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

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

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

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)

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)

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.

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

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1985-07-01

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

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

International Nuclear Information System (INIS)

1985-01-01

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

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

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

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

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

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.)

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)

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

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

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.

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.)

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

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

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.

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.

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)

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)

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.

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.

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

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

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)

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

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

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)

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

Nuclear fission: a review of experimental advances and phenomenology

Science.gov (United States)

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

2018-01-01

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

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).

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)

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)

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

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)

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.

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)

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.

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

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

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)

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.

Experimental approach to fission process of actinides

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-01

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

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)

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)

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.

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

Fission times of excited nuclei: An experimental overview

International Nuclear Information System (INIS)

Morjean, M.; Morjean, M.; Jacquet, D.

2009-01-01

An overview of selected recent experimental results on fission times is presented. Evidences for over-damped motion up to saddle point during the fission process of highly excited nuclei have been obtained independently through fission probability, pre-scission multiplicity and direct time measurements. In addition, strong clues have been found for a temperature dependency of friction. Experiments probing transient effects through fission probabilities are presented and the counterbalanced effects of friction and level density parameters are discussed. Promising perspectives for super-heavy stability studies, based on fission time measurements, are presented. (authors)

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

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

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

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)

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

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)

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

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

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)

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

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

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)

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)

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

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....

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

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.))

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

Experimental survey of the potential energy surfaces associated with fission

International Nuclear Information System (INIS)

Britt, H.C.

1980-01-01

Progress in the experimental determination of the properties of the potential energy surface associated with fission is reviewed. The importance of nuclear symmetry effects on the calculation of fission widths is demonstrated. Evidence is presented for the fragmentation of the mass-asymmetric second barrier in the thorium region and the axial asymmetric first barrier in the californium region. Detailed analyses of experimental data suggest the presence of two parallel second barriers; the normal mass-asymmetric, axial-symmetric barrier and a slightly higher mass-symmetric, axial-asymmetric barrier. Experimental barrier parameters are determined systematically and compared with calculations from various theoretical models. Techniques for expanding fission probability measurements to higher energies are discussed. (author)

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)

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

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

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

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.

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

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.

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)

Development and optimization of neutron measurement methods by fission chamber on experimental reactors - management, treatment and reduction of uncertainties

International Nuclear Information System (INIS)

Blanc-De-Lanaute, N.

2012-01-01

The main objectives of this research thesis are the management and reduction of uncertainties associated with measurements performed by means of a fission-chamber type sensor. The author first recalls the role of experimental reactors in nuclear research, presents the various sensors used in nuclear detection (photographic film, scintillation sensor, gas ionization sensor, semiconducting sensor, other types of radiation sensors), and more particularly addresses neutron detection (activation sensor, gas filling sensor). In a second part, the author gives an overview of the state of the art of neutron measurement by fission chamber in a mock-up reactor (signal formation, processing and post-processing, associated measurements and uncertainties, return on experience of measurements by fission chamber on Masurca and Minerve research reactors). In a third part, he reports the optimization of two intrinsic parameters of this sensor: the thickness of fissile material deposit, and the pressure and nature of the filler gas. The fourth part addresses the improvement of measurement electronics and of post-processing methods which are used for result analysis. The fifth part deals with the optimization of spectrum index measurements by means of a fission chamber. The impact of each parameter is quantified. Results explain some inconsistencies noticed in measurements performed on the Minerve reactor in 2004, and allow the improvement of biases with computed values [fr

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

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

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

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

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.

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

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.

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)

Extraction of potential energy in charge asymmetry coordinate from experimental fission data

Energy Technology Data Exchange (ETDEWEB)

Pasca, H. [Joint Institute for Nuclear Research, Dubna (Russian Federation); ' ' Babes-Bolyai' ' Univ., Cluj-Napoca (Romania); Andreev, A.V.; Adamian, G.G. [Joint Institute for Nuclear Research, Dubna (Russian Federation); Antonenko, N.V. [Joint Institute for Nuclear Research, Dubna (Russian Federation); Tomsk Polytechnic Univ. (Russian Federation). Mathematical Physics Dept.

2016-12-15

For fissioning isotopes of Ra, Ac, Th, Pa, and U, the potential energies as a function of the charge asymmetry coordinate are extracted from the experimental charge distributions of the fission fragment and compared with the calculated scission-point driving potentials. The role of the potential energy surfaces in the description of the fission charge distribution is discussed. (orig.)

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.

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.

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)

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

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).

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

Experimental data on fission and (n,xn) reactions

International Nuclear Information System (INIS)

Belier, G.; Chatillon, A.; Granier, T.; Laborie, J.M.; Laurent, B.; Ledoux, X.; Taieb, J.; Varignon, C.; Bauge, E.; Bersillon, O.; Aupiais, J.; Le Petit, G.; Authier, N.; Casoli, P.

2011-01-01

Investigations on neutron-induced fission of actinides and the deuteron breakup are presented. Neutron-induced fission has been studied for 10 years at the WNR (Weapons Neutron Research) neutron facility of the Los Alamos Neutron Science Center (LANSCE). Thanks to this white neutron source the evolution of the prompt fission neutron energy spectra as a function of the incident neutron energy has been characterized in a single experiment up to 200 MeV incident energy. For some isotopes the prompt neutron multiplicity has been extracted. These experimental results demonstrated the effect on the mean neutron energy of the neutron emission before scission for energies higher than the neutron binding energy. This extensive program ( 235 U and 238 U, 239 Pu, 237 Np and 232 Th were measured) is completed by neutron spectra measurements on the CEA 4 MV accelerator. The D(n,2n) reaction is studied both theoretically and experimentally. The cross-section was calculated for several nucleon-nucleon interactions including the AV18 interaction. It has also been measured on the CEA 7 MV tandem accelerator at incident neutron energies up to 25 MeV. Uncertainties lower than 8% between 5 and 10 MeV were obtained. In particular these experiments have extended the measured domain for cross sections. (authors)

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

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

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.

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

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)

Elise: a new facility for unprecedented experimental nuclear fission studies

International Nuclear Information System (INIS)

Taieb, J.; Belier, G.; Chatillon, A.; Granier, T.; Kelic, A.; Ricciardi, V.; Schmidt, K.H.; Voss, B.; Coste-Delclaux, M.; Diop, C.; Jouanne, C.; Schmitt, C.; Aiche, M.; Czajkowski, S.; Jurado, B.; Audouin, L.; Peyre, J.; Rosier, P.; Tassan-Got, L.; Bertoumieux, E.; Dore, D.; Dupont, E.; Letourneau, A.; Panebianco, S.

2009-01-01

A novel experimental program aiming to study the properties of fragments and neutrons emitted in the fission process has been initiated. The experiment will be held at the ELISe electron-ion collider to be constructed at GSI, Darmstadt in the framework of the FAIR extension of the facility. The experiment will take advantage of the inverse kinematics allowing, in particular, a total mass and charge resolution for all fission fragments. (authors)

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)

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)

Characteristics of prompt fission gamma-ray emission - Experimental results and predictions

International Nuclear Information System (INIS)

Oberstedt, Andreas; Billnert, Robert; Oberstedt, Stephan

2014-01-01

Systematics from 2001, describing prompt fission gamma-ray spectra (PFGS) characteristics as function of mass and atomic number of the fissioning system, has been revisited and parameters have been revised based on recent experimental results. Although originally expressed for spontaneous and thermal neutron-induced fission, validity for fast neutrons was assumed and applied to predict PFGS characteristics for the reaction n + 238 U up to incident neutron energies of E n = 20 MeV. The results from this work are in good agreement with corresponding results from both model calculations and experiments. (authors)

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.)

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

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

International Nuclear Information System (INIS)

Balland, J.

2007-12-01

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

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

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.)

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)

Mathematical processing of experimental data on neutron yield from separate fission fragments

International Nuclear Information System (INIS)

Basova, B.G.; Rabinovich, A.D.; Ryazanov, D.K.

1975-01-01

The algorithm is described for processing the multi-dimensional experiments on measurements of prompt emission of neutrons from separate fission fragments. While processing the data the effect of a number of experimental corrections is correctly taken into account; random coincidence background, neutron spectrum, neutron detector efficiency, instrument angular resolution. On the basis of the described algorithm a program for BESM-4 computer is realized and the treatment of experimental data is performed according to the spontaneous fission of 252 Cf

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

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

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

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

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

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.

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

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)

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

Defect trap model of gas behaviour in UO2 fuel during irradiation

International Nuclear Information System (INIS)

Szuta, A.

2003-01-01

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

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

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.

Methodology and experimental setup for measuring short-lives fission product yields in actinides induced fission by charged particles

International Nuclear Information System (INIS)

Bellido, A.V.

1995-07-01

The theoretical principles and the laboratory set-up for the fission products yields measurements are described. The procedures for the experimental determinations are explain in detail. (author). 43 refs., 5 figs

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

Transient fission product release during reactor shutdown and startup

International Nuclear Information System (INIS)

Hunt, C.E.L.; Lewis, B.J.

1995-01-01

Sweep gas experiments performed at CRL from 1979 to 1985 have been analysed to determine the fraction of the fission product gas inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the xenon release from companion fuel elements and from a well documented experimental fuel bundle irradiated in the NRU reactor. The measured gas release could be matched to within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. (author)

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.)

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)

An experimental investigation of fission product release in SLOWPOKE-2 reactors - Data report

International Nuclear Information System (INIS)

Harnden, A.M.C.

1995-09-01

The results of an investigation into the release of fission products from SLOWPOKE-2 reactors fuelled with a highly-enriched uranium alloy core are detailed in Volume 1. This data report (Volume 2) contains plots of the activity concentrations of the fission products observed in the reactor container at the University of Toronto, Ecole Polytechnique and the Kanata Isotope Production Facility. Release rates from the reactor container water to the gas headspace are also included. (author)

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

Experimental study of fission process by fragment-neutron correlation measurement

Energy Technology Data Exchange (ETDEWEB)

Nishio, Katsuhisa; Yamamoto, Hideki; Kanno, Ikuo; Kimura, Itsuro; Nakagome, Yoshihiro [Kyoto Univ. (Japan). Faculty of Engineering

1997-07-01

Fragment-neutron correlation measurement of {sup 235}U(n{sub th}, f) was carried out. The obtained results showed more statistical accuracy than that of reported thermal neutron reaction. Experimental results and it`s analysis made clear the following facts. The minimum values of <{eta}> (m*) are shown at about 90 and 145 {mu} and <{eta}> (m*) showed the symmetrical form with an axis of symmetrical fission. This tendency is same as the distribution of {sup 252}Cf(s.f). -dV/dTKE(m*) indicates the saw-teethed distribution as same as <{nu}>(m*). The distribution seems depend on stiffness of fission fragment affected by the shell effect. The level density parameter a(m*) of fission fragment obtained from {sup 235}U(n{sub th}, f) expresses the saw-teethed distribution as same as that of {sup 252}Cf(s.f). This distribution can be explained by the empirical equation under consideration of the fission fragment depending on the shell effect and the collective motion. (S.Y.)

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)

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

Fission Product Experimental Program: Validation and Computational Analysis

Energy Technology Data Exchange (ETDEWEB)

Leclaire, N.; Ivanova, T.; Letang, E. [Inst Radioprotect and Surete Nucl, F-92262 Fontenay Aux Roses (France); Girault, E. [CEA Valduc, Serv Rech Neutron and Critcite, 21 - Is-sur-Tille (France); Thro, J. F. [AREVA NC, F-78000 Versailles (France)

2009-02-15

From 1998 to 2004, a series of critical experiments referred to as the fission product (FP) experimental program was performed at the Commissariat a l'Energie Atomique Valduc research facility. The experiments were designed by Institut de Radioprotection et de Surete Nucleaire (IRSN) and funded by AREVA NC and IRSN within the French program supporting development of a technical basis for burnup credit validation. The experiments were performed with the following six key fission products encountered in solution either individually or as mixtures: {sup 103}Rh, {sup 133}Cs, {sup nat}Nd, {sup 149}Sm, {sup 152}Sm, and {sup 155}Gd. The program aimed at compensating for the lack of information on critical experiments involving FPs and at establishing a basis for FPs credit validation. One hundred forty-five critical experiments were performed, evaluated, and analyzed with the French CRISTAL criticality safety package and the American SCALE5. 1 code system employing different cross-section libraries. The aim of the paper is to show the experimental data potential to improve the ability to perform validation of full burnup credit calculation. The paper describes three Phases of the experimental program; the results of preliminary evaluation, the calculation, and the sensitivity/uncertainty study of the FP experiments used to validate the APOLLO2-MORET 4 route in the CRISTAL criticality package for burnup credit applications. (authors)

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.

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)

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

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.

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)

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

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

Experimental Verification of the Fission Chamber Gamma Signal Suppression by the Campbelling Mode

International Nuclear Information System (INIS)

Vermeeren, L.; Weber, M.; Oriol, L.; Breaud, S.; Filliatre, P.; Geslot, B.; Jammes, C.; Normand, S.; Lescop, B.

