WorldWideScience

Sample records for fission fuels

  1. 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)

  2. 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

  3. 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

  4. Fission gas release behaviour in MOX fuels

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  5. Fission gas retention in irradiated metallic fuel

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  6. 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

  7. 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.

  8. Impact of fuel chemistry on fission product behaviour

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  9. Fuel and fission product release from sodium

    International Nuclear Information System (INIS)

    Sauter, H.

    1992-01-01

    The NALA program at Kernforschungszentrum Karlsruhe is concerned with the release of fuel and fission products from hot or boiling sodium pools (radiological secondary source term) in a liquid-metal fast breeder reactor accident scenario with tank failure. The main concern is to determine retention factors (RF), to uncover the most essential parameters that influence the RF values, and to describe the way they do it. In the framework of the last NALA series, NALA IIIc, the influence of sodium-concrete interaction was investigated, partly with subsequent sodium burning. In our experiments, ∼3 kg of sodium and added pieces of concrete reaching from 4 to 40 g was used. The composition of the concrete was suitable for shielding and construction as used in the SNR-300 reactor. Fuel was simulated by 20-μm particles of depleted UO 2 , and CeO 2 , NaI, and TeO 2 were used as fission products. Most experiments were performed in an inert argon gas atmosphere with monitored hydrogen development. In some cases, the preheated pool was allowed to come into contact with ambient air, which caused an ordinary sodium fire. For the latter case, we used the 220-m 3 FAUNA vessel as an outer containment and collected the fire aerosols by a trap and subsequent filters for analysis

  10. 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

  11. Behavior of fission products released from severely damaged fuel during the PBF severe fuel damage tests

    International Nuclear Information System (INIS)

    Osetek, D.J.; Cronenberg, A.W.; Hagrman, D.L.; Broughton, J.M.; Rest, J.

    1984-01-01

    The results of fission product release behavior during the first two Power Burst Facility Severe Fuel Damage tests are presented. Measured fission product release is compared with calculated release using temperature dependent release rate correlations and FASTGRASS analysis. The test results indicate that release from fuel of the high volatility fission products (Xe, Kr, I, Cs, and Te) is strongly influenced by parameters other than fuel temperature; namely fuel/fission product morphology, fuel and cladding oxidation state, extent of fuel liquefaction, and quench induced fuel shattering. Fission product transport from the test fuel through the sample system was strongly influenced by chemical effects. Holdup of I and Cs was affected by fission product chemistry, and transport time while Te release was primarily influenced by the extent of zircaloy oxidation. Analysis demonstrates that such integral test data can be used to confirm physical, chemical, and mechanistic models of fission product behavior for severe accident conditions

  12. 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.

  13. (Fuel, fission product, and graphite technology)

    Energy Technology Data Exchange (ETDEWEB)

    Stansfield, O.M.

    1990-07-25

    Travel to the Forschungszentrum (KFA) -- Juelich described in this report was for the purpose of participating in the annual meeting of subprogram managers for the US/DOE Umbrella Agreement for Fuel, Fission Product, and Graphite Technology. At this meeting the highlights of the cooperative exchange were reviewed for the time period June 1989 through June 1990. The program continues to contribute technology in an effective way for both countries. Revision 15 of the Subprogram Plan will be issued as a result of the meeting. There was interest expressed by KFA management in the level of support received from the NPR program and in potential participation in the COMEDIE loop experiment being conducted at the CEA.

  14. 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)

  15. Separation of fission strontium from nuclear fuels by ammonium molybdatophosphate

    International Nuclear Information System (INIS)

    Hermann, A.

    1984-01-01

    The separation of fission strontium from solutions of prolonged cooled nuclear fuel has been performed using ammoniummolybdatophosphate (AMP) columns. The sorption mechanism of bivalent ions of AMP has been investigated by column and batch experiments. A pure ion exchange of Sr 2+ for two H + or NH 4 + ions has been established. Conditions for the sepasorption and elution of fission strontium and other fission products on AMP columns are described. (author)

  16. An Evaluation of a Fission Product Inventory for CANDU Fuels

    International Nuclear Information System (INIS)

    Jung, Jong Yeob; Park, Joo Hwan

    2007-01-01

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

  17. 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)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  19. Fission-product SiC reaction in HTGR fuel

    International Nuclear Information System (INIS)

    Montgomery, F.

    1981-01-01

    The primary barrier to release of fission product from any of the fuel types into the primary circuit of the HTGR are the coatings on the fuel particles. Both pyrolytic carbon and silicon carbide coatings are very effective in retaining fission gases under normal operating conditions. One of the possible performance limitations which has been observed in irradiation tests of TRISO fuel is chemical interaction of the SiC layer with fission products. This reaction reduces the thickness of the SiC layer in TRISO particles and can lead to release of fission products from the particles if the SiC layer is completely penetrated. The experimental section of this report describes the results of work at General Atomic concerning the reaction of fission products with silicon carbide. The discussion section describes data obtained by various laboratories and includes (1) a description of the fission products which have been found to react with SiC; (2) a description of the kinetics of silicon carbide thinning caused by fission product reaction during out-of-pile thermal gradient heating and the application of these kinetics to in-pile irradiation; and (3) a comparison of silicon carbide thinning in LEU and HEU fuels

  20. Behavior of solid fission products in irradiated fuel

    International Nuclear Information System (INIS)

    Song, Ung Sup; Jung, Yang Hong; Kim, Hee Moon; Yoo, Byun Gok; Kim, Do Sik; Choo, Yong Sun; Hong, Kwon Pyo

    2004-01-01

    Many fission products are generated by fission events in UO 2 fuel under irradiation in nuclear reactor. Concentration of each fission product is changed by conditions of neutron energy spectrum, fissile material, critical thermal power, irradiation period and cooling time. Volatile materials such as Cs and I, the fission products, degrade nuclear fuel rod by the decrease of thermal conductivity in pellet and the stress corrosion cracking in cladding. Metal fission products (white inclusion) make pellet be swelled and decrease volume of pellet by densification. It seems that metal fission products are filled in the pore in pellet and placed between UO 2 lattices as interstitial. In addition, metal oxide state may change structural lattice volume. Considering behavior of fission products mentioned above, concentration of them is important. Fission products could be classified as bellows; solid solution in matrix : Sr, Zr, Nb, Y, La, Ce, Pr, Nd, Pm, Sm - metal precipitates : Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sb, Te - oxide precipitates : Ba, Zr, Nb, Mo, (Rb, Cs, Te) - volatile and gases : Kr, Xe, Br, I, (Rb, Cs, Te)

  1. FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements

    International Nuclear Information System (INIS)

    Pierce, V.H.

    2005-01-01

    1 - Description of problem or function: The FREVAP type of code for estimating the release of longer-lived metallic fission products from HTGR fuel elements has been developed to take into account the combined effects of the retention of metallic fission products by fuel particles and the rather strong absorption of these fission products by the graphite of the fuel elements. Release calculations are made on the basis that the loss of fission product nuclides such as strontium, cesium, and barium is determined by their evaporation from the graphite surfaces and their transpiration induced by the flowing helium coolant. The code is devised so that changes of fission rate (fuel element power), fuel temperature, and graphite temperature may be incorporated into the calculation. Temperature is quite important in determining release because, in general, both release from fuel particles and loss by evaporation (transpiration) vary exponentially with the reciprocal of the absolute temperature. NESC0301/02: This version differs from the previous one in the following points: The source and output files were converted from BCD to ASCII coding. 2 - Method of solution: A problem is defined as having a one-dimensional segment made up of three parts - (1) the fission product source (fuel particles) in series with, (2) a non-source and absorption part (element graphite) and (3) a surface for evaporation to the coolant (graphite-helium interface). More than one segment may be connected (possibly segments stacked axially) by way of the coolant. At any given segment, a continuity equation is solved assuming equilibrium between the source term, absorption term, evaporation at coolant interface and the partial pressure of the fission product isotope in the coolant. 3 - Restrictions on the complexity of the problem - Maxima of: 5 isotopes; 10 time intervals for time-dependent variable; 49 segments (times number of isotopes); 5 different output print time-steps

  2. Radiation effects in fuel materials for fission reactors

    International Nuclear Information System (INIS)

    Matzke, H.

    1983-01-01

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

  3. Fission product release by fuel oxidation after water ingress

    International Nuclear Information System (INIS)

    Schreiber.

    1990-01-01

    On the basis of data obtained by a literature search, a computer code has been established for the calculation of the degree of oxidation of the fuel in the damaged fuel particles, and hence of the fission product release as a function of the time period of steam ingress. (orig.) [de

  4. Fission product release from TRIGA-LEU reactor fuels

    International Nuclear Information System (INIS)

    Baldwin, N.L.; Foushee, F.C.; Greenwood, J.S.

    1980-01-01

    Due to present international concerns over nuclear proliferation, TRIGA reactor fuels will utilize only low-enriched uranium (LEU) (enrichment <20%). This requires increased total uranium loading per unit volume of fuel in order to maintain the appropriate fissile loading. Tests were conducted to determine the fractional release of gaseous and metallic fission products from typical uranium-zirconium hydride TRIGA fuels containing up to 45 wt-% uranium. These tests, performed in late 1977 and early 1978, were similar to those conducted earlier on TRIGA fuels with 8.5 wt-% U. Fission gas release measurements were made on prototypic specimens from room temperature to 1100 deg. C in the TRIGA King Furnace Facility. The fuel specimens were irradiated in the TRIGA reactor at a low power level. The fractional releases of the gaseous nuclides of krypton and xenon were measured under steady-state operating conditions. Clean helium was used to sweep the fission gases released during irradiation from the furnace into a standard gas collection trap for gamma counting. The results of these tests on TRIGA-LEU fuel agree well with data from the similar, earlier tests on TRIGA fuel. The correlation used to calculate the release of fission products from 8.5 wt-% U TRIGA fuel applies equally well for U contents up to 45 wt-%. (author)

  5. Axial transport of fission gas in LWR fuel rods

    International Nuclear Information System (INIS)

    Kinoshita, M.

    1983-01-01

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

  6. Fuel 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

  7. Calculation of fission gases internal pressure in nuclear fuel rods

    International Nuclear Information System (INIS)

    Vasconcelos Santana, M. de.

    1981-12-01

    Models concerning the principal phenomena, particularly thermal expansion, fuel swelling, densification, reestructuring, relocation, mechanical strain, fission gas production and release, direct or indirectly important to calculate the internal pressure in nuclear fuel rods were analysed and selected. Through these analyses a computer code was developed to calculate fuel pin internal pressure evolution. Three different models were utilized to calculate the internal pressure in order to select the best and the most conservative estimate. (Author) [pt

  8. Fission product released experiment of coated fuel particles

    Energy Technology Data Exchange (ETDEWEB)

    Shijiang, Xu; Bing, Yang; Chunhe, Tang; Junguo, Zhu; Jintao, Huang; Binzhong, Zhang [Inst. of Nucl. Energy Technology, Tsinghua Univ., Beijing (China); Jinghan, Luo [Inst. of Atomic Energy, Beijing (China)

    1992-01-15

    Four samples of coated fuel particles were irradiated in the Heavy-Water Research Reactor of the Institute of Atomic Energy. Each of them was divided into two groups and irradiated to the burn up of 0.394% fima and 0.788% fima in two static capsules, respectively. After irradiation and cooling, post irradiation annealing experiment was carried out, the release ratios of the fission product {sup 133}Xe and {sup 131}I were measured, they are in the order of 10{sup -6}{approx}10{sup -7}. The fission product release ratio of naked kernel was also measured under the same conditions as for the coated fuel particles, the ratio of the fission product release of the coated fuel particles and of the naked kernel was in the order of 10{sup -5}{approx}10{sup -4}.

  9. 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.

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

    International Nuclear Information System (INIS)

    1997-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-01-02

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

  12. Fission induced swelling of U–Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Jeong, G.Y. [Ulsan National Institute of Science and Technology, 50 UNIST-gil, Eonyang-eup, Uljoo-gun, Ulsan 689-798 (Korea, Republic of); Park, J.M. [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2015-10-15

    Fission-induced swelling of U–Mo/Al dispersion fuel meat was measured using microscopy images obtained from post-irradiation examination. The data of reduced-size plate-type test samples and rod-type test samples were employed for this work. A model to predict the meat swelling of U–Mo/Al dispersion fuel was developed. This model is composed of several submodels including a model for interaction layer (IL) growth between U–Mo and Al matrix, a model for IL thickness to IL volume conversion, a correlation for the fission-induced swelling of U–Mo alloy particles, a correlation for the fission-induced swelling of IL, and models of U–Mo and Al consumption by IL growth. The model was validated using full-size plate data that were not included in the model development.

  13. Fission product release from HTGR coated microparticles and fuel elements

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  14. 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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  16. Reprocessing free nuclear fuel production via fusion fission hybrids

    Energy Technology Data Exchange (ETDEWEB)

    Kotschenreuther, Mike, E-mail: mtk@mail.utexas.edu [Intitute for Fusion Studies, University of Texas at Austin (United States); Valanju, Prashant; Mahajan, Swadesh [Intitute for Fusion Studies, University of Texas at Austin (United States)

    2012-05-15

    Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively 'new' cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th{sup 232}-U{sup 233} conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO{sub 2} matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U{sup 235} fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

  17. Reprocessing free nuclear fuel production via fusion fission hybrids

    International Nuclear Information System (INIS)

    Kotschenreuther, Mike; Valanju, Prashant; Mahajan, Swadesh

    2012-01-01

    Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th 232 –U 233 conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO 2 matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U 235 fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

  18. Fission product phases in irradiated carbide fuels

    International Nuclear Information System (INIS)

    Ewart, F.T.; Sharpe, B.M.; Taylor, R.G.

    1975-09-01

    Oxide fuels have been widely adopted as 'first charge' fuels for demonstration fast reactors. However, because of the improved breeding characteristics, carbides are being investigated in a number of laboratories as possible advanced fuels. Irradiation experiments on uranium and mixed uranium-plutonium carbides have been widely reported but the instances where segregate phases have been found and subjected to electron probe analysis are relatively few. Several observations of such segregate phases have now been made over a period of time and these are collected together in this document. Some seven fuel pins have been examined. Two of the irradiations were in thermal materials testing reactors (MTR); the remainder were experimental assemblies of carbide gas bonded oxycarbide and sodium bonded oxycarbide in the Dounreay Fast Reactor (DFR). All fuel pins completed their irradiation without failure. (author)

  19. Preliminary design and analysis on nuclear fuel cycle for fission-fusion hybrid spent fuel burner

    International Nuclear Information System (INIS)

    Chen Yan; Wang Minghuang; Jiang Jieqiong

    2012-01-01

    A wet-processing-based fuel cycle and a dry-processing were designed for a fission-fusion hybrid spent fuel burner (FDS-SFB). Mass flow of SFB was preliminarily analyzed. The feasibility analysis of initial loaded fuel inventory, recycle fuel fabrication and spent fuel reprocessing were preliminarily evaluated. The results of mass flow of FDS-SFB demonstrated that the initial loaded fuel inventory, recycle fuel fabrication and spent fuel reprocessing of nuclear fuel cycle of FDS-SFB is preliminarily feasible. (authors)

  20. Performance limits of coated particle fuel. Part III. Fission product migration in HTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nabielek, H.; Hick, H.; Wagner-Loffler, M.; Voice, E. H.

    1974-06-15

    A general introduction and literature survey to the physics and mathematics of fission product migration in HTR fuel is given as well as a review of available experimental results and their evaluation in terms of models and materials data.

  1. 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

  2. Measurement of fission product release during LWR fuel failure

    International Nuclear Information System (INIS)

    Osetek, D.J.; King, J.J.

    1979-01-01

    The PBF is a specialized test reactor consisting of an annular core and a central test space 21 cm in diameter and 91 cm high. A test loop circulates coolant through the central experimental section at typical power reactor conditions. Light-water-reactor-type fuel rods are exposed to power bursts simulating reactivity insertion transients, and to power-cooling-mismatch conditions during which the rods are allowed to operate in film boiling. Fission product concentrations in the test loop coolant are continuously monitored during these transients by a Ge(Li) detector based gamma spectrometer. Automatic batch processing of pulse height spectra results in a list of radionuclide concentrations present in the loop coolant as a function of time during the test. Fission product behavior is then correlated to test parameters and posttest examination of the fuel rods. Data are presented from Test PCM-1

  3. Recoil release of fission products from nuclear fuel

    International Nuclear Information System (INIS)

    Wise, C.

    1985-01-01

    An analytical approximation is developed for calculating recoil release from nuclear fuel into gas filled interspaces. This expression is evaluated for a number of interspace geometries and shown to be generally accurate to within about 10% by comparison with numerical calculations. The results are applied to situations of physical interest and it is demonstrated that recoil can be important when modelling fission product release from low temperature CAGR pin failures. Furthermore, recoil can contribute significantly in experiments on low temperature fission product release, particularly where oxidation enhancement of this release is measured by exposing the fuel to CO 2 . The calculations presented here are one way of allowing for this, other methods are suggested. (orig.)

  4. Release of fission products from miniature fuel plates at elevated temperature

    International Nuclear Information System (INIS)

    Posey, J.C.

    1982-01-01

    Three miniature fuel plates were tested at progressively higher temperatures. A U 3 Si plated blistered and released fission gases at 500 0 C. Two U 3 O 8 filled plates blistered and released fission gases at 550 0 C

  5. 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

  6. Fuel behavior and fission product release under HTGR accident conditions

    International Nuclear Information System (INIS)

    Fukuda, K.; Hayashi, K.; Shiba, K.

    1990-01-01

    In early 1989 a final decision was made over construction of a 30 MWth HTGR called the High Temperature Engineering Test Reactor, HTTR, in Japan in order to utilize it for high temperature gas engineering tests and various nuclear material tests. The HTTR fuel is a pin-in-block type fuel element which is composed of a hexagonal graphite block with dimension of 580 mm in length and 360 mm in face-to-face distance and about 30 of the fuel rods inserted into the coolant channels drilled in the block. The TRISO coated fuel particles for HTTR are incorporated with graphite powder and phenol resin into the fuel compacts, 19 of which are encased into a graphite sleeve as a fuel rod. It is necessary for the HTTR licensing to prove the fuel stability under predicted accidents related to the high temperature events. Therefore, the release of the fission products and the fuel failure have been investigated in the irradiation---and the heating experiments simulating these conditions at JAERI. This report describes the HTTR fuel behavior at extreme temperature, made clear in these experiments

  7. 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)

  8. Fission Product Release from Spent Nuclear Fuel During Melting

    International Nuclear Information System (INIS)

    Howell, J.P.; Zino, J.F.

    1998-09-01

    The Melt-Dilute process consolidates aluminum-clad spent nuclear fuel by melting the fuel assemblies and diluting the 235U content with depleted uranium to lower the enrichment. During the process, radioactive fission products whose boiling points are near the proposed 850 degrees C melting temperature can be released. This paper presents a review of fission product release data from uranium-aluminum alloy fuel developed from Severe Accident studies. In addition, scoping calculations using the ORIGEN-S computer code were made to estimate the radioactive inventories in typical research reactor fuel as a function of burnup, initial enrichment, and reactor operating history and shutdown time.Ten elements were identified from the inventory with boiling points below or near the 850 degrees C reference melting temperature. The isotopes 137Cs and 85Kr were considered most important. This review serves as basic data to the design and development of a furnace off-gas system for containment of the volatile species

  9. 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)

  10. 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

  11. 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

  12. A model for fission product distribution in CANDU fuel

    International Nuclear Information System (INIS)

    Muzumdar, A.P.

    1983-01-01

    This paper describes a model to estimate the distribution of active fission products among the UO 2 grains, grain-boundaries, and the free void spaces in CANDU fuel elements during normal operation. This distribution is required for the calculation of the potential release of activity from failed fuel sheaths during a loss-of-coolant accident. The activity residing in the free spaces (''free'' inventory) is available for release upon sheath rupture, whereas relatively high fuel temperatures and/or thermal shock are required to release the activity in the grain boundaries or grains. A preliminary comparison of the model with the data from in-reactor sweep-gas experiments performed in Canada yields generally good agreement, with overprediction rather than under prediction of radiologically important isotopes, such as I 131 . The model also appears to generally agree with the ''free'' inventory release calculated using ANS-5.4. (author)

  13. Measurement and characterization of fission products released from LWR fuel

    International Nuclear Information System (INIS)

    Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.

    1984-01-01

    Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. This project was sponsored by the USNRC under a broad program of reactor safety studies. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from approx. 2% at 1400 0 C to >50% at 2000 0 C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag

  14. Behavior of Nb fission product during nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Gue, J.P.

    1977-02-01

    Investigations on niobium fission product behavior in nitric acid and tributyl phosphate media have been carried out in order to explain the difficulties encountered in separating this element from fissile materials during spent nuclear fuel reprocessing. The studies have shown that in nitric acid solution, pentavalent niobium has a colloidal hydroxide form. The so-obtained sols were characterized by light scattering, electronic microscopy, electrophoresis and ultracentrifugation methods. In heterogeneous extracting media containing tributyl phosphate and dibutyl phosphoric acid the niobium hydroxide sols could be flocculated by low dibutyl phosphoric acid concentration or extracted into the organic phase containing an excess of dibutyl phosphoric acid [fr

  15. 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

  16. Measurement and characterization of fission products released from LWR fuel

    International Nuclear Information System (INIS)

    Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.

    1984-01-01

    Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from proportional 2% at 1400 0 C to >50% at 2000 0 C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag. (orig./HP)

  17. 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

  18. Comparison of 252Cf time correlated induced fission with AmLi induced fission on fresh MTR reserach reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, Jay Prakash [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-01

    The objectives of this project are to calibrate the Advanced Experimental Fuel Counter (AEFC), benchmark MCNP simulations using experimental results, investigate the effects of change in fuel assembly geometry, and finally to show the boost in doubles count rates with 252Cf active soruces due to the time correlated induced fission (TCIF) effect.

  19. 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

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

    International Nuclear Information System (INIS)

    Dias, Marcio S.

    1982-06-01

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

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

    International Nuclear Information System (INIS)

    Kleykamp, H.

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

  2. Fission induced swelling and creep of U–Mo alloy fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Hofman, G.L. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Cheon, J.S. [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong, Daejeon 305-353 (Korea, Republic of); Robinson, A.B.; Wachs, D.M. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2013-06-15

    Tapering of U–Mo alloy fuel at the end of plates is attributed to lateral mass transfer by fission induced creep, by which fuel mass is relocated away from the fuel end region where fission product induced fuel swelling is in fact the highest. This mechanism permits U–Mo fuel to achieve high burnup by effectively relieving stresses at the fuel end region, where peak stresses are otherwise expected because peak fission product induced fuel swelling occurs there. ABAQUS FEA was employed to examine whether the observed phenomenon can be simulated using physical–mechanical data available in the literature. The simulation results obtained for several plates with different fuel fabrication and loading scheme showed that the measured data were able to be simulated with a reasonable creep rate coefficient. The obtained creep rate constant lies between values for pure uranium and MOX, and is greater than all other ceramic uranium fuels.

  3. 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)

  4. Fission product release from defected nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Lewis, B.J.

    1983-01-01

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

  5. 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

  6. 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.

  7. Preliminary neutronics calculation of fusion-fission hybrid reactor breeding spent fuel assembly

    International Nuclear Information System (INIS)

    Ma Xubo; Chen Yixue; Gao Bin

    2013-01-01

    The possibility of using the fusion-fission hybrid reactor breeding spent fuel in PWR was preliminarily studied in this paper. According to the fusion-fission hybrid reactor breeding spent fuel characteristics, PWR assembly including fusion-fission hybrid reactor breeding spent fuel was designed. The parameters such as fuel temperature coefficient, moderator temperature coefficient and their variation were investigated. Results show that the neutron properties of uranium-based assembly and hybrid reactor breeding spent fuel assembly are similar. The design of this paper has a smaller uniformity coefficient of power at the same fissile isotope mass percentage. The results will provide technical support for the future fusion-fission hybrid reactor and PWR combined with cycle system. (authors)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  9. Release of fission products from miniature fuel plates at elevated temperature

    International Nuclear Information System (INIS)

    Posey, John C.

    1983-01-01

    Three miniature fuel plates were tested at progressively higher temperatures. A U 3 Si filled plate blistered and released fission gases at 500 deg. C. Two U 3 O 8 filled plates blistered and released fission gases at 550 deg. C. (author)

  10. On the behaviour of dissolved fission gases prior to transient testing of fuel pins

    International Nuclear Information System (INIS)

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

    1978-10-01

    The TREAT and CABRI series of reactor safety experiments on irradiated fuel require the transfer of fuel pins from the reactor in which the fuel has achieved some burn-up to the test facility. Subsequently, the fuel is restored to power in the test facility for some time before transient heating is initiated. Such pre-test manoeuvres, where the fuel is subjected to changes in the fission rate and temperature, may have important consequences for the fission gas behaviour during the transient experiment. The results of rate theory calculations are used to assess these effects. (author)

  11. 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

  12. Studies on the fission products behavior during dissolution process of BWR spent fuel

    International Nuclear Information System (INIS)

    Sato, K.; Nakai, E.; Kobayashi, Y.

    1987-01-01

    In order to obtain basic data on fission products behavior in connection with the head end process of fuel reprocessing, especially to obtain better understanding on undissolved residues, small scale dissolution studies were performed by using BWR spent fuel rods which were irradiated as monitoring fuel rods under the monitoring program for LWR fuel assembly performance entitled PROVING TEST ON RELIABILITY OF FUEL ASSEMBLY . The Zircaloy-2 claddings and the fuel pellets were subjected individually to the following studies on 1) release of fission products during dissolution process, 2) characterization of undissolved residues, and 3) analysis of the claddings. This paper presents comprehensive descriptions of the fission products behavior during dissolution process, based on detailed and through PIE conducted by JNFS under the sponsorship of MITI (Ministry of International Trade and Industry)

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

    International Nuclear Information System (INIS)

    Denis, Alicia C.

