products externally. This eliminates the effects of the former detection to the succeeding detection. (Moriyama, K.)

Behaviour of gas cooled reactor fuel under accident conditions

International Nuclear Information System (INIS)

1991-11-01

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

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

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).

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)

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)

Behaviour of trace element concentration in human organs in dependence of age and environment

International Nuclear Information System (INIS)

Persigehl, M.; Schicha, H.; Kasperek, K.; Feinendegen, L.E.; Kernforschungsanlage Juelich G.m.b.H.

1977-01-01

To study the behaviour of trace elements in dependence of age and environment, samples of skin, lung, heart, aorta, kidney, liver and brain were assayed for concentrations of Fe, Zn, Rb, Co, Cr, Se, Sc, Sb, Cs, Al and partly Eu. All samples were dried at 100 deg C for two days. Instrumental neutron-activation analysis was used to determine the element concentrations. The neutron flux was 5 x 10 13 n cm -2 sec -1 . After decay of the short lived radioisotopes, the Al-concentration was measured following a second irradiation of 1 minute and directly comparing with a standard sample. Nearly all element concentrations changed with processing age, but they showed no clear correlation to either parameter assessed. The non-essential elements Se, Sb and Sc were increasingly concentrated in all organs except the skin. Comparing lung samples of patients from highly industrialized regions with those of lesser industrialization, the elements Sc, Al, Sb, Eu and Co were accumulated by a factor of 10 to 100. Thus the concentrations of trace elements in human organism also depend on the degree of industrialization. (T.G.)

Some methods of failed fuel element detection in water cooled reactors

International Nuclear Information System (INIS)

Strindehag, O.M.

1976-01-01

The methods are surveyed using fission products released in the coolant for the detection of failed fuel elements in water cooled reactors. The classification of the detection methods is made with respect to fission product detection in the coolant and to gaseous fission product detection. The detection systems are listed used for the AGESTA power reactor and for the experimental loops of the RA research reactor based on the detection of either gaseous fission products or gaseous daughter products. The AGESTA reactor detection systems using electrostatic precipitators consist of five precipitator channels of which three are intended for detection and two for localization. A special detection unit was developed for the failed fuel element detection in the R-2 reactor experimental steam loop. Its description is listed. In the reactor pressurized-water loop a Cherenkov counter was used in the detection of fission products. An ion exchange monitor whose application is described was used in the total measurement of the main coolant flow in the AGESTA reactor. (J.P.)

Theoretical predictions of properties and gas-phase chromatography behaviour of bromides of group-5 elements Nb, Ta, and element 105, Db.

Science.gov (United States)

Pershina, V; Anton, J

2012-01-21

Fully relativistic, four-component density functional theory electronic structure calculations were performed for MBr(5), MOBr(3), MBr(6)(-), KMBr(6), and MBr(5)Cl(-) of group-5 elements Nb, Ta, and element 105, Db, with the aim to predict adsorption behaviour of the bromides in gas-phase chromatography experiments. It was shown that in the atmosphere of HBr/BBr(3), the pentabromides are rather stable, and their stability should increase in the row Nb Ta > Db. This sequence is in agreement with the one observed in the "one-atom-at-a-time" chromatography experiments. Some other scenarios, such as surface oxide formation were also considered but found to be irrelevant. © 2012 American Institute of Physics

Heat and Fission Product Transport in a Molten U-Zr-O Pool With Crust

International Nuclear Information System (INIS)

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

2002-01-01

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 pool. 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 was estimated by product of the mass concentration and energy conversion factor of each fission product. For the calculation of heat generation rate in the pool, twenty-nine elements were chosen and classified by their chemical properties. 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 was performed for the TMI-2 accident. The pool is assumed to be a partially filled hemispherical geometry and the change of pool geometry during the numerical calculation was neglected. Results of the numerical calculation revealed that the peak temperature of the molten pool significantly decreased and most of the volatile fission products were released from the molten pool during the accident. (authors)

Fifth International Conference on Fission and Properties of Neutron-Rich Nuclei

CERN Document Server

Ramayya, A V; ICFN5

2014-01-01

These proceedings are the fifth in the series of International Conferences covering fission and properties of neutron-rich nuclei, which are at the forefront of nuclear research. The time interval of 5 years between each conference allows for significant new results to be achieved. Recently, world leaders in theory and experiments in research and the development of new facilities for research presented their latest results in areas such as synthesis of superheavy elements, new facilities for and recent results with radioactive ion beams, structure of neutron-rich nuclei, nuclear fission process, fission yields and nuclear astrophysics. This book is a major source of the latest research in these areas and plans for the future. The conference brought together a unique group of over 100 speakers including leaders from the major nuclear laboratories in Canada, China, France, Finland, Germany, Italy, Japan, Russia, Switerzland and the US along with leading research scientists from around the world.