2011-01-01

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

Experimental investigations relevant for hydrogen and fission product issues raised by the Fukushima accident

Directory of Open Access Journals (Sweden)

Sanjeev Gupta

2015-02-01

Full Text Available The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS, unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled

Experimental study of energy dependence of proton induced fission cross sections for heavy nuclei in the energy range 200-1000 MeV

Energy Technology Data Exchange (ETDEWEB)

Kotov, A.A.; Gavrikov, Yu.A.; Vaishnene, L.A.; Vovchenko, V.G.; Poliakov, V.V.; Fedorov, O.Ya.; Chestnov, Yu.A.; Shchetkovskiy, A.I [Petersburg Nuclear Physics Institute, Gatchina, Leningrad district, Orlova roscha 1, 188300 (Russian Federation); Fukahori, T. [Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki 319-1195 (Japan)

2005-07-01

The results of the total fission cross sections measurements for {sup nat}Pb, {sup 209}Bi, {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np and {sup 239}Pu nuclei at the energy proton range 200-1000 MeV are presented. Experiments were carried out at 1 GeV synchrocyclotron of Petersburg Nuclear Physics Institute (Gatchina). The measurement method is based on the registration in coincidence of both complementary fission fragments by two gas parallel plate avalanche counters, located at a short distance and opposite sides of investigated target. The insensitivity of parallel plate avalanche counters to neutron and light charged particles allowed us to place the counters together with target immediately in the proton beam providing a large solid angle acceptance for fission fragment registration and reliable identification of fission events. The proton flux on the target to be studied was determined by direct counting of protons by scintillation telescope. The measured energy dependence of the total fission cross sections is presented. Obtained results are compared with other experimental data as well as with calculation in the frame of the cascade evaporation model. (authors)

Fragment properties in the fission of 237Np with fast neutrons - an experimental investigation of fission dynamics

International Nuclear Information System (INIS)

Naqvi, A.A.

1980-03-01

Fission fragment properties such as mass distribution, kinetic energy distribution or number of prompt emitted neutrons as a function of fragment mass can be used to characterize the scission point configuration. The present experiment allows for the first time to investigate these quantities for neutron induced fission in the MeV range. In this way the influence of excitation energy of the saddle point deformation of the fissioning system ( 237 Np + n) can be studied. Neutrons with energies of 0.8 and 5.5 MeV were produced by the Karlsruhe pulsed 3MV Van de Graaff accelerator. Kinetic energies and velocities of correlated fragments were determined by solid state detectors using the time-of-flight technique. The experimentally determined distributions of fragment properties were compared to a recent model suggested by Wilkins et al. which assumes only relatively weak coupling between internal and collective degrees of freedom. At least qualitative agreement is found for most of the results. (orig.) [de

Experimental data on fission and (n,xn) reactions; Donnees experimentales de fission et de reactions (n,xn)

Energy Technology Data Exchange (ETDEWEB)

Belier, G.; Chatillon, A.; Granier, T.; Laborie, J.M.; Laurent, B.; Ledoux, X.; Taieb, J.; Varignon, C.; Bauge, E.; Bersillon, O.; Aupiais, J.; Le Petit, G. [CEA Bruyeres-le-Chatel, 91 (France); Authier, N.; Casoli, P. [CEA Valduc, 21 - Is-sur-Tille (France)

2011-07-15

Investigations on neutron-induced fission of actinides and the deuteron breakup are presented. Neutron-induced fission has been studied for 10 years at the WNR (Weapons Neutron Research) neutron facility of the Los Alamos Neutron Science Center (LANSCE). Thanks to this white neutron source the evolution of the prompt fission neutron energy spectra as a function of the incident neutron energy has been characterized in a single experiment up to 200 MeV incident energy. For some isotopes the prompt neutron multiplicity has been extracted. These experimental results demonstrated the effect on the mean neutron energy of the neutron emission before scission for energies higher than the neutron binding energy. This extensive program ({sup 235}U and {sup 238}U, {sup 239}Pu, {sup 237}Np and {sup 232}Th were measured) is completed by neutron spectra measurements on the CEA 4 MV accelerator. The D(n,2n) reaction is studied both theoretically and experimentally. The cross-section was calculated for several nucleon-nucleon interactions including the AV18 interaction. It has also been measured on the CEA 7 MV tandem accelerator at incident neutron energies up to 25 MeV. Uncertainties lower than 8% between 5 and 10 MeV were obtained. In particular these experiments have extended the measured domain for cross sections. (authors)

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

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.

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

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

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)

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.

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)

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

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...

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

New experimental approaches to investigate the fission dynamics

Energy Technology Data Exchange (ETDEWEB)

Benlliure, J., E-mail: j.benlliure@usc.es; Rodríguez-Sánchez, J. L.; Alvarez-Pol, H.; Ayyad, Y.; Cortina-Gil, D.; Paradela, C.; Pietras, B.; Ramos, D.; Vargas, J. [Universidade de Santiago de Compostela, 15782 Santiago de Compostela (Spain); Audouin, L.; Boutoux, G. [Institut de Physique Nucléaire d’Orsay, F-91406 Orsay (France); Bélier, G.; Chatillon, A.; Gorbinet, T.; Laurent, B.; Martin, J.-F.; Pellereau, E.; Taïeb, J. [CEA, DAM, DIF, F-91297 Arpajon (France); Casarejos, E. [Universidad de Vigo, E-36200 Vigo (Spain); Heinz, A. [Chalmers University of Technology, SE-412 96 Gothenburg (Sweden); and others

2016-07-07

The first ever achieved full identification of both fission fragments, in atomic and mass number, made it possible to define new observables sensitive to the fission dynamics along the fission path up to the scission point. Moreover, proton-induced fission of {sup 208}Pb at high energies offers optimal conditions for the investigation of dissipative, and transient effects, because of the high-excitation energy of the fissioning nuclei, its low angular momentum, and limited shape distortion by the reaction. In this work we show that the charge distribution of the final fission fragments can constrain the ground-to-saddle dynamics while the mass distribution is sensitive to the dynamics until the scission point.

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.

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

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.)

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

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....

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)

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

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.

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

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

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

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

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

[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

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

International Nuclear Information System (INIS)

Hambsch, F.J.

2004-01-01

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

Coulex fission of 234U, 235U, 237Np, and 238Np studied within the SOFIA experimental program

International Nuclear Information System (INIS)

Martin, Julie-Fiona

2014-01-01

SOFIA (Studies On FIssion with Aladin) is an experimental project which aims at systematically measuring the fission fragments' isotopic yields as well as their total kinetic energy, for a wide variety of fissioning nuclei. The PhD work presented in this dissertation takes part in the SOFIA project, and covers the fission of nuclei in the region of the actinides: 234 U, 235 U, 237 Np and 238 Np. The experiment is led at the heavy-ion accelerator GSI in Darmstadt, Germany. This facility provides intense relativistic primary beam of 238 U. A fragmentation reaction of the primary beam permits to create a secondary beam of radioactive ions, some of which the fission is studied. The ions of the secondary beam are sorted and identified through the FR-S (Fragment Separator), a high resolution recoil spectrometer which is tuned to select the ions of interest.The selected - fissile - ions then fly further to Cave-C, an experimental area where the fission experiment itself takes place. At the entrance of the cave, the secondary beam is excited by Coulomb interaction when flying through an target; the de-excitation process involves low-energy fission. Both fission fragments fly forward in the laboratory frame, due to the relativistic boost inferred from the fissioning nucleus.A complete recoil spectrometer has been designed and built by the SOFIA collaboration in the path of the fission fragments, around the existing ALADIN magnet. The identification of the fragments is performed by means of energy loss, time of flight and deviation in the magnet measurements. Both fission fragments are fully (in mass and charge) and simultaneously identified.This document reports on the analysis performed for (1) the identification of the fissioning system, (2) the identification of both fission fragments, on an event-by-event basis, and (3) the extraction of fission observables: yields, TKE, total prompt neutron multiplicity. These results, concerning the actinides, are discussed, and

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

Energy Technology Data Exchange (ETDEWEB)

Stradal, Karl Alfred

1970-10-15

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

Fission 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)

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

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

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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)

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)

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

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

A compound refining system for separation of gaseous fission products incorporated in a reprocessing pilot plant for spent fuel from neclear power stations

International Nuclear Information System (INIS)

Anon.

1981-01-01

In the V. G. Khlopin Radium Institute, a gas treatment experimental apparatus was installed to the SU-2 type experimental facility. The purpose is to solve variety of problems in the separation, collection and subsequent treatment for gaseous fission products and highly volatile fission products in spent fuel reprocessing. The experimental apparatus has the functions as follows: the measurement of air flow such as flow rate, pressure, total γ activity and krypton-85 content, preliminary air flow cleaning and drying removing aerosol, hydrogen fluoride and nitrogen oxide, and the trapping and analysis of gaseous fission products and highly volatile fission products in air flow. For the collection of these two types of fission products, a liquid absorbent and a solid adsorbent are used in series arrangement. (J.P.N.)

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

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

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

International Nuclear Information System (INIS)

Rao, C.B.; Raj, Baldev

1980-01-01

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

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

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.

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.

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)

Interpretation and modelling of fission product Ba and Mo releases from fuel

Science.gov (United States)

Brillant, G.

2010-02-01

The release mechanisms of two fission products (namely barium and molybdenum) in severe accident conditions are studied using the VERCORS experimental observations. Barium is observed to be mostly released under reducing conditions while molybdenum release is most observed under oxidizing conditions. As well, the volatility of some precipitates in fuel is evaluated by thermodynamic equilibrium calculations. The polymeric species (MoO 3) n are calculated to largely contribute to molybdenum partial pressure and barium volatility is greatly enhanced if the gas atmosphere is reducing. Analytical models of fission product release from fuel are proposed for barium and molybdenum. Finally, these models have been integrated in the ASTEC/ELSA code and validation calculations have been performed on several experimental tests.

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.

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

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

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

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)

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

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

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

Fission energy of uranium isotopes and transuranium elements

International Nuclear Information System (INIS)

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

1981-01-01

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

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)

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.

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

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)

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

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.

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.

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.)

Fission product behaviour in the primary circuit of an HTR

International Nuclear Information System (INIS)

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

1981-01-01

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

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

A model for the release of low-volatility fission products in oxidizing conditions

International Nuclear Information System (INIS)

Cox, D.S.; Hunt, C.E.L.; Liu, Z.; Keller, N.A.; Barrand, R.D.; O'Connor, R.F.

1991-07-01

A thermodynamic and kinetic model has been developed for calculating low-volatility fission-product releases from UO 2 at high temperatures in oxidizing conditions. Volatilization of the UO 2 matrix is assumed to be the rate controlling process. Oxidation kinetics of the UO 2 are modelled by either interfacial rate control, gas phase oxidant transport control, or solid-state diffusion of oxygen. The vapour pressure of UO 3 in equilibrium with the oxidizing fuel is calculated from thermodynamic data, and volatilization rates are determined using a model for forced convective mass transport. Low-volatility fission-product releases are calculated from the volume of vapourized fuel. Model calculations are conservative compared to experimental data for Zr, La, Ce and Nb fission-product releases from irradiated UO 2 exposed to air at 1973-2350 K. The implications of this conservatism are discussed in terms of possible rate control by processes other than convective mass transport of UO 3 . Coefficients for effective surface area (based on experimental data) and for heterogeneous rate controlling reaction kinetics are introduced to facilitate agreement between calculations and the experimental data.

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

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.

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)

Transient fission-product release during reactor shutdown and startup

International Nuclear Information System (INIS)

Hunt, C.E.L.; Lewis, B.J.; Dickson, L.W.

1997-12-01

Sweep-gas experiments performed at AECL's Chalk River Laboratories from 1979 to 1985 have been further analysed to determine the fraction of the gaseous fission-product inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the stable xenon release from companion fuel elements and from a well-documented experimental fuel bundle irradiated in the NRU reactor. The calculated gas release could be matched to the measured values within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. There was also limited information on the fraction of the radioactive iodine that was exposed, but not released, on reactor shutdown. An empirical equation is proposed for calculating this fraction. (author)

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

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.)

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

Validation by theoretical approach to the experimental estimation of efficiency for gamma spectrometry of gas in 100 ml standard flask

International Nuclear Information System (INIS)

Mohan, V.; Chudalayandi, K.; Sundaram, M.; Krishnamony, S.

1996-01-01

Estimation of gaseous activity forms an important component of air monitoring at Madras Atomic Power Station (MAPS). The gases of importance are argon 41 an air activation product and fission product noble gas xenon 133. For estimating the concentration, the experimental method is used in which a grab sample is collected in a 100 ml volumetric standard flask. The activity of gas is then computed by gamma spectrometry using a predetermined efficiency estimated experimentally. An attempt is made using theoretical approach to validate the experimental method of efficiency estimation. Two analytical models named relative flux model and absolute activity model were developed independently of each other. Attention is focussed on the efficiencies for 41 Ar and 133 Xe. Results show that the present method of sampling and analysis using 100 ml volumetric flask is adequate and acceptable. (author). 5 refs., 2 tabs

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.

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.

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

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.)

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.

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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)

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.)