    1999-01-01

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

  14. Fundamental Studies of Irradiation-Induced Defect Formation and Fission Product Dynamics in Oxide Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Stubbins, James

    2012-12-19

    The objective of this research program is to address major nuclear fuels performance issues for the design and use of oxide-type fuels in the current and advanced nuclear reactor applications. Fuel performance is a major issue for extending fuel burn-up which has the added advantage of reducing the used fuel waste stream. It will also be a significant issue with respect to developing advanced fuel cycle processes where it may be possible to incorporate minor actinides in various fuel forms so that they can be 'burned' rather than join the used fuel waste stream. The potential to fission or transmute minor actinides and certain long-lived fission product isotopes would transform the high level waste storage strategy by removing the need to consider fuel storage on the millennium time scale.

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

    International Nuclear Information System (INIS)

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

    1995-03-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  18. Fission product release as a function of chemistry and fuel morphology

    International Nuclear Information System (INIS)

    Hobbins, R.R.; Osetek, D.J.; Petti, D.A.; Hagrman, D.L.

    1989-01-01

    Analysis of the consequences of severe reactor accidents requires knowledge of the location and chemical form of fission products throughout the accident sequence. Two factors that strongly influence the location and chemical form of fission products are the chemistry within the core and the morphology of the fuel or fuel-bearing debris. This paper reviews the current understanding of the these factors garnered from integral and separate effect experiments and the TMI-2 accident, and provides perspective on the significance of contributing phenomena for the analysis of severe accidents, particularly during the in-vessel phase. Information has been obtained recently on phenomena affecting the release of fission products from fuel and the reactor vessel during the in-vessel melt progression phase of a severe accident. The influence of a number of these phenomena will be reviewed, including fuel chemistry, H 2 /H 2 O ratio, fuel liquefaction, molten pools, and debris beds. 13 refs., 1 fig., 1 tab

  19. Fission product release measured during fuel damage tests at the Power Burst Facility

    International Nuclear Information System (INIS)

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

    1985-01-01

    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 quench 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 are offered for the probable reasons for the observed differences and recommendations for further studies are given

  20. 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)

  1. Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

    Science.gov (United States)

    Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

    2017-08-01

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

  2. 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)

  3. 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

  4. 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

  5. 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.

  6. 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.

  7. 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)

  8. Fission products and nuclear fuel behaviour under severe accident conditions part 3: Speciation of fission products in the VERDON-1 sample

    Science.gov (United States)

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

    2017-11-01

    Qualitative and quantitative analyses on the VERDON-1 sample made it possible to obtain valuable information on fission product behaviour in the fuel during the test. A promising methodology based on the quantitative results of post-test characterisations has been implemented to assess the release fraction of non γ-emitter fission products. The order of magnitude of the estimated release fractions for each fission product was consistent with their class of volatility.

  9. 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)

  10. 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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-04-07

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

  13. Fission product release in conditions of a spent fuel pool severe accident

    International Nuclear Information System (INIS)

    Ohai, Dumitru

    2007-01-01

    Full text: Depending on the residence time, fuel burnup, and fuel rack configuration, there may be sufficient decay heat for the fuel clad to heat up, swell, and burst in case of a loss of pool water. Initiating event categories can be: loss of offsite power from events initiated by severe weather, internal fire, loss of pool cooling, loss of coolant inventory, seismic event, aircraft impact, tornado, missile attack. The breach in the clad releases the radioactive gases present in the gap between the fuel and clad, what is called 'gap release'. If the fuel continues to heat up, the zirconium clad will reach the point of rapid oxidation in air. This reaction of zirconium and air, or zirconium and steam is exothermic. The energy released from the reaction, combined with the fuel's decay energy, can cause the reaction to become self-sustaining and ignite the zirconium. The increase in heat from the oxidation reaction can also raise the temperature in adjacent fuel assemblies and propagate the oxidation reaction. Simultaneously, the sintered UO 2 pellets resulting from pins destroying are oxidized. Due to the self-disintegration of pellets by oxidation, fission gases and low volatile fission products are released. The release rate, the chemical nature and the amount of fission products depend on powder granulation distribution and environmental conditions. The zirconium burning and pellets self-disintegration will result in a significant release of spent fuel fission products that will be dispersed from the reactor site. (author)

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

    International Nuclear Information System (INIS)

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

    1992-04-01

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

  15. 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

  16. 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...

  17. Fission product release modelling for application of fuel-failure monitoring and detection - An overview

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B.J., E-mail: lewibre@gmail.com [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, Ontario, K7K 7B4 (Canada); Chan, P.K.; El-Jaby, A. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, Ontario, K7K 7B4 (Canada); Iglesias, F.C.; Fitchett, A. [Candesco Division of Kinectrics Inc., 26 Wellington Street East, 3rd Floor, Toronto, Ontario M5E 1S2 (Canada)

    2017-06-15

    A review of fission product release theory is presented in support of fuel-failure monitoring analysis for the characterization and location of defective fuel. This work is used to describe: (i) the development of the steady-state Visual-DETECT code for coolant activity analysis to characterize failures in the core and the amount of tramp uranium; (ii) a generalization of this model in the STAR code for prediction of the time-dependent release of iodine and noble gas fission products to the coolant during reactor start-up, steady-state, shutdown, and bundle-shifting manoeuvres; (iii) an extension of the model to account for the release of fission products that are delayed-neutron precursors for assessment of fuel-failure location; and (iv) a simplification of the steady-state model to assess the methodology proposed by WANO for a fuel reliability indicator for water-cooled reactors.

  18. Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

    Energy Technology Data Exchange (ETDEWEB)

    Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

    2017-08-01

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (< 20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. This paper discusses the TEM results of the U-10Mo/Zr/Al6061 monolithic fuel plate (Plate ID: L1P09T, ~ 59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory as part of RERTR-9B irradiation campaign with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 C, respectively. A total of 5 TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (> 1 µm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ~ 30 at% and ~ 7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

  19. Theoretical analysis of knock-out release of fission products from nuclear fuels

    International Nuclear Information System (INIS)

    Yamagishi, S.

    1975-01-01

    The knock-out release of fission products is studied theoretically. The general equations of knock-out release are derived, assuming that a fission fragment passing through the surface of nuclear fuels knocks out a local region of the surface with an effective thickness and an effective cross-sectional area. Using these equations, the knock-out release of fission gases is calculated for various cases. The conditions under which the knock-out coefficients (the average number of uranium atoms knocked out by one fission fragment) is obtainable are clarified by experiments on the knock-out release of fission gases. A method of determining the effective thickness and the effective cross-sectional area of a knock-out region is proposed. (Auth.)

  20. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations

    International Nuclear Information System (INIS)

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-01-01

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k eff of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  2. 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)

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

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  4. Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives

    International Nuclear Information System (INIS)

    Foerthmann, R.; Groos, E.; Gruebmeier, H.

    1975-08-01

    Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Temperature Gas-cooled Reactors led to the development of coated particles with improved fission product retention of the kernel, which represent an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates SrAl 2 Si 2 O 8 , BaAl 2 Si 2 O 8 and CsAlSi 2 O 6 in the additional aluminasilica phase of the kernel. (orig.) [de

  5. Modeling of in-vessel fission product release including fuel morphology effects for severe accident analyses

    International Nuclear Information System (INIS)

    Suh, K.Y.

    1989-10-01

    A new in-vessel fission product release model has been developed and implemented to perform best-estimate calculations of realistic source terms including fuel morphology effects. The proposed bulk mass transfer correlation determines the product of fission product release and equiaxed grain size as a function of the inverse fuel temperature. The model accounts for the fuel-cladding interaction over the temperature range between 770 K and 3000 K in the steam environment. A separate driver has been developed for the in-vessel thermal hydraulic and fission product behavior models that were developed by the Department of Energy for the Modular Accident Analysis Package (MAAP). Calculational results of these models have been compared to the results of the Power Burst Facility Severe Fuel Damage tests. The code predictions utilizing the mass transfer correlation agreed with the experimentally determined fractional release rates during the course of the heatup, power hold, and cooldown phases of the high temperature transients. Compared to such conventional literature correlations as the steam oxidation model and the NUREG-0956 correlation, the mass transfer correlation resulted in lower and less rapid releases in closer agreement with the on-line and grab sample data from the Severe Fuel Damage tests. The proposed mass transfer correlation can be applied for best-estimate calculations of fission products release from the UO 2 fuel in both nominal and severe accident conditions. 15 refs., 10 figs., 2 tabs

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

    International Nuclear Information System (INIS)

    Mitenkova, E.; Novikov, N.

    2014-01-01

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

  7. Qualitative and quantitative characteristics of fission products in spent nuclear fuel from RBMK-type reactor

    International Nuclear Information System (INIS)

    Adlys, G.; Adliene, D.

    2002-01-01

    Well-known empirical models or experimental instruments and methods for the estimation of fission product yields do not allow prediction of the behavior and evaluation of the time-dependent qualitative and quantitative characteristics of all fission products in spent nuclear fuel during long-term storage. Several computer codes were developed in different countries to solve this problem. French codes APOLLO1 and PEPIN were used in this work for modeling the characteristics of spent nuclear fuel in the RBMK reactor. The modeling results of qualitative and quantitative characteristics of long-lived fission products for different cooling periods of spent nuclear fuel, including 50-year cooling period, are presented in this paper. The 50-year cooling period conforms to the foreseen time of storage of spent nuclear fuel in CONSTOR and CASTOR casks at the Ignalina NPP. These results correlate well with evaluated quantities for the well-known yields of the nuclides and could be used for the compilation of the database for long-lived fission products in spent nuclear fuel from the RBMK-type reactor. They allow one to predict and to solve effectively safety problems concerning with long-term spent nuclear fuel storage in casks. (author)

  8. 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.)

  9. 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)

  10. Fission product release from nuclear fuel I. Physical modelling in the ASTEC code

    International Nuclear Information System (INIS)

    Brillant, G.; Marchetto, C.; Plumecocq, W.

    2013-01-01

    Highlights: • Physical modeling of FP and SM release in ASTEC is presented. • The release is described as solid state diffusion within fuel for high volatile FP. • The release is described as FP vaporisation for semi volatile FP. • The release is described as fuel vaporisation for low volatile FP. • ASTEC validation is presented in the second paper. - Abstract: This article is the first of a series of two articles dedicated to the mechanisms of fission product release from a degraded core as they are modelled in the ASTEC code. The ASTEC code aims at simulating severe accidents in nuclear reactors from the initiating event up to the radiological consequences on the environment. This code is used for several applications such as nuclear plant safety evaluation including probabilistic studies and emergency preparedness. To cope with the requirements of robustness and low calculation time, the code is based on a semi-empirical approach and only the main limiting phenomena that govern the release from intact rods and from debris beds are considered. For solid fuel, fission products are classified into three groups, depending on their degree of volatility. The kinetics of volatile fission products release depend on the rate-limiting process of solid-state diffusion through fuel grains. For semi-volatile fission products, the release from the open fuel porosities is assumed to be governed by vaporisation and mass transfer processes. The key phenomenon for the release of low volatile fission products is supposed to be fuel volatilisation. A similar approach is used for the release of fission products from a rubble bed. An in-depth validation of the code including both analytical and integral experiments is the subject of the second article

  11. 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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  13. Advanced nuclear fuel production by using fission-fusion hybrid reactor

    International Nuclear Information System (INIS)

    Al-Kusayer, T.A.; Sahin, S.; Abdulraoof, M.

    1993-01-01

    Efforts are made at the College of Engineering, King Saud University, Riyadh to lay out the main structure of a prototype experimental fusion and fusion-fission (hybrid) reactor blanket in cylindrical geometry. The geometry is consistent with most of the current fusion and hybrid reactor design concepts in respect of the neutronic considerations. Characteristics of the fusion chamber, fusion neutrons and the blanket are provided. The studies have further shown that 1 GWe fission-fusion reactor can produce up to 957 kg/year which is enough to fuel five light water reactors of comparable power. Fuel production can be increased further. 29 refs

  14. 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

  15. 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

  16. Fission product release from high gap-inventory LWR fuel under LOCA conditions

    International Nuclear Information System (INIS)

    Lorenz, R.A.; Collins, J.L.; Osborne, M.F.; Malinauskas, A.P.

    1980-01-01

    Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200 0 C. The basis for the model was test data obtained with simulated fuel rods and commercial fuel irradiated to high burnup but containing relatively small amounts of cesium and iodine in the pellet-to-cladding gap space

  17. Measurements of Fission Cross Sections for the Isotopes relevant to the Thorium Fuel Cycle

    CERN Multimedia

    2002-01-01

    The present concern about a sustainable energy supply is characterised by a considerable uncertainty: the green house effect and foreseeable limits in fossil fuel resources on the one hand, the concern about the environmental impact of nuclear fission energy and the long term fusion research on the other hand, have led to the consideration of a variety of advanced strategies for the nuclear fuel cycle and related nuclear energy systems. The present research directories concern such strategies as the extension of the life span of presently operating reactors, the increase of the fuel burn-up, the plutonium recycling, and in particular the incineration of actinides and long-Lived fission products, the accelerator driven systems (ADS), like the "Energy Amplifier" (EA) concept of C. Rubbia, and the possible use of the Thorium fuel cycle. The detailed feasibility study and safety assessment of these strategies requires the accurate knowledge of neutron nuclear reaction data. Both, higher fuel burn-up and especiall...

  18. Contribution of metallic fission product inclusions to axial fuel motion potential

    International Nuclear Information System (INIS)

    Sasa, P.; Cronenberg, A.; Stevenson, M.

    1979-01-01

    In the analysis of postulated nuclear reactor accidents, axial fuel motion within the fuel pin prior to cladding failure can have an important mitigating effect. The question of primary importance is whether or not metallic inclusions have the potential to vaporize during an overheating event and thus contribute to fuel motion. To assess this potential, two limiting calculations were made: 1) The inclusion constituent assumed insoluble in one another and 2) The constituents assumed totally miscible in one another. Thermodynamic considerations indicate that the metallic fission products found within inclusions of fuel rods irradiated in a fast neutron spectrum, would form homogeneous solutions. Therefore, it is concluded that the metallic fission products would not enhance fuel swelling during an overheating event. 16 refs

  19. 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)

  20. Study of DD versus DT fusion fuel cycles for different fusion-fission hybrid energy systems

    International Nuclear Information System (INIS)

    Gohar, Y.; Baker, C.C.

    1981-01-01

    A study was performed to investigate the characteristics of an energy system to produce fissile fuel for fission reactors. DD and DT fusion reactors were examined in this study with either a thorium or uranium blanket for each fusion reactor. Various fuel cycles were examined for light-water reactors including the denatured fuel cycles (which may offer proliferation resistance compared to other fuel cycles); these fuel cycles include a uranium fuel cycle with 239 Pu makeup, a thorium fuel cycle with 239 Pu makeup, a denatured uranium fuel cycle with 233 U makeup, and a denatured thorium fuel cycle with 233 U makeup. Four different blankets were considered for this study. The first two blankets have a tritium breeding capability for DT reactors. Lithium oxide (Li 2 O) was used for tritium breeding due to its high lithium density and high temperature capability; however, the use of Li 2 O may result in higher tritium inventories compared to other solid breeders

  1. 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.

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

    Science.gov (United States)

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

    2018-05-01

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

  3. 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)

  4. 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

  5. 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

  6. An economic parametric analysis of the synthetic fuel produced by a fusion-fission complex

    International Nuclear Information System (INIS)

    Tai, A.S.; Krakowski, R.A.

    1980-01-01

    A simple analytic model is used to examine economic constraints of a fusion-fission complex in which a portion of a thermal energy is used for producing synthetic fuel (synfuel). Since the values of many quantities are not well-known, a parametric analysis has been carried out for testing the sensitivity of the synfuel production cost in relation to crucial economic and technological quantities (investment costs of hybrid and synfuel plants, energy multiplication of the fission blanket, recirculating power fraction of the fusion driver, etc.). In addition, a minimum synfuel selling price has been evaluated, from which the fission-fusion-synfuel complex brings about a higher economic benefit than does the fusion-fission hybrid entirely devoted to fissile-fuel and electricity generation. This paper describes the energy flow diagram of fusion-fission synfuel concept, express the revenue-to-cost formulation and the breakeven synfuel selling price. The synfuel production cost given by the model is evaluated within a range of values of crucial parameters. Assuming an electric cost of 2.7 cents/kWh, an annual investment cost per energy unit of 4.2 to 6 $/FJ for the fusion-fission complex and 1.5 to 3 $/GJ for the synfuel plant, the synfuel production cost lies between 6.5 and 8.5 $/GJ. These production costs can compete with those evaluated for other processes. The study points out a potential use of the fusion-fission hybrid reactor for other than fissile-fuel and electricity generation. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  8. Effects of burnup on fission product release and implications for severe fuel damage events

    International Nuclear Information System (INIS)

    Appelhans, A.D.; Cronenberg, A.W.; Carboneau, M.L.

    1984-01-01

    Xe, Kr, and I fission-product release data from (a) Halden tests where release in intact rods was measured during irradiation at burnups to 18,000 MWd/t and fuel temperatures of 800 to 1800 0 K, and (b) Power Burst Facility (PBF) tests where trace-irradiated fuel (approx. = 90 MWd/t) was driven to temperatures of >2400 0 K and fuel liquefaction occurred are discussed and related to fuel morphology. Results from both indicate that the fission-product morphology and fuel restructuring govern release behavior. The Halden tests show low release at beginning of life with a 10-fold increase at burnups in excess of 10,000 MWd/t, due to the development of grain boundary interlinkage at higher burnups. Such dependence of release on morphology characteristics is consistent with findings from the PBF tests, where for trace-irradiated fuel, the absence of interlinkage accounts for the low release rates observed during initial fuel heatup, with subsequent enhanced Xe, Kr, and I release via liquefaction or quench-induced destruction of the grain structure. Morphology is also shown to influence the chemical release form of I and Cs fission products

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

    International Nuclear Information System (INIS)

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

    1988-12-01

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

  10. 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

  11. Design of containment system of nuclear fuel attacked by corrosion with leaking fission products

    International Nuclear Information System (INIS)

    Poblete Maturana, Tomas

    2015-01-01

    The following report presents the design of an innovative confinement system for the nuclear fuel attacked by corrosion, with leakage of fission products to be used in the RECH-1 nuclear experimental reactor of the Chilean Nuclear Energy Commission, is currently within the framework of the international nuclear waste management program developed by the member countries of the IAEA, including Chile. The main objective of this project is the development of a system that is capable of containing, in the smallest possible volume, the fission products that are released to the reactor coolant medium from the nuclear fuel that are attacked by corrosion. Among the tasks carried out for the development of the project are: the compilation of the necessary bibliography for the selection of the most suitable technology for the retention of the fission products, the calculation of the most important parameters to ensure that the system will operate within ranges that do not compromise the radiological safety, and the design of the hydraulic circuit of the system. The results obtained from the calculations showed that the fuel element confinement system is stable from a thermal point of view since the refrigerant does not under any circumstances reach the saturation temperature and, in addition, from a hydraulic point of view, since the rate at which the refrigerant flows through the hydraulic circuit is low enough so that the deformation of the fuel plates forming the nuclear fuel does not occur. The most appropriate technology for the extraction of fission products according to the literature consulted is by ion exchange. The calculations developed showed that with a very small volume of resins, it is possible to capture all of the non-volatile fission products of a nuclear fuel

  12. Evaluation of DD and DT fusion fuel cycles for different fusion-fission energy systems

    International Nuclear Information System (INIS)

    Gohar, Y.

    1980-01-01

    A study has been carried out in order to investigate the characteristics of an energy system to produce a new source of fissile fuel for existing fission reactors. The denatured fuel cycles were used because it gives additional proliferation resistance compared to other fuel cycles. DT and DD fusion drivers were examined in this study with a thorium or uranium blanket for each fusion driver. Various fuel cycles were studied for light-water and heavy-water reactors. The cost of electricity for each energy system was calculated

  13. Safety analysis on tokamak helium cooling slab fuel fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Jian Hongbing

    1992-01-01

    The thermal analyses for steady state, depressurization and total loss of flow in the tokamak helium cooling slab fuel element fusion-fission hybrid reactor are presented. The design parameters, computed results of HYBRID program and safety evaluation for conception design are given. After all, it gives some recommendations for developing the design

  14. Grain boundary sweeping and dissolution effects on fission product behavior under severe fuel damage accident conditions

    International Nuclear Information System (INIS)

    Rest, J.

    1985-10-01

    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 behavior considers the migration and coalescence of fission gas bubbles in either molten uranium, or a zircaloy-uranium eutectic melt. 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 irradiated fuel are highlighted

  15. 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.)

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

    International Nuclear Information System (INIS)

    Fonteray, Jean

    1965-01-01

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

  17. Fuel depletion analyses for the HEU core of GHARR-1: Part II: Fission product inventory

    International Nuclear Information System (INIS)

    Anim-Sampong, S.; Akaho, E.H.K.; Boadu, H.O.; Intsiful, J.D.K.; Osae, S.

    1999-01-01

    The fission product isotopic inventories have been estimated for a 90.2% highly enriched uranium (HEU) fuel lattice cell of the Ghana Research Reactor-1 (GHARR-1) using the WIMSD/4 transport lattice code. The results indicate a gradual decrease in the Xe 135 inventory, and saturation trend for Sm 149 , Cs 134 and Cs 135 inventories as the fuel is depleted to 10,000 MWd/tU. (author)

  18. 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

  19. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-05-15

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

  1. 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

  2. A delayed neutron technique for measuring induced fission rates in fresh and burnt LWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K.A., E-mail: kajordan@gmail.co [Paul Scherrer Institut, Laboratory for Reactor Physics and System Behaviour, 5232 Villigen (Switzerland); Perret, G. [Paul Scherrer Institut, Laboratory for Reactor Physics and System Behaviour, 5232 Villigen (Switzerland)

    2011-04-01

    The LIFE-PROTEUS program at the Paul Scherrer Institut is being undertaken to characterize the interfaces between burnt and fresh fuel assemblies in modern LWRs. Techniques are being developed to measure fission rates in burnt fuel following re-irradiation in the zero-power PROTEUS research reactor. One such technique utilizes the measurement of delayed neutrons. To demonstrate the feasibility of the delayed neutron technique, fresh and burnt UO{sub 2} fuel samples were irradiated in different positions in the PROTEUS reactor, and their neutron outputs were recorded shortly after irradiation. Fission rate ratios of the same sample irradiated in two different positions (inter-positional) and of two different samples irradiated in the same position (inter-sample) were derived from the measurements and compared with Monte Carlo predictions. Derivation of fission rate ratios from the delayed neutron measured signal requires correcting the signal for the delayed neutron source properties, the efficiency of the measurement setup, and the time dependency of the signal. In particular, delayed neutron source properties strongly depend on the fissile and fertile isotopes present in the irradiated sample and must be accounted for when deriving inter-sample fission rate ratios. Measured inter-positional fission rate ratios generally agree within 1{sigma} uncertainty (on the order of 1.0%) with the calculation predictions. For a particular irradiation position, however, a bias of about 2% is observed and is currently under investigation. Calculated and measured inter-sample fission rate ratios have C/E values deviating from unity by less than 1% and within 2{sigma} of the statistical uncertainties. Uncertainty arising from delayed neutron data is also assessed, and is found to give an additional 3% uncertainty factor. The measurement data indicate that uncertainty is overestimated.

  3. Characterization and chemistry of fission products released from LWR fuel under accident conditions

    International Nuclear Information System (INIS)

    Norwood, K.S.; Collins, J.L.; Osborne, M.F.; Lorenz, R.A.; Wichner, R.P.

    1984-01-01

    Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 2000 0 C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released fission products with emphasis on antimony, cesium, iodine, and silver. The iodine appeared to behave primarily as cesium iodide and the antimony and silver as elements, while cesium behavior was much more complex. 17 refs., 7 figs., 1 tab

  4. Fission product behavior in HTGR fuel particles made from weak-acid resins

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Henson, T.J.

    1979-04-01

    Fission product retention and behavior are of utmost importance in HTGR fuel particles. The present study concentrates on particles made from weak-acid resins, which can vary in composition from 100% UO 2 plus excess carbon to 100% UC 2 plus excess carbon. Five compositions were tested: UC 4 58 O 2 04 , UC 3 68 O 0 01 , UC 4 39 O 1 72 , UC 4 63 O 0 97 , and UC 4 14 O 1 53 . Metallographically sectioned particles were examined with a shielded electron microprobe. The distributions of the fission products were determined by monitoring characteristic x-ray lines while scanning the electron beam over the particle surface

  5. 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)

  6. 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

  7. Potential for the use of hydrochloric acid in fission reactor fuel recycle

    International Nuclear Information System (INIS)

    Mailen, J.C.; Bell, J.T.

