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Sample records for risoe fission gas

  1. IFPE/RISOE-II, Fuel Performance Data from Transient Fission Gas Release

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    1995-01-01

    Description: The RISO National Laboratory in Denmark have carried out three irradiation programs of slow ramp and hold tests, so called 'bump tests' to investigate fission gas release and fuel microstructural changes. The second project took place between 1982 and 1986 and was called 'The RISO Transient Fission Gas Project'. The fuel used in the project was from: IFA-161 irradiated in the Halden BWR (27 to 42 MWd/kgUO 2 ) and GE BWR fuel irradiated in the Millstone 1 reactor 14 to 29 MWd/kgUO 2 . Using the re-fabrication technique, it was possible to back fill the test segment with a choice of gas and gas pressure and to measure the time dependence of fission gas release by continuous monitoring of the plenum pressure. The short length of the test segment was an advantage because, depending on where along the original rod the section was taken, burnup could be chosen variable, and during the test the fuel experienced a single power

  2. Fission gas release and swelling in the fuel pins M1-3 and F9-3: Risoe Fission Gas Project

    Energy Technology Data Exchange (ETDEWEB)

    Walker, C T; Ray, I L.F.; Coquerelle, M; Blank, H

    1982-01-01

    This report presents results for the microscopic swelling local swelling and local gas release in the pin sections M1-3-11 and F9-3-44. The local gas release was derived from the concentration of retained xenon which was measured with the electron microprobe. In addition to xenon, the radial distributions of caesium and neodymium were also determined by EMPA. Caesium is assumed to contribute to microscopic swelling because it results mainly from the decay of /sup 133/Xe, /sup 135/Xe and /sup 137/Xe and, therefore, is trapped together with xenon in bubbles and pores. Neodymium, on the other hand, is soluble in UO/sub 2/ and does not migrate under the influence of the temperature gradients that exist during irradiation. Therefore, the radial distribution of this fission product is an indelible imprint of the burn-up from which the average flux depression can be deduced. 1 ref., 15 figs., 3 tabs.

  3. Fission gas detection system

    International Nuclear Information System (INIS)

    Colburn, R.P.

    1984-01-01

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

  4. Fission gas measuring technology

    International Nuclear Information System (INIS)

    Lee, Hyung Kwon; Kim, Eun Ka; Hwang, Yong Hwa; Lee, Eun Pyo; Chun, Yong Bum; Seo, Ki Seog; Park, Dea Gyu; Chu, Yong Sun; Ahn, Sang Bok.

    1998-02-01

    Safety and economy of nuclear plant are greatly affected by the integrity of nuclear fuels during irradiation reactor core. A series of post-irradiation examination (PIE) including non-destructive and destructive test is to be conducted to evaluate and characterize the nuclear performance. In this report, a principle of the examination equipment to measure and analyse fission gases existing nuclear fuels were described and features of the component and device consisting the fission gas measuring equipment are investigated. (author). 4 refs., 2 tabs., 6 figs

  5. Fission gas measuring technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyung Kwon; Kim, Eun Ka; Hwang, Yong Hwa; Lee, Eun Pyo; Chun, Yong Bum; Seo, Ki Seog; Park, Dea Gyu; Chu, Yong Sun; Ahn, Sang Bok

    1998-02-01

    Safety and economy of nuclear plant are greatly affected by the integrity of nuclear fuels during irradiation reactor core. A series of post-irradiation examination (PIE) including non-destructive and destructive test is to be conducted to evaluate and characterize the nuclear performance. In this report, a principle of the examination equipment to measure and analyse fission gases existing nuclear fuels were described and features of the component and device consisting the fission gas measuring equipment are investigated. (author). 4 refs., 2 tabs., 6 figs.

  6. Fission gas in thoria

    Energy Technology Data Exchange (ETDEWEB)

    Kuganathan, Navaratnarajah, E-mail: n.kuganathan@imperial.ac.uk [Department of Materials, Faculty of Engineering, Imperial College, London, SW7 2AZ (United Kingdom); Ghosh, Partha S. [Material Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Galvin, Conor O.T. [Department of Materials, Faculty of Engineering, Imperial College, London, SW7 2AZ (United Kingdom); Arya, Ashok K. [Material Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Dutta, Bijon K. [Homi Bhabha National Institute, Trombay, Mumbai 400 094 (India); Dey, Gautam K. [Material Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Grimes, Robin W. [Department of Materials, Faculty of Engineering, Imperial College, London, SW7 2AZ (United Kingdom)

    2017-03-15

    The fission gases Xe and Kr, formed during normal reactor operation, are known to degrade fuel performance, particularly at high burn-up. Using first-principles density functional theory together with a dispersion correction (DFT + D), in ThO{sub 2} we calculate the energetics of neutral and charged point defects, the di-vacancy (DV), different neutral tri-vacancies (NTV), the charged tetravacancy (CTV) defect cluster geometries and their interaction with Xe and Kr. The most favourable incorporation point defect site for Xe or Kr in defective ThO{sub 2} is the fully charged thorium vacancy. The lowest energy NTV in larger supercells of ThO{sub 2} is NTV3, however, a single Xe atom is most stable when accommodated within a NTV1. The di-vacancy (DV) is a significantly less favoured incorporation site than the NTV1 but the CTV offers about the same incorporation energy. Incorporation of a second gas atom in a NTV is a high energy process and more unfavourable than accommodation within an existing Th vacancy. The bi-NTV (BNTV) cluster geometry studied will accommodate one or two gas atoms with low incorporation energies but the addition of a third gas atom incurs a high energy penalty. The tri-NTV cluster (TNTV) forms a larger space which accommodates three gas atoms but again there is a penalty to accommodate a fourth gas atom. By considering the energy to form the defect sites, solution energies were generated showing that in ThO{sub 2−x} the most favourable solution equilibrium site is the NTV1 while in ThO{sub 2} it is the DV. - Highlights: • We have considered Xe and Kr in point defects and defect clusters (neutral and charged) using Density Functional Theory (DFT) with a dispersion correction. • The most favourable charge state for a point defect (vacancy or interstitial) is that with full ionic charge and we have found that in all cases gas atoms occupy the fully charged vacancy sites. • The number of fission gas atoms accommodated in ThO{sub 2} is

  7. 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)

  8. 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.

  9. A fission gas release model

    Energy Technology Data Exchange (ETDEWEB)

    Denis, A; Piotrkowski, R [Argentine Atomic Energy Commission, Buenos Aires (Argentina)

    1997-08-01

    The hypothesis contained in the model developed in this work are as follows. The UO{sub 2} is considered as a collection of spherical grains. Nuclear reactions produce fission gases, mainly Xe and Kr, within the grains. Due to the very low solubility of these gases in UO{sub 2}, intragranular bubbles are formed, of a few nanometers is size. The bubbles are assumed to be immobile and to act as traps which capture gas atoms. Free atoms diffuse towards the grain boundaries, where they give origin to intergranular, lenticular bubbles, of the order of microns. The gas atoms in bubbles, either inter or intragranular, can re-enter the matrix through the mechanism of resolution induced by fission fragment impact. The amount of gas stored in intergranular bubbles grows up to a saturation value. Once saturation is reached, intergranular bubbles inter-connect and the gas in excess is released through different channels to the external surface of the fuel. The resolution of intergranular bubbles particularly affects the region of the grain adjacent to the grain boundary. During grain growth, the grain boundary traps the gas atoms, either free or in intragranular bubbles, contained in the swept volume. The grain boundary is considered as a perfect sink, i.e. the gas concentration is zero at that surface of the grain. Due to the spherical symmetry of the problem, the concentration gradient is null at the centre of the grain. The diffusion equation was solved using the implicit finite difference method. The initial solution was analytically obtained by the Laplace transform. The calculations were performed at different constant temperatures and were compared with experimental results. They show the asymptotic growth of the grain radius as a function of burnup, the gas distribution within the grain at every instant, the growth of the gas content at the grain boundary up to the saturation value and the fraction of gas released by the fuel element referred to the total gas generated

  10. Risoe annual report 1979

    International Nuclear Information System (INIS)

    1980-09-01

    Brief notes on several activities of Risoe National Laboratory is given: the effects of radioactive fallout, irradiation of cells, neutron activation analysis, the utilization of nitrogenous fertilizer in cereals, dispersion of hazardous substances through ground water,positron-annihilation, fatigue of metals, neutron-radiography used on fuel elements, analysis of meat consumption and requirement in Den- mark, reliability analysis of oil plat forms and installations for natural gas. Furthermore a summary is presented of current projects at Risoe and a list of selected publications is given. (L.N.)

  11. Fission gas behaviour in water reactor fuels

    International Nuclear Information System (INIS)

    2002-01-01

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

  12. 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)

  13. 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

  14. 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.)

  15. Fission gas retention in irradiated metallic fuel

    International Nuclear Information System (INIS)

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

    1987-01-01

    Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5Fs) are presented. 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 to the measurements shows quantitative agreement with both the magnitude and the axial variation of the retained gas content

  16. Fission gas release behaviour in MOX fuels

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  17. Comparison of the ENIGMA code with experimental data on thermal performance, stable fission gas and iodine release at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Killeen, J C [Nuclear Electric plc, Barnwood (United Kingdom)

    1997-08-01

    The predictions of the ENIGMA code have been compared with data from high burn-up fuel experiments from the Halden and RISO reactors. The experiments modelled were IFA-504 and IFA-558 from Halden and the test II-5 from the RISO power burnup test series. The code has well modelled the fuel thermal performance and has provided a good measure of iodine release from pre-interlinked fuel. After interlinkage the iodine predictions remain a good fit for one experiment, but there is significant overprediction for a second experiment (IFA-558). Stable fission gas release is also well modelled and the predictions are within the expected uncertainly band throughout the burn-up range. This report presents code predictions for stable fission gas release to 40GWd/tU, iodine release measurements to 50GWd/tU and thermal performance (fuel centre temperature) to 55GWd/tU. Fuel ratings of up to 38kW/m were modelled at the high burn-up levels. The code is shown to accurately or conservatively predict all these parameters. (author). 1 ref., 6 figs.

  18. FFTF fission gas monitor computer system

    International Nuclear Information System (INIS)

    Hubbard, J.A.

    1987-01-01

    The Fast Flux Test Facility (FFTF) is a liquid-metal-cooled test reactor located on the Hanford site. A dual computer system has been developed to monitor the reactor cover gas to detect and characterize any fuel or test pin fission gas releases. The system acquires gamma spectra data, identifies isotopes, calculates specific isotope and overall cover gas activity, presents control room alarms and displays, and records and prints data and analysis reports. The fission gas monitor system makes extensive use of commercially available hardware and software, providing a reliable and easily maintained system. The design provides extensive automation of previous manual operations, reducing the need for operator training and minimizing the potential for operator error. The dual nature of the system allows one monitor to be taken out of service for periodic tests or maintenance without interrupting the overall system functions. A built-in calibrated gamma source can be controlled by the computer, allowing the system to provide rapid system self tests and operational performance reports

  19. Gas-phase transport of fission products

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    Knaab, H.; Lang, P.M.

    1985-01-01

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

  2. Antiproton Powered Gas Core Fission Rocket

    International Nuclear Information System (INIS)

    Kammash, Terry

    2005-01-01

    Extensive research in recent years has demonstrated that 'at rest' annihilation of antiprotons in the uranium isotope U238 leads to fission at nearly 100% efficiency. The resulting highly-ionizing, energetic fission fragments can heat a suitable medium to very high temperatures, making such a process particularly suitable for space propulsion applications. Such an ionized medium, which would serve as a propellant, can be confined by a magnetic field during the heating process, and subsequently ejected through a magnetic nozzle to generate thrust. The gasdynamic mirror (GDM) magnetic configuration is especially suited for this application since the underlying confinement principle is that the plasma be of such density and temperature as to make the ion-ion collision mean free path shorter than the plasma length. Under these conditions the plasma behaves like a fluid, and its escape from the system is analogous to the flow of a gas into vacuum from a vessel with a hole. For the system we propose we envisage radially injecting atomic or U238 plasma beam at a pre-determined position and axially pulsing an antiproton beam which upon interaction with the uranium target gives rise to near isotropic ejection of fission fragments with a total mass of 212 amu and total energy of about 160 MeV. These particles, along with the annihilation products (i.e. pions and muons) will heat the background U238 gas - inserted into the chamber just prior to the release of the antiproton - to one keV temperature. Preliminary analysis reveals that such a propulsion system can produce a specific impulse of about 3000 seconds at a thrust of about 50 kN. When applied to a round trip Mars mission, we find that such a journey can be accomplished in about 142 days with 2 days of thrusting and requiring only one gram of antiprotons to achieve it

  3. Risoe annual report 1977/78

    International Nuclear Information System (INIS)

    1978-11-01

    Brief notes on several activities of Risoe National Laboratory is given : reactor safety, uranium in Greenland, radioactive waste, radioactive contamination, windmills, meteorology, new knowledge about selenium, improved barley yield, fusion, behaviour of atoms and molecules, Denmark's energy requirements, gas storage in salt domes. Furthermore a summary is presented of current projects at Risoe, a list of selected publications is given, and design data on research facilities are presented. (BP)

  4. Risoe annual report 1978

    International Nuclear Information System (INIS)

    1979-07-01

    Brief notes on several activities of Risoe National Laboratory is given: battery for the electric car, storage of waste heat, development of fuel elements, reliability, the effect of cooling water temperature in the sea, radioactive chemistry, analysis of different substances in environment and medicine, Risoe library, calibration of thermometers. Furthermore a summary is presented of current projects at Risoe and a list of selected publications is given. (L.N.)

  5. Calculated apparent yields of rare gas fission products

    International Nuclear Information System (INIS)

    Delucchi, A.A.

    1975-01-01

    The apparent fission yield of the rare gas fission products from four mass chains is calculated as a function of separation time for six different fissioning systems. A plot of the calculated fission yield along with a one standard deviation error band is given for each rare gas fission product and for each fissioning system. Those parameters in the calculation that were major contributors to the calculated standard deviation at each separation time were identified and the results presented on a separate plot. To extend the usefulness of these calculations as new and better values for the input parameters become available, a third plot was generated for each system which shows how sensitive the derived fission yield is to a change in any given parameter used in the calculation. (U.S.)

  6. Nondestructive fission gas release measurement and analysis

    International Nuclear Information System (INIS)

    O'Leary, P.M.; Packard, D.R.

    1993-01-01

    Siemens Power Corporation (SPC) has performed reactor poolside gamma scanning measurements of fuel rods for fission gas release (FGR) detection for more than 10 yr. The measurement system has been previously described. Over the years, the data acquisition system, the method of spectrum analysis, and the means of reducing spectrum interference have been significantly improved. A personal computer (PC)-based multichannel analyzer (MCA) package is used to collect, display, and store high-resolution gamma-ray spectra measured in the fuel rod plenum. A PC spread sheet is used to fit the measured spectra and compute sample count rates after Compton background subtraction. A Zircaloy plenum spacer is often used to reduce positron annihilation interference that can arise from the INCONEL reg-sign plenum spring used in SPC-manufactured fuel rods

  7. Risoe annual report 1980

    International Nuclear Information System (INIS)

    1981-09-01

    Brief notes on several activities of Risoe National Laboratory are given: uranium extraction from Kvanefjeld in Greenland, better utilization of uranium as fuel, hydrogen in metals, wind power, lasers, radioactive medicaments, plasma, plant breeding. Furthermore a summary is presented of current projects at Risoe, and a list of selected publications is given. (LN)

  8. Riso na epilepsia

    Directory of Open Access Journals (Sweden)

    Edymar Jardim

    1967-06-01

    Full Text Available São estudados três casos de síndrome convulsiva temporal, com manifestações concomitantes de riso na sua fase inicial. As características principais foram a imotivação e á incoercibilidade do riso. Esses sintomas desapareceram com o uso de anticonvulsivantes.

  9. Risoe annual report 1981

    International Nuclear Information System (INIS)

    1982-08-01

    Brief notes on several activities of Risoe National Laboratory are given: frozen pellets for fusion reactors, reduction of nitrogen fertilizers, surplus heat, energy-economy computer models, environmental chemsitry. Furthermore a summary is presented of current projects at Risoe, and a list of selected publications is given. (LN)

  10. Fission gas release of MOX with heterogeneous structure

    International Nuclear Information System (INIS)

    Nakae, N.; Akiyama, H.; Kamimura, K; Delville, R.; Jutier, F.; Verwerft, M.; Miura, H.; Baba, T.

    2015-01-01

    It is very useful for fuel integrity evaluation to accumulate knowledge base on fuel behavior of uranium and plutonium mixed oxide (MOX) fuel used in light water reactors (LWRs). Fission gas release is one of fuel behaviors which have an impact on fuel integrity evaluation. Fission gas release behavior of MOX fuels having heterogeneous structure is focused in this study. MOX fuel rods with a heterogeneous fuel microstructure were irradiated in Halden reactor (IFA-702) and the BR-3/BR-2 CALLISTO Loop (CHIPS program). The 85 Kr gamma spectrometry measurements were carried out in specific cycles in order to examine the concerned LHR (Linear Heat Rate) for fission gas release in the CHIPS program. The concerned LHR is defined in this paper to be the LHR at which a certain additional fission gas release thermally occurs. Post-irradiation examination was performed to understand the fission gas release behavior in connection with the pellet microstructure. The followings conclusions can be made from this study. First, the concerned LHR for fission gas release is estimated to be in the range of 20-23 kW/m with burnup over 37 GWd/tM. It is moreover guessed that the concerned LHR for fission gas release tends to decrease with increasing burnup. Secondly It is observed that FGR (fission gas release rate) is positively correlated with LHR when the LHR exceeds the concerned value. Thirdly, when burnup dependence of fission gas release is discussed, effective burnup should be taken into account. The effective burnup is defined as the burnup at which the LHR should be exceed the concerned value at the last time during all the irradiation period. And fourthly, it appears that FGR inside Pu spots is higher than outside and that retained (not released) fission gases mainly exist in the fission gas bubbles. Since fission gases in bubbles are considered to be easily released during fuel temperature increase, this information is very important to estimate fission gas release behavior

  11. Evaluation of axial fission gas transport in power ramping experiments

    International Nuclear Information System (INIS)

    Kinoshita, Motoyasu

    1986-01-01

    The LINUS code calculates advective and diffusional transport of fission gas towards an upper plenum through the pellet-cladding gap. The basic equations were modified for analyzing a multi-component gas mixture in the gap and also for dealing with opening and/or closing of the gap, which induces additional axial gas flow. Analysis of the Petten ramp experiment shows that helium pressurization is effective in suppressing an ascending rate of fission gas concentration. After the maximum concentration is achieved through power ramping, the gas concentration could be described by a steady state analytical solution which does not depend on the filling gas pressure. (author)

  12. Calculations of Fission Gas Release During Ramp Tests Using Copernic Code

    Energy Technology Data Exchange (ETDEWEB)

    Tong, Liu [Nuclear Fuel R and D Center, China Nuclear Power Technology Research Institute (CNPRI) (China)

    2013-03-15

    The report performed under IAEA research contract No.15951 describes the results of fuel performance evaluation of LWR fuel rods operated at ramp conditions using the COPERNIC code developed by AREVA. The experimental data from the Third Riso Fission Gas Project and the Studsvik SUPER-RAMP Project presented in the IFPE database of the OECD/NEA has been utilized for assessing the code itself during simulation of fission gas release (FGR). Standard code models for LWR fuel were used in simulations with parameters set properly in accordance with relevant test reports. With the help of data adjustment, the input power histories are restructured to fit the real ones, so as to ensure the validity of FGR prediction. The results obtained by COPERNIC show that different models lead to diverse predictions and discrepancies. By comparison, the COPERNIC V2.2 model (95% Upper bound) is selected as the standard FGR model in this report and the FGR phenomenon is properly simulated by the code. To interpret the large discrepancies of some certain PK rods, the burst effect of FGR which is taken into consideration in COPERNIC is described and the influence of the input power histories is extrapolated. In addition, the real-time tracking capability of COPERNIC is tested against experimental data. In the process of investigation, two main dominant factors influencing the measured gas release rate are described and different mechanisms are analyzed. With the limited predicting capacity, accurate predictions cannot be carried out on abrupt changes of FGR during ramp tests by COPERNIC and improvements may be necessary to some relevant models. (author)

  13. Considerations in modeling fission gas release during normal operation

    International Nuclear Information System (INIS)

    Rumble, E.T.; Lim, E.Y.; Stuart, R.G.

    1977-01-01

    The EPRI LWR fuel rod modeling code evaluation program analyzed seven fuel rods with experimental fission gas release data. In these cases, rod-averged burnups are less than 20,000 MWD/MTM, while the fission gas release fractions range roughly from 2 to 27%. Code results demonstrate the complexities in calculating fission gas release in certain operating regimes. Beyond this work, the behavior of a pre-pressurized PWR rod is simulated to average burnups of 40,000 MWD/MTM using GAPCON-THERMAL-2. Analysis of the sensitivity of fission gas release to power histories and release correlations indicate the strong impact that LMFBR type release correlations induce at high burnup. 15 refs

  14. Reports issued by the Risoe National Laboratory in the series: RISO-R reports and RISO-M reports

    International Nuclear Information System (INIS)

    1982-08-01

    This list includes all scientific and technical reports issued from 1957 - May 1982 by Risoe National Laboratory, former Research Establishment Risoe. The list covers Riso-R and Risoe-M reports, and is arranged according to report numbers. (author)

  15. 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)

  16. Risoe annual report 1987

    International Nuclear Information System (INIS)

    1988-06-01

    An explanation of Risoe National Laboratory's function within the Danish research system is followed by brief accounts of research activities at Risoe during 1987. Energy resources, technology and policy are discussed, the annual accounts are presented, a guide to the National Laboratory and a list of its publications are given. Some of the research activities that took place in 1987 described in more detail are within the fields of chemistry and the environment, superconductivity, new aspects of powdery mildew, polymers and robotics. (AB)

  17. Risoe Publication Activities in 1997; Risoes publikationsvirksomhed i 1997

    Energy Technology Data Exchange (ETDEWEB)

    Alvi, Hanne; Bennov, Solvejg

    1998-08-01

    Risoe`s publication and lecture activities in the last decades are presented through data of total number of publications, distribution of types of publications, number of citations to the international scientific journal articles, and institutions with which Risoe has published the largest number of articles. The data are derived from Risoe`s in-house Publications Database and from the Risoe Institutional Citation Report database produced by the Institute for Scientific Information. The largest part of the report contains a list of references to the scientific and technical journal articles, books, reports, lectures, and to publications for a broader readership authored by researchers at Risoe National Laboratory during the year 1997. The references are organised according to the programme areas of Risoe. (au)

  18. Risoe Research Establishment, Denmark

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1973-07-01

    On the poetic Roskilde Fjord, 40 kilometers from Copenhagen, and near Roskilde, capital of Denmark in the 12th century, stands the Risoe Research Establishment of the Danish Atomic Energy Commission. ere 700 men and women are engaged in searching for ways in which atomic energy can be used to make the world a better and healthier place. The work at Risoe comprises fundamental research, reactor technology and other technological studies, agricultural research and health and safety studies. Nuclear power stations are scheduled to be operative in Denmark some time between 1975 and 1980, and the planning of these stations and development of the many processes this will involve has become a major task at Risoe. Special conditions have to be fulfilled in selecting the site of an atomic research station, and the barren Risoe peninsula had them all: safety, because the site was free from buildings to permit continuous control; closeness to the scientific institutions of the capital, Copenhagen; social amenities in Roskilde; finally, access to an a adequate water supply. his special series of photos covering some aspects of the work and safety conditions at Risoe was commissioned by WHO. (author)

  19. Risoe Research Establishment, Denmark

    International Nuclear Information System (INIS)

    1973-01-01

    On the poetic Roskilde Fjord, 40 kilometers from Copenhagen, and near Roskilde, capital of Denmark in the 12th century, stands the Risoe Research Establishment of the Danish Atomic Energy Commission. ere 700 men and women are engaged in searching for ways in which atomic energy can be used to make the world a better and healthier place. The work at Risoe comprises fundamental research, reactor technology and other technological studies, agricultural research and health and safety studies. Nuclear power stations are scheduled to be operative in Denmark some time between 1975 and 1980, and the planning of these stations and development of the many processes this will involve has become a major task at Risoe. Special conditions have to be fulfilled in selecting the site of an atomic research station, and the barren Risoe peninsula had them all: safety, because the site was free from buildings to permit continuous control; closeness to the scientific institutions of the capital, Copenhagen; social amenities in Roskilde; finally, access to an a adequate water supply. his special series of photos covering some aspects of the work and safety conditions at Risoe was commissioned by WHO. (author)

  20. Determination of fission gas yields from isotope ratios

    DEFF Research Database (Denmark)

    Mogensen, Mogens Bjerg

    1983-01-01

    This paper describes a method of calculating the actual fission yield of Kr and Xe in nuclear fuel including the effect of neutron capture reactions and decay. The bases for this calculation are the cumulative yields (ref. 1) of Kr and Xe isotopes (or pairs of isotopes) which are unaffected...... by neutron capture reactions, and measured Kr and Xe isotope ratios. Also the burnup contribution from the different fissile heavy isotopes must be known in order to get accurate fission gas yields....

  1. Fission-gas bubble modeling for LMFBR accidents

    International Nuclear Information System (INIS)

    Ostensen, R.W.

    1977-01-01

    The behavior of fission-gas bubbles in unrestructured oxide fuel can have a dominant effect on the course of a core disruptive accident in an LMFBR. The paper describes a simplified model of bubble behavior and presents results of that model in analyzing the relevant physical assumptions and predicting gas behavior in molten fuel

  2. 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

  3. The effects of fission gas release on PWR fuel rod design and performance

    International Nuclear Information System (INIS)

    Leech, W.J.; Kaiser, R.S.

    1980-01-01

    The purpose of this investigation was to determine the effects of fission gas release on PWR fuel rod design and performance. Empirical models were developed from fission gas release data. Fission gas release during normal operation is a function of burnup. There is little additional fission gas release during anticipated transients. The empirical models were used to evaluate Westinghouse fuel rod designs. It was determined that fission gas release is not a limiting parameter for obtaining rod average burnups in the range of 50,000 to 60,000 MWD/MTU. Fission gas release during anticipated transients has a negligible effect on the margins to rod design limits. (author)

  4. Risoe publication activities in 1998

    Energy Technology Data Exchange (ETDEWEB)

    Bennov, Solvejg [ed.

    1999-04-01

    The report contains a list of references to the scientific and technical journal articles, books, reports, lectures published in full text, and to publications for a broader readership authored by researchers at Risoe National Laboratory and published in 1998. If the publication mentioned in the reference is electronically available the link to the web-address is added. The references are organised according to the programme areas of Risoe. The text is introduced by total number of publications, distribution of types of publications, number of citations to the international scientific journal articles, institutions with which Risoe has published the largest number of articles, and journals in which Risoe has published most articles. The data are derived from Risoe`s in-house Publications Database and from the Risoe Institutional Citation Report database produced by the Institute for Scientific Information. (au)

  5. Axial transport of fission gas in LWR fuel rods

    International Nuclear Information System (INIS)

    Kinoshita, M.

    1983-01-01

    With regard to fission gas transportation inside the fuel rod, the following three mechanisms are important: (1) a localized and time dependent fission gas release from UO 2 fuel to pellet/clad gap, (2) the consequent gas pressure difference between the gap and the plenum, and (3) the inter-diffusion of initially filled Helium and released fission gas such as Xenon. Among these three mechanisms, the 2nd mechanism would result in the one dimensional flow through P/C gap in the axial direction, while the 3rd would average the local fission gas concentration difference. In this paper, an attempt was made to develop a computerized model, LINUS (LINear flow and diffusion under Un-Steady condition) describing the above two mechanisms, items (2) and (3). The item (1) is treated as an input. The code was applied to analyse short length experimental fuel rods and long length commercial fuel rods. The calculated time evolution of Xe concentration along the fuel column shows that the dilution rate of Xe in commercial fuel rods is much slower than that in short experimental fuel rods. Some other sensitivity studies, such as the effect of pre-pressurization, are also presented. (author)

  6. Fuel performance and fission product behaviour in gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

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

  7. Fuel performance and fission product behaviour in gas cooled reactors

    International Nuclear Information System (INIS)

    1997-11-01

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

  8. Fission gas behavior during fast thermal transients

    International Nuclear Information System (INIS)

    Esteves, R.G.

    1976-01-01

    The behavior of non-equilibrium fission in fuel elements undergoing fast thermal transients is analyzed. To facilitate the analysis, a new variable, the equilibrium variable (EV) is defined. This variable, together with bubble radius, completely specifies a bubble with respect to its size and equilibrium condition. The analysis is coded using a two-variable (radius and EV) multigroup numerical approximation that accepts as input the time-temperature history, the time-fission rate history, and the time-thermal gradient history of the fuel element. Studies were performed to test the code for convergence with respect to the time interval and the number of groups chosen. For a series of transient simulation studies, the measurements obtained at HEDL (microscopic examination of intragranular porosity in oxide fuel transient-tested in TREAT) are used. Two different transient histories were selected; the first, a high-temperature transient (HTT) with a peak at 2477 0 K and the second, a low-temperature transient (LTT) with a peak-temperature at 2000 0 K. The LTT was simulated for three different conditions: Bubbles were allowed to move via (a) only biased migration, (b) via random migration, and (c) via both mechanisms. The HTT was also run for both mechanisms. The agreement with HEDL microscopic observations was fair for bubbles smaller than 964 A in diameter, and poor for larger bubbles. Bubbles that grew during the heat-up part of the transient were frozen at a larger size during the cool down

  9. SPEAR-BETA fuel-performance code system: fission-gas-release module. Final report

    International Nuclear Information System (INIS)

    Christensen, R.

    1983-03-01

    The original SPEAR-BETA general description manual covers both mechanistic and statistical models for fuel reliability, but only mechanistic modeling of fission gas release. This addendum covers the SPEAR-BETA statistical model for fission gas release

  10. Fission gas retention and axial expansion of irradiated metallic fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.; Johanson, E.W.

    1986-05-01

    Out-of-reactor experiments utilizing direct electrical heating and infrared heating techniques were performed on irradiated metallic fuel. The results indicate accelerated expansion can occur during thermal transients and that the accelerated expansion is driven by retained fission gases. The results also demonstrate gas retention and, hence, expansion behavior is a function of axial position within the pin

  11. Role of fission gas release in reactor licensing

    International Nuclear Information System (INIS)

    1975-11-01

    The release of fission gases from oxide pellets to the fuel rod internal voidage (gap) is reviewed with regard to the required safety analysis in reactor licensing. Significant analyzed effects are described, prominent gas release models are reviewed, and various methods used in the licensing process are summarized. The report thus serves as a guide to a large body of literature including company reports and government documents. A discussion of the state of the art of gas release analysis is presented

  12. Fission-gas-bubble mobility in oxide fuel: a critical analysis

    International Nuclear Information System (INIS)

    Tam, S.W.; Johnson, C.E.

    1983-08-01

    The available volatile fission product release data has confirmed the general viability of the scaling model of volatile fission product release in which the fractional release rates of the volatile fission products scale as that of the fission gas. The question of whether fission gas bubbles can move sufficiently fast to be a significant mechanism responsible for fission gas release from the fuel is considered. The mean jump distance per jump of the hopping process in gas bubble motion is analyzed. Surface roughness is also considered

  13. Risoe publication activities in 1998

    International Nuclear Information System (INIS)

    Bennov, Solvejg

    1999-04-01

    The report contains a list of references to the scientific and technical journal articles, books, reports, lectures published in full text, and to publications for a broader readership authored by researchers at Risoe National Laboratory and published in 1998. If the publication mentioned in the reference is electronically available the link to the web-address is added. The references are organised according to the programme areas of Risoe. The text is introduced by total number of publications, distribution of types of publications, number of citations to the international scientific journal articles, institutions with which Risoe has published the largest number of articles, and journals in which Risoe has published most articles. The data are derived from Risoe's in-house Publications Database and from the Risoe Institutional Citation Report database produced by the Institute for Scientific Information. (au)

  14. Risoe Publication Activities in 1997

    International Nuclear Information System (INIS)

    Alvi, Hanne; Bennov, Solvejg

    1998-08-01

    Risoe's publication and lecture activities in the last decades are presented through data of total number of publications, distribution of types of publications, number of citations to the international scientific journal articles, and institutions with which Risoe has published the largest number of articles. The data are derived from Risoe's in-house Publications Database and from the Risoe Institutional Citation Report database produced by the Institute for Scientific Information. The largest part of the report contains a list of references to the scientific and technical journal articles, books, reports, lectures, and to publications for a broader readership authored by researchers at Risoe National Laboratory during the year 1997. The references are organised according to the programme areas of Risoe. (au)

  15. Risoe 1976/77

    International Nuclear Information System (INIS)

    1977-01-01

    A summary of the chief activities of the Risoe National Laboratory is given. The material is presented in a revised sequence following upon changes in the organization of the work at Risoe. These changes partly reflect the widened scope of activities, which now include non-nuclear energy research, that resulted from the Act on Energy Policy Measures from April 1976. The sequence is: reactor technology and safety, nuclear fuel cycle, environmental and safety research, materials research, radiation technology, agricultural research, other energy research, research facilities and auxiliary services. For more detailed descriptions of the work in progress, readers are referred to the annual reports of the various departments, and to the two series of research reports as well as to articles appearing in scientific journals. A list of these publications is given. Design data on research facilities are presented. (BP)

  16. Fuel rod puncturing and fission gas monitoring system examination techniques

    International Nuclear Information System (INIS)

    Song, Woong Sup

    1999-02-01

    Fission gas products accumulated in irradiated fuel rod is 1-2 cm 3 in CANDU and 40-50 cm 3 in PWR fuel rod. Fuel rod puncturing and fission gas monitoring system can be used for both CANDU and PWR fuel rod. This system comprises puncturing device located at in cell part and monitoring device located at out cell part. The system has computerized 9 modes and can calculate both void volume and mass volume only single puncturing. This report describes techniques and procedure for operating fuel rod puncturing and gas monitoring system which can be play an important role in successful operation of the devices. Results obtained from the analysis can give more influence over design for fuel rods. (Author). 6 refs., 9 figs

  17. Transient fission gas release during direct electrical heating experiments

    International Nuclear Information System (INIS)

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.

    1983-12-01

    The gas release behavior of irradiated EBR-II fuel was observed to be dependent on several factors: the presence of cladding, the retained gas content, and the energy absorbed. Fuel that retained in excess of 16 to 17 μmoles/g of fission gas underwent spallation as the cladding melted and released 22 to 45% of its retained gas, while fuel with retained gas levels below approx. 15 to 16 μmoles/g released less than approx. 9% of its gas as the cladding melted. During subsequent direct electrical heating ramps, fuel that did not spall released an additional quantity of gas (up to 4 μmoles/g), depending on the energy absorbed

  18. Reactions of newly formed fission products in the gas phase

    International Nuclear Information System (INIS)

    Strickert, R.G.

    1976-01-01

    A dynamic gas-flow system was constructed which stopped fission products in the gas phase and rapidly separated (in less than 2 sec) volatile compounds from non-volatile ones. The filter assembly designed and used was shown to stop essentially all non-volatile fission products. Between 5 percent and 20 percent of tellurium fission-product isotopes reacted with several hydrocarbon gases to form volatile compounds, which passed through the filter. With carbon monoxide gas, volatile tellurium compound(s) (probably TeCO) were also formed with similar efficiencies. The upper limits for the yields of volatile compounds formed between CO and tin and antimony fission products were shown to be less than 0.3 percent, so tellurium nuclides, not their precursors, reacted with CO. It was found that CO reacted preferentially with independently produced tellurium atoms; the reaction efficiency of beta-produced atoms was only 27 +- 3 percent of that of the independently formed atoms. The selectivity, which was independent of the over-all reaction efficiency, was shown to be due to reaction of independently formed atoms in the gas phase. The gas phase reactions are believed to occur mainly at thermal energies because of the independence of the yield upon argon moderator mole-fraction (up to 80 percent). It was shown in some experiments that about one-half of the TeCO decomposed in passing through a filter and that an appreciable fraction (approximately 20 percent) of the tellurium atoms deposited on the filter reacted agin with CO. Other tellurium atoms on the filter surface (those formed by beta decay and those formed independently but not reacting in the gas phase) also reacted with CO, but probably somewhat less efficiently than atoms formed by TeCO decomposition. No evidence was found for formation of TeCO as a direct result of beta-decay

  19. A new mechanistic and engineering fission gas release model for a uranium dioxide fuel

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Yang, Yong Sik; Kim, Dae Ho; Kim, Sun Ki; Bang, Je Geun

    2008-01-01

    A mechanistic and engineering fission gas release model (MEGA) for uranium dioxide (UO 2 ) fuel was developed. It was based upon the diffusional release of fission gases from inside the grain to the grain boundary and the release of fission gases from the grain boundary to the external surface by the interconnection of the fission gas bubbles in the grain boundary. The capability of the MEGA model was validated by a comparison with the fission gas release data base and the sensitivity analyses of the parameters. It was found that the MEGA model correctly predicts the fission gas release in the broad range of fuel burnups up to 98 MWd/kgU. Especially, the enhancement of fission gas release in a high-burnup fuel, and the reduction of fission gas release at a high burnup by increasing the UO 2 grain size were found to be correctly predicted by the MEGA model without using any artificial factor. (author)

  20. 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).

  1. Fission gas release from fuel at high burnup

    International Nuclear Information System (INIS)

    Meyer, R.O.; Beyer, C.E.; Voglewede, J.C.

    1978-03-01

    The release of fission gas from fuel pellets at high burnup is reviewed in the context of the safety analysis performed for reactor license applications. Licensing actions are described that were taken to correct deficient gas release models used in these safety analyses. A correction function, which was developed by the Nuclear Regulatory Commission staff and its consultants, is presented. Related information, which includes some previously unpublished data, is also summarized. The report thus provides guidance for the analysis of high burnup gas release in licensing situations

  2. Engineering scale tests of an FFTF fission gas delay bed

    International Nuclear Information System (INIS)

    Kabele, T.J.; Bohringer, A.P.

    1975-01-01

    The dynamic adsorption coefficient of 85 Kr on activated charcoal from a nitrogen carrier gas was measured at -80 and -120 0 C at pressures of zero and 30 psig. The effects of the presence of impurities in the nitrogen carrier gas (1 percent oxygen, and 100 vppm carbon dioxide) on the adsorption coefficient of 85 Kr were also measured. The 85 Kr adsorption coefficient increased with decreasing temperature, and increased with increasing pressure. The presence of oxygen and carbon dioxide impurities in the nitrogen carrier gas had no discernible effect upon the adsorption coefficient. The adsorption coefficient for 85 Kr from nitrogen gas was lower than for adsorption of 85 Kr from an argon gas stream. The work concluded a test program which provided design data for the fission gas delay beds which will be installed in the Fast Flux Test Facility (FFTF). (U.S.)

  3. 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)

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  5. Steady-state and transient fission gas release and swelling model for LIFE-4

    International Nuclear Information System (INIS)

    Villalobos, A.; Liu, Y.Y.; Rest, J.

    1984-06-01

    The fuel-pin modeling code LIFE-4 and the mechanistic fission gas behavior model FASTGRASS have been coupled and verified against gas release data from mixed-oxide fuels which were transient tested in the TREAT reactor. Design of the interface between LIFE-4 and FASTGRASS is based on an earlier coupling between an LWR version of LIFE and the GRASS-SST code. Fission gas behavior can significantly affect steady-state and transient fuel performance. FASTGRASS treats fission gas release and swelling in an internally consistent manner and simultaneously includes all major mechanisms thought to influence fission gas behavior. The FASTGRASS steady-state and transient analysis has evolved through comparisons of code predictions with fission-gas release and swelling data from both in- and ex-reactor experiments. FASTGRASS was chosen over other fission-gas behavior models because of its availability, its compatibility with the LIFE-4 calculational framework, and its predictive capability

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

    International Nuclear Information System (INIS)

    Zimmermann, H.

    1980-01-01

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

  7. Fission gas release and pellet microstructure change of high burnup BWR fuel

    International Nuclear Information System (INIS)

    Itagaki, N.; Ohira, K.; Tsuda, K.; Fischer, G.; Ota, T.