Method of measurement of cross sections of heavy nuclei fission induced by intermediate energy protons

International Nuclear Information System (INIS)

Kotov, Alexander; Chtchetkovski, Alexander; Fedorov, Oleg; Gavrikov, Yuri; Chestnov, Yuri; Poliakov, Vladimir; Vaishnene, Larissa; Vovchenko, Vil; Fukahori, Tokio

2003-01-01

The purpose of this work is experimental studies of the energy dependence of the fission cross sections of heavy nuclei, nat Pb, 209 Bi, 232 Th, 233 U, 235 U, 238 U, 237 Np and 239 Pu, by protons at the energies from 200 to 1000 MeV. At present experiment the method based on use of the gas parallel plate avalanche counters (PPACs) for registration of complementary fission fragments in coincidence and the telescope of scintillation counters for direct counting of the incident protons on the target has been used. First preliminary results of the energy dependences of proton induced fission cross sections for nat Pb, 209 Bi, 235 U and 238 U are reported. (author)

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)

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)

Low-energy nuclear fission and our understanding of the nucleus

International Nuclear Information System (INIS)

Hall, H.L.; Hoffman, D.C.

1990-01-01

The interactions between experimental discoveries in low-energy nuclear fission and the theoretical understanding of the structure of the nucleus are reviewed. The history of this synergistic relationship begins with the discovery of fission, the development of the liquid-drop model and the experimental evidence for magic numbers, continues through the development of the shell model, the experimental discovery of shape isomerism, the double-humped fission barrier the spontaneous fission half-life disaster, the discovery of symmetric mass division in spontaneous fission and theoretical treatments based on different paths to scission. It concludes with a brief review of current experimental and theoretical understanding of low-energy fission and the prospects for future developments. (author) 150 refs.; 5 figs.; 1 tab

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)

Gamma Radiation from Fission Fragments Experimental Apparatus-Mass Spectrum Resolution

Energy Technology Data Exchange (ETDEWEB)

Higbie, Jack

1969-08-15

The gamma-radiation from fission fragments was studied as a function of the fragment mass. The mass was determined from the fragment energies using solid state detectors. The mass resolution which can be achieved by this method is treated in detail. The average initial fragment mass and the initial mass resolution is calculated as a function of the measured (apparent) mass yield for three different thicknesses of the fissile material deposit. This treatment gives a clear indication of those factors most important for good mass resolution work. A detailed description of the experimental apparatus is given in the appendices.

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

Theory of nuclear fission: a review

International Nuclear Information System (INIS)

Mosel, U.

1976-01-01

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

Fission fragment yields and total kinetic energy release in neutron-induced fission of235,238U,and239Pu

Science.gov (United States)

Tovesson, F.; Duke, D.; Geppert-Kleinrath, V.; Manning, B.; Mayorov, D.; Mosby, S.; Schmitt, K.

2018-03-01

Different aspects of the nuclear fission process have been studied at Los Alamos Neutron Science Center (LANSCE) using various instruments and experimental techniques. Properties of the fragments emitted in fission have been investigated using Frisch-grid ionization chambers, a Time Projection Chamber (TPC), and the SPIDER instrument which employs the 2v-2E method. These instruments and experimental techniques have been used to determine fission product mass yields, the energy dependent total kinetic energy (TKE) release, and anisotropy in neutron-induced fission of U-235, U-238 and Pu-239.

Fission fragment mass and angular distributions

Indian Academy of Sciences (India)

2015-07-22

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

Fission dynamics with systems of intermediate fissility

Indian Academy of Sciences (India)

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

Physics and chemistry of fission

International Nuclear Information System (INIS)

1979-01-01

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

Fission-induced recrystallization effect on intergranular bubble-driven swelling in U-Mo fuel

Energy Technology Data Exchange (ETDEWEB)

Liang, Linyun; Mei, Zhi-Gang; Yacout, Abdellatif M.

2017-10-01

We have developed a mesoscale phase-field model for studying the effect of recrystallization on the gas-bubble-driven swelling in irradiated U-Mo alloy fuel. The model can simulate the microstructural evolution of the intergranular gas bubbles on the grain boundaries as well as the recrystallization process. Our simulation results show that the intergranular gas-bubble-induced fuel swelling exhibits two stages: slow swelling kinetics before recrystallization and rapid swelling kinetics with recrystallization. We observe that the recrystallization can significantly expedite the formation and growth of gas bubbles at high fission densities. The reason is that the recrystallization process increases the nucleation probability of gas bubbles and reduces the diffusion time of fission gases from grain interior to grain boundaries by increasing the grain boundary area and decreasing the diffusion distance. The simulated gas bubble shape, size distribution, and density on the grain boundaries are consistent with experimental measurements. We investigate the effect of the recrystallization on the gas-bubble-driven fuel swelling in UMo through varying the initial grain size and grain aspect ratio. We conclude that the initial microstructure of fuel, such as grain size and grain aspect ratio, can be used to effectively control the recrystallization and therefore reduce the swelling in U-Mo fuel.

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.

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Fission yield measurements at IGISOL

Science.gov (United States)

Lantz, M.; Al-Adili, A.; Gorelov, D.; Jokinen, A.; Kolhinen, V. S.; Mattera, A.; Moore, I.; Penttilä, H.; Pomp, S.; Prokofiev, A. V.; Rakopoulos, V.; Rinta-Antila, S.; Simutkin, V.; Solders, A.

2016-06-01

The fission product yields are an important characteristic of the fission process. In fundamental physics, knowledge of the yield distributions is needed to better understand the fission process. For nuclear energy applications good knowledge of neutroninduced fission-product yields is important for the safe and efficient operation of nuclear power plants. With the Ion Guide Isotope Separator On-Line (IGISOL) technique, products of nuclear reactions are stopped in a buffer gas and then extracted and separated by mass. Thanks to the high resolving power of the JYFLTRAP Penning trap, at University of Jyväskylä, fission products can be isobarically separated, making it possible to measure relative independent fission yields. In some cases it is even possible to resolve isomeric states from the ground state, permitting measurements of isomeric yield ratios. So far the reactions U(p,f) and Th(p,f) have been studied using the IGISOL-JYFLTRAP facility. Recently, a neutron converter target has been developed utilizing the Be(p,xn) reaction. We here present the IGISOL-technique for fission yield measurements and some of the results from the measurements on proton induced fission. We also present the development of the neutron converter target, the characterization of the neutron field and the first tests with neutron-induced fission.

Fission yield measurements at IGISOL

Directory of Open Access Journals (Sweden)

Lantz M.

2016-01-01

Full Text Available The fission product yields are an important characteristic of the fission process. In fundamental physics, knowledge of the yield distributions is needed to better understand the fission process. For nuclear energy applications good knowledge of neutroninduced fission-product yields is important for the safe and efficient operation of nuclear power plants. With the Ion Guide Isotope Separator On-Line (IGISOL technique, products of nuclear reactions are stopped in a buffer gas and then extracted and separated by mass. Thanks to the high resolving power of the JYFLTRAP Penning trap, at University of Jyväskylä, fission products can be isobarically separated, making it possible to measure relative independent fission yields. In some cases it is even possible to resolve isomeric states from the ground state, permitting measurements of isomeric yield ratios. So far the reactions U(p,f and Th(p,f have been studied using the IGISOL-JYFLTRAP facility. Recently, a neutron converter target has been developed utilizing the Be(p,xn reaction. We here present the IGISOL-technique for fission yield measurements and some of the results from the measurements on proton induced fission. We also present the development of the neutron converter target, the characterization of the neutron field and the first tests with neutron-induced fission.

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)

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)

Nuclear fission and neutron-induced fission cross-sections

CERN Document Server

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

2013-01-01

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

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

Fission in intermediate energy heavy ion reactions

International Nuclear Information System (INIS)

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

1989-01-01

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

Fission in intermediate energy heavy ion reactions

International Nuclear Information System (INIS)

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

1989-01-01

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

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

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

International Nuclear Information System (INIS)

Hinde, D.J.

1991-01-01

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

Fission of heavy nuclei: microscopic study of fission barriers and fragments angular momentum; Fission des noyaux lourds: etude microscopique des barrieres de fission et du moment angulaire des fragments

Energy Technology Data Exchange (ETDEWEB)

Bonneau, L

2003-11-01

A lot of experimental data on nuclear fission has been being collected for the last 65 years, allowing theoreticians to confront their models with reality. The first part of this work is dedicated to the computation of fission barriers. We have extended the HF + BCS (Hartree Fock + Bandeen-Cooper-Schrieffer) method in order to include a new set of polynomials on which wave functions can be broken to, more accurately than on Hermite's polynomials in the 2 fragment configuration. The fission barriers of 26 heavy nuclei from Thorium-230 to Nobelium-256 have been assessed and compared to experimental data, it appears that differences are no greater than 1 MeV. We have discovered a neat correlation between the variation of the experimental fission lifetimes of even Fermium isotopes and the computed heights of second barriers. Moreover our model reproduces the hyper-deformed well of Thorium-230 with a good agreement on the well depth. The second part deals with the scission region. We have performed Hartree-Fock calculations in order to explore different ways of fragmentation. We have shown that the harmonic oscillator gives a valid description of such ways. In order to compute the mean value of J{sup 2} in the fragments we have been driven to propose an adequate definition of that quantity consistent with the non-locality property of the J{sup 2} operator. (A.C.)

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

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…

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

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

International Nuclear Information System (INIS)

Schnez, H.

1977-01-01

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

On the mechanism of fission neutron emission

International Nuclear Information System (INIS)

Maerten, H.; Richter, D.; Seeliger, D.

1986-01-01

This review represents the present knowledge of the mechanism of prompt fission neutron emission. Starting with a brief fission process characterization related with neutron emission, possible emission mechanisms are discussed. It is emphasized that the experimental study of special mechanisms, i.e. scission neutron emission processes, requires a sufficiently correct description of emission probabilities on the base of the main mechanism, i.e. the evaporation from fully accelerated fragments. Adequate statistical-model approaches have to account for the complexity of nuclear fission reflected by an intricate fragment distribution. The present picture of scission neutron emission is not clarified neither experimentally nor theoretically. Deduced data are contradictory and depend on the used analysis procedures often involving rough discriptions of evaporated-neutron distributions. The contribution of two secondary mechanisms of fission neutron emission, i.e. the neutron evaporation during fragment acceleration and neutron emission due to the decay of 5 He after ternary fission, is estimated. We summarize the recent progress of the theoretical description of fission neutron spectra in the framework of statistical models considering the standard spectrum of 252 Cf(sf) neutrons especially. The main experimental basis for the study of fission neutron emission is the accurate measurement of emission probabilities as a function of emission energy and angle (at least) as well as fragment parameters (mass number ratio and kinetic energy). The present status is evaluated. (author)

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

International Nuclear Information System (INIS)

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

1985-01-01

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

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

The study of prompt and delayed muon induced fission III. The ratios of prompt to delayed fission yields

NARCIS (Netherlands)

Rösel, Ch.; Hänscheid, H.; Hartfiel, J.; Mutius, von R.; Achard van Enschut, d' J.F.M.; David, P; Janszen, H.; Johansson, T.; Konijn, J.; Krogulski, T.; Laat, de C.T.A.M.; Paganetti, H.; Petitjean, C.; Polikanov, S.M.; Reist, H.W.; Risse, F.; Schaller, L.A.; Schrieder, W.; Sinha, A.K.; Taal, A.; Theobald, J.P.; Tibell, G.; Trautmann, N.

1993-01-01

The ratios of prompt to delayed fission yields for the isotopes U-233, U-234, U-235, U-236, U-238, Np-237, Pu-242, and Pu-244 and the fission probabilities relative to each other have been investigated experimentally. Using the value of the total fission probability for Np-237 the absolute

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.

Performance and first results of fission product release and transport provided by the VERDON facility

Energy Technology Data Exchange (ETDEWEB)

Gallais-During, A., E-mail: annelise.gallais-during@cea.fr [CEA, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Bonnin, J.; Malgouyres, P.-P. [CEA, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Morin, S. [IRSN, F-13108 Saint-Paul-lez-Durance (France); Bernard, S.; Gleizes, B.; Pontillon, Y.; Hanus, E.; Ducros, G. [CEA, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France)

2014-10-01

Highlights: • A new facility to perform experimental LWR severe accidents sequences is evaluated. • In the furnace a fuel sample is heated up to 2600 °C under a controlled gas atmosphere. • Innovative thermal gradient tubes are used to study fission product transport. • The new VERDON facility shows an excellent consistency with results from VERCORS. • Fission product re-vapourization results confirm the correct functioning of the gradient tubes. - Abstract: One of the most important areas of research concerning a hypothetical severe accident in a light water reactor (LWR) is determining the source term, i.e. quantifying the nature, release kinetics and global released fraction of the fission products (FPs) and other radioactive materials. In line with the former VERCORS programme to improve source term estimates, the new VERDON laboratory has recently been implemented at the CEA Cadarache Centre in the LECA-STAR facility. The present paper deals with the evaluation of the experimental equipment of this new VERDON laboratory (furnace, release and transport loops) and demonstrates its capability to perform experimental sequences representative of LWR severe accidents and to supply the databases necessary for source term assessments and FP behaviour modelling.