    1985-01-01

    The chemistry and the effects of the use of hydrochloric acid as the aqueous phase in fuel recycle are surveyed. Available data are sufficient to suggest that separations of actinides and fission products can be at least as good in an HCl-trialkyl amine system as in Purex. Advantages of the HCl system are simpler operations of the off-gas system, better separation of neptunium from uranium and plutonium, better control of oxidation states of the dissolved species, and simpler recycle of the acid. A possible advantage is the more complete dissolution of the fission products, leaving very little insoluble residue. Disadvantages include lack of development of methods for dissolution of oxide fuel in hydrochloric acid, the requirement for processing equipment constructed of titanium, possible complications in the waste-handling system, and the dissolution of much of the cladding in the case of stainless-steel clad fuel

  8. 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

  9. Economics analysis of fuel cycle cost of fusion–fission hybrid reactors based on different fuel cycle strategies

    Energy Technology Data Exchange (ETDEWEB)

    Zu, Tiejun, E-mail: tiejun@mail.xjtu.edu.cn; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi

    2015-01-15

    Highlights: • Economics analysis of fuel cycle cost of FFHRs is carried out. • The mass flows of different fuel cycle strategies are established based on the equilibrium fuel cycle model. • The levelized fuel cycle costs of different fuel cycle strategies are calculated, and compared with current once-through fuel cycle. - Abstract: The economics analysis of fuel cycle cost of fusion–fission hybrid reactors has been performed to compare four fuel cycle strategies: light water cooled blanket burning natural uranium (Strategy A) or spent nuclear fuel (Strategy B), sodium cooled blanket burning transuranics (Strategy C) or minor actinides (Strategy D). The levelized fuel cycle costs (LFCC) which does not include the capital cost, operation and maintenance cost have been calculated based on the equilibrium mass flows. The current once-through (OT) cycle strategy has also been analyzed to serve as the reference fuel cycle for comparisons. It is found that Strategy A and Strategy B have lower LFCCs than OT cycle; although the LFCC of Strategy C is higher than that of OT cycle when the uranium price is at its nominal value, it would become comparable to that of OT cycle when the uranium price reaches its historical peak value level; Strategy D shows the highest LFCC, because it needs to reprocess huge mass of spent nuclear fuel; LFCC is sensitive to the discharge burnup of the nuclear fuel.

  10. Molten fluoride mixtures as possible fission reactor fuels

    International Nuclear Information System (INIS)

    Grimes, W.R.

    1978-01-01

    Molten mixtures of fluorides with UF 4 as a component have been used as combined fuel and primary heat transfer agent in experimental high-temperature reactors and have been proposed for use in breeders or converters of 233 U from thorium. Such use places stringent and diverse demands upon the fluid fuel. A brief review of chemical behavior of molten fluorides is given to show some of their strengths and weaknesses for such service

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  12. 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.

  13. Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, Kun [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States); Harp, Jason [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-15

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions needs to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions, and therefore evaluate the qualification of U3Si2 as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U3Si2 fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U3Si2. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.

  14. 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

  15. Data sheets of fission product release experiments for light water reactor fuel, (2)

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi; Nagai, Hitoshi; Takeda, Tsuneo; Yamamoto, Katsumune; Nakazaki, Chozaburo.

    1979-07-01

    This is the second data sheets of fission products (FP) release experiments for light water reactor fuel. Results of five FP release experiments from the third to the seventh are presented: results of pre-examinations of UO 2 pellets, photographs of parts of fuel rod assemblies for irradiation and the assemblies, operational conditions of JMTR and OWL-1, variations of radioiodine-131 level in the main loop coolant during experimental periods, and representative results of post-irradiation examinations of respective fuel rods. (author)

  16. Chemical states of fission products in irradiated uranium-plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Kurosaki, Ken; Uno, Masayoshi; Yamanaka, Shinsuke

    1999-01-01

    The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO 2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO 2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

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

    International Nuclear Information System (INIS)

    Verfondern, K.; Mueller, D.

    1991-01-01

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

  18. Impact of fission gas on irradiated PWR fuel behaviour at extended burnup under RIA conditions

    International Nuclear Information System (INIS)

    Lemoine, F.; Schmitz, F.

    1996-01-01

    With the world-wide trend to increase the fuel burnup at discharge of the LWRs, the reliability of high burnup fuel must be proven, including its behaviour under energetic transient conditions, and in particular during RIAs. Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup. The potential for swelling and transient expansion work under rapid heating conditions characterizes the high burnup fuel behaviour by comparison to fresh fuel. This effect is resulting from the steadily increasing amount of gaseous and volatile fission products retained inside the fuel structure. An attempt is presented to quantify the gas behaviour which is motivated by the results from the global tests both in CABRI and in NSRR. A coherent understanding of specific results, either transient release or post transient residual retention has been reached. The early failure of REP Na1 with consideration given to the satisfactory behaviour of the father rod of the test pin at the end of the irradiation (under load follow conditions) is to be explained both by the transient loading from gas driven fuel swelling and from the reduced clad resistance due to hydriding. (R.P.)

  19. 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

  20. 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

  1. 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

  2. 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

  3. Linear accelerator fuel enricher regenerator (LAFER) and fission product transmutor (APEX)

    International Nuclear Information System (INIS)

    Steinberg, M.; Powell, J.R.; Takahashi, H.; Grand, P.; Kouts, H.J.C.

    1979-01-01

    In addition to safety, two other major problems face the nuclear industry today; first is the long-term supply of fissle material and second is the disposal of long-lived fission product waste. The higher energy proton linear accelerator can assist in the solution of each of these problems. High energy protons from the linear accelerator interact with a molten lead target to produce spallation and evaporation neutrons. The neutrons are absorbed in a surrounding blanket of light water power reactor (LWR) fuel elements to produce fissile Pu-239 or U-233 fuel from natural fertile U-238 or Th-232 contained in the elements. The fissile enriched fuel element is used in the LWR power reactor until its reactivity is reduced after which the element is regenerated in the linear accelerator target/blanket assembly and then the element is once again burned (fissioned) in the power LWR. In this manner the natural uranium fuel resource can supply an expanding nuclear power reactor economy without the need for fuel reprocessing, thus satisfying the US policy of non-proliferation. In addition, the quantity of spent fuel elements for long-term disposal is reduced in proportion to the number of fuel regeneration cycles through the accelerator. The limiting factor for in-situ regeneration is the burnup damage to the fuel cladding material. A 300 ma-1.5 GeV (450 MW) proton linear accelerator can produce approximately one ton of fissile (Pu-239) material annually which is enough to supply fuel to three 1000 MW(e) LWR power reactors. With two cycles of enriching and regenerating, the nuclear fuel natural resource can be stretched by a factor of 3.6 compared to present fuel cycle practice without the need for reprocessing. Furthermore, the need for isotopic enrichment facilities is drastically reduced

  4. 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.

  5. Fission gas release from fuels at high burnup

    International Nuclear Information System (INIS)

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

    1982-04-01

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

  6. 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

  7. Retention of gaseous fission products in reprocessing LMFBR fuels

    International Nuclear Information System (INIS)

    Burch, W.D.; Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

    1976-05-01

    The report is devoted to status of the development programme at the Oak Ridge National Laboratory on methods for retaining iodine-131 and 129, Krypton-85, Tritium and Carbon-14 in reprocessing LMFBR fuels. The Iodox process, Fluorocarbon absorption process and Voloxidation process are described for retention of iodine, Krypton-85 and Tritium, respectively. Flowsheets for the different processes are given and results of experimental runs in small engineering-scale equipment are reported

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-03-01

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

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

    International Nuclear Information System (INIS)

    Rao, C.B.; Raj, Baldev

    1980-01-01

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

  10. 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.

  11. 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)

  12. Measuring head for determining the pressure of fission gases released inside bars of nuclear fuel

    International Nuclear Information System (INIS)

    Granata, S.

    1984-01-01

    A measuring head suitable for determining the pressure of fission gases released inside non-instrumented bars of nuclear fuel (which have reached high irradiation levels), and for connection to said bars by a method which allows no escape of said active gases and does not cause appreciable disturbance either to the fuel or to the released fission gases, is disclosed. The head consists of a tubular casing adapted to be welded at one end to the bar, and having a metal bellows at its other end. A pointed metal bar is used to penetrate the bar by a blow to a pin, whereupon pressure variations within the casing are measured by a pressure measuring device having an iron core, the movement of the core, due to such pressure variations, being recorded by a differential transformer. (author)

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

    International Nuclear Information System (INIS)

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

    1978-10-01

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

  14. Calculation characterization of spent fuel hazard related to partitioning and transmutation of minor actinides and fission products

    International Nuclear Information System (INIS)

    Gerasimov, A. S.; Bergelson, B. R.; Tikhomirov, G.V.; Volovik, A.I. . E-mail of corresponding author: geras@itep.ru; Gerasimov, A.S.)

    2005-01-01

    Radiotoxicity is one of important characteristics of radwaste hazard. Radiotoxicity of actinides and fission products from spent fuel of VVER-1000 reactor for processes of burnup, long-term storage, and transmutation is discussed. (author)

  15. A long term radiological risk model for plutonium-fueled and fission reactor space nuclear system

    International Nuclear Information System (INIS)

    Bartram, B.W.; Dougherty, D.K.

    1987-01-01

    This report describes the optimization of the RISK III mathematical model, which provides risk assessment for the use of a plutonium-fueled, fission reactor in space systems. The report discusses possible scenarios leading to radiation releases on the ground; distinctions are made for an intact reactor and a dispersed reactor. Also included are projected dose equivalents for various accident situations. 54 refs., 31 figs., 11 tabs

  16. A high temperature heating device for the study of fission product release from nuclear fuel

    International Nuclear Information System (INIS)

    Svedkauskaite-Le Gore, Jolanta; Kivel, Niko; Guenther-Leopold, Ines

    2010-01-01

    At the Paul Scherrer Institute a high temperature inductive heating furnace, which can heat fuel samples up to 2300 deg. C, has been developed in order to study the release of fission products. The furnace can be directly connected to an inductively coupled plasma mass spectrometer for online monitoring of the released elements and does not require their trapping before measurement. This paper describes the design of the inductive heating furnace, discusses its operating parameters, limitations and illustrates foreseen applications. (authors)

  17. Using New Fission Data with the Multi-detector Analysis System for Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    A. V. Ramayya; A.V. Daniel (Joint Institute for Nuclear Research); C. J. Beyer (Vanderbilt Univ.); E. L. Reber; G. M. Ter-Akopian; G.S. Popeko; J. D. Cole; J. H. Hamilton; J. K. Jewell (INEEL); M. W. Drigert; R. Aryaeinejad; Ts.Yu. Oganessian

    1998-11-01

    New experiments using an array of high purity germanium detectors and fast liquid scintillation detectors has been performed to observe the radiation emitted from the induced fission of 235U with a beam of thermal neutrons. The experiment was performed at the Argonne National Laboratory Intense Pulsed Neutron Source. Preliminary observations of the data are presented. A nondestructive analysis system for the characterization of DOE spent nuclear fuel based on these new data is presented.

  18. Using New Fission Data with the Multi-detector Analysis System for Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cole, Jerald Donald

    1998-11-01

    New experiments using an array of high purity germanium detectors and fast liquid scintillation detectors has been performed to observe the radiation emitted from the induced fission of 235U with a beam of thermal neutrons. The experiment was performed at the Argonne National Laboratory Intense Pulsed Neutron Source. Preliminary observations of the data are presented. A nondestructive analysis system for the characterization of DOE spent nuclear fuel based on these new data is presented.

  19. 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)

  20. Fuel age impacts on gaseous fission product capture during separations

    Energy Technology Data Exchange (ETDEWEB)

    Jubin, Robert T.; Soelberg, Nicolas R.; Strachan, Denis M.; Ilas, G.

    2012-09-21

    As a result of fuel reprocessing, volatile radionuclides will be released from the facility stack if no processes are put in place to remove them. The radionuclides that are of concern in this document are 3H, 14C, 85Kr, and 129 Rosnick 2007 I. The question we attempt to answer is how efficient must this removal process be for each of these radionuclides? To answer this question, we examine the three regulations that may impact the degree to which these radionuclides must be reduced before process gases can be released from the facility. These regulations are 40 CFR 61 (EPA 2010a), 40 CFR 190(EPA 2010b), and 10 CFR 20 (NRC 2012), and they apply to the total radonuclide release and to the dose to a particular organ – the thyroid. Because these doses can be divided amongst all the radionuclides in different ways and even within the four radionuclides in question, several cases are studied. These cases consider for the four analyzed radionuclides inventories produced for three fuel types—pressurized water reactor uranium oxide (PWR UOX), pressurized water reactor mixed oxide (PWR MOX), and advanced high-temperature gascooled reactor (AHTGR)—several burnup values and time out of reactor extending to 200 y. Doses to the maximum exposed individual (MEI) are calculated with the EPA code CAP-88 ( , 1992). Two dose cases are considered. The first case, perhaps unrealistic, assumes that all of the allowable dose is assigned to the volatile radionuclides. In lieu of this, for the second case a value of 10% of the allowable dose is arbitrarily selected to be assigned to the volatile radionuclides. The required decontamination factors (DFs) are calculated for both of these cases, including the case for the thyroid dose for which 14C and 129I are the main contributors. However, for completeness, for one fuel type and burnup, additional cases are provided, allowing 25% and 50% of the allowable dose to be assigned to the volatile radionuclides. Because 3H and 85Kr have

  1. Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives

    Energy Technology Data Exchange (ETDEWEB)

    Foerthmann, R.; Groos, E.; Gruebmeier, H.

    1975-08-15

    Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Temperature Gas-cooled Reactors led to the development of coated particles with improved fission product retention which represent an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates SrAl2Si2O8, BaAl2Si2O8and CsAlSi2O6 in the additional alumina-silica phase of the kernel.

  2. LOFT/LP-FP-2, Loss of Fluid Test, Fission Product Release from Fuel

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The eighth OECD LOFT experiment was conducted on 7 March 1985. It was the second of the two experiments to be performed in the LOFT facility with intentional release of fission products. Its principal objectives were to determine the fission product release from the fuel during a severe fuel damage scenario and the subsequent transport of these fission products in a predominantly vapor/aerosol environment. This was the largest severe fuel damage experiment ever conducted, and serves as an important benchmark between smaller scale tests and the TMI-2 accident. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  3. Grain boundary sweeping and liquefaction-induced fission product behavior in nuclear fuel under severe-core damage accident conditions

    International Nuclear Information System (INIS)

    Rest, J.

    1984-05-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from: (1) irradiated high-burnup LWR fuel in a flowing steam atmosphere during high-temperature, in-cell heating tests performed at Oak Ridge National Laboratory; and (2) trace-irradiated and high-burnup LWR fuel during 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 U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. Results of the analyses demonstrate that intragranular fission product behavior during both types of 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 high-burnup 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 liquefied lamina. The analysis shows that total UO 2 dissolution due to eutectic melting leads to increased release for both trace-irradiated and high-burnup 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 high-burnup fuel are highlighted

  4. Fuel bundle examination techniques for the Phebus fission product test

    International Nuclear Information System (INIS)

    Blanc, J.Y.; Clement, B.; Hardt, P. von der

    1996-01-01

    The paper develops the non-destructive examinations, with a special emphasis on transmission tomography, performed in the Phebus facility, using a linear accelerator associated with a line scan camera based on PCD components. This particular technique enabled the high level of penetration to be obtained, necessary for this high density application. Spatial resolution is not far from the theoretical limit and the density resolution is often adequate. This technique permitted: 1) to define beforehand the cuts on a precise basis, avoiding a long step-by-step choice as in previous in-pile tests; 2) to determine, at an early stage, mass balance, material relocations (in association with axial gamma spectrometry), and FP distribution, as an input into re-calculations of the bundle events. However, classical cuttings, periscopic visual examinations, macrographies, micrographies and EPMA analyses remain essential to give oxidation levels (in the less degraded zones), phase aspect and composition, to distinguish between materials of identical density, and, if possible, to estimate temperatures. Oxidation resistance of sensors (thermocouples or ultrasonic thermometers) is also traced. The EPMA gives access to the molten material chemical analyses, especially in the molten fuel blockage area. The first results show that an important part of the fuel bundle melted (which was one of the objectives of this test) and that the degradation level is close to TIMI-2 with a molten plug under a cavity surrounded by an uranium-rich crust. In lower and upper areas fuel rods are less damaged. Complementaries between these examination techniques and between international teams involved will be major advantages in the Phebus FPT0 test comprehension. 3 refs, 9 figs

  5. Release behavior of fission products from irradiated dispersion fuels at high temperatures

    International Nuclear Information System (INIS)

    Iwai, Takashi; Shimizu, Michio; Nakagawa, Tetsuya

    1990-02-01

    As a framework of reduced enrichment fuel program of JMTR Project, the measurements of fission products release rates at high temperatures (600degC - 1100degC) were performed in order to take the data to use for safety evaluation of LEU fuel. Three type miniplates of dispersion silicide and aluminide fuel, 20% enrichment LEU fuel with 4.8 gU/cc (U 3 Si 2 90 %, USi 10 % and U 3 Si 2 50 %, U 3 Si 50 % dispersed in aluminium) and 45 % enrichment MEU fuel with 1.6 gU/cc, were irradiated in JMTR. The burnups attained by one cycle (22 days) irradiation were within 21.6 % - 22.5 % of initial 235 U. The specimens cut down from miniplates were measured on fission products release rates by means of new apparatus specially designed for this experiment. The specimens were heated up within 600degC - 1100degC in dry air. Then fission products such as 85 Kr, 133 Xe, 131 I, 137 Cs, 103 Ru, 129m Te were collected at each temperature and measured on release rates. In the results of measurement, the release rates of 85 Kr, 133 Xe, 131 I, 129m Te from all specimens were slightly less than that of G.W. Parker's data on U-Al alloy fuel. For 137 Cs and 103 Ru from a silicide specimen (U 3 Si 2 90 %, USi 10 % dispersed in aluminium) and 137 Cs from an aluminide specimen, the release rates were slightly higher than that of G.W. Parker's. (author)

  6. 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)

  7. 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

  8. 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.

  9. Review of tomography technique for 3-dimensional fission product distribution determination in irradiated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pakr, D G; Hong, K P; Joo, Y S; Lee, H K

    2006-04-15

    Tomography algorithm is reviewed in order to determine radial 2-dimensional fission product distribution of irradiated fuel rod and to reconstruct it's image using computer. Main contents are Radon transformation, Fourier central slice theorem, inverse Fourier transform, accompanied FBP(Filtered Back Projection) and BPB(Back Projection Filtering). In addition, another tomography reconstruction algorithm, namely, ART(Algebraic Reconstruction Technique) is reviewed briefly. According to reviewed results, we devise equipment for determining of 2-dimensional distribution of irradiated nuclear fuel using existing gamma scanning apparatus. On results of review, It is necessary to develop computer program of reconstruction algorithm for determining of object function and image reconstruction.

  10. Review of tomography technique for 3-dimensional fission product distribution determination in irradiated fuel

    International Nuclear Information System (INIS)

    Pakr, D. G.; Hong, K. P.; Joo, Y. S.; Lee, H. K.

    2006-04-01

    Tomography algorithm is reviewed in order to determine radial 2-dimensional fission product distribution of irradiated fuel rod and to reconstruct it's image using computer. Main contents are Radon transformation, Fourier central slice theorem, inverse Fourier transform, accompanied FBP(Filtered Back Projection) and BPB(Back Projection Filtering). In addition, another tomography reconstruction algorithm, namely, ART(Algebraic Reconstruction Technique) is reviewed briefly. According to reviewed results, we devise equipment for determining of 2-dimensional distribution of irradiated nuclear fuel using existing gamma scanning apparatus. On results of review, It is necessary to develop computer program of reconstruction algorithm for determining of object function and image reconstruction

  11. Online ICPMS detection of the thermal release of fission products from nuclear fuel samples

    International Nuclear Information System (INIS)

    Guenther-Leopold, I.; Svedkauskaite-Le Gore, J.; Kivel, N.

    2009-01-01

    Full text: The release of volatile and semi-volatile fission products (like Cs, Tc, Mo etc.) from spent nuclear fuel by thermal and thermochemical treatment (oxidative or reductive conditions) as a head-end step for advanced reprocessing scenarios is studied in the Hot Laboratory of the Paul Scherrer Institut. For this purpose, a heated sampling cell online connected to an ICPMS (Element 2, Thermo Fisher Scientific) was designed and tested on simulated fuel samples up to 650 o C. The results of this study as well as technical perspectives for heating experiments up to 2000 o C will be presented. (author)

  12. 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

  13. HTR fuel modelling with the ATLAS code. Thermal mechanical behaviour and fission product release assessment

    International Nuclear Information System (INIS)

    Guillermier, Pierre; Daniel, Lucile; Gauthier, Laurent

    2009-01-01

    To support AREVA NP in its design on HTR reactor and its HTR fuel R and D program, the Commissariat a l'Energie Atomique developed the ATLAS code (Advanced Thermal mechanicaL Analysis Software) with the objectives: - to quantify, with a statistical approach, the failed particle fraction and fission product release of a HTR fuel core under normal and accidental conditions (compact or pebble design). - to simulate irradiation tests or benchmark in order to compare measurements or others code results with ATLAS evaluation. These two objectives aim at qualifying the code in order to predict fuel behaviour and to design fuel according to core performance and safety requirements. A statistical calculation uses numerous deterministic calculations. The finite element method is used for these deterministic calculations, in order to be able to choose among three types of meshes, depending on what must be simulated: - One-dimensional calculation of one single particle, for intact particles or particles with fully debonded layers. - Two-dimensional calculations of one single particle, in the case of particles which are cracked, partially debonded or shaped in various ways. - Three-dimensional calculations of a whole compact slice, in order to simulate the interactions between the particles, the thermal gradient and the transport of fission products up to the coolant. - Some calculations of a whole pebble, using homogenization methods are being studied. The temperatures, displacements, stresses, strains and fission product concentrations are calculated on each mesh of the model. Statistical calculations are done using these results, taking into account ceramic failure mode, but also fabrication tolerances and material property uncertainties, variations of the loads (fluence, temperature, burn-up) and core data parameters. The statistical method used in ATLAS is the importance sampling. The model of migration of long-lived fission products in the coated particle and more

  14. 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.

  15. Standard method of test for atom percent fission in uranium fuel - radiochemical method

    International Nuclear Information System (INIS)

    Anon.

    The determination of the U at. % fission that has occurred in U fuel from an analysis of the 137 Cs ratio to U ratio after irradiation is described. The method is applicable to high-density, clad U fuels (metal, alloys, or ceramic compounds) in which no separation of U and Cs has occurred. The fuels are best aged for several months after irradiation in order to reduce the 13-day 136 Cs activity. The fuel is dissolved and diluted to produce a solution containing a final concentration of U of 100 to 1000 mg U/l. The 137 Cs concentration is determined by ASTM method E 320, for Radiochemical Determination of Cesium-137 in Nuclear Fuel Solutions, and the U concentration is determined by ASTM method E 267, for Determination of Uranium and Plutonium Concentrations and Isotopic Abundances, ASTM method E 318, for Colorimetric Determination of Uranium by Controlled-Potential Coulometry. Calculations are given for correcting the 137 Cs concentration for decay during and after irradiation. The accuracy of this method is limited, not only by the experimental errors with which the fission yield and the half-life of 137 Cs are known

  16. 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.

  17. 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.)

  18. HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1977. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.

  19. 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.

  20. Influence of fission gases on the mechanical state of irradiated oxide fuels

    International Nuclear Information System (INIS)

    Cagna, Celine

    2016-01-01

    The irradiation generates in the fuel, fission gases, mainly xenon and krypton, present in dissolved form and in the form of bubbles. This research objective is to contribute to the fission gas bubbles methodology of characterization and thus to bring elements of reference for the models validation. Two approaches are studied. Based on an existing method of bubbles average pressure evaluation by the coupling of three techniques: EPMA, SEM and SIMS, a new complementary method has been developed on an isolated bubble under the surface. The methodology consists in identifying a closed and filled bubble with xenon by microprobe mapping and SEM images and to measure the amount of present gas by SIMS. 3D observation by FIB abrasion provides an estimation of the bubble volume and thus allows to calculate the bubble pressure. At 300 K, an estimation of the pressure levels is obtained on intragranular micrometric bubbles from the fuel pellets center area. Meanwhile, a method of elastic field strain measurement, produced by the presence of pressurized bubbles, is developed by HR-EBSD. A finite element model evaluates the levels of strain around the fission gas bubbles and shows that only nano-metric bubbles generate measurable elastic strain by this technique. First, the method was calibrated from four points bending tests on monocrystalline silicon and ceramics implanted with xenon, allowing to take into account free strains. This step defines the parameters of acquisition and optimum treatment for its application on irradiated fuels. Measurement of elastic strain with HR-EBSD on irradiated fuel is a relative measure that will require further consideration in the choice of the reference. (author) [fr

  1. 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

  2. Fission product Pd-SiC interaction in irradiated coated particle fuels

    International Nuclear Information System (INIS)

    Tiegs, T.N.