    1998-01-01

    UO 2 fuel, with and without Gadolinium, irradiated for three, five, and six irradiation cycles up to about 60 GWd/t pellet burnup in a commercial BWR were studied. The fission gas release and the rim effect were investigated by the puncture test and gas analysis method, OM (optical microscope), SEM (scanning electron microscope), and EPMA (electron probe microanalyzer). The fission gas release rate of the fuel rods irradiated up to six cycles was below a few percent; there was no tendency for the fission gas release to increase abruptly with burnup. On the other hand, microstructure changes were revealed by OM and SEM examination at the rim position with burnup increase. Fission gas was found depleted at both the rim position and the pellet center region using EPMA. There was no correlation between the fission gas release measured by the puncture test and the fission gas depletion at the rim position using EPMA. However, the depletion of fission gas in the center region had good correlation with the fission gas release rate determined by the puncture test. In addition, because the burnup is very large at the rim position of high burnup fuel and also due to the fission rate of the produced Pu, the Xe/Kr ratio at the rim position of high burnup fuel is close to the value of the fission yield of Pu. The Xe/Kr ratio determined by the gas analysis after the puncture test was equivalent to the fuel average but not to the pellet rim position. From the results, it was concluded that fission gas at the rim position was released from the UO 2 matrix in high burnup, however, most of this released fission gas was held in the porous structure and not released from the pellet to the free volume. (author)

  8. 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)

  9. Multiscale development of a fission gas thermal conductivity model: Coupling atomic, meso and continuum level simulations

    International Nuclear Information System (INIS)

    Tonks, Michael R.; Millett, Paul C.; Nerikar, Pankaj; Du, Shiyu; Andersson, David; Stanek, Christopher R.; Gaston, Derek; Andrs, David; Williamson, Richard

    2013-01-01

    Fission gas production and evolution significantly impact the fuel performance, causing swelling, a reduction in the thermal conductivity and fission gas release. However, typical empirical models of fuel properties treat each of these effects separately and uncoupled. Here, we couple a fission gas release model to a model of the impact of fission gas on the fuel thermal conductivity. To quantify the specific impact of grain boundary (GB) bubbles on the thermal conductivity, we use atomistic and mesoscale simulations. Atomistic molecular dynamic simulations were employed to determine the GB thermal resistance. These values were then used in mesoscale heat conduction simulations to develop a mechanistic expression for the effective GB thermal resistance of a GB containing gas bubbles, as a function of the percentage of the GB covered by fission gas. The coupled fission gas release and thermal conductivity model was implemented in Idaho National Laboratory’s BISON fuel performance code to model the behavior of a 10-pellet LWR fuel rodlet, showing how the fission gas impacts the UO 2 thermal conductivity. Furthermore, additional BISON simulations were conducted to demonstrate the impact of average grain size on both the fuel thermal conductivity and the fission gas release

  10. Technical measurement of small fission gas inventory in fuel rod with laser puncturing system

    International Nuclear Information System (INIS)

    Kim, Hee Moon; Kim, Sung Ryul; Lee, Byoung Oon; Yang, Yong Sik; Baek, Sang Ryul; Song, Ung Sup

    2012-01-01

    The fission gas release cause degradation of fuel rod. It influences fuel temperature and internal pressure due to low thermal conductivity. Therefore, fission gas released to internal void of fuel rod must be measured with burnup. To measure amount of fission gas, fuel rod must be punctured by a steel needle in a closed chamber. Ideal gas law(PV=nRT) is applied to obtain atomic concentration(mole). Steel needle type is good for large amount of fission gas such as commercial spent fuel rod. But, some cases with small fuel rig in research reactor for R/D program are not available to use needle type because of large chamber volume. The laser puncturing technique was developed to solve measurement of small amount of fission gas. This system was very rare equipment in other countries. Fine pressure gage and strong vacuum system were installed, and the chamber volume was reduced at least. Fiber laser was used for easy operation

  11. 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

  12. Fission gas behavior in mixed-oxide fuel during transient overpower

    International Nuclear Information System (INIS)

    Randklev, E.H.; Treibs, H.A.; Mastel, B.; Baldwin, D.L.

    1979-01-01

    Fission gas behavior can be important in determining fuel pin and core performance during a reactor transient. The results are presented of examinations characterizing the changes in microstructural distribution and retention of fission gas in fuel for a series of transient overpower (50 cents/s) tested mixed-oxide fuel pins and their steady state siblings

  13. 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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  15. Transient redistribution of intragranular fission gas in irradiated mixed oxide

    International Nuclear Information System (INIS)

    Hinman, C.A.; Randklev, E.H.

    1981-01-01

    Safety analyses for an LMFBR require a knowledge of the fuel and fission gas behavior under transient conditions. Analyses of microstructural data derived from transiently heated, irradiated, mixed oxide fuel specimens have allowed the calculation of the degree of nonequilibrium of intragranular bubbles formed during the transient. It is hypothesized that the observed over-pressurization of the intragranular bubbles mechanically loads the fuel within the grain, leading to a stress gradient derived force upon near-grain-surface bubbles, driving them preferentially to the grain boundaries. Using existing models for forced diffusion it can be estimated that the stress derived forces on bubbles are within the same magnitude, and possibly greater, than the forces derived from the thermal gradient

  16. Innovation of fission gas release and thermal conductivity measurement methods

    International Nuclear Information System (INIS)

    Van der Meer, K.; Soboler, V.

    1998-01-01

    This presentation described two innovative measurement methods being currently developed at SCK-CEN in order to support the modeling of fuel performance. The first one is an acoustic method to measure the fission gas release in a fuel rod in a non destructive way. The total rod pressure is determined by generating a heat pulse causing a pressure wave that propagates through the gas to an ultrasound transducer. The final pulse width being proportional to the pressure, the latter can thus be determined. The measurement of the acoustic resonance frequency at fixed temperatures enables the distinction between different gas components. The second method is a non-stationary technique to investigate the thermal properties of the fuel rod, like thermal conductivity, diffusivity and heat capacity. These properties are derived from the amplitude and the phase shift of the fuel centre temperature response induced by a periodic temperature variation. These methods did not reveal any physical limitations for the practical applicability. Furthermore, they are rather simple. Preliminary investigations have proven both methods to be more accurate than techniques usually utilized. (author)

  17. Design of pellet surface grooves for fission gas plenum

    International Nuclear Information System (INIS)

    Carter, T.J.; Jones, L.R.; Macici, N.; Miller, G.C.

    1986-01-01

    In the Canada deuterium uranium pressurized heavy water reactor, short (50-cm) Zircaloy-4 clad bundles are fueled on-power. Although internal void volume within the fuel rods is adequate for the present once-through natural uranium cycle, the authors have investigated methods for increasing the internal gas storage volume needed in high-power, high-burnup, experimental ceramic fuels. This present work sought to prove the methodology for design of gas storage volume within the fuel pellets - specifically the use of grooves pressed or machined into the relatively cool pellet/cladding interface. Preanalysis and design of pellet groove shape and volume was accomplished using the TRUMP heat transfer code. Postirradiation examination (PIE) was used to check the initial design and heat transfer assumptions. Fission gas release was found to be higher for the grooved pellet rods than for the comparison rods with hollow or unmodified pellets. This had been expected from the initial TRUMP thermal analyses. The ELESIM fuel modeling code was used to check in-reactor performance, but some modifications were necessary to accommodate the loss of heat transfer surface to the grooves. It was concluded that for plenum design purposes, circumferential pellet grooves could be adequately modeled by the codes TRUMP and ELESIM

  18. 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)

  19. 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.

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

    Science.gov (United States)

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

    2018-06-01

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

  1. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

    Energy Technology Data Exchange (ETDEWEB)

    Pastore, Giovanni, E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Swiler, L.P., E-mail: LPSwile@sandia.gov [Optimization and Uncertainty Quantification, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185-1318 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Spencer, B.W., E-mail: Benjamin.Spencer@inl.gov [Fuel Modeling and Simulation, 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); Van Uffelen, P., E-mail: Paul.Van-Uffelen@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz-Platz 1, D-76344 Karlsruhe (Germany); Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States)

    2015-01-15

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code with a recently implemented physics-based model for fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO{sub 2} single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information in the open literature. The study leads to an initial quantitative assessment of the uncertainty in fission gas behavior predictions with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, significantly higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.

  2. 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

  3. Fission gas release and grain growth in THO2-UO2 fuel irradiated at high temperature

    International Nuclear Information System (INIS)

    Goldberg, I.; Waldman, L.A.; Giovengo, J.F.; Campbell, W.R.

    1979-01-01

    Data are presented on fission gas release and grain growth in ThO 2 -UO 2 fuels irradiated as part of the LWBR fuel element development program. These data for rods that experienced peak linear power outputs ranging from 15 to 22 KW/ft supplement fission gas release data previously reported for 51 rods containing ThO 2 and ThO 2 -UO 2 fuel irradiated at peak linear powers predominantly below 14 KW/ft. Fission gas release was relatively high (up to 15.0 percent) for the rods operated at high power in contrast to the relatively low fission gas release (0.1 to 5.2 percent) measured for the rods operated at lower power. Metallographic examination revealed extensive equiaxed grain growth in the fuel at the high power axial locations of the three rods

  4. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    International Nuclear Information System (INIS)

    1985-01-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity

  5. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity.

  6. 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

  7. Pellet refueling program at Risoe

    International Nuclear Information System (INIS)

    Andersen, V.; Chang, C.T.; Joergensen, L.W.; Nielsen, P.; Sillesen, A.H.

    1978-01-01

    The pellet refueling work at Riso has up to now been concentrated at studying the ablation rate of hydrogen pellets in hydrogen and deuterium plasmas in the Puffatron device. The main results of these studies are well known and we shall only give a brief summary including some more recent results relating to the ablation process. The work on the Puffatron device has been completed and we are presently preparing to start ablation studies in a small Tokamak, Dante. This tokamak has only been constructed this summer and ablation studies are expected to begin in the beginning of 1978. We shall give the expected parameters of the tokamak plasma and indicate some of the planned work. In this presentation we shall also report on the theoretical work on refueling taking place at Riso. We have particularly been interested in the effect of α-particles which could significantly alter the conclusions made from present experiments

  8. Automatic counting of fission fragments tracks using the gas permeation technique

    CERN Document Server

    Yamazaki, I M

    1999-01-01

    An automatic counting system for fission tracks induced in a polycarbonate plastic Makrofol KG (10 mu m thickness) is described. The method is based on the gas transport mechanism proposed by Knudsen, where the gas permeability for a porous membrane is expected to be directly related to its track density. In this work, nitrogen permeabilities for several Makrofol films, with different fission track densities, have been measured using an adequate gas permeation system. The fission tracks were produced by irradiating Makrofol foils with a 252Cf calibrated source in a 2 pi geometry. A calibration curve fission track number versus nitrogen permeability has been obtained, for track densities higher than 1000/cm sup 2 , where the spark gap technique and the visual methods employing a microscope, are not appropriate for track counting.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-06-01

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

  10. Chemical reactions of fission products with ethylene using the gas jet technique

    International Nuclear Information System (INIS)

    Contis, E.T.; Rengan, Krish; Griffin, Henry C.

    1994-01-01

    An understanding of the nature of the chemical reactions taking place between fission products and their carrier gases, and the designing of a fast separation procedure were the purposes of this investigation. Chemical reactions of short-lived (less than one minute half-life) fission products with carrier gases lead to various chemical species which can be separated in the gas phase. The Gas Jet Facility at the Ford Nuclear Reactor was used to study the yields of volatile selenium and bromine fission products of 235 U using a semi-automatic batch solvent extraction technique. Heptane and water were used as organic and inorganic solvents. A carrier gas mixture of ethylene to pre-purified nitrogen (1 : 3) was used to sweep the fission products from the target to the chemistry area for analysis. The results indicated that the volatile selenium products generated by the interaction of selenium fission fragments with ethylene were predominantly organic in nature (84%), possibly organoselenides. The selenium values were used to resolve the fractions of the bromine nuclides, which come from two major sources, viz., directly from fission and from the beta-decay of selenium. The data showed that the fractions of independent bromine fission products in the organic phase were much lower compared to selenium; the bromine values range from 10 to 22% and varied with mass number. Results indicated that the bromine products were inorganic in nature, as possibly hydrogen chloride. ((orig.))

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

  13. The automated Risoe TL dating reader system

    International Nuclear Information System (INIS)

    Boetter-Jensen, L.

    1988-01-01

    The features of the new modified Riso TL dating reader system are described. A vacuum chamber that accommodates the entire 24-position sample changer unit has been designed. The vacuum and N 2 -gas functions are software-controlled. A newly designed heater system is capable of repeated heating cycles to 700 0 C. The sample changer system accommodates fine-grain discs as well as planchettes for coarse grains. Two software-controlled beta irradiators can be attached to the reader, e.g. for predose measurement. The software allows a user without programming expertise to create any desired measuring sequence, and to store and recall data and glow curves for making analyses. (author)

  14. An improved model of fission gas atom transport in irradiated uranium dioxide

    Science.gov (United States)

    Shea, J. H.

    2018-04-01

    The hitherto standard approach to predicting fission gas release has been a pure diffusion gas atom transport model based upon Fick's law. An additional mechanism has subsequently been identified from experimental data at high burnup and has been summarised in an empirical model that is considered to embody a so-called fuel matrix 'saturation' phenomenon whereby the fuel matrix has become saturated with fission gas so that the continued addition of extra fission gas atoms results in their expulsion from the fuel matrix into the fuel rod plenum. The present paper proposes a different approach by constructing an enhanced fission gas transport law consisting of two components: 1) Fick's law and 2) a so-called drift term. The new transport law can be shown to be effectively identical in its predictions to the 'saturation' approach and is more readily physically justifiable. The method introduces a generalisation of the standard diffusion equation which is dubbed the Drift Diffusion Equation. According to the magnitude of a dimensionless Péclet number, P, the new equation can vary from pure diffusion to pure drift, which latter represents a collective motion of the fission gas atoms through the fuel matrix at a translational velocity. Comparison is made between the saturation and enhanced transport approaches. Because of its dependence on P, the Drift Diffusion Equation is shown to be more effective at managing the transition from one type of limiting transport phenomenon to the other. Thus it can adapt appropriately according to the reactor operation.

  15. The migration of intra-granular fission gas bubbles in irradiated uranium dioxide

    International Nuclear Information System (INIS)

    Baker, C.

    1977-05-01

    The mobility of intragranular fission gas bubbles in uranium dioxide irradiated at 1600-1800 0 C has been studied following isothermal annealing at temperatures below 1600 0 C. The intragranular fission gas bubbles, average diameter approximately 2nm, are virtually immobile at temperatures below 1500 0 C. The bubbles have clean surfaces with no solid fission product contamination and are faceted to the highest observed irradiation temperature of 1800 0 C. This bubble faceting is believed to be a major cause of bubble immobility. In fuel operating below 1500 0 C the predominant mechanism allowing the growth of intragranular bubbles and the subsequent gas release must be the diffusion of dissolved gas atoms rather than the movement of entire intragranular bubbles. (author)

  16. Process for separation of inert fission gases for waste gas of a reprocessing plant for nuclear fuel

    International Nuclear Information System (INIS)

    Schnez, H.

    1980-01-01

    The inert fission gases Kr and Xe released in the resolver and other waste gases are taken to an acid regeneration plant. Part of the inert fission gases is separated by compression, cooling and filtering and deposited. The other part flows back to the resolver as flushing gas so that a flushing gas circuit is formed, which prevents explosive gas mixtures occurring. (DG) [de

  17. Modeling fission gas release in high burnup ThO2-UO2 fuel

    International Nuclear Information System (INIS)

    Long, Y.; Yuan, Y.; Pilat, E.E.; Rim, C.S.; Kazimi, M.S.

    2001-01-01

    A preliminary fission gas release model to predict the performance of thoria fuel using the FRAPCON-3 computer code package has been formulated. The following modeling changes have been made in the code: - Radial power/burnup distribution; - Thermal conductivity and thermal expansion; - Rim porosity and fuel density; - Diffusion coefficient of fission gas in ThO 2 -UO 2 fuel and low temperature fission gas release model. Due to its lower epithermal resonance absorption, thoria fuel experiences a much flatter distribution of radial fissile products and radial power distribution during operation as compared to uranian fuel. The rim effect and its consequences in thoria fuel, therefore, are expected to occur only at relatively high burnup levels. The enhanced conductivity is evident for ThO 2 , but for a mixture the thermal conductivity enhancement is small. The lower thermal fuel expansion tends to negate these small advantages. With the modifications above, the new version of FRAPCON-3 matched the measured fission gas release data reasonably well using the ANS 5.4 fission gas release model. (authors)

  18. A review of selected aspects of the effect of water vapor on fission gas release from uranium oxycarbide

    International Nuclear Information System (INIS)

    Myers, B.F.

    1994-04-01

    A selective review is presented of previous measurements and the analysis of experiments on the effect of water vapor on fission gas release from uranium oxycarbide. Evidence for the time-dependent composition of the uranium oxycarbide fuel; the diffusional release of fission gas; and the initial, rapid and limited release of stored fission gas is discussed. In regard to the initial, rapid release of fission gas, clear restrictions on mechanistic hypotheses can be deduced from the experimental data. However, more fundamental experiments may be required to establish the mechanism of the rapid release

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

    International Nuclear Information System (INIS)

    Szuta, M.

    2001-01-01

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

  20. 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)

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

    International Nuclear Information System (INIS)

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

    2006-09-01

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

  2. LOFC fission product release and circulating activity calculations for gas-cooled reactors

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.; Carruthers, L.M.; Lee, C.E.

    1977-01-01

    The inventories of fission products in a gas-cooled reactor under accident and normal steady state conditions are time and temperature dependent. To obtain a reasonable estimate of these inventories it is necessary to consider fuel failure, a temperature dependent variable, and radioactive decay, a time dependent variable. Using arbitrary radioactive decay chains and published fuel failure models for the High Temperature Gas-Cooled Reactor (HTGR), methods have been developed to evaluate the release of fission products during the Loss of Forced Circulation (LOFC) accident and the circulating and plateout fission product inventories during steady state non-accident operation. The LARC-2 model presented here neglects the time delays in the release from the HTGR due to diffusion of fission products from particles in the fuel rod through the graphite matrix. It also neglects the adsorption and evaporation process of metallics at the fuel rod-graphite and graphite-coolant hole interfaces. Any time delay due to the finite time of transport of fission products by convection through the coolant to the outside of the prestressed concrete reactor vessel (PCRV) is also neglected. This model assumes that all fission products released from fuel particles are immediately deposited outside the PCRV with no time delay

  3. A novel method for fission product noble gas sampling

    International Nuclear Information System (INIS)

    Jain, S.K.; Prakash, Vivek; Singh, G.K.; Vinay, Kr.; Awsthi, A.; Bihari, K.; Joyson, R.; Manu, K.; Gupta, Ashok

    2008-01-01

    Noble gases occur to some extent in the Earth's atmosphere, but the concentrations of all but argon are exceedingly low. Argon is plentiful, constituting almost 1 % of the air. Fission Product Noble Gases (FPNG) are produced by nuclear fission and large parts of FPNG is produced in Nuclear reactions. FPNG are b-j emitters and contributing significantly in public dose. During normal operation of reactor release of FPNG is negligible but its release increases in case of fuel failure. Xenon, a member of FPNG family helps in identification of fuel failure and its extent in PHWRs. Due to above reasons it becomes necessary to assess the FPNG release during operation of NPPs. Presently used methodology of assessment of FPNG, at almost all power stations is Computer based gamma ray spectrometry. This provides fission product Noble gases nuclide identification through peak search of spectra. The air sample for the same is collected by grab sampling method, which has inherent disadvantages. An alternate method was developed at Rajasthan Atomic Power Station (RAPS) - 3 and 4 for assessment of FPNG, which uses adsorption phenomena for collection of air samples. This report presents details of sampling method for FPNG and noble gases in different systems of Nuclear Power Plant. (author)

  4. 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.)

  5. Fission gas release from UO2 pellet fuel at high burn-up

    International Nuclear Information System (INIS)

    Vitanza, C.; Kolstad, E.; Graziani, U.

    1979-01-01

    Analysis of in-reactor measurements of fuel center temperature and rod internal pressure at the OECD Halden Reactor Project has led to the development of an empirical fission gas release model, which is described. The model originally derived from data obtained in the low and intermediate burn-up range, appears to give good predictions for rods irradiated to high exposures as well. PIE puncturing data from seven fuel rods, operated at relatively constant powers and peak center temperatures between 1900 and 2000 0 C up to approx. 40,000 MWd/t UO 2 , did not exhibit any burn-up enhancement on the fission gas release rate

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

    International Nuclear Information System (INIS)

    Van Der Merwet, H.; Venter, J.

    2008-01-01

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

  7. The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Amanda J.; Casella, Andrew M.

    2017-04-01

    The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed. - Highlights: •Complementary fission gas release events are reported for U-Mo fuel with and without cladding. •Exothermic reaction between Zr diffusion layer and cladding influences fission gas release. •Mechanisms responsible for fission gas release are similar, but with varying timing and magnitude. •Behavior of samples is similar after 800 °C signaling the onset of superlattice destabilization.

  8. GRSIS program to predict fission gas release and swelling behavior of metallic fast reactor fuel

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Lee, Byung Ho; Nam, Cheol; Sohn, Dong Seong

    1999-03-01

    A mechanistic model of fission gas release and swelling for the U-(Pu)-Zr metallic fuel in the fast reactor, GRSIS (Gas Release and Swelling in ISotropic fuel matrix) was developed. Fission gas bubbles are assumed to nucleate isotropically from the gas atoms in the metallic fuel matrix since they can nucleate at both the grain boundaries and the phase boundaries which are randomly distributed inside the grain. Bubbles can grow to larger size by gas diffusion and coalition with other bubbles so that they are classified as three classes depending upon their sizes. When bubble swelling reaches the threshold value, bubbles become interconnected each other to make the open channel to the external free space, that is, the open bubbles and then fission gases inside the interconnected open bubbles are released instantaneously. During the irradiation, fission gases are released through the open bubbles. GRSIS model can take into account the fuel gap closure by fuel bubble swelling. When the fuel gap is closed by fuel swelling, the contact pressure between fuel and cladding in relation to the bubble swelling and temperature is calculated. GRSIS model was validated by comparison with the irradiation test results of U-(Pu)-Zr fuels in ANL as well as the parametric studies of the key variable in the model. (author). 13 refs., 1 tab., 22 figs

  9. GRSIS program to predict fission gas release and swelling behavior of metallic fast reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Lee, Byung Ho; Nam, Cheol; Sohn, Dong Seong

    1999-03-01

    A mechanistic model of fission gas release and swelling for the U-(Pu)-Zr metallic fuel in the fast reactor, GRSIS (Gas Release and Swelling in ISotropic fuel matrix) was developed. Fission gas bubbles are assumed to nucleate isotropically from the gas atoms in the metallic fuel matrix since they can nucleate at both the grain boundaries and the phase boundaries which are randomly distributed inside the grain. Bubbles can grow to larger size by gas diffusion and coalition with other bubbles so that they are classified as three classes depending upon their sizes. When bubble swelling reaches the threshold value, bubbles become interconnected each other to make the open channel to the external free space, that is, the open bubbles and then fission gases inside the interconnected open bubbles are released instantaneously. During the irradiation, fission gases are released through the open bubbles. GRSIS model can take into account the fuel gap closure by fuel bubble swelling. When the fuel gap is closed by fuel swelling, the contact pressure between fuel and cladding in relation to the bubble swelling and temperature is calculated. GRSIS model was validated by comparison with the irradiation test results of U-(Pu)-Zr fuels in ANL as well as the parametric studies of the key variable in the model. (author). 13 refs., 1 tab., 22 figs.

  10. On the fission gas release from oxide fuels during normal grain growth

    International Nuclear Information System (INIS)

    Paraschiv, M.C.; Paraschiv, A.; Glodeanu, F.

    1997-01-01

    A mathematical formalism for calculating the fission gas release from oxide fuels considering an arbitrary distribution of fuel grain size with only zero boundary condition for gas diffusion at the grain boundary is proposed. It has also been proved that it becomes unnecessary to consider the grain volume distribution function for fission products diffusion when the grain boundary gas resolution is considered, if thermodynamic forces on grain boundaries are only time dependent. In order to highlight the effect of the normal grain growth on fission gas release from oxide fuels Hillert's and Lifshitz and Slyozov's theories have been selected. The last one was used to give an adequate treatment of normal grain growth for the diffusion-controlled grain boundary movement in oxide fuels. It has been shown that during the fuel irradiation, the asymptotic form of the grain volume distribution functions given by Hillert and Lifshitz and Slyozov models can be maintained but the grain growth rate constant becomes time dependent itself. Experimental results have been used to correlate the two theoretical models of normal grain growth to the fission gas release from oxide fuels. (orig.)

  11. 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

  12. The role of fission gas in the analysis of hypothetical core disruptive accidents

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, E A [Gesellschaft fuer Kernforschung mbH, INR Kernforschungszentrum, Karlsruhe (Germany)

    1977-07-01

    This paper summarizes recent work at Karlsruhe with the goal of understanding the effects of fission gas in hypothetical core disruptive accidents. The fission gas behavior model is discussed. The computer programs LANGZEIT and KURZZEIT describe the long-term and the transient gas behavior, respectively. Recent improvements in the modeling and a comparison of results with experimental data are reported. A somewhat detailed study of the role of fission gas in transient overpower (TOP) accidents was carried out. If pessimistic assumptions, like pin failure near the axial midplane are made, these accidents end in core disassembly. The codes HOPE and KADIS were used to analyze the initiating and the disassembly phase in these studies. Improvements of the codes are discussed. They include an automatic data transfer from HOPE to KADIS, and a new equation of state in KADIS, with an improved model for fission gas behavior. The analysis of a 15 cents/sec reactivity ramp accident is presented. Different pin failure criteria are used. In the cases selected, the codes predict an energetic disassembly. For the much discussed loss-of-flow driven TOP, detailed models are presently not available at Karlsruhe. Therefore, only a few comments and the results of a few scoping calculations will be presented.

  13. 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)

  14. 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.

  15. 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)

  16. A microstructure-dependent model for fission product gas release and swelling in UO2 fuel

    International Nuclear Information System (INIS)

    Notley, M.J.F.; Hastings, I.J.

    1979-06-01

    A model for the release of fission gas from irradiated UO2 fuel is presented. It incorporates fission gas diffusion bubble and grain boundary movement,intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO2 fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW/m, burnups between 10 and 300 MW.h/kg U and power histories including constant, high-to-low and low-to-high power periods. The predictions of the model are shown to be most sensitive to fuel power (temperature), the selection of diffusion coefficient for fission gas in UO2 and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth. (author)

  17. 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

  18. Fission Gas Release in LWR Fuel Rods Exhibiting Very High Burn-Up

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...

  19. The Risoe National Laboratory, Denmark

    International Nuclear Information System (INIS)

    Majborn, B.

    2001-01-01

    The Risoe National Laboratory of Denmark started as a nuclear research centre, under the Atomic Energy Commission in 1955, with research reactors, an accelerator and related facilities. The research component, aimed at the introduction of nuclear power plants in Denmark, was wound up in 1985 with the country deciding to forego nuclear power in its energy planning. From 1993 the centre is under the jurisdiction of the Ministry of Research with three main areas of work: i) research on high international level; ii) train researchers; and iii) provide service to industry. The centre is funded up to 53% by the Danish Government and 47% by contract earnings. Some areas of current research include: i) materials science; ii) optics and sensor systems; iii) plant production and ecology; and iv) systems analysis. The nuclear component of the research centre is related to the operation of the nuclear facilities and for maintaining national expertise in nuclear safety and radiation protection. (author)

  20. On the mobility of fission-gas bubbles

    International Nuclear Information System (INIS)

    Nichols, F.A.; Ronchi, C.

    1986-01-01

    The importance of bubble migration in fuel swelling and fission-product release remains a controversial topic in spite of a great deal of research. For steady state analyses some authors ignore bubble motion totally, whereas others use mobilities (based on out-of-pile measurements) which are far below the theoretical diffusion-control predictions. Under transient conditions some continue to use zero or low bubble mobilities, whereas others invoke higher mobilities. Experimental information on mobility of bubbles under irradiation conditions is very limited, but supports the theoretical values for bubble sizes above 1 μm. The authors discuss here some interesting new results which may provide direct evidence for in-pile mobilities comparable with surface-diffusion control predictions for much smaller bubbles (<20nm), where out-of-pile studies indicate greatly reduced mobilities. A brief summary is presented of information available for bubble mobilities, both in- and out-of-pile

  1. Neutron radiography at the Risoe National Laboratory

    International Nuclear Information System (INIS)

    Domanus, J.C.; Gade-Nielsen, P.; Knudsen, P.; Olsen, J.

    1981-11-01

    In this report six papers are collected which will be presented at the First World Conference on Neutron Radiography in San Diego, U.S.A., 7 - 10 December 1981. They are preceded by a short description of the activities of Risoe National Laboratory in the field of post-irradiation examination of nuclear fuel. One of the nondestructive methods used for this examination is neutron radiography. In the six conference papers different aspects of neutron radiography performed at Risoe are presented. (author)

  2. Fission gas release from oxide fuels at high burnups (AWBA development program)

    International Nuclear Information System (INIS)

    Dollins, C.C.

    1981-02-01

    The steady state gas release, swelling and densification model previously developed for oxide fuels has been modified to accommodate the slow transients in temperature, temperature gradient, fission rate and pressure that are encountered in normal reactor operation. The gas release predictions made by the model were then compared to gas release data on LMFBR-EBRII fuels obtained by Dutt and Baker and reported by Meyer, Beyer, and Voglewede. Good agreement between the model and the data was found. A comparison between the model and three other sets of gas release data is also shown, again with good agreement

  3. Acoustic sensor for in-pile fuel rod fission gas release measurement

    International Nuclear Information System (INIS)

    Fourmentel, D.; Villard, J. F.; Ferrandis, J. Y.; Augereau, F.; Rosenkrantz, E.; Dierckx, M.

    2009-01-01

    We have developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in materials testing reactors. In order to perform experimental programs related to the study of the fission gas release kinetics, the CEA (French Nuclear Energy Commission) acquired the ability to equip a pre-irradiated PWR fuel rod with three sensors, allowing the simultaneous on-line measurements of the following parameters: - fuel temperature with a centre-line thermocouple type C, - internal pressure with a specific counter-pressure sensor, - fraction of fission gas released in the fuel rod with an innovative acoustic sensor. The third detector is the subject of this paper. This original acoustic sensor has been designed to measure the molar mass and pressure of the gas contained in the fuel rod plenum. For in-pile instrumentation, the fraction of fission gas, such as Krypton and Xenon, in Helium, can be deduced online from this measurement. The principle of this acoustical sensor is the following: a piezoelectric transducer generates acoustic waves in a cavity connected to the fuel rod plenum. The acoustic waves are propagated and reflected in this cavity and then detected by the transducer. The data processing of the signal gives the velocity of the acoustic waves and their amplitude, which can be related respectively to the molar mass and to the pressure of the gas. The piezoelectric material of this sensor has been qualified in nuclear conditions (gamma and neutron radiations). The complete sensor has also been specifically designed to be implemented in materials testing reactors conditions. For this purpose some technical points have been studied in details: - fixing of the piezoelectric sample in a reliable way with a suitable signal transmission, - size of the gas cavity to avoid any perturbation of the acoustic waves, - miniaturization of the sensor because of narrow in-pile experimental devices

  4. Relative Release-to-Birth Indicators for Investigating TRISO Fuel Fission Gas Release Models

    International Nuclear Information System (INIS)

    Harp, Jason M.; Hawari, Ayman I.

    2008-01-01

    TRISO microsphere fuel is the fundamental fuel unit for Very High Temperature Reactors (VHTR). A single TRISO particle consists of an inner kernel of uranium dioxide or uranium oxycarbide surrounded by layers of pyrolytic carbon and silicon carbide. If the silicon carbide layer fails, fission products, especially the noble fission gases Kr and Xe, will begin to escape the failed particle. The release of fission gas is usually quantified by measuring the ratio of the released activity (R) to the original birth activity (B), which is designated as the R/B ratio. In this work, relative Release-to-Birth indicators (I) are proposed as a technique for interpreting the results of TRISO irradiation experiments. By implementing a relative metric, it is possible to reduce the sensitivity of the indicators to instrumental uncertainties and variations in experimental conditions. As an example, relative R/B indicators are applied to the interpretation of representative data from the Advanced Gas Reactor-1 TRISO fuel experiment that is currently taking place at the Advanced Test Reactor of Idaho National Laboratory. It is shown that the comparison of measured to predicted relative R/B indicators (I) gives insight into the physics of release and helps validate release models. Different trends displayed by the indicators are related to the mechanisms of fission gas release such as diffusion and recoil. The current analysis shows evidence for separate diffusion coefficients for Kr and Xe and supports the need to account for recoil release. (authors)

  5. Analysis of fuel centre temperatures and fission gas release data from the IFPE Database

    International Nuclear Information System (INIS)

    Schubert, A.; Lassmann, K.; Van Uffelen, P.; Van de Laar, J.; Elenkov, D.; Asenov, S.; Boneva, S.; Djourelov, N.; Georgieva, M.

    2003-01-01

    The present work has continued the analysis of fuel centre temperatures and fission gas release, calculated with standard options of the TRANSURANUS code. The calculations are compared to experimental data from the International Fuel Performance Experiments (IFPE) database. It is reported an analysis regarding UO 2 fuel for Western-type reactors: Fuel centre temperatures measured in the experiments Contact 1 and Contact 2 (in-pile tests of 2 rods performed at the Siloe reactor in Grenoble, France, closely simulating commercial PWR conditions); Fission gas release data derived from post-irradiation examinations of 9 fuel rods belonging to the High-Burnup Effects Programme, task 3 (HBEP3). The results allow for a comparison of predictions by TRANSURANUS for the mentioned Western-type fuels with those done previously for Russian-type WWER fuel. The comparison has been extended to include fuel centre temperatures as well as fission gas release. The present version of TRANSURANUS includes a model that calculates the production of Helium. The amount of produced Helium is compared to the measured and to the calculated release of the fission gases Xenon and Krypton

  6. Analysis for In-situ Fission Rate Measurements using 4He Gas Scintillation Detectors

    International Nuclear Information System (INIS)

    Lewis, Jason M.; Raetz, Dominik; Jordan, Kelly A.; Murer, David

    2013-06-01

    Active neutron interrogation is a powerful NDA technique that relies on detecting and analyzing fission neutrons produced in a fuel sample by an interrogating high neutron flux. 4 He scintillation gas fast neutron detectors are investigated in this paper for use in a novel fission rate measurement technique The He-4 detectors have excellent gamma rejection, a fast response time, and give significant information on incident neutron energy allowing for energy cuts to be applied to the detected signal. These features are shown in this work to allow for the detection of prompt fission neutrons in-situ during active neutron interrogation of a 238 U sample. The energy spectrum from three different neutrons sources ( 252 Cf, AmBe, AmLi) is measured using the 4 He detection system and analyzed. An initial response matrix for the detector is determined using these measurements and the kinematic interaction properties of the elastic scattering with the 4 He. (authors)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  8. Utilization of ''CONTACT'' experiments to improve the fission gas release knowledge in PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Charles, M; Abassin, J J; Bruet, M; Baron, D; Melin, P

    1983-03-01

    The CONTACT experiments, which were carried out by the French CEA, within the framework of a CEA-FRAMATOME collaboration agreement, bear on the behaviour of in-pile irradiated PWR fuel rods. We will focus here upon their results dealing with fission gas release. The experimental device is briefly described, then the following results are given: the kinetics of stable fission gas release for various linear ratings; the instantaneous fractional release rates of radioactive gases versus their decay constant in the range 1.5 10/sup -6/-3.6 10/sup -3/s/sup -1/, for various burnups, as also the influence of fuel temperature. Moreover, the influence of the nature and the pressure of the filling gas upon the release is presented for various linear ratings. The experimental results are discussed and analysed with the purpose to model various physical phenomena involved in the release (low-temperature mechanisms, diffusion).

  9. Utilization of ''CONTACT'' experiments to improve the fission gas release knowledge in PWR fuel rods

    International Nuclear Information System (INIS)

    Charles, M.; Abassin, J.J.; Bruet, M.

    1983-01-01

    The CONTACT experiments, which were carried out by the French CEA, within the framework of a CEA-FRAMATOME collaboration agreement, bear on the behaviour of in-pile irradiated PWR fuel rods. We will focus here upon their results dealing with fission gas release. The experimental device is briefly described, then the following results are given: the kinetics of stable fission gas release for various linear ratings; the instantaneous fractional release rates of radioactive gases versus their decay constant in the range 1.5 10 -6 -3.6 10 -3 s -1 , for various burnups, as also the influence of fuel temperature. Moreover, the influence of the nature and the pressure of the filling gas upon the release is presented for various linear ratings. The experimental results are discussed and analysed with the purpose to model various physical phenomena involved in the release (low-temperature mechanisms, diffusion)

  10. 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)

  11. 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)

  12. On-Line Fission Gas Release Monitoring System in the High Flux Reactor Petten

    International Nuclear Information System (INIS)

    Laurie, M.; Fuetterer, M. A.; Appelman, K.H.; Lapetite, J.-M.; Marmier, A.; Knol, S.; Best, J.

    2013-06-01

    For HTR fuel irradiation tests in the HFR Petten a specific installation was designed and installed dubbed the 'Sweep Loop Facility' (SLF). The SLF is tasked with three functions, namely temperature control by gas mixture technique, surveillance of safety parameters (temperature, pressure, radioactivity etc.) and analysis of fission gas release for three individual capsules in two separate experimental rigs. The SLF enables continuous and independent surveillance of all gas circuits. The release of volatile fission products (FP) from the in-pile experiments is monitored by continuous gas purging. The fractional release of these FP, defined as the ratio between release rate of a gaseous fission isotope (measured) to its instantaneous birth rate (calculated), is a licensing-relevant test for HTR fuel. The developed gamma spectrometry station allows for higher measurement frequencies, thus enabling follow-up of rapid and massive release transients. The designed stand-alone system was tested and fully used through the final irradiation period of the HFR-EU1 experiment which was terminated on 18 February 2010. Its robustness allowed the set up to be used as extra safety instrumentation. This paper describes the gas activity measurement technique based on HPGe gamma spectrometry and illustrates how qualitative and quantitative analysis of volatile FP can be performed on-line. (authors)

  13. The temperature distribution in a gas core fission reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E.; Dam, H. van; Kuijper, J.C. (Interuniversitair Reactor Inst., Delft (Netherlands)); Kistemaker, J.; Boersma-Klein, W.; Vitalis, F. (FOM-Instituut voor Atoom-en Molecuulfysica, Amsterdam (Netherlands))

    1991-01-01

    A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author).

  14. The temperature distribution in a gas core fission reactor

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.; Dam, H. van; Kuijper, J.C.; Kistemaker, J.; Boersma-Klein, W.; Vitalis, F.

    1991-01-01

    A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author)

  15. 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

  16. Experimental verification of the gas pumping theory within fission ionisation chambers

    International Nuclear Information System (INIS)

    Bartlett, A.C.

    1975-01-01

    Experimental verification of a theory for gas loss from in-core ionization chambers is reported. A value of the gas pressure within an irradiated miniature fission chamber was derived indirectly by use of published data on Townsend first coefficient/field across the detector as a function of field/pressure. In practice the voltage corresponding to 10% current multiplication is measured. From the current saturation characteristics measured on the detector during irradiation, the change in gas pressure as a function of fluence was derived and compared to theoretically predicted values. Within the limited accuracy obtainable substantial agreement between measurement and theory is obtained. (O.T.)

  17. Fission gas induced deformation model for FRAP-T6 and NSRR irradiated fuel test simulations

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Takehiko; Sasajima, Hideo; Fuketa, Toyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hosoyamada, Ryuji; Mori, Yukihide

    1996-11-01

    Pulse irradiation tests of irradiated fuels under simulated reactivity initiated accidents (RIAs) have been carried out at the Nuclear Safety Research Reactor (NSRR). Larger cladding diameter increase was observed in the irradiated fuel tests than in the previous fresh fuel tests. A fission gas induced cladding deformation model was developed and installed in a fuel behavior analysis code, FRAP-T6. The irradiated fuel tests were analyzed with the model in combination with modified material properties and fuel cracking models. In Test JM-4, where the cladding temperature rose to higher temperatures and grain boundary separation by the pulse irradiation was significant, the fission gas model described the cladding deformation reasonably well. The fuel had relatively flat radial power distribution and the grain boundary gas from the whole radius was calculated to contribute to the deformation. On the other hand, the power density in the irradiated LWR fuel rods in the pulse irradiation tests was remarkably higher at the fuel periphery than the center. A fuel thermal expansion model, GAPCON, which took account of the effect of fuel cracking by the temperature profile, was found to reproduce well the LWR fuel behavior with the fission gas deformation model. This report present details of the models and their NSRR test simulations. (author)

  18. A fission gas release model for MOX fuel and its verification

    International Nuclear Information System (INIS)

    Koo, Y.H.; Sohn, D.S.; Strijov, P.