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

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.)

A study of the consistent and the lumped source approximations in finite element neutron diffusion calculations

International Nuclear Information System (INIS)

Ozgener, B.; Azgener, H.A.

1991-01-01

In finite element formulations for the solution of the within-group neutron diffusion equation, two different treatments are possible for the group source term: the consistent source approximation (CSA) and the lumped source approximation (LSA). CSA results in intra-group scattering and fission matrices which have the same nondiagonal structure as the global coefficient matrix. This situation might be regarded as a disadvantage, compared to the conventional (i.e. finite difference) methods where the intra-group scattering and fission matrices are diagonal. To overcome this disadvantage, LSA could be used to diagonalize these matrices. LSA is akin to the lumped mass approximation of continuum mechanics. We concentrate on two different aspects of the source approximations. Although it has been reported that LSA does not modify the asymptotic h 2 convergence behaviour for linear elements, the effect of LSA on convergence of higher degree elements has not been investigated. Thus, we would be interested in determining, p, the asymptotic order of convergence, in: Δk |k eff (analytical) -k eff (finite element)| = Ch p (1) for finite element approximations of varying degree (N) with both of the source approximations. Since (1) is valid in the asymptotic limit, we must use ultra-fine meshes and quadruple precision arithmetic. For our order of convergence study, we used infinite cylindrical geometry with azimuthal symmetry. Hence, the effects of singularities remain uninvestigated. The second aspect we dwell on is the performance of LSA in bilinear 3-D finite element calculations, compared to CSA. LSA has been used quite extensively in 1- and 2-D even-parity transport and diffusion calculations. In this work, we will try to assess the relative merits of LSA and CSA in 3-D problems. (author)

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.)

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

Sustainable and safe nuclear fission energy technology and safety of fast and thermal nuclear reactors

CERN Document Server

Kessler, Günter

2012-01-01

Unlike existing books of nuclear reactor physics, nuclear engineering and nuclear chemical engineering this book covers a complete description and evaluation of nuclear fission power generation. It covers the whole nuclear fuel cycle, from the extraction of natural uranium from ore mines, uranium conversion and enrichment up to the fabrication of fuel elements for the cores of various types of fission reactors. This is followed by the description of the different fuel cycle options and the final storage in nuclear waste repositories. In addition the release of radioactivity under normal and possible accidental conditions is given for all parts of the nuclear fuel cycle and especially for the different fission reactor types.

Efficient finite element modelling for the investigation of the dynamic behaviour of a structure with bolted joints

Science.gov (United States)

Omar, R.; Rani, M. N. Abdul; Yunus, M. A.; Mirza, W. I. I. Wan Iskandar; Zin, M. S. Mohd

2018-04-01

A simple structure with bolted joints consists of the structural components, bolts and nuts. There are several methods to model the structures with bolted joints, however there is no reliable, efficient and economic modelling methods that can accurately predict its dynamics behaviour. Explained in this paper is an investigation that was conducted to obtain an appropriate modelling method for bolted joints. This was carried out by evaluating four different finite element (FE) models of the assembled plates and bolts namely the solid plates-bolts model, plates without bolt model, hybrid plates-bolts model and simplified plates-bolts model. FE modal analysis was conducted for all four initial FE models of the bolted joints. Results of the FE modal analysis were compared with the experimental modal analysis (EMA) results. EMA was performed to extract the natural frequencies and mode shapes of the test physical structure with bolted joints. Evaluation was made by comparing the number of nodes, number of elements, elapsed computer processing unit (CPU) time, and the total percentage of errors of each initial FE model when compared with EMA result. The evaluation showed that the simplified plates-bolts model could most accurately predict the dynamic behaviour of the structure with bolted joints. This study proved that the reliable, efficient and economic modelling of bolted joints, mainly the representation of the bolting, has played a crucial element in ensuring the accuracy of the dynamic behaviour prediction.

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.

Properties of neutron-rich nuclei studied by fission product nuclear chemistry

International Nuclear Information System (INIS)

Meyer, R.A.; Henry, E.A.; Griffin, H.C.; Lien, O.G. III; Lane, S.M.; Stevenson, P.C.; Yaffe, R.P.; Skarnemark, G.