    1980-04-01

    Silicon carbide is the main barrier to fission product release from coated particle fuels. Consequently, degradation of the SiC must be minimized. Electron microprobe analysis has identified that palladium causes corrosion of the SiC in irradiated coated particles. Further ceramographic and electron microprobe examinations on irradiated particles with kernels ranging in composition from UO 2 to UC 2 , including PuO/sub 2 -x/ and mixed (Th, Pu) oxides, and in enrichment from 0.7 to 93.0% 235 U revealed that temperature is the major factor affecting the penetration rate of SiC by Pd. The effects of kernel composition, Pd concentration, other fission products, and SiC properties are secondary

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  5. Modelling of U-Mo/Al Dispersion fuel fission induced swelling and creep

    International Nuclear Information System (INIS)

    Jeong, Gwan Yoon; Sohn, Dong Seong; Kim, Yeon Soo

    2014-01-01

    In a Dispersion fuel which U-Mo particles are dispersed in Al metal matrix, a similar phenomenon forming a bulge region was observed but it is difficult to quantify and construct a model for explaining creep and swelling because of its complex microstructure change during irradiation including interaction layer (IL) and porosity formation. In a Dispersion fuel meat, fission product induces fuel particles swelling and it has to be accommodated by the deformation of the Al matrix and newly formed IL during irradiation. Then, it is reasonable that stress from fuel swelling in the complex structure should be relaxed by local adjustments of particles, Al matrix, and IL. For analysis of U-Mo/Al Dispersion fuel creep, the creep of U-Mo particle, Al matrix, and IL should be considered. Moreover, not only fuel particle swelling and IL growth, but also fuel and Al matrix consumptions due to IL formation are accounted in terms of their volume fraction changes during irradiation. In this work, fuel particles, Al matrix and IL are treated in a way of homogenized constituents: Fuel particles, Al matrix and IL consist of an equivalent meat during irradiation. Meat volume swelling of two representative plates was measured: One (Plate A) was a pure Al matrix with 6g/cc uranium loading, the other (Plate B) a silicon added Al matrix with 8g/cc uranium loading. The meat swelling of calculated as a function of burnup. The meat swelling of calculation and measurement was compared and the creep rate coefficients for Al and IL were estimated by repetitions. Based on assumption that only the continuous phase of Al-IL combined matrix accommodated the stress from fuel particle swelling and it was allowed to have creep deformation, the homogenization modeling was performed. The meat swelling of two U-Mo/Al Dispersion fuel plates was modeled by using homogenization model

  6. Modelling of U-Mo/Al Dispersion fuel fission induced swelling and creep

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Argonne (United States)

    2014-05-15

    In a Dispersion fuel which U-Mo particles are dispersed in Al metal matrix, a similar phenomenon forming a bulge region was observed but it is difficult to quantify and construct a model for explaining creep and swelling because of its complex microstructure change during irradiation including interaction layer (IL) and porosity formation. In a Dispersion fuel meat, fission product induces fuel particles swelling and it has to be accommodated by the deformation of the Al matrix and newly formed IL during irradiation. Then, it is reasonable that stress from fuel swelling in the complex structure should be relaxed by local adjustments of particles, Al matrix, and IL. For analysis of U-Mo/Al Dispersion fuel creep, the creep of U-Mo particle, Al matrix, and IL should be considered. Moreover, not only fuel particle swelling and IL growth, but also fuel and Al matrix consumptions due to IL formation are accounted in terms of their volume fraction changes during irradiation. In this work, fuel particles, Al matrix and IL are treated in a way of homogenized constituents: Fuel particles, Al matrix and IL consist of an equivalent meat during irradiation. Meat volume swelling of two representative plates was measured: One (Plate A) was a pure Al matrix with 6g/cc uranium loading, the other (Plate B) a silicon added Al matrix with 8g/cc uranium loading. The meat swelling of calculated as a function of burnup. The meat swelling of calculation and measurement was compared and the creep rate coefficients for Al and IL were estimated by repetitions. Based on assumption that only the continuous phase of Al-IL combined matrix accommodated the stress from fuel particle swelling and it was allowed to have creep deformation, the homogenization modeling was performed. The meat swelling of two U-Mo/Al Dispersion fuel plates was modeled by using homogenization model.

  7. MICRO/NANO-STRUCTURAL EXAMINATION AND FISSION PRODUCT IDENTIFICATION IN NEUTRON IRRADIATED AGR-1 TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    van Rooyen, I. J.; Lillo, T. M.; Wen, H. M.; Hill, C. M.; Holesinger, T. G.; Wu, Y. Q.; Aguiara, J. A.

    2016-11-01

    Advanced microscopic and microanalysis techniques were developed and applied to study irradiation effects and fission product behavior in selected low-enriched uranium oxide/uranium carbide TRISO-coated particles from fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA. Although no TRISO coating failures were detected during the irradiation, the fraction of Ag-110m retained in individual particles often varied considerably within a single compact and at the capsule level. At the capsule level Ag-110m release fractions ranged from 1.2 to 38% and within a single compact, silver release from individual particles often spanned a range that extended from 100% retention to nearly 100% release. In this paper, selected irradiated particles from Baseline, Variant 1 and Variant 3 type fueled TRISO coated particles were examined using Scanning Electron Microscopy, Atom Probe Tomography; Electron Energy Loss Spectroscopy; Precession Electron Diffraction, Transmission Electron Microscopy, Scanning Transmission Electron Microscopy (STEM), High Resolution Electron Microscopy (HRTEM) examinations and Electron Probe Micro-Analyzer. Particle selection in this study allowed for comparison of the fission product distribution with Ag retention, fuel type and irradiation level. Nano sized Ag-containing features were predominantly identified in SiC grain boundaries and/or triple points in contrast with only two sitings of Ag inside a SiC grain in two different compacts (Baseline and Variant 3 fueled compacts). STEM and HRTEM analysis showed evidence of Ag and Pd co-existence in some cases and it was found that fission product precipitates can consist of multiple or single phases. STEM analysis also showed differences in precipitate compositions between Baseline and Variant 3 fuels. A higher density of fission product precipitate clusters were identified in the SiC layer in particles from the Variant 3 compact compared with the Variant 1 compact. Trend analysis shows

  8. Mechanistic modelling of gaseous fission product behaviour in UO2 fuel by Rtop code

    International Nuclear Information System (INIS)

    Kanukova, V.D.; Khoruzhii, O.V.; Kourtchatov, S.Y.; Likhanskii, V.V.; Matveew, L.V.

    2002-01-01

    The current status of a mechanistic modelling by the RTOP code of the fission product behaviour in polycrystalline UO 2 fuel is described. An outline of the code and implemented physical models is presented. The general approach to code validation is discussed. It is exemplified by the results of validation of the models of fuel oxidation and grain growth. The different models of intragranular and intergranular gas bubble behaviour have been tested and the sensitivity of the code in the framework of these models has been analysed. An analysis of available models of the resolution of grain face bubbles is also presented. The possibilities of the RTOP code are presented through the example of modelling behaviour of WWER fuel over the course of a comparative WWER-PWR experiment performed at Halden and by comparison with Yanagisawa experiments. (author)

  9. Method and device for fabricating dispersion fuel comprising fission product collection spaces

    Science.gov (United States)

    Shaber, Eric L; Fielding, Randall S

    2015-05-05

    A method of fabricating a nuclear fuel comprising a fissile material, one or more hollow microballoons, a phenolic resin, and metal matrix. The fissile material, phenolic resin and the one or more hollow microballoons are combined. The combined fissile material, phenolic resin and the hollow microballoons are heated sufficiently to form at least some fissile material carbides creating a nuclear fuel particle. The resulting nuclear fuel particle comprises one or more fission product collection spaces. In a preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by forming the fissile material into microspheres. The fissile material microspheres are then overcoated with the phenolic resin and microballoon. In another preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by overcoating the microballoon with the fissile material, and phenolic resin.

  10. 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

  11. Fission product release from HTGR fuel under core heatup accident conditions - HTR2008-58160

    International Nuclear Information System (INIS)

    Verfondern, K.; Nabielek, H.

    2008-01-01

    Various countries engaged in the development and fabrication of modern fuel for the High Temperature Gas-Cooled Reactor (HTGR) have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under operating and accidental conditions of future HTGRs. Within the IAEA directed Coordinated Research Project CRP6 on 'Advances in HTGR Fuel Technology Development' active since 2002, the 13 participating Member States have agreed upon benchmark studies on fuel performance during normal operation and under accident conditions. While the former has been completed in the meantime, the focus is now on the extension of the national code developments to become applicable to core heatup accident conditions. These activities are supported by the fact that core heatup simulation experiments have been resumed recently providing new, highly valuable data. Work on accident performance will be - similar to the normal operation benchmark - consisting of three essential parts comprising both code verification that establishes the correspondence of code work with the underlying physical, chemical and mathematical laws, and code validation that establishes reasonable agreement with the existing experimental data base, but including also predictive calculations for future heating tests and/or reactor concepts. The paper will describe the cases to be studied and the calculational results obtained with the German computer model FRESCO. Among the benchmark cases in consideration are tests which were most recently conducted in the new heating facility KUEFA. Therefore this study will also re-open the discussion and analysis of both the validity of diffusion models and the transport data of the principal fission product species in the HTGR fuel materials as essential input data for the codes. (authors)

  12. Behaviour of short-lived fission products within operating UO2 fuel elements

    International Nuclear Information System (INIS)

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

    1983-01-01

    We have carried out experiments using a ''sweep gas'' technique to determine the behaviour of short-lived fission products within operating, intact UO 2 fuel elements. The Zircaloy-4-clad elements were 500 mm long and contained fuel of density 10.65-10.71 Mg/m 3 . A He-2% H 2 carrier gas swept gaseous or volatile fission products out of the operating fuel element past a gamma spectrometer for measurement. In tests at linear powers of 45 and 60 kW/m to maximum burnups of 70 MW.h/kg U, the species measured directly at the spectrometer were generally the short-lived xenons and kryptons. We did not observe iodine or bromine during normal operation. However, we have deduced the behaviour of I-133 and I-135 from the decay of Xe-133 and Xe-135 during reactor shutdowns. Plots of R/B (released/born) against lambda (decay constant) or effective lambda for all isotopes observed at 45 and 60 kW/m show that a line of slope -0.5, corresponding with diffusion kinetics, is a good fit to the measured xenon and krypton data. Our inferred release of iodine fits the same line. From this we can extrapolate to an R/B for I-131 of about 5x10 -3 . The ANS 5.4 release correlation gives calculated results in good agreement with our measurements. (author)

  13. Determination of burnup in irradiated nuclear fuels by the method of stable 148Nd fission products

    International Nuclear Information System (INIS)

    Souza Sarkis, J.E. de.

    1982-01-01

    A method is described for the isotopic analysis and determination of burnup in irradiated nuclear fuel by mass spectrometric technique. The burnup is calculed from the determination of the concentration of uranium, plutonium and the fission product 148 Nd in the samples of UO 2 irradiated fuel from a P.W.R. type reactor. The method involves the separation of fractions of uranium, plutonium and neodimium from the dissolved irradiated fuel by ion exchange technique. The determination of uranium, plutonium and the fission product 148 Nd is carried out by isotope diluition mass spectrometry technique using as isotope tracers the nuclides 233 U, 242 Pu and 150 Nd. For the chemical processing and handling of the irradiated sample a Hot Chemistry Laboratory was mounted. It consists of glove boxes and equipments for radiation monitoring and protection. The results obtained indicate an atom percent burnup of 2.181 + - 0.035% wich corresponds an 20.937 + - 0.739 Gwd/ton of thermal energy produced. (Author) [pt

  14. Behaviour of fission products in PWR primary coolant and defected fuel rods evaluation

    International Nuclear Information System (INIS)

    Bourgeois, P.; Stora, J.P.

    1979-01-01

    The activity surveillance of the PWR primary coolant by γ spectometry gives some informations on fuel failures. The activity of different nuclides e.g. Xenons, Kryptons, Iodines, can be correlated with the number of the defected fuel rods. Therefore the precharacterization with eventually a prelocalization of the related fuel assemblies direct the sipping-test and allows a saving of time during refueling. A model is proposed to calculate the number of the defected rods from the activity measurements of the primary coolant. A semi-empirical model of the release of the fission products has been built from the activity measurements of the primary coolant in a 900 MWe PWR. This model allows to calculate the number of the defected rods and also a typical parameter of the mean damage. Fission product release is described by three stages: release from uranium dioxide, transport across the gas gap and behaviour in the primary coolant. The model of release from the oxide considers a diffusion process in the grains with trapping. The release then occurs either directly to free surfaces or with a delay due to a transit into closed porosity of the oxide. The amount released is the same for iodine and rare gas. With the gas gap transit is associated a transport time and a probability of trapping for the iodines. In the primary coolant the purification and the radioactive decay are considered. (orig.)

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    International Nuclear Information System (INIS)

    1997-10-01

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

  17. Neutronic performance of a fusion-fission hybrid reactor designed for fuel enrichment for LWRs

    International Nuclear Information System (INIS)

    Yapici, H.; Baltacioglu, E.

    1997-01-01

    In this study, the breeding performance of a fission hybrid reactor was analyzed to provide fissile fuel for Light Water Reactors (LWR) as an alternative to the current methods of gas diffusion and gas centrifuge. LWR fuel rods containing UO 2 or ThO 2 fertile material were located in the fuel zone of the blanket and helium gas or Flibe (Li 2 BeF 4 ) fluid was used as coolant. As a result of the analysis, according to fusion driver (D,T and D,D) and the type of coolant the enrichment of 3%-4% were achieved for operation periods of 12 and 36 months in case of fuel rods containing UO 2 , respectively and for operation periods of 18 and 48 months in case of fuel rods containing ThO 2 , respectively. Depending on the type of fusion driver, coolant and fertile fuel, varying enrichments of between 3% and 8.9% were achieved during operation period of four years

  18. Preliminary assessment of a symbiotic fusion--fission power system using the TH/U refresh fuel cycle

    International Nuclear Information System (INIS)

    Bender, D.J.; Lee, J.D.; Moir, R.W.

    1977-10-01

    Studies of the mirror hybrid reactor by LLL/GA have concluded that the most promising role for this reactor concept is that of a producer of fissile fuel for fission reactors. Studies to date have examined primarily the U/Pu fuel cycle with light-water reactors serving as the consumers of the hybrid-bred fissile fuel; the specific scenarios examined required reprocessing and refabrication of the bred fuel before introduction into the fission reactor. This combination of technologies was chosen to illustrate the manner in which the hybrid reactor concept could serve the needs of, and use the technology of, the fission reactor industry as it now exists (and as it was thought it would evolve). However, the current U.S. Administration has expressed strong concerns about proliferation of nuclear weapons capability and terrorist diversion of weapons-grade nuclear materials. These concerns are based on the projected technology for the light-water reactor/fast breeder reactor using the U/Pu fuel cycle and extensive reprocessing/refabrication. A symbiotic nuclear power generation concept (hybrid fissile producer plus fission burner reactors) is described which eliminates those aspects of the present nuclear fuel cycle that (may) represent significant proliferation/diversion risks. Specifically, the proposed concept incorporates the following features: (1)Th/U 233 fuel cycle, (2) no reprocessing or fabrication of fissile material, and (3) no fissile material in a weapons-grade state

  19. Effect of fission rate on the microstructure of coated UMo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Leenaers, A.; Parthoens, Y.; Cornelis, G.; Kuzminov, V.; Koonen, E.; Van den Berghe, S.; Ye, B.; Hofman, G. L.; Schulthess, Jason

    2017-10-01

    Compared to previous irradiation experiments containing UMo/Al dispersion fuel plates, the SELENIUM irradiation experiment performed at the SCK.CEN BR2 reactor in 2012 showed an improved plate swelling behavior. However, in the high burn-up area of the plates a significant increase in meat thickness was still measured. The origin of this increase is currently not firmly established, but it is clear from the observed microstructure that the swelling rate still is too high for practical purposes and needs to be reduced. It was stipulated that the swelling occurred at the high burnup areas which are also the high power zones at beginning of life. For that reason, an experiment was proposed to investigate the influence of fission rate (i.e. power) on some of the observed phenomena. For this purpose, a sibling plate to a high power (BOL>470 W/cm(2)) SELENIUM plate was irradiated during four BR2 cycles. The SELENIUM 1a fuel plate was submitted to a local maximum heat flux below 350 W/cm(2), throughout the full irradiation. At the end of the last cycle, the SELENIUM 1a fuel plate reached a maximum local burnup value of close to 75%U-235 compared to 70%U-235 for the SELENIUM high power plates. When comparing to the results on the SELENIUM plates, the non-destructive tests clearly show a continued linear swelling behavior of the low power irradiated fuel plate SELENIUM 1a in the high burn-up region. The influence of the fission rate is also evidenced in the microstructural examination of the fuel showing that there is no formation of interaction layer at the high burn-up region.

  20. Effect of fission rate on the microstructure of coated UMo dispersion fuel

    Science.gov (United States)

    Leenaers, A.; Parthoens, Y.; Cornelis, G.; Kuzminov, V.; Koonen, E.; Van den Berghe, S.; Ye, B.; Hofman, G. L.; Schulthess, Jason

    2017-10-01

    Compared to previous irradiation experiments containing UMo/Al dispersion fuel plates, the SELENIUM irradiation experiment performed at the SCK·CEN BR2 reactor in 2012 showed an improved plate swelling behavior. However, in the high burn-up area of the plates a significant increase in meat thickness was still measured. The origin of this increase is currently not firmly established, but it is clear from the observed microstructure that the swelling rate still is too high for practical purposes and needs to be reduced. It was stipulated that the swelling occurred at the high burnup areas which are also the high power zones at beginning of life. For that reason, an experiment was proposed to investigate the influence of fission rate (i.e. power) on some of the observed phenomena. For this purpose, a sibling plate to a high power (BOL>470 W/cm2) SELENIUM plate was irradiated during four BR2 cycles. The SELENIUM 1a fuel plate was submitted to a local maximum heat flux below 350 W/cm2, throughout the full irradiation. At the end of the last cycle, the SELENIUM 1a fuel plate reached a maximum local burnup value of close to 75%235U compared to 70%235U for the SELENIUM high power plates. When comparing to the results on the SELENIUM plates, the non-destructive tests clearly show a continued linear swelling behavior of the low power irradiated fuel plate SELENIUM 1a in the high burn-up region. The influence of the fission rate is also evidenced in the microstructural examination of the fuel showing that there is no formation of interaction layer at the high burn-up region.

  1. 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)

  2. Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

    International Nuclear Information System (INIS)

    Harp, Jason M.; Demkowicz, Paul A.

    2014-01-01

    In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10"-"4 to 10"-"5) of as manufactured defects and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from intentionally failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application was considered. Previous experience utilizing similar techniques, the expected activities in AGR-3/4 rings, and analysis of this work indicate using GECT to evaluate AGR-3/4 will be feasible. The GECT technique was also applied to other irradiated nuclear fuel systems currently available in the HFEF hot cell, including oxide fuel pins, metallic fuel pins, and monolithic plate fuel. Results indicate GECT with the HFEF PGS is effective. (author)

  3. The coupled kinetics of grain growth and fission product behavior in nuclear fuel under degraded-core accident conditions

    International Nuclear Information System (INIS)

    Rest, J.

    1985-01-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, and cesium release from (1) irradiated high-burnup LWR fuel in a flowing steam atmosphere during high-temperature, in-cell heating tests (performed at Oak Ridge National Laboratory) and (2) trace-irratiated LWR fuel during severe-fuel-damage (SFD) tests (performed in the PBF reactor in Idaho). A theory of grain boundary sweeping of gas bubbles has been included within the FASTGRASS-VFP formalism. This theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges, and provides a means of determining whether gas bubbles are caught up and moved along by a moving grain boundary or whether the grain boundary is only temporarily retarded by the bubbles and then breaks away. In addition, as FASTGRASS-VFP provides for a mechanistic calculation of intra- and intergranular fission product behavior, the coupled calculation between fission gas behavior and grain growth is kinetically comprehensive. Results of the analyses demonstrate that intragranular fission product behavior during both types of 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. The effect of fuel oxidation by steam on fission product and grain growth behavior is also considered. 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 high-burnup fuel are highlighted. (orig.)

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

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1980-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Vinjamuri, K.

    1980-01-01

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

  6. 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

  7. 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....

  8. 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)

  9. Irradiation performance of coated fuel particles with fission product retaining kernel additives

    International Nuclear Information System (INIS)

    Foerthmann, R.

    1979-10-01

    The four irradiation experiments FRJ2-P17, FRJ2-P18, FRJ2-P19, and FRJ2-P20 for testing the efficiency of fission product-retaining kernel additives in coated fuel particles are described. The evaluation of the obtained experimental data led to the following results: - zirconia and alumina kernel additives are not suitable for an effective fission product retention in oxide fuel kernels, - alumina-silica kernel additives reduce the in-pile release of Sr 90 and Ba 140 from BISO-coated particles at temperatures of about 1200 0 C by two orders of magnitude, and the Cs release from kernels by one order of magnitude, - effective transport coefficients including all parameters which contribute to kernel release are given for (Th,U)O 2 mixed oxide kernels and low enriched UO 2 kernels containing 5 wt.% alumina-silica additives: 10g sub(K)/cm 2 s -1 = - 36 028/T + 6,261 (Sr 90), 10g Dsub(K)/cm 2 c -2 = - 29 646/T + 5,826 (Cs 134/137), alumina-silica kernel additives are ineffective for retaining Ag 110 m in coated particles. However, also an intact SiC-interlayer was found not to be effective at temperatures above 1200 0 C, - the penetration of the buffer layer by fission product containing eutectic additive melt during irradiation can be avoided by using additives which consist of alumina and mullite without an excess of silica, - annealing of LASER-failed irradiated particles and the irradiation test FRJ12-P20 indicate that the efficiency of alumina-silica kernel additives is not altered if the coating becomes defect. (orig.) [de

  10. On the combination of delayed neutron and delayed gamma techniques for fission rate measurement in nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Perret, G.; Jordan, K. A. [Paul Scherrer Institut, Villigen, 5232 (Switzerland)

    2011-07-01

    Novel techniques to measure newly induced fissions in spent fuel after re-irradiation at low power have been developed and tested at the Proteus zero-power research reactor. The two techniques are based on the detection of high energy gamma-rays emitted by short-lived fission products and delayed neutrons. The two techniques relate the measured signals to the total fission rate, the isotopic composition of the fuel, and nuclear data. They can be combined to derive better estimates on each of these parameters. This has potential for improvement in many areas. Spent fuel characterisation and safeguard applications can benefit from these techniques for non-destructive assay of plutonium content. Another application of choice is the reduction of uncertainties on nuclear data. As a first application of the combination of the delayed neutron and gamma measurement techniques, this paper shows how to reduce the uncertainties on the relative abundances of the longest delayed neutron group for thermal fissions in {sup 235}U, {sup 239}Pu and fast fissions in {sup 238}U. The proposed experiments are easily achievable in zero-power research reactors using fresh UO{sub 2} and MOX fuel and do not require fast extraction systems. The relative uncertainties (1{sigma}) on the relative abundances are expected to be reduced from 13% to 4%, 16% to 5%, and 38% to 12% for {sup 235}U, {sup 238}U and {sup 239}Pu, respectively. (authors)

  11. Rupture of Al matrix in U-Mo/Al dispersion fuel by fission induced creep

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Sohn, Dong Seong [UNIST, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Argonnge (United States); Lee, Kyu Hong [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This phenomenon was found specifically in the dispersion fuel plate with Si addition in the Al matrix to suppress interaction layer (IL) formation between UMo and Al. It is known that the stresses induced by fission induced swelling in U-Mo fuel particles are relieved by creep deformation of the IL, surrounding the fuel particles, that has a much higher creep rate than the Al matrix. Thus, when IL growth is suppressed, the stress is instead exerted on the Al matrix. The observed rupture in the Al matrix is believed to be caused when the stress exceeded the rupture strength of the Al matrix. In this study, the possibility of creep rupture of the Al matrix between the neighboring U-Mo fuel particles was examined using the ABAQUS finite element analysis (FEA) tool. The predicted rupture time for a plate was much shorter than its irradiation life indicating a rupture during the irradiation. The higher stress leads Al matrix to early creep rupture in this plate for which the Al matrix with lower creep strain rate does not effectively relieve the stress caused by the swelling of the U-Mo fuel particles. For the other plate, no rupture was predicted for the given irradiation condition. The effect of creeping of the continuous phase on the state of stress is significant.

  12. Fission product behavior during the first two PBF severe fuel damage tests

    International Nuclear Information System (INIS)

    Osetek, D.J.; Cronenberg, A.W.; Hobbins, R.R.; Vinjamuri, K.

    1984-01-01

    The results of the first two severe fuel damage tests performed in the Power Burst Facility are assessed in terms of fission product release and chemical behavior. On-line gamma spectroscopy and grab sample data indicate limited release during solid-phase fuel heatup. Analysis indicates that the fuel morphology conditions for the trace-irradiated fuel employed in these two tests limit initial release. Only upon high temperature fuel restructuring and liquefaction is significant release indicated. Chemical equilibrium predictions, based on steam oxidation or reduction conditions, indicate I to be the primary iodine species during trnsport in the steam environment of the first test and CsI to be the primary species during transport in the hydrogen environment of the second test. However, the higher steam flow rate conditions of the first test transported the released iodine through the sample system; whereas, low-hydrogen flow rate of the second test apparently allowed the vast majority of iodine-bearing compounds to plateout during transport

  13. Influence of remaining fission products in low-decontaminated fuel on reactor core characteristics

    International Nuclear Information System (INIS)

    Ohki, Shigeo

    2002-07-01

    Design study of core, fuel and related fuel cycle system with low-decontaminated fuel has been performed in the framework of the feasibility study (F/S) on commercialized fast reactor cycle systems. This report summarizes the influence on core characteristics of remaining fission products (FPs) in low-decontaminated fuel related to the reprocessing systems nominated in F/S phase I. For simple treatment of the remaining FPs in core neutronics calculation the representative nuclide method parameterized by the FP equivalent coefficient and the FP volume fraction was developed, which enabled an efficient evaluation procedure. As a result of the investigation on the sodium cooled fast reactor with MOX fuel designed in fiscal year 1999, it was found that the pyrochemical reprocessing with molten salt (the RIAR method) brought the largest influence. Nevertheless, it was still within the allowable range. Assuming an infinite-times recycling, the alternations in core characteristics were evaluated as follows: increment of burnup reactivity by 0.5%Δk/kk', decrement of breeding ratio by 0.04, increment of sodium void reactivity by 0.1x10 -2 Δk/kk' and decrement of Doppler constant (in absolute value) by 0.7x10 -3 Tdk/dT. (author)

  14. 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

  15. Determination of fission products in irradiated fuel by X-ray fluorescence

    International Nuclear Information System (INIS)

    Mogensen, M.; Als-Nielsen, J.; Hessel Andersen, N.