Modification of UO2 grain re-crystallization temperature in function of burn-up as a base for Vitanza experimental curve reconstruction

International Nuclear Information System (INIS)

Szuta, M.; Dąbrowski, L.

2013-01-01

Crossing the experimental critical fuel temperature dependent on burn-up, an onset of fission gas burst release is observed. This observed phenomena can be explained by assumption that the fission gas immobilization in the uranium dioxide irradiated to a fluency of greater than 10 19 fissions/cm 3 is mainly due to radiation induced chemical activity. Application of the “ab initio” method show that the bond energy of Xenon and Krypton is equal to –1.23 eV, and –3.42 eV respectively. Assuming further that the gas chemically bound can be released mainly in the process of re-crystallization and modifying the differential equation of Ainscough of grain growth by including the burn-up dependence and the experimental data of limiting grain size in function of the fuel temperature for the un-irradiated and irradiated fuel we can re-construct the experimental curve of Vitanza. (authors)

Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

Directory of Open Access Journals (Sweden)

Rainer Moormann

2008-01-01

Full Text Available Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA. The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

Fission track dating of volcanic glass: experimental evidence for the validity of the Size-Correction Method

International Nuclear Information System (INIS)

Bernardes, C.; Hadler Neto, J.C.; Lattes, C.M.G.; Araya, A.M.O.; Bigazzi, G.; Cesar, M.F.

1986-01-01

Two techniques may be employed for correcting thermally lowered fission track ages on glass material: the so called 'size-correcting method' and 'Plateau method'. Several results from fission track dating on obsidian were analysed in order to compare the model rising size-correction method with experimental evidences. The results from this work can be summarized as follows: 1) The assumption that mean size of spontaneous and induced etched tracks are equal on samples unaffected by partial fading is supported by experimental results. If reactor effects such as an enhancing of the etching rate in the irradiated fraction due to the radiation damage and/or to the fact that induced fission releases a quantity of energy slightly greater than spontaneous one exist, their influence on size-correction method is very small. 2) The above two correction techniques produce concordant results. 3) Several samples from the same obsidian, affected by 'instantaneous' as well as 'continuous' natural fading to different degrees were analysed: the curve showing decreasing of spontaneous track mean-size vs. fraction of spontaneous tracks lost by fading is in close agreement with the correction curve constructed for the same obsidian by imparting artificial thermal treatements on induced tracks. By the above points one can conclude that the assumptions on which size-correction method is based are well supported, at least in first approximation. (Author) [pt

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

Contribution to the study of nuclear fission

International Nuclear Information System (INIS)

Serot, O.

2009-09-01

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

Fission of heavy nuclei: microscopic study of fission barriers and fragments angular momentum

International Nuclear Information System (INIS)

Bonneau, L.

2003-11-01

A lot of experimental data on nuclear fission has been being collected for the last 65 years, allowing theoreticians to confront their models with reality. The first part of this work is dedicated to the computation of fission barriers. We have extended the HF + BCS (Hartree Fock + Bandeen-Cooper-Schrieffer) method in order to include a new set of polynomials on which wave functions can be broken to, more accurately than on Hermite's polynomials in the 2 fragment configuration. The fission barriers of 26 heavy nuclei from Thorium-230 to Nobelium-256 have been assessed and compared to experimental data, it appears that differences are no greater than 1 MeV. We have discovered a neat correlation between the variation of the experimental fission lifetimes of even Fermium isotopes and the computed heights of second barriers. Moreover our model reproduces the hyper-deformed well of Thorium-230 with a good agreement on the well depth. The second part deals with the scission region. We have performed Hartree-Fock calculations in order to explore different ways of fragmentation. We have shown that the harmonic oscillator gives a valid description of such ways. In order to compute the mean value of J 2 in the fragments we have been driven to propose an adequate definition of that quantity consistent with the non-locality property of the J 2 operator. (A.C.)

Fission-track ages and their geological interpretation

International Nuclear Information System (INIS)

Wagner, G.A.

1981-01-01

In fission-track dating, experimental procedures such as etching and thermal pre-treatment may strongly affect the age values determined and their geological interpretation. This peculiarity is due to the common phenomenon of partial fading of fossil (spontaneous-) fission tracks during a sample's geological history. The proper geological interpretation of the age data must take into account the specific experimental conditions, the stability characteristics and size distribution of fission tracks in the sample, the ages of co-existing minerals, and the independent information about the thermal history of the geological region. (author)

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.

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Fission of nuclei far from stability

International Nuclear Information System (INIS)

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

2000-11-01

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

Fission of heavy nuclei: microscopic study of fission barriers and fragments angular momentum; Fission des noyaux lourds: etude microscopique des barrieres de fission et du moment angulaire des fragments

Energy Technology Data Exchange (ETDEWEB)

Bonneau, L

2003-11-01

A lot of experimental data on nuclear fission has been being collected for the last 65 years, allowing theoreticians to confront their models with reality. The first part of this work is dedicated to the computation of fission barriers. We have extended the HF + BCS (Hartree Fock + Bandeen-Cooper-Schrieffer) method in order to include a new set of polynomials on which wave functions can be broken to, more accurately than on Hermite's polynomials in the 2 fragment configuration. The fission barriers of 26 heavy nuclei from Thorium-230 to Nobelium-256 have been assessed and compared to experimental data, it appears that differences are no greater than 1 MeV. We have discovered a neat correlation between the variation of the experimental fission lifetimes of even Fermium isotopes and the computed heights of second barriers. Moreover our model reproduces the hyper-deformed well of Thorium-230 with a good agreement on the well depth. The second part deals with the scission region. We have performed Hartree-Fock calculations in order to explore different ways of fragmentation. We have shown that the harmonic oscillator gives a valid description of such ways. In order to compute the mean value of J{sup 2} in the fragments we have been driven to propose an adequate definition of that quantity consistent with the non-locality property of the J{sup 2} operator. (A.C.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-25

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

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

Directory of Open Access Journals (Sweden)

Kaplan Abdullah

2015-01-01

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

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.

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

Methodology and experimental setup for measuring short-lives fission product yields in actinides induced fission by charged particles; Metodologia e montagem experimental para a medicao de rendimentos de produtos de fissao de meia vida curta na fissao de actinideos por particulas carregadas

Energy Technology Data Exchange (ETDEWEB)

Bellido, A.V.

1995-07-01

The theoretical principles and the laboratory set-up for the fission products yields measurements are described. The procedures for the experimental determinations are explain in detail. (author). 43 refs., 5 figs.

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.

Experimental Observation and Theoretical Description of Multisoliton Fission in Shallow Water

Science.gov (United States)

Trillo, S.; Deng, G.; Biondini, G.; Klein, M.; Clauss, G. F.; Chabchoub, A.; Onorato, M.

2016-09-01

We observe the dispersive breaking of cosine-type long waves [Phys. Rev. Lett. 15, 240 (1965)] in shallow water, characterizing the highly nonlinear "multisoliton" fission over variable conditions. We provide new insight into the interpretation of the results by analyzing the data in terms of the periodic inverse scattering transform for the Korteweg-de Vries equation. In a wide range of dispersion and nonlinearity, the data compare favorably with our analytical estimate, based on a rigorous WKB approach, of the number of emerging solitons. We are also able to observe experimentally the universal Fermi-Pasta-Ulam recurrence in the regime of moderately weak dispersion.

Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields

Energy Technology Data Exchange (ETDEWEB)

Casoli, P.; Authier, N. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Laurec, J.; Bauge, E.; Granier, T. [CEA, Centre DIF, 91297 Arpajon (France)

2011-07-01

In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

Adsorption properties of fission gases Xe and Kr on pristine and doped graphene: A first principle DFT study

Energy Technology Data Exchange (ETDEWEB)

Vazhappilly, Tijo, E-mail: tijoj@barc.gov.in [Theoretical Chemistry Section, Chemistry Group, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Ghanty, Tapan K., E-mail: tapang@barc.gov.in [Theoretical Chemistry Section, Chemistry Group, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Homi Bhabha National Institute, Training School Complex, Anushakti Nagar, Mumbai 400094 (India); Jagatap, B.N. [Theoretical Chemistry Section, Chemistry Group, Bhabha Atomic Research Centre, Mumbai 400 085 (India)

2017-07-15

Graphene has excellent adsorption properties due to large surface area and has been used in applications related to gas sorption and separation. The separation of radioactive noble gases using graphene is an interesting area of research relevant to nuclear waste management. Radioactive noble gases Xe and Kr are present in the off-gas streams from nuclear fission reactors and spent nuclear fuel reprocessing plants. The entrapment of these volatile fission gases is important in the context of nuclear safety. The separation of Xe from Kr is extremely difficult, and energy intensive cryogenic distillation is generally employed. Physisorption based separation techniques using porous materials is a cost effective alternative to expensive cryogenic distillation. Thus, adsorption of noble gases on graphene is relevant for fundamental understanding of physisorption process. The properties of graphene can be tuned by doping and incorporation of defects. In this regard, we study the binding affinity of Xe and Kr in pristine and doped graphene sheets. We employ first principle calculations using density functional theory, corrected for dispersion interactions. The structural parameters obtained from the current study show excellent agreement with the available theoretical and experimental observations on similar systems. Noble gas adsorption energies on pristine graphene match very well with the available literature. Our results show that the binding energy of fission gases Xe and Kr on graphene can be considerably improved through doping the lattice with a heteroatom. - Graphical abstract: The adsorption of radioactive fission gases Xe and Kr on pristine/doped graphene is an interesting topic in the context of nuclear waste management. Previous experimental and computational studies about Xe/Kr adsorption on graphene were limited to only on pristine graphene. The doping by hetero atom changes the electronic properties of graphene and creates active sites in the lattice. Based

Polyphase diffusion of fission products in graphite

International Nuclear Information System (INIS)

Dannert, V.

1989-05-01

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

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

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.)

Correlated Production and Analog Transport of Fission Neutrons and Photons using Fission Models FREYA, FIFRELIN and the Monte Carlo Code TRIPOLI-4® .

Science.gov (United States)

Verbeke, Jérôme M.; Petit, Odile; Chebboubi, Abdelhazize; Litaize, Olivier

2018-01-01

Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using

Systematics of Fission-Product Yields

International Nuclear Information System (INIS)

Wahl, A.C.

2002-01-01

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

Systematics of Fission-Product Yields

Energy Technology Data Exchange (ETDEWEB)

A.C. Wahl

2002-05-01

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

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)

Study on collection efficiency of fission products by spray: Experimental device and modelling

Energy Technology Data Exchange (ETDEWEB)

Ducret, D.; Roblot, D.; Vendel, J. [Institut de Protection et de Surete Nucleaire, Gif-Sur-Yvette (France); Billarand, Y. [ECCO Pharmacie et Chimie, Neuilly (France)

1997-08-01

Consequences of an hypothetical overheating reactor accident in nuclear power plants can be limited by spraying cold water drops into containment building. The spray reduces the pressure and the temperature levels by condensation of steam and leads to the washout of fission products (aerosols and gaseous iodine). The present study includes a large program devoted to the evaluation of realistic washout rates. An experimental device (named CARAIDAS) was designed and built in order to determine the collection efficiency of aerosols and iodine absorption by drops with representative conditions of post-accident atmosphere. This experimental device is presented in the paper and more particularly: (1) the experimental enclosure in which representative thermodynamic conditions can be achieved, (2) the monosized drops generator, the drops diameter measurement and the drops collector, (3) the cesium iodide aerosols generator and the aerosols measurements. Modelling of steam condensation on drops aerosols collection and iodine absorption are described. First experimental and code results on drops and aerosols behaviour are compared. 8 refs., 18 figs.

Improved fission neutron energy discrimination with {sup 4}He detectors through pulse filtering

Energy Technology Data Exchange (ETDEWEB)

Zhu, Ting, E-mail: ting.zhu@ufl.edu [University of Florida, Gainesville, FL (United States); Liang, Yinong; Rolison, Lucas; Barker, Cathleen; Lewis, Jason; Gokhale, Sasmit [University of Florida, Gainesville, FL (United States); Chandra, Rico [Arktis Radiation Detectors Ltd., Räffelstrasse 11, Zürich (Switzerland); Kiff, Scott [Sandia National Laboratories, CA (United States); Chung, Heejun [Korean Institute for Nuclear Nonproliferation and Control, 1534 Yuseong-daero, Yuseong-gu, Daejeon (Korea, Republic of); Ray, Heather; Baciak, James E.; Enqvist, Andreas; Jordan, Kelly A. [University of Florida, Gainesville, FL (United States)

2017-03-11

This paper presents experimental and computational techniques implemented for {sup 4}He gas scintillation detectors for induced fission neutron detection. Fission neutrons are produced when natural uranium samples are actively interrogated by 2.45 MeV deuterium-deuterium fusion reaction neutrons. Fission neutrons of energies greater than 2.45 MeV can be distinguished by their different scintillation pulse height spectra since {sup 4}He detectors retain incident fast neutron energy information. To enable the preferential detection of fast neutrons up to 10 MeV and suppress low-energy event counts, the detector photomultiplier gain is lowered and trigger threshold is increased. Pile-up and other unreliable events due to the interrogating neutron flux and background radiation are filtered out prior to the evaluation of pulse height spectra. With these problem-specific calibrations and data processing, the {sup 4}He detector's accuracy at discriminating fission neutrons up to 10 MeV is improved and verified with {sup 252}Cf spontaneous fission neutrons. Given the {sup 4}He detector's ability to differentiate fast neutron sources, this proof-of-concept active-interrogation measurement demonstrates the potential of special nuclear materials detection using a {sup 4}He fast neutron detection system.