    2000-01-01

    A fission gas release model for MOX fuel has been developed based on a model for UO 2 fuel. Using the concept of equivalent cell, the model considers the uneven distribution of Pu within the fuel matrix and a number of Pu-rich particles that could lead to a non-uniform fission rate and fission gas distribution across the fuel pellet. The model has been incorporated into a code, COSMOS, and some parametric studies were made to analyze the effect of the size and Pu content of Pu-rich agglomerates. The model was then applied to the experimental data obtained from the FIGARO program, which consisted of the base irradiation of MOX fuels in the BEZNAU-1 PWR and the subsequent irradiation of four refabricated fuel segments in the Halden reactor. The calculated gas releases show good agreement with the measured ones. In addition, the present analysis indicates that the microstructure of the MOX fuel used in the FIGARO program is such that it has produced little difference in terms of gas release compared with UO 2 fuel. (author)

  19. Fission gas release during power change by means of re-irradiation of spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Jinichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    A full length rod irradiated at Tsuruga unit 1 was refabricated to short length rods, and rod inner pressure gauges were re-instrumented to the rods. Re-irradiation tests to study the fission gas release during power change were carried out by means of BOCA/OSF-1 facility at JMTR. In the tests, steady state operation at 40kW/m, power cycling and daily load follow operations between 20 and 40kW/m were conducted for the same high power holding time, and the rod inner pressure change during the tests was measured. The rod inner pressure increase was observed during power change, especially during power reduction. The rod inner pressure increase during a power cycling depended on the length of the high power operation just before the power cycling. The width of the rod inner pressure increase during a power cycling decreased gradually as the power cycling was repeated continuously. When steady state operation and power cycling were repeated at the power levels of 30, 35 and 40kW/m, the power cycling accelerated the fission gas release compared with the steady state operation. The fission gas release during power reduction is estimated to be the release from FP gas bubbles on the grain boundary caused by the thermal stress in the pellet during power reduction. (author)

  20. Fission gas release from fuels at high burnup

    International Nuclear Information System (INIS)

    Kauffmann, Yves; Pointud, M.L.; Vignesoult, Nicole; Atabek, Rosemarie; Baron, Daniel.

    1982-04-01

    Determinations of residual gas concentrations by heating and by X microanalysis were respectively carried out on particles (TANGO program) and on sections of fuel rods, perfectly characterized as to fabrication and irradiation history. A threshold release temperature of 1250 0 C+-100 0 C was determined irrespective of the type of oxide and the irradiation history in the 18,000-45,000 MWdt -1 (U) specific burnup field. The overall analyses of gas released from the fuel rods show that, in the PWR operating conditions, the fraction released remains less than 1% up to a mean specific burnup of 35000 MWdt -1 (U). The release of gases should not be a limiting factor in the increase of specific burnups [fr

  1. Modelling intragranular fission gas release in irradiation of sintered LWR UO2 fuel

    International Nuclear Information System (INIS)

    Loesoenen, Pekka

    2002-01-01

    A model for the release of stable fission gases by diffuion from sintered LWR UO 2 fuel grains is presented. The model takes into account intragranular gas bubble behaviour as a function of grain radius. The bubbles are assumed to be immobile and the gas migrates to grain boundaries by diffusion of single gas atoms. The intragranular bubble population in the model at low burn-ups or temperatures consists of numerous small bubbles. The presence of the bubbles attenuates the effective gas atom diffusion coefficient. Rapid coarsening of the bubble population in increased burn-up at elevated temperatures weakens significantly the attenuation of the effective diffusion coefficient. The solution method introduced in earlier papers, locally accurate method, is enhanced to allow accurate calculation of the intragranular gas behaviour in time varying conditions without excessive computing time. Qualitatively the detailed model can predict the gas retention in the grain better than a more simple model

  2. Local Fission Gas Release and Swelling in Water Reactor Fuel during Slow Power Transients

    DEFF Research Database (Denmark)

    Mogensen, Mogens Bjerg; Walker, C.T.; Ray, I.L.F.

    1985-01-01

    Gas release and fuel swelling caused by a power increase in a water reactor fuel (burn-up 2.7–4.5% FIMA) is described. At a bump terminal level of about 400 W/cm (local value) gas release was 25–40%. The formation of gas bubbles on grain boundaries and their degree of interlinkage are the two...... factors that determine the level of fission gas release during a power bump. Release begins when gas bubbles on grain boundaries start o interlink. This occurred at r/r0 ~ 0.75. Release tunnels were fully developed at r/r0 ~ 0.55 with the result that gas release was 60–70% at this position....

  3. On the behaviour of intragranular fission gas in UO2 fuel

    International Nuclear Information System (INIS)

    Loesoenen, Pekka

    2000-01-01

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

  4. Numerical algorithms for intragranular diffusional fission gas release incorporated in the Transuranus code

    International Nuclear Information System (INIS)

    Lassmann, K.

    2002-01-01

    Complicated physical processes govern diffusional fission gas release in nuclear fuels. In addition to the physical problem there exists a numerical problem, as some solutions of the underlying diffusion equation contain numerical errors that by far exceed the physical details. In this paper the two algorithms incorporated in the TRANSURANUS code, the URGAS and the new FORMAS algorithm are compared. The previously reported deficiency of the most elegant and mathematically sound FORMAS algorithm at low release could be overcome. Both algorithms are simple, fast, without numerical problems, insensitive to time step lengths and well balanced over the entire range of fission gas release. They can be made available on request as FORTRAN subroutines. (author)

  5. Fission gas release and fuel rod chemistry related to extended burnup

    International Nuclear Information System (INIS)

    1993-04-01

    The purpose of the meeting was to review the state of the art in fission gas release and fuel rod chemistry related to extended burnup. The meeting was held in a time when national and international programmes on water reactor fuel irradiated in experimental reactors were still ongoing or had reached their conclusion, and when lead test assemblies had reached high burnup in power reactors and been examined. At the same time, several out-of-pile experiments on high burnup fuel or with simulated fuel were being carried out. As a result, significant progress has been registered since the last meeting, particularly in the evaluation of fuel temperature, the degradation of the global thermal conductivity with burnup and in the understanding of the impact on fission gas release. Fifty five participants from 16 countries and one international organization attended the meeting. 28 papers were presented. A separate abstract was prepared for each of the papers. Refs, figs, tabs and photos

  6. 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

  7. Inherent safety phenomenon of fission-gas induced axial extrusion in oxide and metal fueled LMFBRs

    International Nuclear Information System (INIS)

    Miles, K.J.; Kalimullah.

    1985-01-01

    The current emphasis in LMFBR design is to develop reactor systems that contain as many features as possible to limit the severity of hypothetical accidents and provide the maximum time before corrective action is required while maintaining low capital costs. One feature is the possibility of fission-gas induced axial extrusion of the fuel within the intact cladding. The potential exists for this phenomenon to enable the reactor to withstand most accidents of the TOP variety, or at least provide an extended time for corrective action to be taken. Under transient conditions which produce a heating of the fuel above its nominal operating temperature, thermal expansion of the material axially produces a negative reactivity effect. This effect is presently considered in most accident analysis codes. The phenomenon of fission-gas induced axial extrusion has received renewed interest because of the consideration of metal alloys of uranium and plutonium for the fuel in some current reactor designs

  8. Fission gas bubble identification using MATLAB's image processing toolbox

    Energy Technology Data Exchange (ETDEWEB)

    Collette, R. [Colorado School of Mines, Nuclear Science and Engineering Program, 1500 Illinois St, Golden, CO 80401 (United States); King, J., E-mail: kingjc@mines.edu [Colorado School of Mines, Nuclear Science and Engineering Program, 1500 Illinois St, Golden, CO 80401 (United States); Keiser, D.; Miller, B.; Madden, J.; Schulthess, J. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2016-08-15

    Automated image processing routines have the potential to aid in the fuel performance evaluation process by eliminating bias in human judgment that may vary from person-to-person or sample-to-sample. This study presents several MATLAB based image analysis routines designed for fission gas void identification in post-irradiation examination of uranium molybdenum (U–Mo) monolithic-type plate fuels. Frequency domain filtration, enlisted as a pre-processing technique, can eliminate artifacts from the image without compromising the critical features of interest. This process is coupled with a bilateral filter, an edge-preserving noise removal technique aimed at preparing the image for optimal segmentation. Adaptive thresholding proved to be the most consistent gray-level feature segmentation technique for U–Mo fuel microstructures. The Sauvola adaptive threshold technique segments the image based on histogram weighting factors in stable contrast regions and local statistics in variable contrast regions. Once all processing is complete, the algorithm outputs the total fission gas void count, the mean void size, and the average porosity. The final results demonstrate an ability to extract fission gas void morphological data faster, more consistently, and at least as accurately as manual segmentation methods. - Highlights: •Automated image processing can aid in the fuel qualification process. •Routines are developed to characterize fission gas bubbles in irradiated U–Mo fuel. •Frequency domain filtration effectively eliminates FIB curtaining artifacts. •Adaptive thresholding proved to be the most accurate segmentation method. •The techniques established are ready to be applied to large scale data extraction testing.

  9. 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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  11. Selective Trapping of Volatile Fission Products with an Off-Gas Treatment System

    Energy Technology Data Exchange (ETDEWEB)

    B.R. Westphal; J.J. Park; J.M. Shin; G.I. Park; K.J. Bateman; D.L. Wahlquist

    2008-07-01

    A head-end processing step, termed DEOX for its emphasis on decladding via oxidation, is being developed for the treatment of spent oxide fuel by pyroprocessing techniques. The head-end step employs high temperatures to oxidize UO2 to U3O8 resulting in the separation of fuel from cladding and the removal of volatile fission products. Development of the head-end step is being performed in collaboration with the Korean Atomic Energy Research Institute (KAERI) through an International Nuclear Energy Research Initiative. Following the initial experimentation for the removal of volatile fission products, an off-gas treatment system was designed in conjunction with KAERI to collect specific fission gases. The primary volatile species targeted for trapping were iodine, technetium, and cesium. Each species is intended to be collected in distinct zones of the off-gas system and within those zones, on individual filters. Separation of the volatile off-gases is achieved thermally as well as chemically given the composition of the filter media. A description of the filter media and a basis for its selection will be given along with the collection mechanisms and design considerations. In addition, results from testing with the off-gas treatment system will be presented.

  12. Contribution to the study of the fission-gas release in metallic nuclear fuels

    International Nuclear Information System (INIS)

    Kryger, B.

    1969-10-01

    In order to study the effect of an external pressure on the limitation of swelling due to fission-gas precipitation, some irradiations have been carried out at burn-ups of about 35.000 MWd/ton, and at average sample temperatures of 575 Celsius degrees, of non-alloyed uranium and uranium 8 per cent molybdenum gained in a thick stainless steel can. A cylindrical central hole allows a fuel swelling from 20 to 33 per cent according to the experiment. After irradiation, the uranium samples showed two types of can rupture: one is due to the fuel swelling, and the other, to the pressure of the fission gases, released through a network of microcracks. The cans of the uranium-molybdenum samples are all undamaged and it is shown that the gas release occurs by interconnection of the bubbles for swelling values higher than those obtained in the case of uranium. For each type of fuel, a swelling-fission gas release relationship is established. The results suggest that good performances with a metallic fuel intended for use in fast reactor conditions can be obtained. (author) [fr

  13. Nonequilibrium behavior of fission gas bubbles with emphasis on the effects of the equation of state

    International Nuclear Information System (INIS)

    Steele, W.G.

    1976-12-01

    The paper presents a computer code designed to estimate fission gas behavior during transient fuel conditions, allowing for nonequilibrium bubble states, with emphasis on equation of state sensitivity. The computer code is a modification of the original code by R. G. Esteves, A. R. Wazzan, and D. Okrent, which in its present form includes the following: resolution, coalescence, leakage to the grain boundary, bubble volume adjustment from a nonequilibrium state by vacancy diffusion, a choice of equation of state between the Van der Waals and the perfect gas equation, the incorporation of hydrostatic pressure values, if known, and conservation of gas atoms. Also, there is a version of the code that allows the existence of single gas atoms in solution in the lattice. The original code is discussed to provide a model of the physical processes and to show a general numerical approach to the estimation of the fission gas behavior. The incorporation of various new features into the original work, such as the option of the Van der Waals gas equation, is described. The various physical models are examined for sensitivity to equation of state for both the equilibrium and nonequilibrium bubble descriptions. Selected computer results of a transient simulation are also presented and general conclusions are drawn upon these results

  14. Thermodynamic performance of a gas-core fission reactor

    International Nuclear Information System (INIS)

    Klein, W.

    1987-01-01

    The purpose of this thesis was to investigate the thermodynamic behaviour of a critical quantity of gaseous uranium-fluorides in chemical equilibrium with a graphite wall. From the very beginning a container was considered with cooled walls. As it was evident that a nuclear reactor working with gaseous fuel should run at much higher temperatures than classical LWR or HTGR reactors, most of the investigations were performed for walls with a surface temperature of 1800 to 2000 K. It was supposed that such a surface temperature would be technologically possible for a heat load between 1 and 5 MWatt m -2 . Cooling with high pressure helium-gas has to keep balance with this heat flux. The technical construction of such a wall will be a problem in itself. It is thought that the experiences with re-entry-vessels in space-technology can be used. A basic assumption in all the calculations is that the U-C-F reactor gas 'sees' a graphite wall, possibly graphite tiles supported by heat resistant materials like SiN 2 , SiC 2 and at a lower temperature level by niobium-steel. Such a gastight compound-system is not necessarily of high-tensile strength materials. It has to be surrounded by a cooled neutron moderator-reflector which in its turn must be supported by a steel-wall at room temperature holding pressure of the order of 100 bar (10 MPa). The design of such a compound-wall is a task for the future. 116 refs.; 28 figs.; 29 tabs

  15. Measuring method for amount of fissionable gas in spent fuel pellet

    International Nuclear Information System (INIS)

    Kashibe, Shinji.

    1992-01-01

    The method of the present invention separately measures the amount of both of a fission product (FP) gas accumulated in bubbles at the crystal grain boundary of spent fuel pellets and an FP gas accumulated in the crystal grains. That is, in a radial position of the spent fuel pellet, a microfine region is mechanically destroyed. The amount of the FP gas released by the destruction from the crystal grains is measured by using a mass analyzer. Then, when the destroyed pieces formed by the destruction are recovered and dissolved, FP gas accumulated in the crystal grains of the pellet is released. The amount released is measured by the mass analyzer. With such procedures, the amount of FP gas accumulated in the bubbles at the crystal grain boundary and in the crystal grains at the radial position of the spent fuel pellet can be measured discriminately. Accordingly, the integrity of the fuel pellet can be recognized appropriately. (I.S.)

  16. Risoe DTU annual report 2009. Highlights from Risoe National Laboratory for Sustainable Energy, DTU

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, Birgit; Bindslev, H. (eds.)

    2010-06-15

    Risoe DTU is the National Laboratory for Sustainable Energy at the Technical University of Denmark. The research focuses on development of energy technologies and systems with minimal effect on climate, and contributes to innovation, education and policy. Risoe has large experimental facilities and interdisciplinary research environments, and includes the national centre for nuclear technologies. The 2009 annual report gives highlights on Risoe's research in the following areas: wind energy, bioenergy, solar energy, fusion energy, fuel cells and hydrogen, energy systems and climate change, and nuclear technologies. It also includes information on Education and training, Innovation and business, Research facilities, and Management, Personnel and Operating statements. (LN)

  17. Risoe DTU annual report 2008. Highlights from Risoe National Laboratory for Sustainable Energy, DTU

    International Nuclear Information System (INIS)

    Pedersen, Birgit; Bindslev, H.

    2009-08-01

    Risoe DTU is the National Laboratory for Sustainable Energy at the Technical University of Denmark. The research focuses on development of energy technologies and systems with minimal effect on climate, and contributes to innovation, education and policy. Risoe has large experimental facilities and interdisciplinary research environments, and includes the national centre for nuclear technologies. The 2008 annual report gives highlights on Risoe's research in the following areas: wind energy, bioenergy, solar energy, fusion energy, fuel cells and hydrogen, energy systems and climate change, and nuclear technologies. It also includes information on Education and training, Innovation and business, Research facilities, and Management, Personnel and Operating statements. (LN)

  18. Risoe DTU annual report 2008. Highlights from Risoe National Laboratory for Sustainable Energy, DTU

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, Birgit; Bindslev, H. (eds.)

    2009-08-15

    Risoe DTU is the National Laboratory for Sustainable Energy at the Technical University of Denmark. The research focuses on development of energy technologies and systems with minimal effect on climate, and contributes to innovation, education and policy. Risoe has large experimental facilities and interdisciplinary research environments, and includes the national centre for nuclear technologies. The 2008 annual report gives highlights on Risoe's research in the following areas: wind energy, bioenergy, solar energy, fusion energy, fuel cells and hydrogen, energy systems and climate change, and nuclear technologies. It also includes information on Education and training, Innovation and business, Research facilities, and Management, Personnel and Operating statements. (LN)

  19. Risoe DTU annual report 2009. Highlights from Risoe National Laboratory for Sustainable Energy, DTU

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, Birgit; Bindslev, H [eds.

    2010-06-15

    Risoe DTU is the National Laboratory for Sustainable Energy at the Technical University of Denmark. The research focuses on development of energy technologies and systems with minimal effect on climate, and contributes to innovation, education and policy. Risoe has large experimental facilities and interdisciplinary research environments, and includes the national centre for nuclear technologies. The 2009 annual report gives highlights on Risoe's research in the following areas: wind energy, bioenergy, solar energy, fusion energy, fuel cells and hydrogen, energy systems and climate change, and nuclear technologies. It also includes information on Education and training, Innovation and business, Research facilities, and Management, Personnel and Operating statements. (LN)

  20. Risoe DTU annual report 2009. Highlights from Risoe National Laboratory for Sustainable Energy, DTU

    International Nuclear Information System (INIS)

    Pedersen, Birgit; Bindslev, H.

    2010-06-01

    Risoe DTU is the National Laboratory for Sustainable Energy at the Technical University of Denmark. The research focuses on development of energy technologies and systems with minimal effect on climate, and contributes to innovation, education and policy. Risoe has large experimental facilities and interdisciplinary research environments, and includes the national centre for nuclear technologies. The 2009 annual report gives highlights on Risoe's research in the following areas: wind energy, bioenergy, solar energy, fusion energy, fuel cells and hydrogen, energy systems and climate change, and nuclear technologies. It also includes information on Education and training, Innovation and business, Research facilities, and Management, Personnel and Operating statements. (LN)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-07

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

  2. Characterization of intergranular fission gas bubbles in U-Mo fuel

    International Nuclear Information System (INIS)

    Kim, Y. S.; Hofman, G.; Rest, J.; Shevlyakov, G. V.

    2008-01-01

    This report can be divided into two parts: the first part, which is composed of sections 1, 2, and 3, is devoted to report the analyses of fission gas bubbles; the second part, which is in section 4, is allocated to describe the mechanistic model development. Swelling data of irradiated U-Mo alloy typically show that the kinetics of fission gas bubbles is composed of two different rates: lower initially and higher later. The transition corresponds to a burnup of ∼0 at% U-235 (LEU) or a fission density of ∼3 x 10 21 fissions/cm 3 . Scanning electron microscopy (SEM) shows that gas bubbles appear only on the grain boundaries in the pretransition regime. At intermediate burnup where the transition begins, gas bubbles are observed to spread into the intragranular regions. At high burnup, they are uniformly distributed throughout fuel. In highly irradiated U-Mo alloy fuel large-scale gas bubbles form on some fuel particle peripheries. In some cases, these bubbles appear to be interconnected and occupy the interface region between fuel and the aluminum matrix for dispersion fuel, and fuel and cladding for monolithic fuel, respectively. This is a potential performance limit for U-Mo alloy fuel. Microscopic characterization of the evolution of fission gas bubbles is necessary to understand the underlying phenomena of the macroscopic behavior of fission gas swelling that can lead to a counter measure to potential performance limit. The microscopic characterization data, particularly in the pre-transition regime, can also be used in developing a mechanistic model that predicts fission gas bubble behavior as a function of burnup and helps identify critical physical properties for the future tests. Analyses of grain and grain boundary morphology were performed. Optical micrographs and scanning electron micrographs of irradiated fuel from RERTR-1, 2, 3 and 5 tests were used. Micrographic comparisons between as-fabricated and as-irradiated fuel revealed that the site of

  3. 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

  4. Analysis of fission gas release in LWR fuel using the BISON code

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-01

    Recent advances in the development of the finite-element based, multidimensional fuel performance code BISON of Idaho National Laboratory are presented. Specifically, the development, implementation and testing of a new model for the analysis of fission gas behavior in LWR-UO2 fuel during irradiation are summarized. While retaining a physics-based description of the relevant mechanisms, the model is characterized by a level of complexity suitable for application to engineering-scale nuclear fuel analysis and consistent with the uncertainties pertaining to some parameters. The treatment includes the fundamental features of fission gas behavior, among which are gas diffusion and precipitation in fuel grains, growth and coalescence of gas bubbles at grain faces, grain growth and grain boundary sweeping effects, thermal, athermal, and transient gas release. The BISON code incorporating the new model is applied to the simulation of irradiation experiments from the OECD/NEA International Fuel Performance Experiments database, also included in the IAEA coordinated research projects FUMEX-II and FUMEX-III. The comparison of the results with the available experimental data at moderate burn-up is presented, pointing out an encouraging predictive accuracy, without any fitting applied to the model parameters.

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

    International Nuclear Information System (INIS)

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

    1979-10-01

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

  6. Impact of gas pressure on fission chamber sensitivity in Campbelling mode

    International Nuclear Information System (INIS)

    Geslot, B.; Blaise, P.; Loiseau, P.; Filliatre, P.; Jammes, C.; Breaud, S.; Villard, J-F.; Blanc-de-Lanaute, N.

    2013-06-01

    The study presented in this paper is based on measurements conducted in the MINERVE zero power reactor operated at CEA Cadarache with a CEA-made U-235 miniature fission chamber (8 mm in diameter) and obtained in both pulse and Campbelling modes. Our objective was to investigate the impact of the filling gas mixture and pressure on each operating mode, using the capacity of the chamber to be refilled with gas. Three gas mixtures were tested (pure Ar, Ar+4%N 2 and Ar+10%CH 4 ) with pressure ranging from 1 to 9 bars. The Mean Fission Product Charge (MFPC), which is the mean charge deposited in the gas by fission products, was obtained from pulse mode signals for each detector setting. It is shown the MFPC is another key parameter to optimize the detector neutron sensitivity, after the fissile coating cross section. Campbelling mode signal was acquired with the Fast Neutron Detector System (FNDS) recently developed by CEA and SCK·CEN. Interesting results were obtained which improve our knowledge of the detector operation. Firstly, it was found that the measurements obtained in both modes are very consistent. The MFPC as a function of the gas pressure was found to be not monotonic. Instead, it features a maximum between 3 and 4 bars. This behavior is expected if the detector does not operate in saturation regime. Indeed, our standard voltage bias of 300 V appeared to be not high enough so that the saturation regime is established. Saturation curves measured in Campbelling mode were fitted using a detector modeling in order to extrapolate the saturation regime MFPC, which came to be independent from the gas. Secondly, obtained results show that the measuring range in Campbelling mode with this detector starts from fission rates as low as a few thousand counts per second. So the so called overlapping range, in which both pulse and Campbelling modes are usable, is about one decade with our spectroscopy modules and more than two decades with fast counting electronic

  7. Simulation of pellet-cladding thermomechanical interaction and fission gas release

    International Nuclear Information System (INIS)

    Denis, Alicia; Soba, Alejandro

    2003-01-01

    This paper summarizes the present status of a computer code that describes some of the main phenomena occurring in a nuclear fuel rod throughout its life. Temperature distribution, thermal expansion, elastic and plastic strains, creep, mechanical interaction between pellet and cladding, fission gas release, gas mixing, swelling, and densification are modeled. The modular structure of the code allows for the incorporation of models to simulate different phenomena and material properties. Collapsible rods can be also simulated. The code is bidimensional, assumes cylindrical symmetry for the rod and uses the finite element method to integrate the differential equations. The stress-strain and heat conduction problems are nonlinear due to plasticity and to the temperature dependence of the thermal conductivity. The fission gas inventory is calculated with a diffusion model, assuming spherical grains and using a one-dimensional finite element scheme. Pressure increase, swelling and densification are coupled with the stress field. Good results are obtained for the simulation of the irradiation tests of the first argentine prototypes of MOX fuels, where the bamboo effect is clearly observed, and of the FUMEX series for the fuel centerline temperature, the inside rod pressure and the fractional gas release.

  8. Analysis of fission gas release-to-birth ratio data from the AGR irradiations

    International Nuclear Information System (INIS)

    Einerson, Jeffrey J.; Pham, Binh T.; Scates, Dawn M.; Maki, John T.; Petti, David A.

    2016-01-01

    A series of advanced gas reactor (AGR) irradiation tests is being conducted in the advanced test reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) fuel used in the High temperature gas-cooled reactor (HTGR). Each AGR test consists of multiple independent capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples (TC) embedded in the graphite enabling temperature control. For AGR-1, the first US irradiation of modern TRISO fuel completed in 2009, there were no particle failures detected. For AGR-2, a few exposed kernels existed in the fuel compacts based upon quality control data. For the AGR-3/4 experiment, particle failures in all capsules were expected because of the use of designed-to-fail (DTF) fuel particles whose kernels are identical to the driver fuel kernels and whose coatings are designed to fail under irradiation. The release-rate-to-birth-rate ratio (R/B) for each of krypton and xenon isotopes is calculated from release rates measured by the germanium detectors used in the AGR fission product monitoring (FPM) system installed downstream from each irradiated capsule. Birth rates are calculated based on the fission power in the experiment and fission product generation models. Thus, this R/B is a measure of the ability of fuel particle coating layers and compact matrix to retain fission gas atoms preventing their release into the sweep gas flow. The major factors that govern gaseous diffusion and release processes are found to be fuel material diffusion coefficient, temperature, and isotopic decay constant. To compare the release behavior among the AGR capsules and historic experiments, the R/B per failed particle is used. HTGR designers use this parameter in their fission product behavior models. For the U.S. TRISO fuel, a regression analysis is performed to establish functional relationships

  9. Analysis of Fission Gas Release-to-Birth Ratio Data from the AGR Irradiations

    International Nuclear Information System (INIS)

    Einerson, Jeffrey J.; Pham, Binh T.; Scates, Dawn M.; Maki, John T.; Petti, David A.

    2014-01-01

    A series of Advanced Gas Reactor (AGR) irradiation tests is being conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) fuel used in the High Temperature Gas-cooled Reactor (HTGR). Each AGR test consists of multiple independent capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples (TC) embedded in the graphite enabling temperature control. For AGR-1, the first US irradiation of modern TRISO fuel completed in 2009, there were no particle failures detected. For AGR-2, a few exposed kernels existed in the fuel compacts based upon quality control data. For the AGR-3/4 experiment, particle failures in all capsules were expected because of the use of designed-to-fail (DTF) fuel particles whose kernels are identical to the driver fuel kernels and whose coatings are designed to fail under irradiation. The release-rate-to-birth-rate ratio (R/B) for each of krypton and xenon isotopes is calculated from release rates measured by the germanium detectors used in the AGR Fission Product Monitoring (FPM) System installed downstream from each irradiated capsule. Birth rates are calculated based on the fission power in the experiment and fission product generation models. Thus, this R/B is a measure of the ability of fuel particle coating layers and compact matrix to retain fission gas atoms preventing their release into the sweep gas flow. The major factors that govern gaseous diffusion and release processes are found to be fuel material diffusion coefficient, temperature, and isotopic decay constant. To compare the release behavior among the AGR capsules and historic experiments, the R/B per failed particle is used. HTGR designers use this parameter in their fission product behavior models. For the U.S. TRISO fuel, a regression analysis is performed to establish functional relationships

  10. Analysis of fission gas release-to-birth ratio data from the AGR irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Einerson, Jeffrey J., E-mail: jeffrey.einerson@inl.gov; Pham, Binh T.; Scates, Dawn M.; Maki, John T.; Petti, David A.

    2016-09-15

    A series of advanced gas reactor (AGR) irradiation tests is being conducted in the advanced test reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) fuel used in the High temperature gas-cooled reactor (HTGR). Each AGR test consists of multiple independent capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples (TC) embedded in the graphite enabling temperature control. For AGR-1, the first US irradiation of modern TRISO fuel completed in 2009, there were no particle failures detected. For AGR-2, a few exposed kernels existed in the fuel compacts based upon quality control data. For the AGR-3/4 experiment, particle failures in all capsules were expected because of the use of designed-to-fail (DTF) fuel particles whose kernels are identical to the driver fuel kernels and whose coatings are designed to fail under irradiation. The release-rate-to-birth-rate ratio (R/B) for each of krypton and xenon isotopes is calculated from release rates measured by the germanium detectors used in the AGR fission product monitoring (FPM) system installed downstream from each irradiated capsule. Birth rates are calculated based on the fission power in the experiment and fission product generation models. Thus, this R/B is a measure of the ability of fuel particle coating layers and compact matrix to retain fission gas atoms preventing their release into the sweep gas flow. The major factors that govern gaseous diffusion and release processes are found to be fuel material diffusion coefficient, temperature, and isotopic decay constant. To compare the release behavior among the AGR capsules and historic experiments, the R/B per failed particle is used. HTGR designers use this parameter in their fission product behavior models. For the U.S. TRISO fuel, a regression analysis is performed to establish functional relationships

  11. Effect of axial diffusional delays on the overall fission gas release

    International Nuclear Information System (INIS)

    Pedersen, N.K.

    1983-01-01

    In fission gas release modeling, it is normally assumed that any locally released gas mixes instantly and perfectly with other gases throughout the internal rod void volume. The present work investigates the consequences of the assumption that perfect mixing is dependent on diffusion, although the subassumption is maintained that pressure equilibrium is instantly achieved. In other words, when a burst of gas release occurs at any axial location, sufficient local accommodation takes place throughout the rod to eliminate any pressure gradients, but due to the narrowness of the passages through fuel cracks and fuel-cladding gap, concentration gradients may still prevail. Diffusion coefficients for the subsequent concentration equilibration are derived from classical theories. Application of one-dimensional diffusion theory is straightforward, but the lack of knowledge of the effective width of the axial passage introduces an uncertainty

  12. Fission gas release during post irradiation annealing of large grain size fuels from Hinkley point B

    International Nuclear Information System (INIS)

    Killeen, J.C.

    1997-01-01

    A series of post-irradiation anneals has been carried out on fuel taken from an experimental stringer from Hinkley Point B AGR. The stringer was part of an experimental programme in the reactor to study the effect of large grain size fuel. Three differing fuel types were present in separate pins in the stringer. One variant of large grain size fuel had been prepared by using an MgO dopant during fuel manufactured, a second by high temperature sintering of standard fuel and the third was a reference, 12μm grain size fuel. Both large grain size variants had similar grain sizes around 35μm. The present experiments took fuel samples from highly rated pins from the stringer with local burn-up in excess of 25GWd/tU and annealed these to temperature of up to 1535 deg. C under reducing conditions to allow a comparison of fission gas behaviour at high release levels. The results demonstrate the beneficial effect of large grain size on release rate of 85 Kr following interlinkage. At low temperatures and release rates there was no difference between the fuel types, but at temperatures in excess of 1400 deg. C the release rate was found to be inversely dependent on the fuel grain size. The experiments showed some differences between the doped and undoped large grains size fuel in that the former became interlinked at a lower temperature, releasing fission gas at an increased rate at this temperature. At higher temperatures the grain size effect was dominant. The temperature dependence for fission gas release was determined over a narrow range of temperature and found to be similar for all three types and for both pre-interlinkage and post-interlinkage releases, the difference between the release rates is then seen to be controlled by grain size. (author). 4 refs, 7 figs, 3 tabs

  13. Fission gas release during post irradiation annealing of large grain size fuels from Hinkley point B

    Energy Technology Data Exchange (ETDEWEB)

    Killeen, J C [Nuclear Electric plc, Barnwood (United Kingdom)

    1997-08-01

    A series of post-irradiation anneals has been carried out on fuel taken from an experimental stringer from Hinkley Point B AGR. The stringer was part of an experimental programme in the reactor to study the effect of large grain size fuel. Three differing fuel types were present in separate pins in the stringer. One variant of large grain size fuel had been prepared by using an MgO dopant during fuel manufactured, a second by high temperature sintering of standard fuel and the third was a reference, 12{mu}m grain size fuel. Both large grain size variants had similar grain sizes around 35{mu}m. The present experiments took fuel samples from highly rated pins from the stringer with local burn-up in excess of 25GWd/tU and annealed these to temperature of up to 1535 deg. C under reducing conditions to allow a comparison of fission gas behaviour at high release levels. The results demonstrate the beneficial effect of large grain size on release rate of {sup 85}Kr following interlinkage. At low temperatures and release rates there was no difference between the fuel types, but at temperatures in excess of 1400 deg. C the release rate was found to be inversely dependent on the fuel grain size. The experiments showed some differences between the doped and undoped large grains size fuel in that the former became interlinked at a lower temperature, releasing fission gas at an increased rate at this temperature. At higher temperatures the grain size effect was dominant. The temperature dependence for fission gas release was determined over a narrow range of temperature and found to be similar for all three types and for both pre-interlinkage and post-interlinkage releases, the difference between the release rates is then seen to be controlled by grain size. (author). 4 refs, 7 figs, 3 tabs.

  14. Hot Experiment on Fission Gas Release Behavior from Voloxidation Process using Spent Fuel

    International Nuclear Information System (INIS)

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

    2007-08-01

    Quantitative analysis of the fission gas release characteristics during the voloxidation and OREOX processes of spent PWR fuel was carried out by spent PWR fuel in a hot-cell of the DFDF. The release characteristics of 85 Kr and 14 C fission gases during voloxidation process at 500 .deg. C is closely linked to the degree of conversion efficiency of UO 2 to U 3 O 8 powder, and it can be interpreted that the release from grain-boundary would be dominated during this step. Volatile fission gases of 14 C and 85 Kr were released to near completion during the OREOX process. Both the 14 C and 85 Kr have similar release characteristics under the voloxidation and OREOX process conditions. A higher burn-up spent fuel showed a higher release fraction than that of a low burn-up fuel during the voloxidation step at 500 .deg. C. It was also observed that the release fraction of semi-volatile Cs was about 16% during a reduction at 1,000 .deg. C of the oxidized powder, but over 90% during the voloxidation at 1,250 .deg. C

  15. Fission track astrology of three Apollo 14 gas-rich breccias

    Science.gov (United States)

    Graf, H.; Shirck, J.; Sun, S.; Walker, R.

    1973-01-01

    The three Apollo 14 breccias 14301, 14313, and 14318 all show fission xenon due to the decay of Pu-244. To investigate possible in situ production of the fission gas, an analysis was made of the U-distribution in these three breccias. The major amount of the U lies in glass clasts and in matrix material and no more than 25% occurs in distinct high-U minerals. The U-distribution of each breccia is discussed in detail. Whitlockite grains in breccias 14301 and 14318 found with the U-mapping were etched and analyzed for fission tracks. The excess track densities are much smaller than indicated by the Xe-excess. Because of a preirradiation history documented by very high track densities in feldspar grains, however, it is impossible to attribute the excess tracks to the decay of Pu-244. A modified track method has been developed for measuring average U-concentrations in samples containing a heterogeneous distribution of U in the form of small high-U minerals. The method is briefly discussed, and results for the rocks 14301, 14313, 14318, 68815, 15595, and the soil 64421 are given.

  16. DEPOSITION OF FISSION PRODUCTS FROM HELIUM GAS FLOWING AT HIGH VELOCITIES

    Energy Technology Data Exchange (ETDEWEB)

    Abriss, A.; Ewing, R. A.; Sunderman, D. N.

    1963-11-15

    From American Nuclear Society Meeting, New York, Nov. 1963. Out-of- pile experiments simulating gas cooled reactor flow and temperature conditions were made to correlate by both empirical and theoretical considerations such parameters as Reynolds numbers, velocity, surface conditions, materials of construction, geometry, particulate matter, and fission product diffusion coefficients. It was concluded that all regions of flow disturbance are areas of buildup of activity. No selectivity in deposition among the elements studied, with the exception of I, Te, and Cs, was found. Relative abundances to each other of less volatile isotopes remained constant throughout any particular experiment. Data are tabulated. (P.C.H.)

  17. Characterization of fission gas bubbles in irradiated U-10Mo fuel

    Energy Technology Data Exchange (ETDEWEB)

    Casella, Andrew M.; Burkes, Douglas E.; MacFarlan, Paul J.; Buck, Edgar C.

    2017-09-01

    Irradiated U-10Mo fuel samples were prepared with traditional mechanical potting and polishing methods with in a hot cell. They were then removed and imaged with an SEM located outside of a hot cell. The images were then processed with basic imaging techniques from 3 separate software packages. The results were compared and a baseline method for characterization of fission gas bubbles in the samples is proposed. It is hoped that through adoption of or comparison to this baseline method that sample characterization can be somewhat standardized across the field of post irradiated examination of metal fuels.

  18. 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

  19. 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

  20. 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.)

  1. Recent improvements in modelling fission gas release and rod deformation on metallic fuel in LMR

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, Byoung-Oon; Kim, Young Jin

    2000-01-01

    Metallic fuel design is a key feature to assure LMR core safety goals. To date, a large effort has been devoted to the development of the MACSIS code for metallic fuel rod design and the evaluation of operational limits under irradiation conditions. The updated models of fission gas release, fuel core swelling, and rod deformation are incorporated into the correspondence routines in MACSIS MOD1. The MACSIS MOD1 which is a new version of MACSIS, has been partly benchmarked on FGR, fuel swelling and rod deformation comparing with the results of U-Zr and U-Pu-Zr metal fuels irradiated in LMRs. The MACSIS MOD1 predicts, relatively well, the absolute magnitudes and trends of the gas release and rod deformations depending on burn-up, and it gives better agreement with the experimental data than the previous predictions of MACSIS and the results of the empirical model

  2. Macroscopic calculational model of fission gas release from water reactor fuels

    International Nuclear Information System (INIS)

    Uchida, Masaki

    1993-01-01

    Existing models for estimating fission gas release rate usually have fuel temperature as independent variable. Use of fuel temperature, however, often brings an excess ambiguity in the estimation because it is not a rigorously definable quantity as a function of heat generation rate and burnup. To derive a mathematical model that gives gas release rate explicitly as a function of design and operational parameters, the Booth-type diffusional model was modified by changing the character of the diffusion constant from physically meaningful quantity into a mere mathematical parameter, and also changing its temperature dependency into power dependency. The derived formula was found, by proper choice of arbitrary constants, to satisfactorily predict the release rates under a variety of irradiation histories up to a burnup of 60,000 MWd/t. For simple power histories, the equation can be solved analytically by defining several transcendental functions, which enables simple calculation of release rate using graphs. (author)

  3. Fission gas release from ThO2 and ThO2--UO2 fuels (LWBR development program)

    International Nuclear Information System (INIS)

    Goldberg, I.; Spahr, G.L.; White, L.S.; Waldman, L.A.; Giovengo, J.F.; Pfennigwerth, P.L.; Sherman, J.

    1978-08-01

    Fission gas release data are presented from 51 fuel rods irradiated as part of the LWBR irradiations test program. The fuel rods were Zircaloy-4 clad and contained ThO 2 or ThO 2 -UO 2 fuel pellets, with UO 2 compositions ranging from 2.0 to 24.7 weight percent and fuel densities ranging from 77.8 to 98.7 percent of theoretical. Rod diameters ranged from 0.25 to 0.71 inches and fuel active lengths ranged from 3 to 84 inches. Peak linear power outputs ranged from 2 to 22 kw/ft for peak fuel burnups up to 56,000 MWD/MTM. Measured fission gas release was quite low, ranging from 0.1 to 5.2 percent. Fission gas release was higher at higher temperature and burnup and was lower at higher initial fuel density. No sensitivity to UO 2 composition was evidenced

  4. Quantitative Analysis of Kr-85 Fission Gas Release from Dry Process for the Treatment of Spent PWR Fuel

    International Nuclear Information System (INIS)

    Park, Geun Il; Cho, Kwang Hun; Lee, Dou Youn; Lee, Jung Won; Park, Jang Jin; Song, Kee Chan

    2007-01-01

    As spent UO 2 fuel oxidizes to U 3 O 8 by air oxidation, a corresponding volume expansion separate grains, releasing the grain-boundary inventory of fission gases. Fission products in spent UO 2 fuel can be distributed in three major regions : the inventory in fuel-sheath gap, the inventory on grain boundaries and the inventory in UO 2 matrix. Release characteristic of fission gases depends on its distribution amount in three regions as well as spent fuel burn-up. Oxidation experiments of spent fuel at 500 .deg. C gives the information of fission gases inventory in spent fuel, and further annealing experiments at higher temperature produces matrix inventory of fission gases on segregated grain. In previous study, fractional release characteristics of Kr- 85 during OREOX (Oxidation and REduction of Oxide fuel) treatment as principal key process for recycling spent PWR fuel via DUPIC cycle have already evaluated as a function of fuel burn-up with 27.3, 35 and 65 MWd/tU. In this paper, new release experiment results of Kr-85 using spent fuel with burn- up of 58 GWd/tU are included to evaluate the fission gas release behavior. As a point of summary in fission gases release behavior, the quantitative analysis of Kr- 85 release characteristics from various spent fuels with different burn-up during voloxidation and OREOX process were reviewed

  5. Assessment of dose measurement uncertainty using RisoScan

    International Nuclear Information System (INIS)

    Helt-Hansen, Jakob; Miller, Arne

    2006-01-01

    The dose measurement uncertainty of the dosimeter system RisoScan, office scanner and Riso B3 dosimeters has been assessed by comparison with spectrophotometer measurements of the same dosimeters. The reproducibility and the combined uncertainty were found to be approximately 2% and 4%, respectively, at one standard deviation. The subroutine in RisoScan for electron energy measurement is shown to give results that are equivalent to the measurements with a scanning spectrophotometer

  6. Decommissioning of the Risoe Hot Cell facility

    International Nuclear Information System (INIS)

    Carlsen, H.