    1986-08-01

    X-ray fluorescence is a well established analytical tool for measuring elemental composition of fairly large (approximately 5 cm 2 ) ''cold'' samples. A version of this technique has been developed for analysis of radial distribution of fission products Xe, Cs and Ba in irradiated UO 2 fuel samples. About 0.1 mm thin slices of fuel pellets (full cross sections) are irradiated by 50 keV X-rays. The intensity of the Xe (Cs, Ba) K α fluorescence radiation generated is measured by means of a Ge detector fitted with a collimator. The slit is 0.5 mm wide in the scanning direction and 2 mm long. The measured Xe K α X-ray intensities are converted to absolute concentrations by comparing to the intensity from a Xe gas standard. In the case of Cs and Ba solid standards may be used. The X-ray fluorescence analysis is compared to other techniques used to obtain radial fission product profiles. It is shown how a combination of X-ray fluorescence and electron probe micro analysis is able to reveal the amount of Xe in the grain boundary porosities. (author)

  16. Effects of hyperstoichiometry and fission products on the electrochemical reactivity of UO2 nuclear fuel

    International Nuclear Information System (INIS)

    Betteridge, J.S.; Scott, N.A.M.; Shoesmith, D.W.; Bahen, L.E.; Hocking, W.H.; Lucuta, P.G.

    1997-03-01

    The effects of hyperstoichiometry and fission products on the electrochemical reactivity Of UO 2 nuclear fuel have been systematically investigated using cyclic voltammetry and the O 2 reduction reaction. Significant constraints are placed on the active-site model for O 2 reduction by the modest impact of bulk hyperstoichiometry. Formation of the U 4 O 9 derivative phase was associated with a marked increase in transient surface oxidation/reduction processes, which probably involve localized attack and might be fostered by tensile stresses induced during oxidation. Electrocatalytic reduction Of O 2 on simulated nuclear fuel (SIMFUEL) has been determined to increase progressively with nominal burnup and pronounced enhancement of H 2 O reduction has been observed as well. Substitution of uranium by lower-valence (simulated) fission products, which was formerly considered the probable cause for this behaviour, has now been shown to merely provide good electrical conductivity. Instead, the enhanced reduction kinetics for O 2 and H 2 O on SIMFUEL can be fully accounted for by noble metals, which segregate to the UO 2 grain boundaries as micron-sized particles, despite their low effective surface area. Apparent convergence of the electrochemical properties Of UO 2 and SIMFUEL through natural corrosion likely reflects evolution toward a common active surface. (author)

  17. 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

  18. 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

  19. Scoping study of flowpath of simulated fission products during secondary burning of crushed HTGR fuel in a quartz fluidized-bed burner

    International Nuclear Information System (INIS)

    Rindfleisch, J.A.; Barnes, V.H.

    1976-04-01

    The results of four experimental runs in which isotopic tracers were used to simulate fission products during fluidized bed secondary burning of HTGR fuel were studied. The experimental tests provided insight relative to the flow path of fission products during fluidized-bed burning of HTGR fuel

  20. Fission product release from nuclear fuel II. Validation of ASTEC/ELSA on analytical and large scale experiments

    International Nuclear Information System (INIS)

    Brillant, G.; Marchetto, C.; Plumecocq, W.

    2013-01-01

    Highlights: • A wide range of experiments is presented for the ASTEC/ELSA code validation. • Analytical tests such as AECL, ORNL and VERCORS are considered. • A large-scale experiment, PHEBUS FPT1, is considered. • The good agreement with measurements shows the efficiency of the ASTEC modelling. • Improvements concern the FP release modelling from MOX and high burn-up UO 2 fuels. - Abstract: This article is the second of two articles dedicated to the mechanisms of fission product release from a degraded core. The models of fission product release from nuclear fuel in the ASTEC code have been described in detail in the first part of this work (Brillant et al., this issue). In this contribution, the validation of ELSA, the module of ASTEC that deals with fission product and structural material release from a degraded core, is presented. A large range of experimental tests, with various temperature and conditions for the fuel surrounding atmosphere (oxidising and reducing), is thus simulated with the ASTEC code. The validation database includes several analytical experiments with both bare fuel (e.g. MCE1 experiments) and cladded fuel (e.g. HCE3, VERCORS). Furthermore, the PHEBUS large-scale experiments are used for the validation of ASTEC. The rather satisfactory comparison between ELSA calculations and experimental measurements demonstrates the efficiency of the analytical models to describe fission product release in severe accident conditions

  1. Lanthanide fission product separation from the transuranics in the integral fast reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

    Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Ackerman, J.P.

    1993-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed by Argonne National Laboratory. This reactor uses liquid-metal cooling and metallic fuel. Its spent fuel will be reprocessed using a pyrochemical method employing molten salts and liquid metals in an electrofining operation. The lanthanide fission products are a concern during reprocessing because of heating and fuel performance issues, so they must be removed periodically from the system to lessen their impact. The actinides must first be removed form the system before the lanthanides are removed as a waste stream. This operation requires a relatively good lanthanide-actinide separation to minimize both the amount of transuranic material lost in the waste stream and the amount of lanthanides collected when the actinides are first removed. A computer code, PYRO, that models these operations using thermodynamic and empirical data was developed at Argonne and has been used to model the removal of the lanthanides from the electrorefiner after a normal operating campaign. Data from this model are presented. The results demonstrate that greater that 75% of the lanthanides can be separated from the actinides at the end of the first fuel reprocessing campaign using only the electrorefiner vessel

  2. Nondestructive analysis of the RA fuel burnup, Calculation of the gamma activity ratio of fission products in the fuel - program QU0C1

    International Nuclear Information System (INIS)

    Bulovic, V.F.

    1973-01-01

    The γ radiation of RA reactor fuel element was measured under precisely defined measuring conditions. The spectrum was analysed by spectrometer with semiconductor Ge(Li) detector. The gamma counting rate in the fuel spectrum is defined as a function of fission product activity, gamma energy and yield, fuel thickness and additional absorbers, dimensions of the gamma collimator. Activity ratio of two fission products is defined as a function of counting rate peaks and part of the mentioned quantities. Four options for calculating the activities for fission products are discussed. Three of them are covered by the QU0C1 code written in FORTRAN for the CDC 3600 computer. The code is included in this report [sr

  3. The gamma spectrometry a powerful tool for irradiated fuel and fission products release studies

    International Nuclear Information System (INIS)

    Pontillon, Y.; Roure, C.; Lacroix, B.; Martella, T.; Ducros, G.; Ravel, S.; Gleizes, B.

    2003-01-01

    Over the last decades, due to the potentially severe consequences of a nuclear incident and/or accident for surrounding populations as well as the environment, international safety authorities launched R and D programs in support of general policy on exploitation of nuclear energy. This increasing interest enabled starting of many research programs in CEA and particularly in Nuclear Energy Directorate (DEN). Most of them are devoted to (i) the source term of fission products (including gas) and actinides released from PWR fuel samples in normal or accident conditions, (ii) burn-up determination, (iii) isotopic repartition... by quantitative gamma spectrometry. In this context, the Department of Fuel Studies (DEC), part of the DEN, has acquired considerable experience in this field of research. In order to attain the required capabilities, specific technical facilities set up in shielded hot cells at the CEA-Grenoble and CEA-Cadarache have been developed. In particular, the researchers of the Department have developed several gamma scanning benches and a set of two thermal treatment devices, including the so-called 'VERCORS facility'. These devices are associated to on line quantitative gamma spectrometry, in order to measure emitted gas and fission products (FPs). The greatest asset of such installations is to ensure a high analytical experiments rate, and as a consequence to make parametrical approach of planned studies easier. The first part of the present communication focuses, on the one hand, on the peculiar aspects of the gamma spectrometry applied on irradiated fuel, mad on the other hand, on the technical aspect of the different facilities (i.e. quantitative gamma spectrometry apparatus and corresponding 'home made' software). The last part is devoted to the results which can be obtained with such installation. In particular, it will be explained how experimental programs on FPs and gas release in normal and/or accidental conditions can be conducted

  4. Dosage of fission products in irradiated fuel treatment effluents (radio-chemical method); Dosage des produits de fission dans les effluents du traitement des combustibles irradies (methode radiochimique)

    Energy Technology Data Exchange (ETDEWEB)

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

    1966-07-01

    The dosage methods presented here are applicable to relatively long-lived fission products present in the effluents resulting from irradiated fuel treatment processes (Sr - Cs - Ce - Zr - Nb - Ru - I). The methods are based on the same principle: - addition of a carrying-over agent - chemical separation over several purification stages, - determination of the chemical yield by calorimetry - counting of an aliquot liquid portion. (author) [French] Les methodes de dosage presentees concernent les produits de fission a vie relativement longue presents dans les effluents de traitement des combustibles irradies (Sr - Cs - Ce - Zr - Nb - Ru - I). Elles sont toutes basees sur le meme principe: - addition d'entraineur, - separation chimique en plusieurs stades de purification, - determination du rendement chimique par calorimetrie, - comptage d'une aliquote liquide. (auteur)

  5. Thermochemical Study on the Sulfurization of Fission Products in Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Lee, Jung Won; Yang, M. S.; Park, G. I.; Kim, W. K.; Lee, J. W.

    2005-11-01

    The thermodynamic behavior of the sulfurization of Nd, and Eu element, which are contained in spent nuclear fuel as fission products was investigated through collection and properties analysis of thermodynamic data in sulfurization of uranium oxides, thermodynamic properties analysis for the oxidation and reduction of fission products, and test and analysis for sulfurization characteristics of Nd and Eu oxide. And also, analysis on thermodynamic data, such as M-O-S phase stability diagram and changes of Gibbs free energy for sulfurization of uranium and Nd 2 O 3 and Eu 2 O 3 were carried out. Nd 2 O 3 and Eu 2 O 3 are sulfurized into Nd 2 O 2 S and Eu 2 O 2 S or NdySx and EuySx at a range of 400 to 450 .deg. C, while uranium oxides, such as UO 2 and U 3 O 8 remain unreacted up to 450 .deg. C Formation of UOS at 500 .deg. C is initiated by sulfurization of uranium oxides. Hence, reaction temperature for the sulfurization of the Nd 2 O 3 and Eu 2 O 3 was selected as a 450 .deg. C

  6. A comparative analysis of the effect of gaseous fission products release on the thermal behaviour of oxide fuel rods

    International Nuclear Information System (INIS)

    Totev, T.L.; Kolev, I.G.

    1992-01-01

    Four different models of gaseous fission product release are compared in order to assess the relative effect of thermal characteristics of the fuel rods. The results show that the use of Weisman and EPRI models at a high burnup (over 50000 MW.d/tU) leads to almost the same figures of maximum fuel temperature and gas gap thermal conductivity. The use of Beyer-Hann (Betelle) and Pazdera-Valach (Rzez) models leads to under prediction of the fuel element thermal characteristics. A conclusion has been made that the Weisman model is the most suitable for the WWER-type fuel elements behaviour prediction. 10 refs., 7 figs

  7. Molten Salt Fuel Version of Laser Inertial Fusion Fission Energy (LIFE)

    International Nuclear Information System (INIS)

    Moir, R.W.; Shaw, H.F.; Caro, A.; Kaufman, L.; Latkowski, J.F.; Powers, J.; Turchi, P.A.

    2008-01-01

    Molten salt with dissolved uranium is being considered for the Laser Inertial Confinement Fusion Fission Energy (LIFE) fission blanket as a backup in case a solid-fuel version cannot meet the performance objectives, for example because of radiation damage of the solid materials. Molten salt is not damaged by radiation and therefore could likely achieve the desired high burnup (>99%) of heavy atoms of 238 U. A perceived disadvantage is the possibility that the circulating molten salt could lend itself to misuse (proliferation) by making separation of fissile material easier than for the solid-fuel case. The molten salt composition being considered is the eutectic mixture of 73 mol% LiF and 27 mol% UF 4 , whose melting point is 490 C. The use of 232 Th as a fuel is also being studied. ( 232 Th does not produce Pu under neutron irradiation.) The temperature of the molten salt would be ∼550 C at the inlet (60 C above the solidus temperature) and ∼650 C at the outlet. Mixtures of U and Th are being considered. To minimize corrosion of structural materials, the molten salt would also contain a small amount (∼1 mol%) of UF 3 . The same beryllium neutron multiplier could be used as in the solid fuel case; alternatively, a liquid lithium or liquid lead multiplier could be used. Insuring that the solubility of Pu 3+ in the melt is not exceeded is a design criterion. To mitigate corrosion of the steel, a refractory coating such as tungsten similar to the first wall facing the fusion source is suggested in the high-neutron-flux regions; and in low-neutron-flux regions, including the piping and heat exchangers, a nickel alloy, Hastelloy, would be used. These material choices parallel those made for the Molten Salt Reactor Experiment (MSRE) at ORNL. The nuclear performance is better than the solid fuel case. At the beginning of life, the tritium breeding ratio is unity and the plutonium plus 233 U production rate is ∼0.6 atoms per 14.1 MeV neutron

  8. Research problems of fission product behaviour in fuels of nuclear power plants and ways of their solution

    International Nuclear Information System (INIS)

    Sulaberidze, V.Sh.

    1988-01-01

    The most important problems of studying behaviour of fission products in fuel elements of maneouvrable nuclear power plants units are formulated. In-pile and out-of-pile investigation methods solving these problems are characterized in brief. 12 refs.; 2 figs

  9. 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

  10. Considerations in modelling the melting of fuel containing fission products and solute oxides

    International Nuclear Information System (INIS)

    Akbari, F.; Welland, M.J.; Lewis, B.J.; Thompson, W.T.

    2005-01-01

    It is well known that the oxidation of a defected fuel element by steam gives rise to an increase in O/U ratio with a consequent lowering of the incipient melting temperature. Concurrently, the hyperstoichiometry reduces the thermal conductivity thereby raising the centerline fuel pellet temperature for a fixed linear power. The development of fission products soluble in the UO 2 phase or, more important, the deliberate introduction of additive oxides in advanced CANDU fuel bundle designs further affects and generally lowers the incipient melting temperature. For these reasons, the modeling of the molten (hyperstoichiometric) UO 2 phase containing several solute oxides (ZrO 2 , Ln 2 O 3 and AnO 2 ) is advancing in the expectation of developing a moving boundary heat and mass transfer model aimed at better defining the limits of safe operating practice as burnup advances. The paper describes how the molten phase stability model is constructed. The redistribution of components across the solid-liquid interface that attends the onset of melting of a non-stoichiometric UO 2 containing several solutes will be discussed. The issues of how to introduce boundary conditions into heat transfer calculations consistent with the requirements of the Phase Rule will be addressed. The Stefan problem of a moving boundary associated with the solid/liquid interface sets this treatment apart from conventional heat and mass transfer problems. (author)

  11. 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

  12. 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

  13. Current status of the FASTGRASS/PARAGRASS models for fission product release from LWR fuel during normal and accident conditions

    International Nuclear Information System (INIS)

    Rest, J.; Zawadski, S.A.; Piasecka, M.

    1983-10-01

    The theoretical FASTGRASS model for the prediction of the behavior of the gaseous and volatile fission products in nuclear fuels under normal and transient conditions has undergone substantial improvements. The major improvements have been in the atomistic and bubble diffusive flow models, in the models for the behavior of gas bubbles on grain surfaces, and in the models for the behavior of the volatile fission products iodine and cesium. The thoery has received extensive verification over a wide range of fuel operating conditions, and can be regarded as a state-of-the-art model based on our current level of understanding of fission product behavior. PARAGRASS is an extremely efficient, mechanistic computer code with the capability of modeling steady-state and transient fission-product behavior. The models in PARAGRASS are based on the more detailed ones in FASTGRASS. PARAGRASS updates for the FRAPCON (PNL), FRAP-T (INEL), and SCDAP (INEL) codes have recently been completed and implemented. Results from an extensive FASTGRASS verification are presented and discussed for steady-state and transient conditions. In addition, FASTGRASS predictions for fission product release rate constants are compared with those in NUREG-0772. 21 references, 13 figures

  14. A Status of Art-Report on the Fission Products Behavior Released from Spent Fuel at High Temperature Conditions

    International Nuclear Information System (INIS)

    Park, Geun Il; Kim, J. H.; Lee, J. W.

    2003-04-01

    The experiments on the fission products release behavior from spent fuel at high temperature assuming reactor accident conditions have been carried out at Oak Ridge Nation Laboratory of USA in HI/VI tests, CEA of France in HEVA/VERCOS tests, AEA of England and CRNL of Canada in HOX test. The VEGA program to study the fission product release behavior from LWR irradiated fuel was recently initiated at JAERI. The key parameter affecting the fission product(FP) release behavior is temperature. In addition, other parameters such as fuel oxidation, burnup, pre-transient conditions are found to affect the FP releases considerably in the earlier tests. The atmosphere conditions such as oxidizing atmosphere (steam or air) or reducing atmosphere (hydrogen) can cause significant change of FPs release and transport behavior due to chemical forms of the reactive FPs which is dependent on the oxidation potential. The effect of fuel burnup on the Kr-85 or Cs-137 release showed that the release rates of these radionuclides increased with the increase of burnup, meaning that release rates are dominated by the atomic diffusions in the grains and they are primarily a function of temperature. However, the data on FPs release behavior using higher burnups above 50,000 MWD/MTU are not so many reported up to now. This report summarizes the test results of FPs release behavior in reactor accident conditions produced from other countries mentioned above. This review and analysis on earlier studies would be useful for predicting the release characteristics of FPs from domestic spent fuel. The release rates of fission gas or FPs from spent fuel at high temperature conditions during fabrication process of dry recycling fuel were also analyzed using many data obtained from earlier tests

  15. A summary of the assessment of fuel behaviour, fission product release and pressure tube integrity following a postulated large loss-of-coolant accident

    International Nuclear Information System (INIS)

    Langman, V.J.; Weaver, K.R.

    1984-05-01

    The Ontario Hydro analyses of fuel and pressure tube temperatures, fuel behaviour, fission product release and pressure tube integrity for large break loss-of-coolant accidents in Bruce A or Pickering A have been critically reviewed. The determinations of maximum fuel temperatures and fission product release are very uncertain, and pressure tube integrity cannot be assured where low steam flows are predicted to persist for times on the order of minutes

  16. FIPS: a process for the solidification of fission product solutions using a drum drier. [HTGR fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Halaszovich, St.; Laser, M.; Merz, E.; Thiele, D.

    1976-08-01

    A new process consisting of the steps concentration of the fission product solution, denitration of the solution by addition of formaldehyde, addition of glass-forming additives, drying of the slurry using a drum drier, melting of the dry product in the crucible by rising level in-pot-melting, and off-gas treatment and recovery of nitric acid is under development. A small plant with a capacity of 1 kg glass per hour has been tested in hot cells with fission product solutions from LWR fuel element reprocessing since December 1974. The equipment is very simple to operate and to control. No serious problems arose during operation.

  17. Euratom innovation in nuclear fission: Community research in reactor systems and fuel cycles

    International Nuclear Information System (INIS)

    Goethem, G. van; Hugon, M.; Bhatnagar, V.; Manolatos, P.; Deffrennes, M.

    2007-01-01

    The following questions are naturally at the heart of the current Euratom research and training framework programme:(1)What are the challenges facing the European Union nuclear fission research community in the short (today), medium (2010) and long term (2040)? (2)What kind of research and technological development (RTD) does Euratom offer to respond to these challenges, in particular in the area of reactor systems and fuel cycles? In the general debate about energy supply technologies there are challenges of both a scientific and technological (S/T) as well as an economic and political (E/P) nature. Though the Community research programme acts mainly on the former, there is nevertheless important links with Community policy. These not only exist in the specific area of nuclear policy, but also more generally as is depicted in the following figure. It is shown in the particular area of nuclear fission, to what extent Euratom research, education and innovation ('Knowledge Triangle' in above figure) respond to the following long-term criteria: (1) sustainability, (2) economics, (3) safety, and (4) proliferation resistance. Research and innovation in nuclear fission technology has broad and extended geographical, disciplinary and time horizons:- the community involved extends to all 25 EU Member States and beyond; - the research assembles a large variety of scientific disciplines; - three generations of nuclear power technologies (called II, III and IV) are involved, with the timescales extending from now to around the year 2040. To each of these three generations, a couple of challenges are associated (six in total):- Generation II (1970-2000, today): security of supply+environmental compatibility; - Generation III (around 2010): enhanced safety and competitiveness (economics); - Generation IV (around 2040): cogeneration of heat and power, and full recycling. At the European Commission (EC), the research related to nuclear reactor systems and fuel cycles is

  18. 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)

  19. Assessment of Fuel Analysis Methodology and Fission Product Release for 37-Element Fuel by Using the Latest IST Codes during Stagnation Feeder Break in CANDU

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Jung, Jong Yeob

    2009-09-01

    Feeder break accident is regarded as one of the design basis accident in CANDU reactor which results in a fuel failure. For a particular range of inlet feeder break sizes, the flow in the channel is reduced sufficiently that the fuel and fuel channel integrity can be significantly affected to have damage in the affected channel, while the remainder of the core remains adequately cooled. The flow in the downstream channel can be more or less stagnated due to a balance between pressure at the break on the upstream side and the reverse driving pressure between the break and the downstream end. In the extreme, this can lead to rapid fuel heatup and fuel damage and failure of the fuel channel similar to that associated with a severe channel flow blockage. Such an inlet feeder break scenario is called a stagnation break. In this report, the fuel analysis methodology and the assessment results of fission product inventory and release during the stagnation feeder break are described for conservatively assumed limiting channel. The accident was assumed to be occurred in the refurbished Wolsong unit 1 and the latest safety codes were used in the analysis. Fission product inventories during the steady state were calculated by using ELESTRES-IST 1.2 code. The whole analysis process was carried out by a script file which was programmed by Perl language. The perl script file was programmed to make all ELESTRES input files for each bundle and each ring based on the given power-burnup history and thermal-hydraulic conditions of the limiting channel and to perform the fuel analysis automatically. The fission product release during the transient period of stagnation feeder break was evaluated by applying Gehl model. The amounts of each isotope's release are conservatively evaluated for additional 2 seconds after channel failure. The calculated fission product releases are provided to the following dose assessment as a source term

  20. Fission products and nuclear fuel behaviour under severe accident conditions part 1: Main lessons learnt from the first VERDON test

    Science.gov (United States)

    Pontillon, Y.; Geiger, E.; Le Gall, C.; Bernard, S.; Gallais-During, A.; Malgouyres, P. P.; Hanus, E.; Ducros, G.

    2017-11-01

    This paper describes the first VERDON test performed at the end of September 2011 with special emphasis on the behaviour of fission products (FP) and actinides during the accidental sequence itself. Two other papers discuss in detail the post-test examination results (SEM, EPMA and SIMS) of the VERDON-1 sample. The first VERDON test was devoted to studying UO2 fuel behaviour and fission product releases under reducing conditions at very high temperature (∼2883 K), which was able to confirm the very good performance of the VERDON loop. The fuel sample did not lose its integrity during this test. According to the FP behaviour measured by the online gamma station (fuel sight), the general classification of the FP in relation to their released fraction is very accurate, and the burn-up effect on the release rate is clearly highlighted.

  1. 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)

  2. The role of fission products (noble metal particles) in spent fuel corrosion process in a failed container

    Energy Technology Data Exchange (ETDEWEB)

    Wu, L., E-mail: lwu59@uwo.ca [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada); Shoesmith, D.W. [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada); Univ. of Western Ontario, Surface Science Western, London, Ontario (Canada)

    2013-07-01

    The corrosion/dissolution of simulated spent fuel has been studied electrochemically. Fission products within the UO{sub 2} matrix are found to have significant effect on the anodic dissolution behaviour of the fuel. It is observed that H{sub 2}O{sub 2}oxidation is accelerated on the surfaces of doped noble metal (ε) particles existing in the fuel matrix. It is concluded that H{sub 2}O{sub 2} decomposition rather than UO{sub 2} corrosion should be the dominant reaction under high H{sub 2}O{sub 2} concentrations. (author)

  3. A new technique to measure fission-product diffusion coefficients in UO2 fuel

    International Nuclear Information System (INIS)

    Hocking, W.H.; Verrall, R.A.; Bushby, S.J.

    1999-01-01

    This paper describes a new out-reactor technique for the measurement of fission-product diffusion rates in UO 2 . The technique accurately simulates in-reactor fission-fragment effects: a thermal diffusion that is due to localized mixing in the fission track, radiation-enhanced diffusion that is due to point-defect creation by fission fragments, and bubble resolution. The technique utilizes heavy-ion accelerators - low energy (40 keV to 1 MeV) for fission-product implantation, high energy (72 MeV) to create fission-fragment damage effects, and secondary ion mass spectrometry (SIMS) for measuring the depth profile of the implanted species. Preliminary results are presented from annealing tests (not in the 72 MeV ion flux) at 1465 deg. C and 1650 deg. C at low and high concentrations of fission products. (author)

  4. Fission product retention in TRISO coated UO2 particle fuels subjected to HTR simulated core heating tests

    International Nuclear Information System (INIS)

    Baldwin, C.A.; Kania, M.J.

    1991-01-01

    Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbounded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 deg. C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 deg. C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 deg. C and above may exist. (author). 6 refs, 6 figs, 4 tabs

  5. Fission product retention in TRISCO coated UO2 particle fuels subjected to HTR simulated core heating tests

    International Nuclear Information System (INIS)

    Baldwin, C.A.; Kania, M.J.