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

International Nuclear Information System (INIS)

Nicoli, I.G.

1981-06-01

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

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

Model for fission-product calculations

International Nuclear Information System (INIS)

Smith, A.B.

1984-01-01

Many fission-product cross sections remain unmeasurable thus considerable reliance must be placed upon calculational interpolation and extrapolation from the few available measured cross sections. The vehicle, particularly for the lighter fission products, is the conventional optical-statistical model. The applied goals generally are: capture cross sections to 7 to 10% accuracies and inelastic-scattering cross sections to 25 to 50%. Comparisons of recent evaluations and experimental results indicate that these goals too often are far from being met, particularly in the area of inelastic scattering, and some of the evaluated fission-product cross sections are simply physically unreasonable. It is difficult to avoid the conclusion that the models employed in many of the evaluations are inappropriate and/or inappropriately used. In order to alleviate the above unfortunate situations, a regional optical-statistical (OM) model was sought with the goal of quantitative prediction of the cross sections of the lighter-mass (Z = 30-51) fission products. The first step toward that goal was the establishment of a reliable experimental data base consisting of energy-averaged neutron total and differential-scattering cross sections. The second step was the deduction of a regional model from the experimental data. It was assumed that a spherical OM is appropriate: a reasonable and practical assumption. The resulting OM then was verified against the measured data base. Finally, the physical character of the regional model is examined

Studies on fission with ALADIN. Precise and simultaneous measurement of fission yields, total kinetic energy and total prompt neutron multiplicity at GSI

International Nuclear Information System (INIS)

Martin, Julie-Fiona; Taieb, Julien; Chatillon, Audrey; Belier, Gilbert; Boutoux, Guillaume; Ebran, Adeline; Gorbinet, Thomas; Grente, Lucie; Laurent, Benoit; Pellereau, Eric; Alvarez-Pol, Hector; Ayyad, Yassid; Benlliure, Jose; Cortina Gil, Dolores; Caamano, Manuel; Fernandez Dominguez, Beatriz; Paradela, Carlos; Ramos, Diego; Rodriguez-Sanchez, Jose-Luis; Vargas, Jossitt; Audouin, Laurent; Tassan-Got, Laurent; Aumann, Thomas; Casarejos, Enrique; Farget, Fanny; Rodriguez-Tajes, Carme; Heinz, Andreas; Jurado, Beatriz; Kelic-Heil, Aleksandra; Kurz, Nikolaus; Nociforo, Chiara; Pietri, Stephane; Rossi, Dominic; Schmidt, Karl-Heinz; Simon, Haik; Voss, Bernd; Weick, Helmut

2015-01-01

A novel technique for fission studies, based on the inverse kinematics approach, is presented. Following pioneering work in the nineties, the SOFIA Collaboration has designed and built an experimental set-up dedicated to the simultaneous measurement of isotopic yields, total kinetic energies and total prompt neutron multiplicities, by fully identifying both fission fragments in coincidence, for the very first time. This experiment, performed at GSI, permits to study the fission of a wide variety of fissioning systems, ranging from mercury to neptunium, possibly far from the valley of stability. A first experiment, performed in 2012, has provided a large array of unprecedented data regarding the nuclear fission process. An excerpt of the results is presented. With this solid starter, further improvements of the experimental set-up are considered, which are consistent with the expected developments at the GSI facility, in order to measure more fission observables in coincidence. The completeness reached in the SOFIA data, permits to scrutinize the correlations between the interesting features of fission, offering a very detailed insight in this still unraveled mechanism. (orig.)

Studies on fission with ALADIN. Precise and simultaneous measurement of fission yields, total kinetic energy and total prompt neutron multiplicity at GSI

Energy Technology Data Exchange (ETDEWEB)

Martin, Julie-Fiona; Taieb, Julien; Chatillon, Audrey; Belier, Gilbert; Boutoux, Guillaume; Ebran, Adeline; Gorbinet, Thomas; Grente, Lucie; Laurent, Benoit; Pellereau, Eric [CEA DAM Bruyeres-le-Chatel, Arpajon (France); Alvarez-Pol, Hector; Ayyad, Yassid; Benlliure, Jose; Cortina Gil, Dolores; Caamano, Manuel; Fernandez Dominguez, Beatriz; Paradela, Carlos; Ramos, Diego; Rodriguez-Sanchez, Jose-Luis; Vargas, Jossitt [Universidad de Santiago de Compostela, Santiago de Compostela (Spain); Audouin, Laurent; Tassan-Got, Laurent [CNRS/IN2P3, IPNO, Orsay (France); Aumann, Thomas [Technische Universitaet Darmstadt, Darmstadt (Germany); Casarejos, Enrique [Universidad de Vigo, Vigo (Spain); Farget, Fanny; Rodriguez-Tajes, Carme [CNRS/IN2P3, GANIL, Caen (France); Heinz, Andreas [Chalmers University of Technology, Gothenburg (Sweden); Jurado, Beatriz [CNRS/IN2P3, CENBG, Gradignan (France); Kelic-Heil, Aleksandra; Kurz, Nikolaus; Nociforo, Chiara; Pietri, Stephane; Rossi, Dominic; Schmidt, Karl-Heinz; Simon, Haik; Voss, Bernd; Weick, Helmut [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany)

2015-12-15

A novel technique for fission studies, based on the inverse kinematics approach, is presented. Following pioneering work in the nineties, the SOFIA Collaboration has designed and built an experimental set-up dedicated to the simultaneous measurement of isotopic yields, total kinetic energies and total prompt neutron multiplicities, by fully identifying both fission fragments in coincidence, for the very first time. This experiment, performed at GSI, permits to study the fission of a wide variety of fissioning systems, ranging from mercury to neptunium, possibly far from the valley of stability. A first experiment, performed in 2012, has provided a large array of unprecedented data regarding the nuclear fission process. An excerpt of the results is presented. With this solid starter, further improvements of the experimental set-up are considered, which are consistent with the expected developments at the GSI facility, in order to measure more fission observables in coincidence. The completeness reached in the SOFIA data, permits to scrutinize the correlations between the interesting features of fission, offering a very detailed insight in this still unraveled mechanism. (orig.)

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.)

The multi-step prompt particle emission from fission fragments

International Nuclear Information System (INIS)

Zhivopistsev, A.; Oprea, C.; Oprea, I.

2003-01-01

The purpose of this work is the study of non-equilibrium high-energy gamma emission from 252 Cf. In the framework of the formalism of statistical multi-step compound processes in nuclear reactions. A relation was found between the shape of the high-energy part of the gamma spectrum and different mechanisms of excitation of the fission fragments. Agreement with experimental data for different groups of fission fragments was obtained. The analysis of the experimental high-energy part of gamma spectra yields information about the mechanism of excitation of fission fragments. The influence of dissipation of the deformation excess on intrinsic excitation of fission fragments was studied. (authors)

Simultaneous measurement of neutron-induced fission and capture cross sections for {sup 241}Am at neutron energies below fission threshold

Energy Technology Data Exchange (ETDEWEB)

Hirose, K., E-mail: hirose.kentaro@jaea.go.jp [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Nishio, K.; Makii, H.; Nishinaka, I.; Ota, S. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Nagayama, T. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Graduate School of Science and Engineering, Ibaraki University, Mito 310-0056 (Japan); Tamura, N. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Graduate School of Science and Technology, Niigata University, Niigata 950-2181 (Japan); Goto, S. [Graduate School of Science and Technology, Niigata University, Niigata 950-2181 (Japan); Andreyev, A.N. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Vermeulen, M.J. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Gillespie, S.; Barton, C. [Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Kimura, A.; Harada, H. [Nuclear Science and Engineering Center, JAEA, Tokai, Ibaraki 319-1195 (Japan); Meigo, S. [J-PARC Center, JAEA, Tokai, Ibaraki 319-1195 (Japan); Chiba, S. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo 152-8550 (Japan); Ohtsuki, T. [Research Reactor Institute, Kyoto University, Kumatori-cho S' ennangun,Osaka 590-0494 (Japan)

2017-06-01

Fission and capture reactions were simultaneously measured in the neutron-induced reactions of {sup 241}Am at the spallation neutron facility of the Japan Proton Accelerator Research Complex (J-PARC). Data for the neutron energy range of E{sub n}=0.1–20 eV were taken with the TOF method. The fission events were observed by detecting prompt neutrons accompanied by fission using liquid organic scintillators. The capture reaction was measured by detecting γ rays emitted in the deexcitation of the compound nuclei using the same detectors, where the prompt fission neutrons and capture γ rays were separated by a pulse shape analysis. The cross sections were obtained by normalizing the relative yields at the first resonance to evaluations or other experimental data. The ratio of the fission to capture cross sections at each resonance is compared with those from an evaluated nuclear data library and other experimental data. Some differences were found between the present values and the library/literature values at several resonances.

Dating by fission tracks in archaeology. 1. Principles and experimental methods

International Nuclear Information System (INIS)

Poupeau, G.; Zuleta, E.

1984-01-01

The principles of dating method by uranium fission tracks are shortly exposed. The conditions of application to the archaeology are discussed, in particular for the volcanic glasses, where the fossil fission tracks are often affected of a beginning of effacement. (L.C.) [pt

Natural gas hydrates. Experimental techniques and their applications

Energy Technology Data Exchange (ETDEWEB)

Ye, Yuguang; Liu, Changling (eds.) [Qingdao Institute of Marine Geology (China). Gas Hydrate Laboratory

2013-07-01

Focuses on gas hydrate experiment in laboratory. Intends to provide practical significant parameters for gas hydrate exploration and exploitation in the oceanic and permafrost environments. Consists of different themes that present up-to-date information on hydrate experiments. ''Natural Gas Hydrates: Experimental Techniques and Their Applications'' attempts to broadly integrate the most recent knowledge in the fields of hydrate experimental techniques in the laboratory. The book examines various experimental techniques in order to provide useful parameters for gas hydrate exploration and exploitation. It provides experimental techniques for gas hydrates, including the detection techniques, the thermo-physical properties, permeability and mechanical properties, geochemical abnormalities, stability and dissociation kinetics, exploitation conditions, as well as modern measurement technologies etc.

CACA-2: revised version of CACA-a heavy isotope and fission-product concentration calculational code for experimental irradiation capsules

International Nuclear Information System (INIS)

Allen, E.J.

1976-02-01

A computer program is described which calculates nuclide concentration histories, power or neutron flux histories, burnups, and fission-product birthrates for fueled experimental capsules subjected to neutron irradiations. Seventeen heavy nuclides in the chain from 232 Th to 242 Pu and a user-specified number of fission products are treated. A fourth-order Runge-Kutta calculational method solves the differential equations for nuclide concentrations as a function of time. For a particular problem, a user-specified number of fuel regions may be treated. A fuel region is described by volume, length, and specific irradiation history. A number of initial fuel compositions may be specified for each fuel region. The irradiation history for each fuel region can be divided into time intervals, and a constant power density or a time-dependent neutron flux is specified for each time interval. Also, an independent cross-section set may be selected for each time interval in each irradiation history. The fission-product birthrates for the first composition of each fuel region are summed to give the total fission-product birthrates for the problem

Research on in-pile release of fission products from coated particle fuels

International Nuclear Information System (INIS)

Fukuda, K.; Iwamoto, K.

1985-01-01

Coated particle fuels fabricated in accordance with VHTR (Very High Temperature gas-cooled Reactor) fuel design have been irradiated by both capsules and an in-pile gas loop (OGL-1), and data on the fission products release under irradiation were obtained for loose coated particles, fuel compacts and fuel rods in the temperature range between 800 deg. C and 1600 deg. C. For the fission gases, temperature- and time dependences of the fractional release(R/B) were measured. Relation between release and failure fraction of the coated particles was elucidated on the VHTR reference fuels. Also measured was tritium concentration in the helium coolant of OGL-1. In-pile release behavior of the metallic fission products was studied by measuring the activities of the fission products adsorbed in the graphite sleeves of the OGL-1 fuel rods and the graphite fuel container of the sweep gas capsules in the PIE. Investigation on palladium interaction with SiC coating layer was included. (author)

Energy partition in nuclear fission

International Nuclear Information System (INIS)

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

1990-01-01

A scission point model (two spheroid model TSM) including semi-empirical temperature-dependent shell correction energies for deformed fragments at scission is presented. It has been used to describe the mass-asymmetry-dependent partition of the total energy release on both fragments from spontaneous and induced fission. Characteristic trends of experimental fragment energy and neutron multiplicity data as function of incidence energy in the Th-Cf region of fissioning nuclei are well reproduced. Based on model applications, information on the energy dissipated during the descent from second saddle of fission barrier to scission point have been deduced. (author). 39 refs, 13 figs

Nuclear dynamics in heavy ion induced fusion-fission reactions

International Nuclear Information System (INIS)

Kapoor, S.S.