    1994-02-01

    Concise description of progress in hot cell facility decommissioning at Risoe National Laboratory is presented. Removal of the large contaminated equipment has been completed, all the concrete cells have been finally cleaned. The total contamination left on the concrete walls is of the order of 1850 GBq. Preliminary smear tests proved the stack to be probably clean. The delay in project completion seems to be around 7 months, the remaining work being of rather conventional character. (EG)

  7. Decommissioning of the Risoe Hot Cell facility

    International Nuclear Information System (INIS)

    Carlsen, H.

    1991-08-01

    Concise descriptions of actions taken in relation to the decommissioning of the hot cell facility at Risoe National Laboratory are presented. The removal of fissile material, removal and decontamination of large cell internals, and of large equipment such as glove boxes and steel boxes, in addition to dose commitments, are explained. Tables illustrating the analysis of smear tests, constants for contamination level examination, contamination and radiation levels after cleaning and total contamination versus measured radiation are included. (AB)

  8. 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

  9. Experiments to investigate the effects of small changes in fuel stoichiometry on fission gas release

    Energy Technology Data Exchange (ETDEWEB)

    Copeland, P S; Smith, R C [Windscale Lab., AEA Technology, Seascale, Cumbria (United Kingdom)

    1997-08-01

    Fuel pin failure in-reactor leads to fission product and in the case of a PWR fuel debris release to the coolant. For economic reasons immediate shutdown and discharge of failed fuel needs to be avoided but this needs to be counter-balanced against the increasing dose to operators. PWR practice is to continue running wit failed rods, monitoring coolant activity, and only shutting down the reactor and discharging the fuel when circuit activity levels become unacceptable. The rate of fission product release under failed fuel conditions is of key importance and considerable effort has been directed towards establishing the dependency of release on temperature, heating rate, burn-up, and also the extent of fuel oxidation. As a precursor to a possible wider investigation of this area, a small programme was mounted during 1992/1993 to confirm whether small changes in the oxidation state of the fuel, for example those caused by minor cladding defects, would significantly effect fuel behaviour during postulated design basis faults. The objective of the programme was to determine the effects of small departures from stoichiometric fuel composition on fission gas release, and to compare the results with the current methodology for calculating releases under fault conditions. A total of eight experiments was performed. Two were intended as baseline tests to provide a reference with which to compare the effect of oxidation state influenced behaviour with that of thermal effects. It was found that small changes in stoichiometry of {sup {approx}}1 x 10{sup -6} had little or no effect on release but that changes of {sup {approx}} 1 x 10{sup -4} were observed to increase the diffusion coefficient, for {sup 85}Kr, by up to an order of magnitude and hence greatly increase the release rate. The stoichiometry of the sample used in these tests was, for convenience, adjusted using He/H{sub 2}/H{sub 2}O atmospheres. (Abstract Truncated)

  10. Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

    International Nuclear Information System (INIS)

    Scates, Dawn M.; Hartwell, John K.; Walter, John B.; Drigert, Mark W.; Harp, Jason M.

    2010-01-01

    The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  11. The effect of UO2 density on fission product gas release and sheath expansion

    International Nuclear Information System (INIS)

    Notley, M.J.F.; MacEwan, J.R.

    1965-03-01

    The effect of UO 2 density on fission product gas release and sheath expansion has been determined in an irradiation experiment in which the performance of fuel elements with densities between 10.42 and 10.74 g/cm 3 was compared at ∫λdθ values of 39 and 42 W/cm. The elements were irradiated as clusters of four in a pressurized water loop, hence their irradiation histories were identical. Fission product gas release and the extend of grain growth were greater for the lower density elements. Both effects can be attributed solely to the variation of the thermal conductivity of the fuel with the fractional porosity p, if λ p λ [1 - (2.6 ± 0.8) p] where λ is the thermal conductivity of fully dense UO 2 and λ p is that of the porous UO 2 . This expression is in agreement with laboratory findings. A correlation between the extent of grain growth in the UO 2 and the fractional gas release was found to exist in this test and was shown to apply in a large number of other fuel irradiations. Diametral sheath strain was lower for the low density fuel elements than for those of high density, although the former were deduced to have operated with higher central temperatures. It is supposed that the thermal expansion of the fuel can be partially accommodated by elimination of some of the original porosity. The data are consistent with the assumption that approximately half the porosity in the region of the fuel undergoing grain growth is eliminated. (author)

  12. A review of fission gas release data within the Nea/IAEA IFPE database

    International Nuclear Information System (INIS)

    Turnbull, J.A.; Menut, P.; Sartori, E.

    2002-01-01

    The paper describes the International Fuel Performance Experimental (IFPE) database on nuclear fuel performance. The aim of the project is to provide a comprehensive and well-qualified database on Zr clad UO 2 fuel for model development and code validation in the public domain. The data encompass both normal and off-normal operation and include prototypic commercial irradiations as well as experiments performed in material testing reactors. To date, the database contains some 380 individual cases, the majority of which provide data on FGR either from in-pile pressure measurements or PIE techniques including puncturing, electron probe microanalysis (EPMA) and X-ray fluorescence (XRF) measurements. The paper outlines parameters affecting fission gas release and highlights individual datasets addressing these issues. (authors)

  13. Fission gas and iodine release measured up to 15 GWd/t UO2 burnup

    International Nuclear Information System (INIS)

    Appelhans, A.D.

    1983-01-01

    A summary is presented of the measured release of xenon, krypton and iodine up to 15 GWd/t UO 2 burnup for fuel centerline temperatures ranging from 950 to 1800 K, at average linear heat ratings of 15 to 35 kW/m. The IFA-430 is composed of four 1.28-m-long fuel rods containing 10% enriched UO 2 pellet fuel. Two of the fuel rods are connected, top and bottom, to a gas flow system that permits the fission gases released from the fuel pellets to be swept out of the rods during irradiation and measured via gamma spectrometry. The release/burnup increased significantly between 10 and 15 GWd/t burnup. Fuel temperature did not change. Increased releases were due to physical changes in the fuel-surface area. Changes appeared to be due to higher power operation and burnup

  14. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    International Nuclear Information System (INIS)

    McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel; Wirth, Brian; Kennedy, Rory

    2014-01-01

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  15. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean [Texas A & M Univ., College Station, TX (United States); Shao, Lin [Texas A & M Univ., College Station, TX (United States); Tsvetkov, Pavel [Texas A & M Univ., College Station, TX (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Kennedy, Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-07

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  16. Calibration on Pegase of a selective D.R.G. installation for short life and long life fission gas

    International Nuclear Information System (INIS)

    Vasnier, F.

    1968-01-01

    Pegase irradiation loops are equipped with a detection installation which measures the global activity of short-life and long-life fission gases which are released in CO 2 , but the reduced size of circuits in the loop results in an accumulation of long life fission gases, and therefore in problems in the interpretation of measured signals. Thus, the authors propose an additional detection installation which allows long-life fission gases to be separately measured. The principle is to ensure a partial decay of the sampled gas by imposing an additional transit time in order to get rid of short-life fission gases which have a radioactive period of some tenths of a second. A second detector is then used to measure the residual activity of long-life fission gases. The author describes the installation (the normal circuit and the modified circuit), reports the performed tests and the calibration, presents and discusses the obtained results and the installation sensitivity (for short-life and long-life fission gases), and reports their application to the relationship between DRG (sheath failure detection) signals obtained on Pegase and on EDF and EL4 reactors

  17. Effect of power change on fission gas release. Re-irradiation tests of spent fuel at JMTR

    International Nuclear Information System (INIS)

    Nakamura, Jinichi; Shimizu, Michio; Ishii, Tadahiko; Endo, Yasuichi; Ohwada, Isao; Nabeya, Hideaki; Uetsuka, Hiroshi

    1999-01-01

    A full length rod irradiated at Tsuruga unit 1 was refabricated to short length rods, and rod inner pressure gauges were re-instrumented to the rods. Re-irradiation tests to study the fission gas release during power change were carried out by means of BOCA/OSF-1 facility at the JMTR. In the tests, steady state operation at 40 kW/m and power cycling operations between 20 and 40 kW/m were conducted for the same high power holding time, and the rod inner pressure change during the tests was measured. The rod inner pressure increase was observed during power change, especially during power reduction. The rod inner pressure increase during a power cycling depended on the length of the high power operation just before the power cycling. The fission gas release during power reduction is estimated to be the release from fission gas bubbles on the grain boundary caused by the thermal stress in the pellet during power reduction. When steady state operation and power cycling were repeated at the power levels of 30, 35 and 40 kW/m, the power cycling accelerated the fission gas release compared with the steady state operation. (author)

  18. Risoe's activities in 1999; Risoes virksomhedsregnskab 1999. Opfoelgning paa planerne for aaret 1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-04-01

    This report contains an overview of the results obtained at Risoe National Laboratory in 1999. A performance management contract was agreed with the Ministry of Research. The Board of Governors has the obligation to report the annual progress in obtaining specific goals. (au)

  19. Bubble development and fission gas release during rapid heating of 18 GWd/TeU UO2

    International Nuclear Information System (INIS)

    Small, G.J.

    1985-01-01

    Small samples (approximately 50 mg) of UO 2 irradiated to 18 GWd/TeU have been heated rapidly in an out-of-pile furnace. Ramp rates were in the range 10-80 deg. C.s -1 , peak temperatures varied from 1400 deg. C to 2500 deg. C and dwell times from one to fifteen min. The specimens were sealed in small capsules which were subsequently pierced to determine the total amount of fission gas ( 85 Kr) released during each test. Changes in the size and number of gas bubbles on grain boundaries were examined using SEM, TEM, replication and fractography techniques will be employed later. In this paper are reported the first series of gas release results and some metallography. The results are compared with related experiments and some qualitative conclusions are drawn regarding the mechanisms and kinetics of transient fission gas behaviour. (author)

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

    International Nuclear Information System (INIS)

    Szuta, M.

    2006-01-01

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

  1. Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dawn Scates

    2010-10-01

    A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-30

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

  3. Research Establishment Risoe 1975/76

    International Nuclear Information System (INIS)

    1976-11-01

    A summary of the chief activities of the research establishment Risoe is given. These are roughly divided into sections dealing with nuclear technology, applied research, basic research, and research facilities and auxiliary services. For more detailed descriptions of the work in progress, readers are referred to the annual reports published in the two report series, as well as to articles appearing in scientific journals. A selected list of staff publications is given, and the design data on research facilities are presented. (B.P.)

  4. 40 Years of research at Risoe: A platform for the future - interacting with industry and society

    Energy Technology Data Exchange (ETDEWEB)

    Rosendahl, Lis; Lading, Lars [eds.

    1998-08-01

    Risoe`s 40th anniversary was celebrated June 3, 1998 by a symposium held at Risoe. The interaction of research at Risoe with academia and industry was presented in both national and international perspective. Most of the presentations are in English, a few in Danish. (au)

  5. Evaporation release behavior of volatile fission products from liquid sodium pool to the inert cover gas

    Energy Technology Data Exchange (ETDEWEB)

    Nakagiri, T; Miyahara, S [Oarai Engineering Center, Power Reactor and Nuclear Fuel Development Corp., Oaraimachi, Ibaraki (Japan)

    1996-12-01

    In fuel failure of sodium cooled fast breeder reactors, released volatile fission products (VFPs) such as iodine and cesium from the fuel will be dissolved into the liquid sodium coolant and transferred to the cover vaporization. In the cover gas system of the reactor, natural convection occurs due to temperature differences between the sodium pool and the gas phase. The release rates of VFPs together with sodium vaporization are considered to be controlled by the convection. In this study, three analytical models are developed and examined to calculate the transient release rates using the equilibrium partition coefficients of VFPs. The calculated release rates are compared with experimental results for sodium and sodium iodide. The release rate of sodium is closest to the calculation by the heterogeneous nucleation theory. The release rate of sodium iodide obtained from the experiment is between the release rates calculated by the model based on heat-and-mass transfer analogy and the Hill`s theory. From this study, it is confirmed that the realistic release rate of sodium is able to be calculated by the model based on the heterogeneous nucleation theory. The conservative release rate of sodium iodide is able to be calculated by the model based on the Hill`s theory using the equilibrium partition coefficient of sodium iodide. (author) 7 figs., 1 tab., 3 refs.

  6. Irradiated fuel behavior under accident heating conditions and correlation with fission gas release and swelling model (Chicago)

    International Nuclear Information System (INIS)

    Kryger, B.; Ducamp, F.; Combette, P.

    1981-08-01

    We analyse the mixed oxide fast fuel response to off normal conditions obtained by means of an out-of-pile transient simulation apparatus designed to provide direct observations (temperature, pressure, fuel motion) of fuel fission gas phenomena that might occur during the transients. The results are concerning fast transient tests (0,1 to 1 second) obtained with high gas concentration irradiated fuel (4 to 7 at % burn up, 0,4 cm 3 Xe + Kr /g.UPuO 2 ). The kinetics of fission gas release during the transients have been directly measured and then compared with the calculated results issued of the Chicago model. This model agrees, quite well, with other experiments done in the silene prompt reactor. Other gases than xenon and krypton (such as hydrogen and carbon monoxide) do not play any role in fuel behavior, since they have been completely ruled out

  7. Determination of fission gas release of spent nuclear fuel in puncturing test and in leaching experiments under anoxic conditions

    Energy Technology Data Exchange (ETDEWEB)

    González-Robles, E., E-mail: ernesto.gonzalez-robles@kit.edu [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Metz, V. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Wegen, D.H. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Herm, M. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Papaioannou, D. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Bohnert, E. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Gretter, R. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Müller, N. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Nasyrow, R.; Weerd, W. de; Wiss, T. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Kienzler, B. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany)

    2016-10-15

    During reactor operation the fission gases Kr and Xe are formed within the UO{sub 2} matrix of nuclear fuel. Their quantification is important to evaluate their impact on critical parameters regarding the fuel behaviour during irradiation and (long-term) interim storage, such as internal pressure of the fuel rod and fuel swelling. Moreover the content of Kr and Xe in the plenum of a fuel rod and their content in the UO{sub 2} fuel itself are widely used as indicators for the release properties of {sup 129}I, {sup 137}Cs, and other safety relevant radionuclides with respect to final disposal of spent nuclear fuel. The present study deals with the fission gas release from spent nuclear fuel exposed to simulated groundwater in comparison with the fission gas previously released to the fuel rod plenum during irradiation in reactor. In a unique approach we determined both the Kr and Xe inventories in the plenum by means of a puncturing test and in leaching experiments with a cladded fuel pellet and fuel fragments in bicarbonate water under 3.2 bar H{sub 2} overpressure. The fractional inventory of the fission gases released during irradiation into the plenum was (8.3 ± 0.9) %. The fraction of inventory of fission gases released during the leaching experiments was (17 ± 2) % after 333 days of leaching of the cladded pellet and (25 ± 2) % after 447 days of leaching of the fuel fragments, respectively. The relatively high release of fission gases in the experiment with fuel fragments was caused by the increased accessibility of water to the Kr and Xe occluded in the fuel.

  8. Comparison of a fission-gas effects in a transient overpower test (HUT 5-7A) to FRAS3 code predictions

    International Nuclear Information System (INIS)

    Gruber, E.E.; Randklev, E.H.

    1979-01-01

    Fission gas has an important bearing on fuel dynamics during reactor transients. Fission-gas bubble sizes and densities, both within grains and on grain boundaries, are characterized as functions of radial location at the axial midplane in studies of PNL-9 fuel microstructures before and after the HUT 5-7A (PNL 9-25) TREAT test. The FRAS3 code, being developed to model fission-gas effects in reactor transients, is applied to analyze the results of the HUT 5-7A test are presented to illustrate the observed phenomena and the validity of the modeling approach

  9. 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.)

  10. Fission gas release modelling: developments arising from instrumented fuel assemblies, out-of-pile experiments and microstructural observations

    International Nuclear Information System (INIS)

    Leech, N.A.; Smith, M.R.; Pearce, J.H.; Ellis, W.E.; Beatham, N.

    1990-01-01

    This paper reviews the development of fission gas release modelling in thermal reactor fuel (both steady-state and transient) and in particular, illustrates the way in which experimental data have been, and continue to be, the main driving force behind model development. To illustrate this point various aspects of fuel performance are considered: temperature calculation, steady-state and transient fission gas release, grain boundary gas atom capacity and microstructural phenomena. The sources of experimental data discussed include end-of-life fission gas release measurements, instrumented fuel assemblies (e.g. rods with internal pressure transducers, fuel centre thermocouples), swept capsule experiments, out-of-pile annealing experiments and microstructural techniques applied during post-irradiation evaluation. In the case of the latter, the benefit of applying many observation and analysis techniques on the same fuel samples (the approach adopted at NRL Windscale) is emphasized. This illustrates a shift of emphasis in the modelling field from the development of large, complex thermo-mechanical computer codes to the assessment of key experimental data in order to develop and evaluate sub-models which correctly predict the observed behaviour. (author)

  11. Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin; Gamble, Kyle A.; Andersson, David; Mei, Zhi-Gang; Yacout, Abdellatif M.

    2018-01-01

    Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety of LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.

  12. Studies on fuels with low fission gas release. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-10-01

    For more than a decade, the IAEA has organized various specialists meetings to discuss advances in nuclear fuel technology for non-water cooled reactors. In order to review progress in research and development of fuels with low fission gas release for light water reactors, fast reactors and research reactors, an IAEA Technical Committee meeting was organized in October 1996. At the invitation of the Government of the Russian Federation, the meeting was held in Moscow. Experts from seven Member States and one international organization participated. The objective of the meeting was to exchange topical information on such fuels, to evaluate their advantages and drawbacks, and to explore their commercial utilization. The present volume contains the full text of the sixteen papers presented at the meeting. The information compiled in these proceedings should be useful for engineers, scientists and managers from nuclear fuel development organizations, fuel fabrication plants, utilities and regulatory bodies who are involved in the analysis of fuel behaviour under normal and accident conditions. Refs, figs, tabs

  13. Simulation of the effects of grain boundary fission gas during thermal transients

    International Nuclear Information System (INIS)

    Fenske, G.R.; Emerson, J.E.; Beiersdorf, B.A.

    1984-11-01

    This report presents the results of an initial set of out-of-cell transient heating experiments performed on unirradiated UO 2 pellets fabricated to simulate the effect of grain boundary fission gas on fuel swelling and cladding failure. The fabrication involved trapping high-pressure argon on internal pores by sintering annular UO 2 pellets in a hot isostatic press (HIP). The pellet stack was subjected to two separate transients (DGF83-03A and -03B). Figures show photomicrographs of HIPped and non-HIPped UO 2 , respectively, and the adjacent cladding after DGF83-03B. Fuel melting occurred at the center of both the HIPped and non-HIPped pellets; however, a dark ring is present near the center in the HIPped fuel but not in the non-HIPped fuel. This dark band is a high-porosity region due to increased grain boundary/edge swelling in that pellet. In contrast, grain boundary/edge swelling did not occur in the non-HIPped pellets. Thus, the presence of the high-pressure argon trapped on internal pores during sintering in the HIP altered the microstructural behavior. Results of these preliminary tests indicate that the microstructural behavior of HIPped fuel during thermal transients is different from the behavior of conventionally fabricated fuel

  14. Neutron activation analysis (NAA), radioisotope production via neutron activation (PNA) and fission product gas-jet (GJA)

    Energy Technology Data Exchange (ETDEWEB)

    Gaeggeler, H W [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-11-01

    Three different non-diffractive applications of neutrons are outlined, neutron activation analysis, production of radionuclides, mostly for medical applications, and production of short-lived fission nuclides with a so-called gas-jet. It is shown that all three devices may be incorporated into one single insert at SINQ due to their different requests with respect to thermal neutron flux. Some applications of these three facilities are summarized. (author) 3 figs., 1 tab., 8 refs.

  15. Coupled reaction-diffusion equations to model the fission gas release in the irradiation of the uranium dioxide

    International Nuclear Information System (INIS)

    Moyano, Edgardo A.; Scarpettini, Alberto F.

    2003-01-01

    A semi linear model of weakly coupled parabolic p.d.e. with reaction-diffusion is investigated. The system describes fission gas transfer from grain interior of UO 2 to grain boundaries. The problem is studied in a bounded domain. Using the upper-lower solutions method, two monotone sequences for the finite differences equations are constructed. Reasons are mentioned that allow to affirm that in the proposed functional sector the algorithm converges to the unique solution of the differential system. (author)

  16. Neutron activation analysis (NAA), radioisotope production via neutron activation (PNA) and fission product gas-jet (GJA)

    International Nuclear Information System (INIS)

    Gaeggeler, H.W.

    1996-01-01

    Three different non-diffractive applications of neutrons are outlined, neutron activation analysis, production of radionuclides, mostly for medical applications, and production of short-lived fission nuclides with a so-called gas-jet. It is shown that all three devices may be incorporated into one single insert at SINQ due to their different requests with respect to thermal neutron flux. Some applications of these three facilities are summarized. (author) 3 figs., 1 tab., 8 refs

  17. Decommissioning of the Risoe hot cell facility

    International Nuclear Information System (INIS)

    Carlsen, H.

    1992-02-01

    Concise descriptions of actions taken in relation to the decommissioning of the hot cell facility at Risoe National Laboratory are presented. The removal of fissile material, of large contaminated equipment from the concrete cell line and a separate shielded storage facility, and the removal of large contaminated facilities such as out cell parts of a tube transport system between a concrete cell and a lead shielded steel box and a remotely operated Reichert Telatom microscope housed in a lead shielded glove box is described in addition to the initial mapping of radiation levels related to the decontamination of concrete cells. The dose commitment of 17.7 mSv was ascribed to 12 persons in the 2nd half of 1991. The work resulting in these doses was mainly handling of waste together with the frogman entrances in order to repair the in-cell crane and power manipulator. The overall time schedule for the project still appears to be applicable. (AB)

  18. European synchrotron radiation facility at Risoe

    International Nuclear Information System (INIS)

    1981-07-01

    The results of the feasibility study on a potential European Synchrotron Radiation Facility site at Risoe, Denmark, can be summarized as follows: The site is located in a geologically stable area. The ground is fairly flat, free from vibrations and earth movements, and the foundation properties are considered generally good. The study is based upon the machine concept and main geometry as presented in the ESF feasibility study of May 1979. However, the proposed site could accomodate a larger machine (e.g. 900 m of circumference) or a multi-facility centre. The site is located in the vicinity of Risoe National Laboratory, a R and D establishment with 850 employees and a well-developed technical and scientific infrastructure, which can provide support to the ESRF during the plant construction and operation. In particular the possible combination of synchrotron radiation with the existing neutron scattering facilities in DR 3 is emphasized. The site is located 35 km west of Copenhagen with easy access to the scientific, technological and industrial organizations in the metropolitan area. The regional infrastructure ensures easy and fast communication between the ESRF and locations in the host country as well as abroad. The site is located 35 minutes drive from Copenhagen International Airport and on a main communication route out of Copenhagen. The estimated time duration for the design, construction and commissioning of ESRF phase 1 - taking into account national regulatory procedures - is consistent with that of the ESF feasibility study, i.e. approx. 6 years. The estimated captal costs associated with site-specific structures are consistent with those of the ESF feasibility study, taking into account price increase between 1979 and 1981. It should be emphasized that the study is based upon technical and scientific assessments only, and does not reflect any official position or approval from appropriate authorities. (author)

  19. Decommissioning of the Risoe Hot Cell facility

    International Nuclear Information System (INIS)

    Carlsen, H.

    1994-06-01

    Nuclear fuels have been handled and examined after irradiation by physical and chemical techniques, and radiotherapy sources, mainly 60 Co, have been produced at Risoe National Laboratory since 1964. The aims of decommissioning (during 1990-94, at IAEA Stage 2 level for reactors) were to obtain safe conditions for the remaining parts of the facility and to produce clean space areas for new projects. The facility comprises 6 concrete cells, several lead-shielded steel cells, glove boxes, shielded storage for waste, a remotely operated optical microscope, a frogman area for personnel access to the concrete cells, a decontamination facility, workshops and safety installations. All steel cells, glove boxes and the microscope were emptied and removed. The concrete cells were emptied of fissile material, scientific equipment, general tools and scrap. Decontamination should facilitate waste packing and reduce amount of waste to be stored temporarily at the Risoe waste treatment facility together with highly active waste. The concrete cells were cleaned remotely by wiping, hot spot removal, by mechanical means and vacuum cleaning. The interiors of 2 cells were decontaminated by high pressure water jetting. All master-slave manipulators and part of the contaminated ventilation system at the cells were removed. The cells are left in a non-ventilated state, connected to the atmosphere by an absolute filter. The main contaminants measured before cell closure were 60 Co, 137 Cs and alpha-emitters. Dismantling, decontamination waste disposal and received doses are described. Simple techniques involving low doses were found to be very effective. (AB)

  20. Modelling of fission gas release in rods from the International DEMO-RAMP-II Project at Studsvik

    International Nuclear Information System (INIS)

    Malen, K.

    1983-01-01

    The DEMO-RAMP-II rods had a burn-up of 25-30 MWd/kg U. They were ramped to powers in the range 40-50 kW/m with hold times between 10 s and 4.5 minutes. In spite of the short hold times the fission gas release at the higher powers was more than 1%. With these short hold times it is natural to assume that mixing of released gas with plenum gas is limited. Modelling has been performed using GAPCONSV (a modified GAPCON-THERMAL-2) both with and without mixing of released gas with plenum gas. In particular for the high power-short duration ramps only the ''no mixing'' modelling yields release fractions comparable to the experimental values. (author)

  1. Fission Product Release Behavior of Individual Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Minato, Kazuo [Japan Atomic Energy Research Institute (Japan); Sawa, Kazuhiro [Japan Atomic Energy Research Institute (Japan); Koya, Toshio [Japan Atomic Energy Research Institute (Japan); Tomita, Takeshi [Japan Atomic Energy Research Institute (Japan); Ishikawa, Akiyoshi [Japan Atomic Energy Research Institute (Japan); Baldwin, Charles A; Gabbard, William Alexander [Oak Ridge National Laboratory (United States); Malone, Charlie M [Oak Ridge National Laboratory (United States)

    2000-07-15

    Postirradiation heating tests of TRISO-coated UO{sub 2} particles at 1700 and 1800degC were performed to understand fission product release behavior at accident temperatures. The inventory measurements of the individual particles were carried out before and after the heating tests with gamma-ray spectrometry to study the behavior of the individual particles. The time-dependent release behavior of {sup 85}Kr, {sup 110m}Ag, {sup 134}Cs, {sup 137}Cs, and {sup 154}Eu were obtained with on-line measurements of fission gas release and intermittent measurements of metallic fission product release during the heating tests. The inventory measurements of the individual particles revealed that fission product release behavior of the individual particles was not uniform, and large particle-to-particle variations in the release behavior of {sup 110m}Ag, {sup 134}Cs, {sup 137}Cs, and {sup 154}Eu were found. X-ray microradiography and ceramography showed that the variations could not be explained by only the presence or absence of cracks in the SiC coating layer. The SiC degradation may have been related to the variations.

  2. Fission Product Release Behavior of Individual Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Minato, Kazuo; Sawa, Kazuhiro; Koya, Toshio; Tomita, Takeshi; Ishikawa, Akiyoshi; Baldwin, Charles A.; Gabbard, William Alexander; Malone, Charlie M.

    2000-01-01

    Postirradiation heating tests of TRISO-coated UO 2 particles at 1700 and 1800degC were performed to understand fission product release behavior at accident temperatures. The inventory measurements of the individual particles were carried out before and after the heating tests with gamma-ray spectrometry to study the behavior of the individual particles. The time-dependent release behavior of 85 Kr, 110m Ag, 134 Cs, 137 Cs, and 154 Eu were obtained with on-line measurements of fission gas release and intermittent measurements of metallic fission product release during the heating tests. The inventory measurements of the individual particles revealed that fission product release behavior of the individual particles was not uniform, and large particle-to-particle variations in the release behavior of 110m Ag, 134 Cs, 137 Cs, and 154 Eu were found. X-ray microradiography and ceramography showed that the variations could not be explained by only the presence or absence of cracks in the SiC coating layer. The SiC degradation may have been related to the variations

  3. Krypton and xenon in Apollo 14 samples - Fission and neutron capture effects in gas-rich samples

    Science.gov (United States)

    Drozd, R.; Hohenberg, C.; Morgan, C.

    1975-01-01

    Gas-rich Apollo 14 breccias and trench soil are examined for fission xenon from the decay of the extinct isotopes Pu-244 and I-129, and some samples have been found to have an excess fission component which apparently was incorporated after decay elsewhere and was not produced by in situ decay. Two samples have excess Xe-129 resulting from the decay of I-129. The excess is correlated at low temperatures with excess Xe-128 resulting from neutron capture on I-127. This neutron capture effect is accompanied by related low-temperature excesses of Kr-80 and Kr-82 from neutron capture on the bromine isotopes. Surface correlated concentrations of iodine and bromine are calculated from the neutron capture excesses.

  4. On-line mass spectrometry measurement of fission gas release from nuclear fuel submitted to thermal transients

    International Nuclear Information System (INIS)

    Guigues, E.; Janulyte, A.; Zerega, Y.; Pontillon, Y.

    2013-06-01

    The work presented in this paper has been performed in the framework of a joint research program between Aix-Marseille University and CEA Cadarache. The aim is to develop a mass spectrometer (MS) device for the MERARG facility. MERARG is devoted to the study of fission gas release measurement, from nuclear fuels submitted to annealing tests in high activity laboratory such as LECA-STAR, thanks to gamma spectrometry. The mass spectrometer will then extend the measurement capability from the γ-emitters gases to all the gases involved in the release in order to have a better understanding of the fission gas release dynamics from fuel during thermal transients. Furthermore, the mass spectrometer instrument combines the capabilities and performances of both on-line (for release kinetic) and off-line implementations (for delayed accurate analysis of capacities containing total release gas). The paper deals with two main axes: (1) the modelling of gas sampling inlet device and its performance and (2) the first MS qualification/calibration results. The inlet device samples the gas and also adapts the pressure between MERARG sweeping line at 1.2 bar and mass spectrometer chamber at high vacuum. It is a two-stage device comprising a capillary at inlet, an intermediate vacuum chamber, a molecular leak inlet and a two-stage pumping device. Pressure drops, conductance and throughputs are estimated both for mass spectrometer operation and for exhaust gas recovery. Possible gas segregation is also estimated and device modification is proposed to attain a more accurate calibration. First experimental results obtained from a standard gas bottle show that the quantitative analysis at a few ppm level can be achieved for all isotopes of Kr and Xe, as well as masses 2 and 4 u. (authors)

  5. Prediction of the UO/sub 2/ fission gas release data of Bellamy and Rich using a model recently developed by Combustion Engineering

    International Nuclear Information System (INIS)

    Freeburn, H.R.; Pati, S.R.

    1983-01-01

    The trend in the light water reactor industry to higher discharge burnups of UO/sub 2/ fuel rods has initiated the modification of existing fuel rod models to better account for high burnup effects. The degree to which fission gas release from UO/sub 2/ fuel is enhanced at higher burnup is being addressed in the process. Fission gas release modeling should include the separation of the individual effects of thermal diffusion and any burnup enhancement on the release. Although some modelers have interpreted the Bellamy and Rich data on fission gas release from UO/sub 2/ fuel in this fashion, they have assumed that below about 1250 0 C the gas release is not temperature-dependent, and this has led them to predict a very strong burnup enhancement of gas release above 20 MWd/kgU. More recent data, however, suggest that an appreciable amount of fission gas is released by a thermal diffusion mechanism at even lower temperatures and will add to the fission gas released due to the temperature-independent mechanisms of knockout and recoil

  6. Linking photochemistry in the gas and solution phase: S-H bond fission in p-methylthiophenol following UV photoexcitation.

    Science.gov (United States)

    Oliver, Thomas A A; Zhang, Yuyuan; Ashfold, Michael N R; Bradforth, Stephen E

    2011-01-01

    Gas-phase H (Rydberg) atom photofragment translational spectroscopy and solution-phase femtosecond-pump dispersed-probe transient absorption techniques are applied to explore the excited state dynamics of p-methylthiophenol connecting the short time reactive dynamics in the two phases. The molecule is excited at a range of UV wavelengths from 286 to 193 nm. The experiments clearly demonstrate that photoexcitation results in S-H bond fission--both in the gas phase and in ethanol solution-and that the resulting p-methythiophenoxyl radical fragments are formed with significant vibrational excitation. In the gas phase, the recoil anisotropy of the H atom and the vibrational energy disposal in the p-MePhS radical products formed at the longer excitation wavelengths reveal the operation of two excited state dissociation mechanisms. The prompt excited state dissociation motif appears to map into the condensed phase also. In both phases, radicals are produced in both their ground and first excited electronic states; characteristic signatures for both sets of radical products are already apparent in the condensed phase studies after 50 fs. No evidence is seen for either solute ionisation or proton coupled electron transfer--two alternate mechanisms that have been proposed for similar heteroaromatics in solution. Therefore, at least for prompt S-H bond fissions, the direct observation of the dissociation process in solution confirms that the gas phase photofragmentation studies indeed provide important insights into the early time dynamics that transfer to the condensed phase.

  7. Decommissioning of the Risoe Hot Cell facility

    International Nuclear Information System (INIS)

    Carlsen, H.

    1991-02-01

    The Hot Cell facility at Risoe has been in active use since 1964. During the years several types of nuclear fuels have been handled and examined: test reactor fuel pins from the Danish reactor DR3, the Norwegian Halden reactor, etc; power reactor fuel pins from several foreign reactors, including plutonium enriched pins; HTGR fuel from the Dragon reactor. All kinds of physical and chemical non-destructive and destructive post irradiation examinations have been performed. Besides, different radiotherapy sources have been produced, mainly cobalt sources. The general object of the decommissioning programme for the Hot Cell facility was to obtain a safe condition for the total building that does not require the special safety provisions. The hot cell building will be usable for other purposes after decommissioning. The facilicy comprised six concrete cells, lead cells, glove boxes, a shielded unit for temporary storage of waste, frogman area, decontamination areas, workshops, various installations of importance for safe operation of the plant, offices, etc. The tasks comprised e.g. removal of all irradiated fuel items, removal of other radioactive items, removal of contaminated equipment, and decontamination of all the cells and rooms. The goal was to decontaminate all the concrete cells to a degree where no loose contamination exists in the cells, and where the radiation level is so low, that total removal of the cell structures can be done at any time in the future without significant dose commitments. (AB)

  8. Decommissioning of the Risoe Hot Cell facility

    International Nuclear Information System (INIS)

    Carlsen, H.

    1993-02-01

    A concise description of the current status (December 31st, 1992) regarding the decommissioning of the hot cell facility at Risoe National Laboratory is given in this periodic report. During the second half of the year 1992, all remaining fissile material and a large amount of contaminated material were removed, major repair work was carried out on the in-cell crane, the shielded storage facility was decontaminated and sealed, iodine filters in the cell ventilation system were removed, remote cleaning was carried out on three concrete cells to radiation levels acceptable for final cleaning by frogmen, and the remaining work schedule was planned. These processes are briefly described. Some breakdowns of older, but vital equipment (i.e. the in-cell crane and the power manipulator) that was taken into extensive use led to a certain amount of delay. The collective radiation doses during this half-year were no higher than under normal operation of the facility, and amounted to 12 man-mSv ascribed to 14 persons. It was concluded that, when removing old epoxy paint in the cells using paint strippers applied by hand, personnel can wear polythene oversuits, although a technique for remote handling has been developed. Tables illustrate measured radiation levels in cells number 1,4,5 and 6, and a diagram describes the shielded storage facility. (AB)

  9. Decommissioning of the Risoe Hot Cell facility

    International Nuclear Information System (INIS)

    Carlsen, H.

    1993-10-01

    A concise description of the current status of the decommissioning of the hot cell capacity at Risoe National Laboratory is given in this 6th periodic report covering January 1st to June 30th, 1993. All registered and safeguarded fissile material has been removed and the task of cutting and packing scrap material and experimental equipment from the concrete cell line has been completed. Concrete cells 5 and 6 have been finally cleaned and the master slave manipulators removed from them. The major part of the contamination on the shutters and shutter houses were on their horizontal planes and the main contaminant was 137 Cs. Here the surfaces were cleaned by wiping with wet cloths. The method is described. Tables illustrating the resulting contamination levels are included, the density is now low on the shutters. The method of final inn-cell cleaning is explained, and here again tables represent the resulting contamination levels. The work on ''hot spot'' removal and remote cleaning by vacuuming continues on the remaining cells. A collective dose of ca. 16.3 man-mSv was ascribed to 18 persons in the first half of 1993, arising mainly from in-cell work and waste handling. To sum up, the main results from this period are successful removal of last waste from the cells, remote cleaning of cells 2 and 3, final condition for all shutters and shutter housings and final condition for cells 5 and 6. Tables illustrate measured dose rates in detail. (AB)

  10. Equation of state for L.M.F.B.R. fuel (measurement of fission gas release during transients)

    International Nuclear Information System (INIS)

    Combette, P.; Barthelemy, P.

    1979-01-01

    A sample of fuel (UO 2 or UPuO 2 ) can be heated by fission in a heating transient up to energy deposition 4000 j/g, in the Silene reactor. The Kistler type capsule, the calorimeter device and the radiochemical analysis of fission products enable the pressure pulse and the fuel energy deposition to be measured. So, the relationship between the fuel vapour pressure and the fuel specific energy can be deduced. Peaks pressure (about 1 MPa) coming from fresh UO 2 vaporization, have been measured on a 7 milliseconds time scale. There is a good agreement with the E.O.S. for fresh UO 2 , which is well known for low pressure (1 MPa). Numerous tests have been done with 93% enriched UO 2 and a first test with highly active fuel containing plutonium (15 at %) has been performed. The capsule allows the released gas coming from the irradiated fuel to be retained for measurements and analysis. To investigate the mode of fuel disruption, in-pile fission-heated fuel pellets has been recorded by high speed cinematography

  11. Effect of fission gas leakage on heat transfer within a helium filled spent fuel shipping cask

    International Nuclear Information System (INIS)

    Pope, R.B.; Schimmel, W.P. Jr.

    1978-01-01

    Leakage of Xe from spent fuel elements into a He-filled cask would reduce the thermal conductivity, but it would also increase the nondimensional Grashof and Rayleigh number convection parameters. The thermal performance for various quantities of leaked fission gases was evaluated for a cask containing 9 fuel assemblies, each producing approximately 2 kW, and a He partial pressure of 4 atm. If all pins leaked, the max pin temperature would increase from 761 to 839 0 K. It is concluded that the effect of fission gases is a second order effect

  12. ANS-5.4 fission gas release model. I. Noble gases at high temperature

    International Nuclear Information System (INIS)

    Noble, L.D.

    1979-01-01

    A correlation to describe the release of volatile radioactive fission products has been developed by the ANS Working Group (ANS 5.4) on Fuel Plenum Activity. The model for release at higher temperatures is identical in form to conventional diffusion equations, but the effective diffusion coefficient incorporates an explicit dependence upon exposure. Because applicable radioactive release data is limited, parameters in the model were determined from stable fission measurements, and calculated or measured fuel temperatures. Although the model predicts high release, particularly at higher exposures, values for many cases of interest are considerably less than the 100% assumed in some accident analyses: providing potential for removal of unnecessary conservations

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  14. Analysis of UO2 fuel structure for low and high burn-up and its impact on fission gas release

    International Nuclear Information System (INIS)

    Szuta, M.; El-Koliel, M.S.

    1999-01-01

    During irradiation, uranium dioxide (UO 2 ) fuel undergo important restructuring mainly represented by densification and swelling, void migration, equiaxed grain growth, grain subdivision, and the formation of columnar grains. The purpose of this study is to obtain a comprehensive picture of the phenomenon of equiaxed grain growth in UO 2 ceramic material. The change of the grain size in high-density uranium dioxide as a function of temperature, initial grain size, time, and burnup is calculated. Algorithm of fission gas release from UO 2 fuel during high temperature irradiation at high burnup taking into account grain growth effect is presented. Theoretical results are compared with experimental data. (author)

  15. Heated uranium tetrafluoride target system to release non-rare gas fission products for the TRISTAN isotope separator

    International Nuclear Information System (INIS)

    Gill, R.L.

    1977-10-01

    Off-line experiments indicated that fluorides of As, Se, Br, Kr, Zr, Nb, Mo, Tc, Ru, Sb, Te, I and Xe could be volatilized, but except for Br, Kr, I and Xe, none of these elements were observed after mass separation in the on-line experiments. The results of the on-line experiments indicated a very low level of hydride contamination at ambient temperature and consequently, uranium tetrafluoride replaced uranyl stearate as the primary gaseous fission product target. Possible reasons for the failure of the heated target system to yield non-rare gas activities are discussed and suggestions for designing a new heated target system are presented

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

    International Nuclear Information System (INIS)

    Wood, M.H.