1979-09-01

A review is given of the properties of neutron-rich nuclei studied by fission product nuclear chemistry and includes the techniques used in elemental isolation and current research on the structure of nuclei near 132 Sn, particle emission, and coexisting structure in both neutron-poor and neutron-rich nuclei. 35 references

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.

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.)

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

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

Mo-99 production by fission and future projections

International Nuclear Information System (INIS)

Carranza, E.C.; Novello, A.; Bronca, M.; Cestau, D.; Bavaro, R.; Centurion, R.; Bravo, C.; Bronca, P.; Gualda, E.; Fraguas, F.; Giomi, A.; Ivaldi, L.

2012-01-01

Description of the I-131 and Mo-99 production process: The process starts with the irradiation of uranium-aluminum mini plates in the RA-3, Argentinean Reactor No.3, Ezeiza Atomic Center. In a nuclear reactor there is a constant flow of neutrons and when a neutron with proper energy impacts on a nucleus of U-235, it is absorbed at the same time generate an unstable configuration nuclear. For this reason, the nucleus formed is fission, getting two different atoms. Approximately 6% of the fissions produce Mo-99 and 3% produce I-131; the percentage remaining corresponds to formation of atoms without interest for use in medicine. In conclusion, the objective of the process developed in the Fission Plant, is starting from uranium mini plates, separate the Mo-99 and I-131 generated, the remaining elements formed. - Evolution of Mo-99 Production in the last 10 years: The Fission Mo-99 Plant Production begins routine production of Mo-99 in 1985, using targets made of uranium enriched at 90% U-235. In the 1990s, global concern regarding the use of highly enriched uranium, due to non-proliferation issues, caused the interruption of supply of nuclear material (HEU enriched at 90% of U-235). Following this, Argentina developed target based on low-enriched uranium (less than 20% U-235), becoming in 2002 the first country in the world to produce Mo-99 with LEU targets. From 2002 to date, the activity produced of Mo-99 has been tripled annually (author)

Irradiation behaviour of advanced fuel elements for the helium-cooled high temperature reactor (HTR)

International Nuclear Information System (INIS)

Nickel, H.

1990-05-01

The design of modern HTRs is based on high quality fuel. A research and development programme has demonstrated the satisfactory performance in fuel manufacturing, irradiation testing and accident condition testing of irradiated fuel elements. This report describes the fuel particles with their low-enriched UO 2 kernels and TRISO coating, i.e. a sequence of pyrocarbon, silicon carbide, and pyrocarbon coating layers, as well as the spherical fuel element. Testing was performed in a generic programme satisfying the requirements of both the HTR-MODUL and the HTR 500. With a coating failure fraction less than 2x10 -5 at the 95% confidence level, the results of the irradiation experiments surpassed the design targets. Maximum accident temperatures in small, modular HTRs remain below 1600deg C, even in the case of unrestricted core heatup after depressurization. Here, it was demonstrated that modern TRISO fuels retain all safety-relevant fission products and that the fuel does not suffer irreversible changes. Isothermal heating tests have been extended to 1800deg C to show performance margins. Ramp tests to 2500deg C demonstrate the limits of present fuel materials. A long-term programm is planned to improve the statistical significance of presently available results and to narrow remaining uncertainty limits. (orig.) [de

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.

Fission products detection in irradiated TRIGA fuel by means of gamma spectroscopy and MCNP calculation.

Science.gov (United States)

Cagnazzo, M; Borio di Tigliole, A; Böck, H; Villa, M

2018-05-01

Aim of this work was the detection of fission products activity distribution along the axial dimension of irradiated fuel elements (FEs) at the TRIGA Mark II research reactor of the Technische Universität (TU) Wien. The activity distribution was measured by means of a customized fuel gamma scanning device, which includes a vertical lifting system to move the fuel rod along its vertical axis. For each investigated FE, a gamma spectrum measurement was performed along the vertical axis, with steps of 1 cm, in order to determine the axial distribution of the fission products. After the fuel elements underwent a relatively short cooling down period, different fission products were detected. The activity concentration was determined by calibrating the gamma detector with a standard calibration source of known activity and by MCNP6 simulations for the evaluation of self-absorption and geometric effects. Given the specific TRIGA fuel composition, a correction procedure is developed and used in this work for the measurement of the fission product Zr 95 . This measurement campaign is part of a more extended project aiming at the modelling of the TU Wien TRIGA reactor by means of different calculation codes (MCNP6, Serpent): the experimental results presented in this paper will be subsequently used for the benchmark of the models developed with the calculation codes. Copyright © 2018 Elsevier Ltd. All rights reserved.

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

Muon-induced fission

International Nuclear Information System (INIS)