1992-01-01

Heavy ion induced fission and fission-like reactions evolve through a complex nuclear dynamics encountered in the medium energy nucleus-nucleus collisions. In the recent years, measurements of the fragment-neutron and fragment-charged particle angular correlations in heavy ion induced fusion-fission reactions, have provided new information on the dynamical times of nuclear deformations of the initial dinuclear complex to the fission saddle point and the scission point. From the studies of fragment angular distributions in heavy ion induced fission it has been possible to infer the relaxation times of the dinuclear complex in the K-degree of freedom and our recent measurements on the entrance channel dependence of fragment anisotropies have provided an experimental signature of the presence of fissions before K-equilibration. This paper reviews recent experimental and theoretical status of the above studies with particular regard to the questions relating to dynamical times, nuclear dissipation and the effect of nuclear dissipation on the K-distributions at the fission saddle in completely equilibrated compound nucleus. (author). 19 refs., 9 figs

Neutron-induced fission cross sections

International Nuclear Information System (INIS)

Weigmann, H.

1991-01-01

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

Nuclear data for neutron emission in the fission process

International Nuclear Information System (INIS)

Ganesan, S.

1991-11-01

This document contains the proceedings of the IAEA Consultants' Meeting on Nuclear Data for Neutron Emission in the Fission Process, Vienna, 22 - 24 October 1990. Included are the conclusions and recommendations reached at the meeting and the papers presented by the meeting participants. These papers provide a review of the status of experimental and theoretical data on neutron emission in spontaneous and neutron induced fission with reference to the data needs for reactor applications oriented towards actinide burner studies. The specific topics covered are the following: experimental measurements and theoretical predictions and evaluations of fission neutron energy spectra, average prompt fission neutron multiplicity, correlation in neutron emission from complementary fragments, neutron emission during acceleration of fission fragments, statistical properties of neutron rich nuclei by study of emission spectra of neutrons from the excited fission fragments, integral qualification of nu-bar for the major fissile isotopes, nu-bar total of 239 Pu and 235 U, and related problems. Refs figs and tabs

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

International Nuclear Information System (INIS)

Savic, P.

1989-01-01

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

IFPE/EFE-RO, Experimental Fuel Elements RO89 and RO51 in TRIGA 14 MW Reactor (INR-Pitesti)

International Nuclear Information System (INIS)

Paraschiv, Marius; Turnbull, J.A.

2002-01-01

Description of program or function: Romanian irradiation tests concerned with Candu type fuel elements behavior and with the limits of the design parameters. A particular feature of the Candu fuel project is the small plenum (void volume) added for relaxation of the fission gases, which are inherently released during the fuel irradiation. Two irradiation tests in the C2 device from the TRIGA 14 MW reactor were performed between the years 1985-1987. The tests were done to evaluate the effect of the fuel density on the time-evolution of the fission gas pressure. Experimental fuel elements were adequately instrumented with pressure transducers to follow the fission gas pressure changes during fuel irradiation. The first irradiation test was conducted on the fuel element coded No.89 whose main characteristics were the nominal values of the main fuel design parameters. The second one was conducted on the fuel element coded No.51. Because of the axial flux asymmetry inside the TRIGA reactor core, the experimental elements are shorter in length than the Candu fuel design. The irradiation tests consisted in evaluation of the time-evolution of the internal pressure from two experimental fuel elements having the main design characteristics as the Romanian Candu type fuel element design and to follow the dependence of the internal pressure of the fission gas on the fuel density

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

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

International Nuclear Information System (INIS)

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

1989-05-01

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

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)

Study of fission dynamics with the three-dimensional Langevin equations

Energy Technology Data Exchange (ETDEWEB)

Eslamizadeh, H. [Persian Gulf University, Department of Physics, Bushehr (Iran, Islamic Republic of)

2011-11-15

The dynamics of fission has been studied by solving one- and three-dimensional Langevin equations with dissipation generated through the chaos weighted wall and window friction formula. The average prescission neutron multiplicities, fission probabilities and the mean fission times have been calculated in a broad range of the excitation energy for compound nuclei {sup 210}Po and {sup 224}Th formed in the fusion-fission reactions {sup 4}He+{sup 206}Pb, {sup 16}O+{sup 208}Pb and results compared with the experimental data. The analysis of the results shows that the average prescission neutron multiplicities, fission probabilities and the mean fission times calculated by one- and three-dimensional Langevin equations are different from each other, and also the results obtained based on three-dimensional Langevin equations are in better agreement with the experimental data. (orig.)

Correlated prompt fission data in transport simulations

Science.gov (United States)

Talou, P.; Vogt, R.; Randrup, J.; Rising, M. E.; Pozzi, S. A.; Verbeke, J.; Andrews, M. T.; Clarke, S. D.; Jaffke, P.; Jandel, M.; Kawano, T.; Marcath, M. J.; Meierbachtol, K.; Nakae, L.; Rusev, G.; Sood, A.; Stetcu, I.; Walker, C.

2018-01-01

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n - n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ rays from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in

Correlated prompt fission data in transport simulations

Energy Technology Data Exchange (ETDEWEB)

Talou, P.; Jaffke, P.; Kawano, T.; Stetcu, I. [Los Alamos National Laboratory, Nuclear Physics Group, Theoretical Division, Los Alamos, NM (United States); Vogt, R. [Lawrence Livermore National Laboratory, Nuclear and Chemical Sciences Division, Livermore, CA (United States); University of California, Physics Department, Davis, CA (United States); Randrup, J. [Lawrence Berkeley National Laboratory, Nuclear Science Division, Berkeley, CA (United States); Rising, M.E.; Andrews, M.T.; Sood, A. [Los Alamos National Laboratory, Monte Carlo Methods, Codes, and Applications Group, Los Alamos, NM (United States); Pozzi, S.A.; Clarke, S.D.; Marcath, M.J. [University of Michigan, Department of Nuclear Engineering and Radiological Sciences, Ann Arbor, MI (United States); Verbeke, J.; Nakae, L. [Lawrence Livermore National Laboratory, Nuclear and Chemical Sciences Division, Livermore, CA (United States); Jandel, M. [Los Alamos National Laboratory, Nuclear and Radiochemistry Group, Los Alamos, NM (United States); University of Massachusetts, Department of Physics and Applied Physics, Lowell, MA (United States); Meierbachtol, K. [Los Alamos National Laboratory, Nuclear Engineering and Nonproliferation, Los Alamos, NM (United States); Rusev, G.; Walker, C. [Los Alamos National Laboratory, Nuclear and Radiochemistry Group, Los Alamos, NM (United States)

2018-01-15

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n-n, n-γ, and γ-γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ rays from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX-PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation

Fission fragment angular distributions in proton-induced fission of 209Bi (p,f) and 197Au (p,f)

International Nuclear Information System (INIS)

Soheily, S.; Noshad, H.; Lamehi-Rashti, M.

2002-01-01

The fission fragment angular distributions have been measured for proton-induced fission of 209 B i and 197 A u nuclei using surface barrier detectors at several energies between 25 MeV and 30 MeV. The experimental anisotropies are found to be in agreement with the predictions of the Standard Saddle-Point Statistical Model. The fission cross sections of 209 B i and 197 A u nuclei were also measured and compared with the previous works

Prediction of fission mass-yield distributions based on cross section calculations

International Nuclear Information System (INIS)

Hambsch, F.-J.; G.Vladuca; Tudora, Anabella; Oberstedt, S.; Ruskov, I.

2005-01-01

For the first time, fission mass-yield distributions have been predicted based on an extended statistical model for fission cross section calculations. In this model, the concept of the multi-modality of the fission process has been incorporated. The three most dominant fission modes, the two asymmetric standard I (S1) and standard II (S2) modes and the symmetric superlong (SL) mode are taken into account. De-convoluted fission cross sections for S1, S2 and SL modes for 235,238 U(n, f) and 237 Np(n, f), based on experimental branching ratios, were calculated for the first time in the incident neutron energy range from 0.01 to 5.5 MeV providing good agreement with the experimental fission cross section data. The branching ratios obtained from the modal fission cross section calculations have been used to deduce the corresponding fission yield distributions, including mean values also for incident neutron energies hitherto not accessible to experiment

Delayed β ray spectrum of 235U fission fragments

International Nuclear Information System (INIS)

Pascholati, P.R.

1973-01-01

The time-dependent electron spectra of fission fragments from the thermal-neutron-induced fission of 235 U are calculated. The Gross theory of nuclear beta decay is used to obtain the decay constant and individual electron spectra. The mean energy per fission carried by the electrons and the number of electrons per fission are also calculated. Comparison of these calculated spectra to experimental ones shows good agreements. (Author) [pt

Evaluation of the 235U prompt fission neutron spectrum including a detailed analysis of experimental data and improved model information

Science.gov (United States)

Neudecker, Denise; Talou, Patrick; Kahler, Albert C.; White, Morgan C.; Kawano, Toshihiko

2017-09-01

We present an evaluation of the 235U prompt fission neutron spectrum (PFNS) induced by thermal to 20-MeV neutrons. Experimental data and associated covariances were analyzed in detail. The incident energy dependence of the PFNS was modeled with an extended Los Alamos model combined with the Hauser-Feshbach and the exciton models. These models describe prompt fission, pre-fission compound nucleus and pre-equilibrium neutron emissions. The evaluated PFNS agree well with the experimental data included in this evaluation, preliminary data of the LANL and LLNL Chi-Nu measurement and recent evaluations by Capote et al. and Rising et al. However, they are softer than the ENDF/B-VII.1 (VII.1) and JENDL-4.0 PFNS for incident neutron energies up to 2 MeV. Simulated effective multiplication factors keff of the Godiva and Flattop-25 critical assemblies are further from the measured keff if the current data are used within VII.1 compared to using only VII.1 data. However, if this work is used with ENDF/B-VIII.0β2 data, simulated values of keff agree well with the measured ones.

Fuel performance under normal PWR conditions: A review of relevant experimental results and models

Science.gov (United States)

Charles, M.; Lemaignan, C.

1992-06-01

Experiments conducted at Grenoble (CEA/DRN) over the past 20 years in the field of nuclear fuel behaviour are reviewed. Of particular concern is the need to achieve a comprehensive understanding of and subsequently overcome the limitations associated with high burnup and load-following conditions (pellet-cladding interaction (PCI), fission gas release (FGR), water-side corrosion). A general view is given of the organization of research work as well as some experimental details (irradiation, postirradiation examination — PIE). Based on various experimental programmes (Cyrano, Medicis, Anemone, Furet, Tango, Contact, Cansar, Hatac, Flog, Decor), the main contributions of the thermomechanical behaviour of a PWR fuel rod are described: thermal conductivity, in-pile densification, swelling, fission gas release in steady state and moderate transient conditions, gap thermal conductance, formation of primary and secondary ridges under PCI conditions. Specific programmes (Gdgrif, Thermox, Grimox) are devoted to the behaviour of particular fuels (gadolinia-bearing fuel, MOX fuel). Moreover, microstructure-based studies have been undertaken on fission gas release (fine analysis of the bubble population inside irradiated fuel samples), and on cladding behaviour (PCI related studies on stress-corrosion cracking (SCO, irradiation effects on zircaloy microstructure).

Hefei experimental hybrid fusion-fission reactor conceptual design

International Nuclear Information System (INIS)

Qiu Lijian; Luan Guishi; Xu Qiang

1992-03-01

A new concept of hybrid reactor is introduced. It uses JET-like(Joint European Tokamak) device worked at sub-breakeven conditions, as a source of high energy neutrons to induce a blanket fission of depleted uranium. The solid breeding material and helium cooling technique are also used. It can produce 100 kg of 239 Pu per year by partial fission suppressed. The energy self-sustained of the fusion core is not necessary. Plasma temperature is maintained by external 20 MW ICRF (ion cyclotron resonance frequency) and 10 MW ECRF (electron cyclotron resonance frequency) heating. A steady state plasma current at 1.5 Ma is driven by 10 MW LHCD (lower hybrid current driven). Plasma density will be kept by pellet injection. ICRF can produce a high energy tail in ion distribution function and lead to significant enhancement of D-T reaction rate by 2 ∼ 5 times so that the neutron source strength reaches to the level of 1 x 10 19 n/s. This system is a passive system. It's power density is 10 W/cm 3 and the wall loading is 0.6 W/cm 2 that is the lower limitation of fusion and fission technology. From the calculation of neutrons it could always be in sub-critical and has intrinsic safety. The radiation damage and neutron flux distribution on the first wall are also analyzed. According to the conceptual design the application of this type hybrid reactor earlier is feasible

Fission dynamics in the proton induced fission of heavy nuclei

Energy Technology Data Exchange (ETDEWEB)

Rubchenya, V.A. E-mail: rubchen@phys.jyu.fi; Trzaska, W.H.; Itkis, I.M.; Itkis, M.G.; Kliman, J.; Kniajeva, G.N.; Kondratiev, N.A.; Kozulin, E.M.; Krupa, L.; Pokrovski, I.V.; Voskressenski, V.M.; Hanappe, F.; Materna, T.; Dorvaux, O.; Stuttge, L.; Chubarian, G.; Khlebnikov, S.V.; Vakhtin, D.N.; Lyapin, V.G

2004-04-05

Multi-parameter correlation study of the reaction {sup 242}Pu(p, f) at E{sub p} 13, 20 and 55 MeV has been carried out. Fission fragment mass and kinetic energy distributions and the double differential neutron spectra have been measured. It was observed that the two-humped shape of mass distributions prevailed up to highest proton energy. Manifestation of the nuclear shell Z 28 near fragment mass A{sub fr} = 70 has been detected. The experimental results were analyzed in the framework of a time-dependent statistical model with inclusion of nuclear friction effects in the fission process. The multi-parameter correlation study of the reaction.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Detection of outliers in gas centrifuge experimental data

International Nuclear Information System (INIS)

Andrade, Monica C.V.; Nascimento, Claudio A.O.