    1978-03-01

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

  17. Energy for the future - with Risoe from nuclear power to sustainable energy

    Energy Technology Data Exchange (ETDEWEB)

    Jastrup, M. (ed.)

    2008-07-01

    The title of the book is inspired by Risoe's mission which, at the time of its 50th anniversary, remains uncannily close to that given to Risoe when it was inaugurated in 1958. First and foremost, then as now, Risoe is engaged in the development of tomorrow's energy technologies. In 1958, it was nuclear power. On the occasion of its 50th anniversary, Risoe is working with a palette of sustainable energy sources. (author)

  18. List of selected publications from Risoe's Health Physics Department 1957-1989

    International Nuclear Information System (INIS)

    Heikel Vinther, F.

    1991-01-01

    This list includes scientific and technical papers written by staff members of the former Health Physics Department at Risoe National Laboratory. The first part includes papers in periodicals, proceedings etc. in order of chronology while the second and third part include Riso-R and Riso-M reports respectively arranged according to report numbers. (author)

  19. 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

  20. Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rest, J. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)], E-mail: jrest@anl.gov; Hofman, G.L.; Kim, Yeon Soo [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2009-04-15

    An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than {approx}7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

  1. Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

    Science.gov (United States)

    Rest, J.; Hofman, G. L.; Kim, Yeon Soo

    2009-04-01

    An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

  2. Risoe National Laboratory. List of selected publications 1979

    International Nuclear Information System (INIS)

    1980-11-01

    The list comprises a selection of scientific and technical publications of Risoe National Laboratory and its staff during 1979. Journal articles, conference papers, and reports are included. The publications are arranged in the following broad subject categories: Reactor Safety and Technology, The Nuclear Fuel Cycle, Environmental and General Safety Research, Materials Research, Radiation Technology, Agricultural Research, Non-Nuclear Energy Research, General. (author)

  3. List of selected publications 1981. Risoe National Laboratory

    International Nuclear Information System (INIS)

    1982-07-01

    The list comprises a selection of scientific and technical publications of Risoe National Laboratory and its staff during 1981. Journal articles, conference papers, and reports are included. The publications are arranged in the following broad subject categories: Energy Supply, Environmental and Safety Reseach, Materials Research, Biotechnology and Radiation Research,Experimental Methods and Analyses, Major Research Facilities, General. (author)

  4. List of selected publications 1982. Risoe National Laboratory

    International Nuclear Information System (INIS)

    1983-12-01

    The list comprises a selection of scientific and technical publications of Risoe National Laboratory and its staff during 1982. Journal articles, conference papers, and reports are included. The publications are arranged in the following broad subject categories: Energy Supply and Supporting Technology, Environmental and Safety Research, Materials Research, Biotechnology and Radiation Research, Technical and Administrative Services, General. (author)

  5. Risoe National Laboratory. List of selected publications 1980

    International Nuclear Information System (INIS)

    1981-12-01

    The list comprises a selection of scientific and technical publications of Risoe National Laboratory and its staff during 1980. Journal articles, conference papers, and reports are included. The publications are arranged in the following broad subject categories: Reactor Safety and Technology, The Nuclear Fuel Cycle, Environmental and General Safety Research, Materials Research, and Radiation Technology, Agricultural Research, Non-Nuclear Research, General. (author)

  6. List of selected publications 1978 Risoe National Laboratory

    International Nuclear Information System (INIS)

    1979-09-01

    The list comprises a selection of scientific and technical publications of Risoe National Laboratory and its staff during 1978. Journal articles, conference papers, and reports are included. The publications are arranged in the following broad subject categories: Reactor Safety Technology, The Nuclear Fuel Cycle, Environmental and General Safety Research, Materials Research, Radiation Technology, Agricultural Research, Non-Nuclear Energy Research, General. (author)

  7. List of selected publications 1983. Risoe National Laboratory

    International Nuclear Information System (INIS)

    1985-09-01

    The list comprises a selection of scientific and technical publications of Risoe National Laboratory and its staff during 1983. Journal articles, conference papers, and reports are included. The publications are arranged in the following broad subject categories: Energy Supply and Supporting Technology, Environmental and Safety Research, Materials Research, Biotechnology and Tradiation Research, Technical Support, General. (author)

  8. Diffusion and release of noble gas and halogen fission products with several days half-life in UO2 particle

    International Nuclear Information System (INIS)

    Fang Chao

    2013-01-01

    The exact solutions of diffusion and release model of noble gas and halogen fission products in UO 2 particle of HTGR were built under the conditions of adsorption effect and other physical processes. The corresponding release fractions (F(t)) and the ratio of release and productive amounts (R(t)/B (t)) of fission products were also derived. Furthermore, the F(t) and R(t)/B(t) of 131 I, 131 IXe m , 133 Xe and 133 Xe m whose half-lifes are several days in UO 2 particle with the exact solutions, approximate solutions and corresponding numerical solutions under different temperature histories of reactor core were investigated. The results show that the F(t) and R(t)/B(t) are different in numerical values unless the time of release is long enough. The properties of conservation of exact solutions are much more reasonable than the ones of approximate solutions. It is also found that the results of exact solutions approach the actual working conditions more than the approximate and numerical solutions. (author)

  9. Heat treatments of irradiated uranium oxide in a pressurised water reactor (P.W.R.): swelling and fission gas release

    International Nuclear Information System (INIS)

    Zacharie, I.

    1997-01-01

    In order to keep pressurised water reactors at a top level of safety, it is necessary to understand the chemical and mechanical interaction between the cladding and the fuel pellet due to a temperature increase during a rapid change in reactor. In this process, the swelling of uranium oxide plays an important role. It comes from a bubble precipitation of fission gases which are released when they are in contact with the outside. Therefore, the aim of this thesis consists in acquiring a better understanding of the mechanisms which come into play. Uranium oxide samples, from a two cycles irradiated fuel, first have been thermal treated between 1000 deg C and 1700 deg C for 5 minutes to ten hours. The gas release amount related to time has been measured for each treatment. The comparison of the experimental results with a numerical model has proved satisfactory: it seems that the gases release, after the formation of intergranular tunnels, is controlled by the diffusion phenomena. Afterwards, the swelling was measured on the samples. The microscopic examination shows that the bubbles are located in the grain boundaries and have a lenticular shape. The swelling can be explained by the bubbles coalescence and a model was developed based on this observation. An equation allows to calculate the intergranular swelling in function of time and temperature. The study gives the opportunity to predict the fission gases behaviour during a fuel temperature increase. (author)

  10. A simple method to evaluate the fission gas release at fuel grain boundary including the grain growth both at constant and at transient power histories

    International Nuclear Information System (INIS)

    Paraschiv, M.; Paraschiv, A.

    1991-01-01

    A method to rewrite Fick's second law for a region with a moving boundary when the moving law in time of this boundary is known, has been proposed. This method was applied to Booth's sphere model for radioactive and stable fission product diffusion from the oxide fuel grain in order to take into account the grain growth. The solution of this new equation was presented in the mathematical formulation for power histories from ANS 5.4 model for the stable species. It is very simple to apply and very accurate. The results obtained with this solution for constant and transient temperatures show that the fission gas release (FGR) at grain boundary is strongly dependent on kinetics of grain growth. The utilization of two semiempirical grain growth laws, from published information, shows that the fuel microstructural properties need to be multicitly considered in the fission gas release for every manufacturer of fuel. (orig.)

  11. Separation of the fission product noble gases krypton and xenon from dissolver off-gas in reprocessing HTGR-fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bohnenstingl, J.; Djoa, S. H.; Laser, M.; Mastera, S.; Merz, E.; Morschl, P.

    1976-04-15

    This paper describes a process developed for the retainment and separation of volatile (3H, 129 +131I) and gaseous (85Kr, Xe) fission products from the off-gas produced during dissolution of HTGR-fuel. To prevent unnecessary dilution of liberated noble gases by surrounding atmosphere, a helium purge-gas cycle is applied to enable a coarse fractionating of krypton and xenon by cold-trapping at about 80 deg K after precleaning the gas stream. The process consists of the following steps: deposition of droplets and solid aerosols; chemisorption of iodine on silver impregnated silica gel; catalytic removal of nitrogen oxides and oxygen; drying of the process gas stream; final filtering of abraded solids; deposition of xenon in solid form at 80 deg K and low subpressure; deposition of krypton in solid form at 80 deg K after compression to about 6 bar; decontamination of 85krypton-containing xenon by batch distillation for eventual industrial utilization; and removal of nitrogen and argon enrichment during continuous operation in the purge-gas stream by inleaking air with charcoal. A continuously operating dissolver vessel, closed to the surrounding atmosphere, yields a very high content of noble gases, e.g., 0.35 vol % krypton and 2.0 vol % xenon. The presented off-gas treatment unit is operated in cold runs with 1/3 of the full capacity and can treat about 1 m3 STP/h helium, corresponding to a quantity of about 10,000 MW(e) HTGR-fuel reprocessing plant.

  12. Separation of the fission product noble gases krypton and xenon from dissolver off-gas in reprocessing HTGR-fuel

    International Nuclear Information System (INIS)

    Bohnenstingl, J.; Djoa, S.H.; Laser, M.; Mastera, S.; Merz, E.; Morschl, P.

    1976-01-01

    This paper describes a process developed for the retainment and separation of volatile ( 3 H, 129+131 I) and gaseous ( 85 Kr, Xe) fission products from the off-gas produced during dissolution of HTGR-fuel. To prevent unnecessary dilution of liberated noble gases by surrounding atmosphere, a helium purge-gas cycle is applied to enable a coarse fractionating of krypton and xenon by cold-trapping at about 80 0 K after precleaning the gas stream. The process consists of the following steps: deposition of droplets and solid aerosols; chemisorption of iodine on silver impregnated silica gel; catalytic removal of nitrogen oxides and oxygen; drying of the process gas stream; final filtering of abraded solids; deposition of xenon in solid form at 80 0 K and low subpressure; deposition of krypton in solid form at 80 0 K after compression to about 6 bar; decontamination of 85 Kr-containing xenon by batch distillation for eventual industrial utilization; and removal of nitrogen and argon enrichment during continuous operation in the purge-gas stream by inleaking air with charcoal. A continuously operating dissolver vessel, closed to the surrounding atmosphere, yields a very high content of noble gases, i.e., 0.35 vol % krypton and 2.0 vol % xenon. The presented off-gas treatment unit is operated in cold runs with 1 / 3 of the full capacity and can treat about 1 m 3 STP/h helium, corresponding to a quantity of about 10,000 MW/sub e/ HTGR-fuel reprocessing plant

  13. 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)

  14. 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)

  15. NACOWA experiments on LMFBR cover gas aerosols, heat transfer, and fission product enrichment

    International Nuclear Information System (INIS)

    Minges, J.; Schuetz, W.

    1993-12-01

    Fifteen different NACOWA test series were carried out. The following items were investigated: sodium mass concentration in the cover gas, sodium aerosol particle size, radiative heat transfer across the cover gas, total heat transfer across the cover gas, sodium deposition on the cover plate, temperature profiles across the cover gas, phenomena if the argon cover gas is replaced by helium, enrichment of cesium, iodine, and zinc in the aerosol and in the deposits. The conditions were mainly related to the design parameters of the EFR. According to the first consistent design, a pool temperature of 545 C and a roof temperature of only 120 C were foreseen at a cover gas height of 85 cm. The experiments were carried out in a stainless steel test vessel of 0.6 m diameter and 1.14 m height. Pool temperature (up to 545 C), cover gas height (12.5 cm, 33 cm, and others), and roof temperature (from 110 C to 450 C) were the main parameters. (orig./HP) [de

  16. Risoe National Laboratory - Forty years of research in a changing society

    International Nuclear Information System (INIS)

    Nielsen, H.; Nielsen, K.; Petersen, F.; Siggaard Jensen, H.

    1998-01-01

    The creation of Risoe forty years ago was one of the largest, single investments in Danish research. The intention was to realise Niels Bohr's visions of the peaceful use in Denmark og nuclear energy for electricity production and other purposes. Risoe decided to take the opportunity of its 40th anniversary in 1998 to have its history written in a form that would contribute to the history of modern Denmark. The result was a book in Danish entitled Til samfundets tarv - Forskningscenter Risoes historie. The present text is a slightly reworked translation of the last chapter of that book. It contains a summary of Risoe's history and some reflections on forty years of change. Change in Danish society at large, in research policy, in energy policy, in technological expectations. Changes at Risoe, in leadership, in organisational structure, in strategy and in fields of research. Some of Risoe's largest projects are briefly characterised. (LN)

  17. Wind tunel tests of Risoe-B1-18 and Risoe-B1-24

    Energy Technology Data Exchange (ETDEWEB)

    Fuglsang, P.; Bak, C.; Gaunaa, M.; Antoniou, I.

    2003-01-01

    This report contains 2D measurements of the Risoe-B1-18 and Risoe-B1-24 airfoils. The aerodynamic properties were derived from pressure measurements on the airfoil surface and in the wake. The measurements were conducted in the VELUX open jet wind tunnel, which has a background turbulence intensity of 1%, and an inlet flow velocity of 42 m/s. The airfoil sections had a chord of 0.600 m giving a Reynolds number of 1.6Oe106. The span was 1.9 m and end plates were used to minimize 3D flow effects. The measurements comprised both static and dynamic inflow. Static inflow covered angles of attack from 5o to 30 deg. Dynamic inflow was obtained by pitching the airfoil in a harmonic motion around various mean angles of attack. The test matrix involved smooth flow, various kinds of leading edge roughness, stall strips, vortex generators and Gurney flaps in different combinations. The quality of the measurements was good and the agreement between measurements and numerical CFD predictions with EllipSys2D was good. For both airfoils predictions with turbulent flow captured very well the shapes of lift and drag curves as well as the magnitude of maximum lift. Measurements of Risoe-B1-18 showed that the maximum lift coefficient was 1.64 at an angle of attack of approximately 13 deg. The airfoil was not very sensitive to leading edge roughness despite its high maximum lift. Measurements with stall strips showed that stall strips could control the level of maximum lift. The Risoe-B1-24 measurements showed that the maximum lift coefficient was 1.62 at an angle of attack of approximately 14 deg. The airfoil was only little sensitive to leading edge roughness despite its high relative thickness and high maximum lift. Measurements with delta wing shaped vortex generators increased the maximum lift coefficient to 2.02 and measurements with Gurney flaps increased the maximum lift coefficient to 1.85. Measurements with combination of vortex generators and Gurney flaps showed a maximum

  18. Exhaust gas cleaning system for handling radioactive fission and activation gases

    International Nuclear Information System (INIS)

    Queiser, H.; Schwarz, H.

    1975-01-01

    An exhaust gas cleaning system utilizing the principle of delaying radioactive gases to permit their radioactive decay to a level acceptable for release to the atmosphere, comprising an adsorbent for adsorbing radioactive gas and a container for containing the adsorbent and for constraining gas to flow through the adsorbent, the adsorbent and the container forming simultaneously an adsorptive delay section and a mechanical delay section, by means of a predetermined ratio of volume of voids in the adsorbent to total volume of the container containing the adsorbent, for delaying radioactive gas to permit its radioactive decay to a level acceptable for release to the atmosphere is described. A method of using an adsorbent for cleaning a radioactive gas containing an isotope which is adsorbed by the adsorbent and containing an isotope whose adsorption by the adsorbent is low as compared to the isotope which is adsorbed and which is short-lived as compared to the isotope which is adsorbed, comprising constraining the gas to flow through the adsorbent with the retention time for the isotope which is adsorbed being at least the minimum for permitting radioactive decay to a level acceptable for release to the atmosphere and with the retention time for the isotope of relatively low adsorption and relatively short life being at least the minimum for permitting radioactive decay to a level acceptable for release to the atmosphere is also described. (U.S.)

  19. Fuel efficient hydrodynamic containment for gas core fission reactor rocket propulsion. Final report, September 30, 1992--May 31, 1995

    International Nuclear Information System (INIS)

    Sforza, P.M.; Cresci, R.J.

    1997-01-01

    Gas core reactors can form the basis for advanced nuclear thermal propulsion (NTP) systems capable of providing specific impulse levels of more than 2,000 sec., but containment of the hot uranium plasma is a major problem. The initial phase of an experimental study of hydrodynamic confinement of the fuel cloud in a gas core fission reactor by means of an innovative application of a base injection stabilized recirculation bubble is presented. The development of the experimental facility, a simulated thrust chamber approximately 0.4 m in diameter and 1 m long, is described. The flow rate of propellant simulant (air) can be varied up to about 2 kg/sec and that of fuel simulant (air, air-sulfur hexafluoride) up to about 0.2 kg/sec. This scale leads to chamber Reynolds numbers on the same order of magnitude as those anticipated in a full-scale nuclear rocket engine. The experimental program introduced here is focused on determining the size, geometry, and stability of the recirculation region as a function of the bleed ratio, i.e. the ratio of the injected mass flux to the free stream mass flux. A concurrent CFD study is being carried out to aid in demonstrating that the proposed technique is practical

  20. Fission gas release from the sintered UO{sub 2} fuel; Oslobadjanje fisionih gasova iz goriva od sinterovanog UO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Sigulinski, F; Stevanovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1966-11-15

    This paper shoes the phenomena which control fission gases release from the sintered UO{sub 2} dependent of the burnup rate: ejection, release, diffusion, increased fission gas accumulation causing structural changes in the fuel. release of fission gases from the fuel for power reactors was studied as well. The influence of factors as temperature, characteristics of fuel, burnup rate and burnup level was analyzed. Prikazani su mehanizmi koji kontrolisu izdvajanje fisionih gasova iz sinterovanog UO{sub 2} pri razlicitim brzinama izgaranja: izletanje, izbijanje, difuzija, povecano izdvajanje fisionih gasova koje prati strukturne promene u gorivu. Razmatrano je proucavanje izdvajanja fisionih gasova iz goriva za reaktore snage. Analiziran je uticaj faktora kao sto su temperatura, karakteristike goriva, brzina i stepen izgaranja (author)

  1. Towards the inclusion of open fabrication porosity in a fission gas release model

    Energy Technology Data Exchange (ETDEWEB)

    Claisse, Antoine, E-mail: claisse@kth.se [KTH Royal Institute of Technology, Reactor Physics, AlbaNova University Centre, 106 91, Stockholm (Sweden); Van Uffelen, Paul [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125, Karlsruhe (Germany)

    2015-11-15

    A model is proposed for fission product release in oxide fuels that takes into account the open porosity in a mechanistic manner. Its mathematical framework, assumptions and limitations are presented. It is based on the model for open porosity in the sintering process of crystalline solids. More precisely, a grain is represented by a tetrakaidecahedron and the open porosity is represented by a continuous cylinder along the grain edges. It has been integrated in the TRANSURANUS fuel performance code and applied to the first case of the first FUMEX project as well as to neptunium and americium containing pins irradiated during the SUPERFACT experiment and in the JOYO reactor. The results for LWR and FBR fuels are consistent with the experimental data and the predictions of previous empirical models when the thermal mechanisms are the main drivers of the release, even without using a fitting parameter. They also show a different but somewhat expected behaviour when very high porosity fuels are irradiated at a very low burn-up and at low temperature. - Highlights: • We developed a new athermal FGR model based on the porosity. • We present the model, its framework, assumptions and limitations. • We test it out on several irradiation experiments. • Results are comparable to previous models but without using an empirical parameter.

  2. Risoe energy report 6. Future options for energy technologies

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, Hans; Soenderberg Petersen, L [eds.

    2007-11-15

    Fossil fuels provide about 80% of the global energy demand, and this will continue to be the situation for decades to come. In the European Community we are facing two major energy challenges. The first is sustainability, and the second is security of supply, since Europe is becoming more dependent on imported fuels. These challenges are the starting point for the present Risoe Energy Report 6. It gives an overview of the energy scene together with trends and emerging energy technologies. The report presents status and trends for energy technologies seen from a Danish and European perspective from three points of view: security of supply, climate change and industrial perspectives. The report addresses energy supply technologies, efficiency improvements and transport. The report is volume 6 in a series of reports covering energy issues at global, regional and national levels. The individual chapters of the report have been written by staff members from the Technical University of Denmark and Risoe National Laboratory together with leading Danish and international experts. The report is based on the latest research results from Risoe National Laboratory, Technical University of Denmark, together with available internationally recognized scientific material, and is fully referenced and refereed by renowned experts. Information on current developments is taken from the most up-to-date and authoritative sources available. Our target groups are colleagues, collaborating partners, customers, funding organizations, the Danish government and international organizations including the European Union, the International Energy Agency and the United Nations. (au)

  3. Write-up for the diffractometer D1 at Risoe

    International Nuclear Information System (INIS)

    Bundgaard, J.; Krebs Larsen, F.; Lebech, B.; Nielsen, M.H.; Skaarup, P.

    1982-05-01

    Manual for the crystallographic program system used to control the 4-circle neutron diffractometer D1/TASII at DR3, Risoe. The mechanical part of the diffractometer consists of a monochromator part which allows an easy change of incident neutron wavelenght and a four-circle HUBER goniostate consisting of an Euler cradle (HUBER 512) and two horizontal goniometers (HUBER 440 and HUBER 430). The goniostate is computer controlled by a PDP-11/34 interfaced via CAMAC modules. The PDP-11/34 computer has a 128 k byte memory, two hard magnetic disc stations, a fast DEC-writer terminal and a screen terminal. The diffractometer can be operated remotely via modem and telephone line connections from remote stations such as the University of Aarhus and ILL, Grenoble. Minor parts of the software used to control the diffractometer were developed at Risoe while the major parts were a generous gift to Risoe from College 5, the diffraction group, at the Institute Laue-Langevin, Grenoble, France. (editors)

  4. Development of gas-jet transport systems for fission products and coupling these with methods for continuous separation of short-lived product nuclides

    International Nuclear Information System (INIS)

    Stender, E.

    1979-01-01

    The development of gas-jet transport systems for fission products as well as the coupling of these with continuous separation methods from aqueous solutions (SISAK) and with a mass separator for on-line separation of neutron-rich nuclides are described in this work. Nuclides from the fission of 235 U or other fission materials can be transported using gas-jet systems with thermal neutrons over larger distances (100 m and over). Aerosols (clusters) of either organic (e.g. ethylene) or inorganic nature (e.g. potassium chloride) serve as carrier for the nuclides. The clusters are passed through 1 mm capillaries with a transport gas (nitrogen, helium etc.) under laminar flow conditions. The diameter of the cluster fluctuates between 10 -7 and 10 -6 m. The time required from the production of a nuclide to its detection at the end of a 8 m long capillary tube is 0.8 s for the ethylene/nitrogen and potassium chloride/helium gas-jet systems. By coupling various gas-jet systems with the continuous extraction technique SISAK working with H centrifuges, the elements lanthanum, cerium, praseodymium, zirconium, niobium and technetium can be separated out of the complex fission product mixtures. The on-line technetium chemistry was used with neutron-rich 106 Tc (36 s), 107 Tc (21 s) and 108 Tc (5 s) for γγ(t) measurements. The coupling of a potassium chloride/helium gas jet with a mass separator equiped with a plasma ion source is described. The dependence of the transmission rate of various test parameters is investigated to optimize the system. (orig.) [de

  5. Modelling the behaviour of intergranular fission gas during out-of-pile annealing

    International Nuclear Information System (INIS)

    Valin, S.; Mocellin, A.; Eminet, G.; Ravel, S.

    2002-01-01

    Standard UO 2 fuel irradiated in a commercial PWR (burn-up: 14 GWd/t u ) and hyper-stoichiometric large-grained fuel irradiated in an experimental reactor (burn-up: 9 GWd/t u ) were submitted to out-of-pile annealing treatments under sweeping helium. The 85 Kr release was continuously monitored during the annealing experiments, with temperatures ranging from 1 400 to 1 760 deg C. The burst release was not systematically observed during the heating ramp but did sometimes appear during the isothermal period following the ramp. A model based on microstructural observations was developed to explain the experimental results. The release was shown to depend on gas diffusion towards grain boundaries and on temporary gas trapping in grain boundary bubbles. The gas saturation on grain boundaries, leading to release, depends on their crystallographic orientation. (authors)

  6. Contribution to the study of the fission-gas release in metallic nuclear fuels; Contribution a l'etude du degagement des gaz de fission dans les combustibles nucleaires metalliques

    Energy Technology Data Exchange (ETDEWEB)

    Kryger, B [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1969-10-01

    In order to study the effect of an external pressure on the limitation of swelling due to fission-gas precipitation, some irradiations have been carried out at burn-ups of about 35.000 MWd/ton, and at average sample temperatures of 575 Celsius degrees, of non-alloyed uranium and uranium 8 per cent molybdenum gained in a thick stainless steel can. A cylindrical central hole allows a fuel swelling from 20 to 33 per cent according to the experiment. After irradiation, the uranium samples showed two types of can rupture: one is due to the fuel swelling, and the other, to the pressure of the fission gases, released through a network of microcracks. The cans of the uranium-molybdenum samples are all undamaged and it is shown that the gas release occurs by interconnection of the bubbles for swelling values higher than those obtained in the case of uranium. For each type of fuel, a swelling-fission gas release relationship is established. The results suggest that good performances with a metallic fuel intended for use in fast reactor conditions can be obtained. (author) [French] Afin d'etudier l'effet d'une pression exterieure sur la limitation du gonflement due a la precipitation des gaz de fission, on a irradie a des taux de combustion d'environ 35.000 MWj/t et a des temperatures moyennes de 575 degres des echantillons d'uranium non allie et d'uranium-molybdene 8 pour cent contenus dans une gaine en acier inoxydable epaisse. Un trou cylindrique central permet au combustible de gonfler librement de 20 a 33 pour cent suivant les cas. Apres irradiation les echantillons d'uranium presentent deux types de ruptures de gaine: l'une due au gonflement du combustible, l'autre a la pression des gaz degages, ce degagement des gaz etant provoque par un reseau de micro-fissures. Les gaines des echantillons d'alliage uranium-molybdene sont toutes intactes et l'on montre que le relachement des gaz opere par interconnexion des bulles pour des valeurs de gonflement plus elevees que dans

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

    International Nuclear Information System (INIS)

    Vaeth, L.

    1980-12-01

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

  8. Comparison between two gas-cooled TRU burner subcritical reactors: fusion-fission and ADS

    International Nuclear Information System (INIS)

    Carluccio, T.; Rossi, P.C.R.; Angelo, G.; Maiorino, J.R.

    2011-01-01

    This work shows a preliminary comparative study between two gas cooled subcritical fast reactor as dedicated transuranics (TRU) transmuters: using a spallation neutron source or a D-T fusion neutron source based on ITER. The two concepts are compared in terms of a minor actinides burning performance. Further investigations are required to choose the best partition and transmutation strategy. Mainly due to geometric factors, the ADS shows better neutron multiplication. Other designs, like SABR and lead cooled ADS may show better performances than a Gas Coolead Subcritical Fast Reactors and should be investigated. We noticed that both designs can be utilized to transmutation. Besides the diverse source neutron spectra, we may notice that the geometric design and cycle parameters play a more important role. (author)

  9. 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

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

    International Nuclear Information System (INIS)

    Zweifel, Tobias

    2014-01-01

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

  11. A method of calculating fission gas diffusion from UO{sub 2} fuel and its application to the X-2-f loop test

    Energy Technology Data Exchange (ETDEWEB)

    Booth, A H

    1957-09-15

    A method for calculating the fraction of the rare gas fission products that diffuses out of a UO{sub 2} fuel element under conditions In a reactor is outlined, The method is based on the values of the diffusion constant found in laboratory experiments, as described In CRDC-718, and assumes that these remain unaltered during the period that the fuel is in the reactor, The method has been applied to two types of oxide in the X-2-f loop test of 1956 and the results compared with the amounts of fission gas found by analysis of the gases collected in sheath puncture experiments, as described in CRDC-719. The calculated values depend heavily on the estimated temperatures In the fuel. They are in close agreement with the experimental values provided that, in calculating the temperature, certain assumptions are made regarding the thermal expansion of the oxide cylinder. (author)

  12. 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)

  13. Methods and devices prepared to eliminate activation and fission products from PEC reactor cover gas

    International Nuclear Information System (INIS)

    Caponetti, R.; Gherardi, G.; Petrazzuolo, F.

    1987-01-01

    The major effort made in Italy for the development of fast nuclear reactor is concentrated in the PEC reactor, whose construction is now in the completion stage. The PEC reactor (Prova Elementi di Combustibile - Fuel Element Testing ) is a sodium-cooled reactor with a power rating of 120 MWt, being built for the purpose of studying the behavior of fuel elements under thermal and neutronic conditions similar to those of fast reactor power stations, whit particular attention to safety aspects. The PEC reactor represents a research instrument particularly suitable for studies and experiments in the following fields: performances of the fuel element and its economical optimization (also with the possibility of testing fuel elements not necessarily based on mixed oxides); experiments in the safety field, not only referred to fuel elements, but also to plant subsystems. The experimental program will cover the research of the limit conditions of the typical parameters, such as cladding temperature, linear power, radiation rate, etc. PEC will also allow researches on new-concept fuel elements and thermal, hydraulic and power transients and cycles foreseen in the commercial power plants under normal, upset and emergency conditions. A number of the solutions regarding the PEC reactor and preparatory approaches to its operation are reported in this paper. In particular the following items are discussed: a description of three cover-gas circuits present in the reactor; an estimate of the contamination conditions foreseen under operating conditions; a description of the equipment for the purification of the cover gas and relative operating conditions. There are three cover-gas circuits present in the PEC reactor. They concern the following sodium circuits: primary reactor, primary emergency reactor and sodium purification primary reactor; secondary reactor, test channel and emergency reactor; primary test channel

  14. Feasibility of underground storage/disposal of noble gas fission products

    International Nuclear Information System (INIS)

    Winar, R.M.; Trevorrow, L.E.; Steindler, M.J.

    1979-08-01

    The quantities of 85 Kr that can be released to the environment from nuclear energy production are to be limited after 1983 by Federal regulations. Although procedures for collecting the 85 Kr released in the nuclear fuel cycle have been developed to the point that they are commercially available, procedures for terminal disposal of the collected gas are still being examined for their feasibility. In this work, the possibilities of underground disposal of 85 Kr by several techniques were evaluated. It was concluded that (1) disposal of 85 Kr as a solution in water or other solvents in deep wells would have the major disadvantages of liquid migration and the requirement of extremely large volumes of solvent; (2) disposal as bubbles entrained in cement grout injected underground presents the uncertainty of gaseous migration through permeable solid grout; (3) disposal by injection into abandoned oil fields would be favored by solubility of krypton in residual hydrocarbons, but has the disadvantages that such fields contain numerous shafts offering avenues of escape and also that the fields may be reworked in the future for their hydrocarbon residues; (4) underground retention of 85 Kr injected as a gas may be promising, given the right lithology, through entrapment in interstices between fine sand grains held together by the interfacial tension of wetted surfaces. 9 figures, 5 tables

  15. RisoeScan 1.0 - User manual and toolset for retrospective validation

    Energy Technology Data Exchange (ETDEWEB)

    Helt-Hansen, J

    2004-12-01

    The RisoeScan software is used for dose measurements with radiochromic films that color visibly. This report consists of two documents for use with the RisoeScan software. The User Manual tells how to use the program and the Toolset for Retrospective Validation describes how to perform a retrospective validation of the software. (au)

  16. Fission gas induced fuel swelling in low and medium burnup fuel during high temperature transients

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1980-01-01

    The behavior of light water reactor fuel elements under postulated accident conditions is being studied by the EG and G Idaho, Inc., Thermal Fuels Behavior Program for the Nuclear Regulatory Commission. As a part of this program, unirradiated and previously irradiated, pressurized-water-reactor type fuel rods were tested under power-cooling-mismatch (PCM) conditions in the Power Burst Facility (PBF). During these integral in-reactor experiments, film boiling was produced on the fuel rods which created high fuel and cladding temperatures. Fuel rod diameters increased in the film boiling region to a greater extent for irradiated rods than for unirradiated rods. The purpose of the study was to investigate and assess the fuel swelling which caused the fuel rod diameter increases and to evaluate the ability of an analytical code, the Gas Release and Swelling Subroutine - Steady-State and Transient (GRASS-SST), to predict the results

  17. Fission gas induced fuel swelling in low and medium burnup fuel during high temperature transients. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Vinjamuri, K.

    1980-01-01

    The behavior of light water reactor fuel elements under postulated accident conditions is being studied by the EG and G Idaho, Inc., Thermal Fuels Behavior Program for the Nuclear Regulatory Commission. As a part of this program, unirradiated and previously irradiated, pressurized-water-reactor type fuel rods were tested under power-cooling-mismatch (PCM) conditions in the Power Burst Facility (PBF). During these integral in-reactor experiments, film boiling was produced on the fuel rods which created high fuel and cladding temperatures. Fuel rod diameters increased in the film boiling region to a greater extent for irradiated rods than for unirradiated rods. The purpose of the study was to investigate and assess the fuel swelling which caused the fuel rod diameter increases and to evaluate the ability of an analytical code, the Gas Release and Swelling Subroutine - Steady-State and Transient (GRASS-SST), to predict the results.

  18. In the interest of society - The history of Risoe National Laboratory

    International Nuclear Information System (INIS)

    Nielsen, H.; Nielsen, K.; Petersen, F.; Siggaard Jensen, H.

    1998-01-01

    The creation of Risoe forty years ago was one of the largest, single investments in Danish research. The intention was to realise Niels Bohr's visions of the peaceful use in Denmark of nuclear energy for electricity production and other purposes. Risoe decided to take the opportunity of its 40th anniversary in 1998 to have its history written in a form that would contribute to the history of modern Denmark. Four historians of science and technology were asked to carry out this task, and for almost two years they studied records and publications and interviewed present and former staff members of Risoe. The book recounts Risoe's history of the political and administrative level and presents selected and characteristic aspects of the comprehensive research that has been carried out at Risoe. (LN)

  19. Association Euratom - Risoe National Laboratory annual progress report 1996

    Energy Technology Data Exchange (ETDEWEB)

    Lynov, J.P.; Singh, B.N. [eds.

    1997-05-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics group has activities within development of laser diagnostics for fusion plasmas and studies of nonlinear dynamical processes related to electrostatic turbulence and turbulent transport in magnetized plasmas. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. A summary is presented of the results obtained in the Research Unit during 1996. (au) 5 tabs., 25 ills., 11 refs.

  20. Association Euratom - Risoe National Laboratory annual progress report 2005

    Energy Technology Data Exchange (ETDEWEB)

    Bindslev, H.; Singh, B.N. (eds.)

    2006-11-15

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. A summary is presented of the results obtained in the Research Unit during 2005. (au)

  1. Association Euratom - Risoe National Laboratory annual progress report 1995

    International Nuclear Information System (INIS)

    Lynov, J.P.; Singh, B.N.

    1996-05-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics group has activities within studies of nonlinear dynamical processes in magnetized plasmas, and development of pellet injectors for fusion experiments. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step and the Long-term Technology programme. A summary is presented of the results obtained in the Research Unit during 1995. (au) 5 tabs., 32 ills., 33 refs

  2. Association Euratom - Risoe National Laboratory annual progress report 1994

    International Nuclear Information System (INIS)

    Lynov, J.P.; Michelsen, P.; Singh, B.N.

    1995-06-01

    The program of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics group has activities within (a) studies of nonlinear dynamical processes in magnetized plasmas, (b) development of laser diagnostics for fusion plasmas, and (c) development of pellet injectors for fusion experiments. The activities in technology cover (a) radiation damage of fusion reactor materials and (b) water radiolysis under ITER conditions. A summary of the activities in 1994 is presented. (au) 20 ills., 19 refs

  3. Association Euratom - Risoe National Laboratory annual progress report 1996

    International Nuclear Information System (INIS)

    Lynov, J.P.; Singh, B.N.

    1997-05-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics group has activities within development of laser diagnostics for fusion plasmas and studies of nonlinear dynamical processes related to electrostatic turbulence and turbulent transport in magnetized plasmas. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. A summary is presented of the results obtained in the Research Unit during 1996. (au) 5 tabs., 25 ills., 11 refs

  4. Association Euratom - Risoe National Laboratory annual progress report 2003

    Energy Technology Data Exchange (ETDEWEB)

    Bindslev, H; Singh, B N

    2004-05-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. A summary is presented of the results obtained in the Research Unit during 2003. (au)

  5. Numerical simulation of the RISOe1-airfoil dynamic stall

    Energy Technology Data Exchange (ETDEWEB)

    Bertagnolio, F.; Soerensen, N. [Risoe National Lab., Wind Energy and Atmospheric Physics Dept., Roskilde (Denmark)

    1997-12-31

    In this paper we are concerned with the numerical computation of the dynamic stall that occur in the viscous flowfield over an airfoil. These results are compared to experimental data that were obtained with the new designed RISOe1-airfoil, both for a motionless airfoil and for a pitching motion. Moreover, we present some numerical computations of the plunging and lead-lag motions. We also investigate the possibility of using the pitching motion to simulate the plunging and lead-lag situations. (au)

  6. Association Euratom - Risoe National Laboratory annual progress report 2004

    Energy Technology Data Exchange (ETDEWEB)

    Bindslev, H.; Singh, B.N (eds.)

    2005-06-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. A summary is presented of the results obtained in the Research Unit during 2004. (au)

  7. Association Euratom - Risoe National Laboratory annual progress report 2000

    International Nuclear Information System (INIS)

    Lynov, J.P.; Singh, B.N.

    2001-08-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics group has activities within development of laser diagnostics for fusion plasmas and studies of nonlinear dynamical processes related to turbulence and turbulent transport in the edge region of magnetised fusion plasmas. The activities in technology cover investigations of radiation damage of fusion rector materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. A summary is presented of the results obtained in the Research Unit during 2000. (au)

  8. Association Euratom - Risoe National Laboratory annual progress report 1994

    Energy Technology Data Exchange (ETDEWEB)

    Lynov, J P; Michelsen, P; Singh, B N [eds.

    1995-06-01

    The program of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics group has activities within (a) studies of nonlinear dynamical processes in magnetized plasmas, (b) development of laser diagnostics for fusion plasmas, and (c) development of pellet injectors for fusion experiments. The activities in technology cover (a) radiation damage of fusion reactor materials and (b) water radiolysis under ITER conditions. A summary of the activities in 1994 is presented. (au) 20 ills., 19 refs.

  9. Association Euratom - Risoe National Laboratory. Annual progress report 2002

    International Nuclear Information System (INIS)

    Bindslev, H.; Singh, B.N.

    2003-05-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. (au)

  10. Association Euratom - Risoe National Laboratory annual progress report 1999

    International Nuclear Information System (INIS)

    Lynov, J.P.; Singh, B.N.

    2001-01-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics group has activities within development of laser diagnostics for fusion plasmas and studies of nonlinear dynamical processes related to electrostatic turbulence and turbulent transport in magnetised plasmas. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. A summary is presented of the results obtained in the Research Unit during 1999. (au)

  11. Association Euratom - Risoe National Laboratory annual progress report 2005

    International Nuclear Information System (INIS)

    Bindslev, H.; Singh, B.N.

    2006-11-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. A summary is presented of the results obtained in the Research Unit during 2005. (au)

  12. Association Euratom - Risoe National Laboratory. Annual progress report 2001

    International Nuclear Information System (INIS)

    Bindslev, H.; Singh, B.N.

    2002-06-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. A summary is presented of the results obtained in the Research Unit during 2001. (au)

  13. 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

  14. 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)

  15. A prediction of the UO2 fission gas release data of Bellamy and Rich using a model recently developed by combustion engineering

    International Nuclear Information System (INIS)

    Freeburn, H.R.; Pati, S.R.