    1990-11-01

    Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbonded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 degree C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 degree C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 degree C and above may exist. 6 refs., 6 figs., 4 tabs

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

    International Nuclear Information System (INIS)

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

    1985-06-01

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

  7. Storage of spent fuels: implementation of a research program on the risk of waste container rupture due to stress corrosion induced by fission products

    International Nuclear Information System (INIS)

    Parise, M.; Walle, E.; Foct, J.

    2001-01-01

    The following topics were dealt with: research programm on stress corrosion of spent fuel casks materials due to fission products, such as iodine, chemical interactions with zirconium, chemical aspects of stress corrosion, rupture risk assessment

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

    International Nuclear Information System (INIS)

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

    1986-10-01

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

  9. Long Term Fuel Sustainable Fission Energy Perspective Relevant for Combating Climate Change

    International Nuclear Information System (INIS)

    Knapp, V.; Matijevic, M.; Pevec, D.; Crnobrnja, B.; Lale, D.

    2016-01-01

    In recent research we outlined climate relevant and immediately available proven light water nuclear strategy with a potential to contribute essentially and timely to reduction of carbon dioxide emission to the year 2065. We consider in this paper what is the perspective of fission energy after that year should its contribution be needed. Singling out two technologies with long term perspective which need no or small amounts of uranium, fast breeders and molten salt thorium reactors, we consider their main technical and safety characteristics. In both of these technologies it is essential to have starter nuclides as neither U238 nor Th232 are fissile. We investigated whether plutonium from spent fuel of light water reactors generated to the year 2065 would be present in quantities sufficient to continue operation on the same or similar level in both technologies. However in operational safety, proliferation risks, waste production, in our judgement we must give preference to thorium technology, if it would be ready in second half of the century.(author).

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

    International Nuclear Information System (INIS)

    1987-09-01

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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  12. Fission product determination in irradiated fuel processing waste (electrophoresis); Dosage des produits de fission dans les effluents de traitement des combustibles irradies (electrophorese)

    Energy Technology Data Exchange (ETDEWEB)

    Auchapt, J M; Tret, J [Commissariat a l' Energie Atomique, Centre de Marcoule, 30 - Bagnols-sur-Ceze (France). Centre de Production de Plutonium de Marcoule. Services d' Extraction du Plutonium

    1966-07-01

    This dosage method concerns fission products present in the waste produced from the processing of cooled irradiated fuels. - Sr, Cs, Ce, Y, Ru by quantitative analysis; - Zr, Nb by qualitative analysis. It includes electrophoresis on paper strips one meter long which is then analysed between two window-less Geiger counters. For an activity of 10{sup -2} {mu}Ci of any cation in a 10 {mu}l spot, the standard error {sigma} if 3 to 4 per cent. complete analysis lasts about 5 hours. (authors) [French] Cette methode de dosage concerne les produits de fission presents dans les effluents de traitement des combustibles irradies refroidis: - Sr, Cs, Ce, Y, Ru en analyse quantitative; - Zr, Nb en analyse qualitative. Elle comporte une electrophorese sur bande de papier de un metre de longueur suivie d'un depouillement entre deux compteurs Geiger sans fenetre. Pour une activite de 10{sup -2} {mu}Ci d'un cation quelconque dans une tache de 10 {mu}l l'erreur standard {sigma} est de 3 a 4 pour cent. L'analyse complete demande environ 5 heures. (auteurs)

  13. Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium-Uranium-Plutonium Fuel Cycle

    Science.gov (United States)

    Marshalkin, V. Ye.; Povyshev, V. M.

    2017-12-01

    It is shown for a closed thorium-uranium-plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain 54 kg of fission products, 0.8 kg of thorium, 0.10 kg of uranium isotopes, 0.005 kg of plutonium isotopes, 0.002 kg of neptunium, and "trace" amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  15. Preliminary results of the BTF-104 experiment: an in-reactor test of fuel behaviour and fission-product release and transport under LOCA/LOECC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Dickson, L W; Elder, P H; Devaal, J W; Irish, J D; Yamazaki, A R [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1996-12-31

    The BTF-104 experiment is one of a series of in-reactor tests being performed to measure fuel behaviour and fission-product release from nuclear fuel subjected to accident conditions. The primary objective of the BTF-104 experiment was to measure fission-product releases from a CANDU-sized fuel element under combined Loss-of-Coolant Accident (LOCA) and Loss-of-Emergency-Core-Cooling (LOECC) conditions at an average fuel temperature of about 1550 deg C. The preliminary results of the BTF-104 experiment are presented in this paper. (author). 6 refs., 12 figs.

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  17. A separate effect study of the influence of metallic fission products on CsI radioactive release from nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Di Lemma, F.G., E-mail: fidelma.dilemma@gmail.com [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Department of Radiation Science and Technology, Faculty of Applied Sciences, Delft University of Technology, Delft, 2629 JB (Netherlands); Colle, J.Y., E-mail: jean-yves.colle@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Beneš, O. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Konings, R.J.M. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Department of Radiation Science and Technology, Faculty of Applied Sciences, Delft University of Technology, Delft, 2629 JB (Netherlands)

    2015-10-15

    The chemistry of cesium and iodine is of main importance to quantify the radioactive release in case of a nuclear reactor accident, or sabotage involving irradiated nuclear materials. We studied the interaction of CsI with different metallic fission products such as Mo and Ru. These elements can be released from nuclear fuel when exposed to oxidising conditions, as in the case of contact of overheated nuclear fuel with air (e.g. in a spent fuel cask sabotage, uncovering of a spent fuel pond, or air ingress accidents). Experiments were performed by vaporizing mixtures of the compounds in air, and analysing the produced aerosols in view of a possible gas–gas and gas–aerosol reactions between the compounds. These results were compared with the gaseous species predicted by thermochemical equilibrium calculations and experimental equilibrium vaporization tests using Knudsen Effusion Mass Spectrometry.

  18. 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)

  19. Delayed Fission Product Gamma-Ray Transmission Through Low Enriched UO2 Fuel Pin Lattices in Air

    Energy Technology Data Exchange (ETDEWEB)

    Trumbull, TH [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2004-10-18

    The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray dose on the outside of the lattice as a function of radial position, the so-called “channeling” effect, was analyzed. Techniques for performing experimental measurements and data reduction at Rensselaer Polytechnic Institute’s Reactor Critical Facility (RCF) were derived. An experimental apparatus was constructed to hold the arrangements of fuel pins for the measurements. A gamma-ray spectroscopy system consisting of a sodium-iodide scintillation detector was used to collect data. Measurements were made with and without a collimator installed. A point-kernel transport code was developed to map the radial dependence of the gamma-ray flux. Input files for the Monte Carlo code, MCNP, were also developed to accurately model the experimental measurements. The results of the calculations were compared to the experimental measurements. In order to determine the delayed fission-product gamma-ray source for the calculations, a technique was developed using a previously written code, DELBG and the reactor state-point data obtained during the experimental measurements. Calculations were performed demonstrating the effects of material homogenization on the gamma-ray transmission through the fuel pin lattice.Homogeneous and heterogeneous calculations were performed for all RCF fuel pin lattices as well as for a typical commercial pressurized water reactor fuel bundle. The results of the study demonstrated the effectiveness of the experimental measurements to isolate the channeling effect of delayed fission-product gamma-rays through lattices of RCF fuel pins

  20. 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)

  1. Assessment of pin-by-pin fission rate distribution within MOX/UO{sub 2} fuel assembly using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Louis, Heba Kareem; Amin, Esmat [Nuclear and Radiological Regulation Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2016-03-15

    The aim of the present paper is to assess the calculations of pin-by-pin group integrated fission rates within MOX/UO{sub 2} Fuel assemblies using the Monte Carlo code MCNP2.7c with two sets of the available latest nuclear data libraries used for calculating MOX-fueled systems. The data that are used in this paper are based on the benchmark by the NEA Nuclear Science Committee (NSC). The k{sub ∞} and absorption/fission reaction rates per isotope, k{sub eff} and pin-by-pin group integrated fission rates on 1/8 fraction of the geometry are determined. To assess the overall pin-by-pin fission rate distribution, the collective per cent error measures were investigated. The results of AVG, MRE and RMS error measures were less than 1 % error. The present results are compared with other participants using other Monte Carlo codes and with CEA results that were taken in the benchmark as reference. The results with ENDF/B-VI.6 are close to the results received by MVP (JENDL3.2) and SCALE 4.2 (JEF2.2). The results with ENDF/BVII.1 give higher values of k{sub ∞} reflecting the changes in the newer evaluations. In almost all results presented here, the MCNP calculated results with ENDF/B VII.1 should be considered more than those obtained by using other Monte Carlo codes and nuclear data libraries. The present calculations may be consider a reference for evaluating the numerical schemes in production code systems, as well as the global performance including cross-section data reduction methods as the calculations used continuous energy and no geometrical approximations.

  2. 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.

  3. Gamma-ray spectrometric measurements of fission rate ratios between fresh and burnt fuel following irradiation in a zero-power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kröhnert, H., E-mail: hanna.kroehnert@ensi.ch [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); École Polytechnique Fédérale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland); Perret, G.; Murphy, M.F. [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Chawla, R. [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); École Polytechnique Fédérale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

    2013-01-11

    The gamma-ray activity from short-lived fission products has been measured in fresh and burnt UO{sub 2} fuel samples after irradiation in a zero-power reactor. For the first time, short-lived gamma-ray activity from fresh and burnt fuel has been compared and fresh-to-burnt fuel fission rate ratios have been derived. For the measurements, well characterized fresh and burnt fuel samples, with burn-ups up to 46 GWd/t, were irradiated in the zero-power research reactor PROTEUS. Fission rate ratios were derived based on the counting of high-energy gamma-rays above 2200 keV, in order to discriminate against the high intrinsic activity of the burnt fuel. This paper presents the measured fresh-to-burnt fuel fission rate ratios based on the {sup 142}La (2542 keV), {sup 89}Rb (2570 keV), {sup 138}Cs (2640 keV) and {sup 95}Y (3576 keV) high-energy gamma-ray lines. Comparisons are made with the results of Monte Carlo modeling of the experimental configuration, carried out using the MCNPX code. The measured fission rate ratios have 1σ uncertainties of 1.7–3.4%. The comparisons with calculated predictions show an agreement within 1–3σ, although there appears to be a slight bias (∼3%).

  4. 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.)

  5. Fission yields data generation and benchmarks of decay heat estimation of a nuclear fuel

    Science.gov (United States)

    Gil, Choong-Sup; Kim, Do Heon; Yoo, Jae Kwon; Lee, Jounghwa

    2017-09-01

    Fission yields data with the ENDF-6 format of 235U, 239Pu, and several actinides dependent on incident neutron energies have been generated using the GEF code. In addition, fission yields data libraries of ORIGEN-S, -ARP modules in the SCALE code, have been generated with the new data. The decay heats by ORIGEN-S using the new fission yields data have been calculated and compared with the measured data for validation in this study. The fission yields data ORIGEN-S libraries based on ENDF/B-VII.1, JEFF-3.1.1, and JENDL/FPY-2011 have also been generated, and decay heats were calculated using the ORIGEN-S libraries for analyses and comparisons.

  6. Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

    2016-06-01

    Low-enriched (U-235 RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

  7. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations

  8. Release of fission products from irradiated SRP fuels at elevated temperature. Data report on the first stage of the SRP source term study

    International Nuclear Information System (INIS)

    Woodley, R.E.

    1986-06-01

    For a sound evaluation of the consequences of a hypothetical nuclear reactor accident, a knowledge of the extent of fission product release from the fuel at anticipated temperatures and atmosphere conditions is required. Measurements of fission product release have been performed with a variety of nuclear fuels under various conditions of temperature and atmosphere. While the use of data obtained on fuels similar to the fuel of interest may provide a reasonable estimate of release fractions, precise information of this nature can only be obtained from measurements employing specimens of the actual fuels used in the nuclear reactor under consideration. The two fuels of interest in the present study are an alloy, a dispersion of UAl 4 in an aluminum matrix, and a cermet, a dispersion of U 3 O 8 in an aluminum matrix. Both fuels are clad in aluminum

  9. 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.)

  10. 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.

  11. 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.

  12. Computational analysis of modern HTGR fuel performance and fission product release during the HFR-EU1 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl, E-mail: k.verfondern@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Xhonneux, André, E-mail: xhonneux@lrst.rwth-aachen.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Nabielek, Heinz, E-mail: heinznabielek@me.com [Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Allelein, Hans-Josef, E-mail: h.j.allelein@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, Chair for Reactor Safety and Reactor Technology, 52072 Aachen (Germany)

    2014-07-01

    Highlights: • HFR-EU1 irradiation test demonstrates high quality of HTGR spherical fuel elements. • Irradiation performance is in good agreement with German fuel performance modeling. • International benchmark exercise expected first particle to fail at ∼13–17% FIMA. • EOL silver release is predicted to be in the percentage range. • EOL cesium and strontium are expected to remain at a low level. - Abstract: Various countries engaged in the development and fabrication of modern HTGR fuel have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under HTGR operating and accident conditions. Verification and validation studies are conducted by code-to-code benchmarking and code-to-experiment comparisons as part of international exercises. The methodology developed in Germany since the 1980s represents valuable and efficient tools to describe fission product release from spherical fuel elements and TRISO fuel performance, respectively, under given conditions. Continued application to new results of irradiation and accident simulation testing demonstrates the appropriateness of the models in terms of a conservative estimation of the source term as part of interactions with HTGR licensing authorities. Within the European irradiation testing program for HTGR fuel and as part of the former EU RAPHAEL project, the HFR-EU1 irradiation experiment explores the potential for high performance of the presently existing German and newly produced Chinese fuel spheres under defined conditions up to high burnups. The fuel irradiation was completed in 2010. Test samples are prepared for further postirradiation examinations (PIE) including heatup simulation testing in the KÜFA-II furnace at the JRC-ITU, Karlsruhe, to be conducted within the on-going ARCHER Project of the European Commission. The paper will describe the application of the German computer models to the HFR-EU1 irradiation test and

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  14. Releases of radioiodine from the Karlsruhe nuclear fuel reprocessing plant as a result of spontaneous fission of actinides

    International Nuclear Information System (INIS)

    Schuettelkopf, H.

    1977-02-01

    Fro, 23,7,1976 to 28.7.76 and from 8.3.76 to 9.16.76 50 pCi 131 I/m 3 , 116 pCi 133 I/m 3 und 195 pCi 135 I/m 3 were measured on an average in 11 samples of waste air from the Karlsruhe Nuclear Fuel Reprocessing Plant (WAK). During these time intervals no dissolution of fuel material was performed. From 16.9.76 to 27.10.76 18 charges of nuclear fuel were dissolved. During this period 3.3 pCi 131 I/m 3 and 7.9 pCi 133 I/m 3 were obtained as mean waste air concentrations which were higher than the lower detection limit of the method of measurement used. 244 Cm, 242 Cm, 242 Pu, 240 Pu and 238 Pu are responsible for the production of radioiodine in nuclear fuel by spontaneous fission. 244 Cm is the most important nuclide in highly active waste solutions (HAL). The cumulative fission yield is well approximated by 3% for 13 I and by 6% for 133 I. The radioiodine is set free during fuel dissolution by venting of tanks and HAL pipes and during the vritification of such solutions. The radioiodine produced by spontaneous fission is released from WAK only by venting of tanks and HAL pipes. Corresponding to the conditions of venting, air concentrations as high as 4.4 pCi 131 I/m 3 and 8.2 pCi 133 I/m 3 are expected. These concentrations agree well with air concentrations measured during the period of fuel dissolution. Based on plausible assumptions the 131 I and 133 I waste air concentrations for the period of outage are calculated from an evaporated volume of HAL in the pipes corresponding to about 10 g of 244 Cm and with 40% equilibrium between I 2 in evaporated HAL and in waste air. In the worst case 131 I-concentrations in the waste air of WAK result in an annual release of 0.2 mCi 131 I. This value is less than 1% of the authorized annual releases of 1976. For a reprocessing plant of 1,400 t/a capacity the annual expected release of 131 I lies in the mCi range. (orig.) [de

  15. 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

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

    International Nuclear Information System (INIS)

    Allelein, H.J.

    1980-12-01

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

  17. 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.)

  18. 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)

  19. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  20. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  1. Effects of fuel particle size and fission-fragment-enhanced irradiation creep on the in-pile behavior in CERCER composite pellets

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yunmei [Institute of Mechanics and Computational Engineering, Department of Aeronautics and Astronautics, Fudan University, Shanghai 200433 (China); Ding, Shurong, E-mail: dsr1971@163.com [Institute of Mechanics and Computational Engineering, Department of Aeronautics and Astronautics, Fudan University, Shanghai 200433 (China); Zhang, Xunchao; Wang, Canglong; Yang, Lei [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China)

    2016-12-15

    The micro-scale finite element models for CERCER pellets with different-sized fuel particles are developed. With consideration of a grain-scale mechanistic irradiation swelling model in the fuel particles and the irradiation creep in the matrix, numerical simulations are performed to explore the effects of the particle size and the fission-fragment-enhanced irradiation creep on the thermo-mechanical behavior of CERCER pellets. The enhanced irradiation creep effect is applied in the 10 μm-thick fission fragment damage matrix layer surrounding the fuel particles. The obtained results indicate that (1) lower maximum temperature occurs in the cases with smaller-sized particles, and the effects of particle size on the mechanical behavior in pellets are intricate; (2) the first principal stress and radial axial stress remain compressive in the fission fragment damage layer at higher burnup, thus the mechanism of radial cracking found in the experiment can be better explained. - Highlights: • A grain-scale gas swelling model considering the development of recrystallization and resolution is adopted for particles. • The influence of fission-gas-induced porosity is considered in the constitutive relations for particles. • A simulation method is developed for the multi-scale thermo-mechanical behavior. • The effects of fuel particle size and fission-fragment-enhanced irradiation creep are investigated in pellets.

  2. 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.

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

    International Nuclear Information System (INIS)

    Dumas, J.M.; Lhiaubet, G.

    1989-07-01

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

  4. Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

    Science.gov (United States)

    Seibert, Rachel L.; Terrani, Kurt A.; Velázquez, Daniel; Hunn, John D.; Baldwin, Charles A.; Montgomery, Fred C.; Terry, Jeff

    2018-03-01

    The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Program and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form PdxSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. They may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.

  5. Freshly induced short-lived gamma-ray activity as a measure of fission rates in lightly re-irradiated spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kroehnert, H., E-mail: hanna.kroehnert@psi.c [Paul Scherrer Institut (PSI), OPRA-E07, CH-5232 Villigen (Switzerland); Perret, G., E-mail: gregory.perret@psi.c [Paul Scherrer Institut (PSI), OPRA-E07, CH-5232 Villigen (Switzerland); Murphy, M.F., E-mail: mike.murphy@psi.c [Paul Scherrer Institut (PSI), OPRA-E07, CH-5232 Villigen (Switzerland); Chawla, R., E-mail: rakesh.chawla@epfl.c [Paul Scherrer Institut (PSI), OPRA-E07, CH-5232 Villigen (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

    2010-12-01

    A new measurement technique has been developed to determine fission rates in burnt fuel, following re-irradiation in a zero-power research reactor. The development has been made in the frame of the LIFE-PROTEUS program at the Paul Scherrer Institute, which aims at characterizing the interfaces between fresh and highly burnt fuel assemblies in modern LWRs. To discriminate against the high intrinsic gamma-ray activity of the burnt fuel, the proposed measurement technique uses high-energy gamma-rays, above 2000 keV, emitted by short-lived fission products freshly produced in the fuel. To demonstrate the feasibility of this technique, a fresh UO{sub 2} sample and a 36 GWd/t burnt UO{sub 2} sample were irradiated in the PROTEUS reactor and their gamma-ray activities were recorded directly after irradiation. For both fresh and the burnt fuel samples, relative fission rates were derived for different core positions, based on the short-lived {sup 142}La (2542 keV), {sup 89}Rb (2570 keV), {sup 138}Cs (2640 keV) and {sup 95}Y (3576 keV) gamma-ray lines. Uncertainties on the inter-position fission rate ratios were mainly due to the uncertainties on the net-area of the gamma-ray peaks and were about 1-3% for the fresh sample, and 3-6% for the burnt one. Thus, for the first time, it has been shown that the short-lived gamma-ray activity, induced in burnt fuel by irradiation in a zero-power reactor, can be used as a quantitative measure of the fission rate. For both fresh and burnt fuel, the measured results agreed, within the uncertainties, with Monte Carlo (MCNPX) predictions.

  6. Utilization of freshly induced high-energy gamma-ray activity as a measure of fission rates in re-irradiated burnt UO{sub 2} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, M. F.; Perret, G. [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Krohnert, H.; Chawla, R. [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

    2009-07-01

    In the frame of the LIFE-PROTEUS (Large-scale Irradiation Fuel Experiments at PROTEUS) program, a measurement technique is being developed to measure fission rates in burnt fuel, following re-irradiation in a zero-power research reactor. In the presented approach, the fission rates are estimated by measuring high energy gamma-rays (above 2000 keV) emitted by short-lived fission products freshly produced in the fuel. Due to their high energies, these gamma-rays can be discriminated against the high intrinsic gamma-ray activity of the burnt fuel, which reaches energies up to 2000 keV. To demonstrate the feasibility of this approach, fresh and burnt fuel samples (with burn-ups varying from 36 to 64 MWd/kg) were irradiated in the PROTEUS reactor at the Paul Scherrer Institut, and their emitted gamma-ray spectra were recorded shortly after irradiation. It was possible, for the first time, to detect the short-lived gamma-ray activity in the high-energy region, even in the presence of the intrinsic gamma-ray background of the burnt fuel samples. Using the short-lived gamma-ray lines {sup 142}La (2542 keV), {sup 89}Rb (2570 keV), 95Y (2632 keV), {sup 138}Cs (2640 keV) and {sup 95}Y (3576 keV), relative fission rates between different core positions were derived for a fresh sample as well as for a burnt sample with a burn-up of 36 MWd/kg. It was shown that, for both the fresh and burnt fuel samples, the measured fission rate ratios agreed well, i.e. within the statistical uncertainties, with calculation results obtained by Monte Carlo simulations. (authors)

  7. Minimization of the fission product waste by using thorium based fuel instead of uranium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, A. Abdelghafar, E-mail: Agalahom@yahoo.com

    2017-04-01

    This research discusses the neutronic characteristics of VVER-1200 assembly fueled with five different fuel types based on thorium. These types of fuel based on mixing thorium as a fertile material with different fissile materials. The neutronic characteristics of these fuels are investigated by comparing their neutronic characteristics with the conventional uranium dioxide fuel using the MCNPX code. The objective of this study is to reduce the production of long-lived actinides, get rid of plutonium component and to improve the fuel cycle economy while maintaining acceptable values of the neutronic safety parameters such as moderator temperature coefficient, Doppler coefficient and effective delayed neutrons (β). The thorium based fuel has a more negative Doppler coefficient than uranium dioxide fuel. The moderator temperature coefficient (MTC) has been calculated for the different proposed fuels. Also, the fissile inventory ratio has been calculated at different burnup step. The use of Th-232 as a fertile material instead of U-238 in a nuclear fuel is the most promising fuel in VVER-1200 as it is the ideal solution to avoid the production of more plutonium components and long-lived minor actinides. The reactor grade plutonium accumulated in light water reactor with burnup can be recycled by mixing it with Th-232 to fuel the VVER-1200 assembly. The concentrations of Xe-135 and Sm-151 have been investigated, due to their high thermal neutron absorption cross section.

  8. 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)

  9. A study of fission product transport from failed fuel during N reactor postulated accidents

    International Nuclear Information System (INIS)

    Hagrman, D.L.

    1989-09-01

    This report presents a study of fission product transport behavior in N Reactor during a severe accident. More detail about fission product behavior than has previously been available is provided and key parameters that control this behavior are identified. The current report is an extension to a previous interum study that has added an aerosol formation model, replaced an older aerosol deposition model with an improved correlation, and incorporated results of a revised analysis of the process tubes. The LACE LA1 and LA3 tests are used to assess the revised model applied to determine aerosol deposition. The study concludes that a cesium iodide aerosol is likely to form near the downstream end of the process tubes. Transport of most of the released cesium and iodine as well as less volatile material depends on the behavior of this aerosol and the behavior is sensitive to several parameters that are not well known. If the environment is very clean and effluent flow is sufficient to support oxidation of the zircaloy and uranium of the process tubes, almost none of the aerosol deposits in the riser. Reduction of the effluent flow or the presence of high concentrations of aerosols of very low volatile material like zirconium, uranium, or their oxides causes deposition of the fission products in the riser piping. 24 refs., 18 figs., 11 tabs

  10. Modeling defect and fission gas properties in U-Si fuels

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Noordhoek, Mark [Univ. of South Carolina, Columbia, SC (United States); Besmann, Theodore [Univ. of South Carolina, Columbia, SC (United States); Middleburgh, Simon C. [Westinghouse Electric Sweden, Vasteras (Sweden); Lahoda, E. J. [Westinghouse Electric Company LLC, Cranberry Woods, PA (United States); Chernatynskiy, Aleksandr [Missouri University of Science and Technology; Grimes, Robin W. [Imperial College, London (United Kingdom)

    2017-04-27

    Uranium silicides, in particular U3Si2, are being explored as an advanced nuclear fuel with increased accident tolerance as well as competitive economics compared to the baseline UO2 fuel. They benefit from high thermal conductivity (metallic) compared to UO2 fuel (insulator or semi-conductor) used in current Light Water Reactors (LWRs). The U-Si fuels also have higher fissile density. In order to perform meaningful engineering scale nuclear fuel performance simulations, the material properties of the fuel, including the response to irradiation environments, must be known. Unfortunately, the data available for USi fuels are rather limited, in particular for the temperature range where LWRs would operate. The ATF HIP is using multi-scale modeling and simulations to address this knowledge gap.