2005-01-01

Isotope separation in a gas centrifuge is a very complex process. Development and optimization of a gas centrifuge requires experimentation. These data contain experimental errors, and like other experimental data, there may be some gross errors, also known as outliers. The detection of outliers in gas centrifuge experimental data may be quite complicated because there is not enough repetition for precise statistical determination and the physical equations may be applied only on the control of the mass flows. Moreover, the concentrations are poorly predicted by phenomenological models. This paper presents the application of a three-layer feed-forward neural network to the detection of outliers in a very extensive experiment for the analysis of the separation performance of a gas centrifuge. (author)

The fast fission effect in a cylindrical fuel element

Energy Technology Data Exchange (ETDEWEB)

Carlvik, I; Pershagen, B

1959-06-15

A new formula for the fast fission factor is derived, which takes proper account to fast capture. The fission neutron spectrum is divided into two groups with constant fission cross section in one group and zero fission cross section in the other. The average total, elastic, inelastic and capture cross sections in the two groups are calculated. Different assumptions regarding anisotropic and inelastic scattering are investigated. The effects of backscattering from the moderator and fast fission in neighbouring fuel elements are pointed out. Formulas for the fast fission ratio and for the fast conversion ratio are derived. The calculated fast fission ratios are compared with experimental values. Curves are given for the fast fission factor in uranium metal and uranium oxide.

Dispersion of the Neutron Emission in U{sup 235} Fission

Science.gov (United States)

Feynman, R. P.; de Hoffmann, F.; Serber, R.

1955-01-01

Equations are developed which allow the calculation of the average number of neutrons per U{sup235} fission from experimental measurements. Experimental methods are described, the results of which give a value of (7.8 + 0.6){sup ½} neutrons per U{sup 235} thermal fission.

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

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)

An experimental investigation of fission product release in SLOWPOKE-2 reactors

International Nuclear Information System (INIS)

Harnden, A.M.C.

1995-09-01

Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

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)

14. International workshop on nuclear fission physics. Proceedings

International Nuclear Information System (INIS)

2000-01-01

The meetings on nuclear fission took place 12-15 October 1998 and was organized by Institute of Physics and Power Engineering. The aim of the workshop was to present and discuss main new both theoretical and experimental results obtained in the area of nuclear fission, dynamical feature, properties of fission fragments and complementary radiation. As usual the program of the workshop was designed to cover a wide range of physical phenomena - from low energy and spontaneous fission to fission of hot rotating nuclei and multifragmentation at intermediate and high energies. Reaction induced by slow and fast neutron, light and heavy ions were discussed [ru

Fission rates measured using high-energy gamma-rays from short half-life fission products in fresh and spent nuclear fuel

International Nuclear Information System (INIS)

Kroehnert, H.

2011-02-01

In recent years, higher discharge burn-ups and initial fuel enrichments have led to more and more heterogeneous core configurations in light water reactors (LWRs), especially at the beginning of cycle when fresh fuel assemblies are loaded next to highly burnt ones. As this trend is expected to continue in the future, the Paul Scherrer Institute has, in collaboration with the Swiss Association of Nuclear Utilities, swissnuclear, launched the experimental programme LIFE(at)PROTEUS. The LIFE(at)PROTEUS programme aims to better characterise interfaces between burnt and fresh UO 2 fuel assemblies in modern LWRs. Thereby, a novel experimental database is to be made available for enabling the validation of neutronics calculations of strongly heterogeneous LWR core configurations. During the programme, mixed fresh and highly burnt UO 2 fuel lattices will be investigated in the zero-power research reactor PROTEUS. One of the main types of investigations will be to irradiate the fuel in PROTEUS and measure the resulting fission rate distributions across the interface between fresh and burnt fuel zones. The measurement of fission rates in burnt fuel re-irradiated in a zero-power reactor requires, however, the development of new experimental techniques which are able to discriminate against the high intrinsic activity of the fuel. The principal goal of the present research work has been to develop such a new measurement technique. The selected approach is based on the detection of high-energy gamma-ray lines above the intrinsic background (i.e. above 2200 keV), which are emitted by short-lived fission products freshly created in the fuel. The fission products 88 Kr, 142 La, 138 Cs, 84 Br, 89 Rb, 95 Y, 90m Rb and 90 Rb, with half-lives between 2.6 min and 2.8 h, have been identified as potential candidates. During the present research work, the gamma-ray activity of short-lived fission products has, for the first time, been measured and quantitatively evaluated for re

Comparison of profess predictions of D-COM blind problem with the experimental observations

International Nuclear Information System (INIS)

Sah, D.N.; Venkatesh, D.; Ram Adasan, E.

1984-01-01

As a part of an IAEA co-ordinated research program D-COM, a code exercise was organised in May 1983, to investigate the predictive capability of fuel performance codes with respect to transient fission gas release. In this exercise the computer code PROFESS was used to calculate the irradiation performance of fuel pins of the D-COM Blind problem circulated to the participants in the exercise. Calculations of fuel centre temperature, fuel-clad gap conductance, fission gas release during steady state and transient, and fuel restructurings for all the fuel pins were made by PROFESS. Comparison of predictions with experimental findings showed good agreement for several performance parameters. The comparison also revealed some areas where improvement was desired in the modelling of fuel behaviour. A recalculation was carried out for all fuel pins after incorporating modifications and adding additional models in the code. This allowed better agreement between the predicted and observed values of fission gas release in the fuel pins. This paper presents the results of blind calculation, recalculation and sensitivity analysis carried out by PROFESS on the D-COM Blind Problem. The paper also gives a brief description of the models of fission gas release and fuel restructuring used in the calculation. (author)

Tip model of cold fission

International Nuclear Information System (INIS)

Goennenwein, F.; Boersig, B.

1991-01-01

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

Nuclear isomerism in fission fragments produced by the spontaneous fission of {sup 252}Cf; Isomerisme nucleaire dans les fragments de fission produits dans la fission spontanee du {sup 252}Cf

Energy Technology Data Exchange (ETDEWEB)

Gautherin, C

1997-09-01

This thesis is devoted to the study of the nuclear structure of neutron-rich nuclei, via the search of isomeric nuclear states. Neutron-rich nuclei were produced in the spontaneous fission of {sup 252}Cf. The experimental study of isomeric states in these nuclei was performed with the {gamma}-array EUROGAM II, coupled to an additional and original fission fragment detector composed by photovoltaic cells, SAPhIR. The photovoltaic cells are well adapted to detect low energy heavy ions and have good energy and time resolutions to obtain a good fission fragment detection. This experiment led to the discovery of new isomeric states in {sup 135}Xe, {sup 104}Mo, {sup 146,147,148}Ce and {sup 152,154,156}Nd, with lifetimes between 60 ns and 2 {mu}s. Level schemes of these nuclei have been completed. An interpretation of the isomeric states in the nuclei {sup 154,156}Nd and {sup 156,158}Sm was performed by Hartree-Fock-Bogolyubov calculations using the DIS Gogny force with two quasi-particles excitations. The confrontation with the experimental results led to an interpretation of these isomeric states as K-isomers. (author)

First wall material damage induced by fusion-fission neutron environment

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

Transfer-induced fission in inverse kinematics: Impact on experimental and evaluated nuclear data bases

Energy Technology Data Exchange (ETDEWEB)

Farget, F.; Schmidt, K.H.; Clement, E.; Delaune, O.; Derkx, X.; Dijon, A.; Golabek, C.; Lemasson, A.; Roger, T.; Schmitt, C. [CEA/DSM-CNRS/IN2P3, GANIL, Caen (France); Caamano, M.; Ramos, D.; Benlliure, J.; Cortina, D.; Fernandez-Dominguez, B.; Paradela, C. [Universidade de Santiago de Compostela, Santiago de Compostela (Spain); Rodriguez-Tajes, C. [CEA/DSM-CNRS/IN2P3, GANIL, Caen (France); Universidade de Santiago de Compostela, Santiago de Compostela (Spain); Audouin, L. [Universite Paris-Sud 11, CNRS/IN2P3, Institut de Physique Nucleaire, Orsay (France); Casarejos, E. [Universidade de Vigo, Vigo (Spain); Dore, D.; Salsac, M.D. [Centre de Saclay, CEA, Irfu, Gif-sur-Yvette (France); Gaudefroy, L. [CEA DAM Ile-de-France, BP 12, Bruyeres-le-Chatel (France); Heinz, A. [Chalmers Tekniska Hoegskola, Fundamental Fysik, Goeteborg (Sweden); Jurado, B. [Universite Bordeaux, CENBG, UMR 5797 CNRS/IN2P3, Gradignan (France)

2015-12-15

Inverse kinematics is a new tool to study nuclear fission. Its main advantage is the possibility to measure with an unmatched resolution the atomic number of fission fragments, leading to new observables in the properties of fission-fragment distributions. In addition to the resolution improvement, the study of fission based on nuclear collisions in inverse kinematics beneficiates from a larger view with respect to the neutron-induced fission, as in a single experiment the number of fissioning systems and the excitation energy range are widden. With the use of spectrometers, mass and kinetic-energy distributions may now be investigated as a function of the proton and neutron number sharing. The production of fissioning nuclei in transfer reactions allows studying the isotopic yields of fission fragments as a function of the excitation energy. The higher excitation energy resulting in the fusion reaction leading to the compound nucleus {sup 250}Cf at an excitation energy of 45MeV is also presented. With the use of inverse kinematics, the charge polarisation of fragments at scission is now revealed with high precision, and it is shown that it cannot be neglected, even at higher excitation energies. In addition, the kinematical properties of the fragments inform on the deformation configuration at scission. (orig.)

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)

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.

Fission fragment angular distributions and fission cross section validation

International Nuclear Information System (INIS)

Leong, Lou Sai

2013-01-01

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

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

Fission cross section measurements for minor actinides

Energy Technology Data Exchange (ETDEWEB)

Fursov, B. [IPPE, Obninsk (Russian Federation)

1997-03-01

The main task of this work is the measurement of fast neutron induced fission cross section for minor actinides of {sup 238}Pu, {sup 242m}Am, {sup 243,244,245,246,247,248}Cm. The task of the work is to increase the accuracy of data in MeV energy region. Basic experimental method, fissile samples, fission detectors and electronics, track detectors, alpha counting, neutron generation, fission rate measurement, corrections to the data and error analysis are presented in this paper. (author)

Analysis of effects of pellet-cladding bonding on trapping of the released fission gases in high burnup KKL BWR fuels

Energy Technology Data Exchange (ETDEWEB)

Brankov, Vladimir [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Khvostov, Grigori; Mikityuk, Konstantin [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Pautz, Andreas [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Restani, Renato; Abolhassani, Sousan [Laboratory for Nuclear Materials at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Ledergerber, Guido [Kernkraftwerk Leibstadt, 5325 Leibstadt (Switzerland); Wiesenack, Wolfgang [Institutt for Energiteknikk - OECD Halden Reactor Project, Os Allé 5, 1777 Halden (Norway)

2016-08-15

Highlights: • Explanation for the scatter in measured fission gas release in high-BU BWR fuel rods. • Partial fuel-clad bond layer formation in high-BU BWR fuel. • Hypothesis for fission gas trapping facilitated by the pellet-cladding bond layer. • Correlation between burnup asymmetry and the quantity of trapped fission gas. • Implications of the trapped FG in LOCA transient. - Abstract: The first part of the paper presents results of a numerical analysis of the fuel behavior during base irradiation in the Kernkraftwerk Leibstadt Boiling Water Reactor (KKL BWR) using EPRI’s FALCON code coupled to GRSW-A – an advanced model for fuel swelling and fission gas release. Post-irradiation examinations conducted at the Paul Scherrer Institute’s (PSI) hot laboratory gave evidence of a distinct circumferential non-uniformity of local burnup at pellet surfaces. For several fuel samples, intact pellet-cladding bonding areas on the high burnup sides of the pellets at high burnup above ∼70 MWd/kgU were observed. It is hypothesized that a part of the fission gases, which are expected to be released by those areas, can be trapped and do not reach the rod plenum. In this paper, a simple approach to modeling of fission gas trapping is employed which reveals a potential correlation between the position of the rod within the fuel assembly (and therefore the degree of circumferential burnup non-uniformity) and the degree of fission gas trapping. A model is suggested to correlate the amount of locally trapped gas with the integral of the local contact pressure and the degree of circumferential burnup non-uniformity. The model is calibrated with available measurements of FGR from rod puncturing at the level of the plenums. In future work, the hypothesis about the axial distribution of trapped fission gas will be extrapolated to the Loss-Of-Coolant Accident (LOCA) analysis as an attempt to explain the fission gas release observed in some samples fabricated from

Calculating the mass distribution of heavy nucleus fission product by neutrons

International Nuclear Information System (INIS)

Gudkov, A.N.; Koldobskij, A.B.; Kolobashkin, V.M.; Semenova, E.V.