    1983-01-01

    The trend in the Light Water Reactor industry to higher discharge burnups of UO 2 fuel rods has initiated the modification of existing fuel rod models to better account for high burnup effects. A model recently developed by Combustion Engineering, Inc. (C-E) for fission gas release from UO 2 fuel recognizes the separate effects of temperature-dependent and temperature-independent release mechanisms. This model accounts for a moderate burnup enhancement that is based on a concept of a saturation inventory existing for the intra- and inter-grannular storage of fission gas within the fuel pellet. The saturation inventory, as modelled, is strongly dependent on the local temperature and the changing grain size of the fuel with burnup. Although the fitting constants of the model were determined solely from more current gas release data from fuel more typical of the C-E product line, the model, nonetheless, provides an excellent prediction of the Bellamy and Rich data over the entire burnup range represented by the data (+-1.6% gas release at a 1σ level). The ability to obtain a good comparison with this data base provides additional support for the use of the particular separation of the effects of thermal diffusion and burnup enhancement on fission gas release that is embodied in the model. Furthermore, the degree of burnup enhancement in the model is believed to be moderate enough to suggest that this high burnup effect should not impede the extension of discharge burnup limits associated with current design fuel rods for Pressurized Water Reactors

  16. High temperature ultrasonic sensor for fission gas characterization in MTR harsh environment

    Science.gov (United States)

    Gatsa, O.; Combette, P.; Rozenkrantz, E.; Fourmentel, D.; Destouches, C.; Ferrandis, J. Y. AD(; )

    2018-01-01

    In the contemporary world, the measurements in hostile environment is one of the predominant necessity for automotive, aerospace, metallurgy and nuclear plant. The measurement of different parameters in experimental reactors is an important point in nuclear power strategy. In the near past, IES (Institut d'Électronique et des Systèmes) on collaboration with CEA (Commissariat à l'Energie Atomique et aux Energies Alternatives) have developed the first ultrasonic sensor for the application of gas quantity determination that has been tested in a Materials Testing Reactor (MTR). Modern requirements state to labor with the materials that possess stability on its parameters around 350°C in operation temperature. Previous work on PZT components elaboration by screen printing method established the new basis in thick film fabrication and characterization in our laboratory. Our trials on Bismuth Titanate ceramics showed the difficulties related to high electrical conductivity of fabricated samples that postponed further research on this material. Among piezoceramics, the requirements on finding an alternative solution on ceramics that might be easily polarized and fabricated by screen printing approach were resolved by the fabrication of thick film from Sodium Bismuth Titanate (NBT) piezoelectric powder. This material exhibits high Curie temperature, relatively good piezoelectric and coupling coefficients, and it stands to be a good solution for the anticipated application. In this paper, we present NBT thick film fabrication by screen printing, characterization of piezoelectric, dielectric properties and material parameters studies in dependence of temperature. Relatively high resistivity in the range of 1.1013 Ohm.cm for fabricated thick film is explained by Aurivillius structure in which a-and b-layers form perovskite structure between oxides of c-layer. Main results of this study are presented and discussed in terms of feasibility for an application to a new sensor

  17. Overview of results from 2D airfoil testing at Risoe

    Energy Technology Data Exchange (ETDEWEB)

    Fuglsang, P. [Risoe National Lab., Wind Energy and Atmospheric Physics Dept., Roskilde (Denmark)

    1997-12-31

    This paper gives an overview of the results from two dimensional airfoil testing at Risoe. A two dimensional testing method was recently developed where a test rig is inserted into an open jet flow in a wind tunnel of the close return loop type with an open test section. Pressure measurements provide the lift and drag forces. Both stationary flow and dynamic inflow from pitch motion are possible. The wind tunnel static pressure and total dynamic pressures were calibrated and wind tunnel boundary corrections were found. So far, the testing method was verified by comparison of NACA 63-215 airfoil measurements to numerical predictions and to measurements. Furthermore, the Risoe-1, FFA-W3-241, FFA-W3-301 and NACA 63-430 airfoils were measured. Different types of leading edge roughness and vortex generators were investigated. For all airfoils, good agreements with predictions were obtained on both pressure distribution and on lift coefficient. The drag coefficients were slightly higher than predicted. (eg) 10 refs.

  18. Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Behafarid, F.; Shaver, D. R. [Rensselaer Polytechnic Inst., Troy, NY (United States); Bolotnov, I. A. [North Carolina State Univ., Raleigh, NC (United States); Jansen, K. E. [Univ. of Colorado, Boulder, CO (United States); Antal, S. P.; Podowski, M. Z. [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2012-07-01

    The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

  19. 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

  20. 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)

  1. Fission gas and iodine release measured in IFA-430 up to 15 GWd/t UO2 burnup

    International Nuclear Information System (INIS)

    Appelhans, A.D.; Turnbull, J.A.; White, R.J.

    1983-01-01

    The release of fission products from fuel pellets to the fuel-cladding gap is dependent on the fuel temperature, the power (fission rate) and the burnup (fuel structure). As part of the US Nuclear Regulatory Commission's Fuel Behavior Program, EG and G Idaho, Inc., is conducting fission product release studies in the Heavy Boiling Water Reactor in Halden, Norway. This paper presents a summary of the results up to December, 1982. The data cover fuel centerline temperatures ranging from 700 to 1500 0 C for average linear heat ratings of 16 to 35 kW/m. The measurements have been performed for the period between 4.2 and 14.8 GWd/t UO 2 of burnup of the Instrumented Fuel Assembly 430 (IFA-430). The measurement program has been directed toward quantifying the release of the short-lived radioactive noble gases and iodines

  2. Pilot experiments with relativistic uranium projectile and fission fragments thermalized in a cryogenic gas-filled stopping cell

    Energy Technology Data Exchange (ETDEWEB)

    Reiter, Moritz Pascal

    2015-07-01

    online experiments, which is about three times larger than any stopping cell, using RF structures for the extraction of ions, has demonstrated. The area density and therefore the stopping power of the CSC is limited by the differential pumping. To overcome this limitation the CSC was tested with neon as a stopping gas with area densities of up to 11.3 mg/cm{sup 2} helium equivalent, demonstrating a unprecedented area density for stopping cells based on RF structures. The RF carpet performed reliably and its potential for the future FAIR stopping cell was shown. During the experiments at GSI the mean extraction time of {sup 221}Ac ions from the CSC to a silicon surface detector was measured, it amounts to 24 ms. This value is well in agreement with offline measurements using a pulsed {sup 223}Ra recoil ion source. The combination of a high density stopping cell with high total efficiencies and a non-scanning high-resolution mass spectrometer can be used as an independent identification detector for exotic nuclei by their mass, allowing a recalibration of the in-flight detectors of any fragment separator. As a proof-of-principal experiment the CSC and a MR-TOF-MS have been used as a mass tagger for the FRS at GSI. 134-I ions were produced by in-flight fission from an {sup 238}U primary beam at 1000 MeV/u and identified by the mass tagger. The new method does not rely on specific decay properties and therefore allows a recalibration of the fragment separator independent of the fragment and can also be used with stable nuclides. The usage of the CSC and a MR-TOF-MS will allow fast recalibration and a more effective usage of the limited amount of beam time for all experiments with exotic nuclei even in the case the nuclide of interest is not clearly identified by the in-flight detection scheme. With the CSC low energy experiments such as high-precision mass measurements and decay spectroscopy were made possible, the half lifes of {sup 221}Ac and {sup 223}Th have been measured

  3. Pilot experiments with relativistic uranium projectile and fission fragments thermalized in a cryogenic gas-filled stopping cell

    International Nuclear Information System (INIS)

    Reiter, Moritz Pascal

    2015-01-01

    experiments, which is about three times larger than any stopping cell, using RF structures for the extraction of ions, has demonstrated. The area density and therefore the stopping power of the CSC is limited by the differential pumping. To overcome this limitation the CSC was tested with neon as a stopping gas with area densities of up to 11.3 mg/cm"2 helium equivalent, demonstrating a unprecedented area density for stopping cells based on RF structures. The RF carpet performed reliably and its potential for the future FAIR stopping cell was shown. During the experiments at GSI the mean extraction time of "2"2"1Ac ions from the CSC to a silicon surface detector was measured, it amounts to 24 ms. This value is well in agreement with offline measurements using a pulsed "2"2"3Ra recoil ion source. The combination of a high density stopping cell with high total efficiencies and a non-scanning high-resolution mass spectrometer can be used as an independent identification detector for exotic nuclei by their mass, allowing a recalibration of the in-flight detectors of any fragment separator. As a proof-of-principal experiment the CSC and a MR-TOF-MS have been used as a mass tagger for the FRS at GSI. 134-I ions were produced by in-flight fission from an "2"3"8U primary beam at 1000 MeV/u and identified by the mass tagger. The new method does not rely on specific decay properties and therefore allows a recalibration of the fragment separator independent of the fragment and can also be used with stable nuclides. The usage of the CSC and a MR-TOF-MS will allow fast recalibration and a more effective usage of the limited amount of beam time for all experiments with exotic nuclei even in the case the nuclide of interest is not clearly identified by the in-flight detection scheme. With the CSC low energy experiments such as high-precision mass measurements and decay spectroscopy were made possible, the half lifes of "2"2"1Ac and "2"2"3Th have been measured, alpha spectroscopy of short

  4. The release code package REVOLS/RENONS for fission product release from a liquid sodium pool into an inert gas atmosphere

    International Nuclear Information System (INIS)

    Starflinger, J.; Scholtyssek, W.; Unger, H.

    1994-12-01

    For aerosol source term considerations in the field of nuclear safety, the investigation of the release of volatile and non-volatile species from liquid surfaces into a gas atmosphere is important. In case of a hypothetical liquid metal fast breeder reactor accident with tank failure, primary coolant sodium with suspended or solved fuel particles and fission products may be released into the containment. The computer code package REVOLS/RENONS, based on a theoretical mechanistic model with a modular structure, has been developed for the prediction of sodium release as well as volatile and non-volatile radionuclide release from a liquid pool surface into the inert gas atmosphere of the inner containment. Hereby the release of sodium and volatile fission products, like cesium and sodium iodide, is calculated using a theoretical model in a mass transfer coefficient formulation. This model has been transposed into the code version REVOLS.MOD1.1, which is discussed here. It enables parameter analysis under highly variable user-defined boundary conditions. Whereas the evaporative release of the volatile components is governed by diffusive and convective transport processes, the release of the non-volatile ones may be governed by mechanical processes which lead to droplet entrainment from the wavy pool surface under conditions of natural or forced convection into the atmosphere. The mechanistic model calculates the liquid entrainment rate of the non-volatile species, like the fission product strontium oxide and the fuel (uranium dioxide) from a liquid pool surface into a parallel gas flow. The mechanistic model has been transposed into the computer code package REVOLS/RENONS, which is discussed here. Hereby the module REVOLS (RElease of VOLatile Species) calculates the evaporative release of the volatile species, while the module RENONS (RElease of NON-Volatile Species) computes the entrainment release of the non-volatile radionuclides. (orig./HP) [de

  5. Fission fragment driven neutron source

    Science.gov (United States)

    Miller, Lowell G.; Young, Robert C.; Brugger, Robert M.

    1976-01-01

    Fissionable uranium formed into a foil is bombarded with thermal neutrons in the presence of deuterium-tritium gas. The resulting fission fragments impart energy to accelerate deuterium and tritium particles which in turn provide approximately 14 MeV neutrons by the reactions t(d,n).sup.4 He and d(t,n).sup.4 He.

  6. Association Euratom - Risoe National Laboratory Annual Progress Report 1998

    International Nuclear Information System (INIS)

    Lynov, J.P.; Singh, B.N.

    1999-08-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics group has activities within development of laser diagnostics for fusion plasmas and studies of nonlinear dynamical processes related to electrostatic turbulence and turbulent transport in magnetised plasmas. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. The technology activities also include contributions to macrotasks, which are carried out under the programme for Socio-Economic Research on Fusion (SERF). A summary is presented of the results obtained in the Research Unit during 1998. (au)

  7. Association Euratom - Risoe National Laboratory Annual Progress Report 1998

    Energy Technology Data Exchange (ETDEWEB)

    Lynov, J.P.; Singh, B.N. [eds.

    1999-08-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics group has activities within development of laser diagnostics for fusion plasmas and studies of nonlinear dynamical processes related to electrostatic turbulence and turbulent transport in magnetised plasmas. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. The technology activities also include contributions to macrotasks, which are carried out under the programme for Socio-Economic Research on Fusion (SERF). A summary is presented of the results obtained in the Research Unit during 1998. (au) 27 ills., 18 refs.

  8. Association Euratom - Risoe National Laboratory annual progress report 1997

    International Nuclear Information System (INIS)

    Lynov, J.P.; Singh, B.N.

    1998-11-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics group has activities within development of laser diagnostics for fusion plasmas and studies of nonlinear dynamical processes related to electrostatic turbulence and turbulent transport in magnetised plasmas. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. The technology activities also include contributions to macrotasks carried out under the programme for Socio-Economic Research on Fusion (SERF). A summary is presented of the results obtained in the Research Unit during 1997. (au)

  9. Association Euratom - Risoe National Laboratory annual progress report 2006

    Energy Technology Data Exchange (ETDEWEB)

    Michelsen, P.K.; Singh, B.N. (eds.)

    2007-09-15

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory, Technical University of Denmark, covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. A summary is presented of the results obtained in the Research Unit during 2006. (au)

  10. Risoe energy report 5. Renewable energy for power and transport

    International Nuclear Information System (INIS)

    Larsen, Hans; Soenderberg Petersen, L.

    2006-11-01

    The global energy policy scene today is dominated by three concerns, namely security of supply, climate change and energy for development and poverty alleviation. This is the starting point for Risoe Energy Report 5 that addresses status and trends in renewable energy, and gives an overview of global driving forces for transformation of the energy systems in the light of security of supply, climate change and economic growth. More specifically status and trends in renewable energy technologies, for broader applications in off grid power production (and heat) will be discussed. Furthermore the report will address wider introduction of renewable energy in the transport sector, for example renewable based fuels, hybrid vehicles, electric vehicles and fuel cell driven vehicles. (au)

  11. Risoe energy report 5. Renewable energy for power and transport

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, Hans; Soenderberg Petersen, L. (eds.)

    2006-11-15

    The global energy policy scene today is dominated by three concerns, namely security of supply, climate change and energy for development and poverty alleviation. This is the starting point for Risoe Energy Report 5 that addresses status and trends in renewable energy, and gives an overview of global driving forces for transformation of the energy systems in the light of security of supply, climate change and economic growth. More specifically status and trends in renewable energy technologies, for broader applications in off grid power production (and heat) will be discussed. Furthermore the report will address wider introduction of renewable energy in the transport sector, for example renewable based fuels, hybrid vehicles, electric vehicles and fuel cell driven vehicles. (au)

  12. Characterisation and classification of RISOe P2546 cup anemometer

    Energy Technology Data Exchange (ETDEWEB)

    Friis Pedersen, T.

    2003-04-01

    The characteristics of the RISOe P2546 cup anemometer were investigated in detail by wind tunnel and laboratory tests. The characteristics include accredited calibration, tilt response measurements for tilt angles between -40 degC to 40 degC, gust response measurements at 8m/s and turbulence intensities of 10%, 16% and 23%, step response measurements at step wind speeds 3,7, 8, 11,9 and 15,2m/s, measurement of torque characteristics at 8m/s, rotor inertia measurements and measurements of friction of bearings at temperatures -20 degC to 40 degC. Characteristics were fitted to a time domain cup anemometer model. The characteristics were transformed into the CLASSCUP classification scheme, and were related to the cup anemometer requirements in the Danish certification system and in the IEC 61400-121 Committee Draft. (au)

  13. Association Euratom - Risoe National Laboratory annual progress report 1997

    Energy Technology Data Exchange (ETDEWEB)

    Lynov, J.P.; Singh, B.N. [eds.

    1998-11-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory covers work in fusion plasma physics and in fusion technology. The fusion plasma physics group has activities within development of laser diagnostics for fusion plasmas and studies of nonlinear dynamical processes related to electrostatic turbulence and turbulent transport in magnetised plasmas. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. The technology activities also include contributions to macrotasks carried out under the programme for Socio-Economic Research on Fusion (SERF). A summary is presented of the results obtained in the Research Unit during 1997. (au) 5 tabs., 30 ills., 12 refs.

  14. Association Euratom - Risoe National Laboratory annual progress report 2006

    International Nuclear Information System (INIS)

    Michelsen, P.K.; Singh, B.N.

    2007-09-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory, Technical University of Denmark, covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. The activities in technology cover investigations of radiation damage of fusion reactor materials. These activities contribute to the Next Step, the Long-term and the Underlying Fusion Technology programme. A summary is presented of the results obtained in the Research Unit during 2006. (au)

  15. Experimental verification of the new RISOe-A1 airfoil family for wind turbines

    Energy Technology Data Exchange (ETDEWEB)

    Dahl, K S; Fuglsang, P; Antoniou, I [Risoe National Lab., Roskilde (Denmark)

    1999-03-01

    This paper concerns the experimental verification of a new airfoil family for wind turbines. The family consist of airfoils in the relative thickness range from 15% to 30%. Three airfoils, Risoe-A1-18, Risoe-A1-21, and Risoe-A1-24 were tested in a wind tunnel. The verification consisted of both static and dynamic measurements. Here, the static results are presented for a Reynolds number of 1.6x10{sup 6} for the following airfoil configurations: smooth surface (all three airfoils) and Risoe-A1-24 mounted with leading edge roughness, vortex generators, and Gurney-flaps, respectively. All three airfoils have constant lift curve slope and almost constant drag coefficient until the maximum lift coefficient of about 1.4 is reached. The experimental results are compared with corresponding computational from the general purpose flow solver, EllipSys2D, showing good agreement. (au)

  16. Process for separation of fission and activation products from a gas atmosphere and device for carrying out the process

    International Nuclear Information System (INIS)

    Schwarzer, K.

    1980-01-01

    Dust particles of concrete, earth, cement or silica gel are blown into the reactor containment to deposit fission and activation products which are situated in the reactor containment of the HTR after a fault. The container for storing the dust is connected via a pneumatic transport pipe through the reinforced concrete wall with the inside of the reactor building for this purpose. (RW) [de

  17. Decommissioning of the nuclear facilities at Risoe National Laboratory. Descriptions and cost assessment. Danish summary

    International Nuclear Information System (INIS)

    Lauridsen, Kurt

    2001-02-01

    The report gives a brief description of relevant aspects of the decommissioning of all nuclear facilities at Risoe National Laboratory, including the necessary operations to be performed and the associated costs. Together with a more detailed report, written in English, this report is the result of a project initiated by Risoe in the summer of 2000. The English report has undergone an international review, the results of which are summarised in the present report. (au)

  18. Process for separation of fission and activation products from a gas atmosphere and device for carrying out the process

    International Nuclear Information System (INIS)

    Schwarzer, K.

    1981-01-01

    Dust particles of concrete, earth, cement or silica gel are blown into the reactor containment to deposit fission and activation products which are situated in the reactor containment of the HTR after a fault. The container for storing the dust is connected via a pneumatic transport pipe through the reinforced concrete wall with the inside of the reactor building for this purpose. (orig./HP)

  19. 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

  20. Risoe energy report 7. Future low carbon energy systems

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, Hans; Soenderberg Petersen, L. (eds.)

    2008-10-15

    This Risoe Energy Report, the seventh of a series that began in 2002, takes as its point of reference the recommendations of the Intergovernmental Panel on Climate Change (IPCC) in 2007. The IPCC states that if anticipated climate change is to remain in the order of 2 to 3 degrees centigrades over the next century, the world's CO{sub 2} emissions would have to peak within the next 10-15 years and ultimately be reduced to approximately 50% of their present level by the middle of the century. The IPCC states further that this would be possible, provided that serious action is taken now. The different regions and countries of the world are in various states of development, and hence have different starting points for contributing to these reductions in CO{sub 2} emissions. This report presents state-of-the-art and development perspectives for energy supply technologies, new energy systems, end-use energy efficiency improvements and new policy measures. It also includes estimates of the CO{sub 2} reduction potentials for different technologies. The technologies are characterized with regard to their ability to contribute either to ensuring a peak in CO{sub 2} emissions within 10-15 years, or to long-term CO{sub 2} reductions. The report outlines the current and likely future composition of energy systems in Denmark, and examines three groups of countries: i) Europe and the other OECD member nations; ii) large and rapidly growing developing economies, notably India and China; iii) typical least developed countries, such as many African nations. The report emphasises how future energy developments and systems might be composed in these three country groupings, and to what extent the different technologies might contribute. The report addresses the need for research and demonstration together with market incentives, and policy measures with focus on initiatives that can promote the development towards CO{sub 2} reductions. Specifically, the report identifies system

  1. 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.

  2. 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.

  3. Mesoscale Benchmark Demonstration Problem 1: Mesoscale Simulations of Intra-granular Fission Gas Bubbles in UO2 under Post-irradiation Thermal Annealing

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert; Gao, Fei; Sun, Xin; Tonks, Michael; Biner, Bullent; Millet, Paul; Tikare, Veena; Radhakrishnan, Balasubramaniam; Andersson , David

    2012-04-11

    A study was conducted to evaluate the capabilities of different numerical methods used to represent microstructure behavior at the mesoscale for irradiated material using an idealized benchmark problem. The purpose of the mesoscale benchmark problem was to provide a common basis to assess several mesoscale methods with the objective of identifying the strengths and areas of improvement in the predictive modeling of microstructure evolution. In this work, mesoscale models (phase-field, Potts, and kinetic Monte Carlo) developed by PNNL, INL, SNL, and ORNL were used to calculate the evolution kinetics of intra-granular fission gas bubbles in UO2 fuel under post-irradiation thermal annealing conditions. The benchmark problem was constructed to include important microstructural evolution mechanisms on the kinetics of intra-granular fission gas bubble behavior such as the atomic diffusion of Xe atoms, U vacancies, and O vacancies, the effect of vacancy capture and emission from defects, and the elastic interaction of non-equilibrium gas bubbles. An idealized set of assumptions was imposed on the benchmark problem to simplify the mechanisms considered. The capability and numerical efficiency of different models are compared against selected experimental and simulation results. These comparisons find that the phase-field methods, by the nature of the free energy formulation, are able to represent a larger subset of the mechanisms influencing the intra-granular bubble growth and coarsening mechanisms in the idealized benchmark problem as compared to the Potts and kinetic Monte Carlo methods. It is recognized that the mesoscale benchmark problem as formulated does not specifically highlight the strengths of the discrete particle modeling used in the Potts and kinetic Monte Carlo methods. Future efforts are recommended to construct increasingly more complex mesoscale benchmark problems to further verify and validate the predictive capabilities of the mesoscale modeling

  4. Measurement of fission gas release, internal pressure and cladding creep rate in the fuel pins of PHWR bundle of normal discharge burnup

    Energy Technology Data Exchange (ETDEWEB)

    Viswanathan, U.K. [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Sah, D.N., E-mail: dnsah@barc.gov.i [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Rath, B.N.; Anantharaman, S. [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2009-08-01

    Fuel pins of a Pressurised Heavy Water Reactor (PHWR) fuel bundle discharged from Narora Atomic Power Station unit no. 1 after attaining a fuel burnup of 7528 MWd/tU have been subjected to two types of studies, namely (i) puncture test to estimate extent of fission gas release and internal pressure in the fuel pin and (ii) localized heating of the irradiated fuel pin to measure the creep rate of the cladding in temperature range 800 deg. C - 900 deg. C. The fission gas release in the fuel pins from the outer ring of the bundle was found to be about 8%. However, only marginal release was found in fuel pins from the middle ring and the central fuel pin. The internal gas pressure in the outer fuel pin was measured to be 0.55 +- 0.05 MPa at room temperature. In-cell isothermal heating of a small portion of the outer fuel pins was carried out at 800 deg. C, 850 deg. C and 900 deg. C for 10 min and the increase in diameter of the fuel pin was measured after heat treatment. Creep rates of the cladding obtained from the measurement of the diameter change of the cladding due to heating at 800 deg. C, 850 deg. C and 900 deg. C were found respectively to be 2.4 x 10{sup -5} s{sup -1}, 24.6 x 10{sup -5} s{sup -1} and 45.6 x 10{sup -5} s{sup -1}.

  5. On the rate determining step in fission gas release from high burn-up water reactor fuel during power transients

    International Nuclear Information System (INIS)

    Walker, C.T.; Mogensen, M.

    1987-01-01

    The radial distribution of grain boundary gas in a PWR and a BWR fuel is reported. The measurements were made using a new approach involving X-ray fluorescence analysis and electron probe microanalysis. In both fuels the concentration of grain boundary gas was much higher than hitherto suspected. The gas was mainly contained in the bubble/pore structure. The factors that determined the fraction of gas released from the grains and the level of gas retention on the grain boundaries are identified and discussed. The variables involved are the local fuel stoichiometry, the amount of open porosity, the magnitude of the local compressive hydrostatic stress and the interaction of metallic precipitates with gas bubbles on the grain faces. It is concluded that under transient conditions the interlinkage of gas bubbles on the grain faces and the subsequent formation of grain edge tunnels is the rate determining step for gas release; at least when high burn-up fuel is involved. (orig.)

  6. Por uma ciência do riso e da sabedoria

    Directory of Open Access Journals (Sweden)

    Paula Corrêa Henning

    2010-06-01

    Full Text Available

    O artigo busca problematizar a história das ciências, especialmente o ethos da Modernidade e Pós-modernidade. Para isso, utiliza autores que se anteciparam à Pós-modernidade como Friedrich Nietzsche e Michel Foucault. Na esteira dessa discussão busca uma problematização acerca da Ciência nos rastros da Modernidade e algumas fissuras e fragilidades produzidas na ciência num cenário contemporâneo. Apresenta ainda alguns discursos midiáticos acerca de propagandas televisivas que trazem o discurso científico como legítimo e inabalável. Traz para o campo de discussão uma ciência alegre, como aprendemos com Nietzsche ou prosaica por vezes, anunciando o riso e a sabedoria na produção do conhecimento científico.

     

  7. Estranhamento e riso no cinema contemporâneo

    Directory of Open Access Journals (Sweden)

    Pablo Augusto Silva

    2010-01-01

    Full Text Available O artigo é uma análise crítica do filme A hora do show (Bamboozled, lançado em 2000, do diretor estadunidense Spike Lee. Há em sua obra, e particularmente nesse filme, influências conceituais de dois importantes autores. De Bertold Brecht e o seu efeito de estranhamento ou efeito V (do alemão Verfremdungseffekt; e de Henri Bergson e sua concepção do riso como portador de determinada função e significado social. A obra de Spike Lee, aberta no sentido modernista, perpassa uma intenção didático-pedagógica do uso e da desconstrução da imagem eurocêntrica que ainda não foi suficientemente analisada e compreendida pela crítica.This article is a critical analysis from Bamboozled film, launched in 2000 by Spike Lee. In his work, particularly in this film, there are conceptual influences from two of the most important authors: Bertold Brecht and his alienation effect or "V-effekt" (from the German Verfremdungseffekt , and Henri Bergson with his conception of laughter and its specific function and social meaning. The Spike Lee's work, opened in the modernist sense, brings a didactic-pedagogic intention about the using and the deconstruction of Eurocentric image - which has not been sufficiently examined and understood by the criticism yet.

  8. Characterisation and classification of RISOe P2546 cup anemometer

    Energy Technology Data Exchange (ETDEWEB)

    Friis Pedersen, T.

    2004-03-01

    The characteristics of the RISOe P2546 cup anemometer were investigated in detail, and all data presented in figures and tables. The characteristics include: wind tunnel calibrations, including an accredited calibration; tilt response meas-urements for tilt angles from -40 deg. C to 40 deg. C; gust response measurements at 8m/s, 10,5m/s and 13m/s and turbulence intensities of 10%, 16% and 23%; step response measurements at step wind speeds 4, 8, 12 and 15m/s; measurement of torque characteristics at 8m/s; rotor inertia measurements and measurements of friction of bearings at temperatures -20 deg. C to 40 deg. C. The characteristics are fitted to a time domain cup anemometer model, and the cup anemometer is put into the CLASSCUP classification scheme. The characteristics are also compared to the requirements to cup anemometers in the Danish wind turbine certification system and the CD and CDV of the revision of the standard IEC 61400-12. (au)

  9. Fission product induced swelling of U–Mo alloy fuel

    International Nuclear Information System (INIS)

    Kim, Yeon Soo; Hofman, G.L.

    2011-01-01

    Highlights: ► We measured fuel swelling of U–Mo alloy by fission products at temperatures below 250 °C. ► We quantified the swelling portion of U–Mo by fission gas bubbles. ► We developed an empirical model as a function of fission density. - Abstract: Fuel swelling of U–Mo alloy was modeled using the measured data from samples irradiated up to a fission density of ∼7 × 10 27 fissions/m 3 at temperatures below ∼250 °C. The overall fuel swelling was measured from U–Mo foils with as-fabricated thickness of 250 μm. Volume fractions occupied by fission gas bubbles were measured and fuel swelling caused by the fission gas bubbles was quantified. The portion of fuel swelling by solid fission products including solid and liquid fission products as well as fission gas atoms not enclosed in the fission gas bubbles is estimated by subtracting the portion of fuel swelling by gas bubbles from the overall fuel swelling. Empirical correlations for overall fuel swelling, swelling by gas bubbles, and swelling by solid fission products were obtained in terms of fission density.

  10. 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)

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

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujita, Misao

    1984-01-01

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

  12. Intercomparison of in vivo monitoring systems in Europe. Results from Risoe National Laboratory

    International Nuclear Information System (INIS)

    Lauridsen, B.; Soegaard-Hansen, J.

    1996-12-01

    This report contains the contribution from Risoe National Laboratory to the European project: 'Intercomparison of in Vivo Monitoring Systems in Europe'. The whole-body counter at Risoe and the measurement on a phantom used as an intercalibration object in the project is described. In four case studies, prepared by the project coordinator, intakes of radionuclides and resulting doses are calculated. These calculations are based on informations on the radioactive materials taken into the body, routes of intake and on body contents of radionuclides from simulated single or multiple whole-body measurement. The answer from Risoe National Laboratory to two questionnaires - one on the whole-body counting facility and calibration methods and one on the legal requirements is the country - is listed. (au)

  13. S’ o riso tocasse meu coração

    Directory of Open Access Journals (Sweden)

    Fernando LIra Ximenes

    2014-10-01

    Full Text Available Este artigo, que está mais para um ensaio de um ensaio, poderia chamar-se: “Quatro propostas para uma próxima gargalhada”. Assim, como nas seis propostas de Italo Calvino que eram cinco em vez de seis, aqui temos apenas três. No entanto, todas as três propostas, na essência, têm o mesmo objetivo: identificar possibilidades de alcançar o risível, através da imaginação e da experiência que possuem a capacidade de nos tocar. Não há aqui a pretensão de apresentar fórmulas. Racionalizar sobre a causa do riso é de fato uma atividade inglória. É como se quiséssemos explicar a piada, perde-se a graça! Deseja-se apenas apontar pistas para um olhar que, às vezes, parece em gestação, mas ainda não reparou que nascera deveras. ABSTRACT This paper, which is more like an essay to an essay, could be called: four proposals for a next laugh. Thus, as in the six proposals of Italo Calvino which were five instead of six, here we have only three. However, all them, in essence, have the same objective: to identify opportunities to achieve the laughable, through imagination and experience that have the ability to touch us. There’s not the pretension to present rules. Rationalize about the cause of laughter is indeed an inglorious activity. As if we wanted to explain the joke, you lose the grace! The wish is just to point out clues for a look that sometimes seems in gestation, but haven’t noticed that indeed was born. KEYWORDS Humour, laughter, imagination.

  14. Risoe energy report 3. Hydrogen and its competitors

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, H; Feidenhans' l, R; Soenderberg Petersen, L [eds.

    2004-10-01

    Interest in the hydrogen economy has grown rapidly in recent years. Countries with long traditions of activity in hydrogen research and development have now been joined by a large number of newcomers. The main reason for this surge of interest is that the hydrogen economy may be an answer to the two main challenges facing the world in the years to come: climate change and the need for security of energy supplies. Both these challenges require the development of new, highly-efficient energy technologies that are either carbon-neutral or low emitting technologies. Another reason for the growing interest in hydrogen is the strong need for alternative fuels, especially in the transport sector. Alternative fuels could serve as links between the power system and the transport sector, to facilitate the uptake of emerging technologies and increase the flexibility and robustness of the energy system as a whole. This Risoe Energy Report provides a perspective on energy issues at global, regional and national levels. The following pages provide a critical examination of the hydrogen economy and its alternatives. The report explains the current R and D situation addresses the challenges facing the large-scale use of hydrogen, and makes some predictions for the future. The current and future role of hydrogen in energy systems is explored at Danish, European and global levels. The report discusses the technologies for producing, storing and converting hydrogen, the role of hydrogen in the transport sector and in portable electronics, hydrogen infrastructure and distribution systems, and environmental and safety aspects of the hydrogen economy. (BA)

  15. The Risoe model for calculating the consequences of the release of radioactive material to the atmosphere

    International Nuclear Information System (INIS)

    Thykier-Nielsen, S.

    1980-07-01

    A brief description is given of the model used at Risoe for calculating the consequences of releases of radioactive material to the atmosphere. The model is based on the Gaussian plume model, and it provides possibilities for calculation of: doses to individuals, collective doses, contamination of the ground, probability distribution of doses, and the consequences of doses for give dose-risk relationships. The model is implemented as a computer program PLUCON2, written in ALGOL for the Burroughs B6700 computer at Risoe. A short description of PLUCON2 is given. (author)

  16. Ternary fission

    Indian Academy of Sciences (India)

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

  17. 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)

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

    International Nuclear Information System (INIS)

    Mogensen, M.; Walker, C.T.

    1999-01-01

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

  19. Fission-product retention in HTGR fuels

    International Nuclear Information System (INIS)

    Homan, F.J.; Kania, M.J.; Tiegs, T.N.

    1982-01-01

    Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed

  20. 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

  1. Pilot plant production at Riso of LEU silicide fuel for the Danish reactor DR3

    International Nuclear Information System (INIS)

    Toft, P.; Borring, J.; Adolph, E.

    1988-01-01

    A pilot plant for fabricating LEU silicide fuel elements has been established at Riso National Laboratory. Three test elements for the Danish reactor DR3 have been fabricated, based on 19.88% enriched U 3 Si 2 powder that has been purchased elsewhere. The pilot plant has been set up and 3 test elements fabricated without any major difficulties

  2. Fission meter

    Science.gov (United States)

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

    2012-04-10

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

  3. Chemical Production using Fission Fragments

    International Nuclear Information System (INIS)

    Dawson, J. K.; Moseley, F.

    1960-01-01

    Some reactor design considerations of the use of fission recoil fragment energy for the production of chemicals of industrial importance have been discussed previously in a paper given at the Second United Nations International Conference on the Peaceful Uses of Atomic Energy [A/Conf. 15/P.76]. The present paper summarizes more recent progress made on this topic at AERE, Harwell. The range-energy relationship for fission fragments is discussed in the context of the choice of fuel system for a chemical production reactor, and the experimental observation of a variation of chemical effect along the length of a fission fragment track is described for the irradiation of nitrogen-oxygen mixtures. Recent results are given on the effect of fission fragments on carbon monoxide-hydrogen gas mixtures and on water vapour. No system investigated to date shows any outstanding promise for large-scale chemical production. (author) [fr

  4. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  5. Total surface area change of Uranium dioxide fuel in function of burn-up and its impact on fission gas release during neutron irradiation for small, intermediate and high burn-up

    International Nuclear Information System (INIS)

    Szuta, M.

    2011-01-01

    In the early published papers it was observed that the fractional fission gas release from the specimen have a tendency to increase with the total surface area of the specimen - a fairy linear relationship was indicated. Moreover it was observed that the increase of total surface area during irradiation occurs in the result of connection the closed porosity with the open porosity what in turn causes the increase of fission gas release. These observations let us surmise that the process of knock-out release is the most significant process of fission gas release since its quantity is proportional to the total surface area. Review of the experiments related to the increase of total surface area in function of burn-up is presented in the paper. For very high burn-up the process of grain sub-division (polygonization) occurs under condition that the temperature of irradiated fuel lies below the temperature of grain re-crystallization. Simultaneously with the process of polygonization, the increase in local porosity and the decrease in local density in function of burn-up occurs, which leads to the increase of total surface area. It is suggested that the same processes take place in the transformed fuel as in the original fuel, with the difference that the total surface area is so big that the whole fuel can be treated as that affected by the knock-out process. This leads to explanation of the experimental data that for very high burn-up (>120 MWd/kgU) the concentration of xenon is constant. An explanation of the grain subdivision process in function of burn-up in the 'athermal' rim region in terms of total surface area, initial grain size and knock-out release is undertaken. Correlation of the threshold burn-up, the local fission gas concentration, local total surface area, initial and local grain size and burn-up in the rim region is expected. (author)

  6. Ideological Fission

    DEFF Research Database (Denmark)

    Christiansen, Steen Ledet

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

  7. Decommissioning of the nuclear facilities at Risoe National Laboratory. Descriptions and cost assessment. Danish summary[Denmark]; Dekommissionering af Risoes nukleare anlaeg - vurdering af opgaver og omkostninger. Dansk sammenfatning

    Energy Technology Data Exchange (ETDEWEB)

    Lauridsen, Kurt

    2001-02-01

    The report gives a brief description of relevant aspects of the decommissioning of all nuclear facilities at Risoe National Laboratory, including the necessary operations to be performed and the associated costs. Together with a more detailed report, written in English, this report is the result of a project initiated by Risoe in the summer of 2000. The English report has undergone an international review, the results of which are summarised in the present report. (au)

  8. 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

  9. The contribution of Risoe National Laboratory to the research and development programs of the Danish Ministry of Energy

    International Nuclear Information System (INIS)

    Skjerk Christensen, P.; Brown Joergensen, B.

    1990-07-01

    Since 1978 Risoe has been responsible for a number of projects in the research and development programs of the Danish Ministry of Energy. This report gives a review of current and finished projects. All current projects are described briefly, stating status and results obtained, while the results of finished projects are described in more detail. Risoe's contribution to the organization and the administraton of the programs is mentioned. Finally a list of references is given. (author) 3 tabs., 24 ills.; 45 refs

  10. The contribution of Risoe National Laboratory to the research development programs of the Danish Ministry of Energy

    International Nuclear Information System (INIS)

    Skjerk Christensen, P.

    1986-05-01

    Since 1978 Risoe has been responsible for a number of projects in the research and development programs of the Danish Ministry of Energy. This report gives a review of current abd finished projects. All current projects are described briefly, stating status and results obtained, whole the results of finished projects are described in more detail. Risoe's contribution of the organization and the administration of the programs is mentioned. Finally, a list of references is given. (Author)

  11. The contribution of Risoe National Laboratory to the research and development programs of the Danish Ministry of Energy

    International Nuclear Information System (INIS)

    Skjerk Christensen, P.; Petersen, S.

    1988-06-01

    Since 1978 Risoe has been responsible for a number of projects in the research and development programs of the Danish Ministry of Energy. This report gives a review of current and finished projects. All current projects are described briefly, stating status and results obtained, while the results of finished projects are described in more detail. Risoe's contribution to the organization and the administration of the programs is mentioned. Finally a list of references is given. 11 ills., 34 refs. (author)

  12. The contribution of Risoe National Laboratory to the research development programs of the Danish Ministry of Energy

    International Nuclear Information System (INIS)

    1985-07-01

    Since 1978 Risoe has been responsible for a number of projects in the research and development programs of the Danish Ministry of Energy. This report gives a review of current and finished projects. All current projects are described briefly, stating status and results obtained, while the results of finished projects are described in more detail. Risoe's contribution of the organization and the administration of the programs is mentioned. Finally a list of references is given. (author)

  13. The contribution of Risoe National Laboratory to the research and development programs of the Danish Ministry of Energy

    International Nuclear Information System (INIS)

    Skjerk Christensen, P.; Petersen, S.

    1989-04-01

    Since 1978 Risoe has been responsible for a number of projects in the research and development programs of the Danish Ministry of Energy. This report gives a review of current and finished projects. All current projects are described briefly, stating status and results obtained, while the results of finished projects are described in more detail. Risoe's contribution to the organization and the administration of the programs is mentioned. Finally a list of references is given. (author) 4 tabs., 22 ills., 33 refs

  14. Diffusion of gases in solids: rare gas diffusion in solids; tritium diffusion in fission and fusion reactor metals. Final report

    International Nuclear Information System (INIS)

    Abraham, P.M.; Chandra, D.; Mintz, J.M.; Elleman, T.S.; Verghese, K.

    1976-01-01

    Major results of tritium and rare gas diffusion research conducted under the contract are summarized. The materials studied were austenitic stainless steels, Zircaloy, and niobium. In all three of the metal systems investigated, tritium release rates were found to be inhibited by surface oxide films. The effective diffusion coefficients that control tritium release from surface films on Zircaloy and niobium were determined to be eight to ten orders of magnitude lower than the bulk diffusion coefficients. A rapid component of diffusion due to grain boundaries was identified in stainless steels. The grain boundary diffusion coefficient was determined to be about six orders of magnitude greater than the bulk diffusion coefficient for tritium in stainless steel. In Zircaloy clad fuel pins, the permeation rate of tritium through the cladding is rate-limited by the extremely slow diffusion rate in the surface films. Tritium diffusion rates through surface oxide films on niobium appear to be controlled by cracks in the surface films at temperatures up to 600 0 C. Beyond 600 0 C, the cracks appear to heal, thereby increasing the activation energy for diffusion through the oxide film. The steady-state diffusion of tritium in a fusion reactor blanket has been evaluated in order to calculate the equilibrium tritium transport rate, approximate time to equilibrium, and tritium inventory in various regions of the reactor blanket as a function of selected blanket parameters. Values for these quantities have been tabulated

  15. TEM and XAS investigation of fission gas behaviors in U-Mo alloy fuels through ion beam irradiation

    Science.gov (United States)

    Zang, Hang; Yun, Di; Mo, Kun; Wang, Kunpeng; Mohamed, Walid; Kirk, Marquis A.; Velázquez, Daniel; Seibert, Rachel; Logan, Kevin; Terry, Jeffrey; Baldo, Peter; Yacout, Abdellatif M.; Liu, Wenbo; Zhang, Bo; Gao, Yedong; Du, Yang; Liu, Jing

    2017-10-01

    In this study, smaller-grained (hundred nano-meter size grain) and larger-grained (micro-meter size grain) U-10Mo specimens have been irradiated (implanted) with 250 keV Xe+ beam and were in situ characterized by TEM. Xe bubbles were not seen in the specimen after an implantation fluence of 2 × 1020 ions/m2 at room temperature. Nucleation of Xe bubbles happened during heating of the specimen to a final temperature of 300 °C. By comparing measured Xe bubble statistics, the nucleation and growth behaviors of Xe bubbles were investigated in smaller-grained and larger-grained U-10Mo specimens. A multi-atom kind of nucleation mechanism has been observed in both specimens. X-ray Absorption spectroscopy showed the edge position in the bubbles to be the same as that of Xe gas. The size of Xe bubbles has been shown to be bigger in larger-grained specimens than in smaller-grained specimens at the same implantation conditions.