  11. Modeling defect and fission gas properties in U-Si fuels

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Noordhoek, Mark J. [Univ. of South Carolina, Columbia, SC (United States); Besmann, Theodore M. [Univ. of South Carolina, Columbia, SC (United States); Middleburgh, Simon C. [Westinghouse Electric Sweden, Vasteras (Sweden); Lahoda, E. J. [Westinghouse Electric Company LLC, Cranberry Woods, PA (United States); Chernatynskiy, Aleksandr [Missouri Univ. of Science and Technology, Rolla, MO (United States); Grimes, Robin W. [Imperial College, London (United Kingdom)

    2017-04-14

    Uranium silicides, in particular U3Si2, are being explored as an advanced nuclear fuel with increased accident tolerance as well as competitive economics compared to the baseline UO2 fuel. They benefit from high thermal conductivity (metallic) compared to UO2 fuel (insulator or semi-conductor) used in current Light Water Reactors (LWRs). The U-Si fuels also have higher fissile density. In order to perform meaningful engineering scale nuclear fuel performance simulations, the material properties of the fuel, including the response to irradiation environments, must be known. Unfortunately, the data available for USi fuels are rather limited, in particular for the temperature range where LWRs would operate. The ATF HIP is using multi-scale modeling and simulations to address this knowledge gap.

  12. Proportioning of 79Se and 126Sn long life radionuclides in the fission products solutions coming from spent fuels processing

    International Nuclear Information System (INIS)

    Comte, J.

    2001-11-01

    The determination of radionuclides present in waste resulting from the nuclear fuel reprocessing is a request from the regulatory authorities to ensure an optimal management of the storage sites. Long-lived radionuclides (T 1/2 > 30 years) are particularly concerned owing to the fact that their impact must be considered for the long term. Safety studies have established a list of long-lived radionuclides (LLRN) whose quantification is essential for the management of the disposal site. Among these, several are pure β emitters, present at low concentration levels in complex matrices. Their determination, by radiochemical method or mass spectrometry, involves selective chemical separations from the others β/γ emitters and from the measurement interfering elements. The work undertaken in this thesis relates to the development of analytical methods for the determination of two long-lived radionuclides: selenium 79 and tin 126, in acid solutions of fission products present in nuclear fuel reprocessing plant. For selenium 79, a β emitter with a half live estimated to be 10 6 years, the bibliography describes different chemical separation methods including precipitation, liquid-liquid extraction and chromatography on ionic resins. After optimisation on a synthetic solution, two of these techniques, precipitation by potassium iodine and separation with ion exchange resins were applied to a genuine solution of fission products at Cogema La Hague. The results showed that only the ion exchange method allows us to obtain a solution sufficiently decontaminated (FDβγ = 250) with a significant selenium recovery yield (85%). This separation allows the measurement of the 79 Se by electrothermal vaporization coupled with inductively coupled plasma mass spectrometry (ETV-ICP/MS), after transfer of the samples to CEA/Cadarache. The concentration of 79 Se measured is 0,42 mg/L in the solution of fission products with an isotopic ratio 79 Se/ 82 Se equal to that recommended by the

  13. Experimental validation of CASMO-4E and CASMO-5M for radial fission rate distributions in a westinghouse SVEA-96 Optima2 BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Grimm, P.; Perret, G. [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland)

    2012-07-01

    Measured and calculated radial total fission rate distributions are compared for the three axial sections of a Westinghouse SVEA-96 Optima2 BWR fuel assembly, comprising 96, 92 and 84 fuel rods, respectively. The measurements were performed on a full-size fuel assembly in the PROTEUS zero-power experimental facility. The measured fission rates are compared to the results of the CASMO-4E and CASMO-5M fuel assembly codes. Detailed measured geometrical data were used in the models, and effects of the surrounding zones of the reactor were taken into account by correction factors derived from MCNPX calculations. The results of the calculations agree well with those of the experiments, with root-mean-square deviations between 1.2% and 1.5% and maximum deviations of 3-4%. The quality of the predictions by CASMO-4E and CASMO-5M is comparable. (authors)

  14. Criticality accident in uranium fuel processing plant. The estimation of the total number of fissions with related reactor physics parameters

    International Nuclear Information System (INIS)

    Nishina, Kojiro; Oyamatsu, Kazuhiro; Kondo, Shunsuke; Sekimoto, Hiroshi; Ishitani, Kazuki; Yamane, Yoshihiro; Miyoshi, Yoshinori

    2000-01-01

    This accident occurred when workers were pouring a uranium solution into a precipitation tank with handy operation against the established procedure and both the cylindrical diameter and the total mass exceeded the limited values. As a result, nuclear fission chain reactor in the solution reached not only a 'criticality' state continuing it independently but also an instantly forming criticality state exceed the criticality and increasing further nuclear fission number. The place occurring the accident at this time was not reactor but a place having not to form 'criticality' called by a processing process of uranium fuel. In such place, as because of relating to mechanism of chain reaction, it is required naturally for knowledge on the reactor physics, it is also necessary to understand chemical reaction in chemical process, and functions of tanks, valves and pumps mounted at the processes. For this purpose, some information on uranium concentration ratio, atomic density of nuclides largely affecting to chain reaction such as uranium, hydrogen, and so forth in the solution, shape, inner structure and size of container for the solution, and its temperature and total volume, were necessary for determining criticality volume of the accident uranium solution by using nuclear physics procedures. Here were described on estimation of energy emission in the JCO accident, estimation from analytical results on neutron and solution, calculation of various nuclear physics property estimation on the JCO precipitation tank at JAERI. (G.K.)

  15. Determination of fission product and heavy metals inventories in FTE-4 fuel rods by a grind-burn-leach flowsheet

    International Nuclear Information System (INIS)

    Fitzgerald, C.L.; Vaughen, V.C.A.; Lamb, C.E.

    1977-07-01

    Experiments using High-Temperature Gas-Cooled Reactor (HTGR) fuel material, TRISO-coated (2.75 Th/U)C 2 --TRISO-coated ThC 2 and TRISO-coated UO 2 --BISO-coated ThO 2 , were performed in Building 4507 (the High-Level Chemical Development Facility) to determine the inventory and transport behavior of fission products and heavy metals from a grind-burn-leach process flowsheet. In addition, values calculated by the ORNL Isotope Generation and Depletion Code (ORIGEN, a computer program used for predicting quantities of activation products, actinides, and fission products from irradiation data and nuclear data libraries) are compared with values derived by chemical analyses (CA) and those measured by a gamma-scan nondestructive analytical (NDA) technique. Reasonable agreement was obtained between ORIGEN and NDA results for one of the tests, but the values obtained by chemical analysis were lower than either of the two other sets of values. With the exception of 234 U, isotopic uranium values determined by chemical analysis (mass spectrometry) agreed within 15 percent of the ORIGEN prediction

  16. Benchmark criticality experiments for fast fission configuration with high enriched nuclear fuel

    International Nuclear Information System (INIS)

    Sikorin, S.N.; Mandzik, S.G.; Polazau, S.A.; Hryharovich, T.K.; Damarad, Y.V.; Palahina, Y.A.

    2014-01-01

    Benchmark criticality experiments of fast heterogeneous configuration with high enriched uranium (HEU) nuclear fuel were performed using the 'Giacint' critical assembly of the Joint Institute for Power and Nuclear Research - Sosny (JIPNR-Sosny) of the National Academy of Sciences of Belarus. The critical assembly core comprised fuel assemblies without a casing for the 34.8 mm wrench. Fuel assemblies contain 19 fuel rods of two types. The first type is metal uranium fuel rods with 90% enrichment by U-235; the second one is dioxide uranium fuel rods with 36% enrichment by U-235. The total fuel rods length is 620 mm, and the active fuel length is 500 mm. The outer fuel rods diameter is 7 mm, the wall is 0.2 mm thick, and the fuel material diameter is 6.4 mm. The clad material is stainless steel. The side radial reflector: the inner layer of beryllium, and the outer layer of stainless steel. The top and bottom axial reflectors are of stainless steel. The analysis of the experimental results obtained from these benchmark experiments by developing detailed calculation models and performing simulations for the different experiments is presented. The sensitivity of the obtained results for the material specifications and the modeling details were examined. The analyses used the MCNP and MCU computer programs. This paper presents the experimental and analytical results. (authors)

  17. Detection of fission products release in the research reactor 'RA' spent fuel storage pool

    International Nuclear Information System (INIS)

    Matausek, M.V.; Vukadin, Z.; Pavlovic, S.; Maksin, T.; Idakovic, Z.; Marinkovic, N.

    1997-05-01

    Spent fuel resulting from 25 years of operating the 6.5/10 MW thermal heavy water moderated and cooled research reactor RA at the VINCA Institute is presently all stored in the temporary spent fuel storage pool in the basement of the reactor building. In 1984, the reactor was shut down for refurbishment, which for a number of reasons has not yet been completed. Recent investigations show that independent of the future status of the research reactor, safe disposal of the so far irradiated fuel must be the subject of primary concern. The present status of the research reactor RA spent fuel storage pool at the VINCA Institute presents a serious safety problem. Action is therefore initiated in two directions. First, safety of the existing spent fuel storage should be improved. Second, transferring spent fuel into another, presumably dry storage space should be considered. By storing the previously irradiated fuel of the research reactor RA in a newly built storage space, sufficient free space will be provided in the existing spent fuel storage pool for the newly irradiated fuel when the reactor starts operation again. In the case that it would be decided to decommission the research reactor RA, the newly built storage space would provide safe disposal for the fuel irradiated so far

  18. Micrographic study on distribution of fission products in high burn-up metallic alloy fuel

    International Nuclear Information System (INIS)

    Kolay, S.; Basu, M.; Das, D.

    2012-01-01

    One of the important mandates in the three-stage nuclear power generation programme of India is to utilize uranium-plutonium based alloy fuels in enabling shorter doubling time for breeding of the fissile isotopes ( 239 Pu and 233 U ) to be used in thorium based driver fuel in the third stage. Reported information shows the successful performance of alloy fuel with somewhat porous matrix in achieving 10-15 atom% burnup. The porosity and microstructure of these alloys are strongly dependent on their composition and phases present. Porosity also influences the extent of fuel swelling and gas release. So to assess fuel performance and fuel integrity under high burn-up condition it is essential to have knowledge about the new phases formed and their redistribution that occurs as a result of inter-diffusion and temperature gradient. This study addresses these issues taking the base alloy U-10 wt %Zr

  19. Uranium content of coal ashes from Southern Brazil coal fueled power stations, by the fission track registration technique

    International Nuclear Information System (INIS)

    Morales, R.K.

    1981-01-01

    The feasibility of the application of the fission track registration technique for the determination of uranium in coal ashes was shown. The wet method was employed using as detector the Makrofol KG=10 μm, manufactured by Bayer. The coal ashes were originated from coal-fueled power stations localized in Southern Brazil. The results obtained ranged from 10 to 27 mg U/kg. Since the total error variation was from 18,4% to 23,8%, the method used was considered excellent. The determination of the uranium content in coal ashes is of considerable interest in environmental control in power stations, in their vicinity and wherever these ashes are used or stored. The technique used is the work proved to be very appropriate for the purpose aimed at. (Author) [pt

  20. The future implications of some long-lived fission product nuclides discharged to the environment in fuel reprocessing wastes

    International Nuclear Information System (INIS)

    Bryant, P.M.; Jones, J.A.

    1972-12-01

    Current reprocessing practice leads to the discharge to the environment of virtually all the krypton-85 and tritium, and a large fraction of the iodine-129, formed as fission products in reactor fuel. As nuclear power programmes expand the global inventory of these long-lived nuclides is increasing. The radiological significance of these discharges is assessed in terms of radiation exposure of various population groups during the next few decades. The results of this assessment show that krypton-85 will give higher dose rates than tritium or iodine-129, but that on conventional radiological protection criteria these do not justify taking action to remove krypton-85 from reprocessing plant effluents before the 21st century. (author)

  1. Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

    Energy Technology Data Exchange (ETDEWEB)

    Lemmens, K., E-mail: klemmens@sckcen.be [Waste and Disposal Expert Group, Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); González-Robles, E.; Kienzler, B. [Karlsruhe Institute of Technology Institute for Nuclear Waste Disposal (KIT-INE), PO Box 3640, D-76021 Karlsruhe (Germany); Curti, E. [Laboratory for Waste Management, Nuclear Energy and Safety Dept., Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Serrano-Purroy, D. [European Commission, DG Joint Research Centre - JRC, Directorate G - Nuclear Safety & Security, Department G.III, PO Box 2340, D-76125 Karlsruhe (Germany); Sureda, R.; Martínez-Torrents, A. [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Roth, O. [Studsvik, Nuclear AB, 611 82 Nyköping (Sweden); Slonszki, E. [Magyar Tudományos Akadémia Energiatudományi Kutatóközpont (MTA EK), PO Box 49, H-1525 Budapest (Hungary); Mennecart, T. [Waste and Disposal Expert Group, Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); Günther-Leopold, I. [Laboratory for Waste Management, Nuclear Energy and Safety Dept., Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Hózer, Z. [Magyar Tudományos Akadémia Energiatudományi Kutatóközpont (MTA EK), PO Box 49, H-1525 Budapest (Hungary)

    2017-02-15

    The instant release of fission products from high burn-up UO{sub 2} fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45–63 GWd/t{sub HM} and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride – bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H{sub 2} atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways. - Highlights: • Leach tests were performed to study the instant release of fission products from high burn-up UO{sub 2} fuels and one MOX fuel. • In these tests, the fission gas release given by the operator was a pessimistic estimator of the iodine and cesium release. • Iodine and cesium release is proportional to linear power rating beyond 200 W cm{sup −1}. • Closure of the fuel-cladding gap at high burn-up slows down the release. • The release rate decreases following an exponential equation.

  2. Nuclear fission energy: new build, operation, fuel cycle and decommissioning in the international perspective

    Energy Technology Data Exchange (ETDEWEB)

    Niessen, Stefan [AREVA GmbH, Erlangen (Germany)

    2015-07-01

    Over 60 nuclear power reactors are in construction today and over 400 are connected to the grid. The presentation will show where. A nuclear new build project involves a team of several thousand people. Some pictures from ongoing new build projects will illustrate this. Using concrete examples from the AREVA group, the nuclear fuel cycle from uranium mines in Niger, Kazakhstan or Canada to chemical conversion, enrichment and fuel manufacturing will be explained. Also the recycling of used fuel and the fabrication of MOX fuel is addressed. The presentation closes with an overview on decommissioning and final storage projects.

  3. 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)

  4. Fission gas detection system

    International Nuclear Information System (INIS)

    Colburn, R.P.

    1984-01-01

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

  5. Tokamak hybrid thermonuclear reactor for the production of fissionable fuel and electric power

    International Nuclear Information System (INIS)

    Velikhov, E.P.; Glukhikh, V.A.; Gur'ev, V.V.

    1978-01-01

    The results of feasibility studies of a tokamak- based hybrid reactor concept are presented. The system selected has a D-T plasma volume of 575 m 3 with additional plasma heating by injection of fast neutral particles. The method of heating makes it possible to achieve an economical two-component tokamak regime at ntau=(4-6)x10 13 sxcm -3 , i e. far below the Lawson criterion. Plasma and vacuum chamber are surrounded by a blanket where fissionable plutonium is produced and heat transformed into electric power is generated. Major plasma-neutron-physical characteristics of the 6905 MWth (2500 MWe) reactor and its electromagnetic system are presented. Evaluations show that the hybrid reactor can produce about 800 kg of Pu per 1GWth/yr as compared to 70-150 kg of Pu for fast breeder reactors. The increased Pu production rate is the major merit of the concept promising for both power generation and fuelling thermal fission reactions

  6. Comparison Of 252Cf Time Correlated Induced Fisssion With AmLi Induced Fission On Fresh MTR Research Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, Jay Prakash [Los Alamos National Laboratory

    2017-03-30

    The effective application of international safeguards to research reactors requires verification of spent fuel as well as fresh fuel. To accomplish this goal various nondestructive and destructive assay techniques have been developed in the US and around the world. The Advanced Experimental Fuel Counter (AEFC) is a nondestructive assay (NDA) system developed at Los Alamos National Laboratory (LANL) combining both neutron and gamma measurement capabilities. Since spent fuel assemblies are stored in water, the system was designed to be watertight to facilitate underwater measurements by inspectors. The AEFC is comprised of six 3He detectors as well as a shielded and collimated ion chamber. The 3He detectors are used for active and passive neutron coincidence counting while the ion chamber is used for gross gamma counting. Active coincidence measurement data is used to measure residual fissile mass, whereas the passive coincidence measurement data along with passive gamma measurement can provide information about burnup, cooling time, and initial enrichment. In the past, most of the active interrogation systems along with the AEFC used an AmLi neutron interrogation source. Owing to the difficulty in obtaining an AmLi source, a 252Cf spontaneous fission (SF) source was used during a 2014 field trail in Uzbekistan as an alternative. In this study, experiments were performed to calibrate the AEFC instrument and compare use of the 252Cf spontaneous fission source and the AmLi (α,n) neutron emission source. The 252Cf source spontaneously emits bursts of time-correlated prompt fission neutrons that thermalize in the water and induce fission in the fuel assembly. The induced fission (IF) neutrons are also time correlated resulting in more correlated neutron detections inside the 3He detector, which helps reduce the statistical errors in doubles when using the 252Cf interrogation source instead of

  7. Fission products in the spent nuclear fuel from czech nuclear power plants

    International Nuclear Information System (INIS)

    Lelek, V.; Mikisek, M.; Marek, T.

    1999-01-01

    The nuclear power is expected to become a supply able to cover a significant part of the world energetic demand in future. But its big disadvantage, the risk of the spent nuclear fuel, has to be solved. The aim of this paper is to make simple estimates of the upper limits of amounts of the most dangerous spent fuel components and their compounds produced in Czech Republic until 2040. Our estimates are independent on particular type reactor (only on its power) and so they can be carried out for any nuclear fuel cycle. (Authors)

  8. 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

  9. US/FRG umbrella agreement for cooperation in GCR Development. Fuel, fission products, and graphite subprogram. Quarterly status report, July 1, 1982-September 30, 1982

    International Nuclear Information System (INIS)

    Turner, R.F.

    1982-10-01

    This report describes the status of the cooperative work being performed in the Fuel, Fission Product, and Graphite Subprogram under the HTR-Implementing Agreement of the United States/Federal Republic of Germany Umbrella Agreement for Cooperation in GCR Development. The status is described relative to the commitments in the Subprogram Plan for Fuel, Fission Products, and Graphite, Revision 5, April 1982. The work described was performed during the period July 1, 1982 through September 30, 1982 in the HTGR Base Technology Program at Oak Ridge National Laboratory, the HTGR Fuel and Plant Technology Programs at General Atomic Company (GA), and the Project HTR-Brennstoffkreislauf of the Entwicklungsgemeinschaft HTR at KFA Julich, HRB Mannheim, HOBEG Hanau, and SIGRI Meitingen. The requirement for and format of this quarterly status report are specified in the HTR Implementing Agreement procedures for cooperation. Responsibility for preparation of the quarterly report alternates between GA and KFA

  10. 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)

  11. Fuel and fission product behaviour in early phases of a severe accident. Part I: Experimental results of the PHEBUS FPT2 test

    Energy Technology Data Exchange (ETDEWEB)

    Barrachin, M., E-mail: marc.barrachin@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Gavillet, D. [Paul Scherrer Institute, Würenlingen and Villigen, CH-5232 Villigen PSI (Switzerland); Dubourg, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); De Bremaecker, A. [Institute for Nuclear Materials Sciences, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2014-10-15

    One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO{sub 2} fuel test section and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 and 900 mm) of the 1-m long test section are presented in this paper. Material interactions leading to local corium formation were identified: firstly between fuel and Zircaloy-4 cladding, notably at 823 mm, where the cladding melting temperature was reached, and secondly between fuel and stainless steel oxides. Regarding fission products, molybdenum left so-called metallic precipitates mainly composed of ruthenium. Xenon and caesium behave similarly whereas barium and molybdenum often seems to be associated in precipitates.

  12. Non-electrical uses of thermal energy generated in the production of fissile fuel in fusion--fission reactors: a comparative economic parametric analysis for a hybrid with or without synthetic fuel production

    International Nuclear Information System (INIS)

    Tai, A.S.; Krakowski, R.A.

    1979-01-01

    A parametric analysis has been carried out for testing the sensitivity of the synfuel production cost in relation to crucial economic and technologic quantities (investment costs of hybrid and synfuel plant, energy multiplication of the fission blanket, recirculating power fraction of the fusion driver, etc.). In addition, a minimum synfuel selling price has been evaluated, from which the fission--fusion--synfuel complex brings about a higher economic benefit than does the fusion--fission hybrid entirely devoted to fissile-fuel and electricity generation. Assuming an electricity cost of 2.7 cents/kWh, an annual investment cost per power unit of 4.2 to 6 $/GJ (132 to 189 k$/MWty) for the fission--fusion complex and 1.5 to 3 $/GJ (47 to 95 k$/MWty) for the synfuel plant, the synfuel production net cost (i.e., revenue = cost) varies between 6.5 and 8.6 $/GJ. These costs can compete with those obtained by other processes (natural gas reforming, resid partial oxidation, coal gasification, nuclear fission, solar electrolysis, etc.). This study points out a potential use of the fusion--fission hybrid other than fissile-fuel and electricity generation

  13. Analysis of irradiated Magnox fuel for fission product and actinide elements

    International Nuclear Information System (INIS)

    Foster, E.; Fudge, A.J.; Glover, K.M.; Wiltshire, R.A.P.

    1980-03-01

    This report describes the methods used for, and the results obtained from, chemical and isotopic analysis of four Magnox fuel samples of known irradiation history. The analysis of this material was carried out in conjunction with CEGB, Berkeley Nuclear Laboratories and BNFL. The data obtained was used to evaluate the accuracy of fuel composition as predicted by computer codes such as RICE (Reactor Inventory Code). (author)

  14. Selective separation of long-lived fission products in nuclear spent fuel

    International Nuclear Information System (INIS)

    Ozawa, Masaki; Ikegami, Tetsuo; Asakura, Toshihide; Morita, Yasuji

    2003-01-01

    A catalytic electrolytic extraction (CEE) method has been studied to separate rare metal fission products (Pd, Ru, Rh, Tc, Te and Se). Some of them are long-lived radioactive, but are potentially strategic materials. In CEE process, Pd 2+ cation itself would not only be easily deposited from various nitric acid solutions but also enhance the deposition of RuNO 3+ and ReO 4 by acting as Pd adatom . Monte than 99% of the recovery ratio was attained for RuNo 3+ and Pd 2+ . Fundamental radio-electrochemical property of FPs/TRU in the synthetic and real dissolver solution is being investigated in the new JAERI-JNC collaboration frame. For the extended recycling of such FPs (including Cs/Sr), new precedent research programs are on-going involving universities. (author)

  15. 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

  16. Modelling the release of volatile fission product cesium from CANDU fuel under severe accident conditions using artificial neural networks

    International Nuclear Information System (INIS)

    Andrews, W.S.; Lewis, B.J.; Cox, D.S.

    1997-01-01

    An artificial neural network (ANN) model has been developed to predict the release of volatile fission products from CANDU fuel under severe accident conditions. The model was based on data for the release Of 134 Cs measured during three annealing experiments (Hot Cell Experiments 1 and 2, or HCE- 1, HCE-2 and Metallurgical Cell Experiment 1, or MCE- 1) at Chalk River Laboratories. These experiments were comprised of a total of 30 separate tests. The ANN established a correlation among 14 separate input variables and predicted the cumulative fractional release for a set of 386 data points drawn from 29 tests to a normalized error, E n , of 0.104 and an average absolute error, E abs , of 0.064. Predictions for a blind validation set (test HCE2-CM6) had an E n of 0.064 and an E abs of 0.054. A methodology is presented for deploying the ANN model by providing the connection weights. Finally, the performance of an ANN model was compared to a fuel oxidation model developed by Lewis et al. and to the U.S. Nuclear Regulatory Commission's CORSOR-M. (author)

  17. 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

  18. 1: the atom. 2: radioactivity. 3: man and radiations. 4: the energy. 5: nuclear energy: fusion and fission. 6: the operation of a nuclear reactor. 7: the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2002-01-01

    This series of 7 digest booklets present the bases of the nuclear physics and of the nuclear energy: 1 - the atom (structure of matter, chemical elements and isotopes, the four fundamental interactions, nuclear physics); 2 - radioactivity (definition, origins of radioelements, applications of radioactivity); 3 - man and radiations (radiations diversity, biological effects, radioprotection, examples of radiation applications); 4 - energy (energy states, different forms of energy, characteristics); 5 - nuclear energy: fusion and fission (nuclear energy release, thermonuclear fusion, nuclear fission and chain reaction); 6 - operation of a nuclear reactor (nuclear fission, reactor components, reactor types); 7 - nuclear fuel cycle (nuclear fuel preparation, fuel consumption, reprocessing, wastes management). (J.S.)

  19. Microstructural change and its influence on fission gas release in high burnup UO 2 fuel

    Science.gov (United States)

    Une, K.; Nogita, K.; Kashibe, S.; Imamura, M.