1981-01-01

The technique of calculating the fission product mass yields by neutrons which are necessary for performing nucleus physical calculations in designing nuclear reactor cores is considered. The technique is based on the approximation of fission product mass distribution over the whole mass range by five Gauss functions. New analytical expressions for determining energy weights of used gaussians are proposed. The results of comparison of experimental data with calculated values for fission product mass obtained for reference processes in the capacity of which the fission reactions are chosen: 233 U, 235 U fission by thermal neutrons, 232 Th, 233 U, 235 U, 238 U by fission spectrum neutrons and 14 MeV neutrons and for 232 Th fission reactions by 11 MeV neutrons and 238 U by 7.7 MeV neutrons. On the basis of the analysis of results obtained the conclusion is drawn on a good agreement of fission product mass yield calculation values obtained using recommended values of mass distribution parameters with experimental data [ru

Methods to Collect, Compile, and Analyze Observed Short-lived Fission Product Gamma Data

Energy Technology Data Exchange (ETDEWEB)

Finn, Erin C.; Metz, Lori A.; Payne, Rosara F.; Friese, Judah I.; Greenwood, Lawrence R.; Kephart, Jeremy D.; Pierson, Bruce D.; Ellis, Tere A.

2011-09-29

A unique set of fission product gamma spectra was collected at short times (4 minutes to 1 week) on various fissionable materials. Gamma spectra were collected from the neutron-induced fission of uranium, neptunium, and plutonium isotopes at thermal, epithermal, fission spectrum, and 14-MeV neutron energies. This report describes the experimental methods used to produce and collect the gamma data, defines the experimental parameters for each method, and demonstrates the consistency of the measurements.

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

Manifestation of transient effects in fission induced by relativistic heavy-ion collisions

Energy Technology Data Exchange (ETDEWEB)

Jurado, B.; Schmitt, C.; Schmidt, K.H.; Benlliure, J. [Universidad de Santiago de Compostela (Spain); Junghans, A.R. [Forschungszentrum Rossendorf e.V. (FZR), Dresden (Germany)

2004-03-01

We examine the manifestation of transient effects in fission by analysing experimental data where fission is induced by peripheral heavy-ion collisions at relativistic energies. Available total nuclear fission cross sections of {sup 238}U at1.A GeV on gold and uranium targets are compared with a nuclear-reaction code, where transient effects in fission are modelled using different approximations to the numerical time-dependent fission-decay width: a new analytical description based on the solution of the Fokker-Planck equation and two widely used but less realistic descriptions, a step function and an exponential-like function. The experimental data are only reproduced when transient effects are considered. The deduced value of the dissipation strength {beta} depends strongly on the approximation applied for the time-dependent fission-decay width and is estimated to be of the order of 2 x 10{sup 21} s{sup -1}. A careful analysis sheds severe doubts on the use of the exponential-like in-growth function largely used in the past. Finally, we discuss which should be the characteristics of experimental observables to be most sensitive to transient effects in fission. (orig.)

Fuel performance evaluation for the CAFE experimental device

International Nuclear Information System (INIS)

Giovedi, Claudia; Hirota, Leandro T.

2011-01-01

Fuel rod cladding material is the second barrier to prevent the release of radioactive inventories in a PWR reactor. In this sense, an important safety aspect is to assess the fuel behavior under operational conditions. This can be made by means of fuel performance codes and confirmed by experimental measurements. In order to evaluate the fuel behavior of fuel rods in steady-state conditions, it was designed an experimental irradiation device, the Nuclear Fuel Irradiation Circuit (CAFE-Mod1). This device will allow controlling the surface rod temperature, to measure the power associated to the rod and the evolution of fission gas release for a typical PWR fuel pin. However, to support the experimental irradiation program, it is extremely important to simulate the experimental conditions using a fuel performance code. The aim of this paper is to evaluate some parameters and aspects related to the fuel rod behavior during the irradiation program. This evaluation was carried out by means of an adapted fuel performance code. Obtained results have shown that besides of the variation observed for parameters, such as, fuel temperature and fission gas release as a function of fuel enrichment level, the fuel rod integrity was preserved in all studied conditions. (author)

Qualitative assessment of the fission product release capability of ELOCA.Mk5

International Nuclear Information System (INIS)

Klein, M.E.; Carlucci, L.N.; Arimescu, V.I.

1995-01-01

A qualitative assessment of the fission product release capability of the ELOCA.Mk5 computer code was performed by simulating two transients from the sweep-gas experiment, FIO-133. Improved agreement between calculated and experimental trends in release was obtained by applying an interface pressure stress component to the pellet center. As well, results show that the current system for defining the reference temperature distribution for the thermal stress component is not always realistic. These results are being used in the development of a new, mechanistic pellet stress model. (author)

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)

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.)

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)

Yields of products from thermal-neutron induced fission of 235U

International Nuclear Information System (INIS)

Rudstam, G.; Aagaard, P.; Zwicky, H.U.

1985-01-01

Methods for fission yield determinations at an ISOL-system connected to a nuclear reactor have been developed. The present report contains detailed descriptions both of the experimental techniques and of the method used to correct the experimental yields for the decay of short-lived nuclear species in the delay between production and measurement. The methods have been applied to the determination of the fission yields of 40 fission products including 2 isometric pairs in the light mass region and those of 99 fission products including 25 isometric pairs or triplets in the heavy mass region. For 64 cases this is the first determination published. (author)

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

Measurements of isomeric yield ratios of fission products from proton-induced fission on natU and 232Th via direct ion counting

Directory of Open Access Journals (Sweden)

Rakopoulos Vasileios

2017-01-01

Full Text Available Independent isomeric yield ratios (IYR of 81Ge, 96Y, 97Y, 97Nb, 128Sn and 130Sn have been determined in the 25 MeV proton-induced fission of natU and 232Th. The measurements were performed at the Ion Guide Isotope Separator On-Line (IGISOL facility at the University of Jyväskylä. A direct ion counting measurement of the isomeric fission yield ratios was accomplished for the first time, registering the fission products in less than a second after their production. In addition, the IYRs of natU were measured by means of γ-spectroscopy in order to verify the consistency of the recently upgraded experimental setup. From the obtained results, indications of a dependence of the production rate on the fissioning system can be noticed. These data were compared with data available in the literature, whenever possible. Using the TALYS code and the experimentally obtained IYRs, we also deduced the average angular momentum of the fission fragments after scission.

A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

International Nuclear Information System (INIS)

Koch, M.; Kazimi, M.S.

1991-04-01

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

Energy Technology Data Exchange (ETDEWEB)

Koch, M.; Kazimi, M.S.

1991-04-01

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

Dynamic effect analysis in 240Pu fission at low energy

International Nuclear Information System (INIS)

Patin, Y.; Lachkar, J.; Sigaud, J.

1975-01-01

The variations of kinetic and excitation energies and fragment masses have been analyzed as a function of the fissioning nucleus excitation energy. Most interest has been taken in the fission of 240 Pu where many experimental data have been reported. The results tend, in the whole, to illustrate the existence of two modes of fission; the first one is superfluid, the other is strongly damped in the last stage of the fission process [fr

Theoretical Description of the Fission Process

International Nuclear Information System (INIS)

Nazarewicz, Witold

2009-01-01

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

Role of ternary fission in synthesis of bypassed nuclei

International Nuclear Information System (INIS)

Kramarovskij, Ya.M.; Chechev, V.P.

1983-01-01

A possible influence of ternary fission with escape of neutron-enriched light charged particles on the synthesis of bypassed nuclides is considered. It is shown that this concept cannot give explanation of bypassed isotope concentrations, but it can make some contribution, if the probability of ternary fission for superheavy nuclei grows sharply with Z 2 /A parameter. The account of β-delayed fission contributes to the shift of ternary fission fragments into the region of neutron-deficient isotopes. Consistent consideration of the ternary fission role in the nucleosynthesis is possible only with an important accumulation of experimental and theoretical data on this process, particularly for the nuclei with Z > 100

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)

COMEDIE BD1 experiment: Fission product behaviour during depressurization transients

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

Shmid, M.

1979-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-01

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

Simultaneous investigation of fission fragments and neutrons in 252Cf(s,f)

International Nuclear Information System (INIS)

Budtz-Joergensen, C.; Knitter, H.H.

1986-01-01

The gridded twin ion chamber developed at CBNM is used to measure the kinetic energy-, mass- and angular distributions of the fission fragments of 252 Cf in an advantageous 4π-geometry. Together with a neutron time-of-flight detector this experimental arrangement permits to measure the correlation between neutron emission, fragment angle, mass and energy in the spontaneous fission of 252 Cf. With the present experimental set-up a mass resolution for fission fragments of 0.5 a.m.u., an angular resolution of Δcosθ = 0.05 and a timing resolution of 0.7 ns FWHM were observed. Preliminary evaluations of the raw experimental data are presented for the fission fragment mass distribution, the average total kinetic energy and their variance as function of mass, the angular distribution between fragments and neutrons, the number of neutrons emitted per fragment as function of fragment mass, the average neutron emission energies as function of mass, and the prompt fission neutron spectrum averaged over all fragments. (author)

Simulation of fusion first-wall environment in a fission reactor

International Nuclear Information System (INIS)

Hassanein, A.M.; Kulcinski, G.L.; Longhurst, G.R.

1982-01-01

A novel concept to produce a realistic simulation of a fusion first-wall test environment has been proposed recently. This concept takes advantage of the (/eta/, α) reaction in 59 Ni to produce a high internal helium content in the metal while using the 3 He (/eta/, /rho/)T reaction in the gas surrounding the specimen to produce an external heat and particle flux. Models to calculate heat flux, erosion rate, implantation, and damage rate to the walls of the test module are presented. Preliminary results show that a number of important fusion technology issues could be tested experimentally in a fission reactor such as the Engineering Test Reactor

Fission rates measured using high-energy gamma-rays from short half-life fission products in fresh and spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Kroehnert, H.

2011-02-15

In recent years, higher discharge burn-ups and initial fuel enrichments have led to more and more heterogeneous core configurations in light water reactors (LWRs), especially at the beginning of cycle when fresh fuel assemblies are loaded next to highly burnt ones. As this trend is expected to continue in the future, the Paul Scherrer Institute has, in collaboration with the Swiss Association of Nuclear Utilities, swissnuclear, launched the experimental programme LIFE(at)PROTEUS. The LIFE(at)PROTEUS programme aims to better characterise interfaces between burnt and fresh UO{sub 2} fuel assemblies in modern LWRs. Thereby, a novel experimental database is to be made available for enabling the validation of neutronics calculations of strongly heterogeneous LWR core configurations. During the programme, mixed fresh and highly burnt UO{sub 2} fuel lattices will be investigated in the zero-power research reactor PROTEUS. One of the main types of investigations will be to irradiate the fuel in PROTEUS and measure the resulting fission rate distributions across the interface between fresh and burnt fuel zones. The measurement of fission rates in burnt fuel re-irradiated in a zero-power reactor requires, however, the development of new experimental techniques which are able to discriminate against the high intrinsic activity of the fuel. The principal goal of the present research work has been to develop such a new measurement technique. The selected approach is based on the detection of high-energy gamma-ray lines above the intrinsic background (i.e. above 2200 keV), which are emitted by short-lived fission products freshly created in the fuel. The fission products {sup 88}Kr, {sup 142}La, {sup 138}Cs, {sup 84}Br, {sup 89}Rb, {sup 95}Y, {sup 90m}Rb and {sup 90}Rb, with half-lives between 2.6 min and 2.8 h, have been identified as potential candidates. During the present research work, the gamma-ray activity of short-lived fission products has, for the first time, been

Studies of short-lived products of spallation fission reactions at TRIUMF

CERN Document Server

Bischoff, G; D'Auria, J M; Dautet, H; Lee, J K P; Pate, B D; Wiesehahn, W

1976-01-01

The gas-jet recoil transport technique has been used to transport products from spallation and fission reactions from a target chamber to a shielded location for nuclear spectroscopic studies. These involve X- beta - gamma coincidence measurements and (shortly) time- of-flight mass spectroscopy. It has been deduced that the proton beam at present intensities has no appreciable effect on the ability of ethylene and other cluster-producing gases to transport radioactivity. Preliminary results will be presented for shortlived fission products from uranium, and for spallation products of iodine and argon. The latter were obtained from the bombardment of gas and aerosol targets mixed with the transporting gas in the target chamber, which appears to be a generally useful technique.

Fission rate measurements in fuel plate type assembly reactor cores

International Nuclear Information System (INIS)

Rogers, J.W.

1988-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

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