  16. Decommissioning of the nuclear facilities at Risoe National Laboratory. Descriptions and cost assessment

    International Nuclear Information System (INIS)

    Lauridsen, Kurt

    2001-02-01

    The report is the result of a project initiated by Risoe National Laboratory in June 2000 on request from the Minister of Research and Information Technology. It describes the nuclear facilities at Risoe National Laboratory to be decommissioned and gives an assessment of the work to be done and the costs incurred. Three decommissioning scenarios were considered with decay times of 10, 25 and 40 years for the DR 3 reactor. The assessments conclude, however, that there will not be much to gain by allowing for the longer decay periods; some operations still will need to be performed remotely. Furthermore, the report describes some of the legal and licensing framework for the decommissioning and gives an assessment of the amounts of radioactive waste to be transferred to a Danish repository. (au)

  17. Decommissioning of the nuclear facilities at Risoe National Laboratory. Descriptions and cost assessment[Denmark

    Energy Technology Data Exchange (ETDEWEB)

    Lauridsen, Kurt [ed.

    2001-02-01

    The report is the result of a project initiated by Risoe National Laboratory in June 2000 on request from the Minister of Research and Information Technology. It describes the nuclear facilities at Risoe National Laboratory to be decommissioned and gives an assessment of the work to be done and the costs incurred. Three decommissioning scenarios were considered with decay times of 10, 25 and 40 years for the DR 3 reactor. The assessments conclude, however, that there will not be much to gain by allowing for the longer decay periods; some operations still will need to be performed remotely. Furthermore, the report describes some of the legal and licensing framework for the decommissioning and gives an assessment of the amounts of radioactive waste to be transferred to a Danish repository. (au)

  18. 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.

  19. Main results from Risoe's wind-diesel programme 1984-1990

    International Nuclear Information System (INIS)

    Lundsager, P.; Christensen, C.J.

    1991-12-01

    The report presents the results of the wind-diesel work done in projects at Risoe National Laboratory during the years 1984-90, including important earlier publications as appendices. The partners in the original joint project were Risoe National Laboratory, Denmark, and Chalmers University of Technology, Sweden. Chalmers has constructed and laboratory tested an advanced wind-diesel-battery system with variable speed operation of the wind turbine, while Risoe has established a flexible and versatile wind-diesel test facility and field tested Chalmers system. As part of the subsequent EFP projects Risoe designed and constructed a simple wind-diesel system without storage, characterized by several innovative features. This concept was part of a ''simple wind-diesel systems strategy'', in which immediate cost-effectiveness is ensured by the simplicity and reliability of the design. Dynamic computer models were developed for system design and analysis purposes, and a general logistic computer model was developed for the determination of fuel savings and power supply capabilities for a number of system configurations. In addition to a considerable body of experience the main results of activities are: A versatile wind-diesel test facility and a proposed standard wind-diesel test procedure. Two wind-diesel systems at each end of the spectrum of configurations. Computer models for logistic and dynamic modelling. The two systems represent the very simple system concept, believed to be a presently economically optimal configuration, and the very sophisticated concept believed to be a future optimal configuration. (au) (6 tabs., 67 ills., 25 refs.)

  20. 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)

  1. 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.)

  2. Advances on fission chamber modelling

    International Nuclear Information System (INIS)

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

    2013-06-01

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

  3. 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)

  4. 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)

  5. Nuclear fission studies: from LOHENGRIN to FIPPS

    International Nuclear Information System (INIS)

    Chebboubi, Abdelaziz

    2015-01-01

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

  6. 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)

  7. 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.)

  8. 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.

  9. Risoe energy report 1. New and emerging technologies - options for the future

    International Nuclear Information System (INIS)

    Larsen, H.; Soenderberg Petersen, L.

    2002-10-01

    All over the world, increasing energy consumption, liberalisation of energy markets and the need to take action on climate change are producing new challenges for the energy sector. At the same time there is increasing pressure for research, new technology and industrial products to be socially acceptable and to generate prosperity. The result is a complex and dynamic set of conditions affecting decisions on investment in research and new energy technology. To meet these challenges in the decades ahead, industrialists and policymakers need appropriate analyse energy systems, plus knowledge of trends for existing technologies and prospects for emerging technologies. This is the background for this first Risoe Energy Report, which sets out the global, European and Danish energy scene together with trends in development and emerging technologies. The report is the first in a new series from Risoe National Laboratory. The global energy developments are presented based on the latest available information from authoritative sources like IEA, WEC, World Energy Assessment etc. Some of the major challenges are presented in terms of the changing energy markets in all regions, the focus on environmental concerns in the industrialised countries, and energy for development and access to energy for the poor in developing countries. The report presents the status of R and D in progress for supply technologies. The various technologies are assessed with respect to status, trends and perspectives for the technology, and international R and D plans. For the technologies where Risoe is undertaking R and D this is highlighted in a separate section. Recent studies of emerging energy technologies from international organisations and leading research organisations are reviewed. There are reviews of national research activities on new energy technologies in a number of countries as well as in Risoe National Laboratory. Conclusions for Danish energy supply, Danish industry, and Danish

  10. 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.

  11. Fission gas pressure build-up and fast-breeder economy; Accumulation de la pression des gaz de fission et economie des reacteurs surgenerateurs a neutrons rapides; Nakoplenie davleniya gazov produktov deleniya i ehkonomika reaktorov-razmnozhitelej na bystrykh nejtronakh; Aumento de la presion de los gases de fision y economia de los reactores reproductores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Engelmann, P [Kernforschungszentrum, Karlsruhe (Germany)

    1962-03-15

    Fuel-cycle costs and doubling time of fast-breeder reactors are strongly affected by the fuel-burn-up obtainable. Use of oxide or carbide fuel offers the possibility of reaching a burn-up of 100 000 MWd/t. In fuel-clad elements, a limiting factor is the fission-gas-pressure build-up. At the high burn-up considered, an appreciable fraction of the fission gases gets into the pores and thus contributes to the pressure on the can. Starting from the known fission-product yields and decay chains, gas production and pressure build-up have been calculated. Three physical models have been employed in calculating the pressure acting upon the can : the gas is contained either in interconnected pores, in separate pores, or in a central hole. The pressure-dependence upon free volume (fuel density) and temperature will be discussed. Cans made of high-strength materials as Ineonel-X and molybdenum could stand the fission-gas pressure at operating temperatures. Unfortunately, these materials have higher absorption cross-sections than stainless steel. Results of a multi-group calculation are given, showing the effect of using these can materials and of decreasing the fuel density on critical mass and breeding ratio in small and medium-size breeders. (author) [French] Le cout du cycle de combustible et la periode de doublement des reacteurs surgenerateurs a neutrons rapides dependent etroitement du taux de combustion. En utilisant pour combustible un oxyde ou un carbure, on peut atteindre un taux de combustion de 100 000 MW j/t. Avec des combustibles gaines, l'accumulation de la pression des gaz de fission est un facteur limitatif. Pour le fort taux de combustion envisage, une fraction non negligeable des gaz de fission penetre dans les interstices et contribue ainsi a la pression sur la gaine. A partir des rendements en produits de fission et des chaines de desintegration connus, l'auteur a calcule la production de gaz et l'accumulation de pression. Pour calculer la pression

  12. 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

  13. Fission fragment excited laser system

    Science.gov (United States)

    McArthur, David A.; Tollefsrud, Philip B.

    1976-01-01

    A laser system and method for exciting lasing action in a molecular gas lasing medium which includes cooling the lasing medium to a temperature below about 150 K and injecting fission fragments through the lasing medium so as to preferentially excite low lying vibrational levels of the medium and to cause population inversions therein. The cooled gas lasing medium should have a mass areal density of about 5 .times. 10.sup.-.sup.3 grams/square centimeter, relaxation times of greater than 50 microseconds, and a broad range of excitable vibrational levels which are excitable by molecular collisions.

  14. Risoe energy report 9. Non-fossil energy technologies in 2050 and beyond

    International Nuclear Information System (INIS)

    Larsen, Hans; Soenderberg Petersen, L.

    2010-11-01

    This Risoe Energy Report, the ninth in a series that began in 2002, analyses the long-term outlook for energy technologies in 2050 in a perspective where the dominating role of fossil fuels has been taken over by non-fossil fuels, and CO 2 emissions have been reduced to a minimum. Against this background, the report addresses issues like: 1) How much will today's non-fossil energy technologies have evolved up to 2050? 2) Which non-fossil energy technologies can we bring into play in 2050, including emerging technologies? 3) What are the implications for the energy system? Further, Volume 9 analyses other central issues for the future energy supply: 4) The role of non-fossil energy technologies in relation to security of supply and sustainability 5) System aspects in 2050 6) Examples of global and Danish energy scenarios in 2050 The report is based on the latest research results from Risoe DTU, together with available international literature and reports. (Author)

  15. Risoe energy report 9. Non-fossil energy technologies in 2050 and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, Hans; Soenderberg Petersen, L. (eds.)

    2010-11-15

    This Risoe Energy Report, the ninth in a series that began in 2002, analyses the long-term outlook for energy technologies in 2050 in a perspective where the dominating role of fossil fuels has been taken over by non-fossil fuels, and CO{sub 2} emissions have been reduced to a minimum. Against this background, the report addresses issues like: 1) How much will today's non-fossil energy technologies have evolved up to 2050? 2) Which non-fossil energy technologies can we bring into play in 2050, including emerging technologies? 3) What are the implications for the energy system? Further, Volume 9 analyses other central issues for the future energy supply: 4) The role of non-fossil energy technologies in relation to security of supply and sustainability 5) System aspects in 2050 6) Examples of global and Danish energy scenarios in 2050 The report is based on the latest research results from Risoe DTU, together with available international literature and reports. (Author)

  16. Heat treatments of irradiated uranium oxide in a pressurised water reactor (P.W.R.): swelling and fission gas release; Traitements thermiques de l`oxyde d`uranium irradie dans un reacteur a eau pressurisee (R.E.P.): gonflement et relachement des gaz de fission

    Energy Technology Data Exchange (ETDEWEB)

    Zacharie, I

    1997-03-27

    In order to keep pressurised water reactors at a top level of safety, it is necessary to understand the chemical and mechanical interaction between the cladding and the fuel pellet due to a temperature increase during a rapid change in reactor. In this process, the swelling of uranium oxide plays an important role. It comes from a bubble precipitation of fission gases which are released when they are in contact with the outside. Therefore, the aim of this thesis consists in acquiring a better understanding of the mechanisms which come into play. Uranium oxide samples, from a two cycles irradiated fuel, first have been thermal treated between 1000 deg C and 1700 deg C for 5 minutes to ten hours. The gas release amount related to time has been measured for each treatment. The comparison of the experimental results with a numerical model has proved satisfactory: it seems that the gases release, after the formation of intergranular tunnels, is controlled by the diffusion phenomena. Afterwards, the swelling was measured on the samples. The microscopic examination shows that the bubbles are located in the grain boundaries and have a lenticular shape. The swelling can be explained by the bubbles coalescence and a model was developed based on this observation. An equation allows to calculate the intergranular swelling in function of time and temperature. The study gives the opportunity to predict the fission gases behaviour during a fuel temperature increase. (author) 56 refs.

  17. Status of fission power

    International Nuclear Information System (INIS)

    Levenson, M.

    1977-01-01

    Fission energy is reviewed from the viewpoints of technology, economics, politics, manufacturers, consumers, and foreign countries. Technically, the reactor program is operating and the light water reactor industry shows signs of maturing, although recent business has been disappointing. Marketing of gas-cooled reactors depends, not on technical, but economic and political issues. Liquid metal fast breeder reactors have been demonstrated worldwide, while the gas-cooled fast breeder remains an undemonstrated option. Nuclear plants, currently costing the same as coal plants with scrubbers, are the cheapest option for utilities because most of the cost is imbedded. The defeat of nuclear initiatives in seven states indicates that public feeling is not as anti-nuclear as opponents to nuclear power claim. The harshness of last winter demonstrated the advantages of a power source that is not so sensitive to the weather for reliable operation and transport, as well as low cost energy. Other nations are proceeding to build a nuclear capability, which the U.S. may jeopardize because of concerns about the fuel cycle, nuclear waste disposal, uranium reserves, and nuclear proliferation

  18. IFPE/IFA-597.3, centre-line temperature, fission gas release and clad elongation at high burn-up (60-62 MWd/kg)

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    2003-01-01

    grain boundary porosity in the central regions of the pellet. The back scattered electron image of the pellet periphery showed an area of high porosity as would be associated with re-structuring caused by formation of the Hobs. The depth of this region was typically 200 microns with traces of Hobs at greater depths. After transfer to Kjeller hot laboratories in Norway, three segments were cut from rod D5 33-25065. All were from regions of uniform burn-up as measured by axial gamma scanning. The sections were identified as D5:1, rod 7, D5:2, rod 8 and D5:3, rod 9. Whilst drilling rod 7 to insert a thermocouple, the drill broke and thermocouple insertion was abandoned. The rod was finally sealed with a clad elongation detector (EC). Rods 8 and 9 were successfully drilled and thermocouples inserted. They were then sealed with pressure transducers (PF). Rods 8 and 9 were loaded into positions 2 and 5 in IFA-597.2 (second loading) and irradiated in Halden for some 20 days in July 1995. After a small number of power ramps, rod 9 failed and the assembly withdrawn. During this time, useful data were generated on centreline temperature as a function of power. Rod 9 was removed and replaced by rod 7. The assembly was returned to the reactor as IFA-597.3 (third loading); the irradiation started in January 1997 and continued to May of that year having accrued a further ∼2 MWd/kg UO 2 . Data obtained included, centreline temperature as a function of power and burn-up, (rod 8), FGR from the increase in rod internal pressure (rod 8) and clad elongation (rod 7). The assembly was discharged and transported to Kjeller for PIE. FGR of 12.6% and 15.8% were measured from puncturing and gas extraction from rods 7 and 8 respectively. Further PIE was carried out on rod 8 after it had been transported to Studsvik. Here, additional ceramography was performed, axial gamma scans, radial scans of selected fission products and the radial porosity distribution. On account of the high burn-up of

  19. Shipments of irradiated DIDO fuel from Risoe National Laboratory to the Savannah River Site - Challenges and achievements

    International Nuclear Information System (INIS)

    Anne, C.; Patterson, J.

    2003-01-01

    On September 28, 2000, the Board of Governors of Risoe National Laboratory decided to shut down the Danish research reactor DR3 due to technical problems (corrosion on the reactor aluminum tank). Shortly thereafter, the Danish Government asked the National Laboratory to empty the reactor and its storage pools containing a total of 255 DIDO irradiated elements and ship them to Savannah River Site in the USA as soon as possible. Risoe National Laboratory had previously contracted with Cogema Logistics to ship DR3 DIDO fuel elements to SRS through the end of the return program. The quantity of fuel was less than originally intended but the schedule was significantly shorter. It was agreed in June 2001 that a combination of Cogema Logistics' and NAC casks would be preferable, as it would allow Risoe to ship all the irradiated fuel in two shipments and complete the shipments by June 2002. Risoe National Laboratory, Cogema Logistics and NAC International had twelve months to perform the shipments including licensing, basket fabrication for the NAC-LWT casks and actual transport. The paper describes the challenging work that was accomplished to meet the date of June 2002. (author)

  20. Fission neutron spectra measurements at LANSCE - Status and plans

    International Nuclear Information System (INIS)

    Haight, R. C.; Noda, S.; Nelson, R. O.; O'Donnell, J. M.; Devlin, M.; Chatillon, A.; Granier, T.; Taiebb, J.; Laurent, B.; Belier, G.; Becker, J. A.; Wu, C. Y.

    2010-01-01

    A program to measure fission neutron spectra from neutron-induced fission of actinides is underway at the Los Alamos Neutron Science Center (LANSCE) in a collaboration among the CEA laboratory at Bruyeres-le-Chatel, Lawrence Livermore National Laboratory and Los Alamos National Laboratory. The spallation source of fast neutrons at LANSCE is used to provide incident neutron energies from less than 1 MeV to 100 MeV or higher. The fission events take place in a gas-ionization fission chamber, and the time of flight from the neutron source to that chamber gives the energy of the incident neutron. Outgoing neutrons are detected by an array of organic liquid scintillator neutron detectors, and their energies are deduced from the time of flight from the fission chamber to the neutron detector. Measurements have been made of the fission neutrons from fission of 235 U, 238 U, 237 Np and 239 Pu. The range of outgoing energies measured so far is from 0.7 MeV to approximately 8 MeV. These partial spectra and average fission neutron energies are compared with evaluated data and with models of fission neutron emission. Results to date are summarized in this presentation. Future plans are to make significant improvements in the fission chambers, neutron detectors, signal processing, data acquisition and the experimental environment to provide high fidelity data including measurements of fission neutrons below 0.7 MeV and improvements in the data above 8 MeV. (authors)

  1. Determination of Fission Gas Inclusion Pressures in High Burnup Nuclear Fuel using Laser Ablation ICP-MS combined with SEM/EPMA and Optical Microscopy

    International Nuclear Information System (INIS)

    Horvath, Matthias I.; Guenther-Leopold, Ines; Kivel, Niko; Restani, Renato; Guillong, Marcel; Izmer, Andrei; Hellwig, Christian; Guenther, Detlef

    2008-01-01

    In approximately 20% of all fissions at least one of the fission products is gaseous. These are mainly xenon and krypton isotopes contributing up to 90% by the xenon isotopes. Upon reaching a burn-up of 60 - 75 GWd/tHM a so called High Burnup Structure (HBS) is formed in the cooler rim of the fuel. In this region a depletion of the noble fission gases (FG) in the matrix and an enrichment of FG in μm-sized pores can be observed. Recent calculations show that in these pores the pressure at room temperature can be as large as 30 MPa. The knowledge of the FG pressure in pores is important to understand the high burn-up fuel behavior under accident conditions (i.e. RIA or LOCA). With analytical methods routinely used for the characterization of solid samples, i.e. Electron Probe Micro Analysis (EPMA), Secondary Ion Mass Spectrometry (SIMS), the quantification of gaseous inclusions is very difficult to almost impossible. The combination of a laser ablation system (LA) with an inductively coupled plasma mass spectrometer (ICP-MS) offers a powerful tool for quantification of the gaseous pore inventory. This method offers the advantages of high spatial resolution with laser spot sizes down to 10 μm and low detection limits. By coupling with scanning electron microscopy (SEM) for the pore size distribution, EPMA for the FG inventory in the fuel matrix and optical microscopy for the LA-crater sizes, the pressures in the pores and porosity was calculated. As a first application of this calibration technique for gases, measurements were performed on pressurized water reactor (PWR) fuel with a rod average of 105 GWd/tHM to determine the local FG pressure distribution. (authors)

  2. Birth to death analysis of the energy payback ratio and CO2 gas emission rates from coal, fission, wind, and DT-fusion electrical power plants

    International Nuclear Information System (INIS)

    White, Scott W.; Kulcinski, Gerald L.

    2000-01-01

    The amount of electrical energy produced over the lifetime of coal, LWR fission, UP fusion, and wind power plants is compared to the total amount of energy required to procure the fuel, build, operate, and decommission the power plants. The energy payback ratio varies from a low of 11 for coal plants to a high of 27 for DT-fusion plants. The magnitude of the energy investment and the source of the various energy inputs determine the CO 2 emission factor. This number varies from a low of 9 to a high of 974 tonnes of CO 2 per GW e h for DT-fusion and coal plants, respectively

  3. 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

  4. 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)

  5. 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

  6. 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)

  7. Association Euratom - Risoe National Laboratory for Sustainable Energy, Technical University of Denmark. Annual progress report 2010

    Energy Technology Data Exchange (ETDEWEB)

    Korsholm, S.B.; Michelsen, P.K.; Rasmussen, J.J.; Westergaard, C.M. (eds.)

    2011-04-15

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory for Sustainable Energy, Technical University of Denmark, covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. Within fusion technology there are activities related to development of high temperature superconductors. Other activities are system analysis, initiative to involve Danish industry in ITER contracts and public information. A summary is presented of the results obtained in the Research Unit during 2010. (Author)

  8. Association Euratom - Risoe National Laboratory for Sustainable Energy, Technical University of Denmark. Annual progress report 2009

    International Nuclear Information System (INIS)

    Korsholm, S.B.; Michelsen, P.K.; Rasmussen, J.J.; Westergaard, C.M.

    2010-04-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory for Sustainable Energy, Technical University of Denmark, covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. Within fusion technology there are activities related to development of high temperature superconductors. Minor activities are system analysis, initiative to involve Danish industry in ITER contracts and public information. A summary is presented of the results obtained in the Research Unit during 2009. (Author)

  9. Association Euratom - Risoe National Laboratory for Sustainable Energy, Technical University of Denmark. Annual progress report 2009

    Energy Technology Data Exchange (ETDEWEB)

    Korsholm, S B; Michelsen, P K; Rasmussen, J J; Westergaard, C M [eds.

    2010-04-15

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory for Sustainable Energy, Technical University of Denmark, covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. Within fusion technology there are activities related to development of high temperature superconductors. Minor activities are system analysis, initiative to involve Danish industry in ITER contracts and public information. A summary is presented of the results obtained in the Research Unit during 2009. (Author)

  10. Association Euratom - Risoe National Laboratory for Sustainable Energy, Technical University of Denmark. Annual progress report 2008

    Energy Technology Data Exchange (ETDEWEB)

    Korsholm, S.B.; Michelsen, P.K.; Rasmussen, J.J.; Westergaard, C.M. (eds.)

    2009-04-15

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory for Sustainable Energy, Technical University of Denmark, covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. New activities in technology related to development of high temperature superconductors have been initiated in 2008. Minor activities are system analysis, initiative to involve Danish industry in ITER contracts and public information. A summary is presented of the results obtained in the Research Unit during 2008. (Author)

  11. Association Euratom - Risoe National Laboratory for Sustainable Energy, Technical University of Denmark. Annual progress report 2008

    International Nuclear Information System (INIS)

    Korsholm, S.B.; Michelsen, P.K.; Rasmussen, J.J.; Westergaard, C.M.

    2009-04-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory for Sustainable Energy, Technical University of Denmark, covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. New activities in technology related to development of high temperature superconductors have been initiated in 2008. Minor activities are system analysis, initiative to involve Danish industry in ITER contracts and public information. A summary is presented of the results obtained in the Research Unit during 2008. (Author)

  12. Association Euratom - Risoe National Laboratory, Technical Univ. of Denmark. Annual progress report 2007

    International Nuclear Information System (INIS)

    Michelsen, P.K.; Korsholm, S.B.; Rasmussen, J.J.

    2008-04-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory, Technical University of Denmark, covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. The activities in technology on investigations of radiation damage of fusion reactor materials have been phased out during 2007. Minor activities are system analysis, initiative to involve Danish industry in ITER contracts and public information. A summary is presented of the results obtained in the Research Unit during 2007. (Author)

  13. Geological characterisation of potential disposal areas for radioactive waste from Risoe, Denmark

    International Nuclear Information System (INIS)

    Gravesen, P.; Binderup, M.; Nilsson, B.; Schack Pedersen, S.A.

    2011-01-01

    Low- and intermediate-level radioactive waste from the Danish nuclear research facility, Risoe, includes construction materials from the reactors, different types of contaminated material from the research projects and radioactive waste from hospitals, industry and research institutes. This material must be stored in a permanent disposal site in Denmark for at least 300 years. The latter study was conducted by the Geological Survey of Denmark and Greenland (GEUS) and the aim was to locate a sediment or rock body with low permeability down to 100-300 m below the ground surface. GEUS was given the task to locate approximately 20 potential disposal areas. The survey resulted in the selection of 22 areas throughout Denmark. Six of these areas are preferred on geological and hydrogeological criteria. (LN)

  14. Association Euratom - Risoe National Laboratory, Technical Univ. of Denmark. Annual progress report 2007

    Energy Technology Data Exchange (ETDEWEB)

    Michelsen, P.K.; Korsholm, S.B.; Rasmussen, J.J. (eds.)

    2008-04-15

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory, Technical University of Denmark, covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. The activities in technology on investigations of radiation damage of fusion reactor materials have been phased out during 2007. Minor activities are system analysis, initiative to involve Danish industry in ITER contracts and public information. A summary is presented of the results obtained in the Research Unit during 2007. (Author)

  15. Association Euratom - Risoe National Laboratory for Sustainable Energy, Technical University of Denmark. Annual progress report 2010

    International Nuclear Information System (INIS)

    Korsholm, S.B.; Michelsen, P.K.; Rasmussen, J.J.; Westergaard, C.M.

    2011-04-01

    The programme of the Research Unit of the Fusion Association Euratom - Risoe National Laboratory for Sustainable Energy, Technical University of Denmark, covers work in fusion plasma physics and in fusion technology. The fusion plasma physics research focuses on turbulence and transport, and its interaction with the plasma equilibrium and particles. The effort includes both first principles based modelling, and experimental observations of turbulence and of fast ion dynamics by collective Thomson scattering. Within fusion technology there are activities related to development of high temperature superconductors. Other activities are system analysis, initiative to involve Danish industry in ITER contracts and public information. A summary is presented of the results obtained in the Research Unit during 2010. (Author)

  16. Risoe energy report 8. The intelligent energy system infrastructure for the future

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, Hans; Soenderberg Petersen, L. (eds.)

    2009-09-15

    This report is volume 8 in a series started in 2002, and will take its point of reference in the need for the development of a highly flexible and intelligent energy system infrastructure which facilitates substantial higher amounts of renewable energy than today's energy systems. This intelligent and flexible infrastructure is a prerequisite in achieving the goals set up by IPCC in 2007 on CO{sub 2} reductions as well as ensuring the future security of energy supply in all regions of the world. The report presents a generic approach for future infrastructure issues on local, regional and global scale with focus on the energy system. The report is based on chapters and updates from Risoe Energy Report 1 - 7, as well as input from contributors to the DTU Climate Change Technology workshops and available international literature and reports. (author)

  17. Rare-gas yields in 238U and 232Th fission by 14MeV neutrons, measured by an emanating method

    International Nuclear Information System (INIS)

    Feu Alvim, C.A.; Bocquet, J.P.; Brissot, R.; Crancon, J.; Moussa, A.

    1977-01-01

    A direct method, using emanation of rare gases by uranyle stearate and thorium stearate, has been applied to the measurement of cumulative fractional yields of certain isotopes of krypton and xenon, in the fissions of 238 U and 232 Th by 14MeV-neutrons. The independent yields of the same isotopes were measured previously by means of isotopic on-line separation. From these results, the widths of the mass and charge distributions, the relative chain yields, the fractional cumulative yields of certain bromine and iodine isotopes, the values of Zsub(p) the most probable charge, in the isobaric chains 87-93 and 137-142, and the elemental yields of krypton and xenon were calculated [fr

  18. The nuclear fission

    International Nuclear Information System (INIS)

    Fiorentino, J.

    1983-01-01

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

  19. Muon-induced fission

    International Nuclear Information System (INIS)

    Polikanov, S.

    1980-01-01

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

  20. Relativistic Coulomb Fission

    Science.gov (United States)

    Norbury, John W.

    1992-01-01

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

  1. 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

  2. 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

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

    International Nuclear Information System (INIS)

    2009-01-01

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

  4. Acoustic sensors for fission gas characterization: R and D skills devoted to innovative instrumentation in MTR, non-destructive devices in hot lab facilities and specific transducers for measurements of LWR rods in nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Ferrandis, J.Y.; Leveque, G.; Rosenkrantz, E.; Augereau, F.; Combette, P. [University Montpellier, IES, UMR 5214, F-34000, Montpellier (France); CNRS, IES, UMR 5214, F-34000, Montpellier (France)

    2015-07-01

    First of all, we will present the main principle of the method. A piezoelectric transducer, driven by a pulse generator, generates the acoustic waves in a cavity that may be the fuel rod or a chamber connected to an instrumented rod. The composition determination consists in measuring the time of flight of the acoustic signal emitted. The pressure can be estimated by a calibration process, above the measurement of the amplitude of the signal. Two projects will then be detailed. The first project consists in the development of advanced instrumentation for in-pile experiments in Material Testing Reactor. It constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). As a first step of the development program, we performed in-pile tests on the most sensitive component, i.e., the piezoelectric transducer. For this purpose, the active part of this sensor has been qualified on gamma and neutron radiations and at high temperature. Various industrial piezo-ceramics were exposed to a high activity Cobalt source for few days. The cumulated dose was ranged from 50 kGy up to 2 MGy. Next, these devices were placed inside a Material Test Reactor to investigate their reliability towards neutron fluence. The final fluence after 150 days of irradiation was up to 1.6.10{sup 21}n/cm{sup 2} (for thermal neutron). Irreversible variations have been measured. Next, a specific sensor has been implemented on an instrumented fuel rod and tested in the frame of a REMORA 3 Irradiation test. It was the first experiment under high mixed, temperature neutron and gamma flux. A first irradiation phase took place in March 2010 in the OSIRIS reactor and in November 2010 for the second step of the

  5. Final report on the Risoe monitoring programme after the Chernobyl accident for the period Oct 1, 1986 - Sept 30, 1987

    International Nuclear Information System (INIS)

    Aarkrog, A.; Nielsen, S.P.; Dahlgaard, H.; Lauridsen, B.; Soegaard-Hansen, J.

    1988-01-01

    In cooperation with the National Agency of Environmental Protection in Denmark, Risoe National Laboratory has examined the radioactive contamination from the Chernobyl accident. The programme for these investigations was an expansion of the countrywide monitoring programme operated since 1962 by Risoe National Laboratory. The present report cover the period Oct 1, 1986 to Sept. 30, 1987. All types of environmental samples relevant for radioactive contamination has been analysed. Most samples were collected countrywide and all samples were analysed for radiocaesium ( 134 Cs and 137 Cs). Many samples were furthermore anlaysed for 90 Sr and in a few samples transuranic elements ( 29,240 Pu, 241 Am and 242 Cm) were determined. On the basis of the diet and wholebody measurements of radiocaesium the individual mean dose equivalent commitment from Danish diet consumed in the first two years after the Chernobyl accident was calculated to 27 μ Sv. (author)

  6. Downstream behavior of fission products

    International Nuclear Information System (INIS)

    Johnson, I.; Farahat, M.K.; Settle, J.L.; Johnson, C.E.; Ritzman, R.

    1986-01-01

    The downstream behavior of fission products has been investigated by injecting mixtures of CsOH, CsI, and Te into a flowing steam/hydrogen stream and determining the physical and chemical changes that took place as the gaseous mixture flowed down a reaction duct on which a temperature gradient (1000 0 to 200 0 C) had been imposed. Deposition on the wall of the duct occurred by vapor condensation in the higher temperature regions and by aerosol deposition in the remainder of the duct. Reactions in the gas stream between CsOH and CsI and between CsOH and Te had an effect on the vapor condensation. The aerosol was characterized by the use of impingement tabs placed in the gas stream

  7. Fission gas release behavior of MOX fuels under simulated daily-load-follow operation condition. IFA-554/555 test evaluation with FASTGRASS code

    International Nuclear Information System (INIS)

    Ikusawa, Yoshihisa; Ozawa, Takayuki

    2008-03-01

    IFA-554/555 load-follow tests were performed in HALDEN reactor (HBWR) to study the MOX fuel behavior under the daily-load-follow operation condition in the framework of ATR-MOX fuel development in JAEA. IFA-554/555 rig had the instruments of rod inner pressure, fuel center temperature, fuel stack elongation, and cladding elongation. Although the daily-load-follow operation in nuclear power plant is one of the available options for economical improvement, the power change in a short period in this operation causes the change of thermal and mechanical irradiation conditions. In this report, FP gas release behavior of MOX fuel rod was evaluated under the daily-load-follow operation condition with the examination data from IFA-554/555 by using the computation code 'FASTGRASS'. From the computation results of FASTGRASS code which could compute the FP gas release behavior under the transient condition, it could be concluded that FP gas was released due to the relaxation of fuel pellet inner stress and pellet temperature increase, which were caused by the cyclic power change during the daily-load-follow operation. In addition, since the amount of released FP gas decreased during the steady operation after the daily-load-follow, it could be mentioned that the total of FP gas release at the end of life with the daily-load-follow is not so much different from that without the daily-load-follow. (author)

  8. Isotopic composition of fission gases in LWR fuel

    International Nuclear Information System (INIS)

    Jonsson, T.

    2000-01-01

    Many fuel rods from power reactors and test reactors have been punctured during past years for determination of fission gas release. In many cases the released gas was also analysed by mass spectrometry. The isotopic composition shows systematic variations between different rods, which are much larger than the uncertainties in the analysis. This paper discusses some possibilities and problems with use of the isotopic composition to decide from which part of the fuel the gas was released. In high burnup fuel from thermal reactors loaded with uranium fuel a significant part of the fissions occur in plutonium isotopes. The ratio Xe/Kr generated in the fuel is strongly dependent on the fissioning species. In addition, the isotopic composition of Kr and Xe shows a well detectable difference between fissions in different fissile nuclides. (author)

  9. Simulation of Fission Product Liftoff Behavior During Depressurization Transients

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  10. Fuel morphology effects on fission product release

    International Nuclear Information System (INIS)

    Osetek, D.J.; Hartwell, J.K.; Cronenberg, A.W.

    1986-01-01

    Results are presented of fission product release behavior observed during four severe fuel damage tests on bundles of UO 2 fuel rods. Transient temperatures up to fuel melting were obtained in the tests that included both rapid and slow cooldown, low and high (36 GWd/t) burnup fuel and the addition of Ag-In-Cd control rods. Release fractions of major fission product species and release rates of noble gas species are reported. Significant differences in release behavior are discussed between heatup and cooldown periods, low and high burnup fuel and long- and short-lived fission products. Explanations for the observed differences are offered that relate fuel morphology changes to the releases

  11. Fission in a Plasma

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-26

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

  12. Microscopic Theory of Fission

    International Nuclear Information System (INIS)

    Younes, W; Gogny, D

    2008-01-01

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

  13. Fusion-fission dynamics

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  14. Correlation of recent fission product release data

    International Nuclear Information System (INIS)

    Kress, T.S.; Lorenz, R.A.; Nakamura, T.; Osborne, M.F.

    1989-01-01

    For the calculation of source terms associated with severe accidents, it is necessary to model the release of fission products from fuel as it heats and melts. Perhaps the most definitive model for fission product release is that of the FASTGRASS computer code developed at Argonne National Laboratory. There is persuasive evidence that these processes, as well as additional chemical and gas phase mass transport processes, are important in the release of fission products from fuel. Nevertheless, it has been found convenient to have simplified fission product release correlations that may not be as definitive as models like FASTGRASS but which attempt in some simple way to capture the essence of the mechanisms. One of the most widely used such correlation is called CORSOR-M which is the present fission product/aerosol release model used in the NRC Source Term Code Package. CORSOR has been criticized as having too much uncertainty in the calculated releases and as not accurately reproducing some experimental data. It is currently believed that these discrepancies between CORSOR and the more recent data have resulted because of the better time resolution of the more recent data compared to the data base that went into the CORSOR correlation. This document discusses a simple correlational model for use in connection with NUREG risk uncertainty exercises. 8 refs., 4 figs., 1 tab

  15. Risoe energy report 10. Energy for smart cities in an urbanised world

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, Hans; Soenderberg Petersen, L [eds.

    2011-11-15

    This Risoe Energy Report is the tenth in a series which began in 2002. Volume 10 takes as its point of reference the rapid urbanisation of the world. UN population statistics show that global population is expected to surpass 9 billion by 2050, and nearly 6.3 billion people will be living in urban areas. Urban regions will thus absorb most of the world's population increase in the next four decades while drawing in some of the rural population as well: by 2050 there will be 600 million fewer people in rural areas. The large cities and megacities created by this rapid urbanisation contribute to climate change, and in turn are affected by its consequences. For these and other reasons we need a new approach to what cities should do to become more liveable, economically successful, and environmentally responsible. Megacities of the future need to be smart cities: that is, energy-efficient, consumer-focused and technologydriven. This cannot be achieved simply by improving existing technologies. Instead we need a new smart approach based on smart solutions. With this background the report addresses energy related issues for smart cities, including energy infrastructure, onsite energy production, transport, economy, sustainability, housing, living and governance, including incentives and barriers influencing smart energy for smart cities. (LN)

  16. Risoe energy report 10. Energy for smart cities in an urbanised world

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, Hans; Soenderberg Petersen, L. (eds.)

    2011-11-15

    This Risoe Energy Report is the tenth in a series which began in 2002. Volume 10 takes as its point of reference the rapid urbanisation of the world. UN population statistics show that global population is expected to surpass 9 billion by 2050, and nearly 6.3 billion people will be living in urban areas. Urban regions will thus absorb most of the world's population increase in the next four decades while drawing in some of the rural population as well: by 2050 there will be 600 million fewer people in rural areas. The large cities and megacities created by this rapid urbanisation contribute to climate change, and in turn are affected by its consequences. For these and other reasons we need a new approach to what cities should do to become more liveable, economically successful, and environmentally responsible. Megacities of the future need to be smart cities: that is, energy-efficient, consumer-focused and technologydriven. This cannot be achieved simply by improving existing technologies. Instead we need a new smart approach based on smart solutions. With this background the report addresses energy related issues for smart cities, including energy infrastructure, onsite energy production, transport, economy, sustainability, housing, living and governance, including incentives and barriers influencing smart energy for smart cities. (LN)

  17. O PAÍS DO CARNAVAL: O REFINADO RISO DA EUTRAPELIA

    Directory of Open Access Journals (Sweden)

    Júlia Ciasca Brandão

    2017-07-01

    Full Text Available Este artigo busca compreender o novo significado atribuído ao riso e ao Carnaval pelas autoridades religiosas e civis no século XVII, tendo como objeto de investigação a Utopia, obra homônima à de Thomas Morus, publicada em 1640 e escrita por Jakob Bidermann, jesuíta alemão, dramaturgo e futuro inquisidor da Igreja Católica em Roma. O texto relata a viagem de três amigos para um país imaginário chamado Utopia, onde os habitantes festejam um eterno Carnaval, o caos é institucionalizado, as regras da cultura estão suspensas, os cidadãos andam mascarados, e bebe-se e come-se em demasia. Os viajantes experimentam as terríveis consequências deste “mundo de cabeça para baixo” e as inseguranças de um lugar que vive imerso em vício, vaidade, violência e excesso; descobrem então que o mundo real, se baseado nas virtudes e na moral cristã, e principalmente se guiado pela Igreja Católica, torna-se um lugar muito mais verdadeiro, tranquilo e belo de se viver. Este artigo também discute a escolha do título e o significado de “utopia” concebido pelo texto.

  18. Numerical study of the static and pitching RISOe-B1-18 airfoil[STALL

    Energy Technology Data Exchange (ETDEWEB)

    Bertagnolio, F.

    2004-01-01

    The objective of this report is the better understanding of the physics of the aeroelastic motion of wind turbine blades in order to improve the numerical models used for their design. In this study, the case of the RISOe-B1-18 airfoil which was equipped and measured in an open jet wind tunnel is studied. Two and three dimensional Navier-Stokes calculations using the k-w SST and Detached Eddy Simulation turbulence models are conducted. An engineering semi-empirical dynamic stall model is also used for performing calculations. Computational results are compared to the experimental results that are available both for the static airfoil and in the case of pitching motions. It is shown that the Navier-Stokes simulations can reproduced the main characteristic features of the flow. The DES model seems also to be able to reproduce some details of the unsteady aerodynamics. The Navier-Stokes computations can then be used to improve the performance of the engineering model. (au)

  19. 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.)

  20. 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

  1. Fission Surface Power Technology Development Update

    Science.gov (United States)

    Palac, Donald T.; Mason, Lee S.; Houts, Michael G.; Harlow, Scott

    2011-01-01

    Power is a critical consideration in planning exploration of the surfaces of the Moon, Mars, and places beyond. Nuclear power is an important option, especially for locations in the solar system where sunlight is limited or environmental conditions are challenging (e.g., extreme cold, dust storms). NASA and the Department of Energy are maintaining the option for fission surface power for the Moon and Mars by developing and demonstrating technology for a fission surface power system. The Fission Surface Power Systems project has focused on subscale component and subsystem demonstrations to address the feasibility of a low-risk, low-cost approach to space nuclear power for surface missions. Laboratory demonstrations of the liquid metal pump, reactor control drum drive, power conversion, heat rejection, and power management and distribution technologies have validated that the fundamental characteristics and performance of these components and subsystems are consistent with a Fission Surface Power preliminary reference concept. In addition, subscale versions of a non-nuclear reactor simulator, using electric resistance heating in place of the reactor fuel, have been built and operated with liquid metal sodium-potassium and helium/xenon gas heat transfer loops, demonstrating the viability of establishing system-level performance and characteristics of fission surface power technologies without requiring a nuclear reactor. While some component and subsystem testing will continue through 2011 and beyond, the results to date provide sufficient confidence to proceed with system level technology readiness demonstration. To demonstrate the system level readiness of fission surface power in an operationally relevant environment (the primary goal of the Fission Surface Power Systems project), a full scale, 1/4 power Technology Demonstration Unit (TDU) is under development. The TDU will consist of a non-nuclear reactor simulator, a sodium-potassium heat transfer loop, a power

  2. Fission neutron spectra measurements at LANSCE - status and plans

    International Nuclear Information System (INIS)

    Haight, Robert C.; Noda, Shusaku; Nelson, Ronald O.; O' Donnell, John M.; Devlin, Matt; Chatillon, Audrey; Granier, Thierry; Taieb, Julien; Laurent, Benoit; Belier, Gilbert; Becker, John A.; Wu, Ching-Yen

    2009-01-01

    A program to measure fission neutron spectra from neutron-induced fission of actinides is underway at the Los Alamos Neutron Science Center (LANSCE) in a collaboration among the CEA laboratory at Bruyeres-le-Chatel, Lawrence Livermore National Laboratory and Los Alamos National Laboratory. The spallation source of fast neutrons at LANSCE is used to provide incident neutron energies from less than 1 MeV to 100 MeV or higher. The fission events take place in a gas-ionization fission chamber, and the time of flight from the neutron source to that chamber gives the energy of the incident neutron. Outgoing neutrons are detected by an array of organic liquid scintillator neutron detectors, and their energies are deduced from the time of flight from the fission chamber to the neutron detector. Measurements have been made of the fission neutrons from fission of 235 U, 238 U, 237 Np and 239 Pu. The range of outgoing energies measured so far is from 1 MeV to approximately 8 MeV. These partial spectra and average fission neutron energies are compared with evaluated data and with models of fission neutron emission. Results to date will be presented and a discussion of uncertainties will be given in this presentation. Future plans are to make significant improvements in the fission chambers, neutron detectors, signal processing, data acquisition and the experimental environment to provide high fidelity data including mea urements of fission neutrons below 1 MeV and improvements in the data above 8 MeV.

  3. 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.

  4. 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)

  5. The evaluation for reference fission yield of 238U fission

    International Nuclear Information System (INIS)

    Liang Qichang; Liu Tingjin

    1998-01-01

    In the fission yield data evaluation and measurement, the reference yield is very important, good or poor recommended or measurement values depend upon the reference data to a great extent. According to the CRP's requirement, the evaluation of reference fission yields have been and will be carried out in CNDC, as a part of the whole work (contract No.9504/R 0 /Regular Budget Fund), the evaluation for 29 reference fission yields of 15 product nuclides from 238 U fission have been completed

  6. Fission fragment angular momentum

    International Nuclear Information System (INIS)

    Frenne, D. De

    1991-01-01

    Most of the energy released in fission is converted into translational kinetic energy of the fragments. The remaining excitation energy will be distributed among neutrons and gammas. An important parameter characterizing the scission configuration is the primary angular momentum of the nascent fragments. Neutron emission is not expected to decrease the spin of the fragments by more than one unit of angular momentum and is as such of less importance in the determination of the initial fragment spins. Gamma emission is a suitable tool in studying initial fragment spins because the emission time, number, energy, and multipolarity of the gammas strongly depend on the value of the primary angular momentum. The main conclusions of experiments on gamma emission were that the initial angular momentum of the fragments is large compared to the ground state spin and oriented perpendicular to the fission axis. Most of the recent information concerning initial fragment spin distributions comes from the measurement of isomeric ratios for isomeric pairs produced in fission. Although in nearly every mass chain isomers are known, only a small number are suitable for initial fission fragment spin studies. Yield and half-life considerations strongly limit the number of candidates. This has the advantage that the behavior of a specific isomeric pair can be investigated for a number of fissioning systems at different excitation energies of the fragments and fissioning nuclei. Because most of the recent information on primary angular momenta comes from measurements of isomeric ratios, the global deexcitation process of the fragments and the calculation of the initial fragment spin distribution from measured isomeric ratios are discussed here. The most important results on primary angular momentum determinations are reviewed and some theoretical approaches are given. 45 refs., 7 figs., 2 tabs

  7. Fission product detection

    International Nuclear Information System (INIS)

    Liatard, E.; Akrouf, S.; Bruandet, J.F

    1987-01-01

    The response of photovoltaic cells to heavy ions and fission products have been tested on beam. Their main advantages are their extremely low price, their low sensitivity to energetic light ions with respect to fission products, and the possibility to cut and fit them together to any shape without dead zone. The time output signals of a charge sensitive preamplifier connected to these cells allows fast coincidences. A resolution of 12ns (F.W.H.M.) have been measured between two cells [fr

  8. Low energy nuclear fission

    International Nuclear Information System (INIS)

    Nifenecker, H.

    1982-02-01

    In these lectures we present the liquid drop model of fission and compare some of its prediction with experiment. The liquid drop analogy allows to define in a rather simple and intuitive way a number of useful concepts and possible observables. We then discuss, using the example of the oscillator model, the generality of shell effects. We show how a synthesis of the liquid drop model and of the shell model can be made using the Strutinsky shell averaging procedure. Some experimental data related to the existence of shape isomers are presented and discussed. We conclude by discussing some aspects, both experimental and theoretical, of fission dynamics

  9. Fission of heavy hypernuclei

    International Nuclear Information System (INIS)

    Nifenecker, H.

    1993-01-01

    The results on delayed and prompt fission of heavy hypernuclei obtained by the LEAR PS177 collaboration are recalled and discussed. It is shown that the hypernuclei life-times can be explained in term of a weak strangeness violating lambda-nucleon interaction with a cross section close to 6.0 10 -15 barns. The lambda attachment function is shown to be sensitive to the scission configuration, just before fission, and to the neck dynamics. This function provides a new way to study the nuclear scission process. (author)

  10. Fission modelling with FIFRELIN

    International Nuclear Information System (INIS)

    Litaize, Olivier; Serot, Olivier; Berge, Leonie

    2015-01-01

    The nuclear fission process gives rise to the formation of fission fragments and emission of particles (n,γ, e - ). The particle emission from fragments can be prompt and delayed. We present here the methods used in the FIFRELIN code, which simulates the prompt component of the de-excitation process. The methods are based on phenomenological models associated with macroscopic and/or microscopic ingredients. Input data can be provided by experiment as well as by theory. The fission fragment de-excitation can be performed within Weisskopf (uncoupled neutron and gamma emission) or a Hauser-Feshbach (coupled neutron/gamma emission) statistical theory. We usually consider five free parameters that cannot be provided by theory or experiments in order to describe the initial distributions required by the code. In a first step this set of parameters is chosen to reproduce a very limited set of target observables. In a second step we can increase the statistics to predict all other fission observables such as prompt neutron, gamma and conversion electron spectra but also their distributions as a function of any kind of parameters such as, for instance, the neutron, gamma and electron number distributions, the average prompt neutron multiplicity as a function of fission fragment mass, charge or kinetic energy, and so on. Several results related to different fissioning systems are presented in this work. The goal in the next decade will be i) to replace some macroscopic ingredients or phenomenological models by microscopic calculations when available and reliable, ii) to be a support for experimentalists in the design of detection systems or in the prediction of necessary beam time or count rates with associated statistics when measuring fragments and emitted particle in coincidence iii) extend the model to be able to run a calculation when no experimental input data are available, iv) account for multiple chance fission and gamma emission before fission, v) account for the

  11. Low energy nuclear fission

    International Nuclear Information System (INIS)

    Nifenecker, H.

    1980-08-01

    In these lectures the liquid drop model of fission is presented and some of its predictions compared with experiment. The liquid drop analogy allows to define in a rather simple and intuitive way a number of useful concepts and possible observables. It is shown how a synthesis of the liquid drop model and of the shell model can be made using the Strutinsky shell averaging procedure. Some experimental data related to the existence of shape isomers are presented and discussed. We conclude by discussing some aspects, both experimental and theoretical, of fission dynamics

  12. Angular momentum distribution of primary fission fragments by measurement of the relative yield of isomeric fission products

    International Nuclear Information System (INIS)

    Dornhoefer, H.

    1980-01-01

    The fission products 132 I and 136 I produced in the fission reactions 238 U(α,f) and 238 U(d,f) were spectroscoped using a gas transport system. Thereby was taken advantage of the fact that at the transport with pure helium without aerosols only iodine activities were collected in a membrane filter. The relative independent yields of the isomeric fission products of 132 I and 136 I were determined for different excitation energies. Thereby was taken advantage of the fact that the transport yield of the gas transport system for 136 I directly produced from the fission was greater than for iodine indirectly produced by β-decay. (orig./HSI) [de

  13. Fission Product Library and Resource

    Energy Technology Data Exchange (ETDEWEB)

    Burke, J. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Padgett, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-09-29

    Fission product yields can be extracted from an irradiated sample by performing gamma ray spectroscopy on the whole sample post irradiation. There are several pitfalls to avoid when trying to determine a specific isotope's fission product yield.

  14. Spontaneous fission of superheavy nuclei

    Indian Academy of Sciences (India)

    The fission-like configurations are used for the total deformation energy calculations. A ... oscillator potential for the two fission fragment regions reads as ... Beyond this limit, the contribution of more remote levels is negligible. Once the density ...

  15. Mirror fusion--fission hybrids

    International Nuclear Information System (INIS)

    Lee, J.D.

    1978-01-01

    The fusion-fission concept and the mirror fusion-fission hybrid program are outlined. Magnetic mirror fusion drivers and blankets for hybrid reactors are discussed. Results of system analyses are presented and a reference design is described

  16. 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

  17. 50 years of nuclear fission

    International Nuclear Information System (INIS)

    Hilscher, D.

    1989-01-01

    The article tells the story of the discovery of nuclear fission in Berlin 50 years ago by Otto Hahn and Fritz Strassmann in cooperation with Lise Meitner. 50 years later nuclear fission is still a subject of research. Some question remain unanswered. Selected new research results are used to discuss the dynamics of the collective movement of the elementary nuclear fission process. (orig.) [de

  18. Fission dynamics of hot nuclei

    Indian Academy of Sciences (India)

    2014-04-05

    Apr 5, 2014 ... across the fission barrier is very small or in other words, the fission barrier is much ... of this shape evolution, the gross features of the fissioning nucleus can be described ..... [7] Y Abe, C Gregoire and H Delagrange, J. Phys.

  19. Status of fission yield measurements

    International Nuclear Information System (INIS)

    Maeck, W.J.

    1979-01-01

    Fission yield measurement and yield compilation activities in the major laboratories of the world are reviewed. In addition to a general review of the effort of each laboratory, a brief summary of yield measurement activities by fissioning nuclide is presented. A new fast reactor fission yield measurement program being conducted in the US is described

  20. The discovery of fission

    International Nuclear Information System (INIS)

    McKay, H.A.C.

    1978-01-01

    In this article by the retired head of the Separation Processes Group of the Chemistry Division, Atomic Energy Research Establishment, Harwell, U.K., the author recalls what he terms 'an exciting drama, the unravelling of the nature of the atomic nucleus' in the years before the Second World War, including the discovery of fission. 12 references. (author)

  1. 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.

  2. 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

  3. 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)

  4. 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

  5. Prompt fission neutron spectra and average prompt neutron multiplicities

    International Nuclear Information System (INIS)

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

    1983-01-01

    We present a new method for calculating the prompt fission neutron spectrum N(E) and average prompt neutron multiplicity anti nu/sub p/ as functions of the fissioning nucleus and its excitation energy. The method is based on standard nuclear evaporation theory and takes into account (1) the motion of the fission fragments, (2) the distribution of fission-fragment residual nuclear temperature, (3) the energy dependence of the cross section sigma/sub c/ for the inverse process of compound-nucleus formation, and (4) the possibility of multiple-chance fission. We use a triangular distribution in residual nuclear temperature based on the Fermi-gas model. This leads to closed expressions for N(E) and anti nu/sub p/ when sigma/sub c/ is assumed constant and readily computed quadratures when the energy dependence of sigma/sub c/ is determined from an optical model. Neutron spectra and average multiplicities calculated with an energy-dependent cross section agree well with experimental data for the neutron-induced fission of 235 U and the spontaneous fission of 252 Cf. For the latter case, there are some significant inconsistencies between the experimental spectra that need to be resolved. 29 references

  6. Mechanistic prediction of fission product release under normal and accident conditions: key uncertainties that need better resolution

    International Nuclear Information System (INIS)

    Rest, J.

    1983-09-01

    A theoretical model has been used for predicting the behavior of fission gas and volatile fission products (VFPs) in UO 2 -base fuels during steady-state and transient conditions. This model represents an attempt to develop an efficient predictive capability for the full range of possible reactor operating conditions. Fission products released from the fuel are assumed to reach the fuel surface by successively diffusing (via atomic and gas-bubble mobility) from the grains to grain faces and then to the grain edges, where the fission products are released through a network of interconnected tunnels of fission-gas induced and fabricated porosity. The model provides for a multi-region calculation and uses only one size class to characterize a distribution of fission gas bubbles

  7. Ternary Fission of U235 by Resonance Neutrons

    International Nuclear Information System (INIS)

    Kvitek, I.; Popov, Ju.P.; Rjabov, Ju.V.

    1965-01-01

    Recently a number of papers have appeared indicating considerable variations in the ratio of the ternary-fission cross-section to the binary-fission cross-section of U 235 on transition from one neutron resonance to another. However, such variations have not been discovered in U 233 and Pu 239 . The paper reports investigations of the ternary fission of U 235 by neutrons with an energy of 0.1 to 30 eV. Unlike other investigators of the ternary fission of U 235 , we identified the ternary-fission event by the coincidence of one of the fission fragments with a light long-range particle. This made it passible to separate ternary fissions from the possible contribution of the (n, α)reaction. The measurements were performed at the fast pulsed reactor of the Joint Institute for Nuclear Research by the time-of-flight method. A flight length of 100 m was used, giving a resolution of 0.6 μs/m. Gas scintillation counters filled with xenon at a pressure of 2 atm were used to record the fission fragments and the light long-range particle. A layer of enriched U 235 ∼2 mg/cm 2 thick and ∼300 cm 2 in area was applied to an aluminium foil 20-fim thick. The scintillations from the fission fragments were recorded in the gas volume on one side of the foil and those from the light long-range particles in that on the other. In order to assess the background (e.g . coincidences of the pulse from a fragment with that from a fission gamma quantum or a proton from the (n, p) reaction in the aluminium foil), a measurement was carried out in which the volume recording the long-range particle was shielded with a supplementary aluminium filter 1-mm thick. The results obtained indicate the absence of the considerable variations in the ratio between the ternary-and binary- fission cross-sections for U 235 that have been noted by other authors. Measurements showed no irregularity in the ratio of the cross-sections in the energy range 0.1 to 0.2 eV. The paper discusses the possible effect of

  8. Fusion-fission dynamics

    International Nuclear Information System (INIS)

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

    1991-04-01

    Classical dynamical calculations of the heavy ion induced fission process for the reactions 40 Ar+ 141 Pr, 20 Ne+ 165 Ho and 12 C+ 175 Lu leading to the iridium like nucleus have been performed. As a result prescission lifetimes were obtained and compared with the experimental values. The agreement between the calculated and experimental lifetimes indicates that the one-body dissipation picture is much more relevant in describing the fusion-fission dynamics than the two-body one. Somewhat bigger calculated times than the experimental ones in case of the C+Lu reaction at 16 MeV/nucleon may be a signal on the energy range applicability of the one-body dissipation model. (author)

  9. Investigation of pellet acceleration by an arc heated gas gun

    International Nuclear Information System (INIS)

    Andersen, S.A.; Baekmark, L.; Jensen, V.O.; Michelsen, P.; Weisberg, K.V.

    1988-10-01

    This report describes work on pellet acceleration by means of an arc heated gas gun. Preliminary results were described in Riso-M-2536 and in Riso-M-2650. This final report describes the work carried out from 1987.03.31 to 1988.09.30. An arc heated hydrogen gas source, for pneumatic acceleration of deuterium pellets to velocities above 2 km/s, was developed. Experiments were performed with an arc chamber to which different methods of hydrogen supply were possible, and to which the input of electrical power could be programmed. Results in terms of pressure transients and acceleration curves are presented. Maximum pellet velocities approaching 2 km/s were obtained. This limit is discussed in relation to the presented data. Finally this report contains a summary and a conclusion for the entire project. (author) 34 ills., 3 refs

  10. The fission track method

    International Nuclear Information System (INIS)

    Hansen, K.

    1990-01-01

    During the last decade fission track (FT) analysis has evolved as an important tool in exploration for hydrocarbon resources. Most important is this method's ability to yield information about temperatures at different times (history), and thus relate oil generation and time independently of other maturity parameters. The purpose of this paper is to introduce the basics of the method and give an example from the author's studies. (AB) (14 refs.)

  11. 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.

  12. 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.

  13. [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

  14. 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.)

  15. 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

  16. Gas

    International Nuclear Information System (INIS)

    1996-01-01

    The French government has decided to modify the conditions of extension of local natural gas authorities to neighbouring districts. The European Union is studying the conditions of internal gas market with the objective of more open markets although considering public service requirements

  17. Environmental life cycle assessment of high temperature nuclear fission and fusion biomass gasification plants

    International Nuclear Information System (INIS)

    Takeda, Shutaro; Sakurai, Shigeki; Kasada, Ryuta; Konishi, Satoshi

    2017-01-01

    The authors propose nuclear biomass gasification plant as an advancement of conventional gasification plants. Environmental impacts of both fission and fusion plants were assessed through life cycle assessment. The result suggested the reduction of green-house gas emissions would be as large as 85.9% from conventional plants, showing a potential for the sustainable future for both fission and fusion plants. (author)

  18. Experimental determination of fission gas adsorption coefficients

    International Nuclear Information System (INIS)

    Lovell, R.; Underhill, D.W.

    1979-01-01

    Large charcoal beds have been used for a number of years for the holdup and decay of radioactive isotopes of krypton and xenon. Reliable design of these beds depends on an accurate knowledge of the adsorption coefficient of krypton and xenon on the adsorbents used in these beds. It is somewhat surprising that there is no standard procedure of determining the adsorption coefficient for krypton and xenon. Fundamental information needed to establish a standardized reproducible test procedure is given emphasizing the breakthrough curves commonly used to analyze dynamic adsorption data can lead to serious systematic errors and the fact that the adsorption coefficient, if calculated from the arithmetic holding time, is independent of geometric factors such as the shape of the adsorption bed and the irregular shape of the adsorbent

  19. Fission fragment pumping of CO gas mixtures

    International Nuclear Information System (INIS)

    McArthur, D.A.

    1978-01-01

    Some initial experiments on the spectral properties of the room-temperature CO nuclear pumped laser (NPL), and also the results of some parameter optimization experiments are presented. The parameter optimization has resulted in peak laser powers approximately 20 W and laser energies > or approximately equal to 1 mJ, which are greater than those reported thus far for any other room temperature NPL (however, the laser powers and energies of the low-temperature CO NPL are still much higher than those reported here). The spectral properties of the CO laser were studied because they may reveal how energy enters the vibrational level system of the CO molecule, and thus help quantify the amount of vibrational excitation which is occurring

  20. Low- and intermediate level radioactive waste from Risoe, Denmark. Site studies. Report no. 5. Thise, Skive Municipality; Lav- og mellem radioaktivt affald fra Risoe, Danmark. Omegnsstudier. Rapport nr. 5. Omraede Thise, Skive Kommune

    Energy Technology Data Exchange (ETDEWEB)

    Gravesen, P.; Nilsson, B.; Binderup, M.; Larsen, Tine; Schack Pedersen, S.A.

    2012-07-01

    The low- and intermediate-level radioactive wastes from Risoe (the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes) have to be stored in a final disposal in Denmark for at least 300 years. In 2011, the results of the first analyses of 20 potential areas for siting a waste disposal were published. Of these potential areas, 6 specific sites were selected for further detailed studies. The site studies include information about geology, land use, nature preservation, archaeology, drinking water supply etc. The 5 municipalities with the 6 selected sites have been visited to obtain as much information about local conditions as possible. The present report describes the results for the area Thise, in the Municipality of Skive, northern Jutland. (LN)

  1. Low- and intermediate level radioactive waste from Risoe, Denmark. Site studies. Report no. 3. Kertinge Mark, Kerteminde Municipality; Lav- og mellem radioaktivt affald fra Risoe, Danmark. Omegnsstudier. Rapport nr. 3. Omraede Kertinge Mark, Kerteminde Kommune

    Energy Technology Data Exchange (ETDEWEB)

    Gravesen, P.; Nilsson, B.; Binderup, M.; Larsen, Tine; Schack Pedersen, S.A.

    2012-07-01

    The low- and intermediate-level radioactive wastes from Risoe (the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes) have to be stored in a final disposal in Denmark for at least 300 years. In 2011, the results of the first analyses of 20 potential areas for siting a waste disposal were published. Of these potential areas, 6 specific sites were selected for further detailed studies. The site studies include information about geology, land use, nature preservation, archaeology, drinking water supply etc. The 5 municipalities with the 6 selected sites have been visited to obtain as much information about local conditions as possible. The present report describes the results for the area Kertinge Mark in the Municipality of Kerteminde, the island Funen. (LN)

  2. Low- and intermediate level radioactive waste from Risoe, Denmark. Site studies. Report no. 4. Hvidbjerg, Thyholm, Struer Municipality; Lav- og mellem radioaktivt affald fra Risoe, Danmark. Omegnsstudier. Rapport nr. 4. Omraede Hvidbjerg, Thyholm, Struer Kommune

    Energy Technology Data Exchange (ETDEWEB)

    Gravesen, P.; Nilsson, B.; Binderup, M.; Larsen, Tine; Schack Pedersen, S.A.

    2012-07-01

    The low- and intermediate-level radioactive wastes from Risoe (the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes) have to be stored in a final disposal in Denmark for at least 300 years. In 2011, the results of the first analyses of 20 potential areas for siting a waste disposal were published. Of these potential areas, 6 specific sites were selected for further detailed studies. The site studies include information about geology, land use, nature preservation, archaeology, drinking water supply etc. The 5 municipalities with the 6 selected sites have been visited to obtain as much information about local conditions as possible. The present report describes the results for the area Hvidbjerg, Thyholm, in the Municipality of Struer, northern Jutland. (LN)

  3. Low- and intermediate level radioactive waste from Risoe, Denmark. Site studies. Report no. 6. Skive Vest, Skive Municipality; Lav- og mellem radioaktivt affald fra Risoe, Danmark. Omegnsstudier. Rapport nr. 6. Omraede Skive Vest, Skive Kommune

    Energy Technology Data Exchange (ETDEWEB)

    Gravesen, P.; Nilsson, B.; Binderup, M.; Larsen, Tine; Schack Pedersen, S.A.

    2012-07-01

    The low- and intermediate-level radioactive wastes from Risoe (the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes) have to be stored in a final disposal in Denmark for at least 300 years. In 2011, the results of the first analyses of 20 potential areas for siting a waste disposal were published. Of these potential areas, 6 specific sites were selected for further detailed studies. The site studies include information about geology, land use, nature preservation, archaeology, drinking water supply etc. The 5 municipalities with the 6 selected sites have been visited to obtain as much information about local conditions as possible. The present report describes the results for the area Skive Vest, in the Municipality of Skive, northern Jutland. (LN)

  4. Low- and intermediate level radioactive waste from Risoe, Denmark. Site studies. Report no. 2. Roedbyhavn, Lolland Municipality; Lav- og mellem radioaktivt affald fra Risoe, Danmark. Omegnsstudier. Rapport nr. 2. Omraede Roedbyhavn, Lolland Kommune

    Energy Technology Data Exchange (ETDEWEB)

    Gravesen, P.; Nilsson, B.; Binderup, M.; Larsen, Tine; Schack Pedersen, S.A.

    2012-07-01

    The low- and intermediate-level radioactive wastes from Risoe (the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes) have to be stored in a final disposal in Denmark for at least 300 years. In 2011, the results of the first analyses of 20 potential areas for siting a waste disposal were published. Of these potential areas, 6 specific sites were selected for further detailed studies. The site studies include information about geology, land use, nature preservation, archaeology, drinking water supply etc. The 5 municipalities with the 6 selected sites have been visited to obtain as much information about local conditions as possible. The present report describes the results for the area at Roedbyhavn in the Municipality of Lolland, southern Denmark. (LN)

  5. Low- and intermediate level radioactive waste from Risoe, Denmark. Site studies. Report no. 1. Oestermarie - Paradisbakkerne, Bornholm Region; Lav- og mellem radioaktivt affald fra Risoe, Danmark. Omegnsstudier. Rapport nr. 1. Omraede Oestermarie - Paradisbakkerne, Bornholms Regionskommune

    Energy Technology Data Exchange (ETDEWEB)

    Gravesen, P.; Nilsson, B.; Binderup, M.; Larsen, Tine; Schack Pedersen, S.A.

    2012-07-01

    The low- and intermediate-level radioactive wastes from Risoe (the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes) have to be stored in a final disposal in Denmark for at least 300 years. In 2011, the results of the first analyses of 20 potential areas for siting a waste disposal were published. Of these potential areas, 6 specific sites were selected for further detailed studies. The site studies include information about geology, land use, nature preservation, archaeology, drinking water supply etc. The 5 municipalities with the 6 selected sites have been visited to obtain as much information about local conditions as possible. The present report describes the results for the area at Oestermarie-Paradisbakkerne in the region of Bornholm, East Denmark. (LN)

  6. A small flat fission chamber

    International Nuclear Information System (INIS)

    Li Yijun; Wang Dalun; Chen Suhe

    1999-01-01

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

  7. Fission barriers of light nuclei

    International Nuclear Information System (INIS)

    Grotowski, K.; Planeta, R.; Blann, M.; Komoto, T.

    1989-01-01

    Experimental fission excitation functions for compound nuclei /sup 52/Fe, /sup 49/Cr, /sup 46/V, and /sup 44/Ti formed in heavy-ion reactions are analyzed in the Hauser-Feshbach/Bohr-Wheeler formalism using fission barriers based on the rotating liquid drop model of Cohen et al. and on the rotating finite range model of Sierk. We conclude that the rotating finite range approach gives better reproduction of experimental fission yields, consistent with results found for heavier systems

  8. Nuclear Forensics and Radiochemistry: Fission

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-07

    Radiochemistry has been used to study fission since it’ discovery. Radiochemical methods are used to determine cumulative mass yields. These measurements have led to the two-mode fission hypothesis to model the neutron energy dependence of fission product yields. Fission product yields can be used for the nuclear forensics of nuclear explosions. The mass yield curve depends on both the fuel and the neutron spectrum of a device. Recent studies have shown that the nuclear structure of the compound nucleus can affect the mass yield distribution.

  9. Micro plate fission chamber development

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  10. 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.)

  11. 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

  12. Fission fragment simulation of fusion neutron radiation effects on bulk mechanical properties

    International Nuclear Information System (INIS)

    Van Konynenburg, R.A.; Mitchell, J.B.; Guinan, M.W.; Stuart, R.N.; Borg, R.J.

    1976-01-01

    This research demonstrates the feasibility of using homogeneously-generated fission fragments to simulate high-fluence fusion neutron damage in niobium tensile specimens. This technique makes it possible to measure radiation effects on bulk mechanical properties at high damage states, using conveniently short irradiation times. The primary knock-on spectrum for a fusion reactor is very similar to that produced by fission fragments, and nearly the same ratio of gas atoms to displaced atoms is produced in niobium. The damage from fission fragments is compared to that from fusion neutrons and fission reactor neutrons in terms of experimentally measured yield strength increase, transmission electron microscopy (TEM) observations, and calculated damage energies

  13. Energy from nuclear fission()

    Science.gov (United States)

    Ripani, M.

    2015-08-01

    The main features of nuclear fission as physical phenomenon will be revisited, emphasizing its peculiarities with respect to other nuclear reactions. Some basic concepts underlying the operation of nuclear reactors and the main types of reactors will be illustrated, including fast reactors, showing the most important differences among them. The nuclear cycle and radioactive-nuclear-waste production will be also discussed, along with the perspectives offered by next generation nuclear assemblies being proposed. The current situation of nuclear power in the world, its role in reducing carbon emission and the available resources will be briefly illustrated.

  14. Low- and intermediate level radioactive waste from Risoe, Denmark. Site studies. Report no. 5. Thise, Skive Municipality

    International Nuclear Information System (INIS)

    Gravesen, P.; Nilsson, B.; Binderup, M.; Larsen, Tine; Schack Pedersen, S.A.

    2012-01-01

    The low- and intermediate-level radioactive wastes from Risoe (the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes) have to be stored in a final disposal in Denmark for at least 300 years. In 2011, the results of the first analyses of 20 potential areas for siting a waste disposal were published. Of these potential areas, 6 specific sites were selected for further detailed studies. The site studies include information about geology, land use, nature preservation, archaeology, drinking water supply etc. The 5 municipalities with the 6 selected sites have been visited to obtain as much information about local conditions as possible. The present report describes the results for the area Thise, in the Municipality of Skive, northern Jutland. (LN)

  15. Low- and intermediate level radioactive waste from Risoe, Denmark. Site studies. Report no. 2. Roedbyhavn, Lolland Municipality

    International Nuclear Information System (INIS)

    Gravesen, P.; Nilsson, B.; Binderup, M.; Larsen, Tine; Schack Pedersen, S.A.

    2012-01-01

    The low- and intermediate-level radioactive wastes from Risoe (the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes) have to be stored in a final disposal in Denmark for at least 300 years. In 2011, the results of the first analyses of 20 potential areas for siting a waste disposal were published. Of these potential areas, 6 specific sites were selected for further detailed studies. The site studies include information about geology, land use, nature preservation, archaeology, drinking water supply etc. The 5 municipalities with the 6 selected sites have been visited to obtain as much information about local conditions as possible. The present report describes the results for the area at Roedbyhavn in the Municipality of Lolland, southern Denmark. (LN)

  16. Low- and intermediate level radioactive waste from Risoe, Denmark. Site studies. Report no. 3. Kertinge Mark, Kerteminde Municipality

    International Nuclear Information System (INIS)

    Gravesen, P.; Nilsson, B.; Binderup, M.; Larsen, Tine; Schack Pedersen, S.A.

    2012-01-01

    The low- and intermediate-level radioactive wastes from Risoe (the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes) have to be stored in a final disposal in Denmark for at least 300 years. In 2011, the results of the first analyses of 20 potential areas for siting a waste disposal were published. Of these potential areas, 6 specific sites were selected for further detailed studies. The site studies include information about geology, land use, nature preservation, archaeology, drinking water supply etc. The 5 municipalities with the 6 selected sites have been visited to obtain as much information about local conditions as possible. The present report describes the results for the area Kertinge Mark in the Municipality of Kerteminde, the island Funen. (LN)

  17. Low- and intermediate level radioactive waste from Risoe, Denmark. Site studies. Report no. 4. Hvidbjerg, Thyholm, Struer Municipality

    International Nuclear Information System (INIS)

    Gravesen, P.; Nilsson, B.; Binderup, M.; Larsen, Tine; Schack Pedersen, S.A.

    2012-01-01

    The low- and intermediate-level radioactive wastes from Risoe (the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes) have to be stored in a final disposal in Denmark for at least 300 years. In 2011, the results of the first analyses of 20 potential areas for siting a waste disposal were published. Of these potential areas, 6 specific sites were selected for further detailed studies. The site studies include information about geology, land use, nature preservation, archaeology, drinking water supply etc. The 5 municipalities with the 6 selected sites have been visited to obtain as much information about local conditions as possible. The present report describes the results for the area Hvidbjerg, Thyholm, in the Municipality of Struer, northern Jutland. (LN)

  18. Low- and intermediate level radioactive waste from Risoe, Denmark. Site studies. Report no. 6. Skive Vest, Skive Municipality

    International Nuclear Information System (INIS)

    Gravesen, P.; Nilsson, B.; Binderup, M.; Larsen, Tine; Schack Pedersen, S.A.

    2012-01-01

    The low- and intermediate-level radioactive wastes from Risoe (the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes) have to be stored in a final disposal in Denmark for at least 300 years. In 2011, the results of the first analyses of 20 potential areas for siting a waste disposal were published. Of these potential areas, 6 specific sites were selected for further detailed studies. The site studies include information about geology, land use, nature preservation, archaeology, drinking water supply etc. The 5 municipalities with the 6 selected sites have been visited to obtain as much information about local conditions as possible. The present report describes the results for the area Skive Vest, in the Municipality of Skive, northern Jutland. (LN)

  19. Low- and intermediate level radioactive waste from Risoe, Denmark. Site studies. Report no. 1. Oestermarie - Paradisbakkerne, Bornholm Region

    International Nuclear Information System (INIS)

    Gravesen, P.; Nilsson, B.; Binderup, M.; Larsen, Tine; Schack Pedersen, S.A.

    2012-01-01

    The low- and intermediate-level radioactive wastes from Risoe (the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes) have to be stored in a final disposal in Denmark for at least 300 years. In 2011, the results of the first analyses of 20 potential areas for siting a waste disposal were published. Of these potential areas, 6 specific sites were selected for further detailed studies. The site studies include information about geology, land use, nature preservation, archaeology, drinking water supply etc. The 5 municipalities with the 6 selected sites have been visited to obtain as much information about local conditions as possible. The present report describes the results for the area at Oestermarie-Paradisbakkerne in the region of Bornholm, East Denmark. (LN)

  20. 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.

  1. 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.)

  2. 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

  3. Axisymmetric Magnetic Mirror Fusion-Fission Hybrid

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-05-13

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

  4. Weld Development for Aluminum Fission Chamber

    Energy Technology Data Exchange (ETDEWEB)

    Cross, Carl Edward [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez, Jesse Norris [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-16

    The Sigma welding team was approached to help fabricate a small fission chamber (roughly ½ inch dia. x ½ inch tall cylinder). These chambers are used as radiation sensors that contain small traces of radionuclides (Cf 252, U 235, and U 238) that serve to ionize gas atoms in addition to external radiation. When a voltage is applied within the chamber, the resulting ion flow can be calibrated and monitored. Aluminum has the advantage of not forming radioactive compounds when exposed to high external radiation (except from minor Na alloy content). Since aluminum has not been used before in this application, this presented an unexplored challenge.

  5. Neutron and thermal dynamics of a gaseous core fission reactor

    International Nuclear Information System (INIS)

    van Dam, H.; Kuijper, J.C.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

    1989-01-01

    In this paper neutron kinetics and thermal dynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focused on the properties of the fuel gas, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

  6. Reactor physics and thermodynamics of a gaseous core fission reactor

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Van Dam, H.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

    1990-01-01

    Neutron kinetics and thermodynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focussed on the properties of the fuel gas, the stationary temperature distribution, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

  7. Neutron multiplicity of fission fragments

    Energy Technology Data Exchange (ETDEWEB)

    Abdelrahman, Y S [Physics department, mu` rah university Al-Karak, (Jordan)

    1995-10-01

    The total average neutron multiplicity of the fission fragments produced by the spontaneous fission of {sup 248} Cm has been measured. This measurement has been done by using a new experimental technique. This technique mainly depends on {gamma}-{gamma} coincidence using a very high resolution high purity germanium (HPGe) detector. 2 figs.

  8. Fission throughout the periodic table

    International Nuclear Information System (INIS)

    Moretto, L.G.; Wozniak, G.J.

    1989-04-01

    The dualistic view of fission and evaporation as two distinct compound nucleus processes is substituted with a unified view in which fission, complex fragment emission, and light particle evaporation are seen as different aspects of a single process. 47 refs., 22 figs

  9. Trapping technology for gaseous fission products from voloxidation process

    International Nuclear Information System (INIS)

    Shin, Jin Myeong; Park, J. J.; Park, G. I.; Jung, I. H.; Lee, H. H.; Kim, G. H.; Yang, M. S.

    2005-05-01

    The objective of this report is to review the different technologies for trapping the gaseous wastes containing Cs, Ru, Tc, 14 C, Kr, Xe, I and 3 H from a voloxidation process. Based on literature reviews and KAERI's experimental results on the gaseous fission products trapping, appropriate trapping method for each fission product has been selected considering process reliability, simplicity, decontamination factor, availability, and disposal. Specifically, the most promising trapping method for each fission product has been proposed for the development of the INL off-gas trapping system. A fly ash filter is proposed as a trapping media for a cesium trapping unit. In addition, a calcium filter is proposed as a trapping media for ruthenium, technetium, and 14 C trapping unit. In case of I trapping unit, AgX is proposed. For Kr and Xe, adsorption on solid is proposed. SDBC (Styrene Divinyl Benzene Copolymer) is also proposed as a conversion media to HTO for 3 H. This report will be used as a useful means for analyzing the known trapping technologies and help selecting the appropriate trapping methods for trapping volatile and semi-volatile fission products, long-lived fission products, and major heat sources generated from a voloxidation process. It can also be used to design an off-gas treatment system

  10. Trapping technology for gaseous fission products from voloxidation process

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Jin Myeong; Park, J. J.; Park, G. I.; Jung, I. H.; Lee, H. H.; Kim, G. H.; Yang, M. S

    2005-05-15

    The objective of this report is to review the different technologies for trapping the gaseous wastes containing Cs, Ru, Tc, {sup 14}C, Kr, Xe, I and {sup 3}H from a voloxidation process. Based on literature reviews and KAERI's experimental results on the gaseous fission products trapping, appropriate trapping method for each fission product has been selected considering process reliability, simplicity, decontamination factor, availability, and disposal. Specifically, the most promising trapping method for each fission product has been proposed for the development of the INL off-gas trapping system. A fly ash filter is proposed as a trapping media for a cesium trapping unit. In addition, a calcium filter is proposed as a trapping media for ruthenium, technetium, and {sup 14}C trapping unit. In case of I trapping unit, AgX is proposed. For Kr and Xe, adsorption on solid is proposed. SDBC (Styrene Divinyl Benzene Copolymer) is also proposed as a conversion media to HTO for {sup 3}H. This report will be used as a useful means for analyzing the known trapping technologies and help selecting the appropriate trapping methods for trapping volatile and semi-volatile fission products, long-lived fission products, and major heat sources generated from a voloxidation process. It can also be used to design an off-gas treatment system.

  11. Post-irradiation studies on knock-out and pseudo-recoil releases of fission products from fissioning UO2

    International Nuclear Information System (INIS)

    Yamagishi, S.; Tanifuji, T.

    1976-01-01

    By using post-irradiation techniques, in-pile releases of 133 Xe, sup(85m)Kr, 88 Kr, 87 Kr and 138 Xe from UO 2 fissioning at low temperatures below about 200 0 C are studied: these are analyzed into a time-dependent knock-out and time-independent pseudo-recoil releases. For the latter, a 'self knock-out' mechanism is proposed: when a fission fragment loses thoroughly its energy near the UO 2 surface and stops there, it will knock out the surface substances and accordingly the fragment (i.e. the fission product) will be released. The effective thickness of the layer where the self knock-out occurs is found to be approximately 7A. As for the knock-out release, the following is estimated from its dependence on various factors: the knock-out release of fission products occurs from the surface layer with the effective thickness of approximately 20A: the shape of UO 2 matrix knocked out by one fission fragment passing through the surface is equivalent to a cylinder approximately 32A diameter by approximately 27A thick, (i.e. the knock-out coefficient for UO 2 is approximately 660 uranium atoms per knock-out event). On the basis of the above estimations, the conclusions derived from the past in-pile studies of fission gas releases are evaluated. (Auth.)

  12. 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

  13. Mechanisms of fission neutron emission

    International Nuclear Information System (INIS)

    Maerten, H.

    1991-01-01

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

  14. 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

  15. How fission was discovered

    International Nuclear Information System (INIS)

    Fluegge, S.

    1989-01-01

    After the great survey of neutron induced radioactivity by Fermi and co-workers, the laboratories in Paris and Berlin-Dahlen tried to disentangle the complex results found in uranium. At that time neutron sources were small, activities low, and equipment very simple. Chemistry beyond uranium still was unknown. Hahn and Meitner believed to have observed three transuranic isomeric chains, a doubtful result even then. Early in 1938, Curie and Savic in Paris found an activity interpreted to be actinium, and Hahn and Meitner another to be radium. Both interpretations seemed impossible from energy considerations. Hahn and Strassmann, therefore, continued this work and succeeded to separate the new activity from radium. There remained no doubt that a barium isotope had been produced, the uranium nucleus splitting in the yet-unknown process we now call fission

  16. Fifty years with nuclear fission

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  17. 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

  18. 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.

  19. 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.)

  20. 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