    1992-06-01

    The microstructural change of UO 2 fuel pellets (burnup: 6-83 GWd/t), base irradiated under LWR conditions, has been studied by detailed postirradiation examinations. The lattice parameter near the fuel rim in the irradiated UO 2 increased with burnup and appeared to become constant beyond about 50 GWd/t. This lattice dilation was mainly due to the accumulation of radiation induced point defects. Moreover, the dislocation density in the UO 2 matrix developed progressively with burnup, and eventually the tangled dislocations organized many sub-grain boundaries in the highest burnup fuel of 83 GWd/t. This sub-grain structure induced by accumulated radiation damage was compatible in appearance with SEM fractography results which revealed sub-divided grains of sub-micron size in as-fabricated grains. The influence of burnup on 85Kr release from the UO 2 fuels has been examined by means of a postirradiation annealing technique. The higher fractional release of high burnup fuels was mainly due to the burnup dependence of the fractional burst release evolved on temperature ramp. The fractional burst release was represented in terms of the square root of burnup from 6 to 83 GWd/t.

  20. TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey James [Univ. of California, Berkeley, CA (United States)

    2011-11-30

    This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated

  1. Non-fertile fuels for burning weapons plutonium in thermal fission reactors

    International Nuclear Information System (INIS)

    Lombardi, C.; Mazzola, A.; Vettraino, F.

    1996-01-01

    In the last few years, the excess plutonium disposition has become ever more a topical and critical issue. As a matter of fact, more than 200 MT of plutonium coming from spent fuel reprocessing have been already stockpiled and over the next decade, under the already ratified agreements, another about 200 MT of weapon-grade plutonium are expected to be available from nuclear weapons dismantlement. On this basis, an ever growing plutonium production is no longer the goal and the already stored quantities should be burnt in power reactors by taking care that no new plutonium is generated under irradiation. This new outlook in considering plutonium has led many designers to reassess the Fast Breeder Reactors (FBR) role and shifting from breeder to burner machines perspective. Several solutions for burning plutonium have been so far proposed and discussed from the safeguards, proliferation resistance, environmental safety, technological background, economy and time schedule standpoint. A proposal for plutonium burning in commercial Pressurized Water Reactors (PWR) by using a non-fertile oxide-type fuel consisting of PuO 2 diluted in an inert matrix is reported hereafter. This solution appears to receive an ever growing interest in the nuclear community. In order not to produce new plutonium during irradiation an innovative U-free fuel is being researched, based on an inert matrix which will consist in a mixed compound of inert oxides, such as ZrO 2 , Al2O 3 , MgO, CeO 2 where the plutonium oxide is dispersed in. The matrix will fulfill the following requirements: good chemical compatibility, acceptable thermal conductivity, good nuclear properties, good stability under irradiation, good dissolution resistance. The plutonium relative content will be comparable to that used in MOX fuel. The fuel is expected to be characterized by a high chemical stability (rock-like fuel), so that after discharge from reactor and adequate cooling time, it can be considered a High Level

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  3. Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

    International Nuclear Information System (INIS)

    Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

    1986-05-01

    In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant

  4. Irradiation tests of U3Si2-Al fuels up to very high fission densities

    International Nuclear Information System (INIS)

    Nuding, M.; Boening, K.

    2001-01-01

    The new research reactor of the Munich Technical University (TUM), the FRM-II, will have U 3 Si 2 -Al as the fuel. This fuel is considered qualified and optimally usable in the light of findings obtained in the RERTR program (Reduced Enrichment for Research and Test Reactors). The RERTR program was conducted to develop new fuel for the use of low enriched uranium (LEU) in research reactors. As the unique properties of the FRM-II in research and application are based also on achieving a very compact reactor core with highly enriched uranium (HEU), additional irradiation tests were performed on the basis of the RERTR program. They were run in close cooperation with the French Commissariat a l'Energie Atomique (CEA) in its SILOE and OSIRIS facilities, among others. After extensive evaluation, also of other studies, these tests confirm the RERTR findings about fuel swelling behavior and, consequently, the suitability of U 3 Si 2 -Al (HEU) for use in the compact core of the FRM-II. (orig.) [de

  5. Preconceptual design of a packed fluidized bed blanket for a fission suppressed thorium-fueled CTHR

    International Nuclear Information System (INIS)

    Chi, J.W.H.; Karbowski, J.S.; Chapin, D.L.

    1981-01-01

    This paper describes a thorium-fueled PFB blanket concept for a Commercial Tokamak Hybrid Reactor. A preliminary mechanical concept is presented and the results of neutronics, thermal-hydraulics and economics analyses are discussed. Futher work needed to design and advance the concept is recommended

  6. Concept and experimental studies on fuel and target for minor actinides and fission products transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Prunier, C; Guerin, Y [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d` Etudes des Combustibles; Salvatores, M [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires; Zaetta, A [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d` Etudes des Reacteurs

    1994-12-31

    High activity long-lived radionuclides in nuclear wastes, namely minor actinides (americium and neptunium) are in large amount generated by current nuclear reactive. The destruction of these radionuclides is a part of the French SPIN (Partitioning and Burning) program consistent with the determination to send a minimum amount of harmful products for final storage. Transmutation concepts are defined for neptunium and americium taking into account fuel cycle strategies. Neptunium destruction does not pose any major problems. It`s a by-product of uranium consumption, as plutonium and in despite of a slight gamma activity due to the protactinium 233 it`s quite easy to handle. Diluting neptunium in the mixed oxide fuels (MOX) should not be an obstacle for fabrication, in-pile behaviour and reprocessing either. Consequently we make the proposal of homogeneous mode of neptunium in MOX which should be soon explored in the experimental OSIRIS reactor and in the Phenix and Superphenix reactors. The analysis is more complex for the multi isotope americium. Its destruction is difficult because of gamma radioactivity which complicates fabrication. Experiments in Phenix and calculation showed that Phenix reactor offers a good potential for americium incineration, but similar data do not exist for PWR. It will remain a well known difficulty for fabrication and reprocessing. In this case we have to put a real new face to the fabrication flow-sheet of americium compounds and we propose to develop the heterogeneous mode. Targets choice are defined in term of: -safety, considering fuel reaction with cladding and water sodium, -transmutation rate, limited by target behaviour, in FR`s (Phenix), PWR`s (OSIRIS) and HFR (Petten), -reprocessing, checking the solubility of such targets by Purex process. So, at the beginning of our program the account has been on improving fuel and targets properties related to safety and fuel cycle. (authors). 4 figs.

  7. Prediction of fission product and actinide levels in irradiated fuel and cladding

    International Nuclear Information System (INIS)

    Burstall, R.F.; Thornton, D.E.J.

    1977-01-01

    The production of radioactive isotopes and their subsequent decay is of crucial importance in the nuclear industry, dominating the shield design of chemical reprocessing plants, transport flasks and waste disposal facilities which account for a large part of the capital investment in a nuclear programme. The isotopes are also important in studies of reactor shielding. The computation of the level and behavior of such nuclides has been practiced for many years in countries with nuclear industries, with ever-increasing sophistication in methods of calculation and in improving the accuracy of the basic nuclide data. Calculation is usually made for three groups of nuclides, the actinides or transuranics, the fission products, and nuclides present in the cladding. The currently accepted computer code within the UKAEA for such calculations is FISPIN. This code calculates activities for all the above groups either separately or in combination. As well as individual nuclide concentrations and activities integral information is produced. The paper describes the methods of calculation. The code has been compared with other codes which have a similar function, and it is concluded that the only significant differences are those associated with data. A number of different data sets, to a large degree independent, have been compared using the code, and the paper describes some of the results obtained

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-08-01

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

  9. 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)

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

    Science.gov (United States)

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

    2017-11-01

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

  11. 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)

  12. The potential of fission nuclear power in resolving global climate change under the constraints of nuclear fuel resources and once-through fuel cycles

    International Nuclear Information System (INIS)

    Knapp, Vladimir; Pevec, Dubravko; Matijevic, Mario

    2010-01-01

    Nuclear fission is receiving new attention as a developed source of carbon-free energy. A much larger number of nuclear reactors would be needed for a major impact on carbon emission. The crucial question is whether it can be done without increasing the risk of nuclear proliferation. Specifically, can a larger nuclear share in world energy production, well above the present 6%, be achieved in the next few decades without adding the proliferation-sensitive technologies of reprocessing spent fuel and recycling plutonium to the problems of the unavoidable use of enrichment technology? The answer depends on the available uranium resources. We first looked for the maximum possible nuclear build-up in the 2025-2065 period under the constraints of the estimated uranium resources and the use of once-through nuclear fuel technology. Our results show that nuclear energy without reprocessing could reduce carbon emission by 39.6% of the total reduction needed to bring the WEO 2009 Reference Scenario prediction of total GHG emissions in 2065 to the level of the WEO 450 Scenario limiting global temperature increase to 2 deg. C. The less demanding strategy of the nuclear replacement of all non-CCS coal power plants retiring during the 2025-2065 period would reduce emission by 26.1%.

  13. Delayed fission

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  14. Fuel cycles and fission products clean up for the MOSART concept

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V. V.; Feynberg, O. S.; Zakirov, R. Y. [Russian Research Center, Kurchatov Institute, Moscow (Russian Federation)

    2008-08-15

    Analysis of different methods for MOSART fuel salt clean up and experimental results produced within ISTC 1606 permits to make the following conclusions: {center_dot} Fuel salt clean up steps of Li,Be/F MOSART system could be based on the method of reductive extraction in liquid bismuth. {center_dot} Electrodeposition on solid indifferent electrodes (Mo, W) is not effective for separation of actinides and lanthanides in Li,Na,Be/F melt. However, preliminary experiments on electrodeposition on solid non indifferent electrodes (nickel) give foundation to suppose that An/Ln separation on these electrodes could be more effective. In order to determine characteristics of An/Ln electrochemical separation, it is necessary to conduct additional studies in Li,Na,Be/F melt on solid non indifferent electrodes.

  15. Release of fission products from irradiated SRP fuels at elevated temperatures: Data report on the second stage of the SRP source term study

    International Nuclear Information System (INIS)

    Woodley, R.E.

    1987-03-01

    The measurements of the release of fission products from irradiated Savannah River Plant (SRP) fuels at elevated temperatures reported herein extend the results of the first stage of the investigation to two additional fuel temperatures. In the first stage, two types of SRP fuels, a uranium-aluminum alloy designated MK-16 and a U 3 O 8 -aluminum cermet designated OX-2, were exposed to one of three different atmospheres, argon, air, or 80% steam-20% argon, at either of two different temperatures, 700 or 1100 0 C. In the second stage, the two fuels and three atmospheres remained the same, but the fuel temperatures, 850 and 1000 0 C, were intermediate to those previously employed. For each set of conditions, the measurements were repeated and, thus, the second stage of the study, like the first, consisted of 24 separate runs. This report presents the results of the 24 second-stage measurements

  16. Effects of solid fission products forming dissolved oxide (Nd) and metallic precipitate (Ru) on the thermal conductivity of uranium base oxide fuel

    International Nuclear Information System (INIS)

    Kim, Dong-Joo; Yang, Jae-Ho; Kim, Jong-Hun; Rhee, Young-Woo; Kang, Ki-Won; Kim, Keon-Sik; Song, Kun-Woo

    2007-01-01

    The effects of solid fission products on the thermal conductivity of uranium base oxide nuclear fuel were experimentally investigated. Neodymium (Nd) and ruthenium (Ru) were added to represent the physical states of solid fission products such as 'dissolved oxide' and 'metallic precipitate', respectively. Thermal conductivity was determined on the basis of the thermal diffusivity, density and specific heat values. The effects of the additives on the thermal conductivity were quantified in the form of the thermal resistivity equation - the reciprocal of the phonon conduction equation - which was determined from the measured data. It is concluded that the thermal conductivity of the irradiated nuclear fuel is affected by both the 'dissolved oxide' and the 'metallic precipitate', however, the effects are in the opposite direction and the 'dissolved oxide' influences the thermal conductivity more significantly than that of the 'metallic precipitate'

  17. Use of fluoride systems for some fission product separation from residues of fast reactor spent fuel fluorination

    International Nuclear Information System (INIS)

    Shishkov, Yu.D.; Khomyakov, V.I.

    1977-01-01

    Investigated has been a possibility of the use of fluoride systems (acid nitrozyl fluoride and molten salts) for americium extraction from residues of fluorination of irradiated fuel containing mainly fluorides of rare earth compounds, alkali and alkaline earth elements. At treatment of fission product fluorides by acid nitrozyl fluoride only cesium and uranium fluorides dissolve, while americium and rare earth fluorides are practically non-soluble in it. The solubility of cesium, strontium, barium and fluorides of some other rare earth elements in molten cryolite at the temperature of 1000 deg C, Li-NaF and LiF-CaF 2 of eutectic content at 750 and 800 deg C are respectively 15-77 %. Cerium fluoride presents an exception, its solubility in cryolite being only 0.73%. At treatment of mixture of americium and lanthanum fluorides by molten salts in the weight ratio of 1:1, approximately 50% of lanthanum and 65-70% of americium turn into melt independent of the type of melt. The maximum melt output of americium is obtained at treatment of lanthanum and americium fluoride mixture by cryolite melt at the temperature of 1000 deg C. It is shown that the presence of rare earth of fluorides, except lanthanum fluoride, effect significantly of americium distribution over phases in the process of fluoride processing by the fluoride molten salts

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-04-15

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

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  20. Heat transfer in a spent fuel pool concept containing PWR, Hybrid ADS-Fission, and VHTR spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Fernando P.; Cardoso, Fabiano; Salomé, Jean A.D.; Velasquez, Carlos E.; Pereira, Claubia, E-mail: fernandopereirabh@gmail.com, E-mail: fabinuclear@yahoo.com.br, E-mail: jadsalome@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Thermal evaluation under wet storage conditions of spent fuels (SF) of the types UO{sub 2} discharged from Pressurized Water Reactor (PWR) and Very High-temperature Reactor (VHTR), and (Th,TRU)O{sub 2} from Accelerator-Driven Subcritical Reactor System (ADS) and VHTR are presented. The analyzes are in the absence of an external cooling system of the pool, and the goal is to compare the water boiling time of the pool storing these different types of SF, at time t=0 year after reactor discharge. Two techniques were implemented. In the first one, all the materials of the fuel elements are considered. In the second, the SF is treated as holes inside the pool, assuming the heat transfer directly from the SF to the water. Results from first technique show that the boiling time (T{sub b}) ranged from 23 minutes for (Th,TRU)O{sub 2} from VHTR to 3 hours for UO{sub 2} from VHTR, while for the second technique, T{sub b} ranged from 10 minutes for (Th,TRU)O{sub 2} from VHTR to 2.7 hours for UO{sub 2} from VHTR. The discrepancies between Tb from both techniques reveal that the pathways considered for the heat transfer are crucial to the results. The thermal studies used the module CFX of the ANSYS Workbench 16.2 - student version. (author)

  1. Development of test apparatus for fission product release from overheated fuel element

    International Nuclear Information System (INIS)

    Takai, T.; Hirosawa, T.; Funabashi, H.; Miyahara, S.

    1996-01-01

    Evaluation of the source term released to environment under the accident conditions is important to the safety evaluation and design of reactor containment vessel. However, available data related to FBR source term are very limited, especially for the FPs release data from overheated FBR fuel. The present, source term evaluation of FBR is based on assumption from that of LWR. Though, this evaluation is very conservative. Evaluation large scale FBR source term using this method is result in extremely conservative and lead construction of large scale plant becomes doubtful from the viewpoints of cost and safety system. Though, it is necessary to evaluate source term from the realistic and rational scenario considering a characteristic of FBR. Preparation of FPs release experiment from irradiated fuel is going on to investigate the FPs release and transport and to develop the analysis code for in-vessel source term evaluation. Fabrication of this apparatus was started in 1992, and the installation was completed in 1994. This apparatus passed the facility inspection by Science and Technology Agency in March 1995. This apparatus consists of a high frequency induction furnace, thermal gradient tube (TGT), sintered metal filters, cold traps, gas-analyzer, γ-ray spectrometry system and so on. In the experiment, FPs release rate and behavior will be investigated using gamma-ray spectrogram and FP gas analysis. Physical and chemical composition of released FP would be investigated from FPs deposited profiles on TGT. Now, cold experiment using simulant FP materials are conducted. (author)

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

    International Nuclear Information System (INIS)

    Waterbury, G.R.

    1975-10-01

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

  3. Study of the Effect of (U0.8Pu0.2O2 Uranium–Plutonium Mixed Fuel Fission Products on a Living Organism

    Directory of Open Access Journals (Sweden)

    Ayagoz Baimukhanova

    2016-08-01

    Full Text Available The article describes the results of experiments conducted on pigs to determine the effect of plutonium, which is the most radiotoxic and highly active element in the range of mixed fuel (U0.8Pu0.2O2 fission products, on living organisms. The results will allow empirical prediction of the emergency plutonium radiation dose for various organs and tissues of humans in case of an accident in a reactor running on mixed fuel (U0.8Pu0.2O2.

  4. Effect of fission fragment on thermal conductivity via electrons with an energy about 0.5 MeV in fuel rod gap

    Directory of Open Access Journals (Sweden)

    F Golian

    2017-02-01

    Full Text Available The heat transfer process from pellet to coolant is one of the important issues in nuclear reactor. In this regard, the fuel to clad gap and its physical and chemical properties are effective factors on heat transfer in nuclear fuel rod discussion. So, the energy distribution function of electrons with an energy about 0.5 MeV in fuel rod gap in Busherhr’s VVER-1000 nuclear reactor was investigated in this paper. Also, the effect of fission fragments such as Krypton, Bromine, Xenon, Rubidium and Cesium on the electron energy distribution function as well as the heat conduction via electrons in the fuel rod gap have been studied. For this purpose, the Fokker- Planck equation governing the stochastic behavior of electrons in absorbing gap element has been applied in order to obtain the energy distribution function of electrons. This equation was solved via Runge-Kutta numerical method. On the other hand, the electron energy distribution function was determined by using Monte Carlo GEANT4 code. It was concluded that these fission fragments have virtually insignificant effect on energy distribution of electrons and therefore, on thermal conductivity via electrons in the fuel to clad gap. It is worth noting that this result is consistent with the results of other experiments. Also, it is shown that electron relaxation in gap leads to decrease in thermal conductivity via electrons

  5. Normalizing the maximum permissible seal failure of the fuel cladding of VVER and the activity of the fission products in the coolant

    International Nuclear Information System (INIS)

    Luzanova, L.M.; Miglo, V.N.; Slavyagin, P.D.

    1993-01-01

    In most countries developing a nuclear power industry based on pressurized water reactors, one of the conditions for issuing a license under normal operating conditions for issuing a license stipulates that the fuel elements may not lose their hermetic seal either under normal operating conditions or during presumable disturbances of the conditions of normal use. At a conference on radiation safety the ALARA principle was taken to be fundamental, it being attempted to keep the activity of the coolant of the primary circuit, including the fission products emerging from unsealed fuel elements, to a level as low as reasonably possible. As many years of experience in the nuclear power industry have shown, nuclear power stations are in many cases operated with nonhermetic fuel elements in the core. Therefore, from the point of view of safety and economy, the best way to operate a power plant is to try to ensure maximum burnup of the fuel of the unsealed elements as they operate within the limits of safe activity of the fission products in the fuel circuits

  6. 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

  7. Techniques and results of examination of fission product release from VVER fuel rods with artificial defects and a burnup of ∼60 MWd/kgU at the MIR loop facility

    International Nuclear Information System (INIS)

    Burukin, A.; Goryachev, A.; Ilyenko, S.; Izhutov, A.; Konyashov, V.; Shishin, V.; Shulimov, V.; Luzanova, L.; Miglo, V.

    2009-01-01

    Complex of equipment and several techniques for examination of radioactive fission product release from defective fuel rods were developed, prepared and tested at the PV-1 loop facility of the MIR reactor. During the first test, which was conducted at the PV-1 loop facility and aimed at testing of developed equipment and techniques, measurement of radioactive fission product release from an experimental re-fabricated fuel rod with a burnup of ∼60 MWd/kgU and an artificial defect was performed under design-basis steady-state operating conditions of the VVER-1000 reactor. PIE of all main parameters of the experimental defective fuel rod did not reveal any state peculiarities which could be caused by the artificial defect, i.e. fuel and cladding characteristics in the defect area did not differ from the initial ones (before testing) as well as their characteristics in areas distant from the defect; they are typical for fuel rods with a similar irradiation history in the VVER NPP. The gap in the experimental fuel rod was bridged due to close contact between fuel and cladding at increased fuel burnup; it can appreciable reduce release of radioactive fission products into the PV-1 primary coolant. This suggestion and quantitative characteristics of effect of gap bridging in a high-burnup fuel rod on radioactive fission product release should be investigated during the next tests performed at the PV-1 loop facility. Values of radioactive fission product release measured during the first test at the PV-1 loop facility in the MIR reactor will be used for development of an empirical engineering model in order to take into account high burnup effects and their impact on fission product release from fuel and defective fuel rods

  8. Chemistry of actinides and fission products in the nuclear-fuel cycle

    International Nuclear Information System (INIS)

    Anon.

    2004-01-01

    This colloquium was held under the auspices of the French and Russian Academies of Sciences, from 21 to 23 May 2003, at the 'Ecole nationale superieure de chimie de Paris' (ENSCP), under the cooperative framework agreed between the two Academies. Fifteen specialists from each country were brought together to present their results concerning research in their respective fields (industrial considerations, fundamental chemistry, the environment, new conditioning systems, hydro- and pyro-chemical separation techniques), situating the results in the general context of the two countries'common strategy for closing the nuclear fuel cycle and for the management of radioactive waste. The colloquium brought together 26 oral presentations, and three round table discussions (theoretical chemistry and modelling, the frontiers of research on the nuclear cycle, elemental characterisation). The speakers chosen represented a large section of the organisations involved in the research on these topics, from each country. This thematic issue of the Comptes Rendus Chimie presents some new insights into these topics and some original results. The colloquium was supported financially par the DRI of the French Academy des sciences, CNRS, IN2P3, CEA, Cogema, EDF, and ENSCP. (authors)

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

    Science.gov (United States)

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

    1967-01-01

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

  10. 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.

  11. Behavior of fission gases in nuclear fuel: XAS characterization of Kr in UO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Martin, P.M., E-mail: Philippe-m.martin@cea.fr [CEA, DEN, Cadarache DEC/SESC, F-13108 St-Paul-Lez-Durance Cedex (France); Vathonne, E.; Carlot, G.; Delorme, R.; Sabathier, C.; Freyss, M.; Garcia, P.; Bertolus, M. [CEA, DEN, Cadarache DEC/SESC, F-13108 St-Paul-Lez-Durance Cedex (France); Glatzel, P. [European Synchrotron Radiation Facility, 6 Rue Jules Horowitz, 38043 Grenoble (France); Proux, O. [OSUG, Observatoire des Sciences de l’Univers de Grenoble, CNRS and Université Joseph Fourier, BP 53, 38041 Grenoble Cedex 9 (France)

    2015-11-15

    X-ray Absorption Spectroscopy (XAS) was used to study the behavior of krypton as a function of its concentration in UO{sub 2} samples implanted with Kr ions. For a 0.5 at.% krypton local concentration, by combining XAS results and DFT + U calculations, we show that without any thermal treatment Kr atoms are mainly incorporated in the UO{sub 2} lattice as single atoms inside a neutral bound Schottky defect with O vacancies aligned along the (100) direction (BSD1). A thermal treatment at 1273 K induces the precipitation of dense Kr nano-aggregates, most probably solid at room temperature. In addition, 26 ± 2% of the Kr atoms remain inside BSD1 showing that Kr-BSD1 complex is stable up to this temperature. Consequently, the (in-)solubility of krypton in UO{sub 2} has to be re-evaluated. For high Kr concentration (8 at.%), XAS signals show that Kr atoms have precipitated in nanometer-sized aggregates with internal densities ranging between 4.15(7) g cm{sup −3} and 3.98(5) g cm{sup −3} even after annealing at 873 K. By neglecting the effect due to the UO{sub 2} matrix, the corresponding krypton pressures at 300 K were equal to 2.6(3) GPa and 2.0(2) GPa, respectively. After annealing at 1673 K, regardless of the initial Kr concentration, a bi-modal distribution is observed with solid nano-aggregates even at room temperature and larger cavities only partially filled with Kr. These results are very close to those observed in UO{sub 2} fuel irradiated in reactor. In this study we show that a rare gas can be used as a probe to investigate the defect creation and their stability in UO{sub 2}.

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

    Energy Technology Data Exchange (ETDEWEB)

    Dubourg, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Barrachin, M., E-mail: marc.barrachin@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Ducher, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Gavillet, D. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); De Bremaecker, A. [Institute for Nuclear Materials Sciences, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2014-10-15

    One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO{sub 2} fuel bundle and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 mm and 900 mm) of the test section previously reported are interpreted in the present paper. Solid state interactions between fuel and cladding have been compared with the characteristics of interaction identified in the previous separate-effect tests. Corium resulting from the interaction between fuel and cladding was formed. The uranium concentration in the corium is compared to analytical tests and a scenario for the corium formation is proposed. The analysis showed that, despite the rather low fuel burn up, the conditions of temperature and oxygen potential reached during the starvation phase are able to give an early very significant release fraction of caesium. A significant part (but not all) of the molybdenum was segregated at grain boundaries and trapped in metallic inclusions from which they were totally removed in the final part of the experiment. During the steam starvation phase, the conditions of oxygen potential were favourable for the formation of simple Ba and BaO chemical forms but the temperature was too low to provoke their volatility. This is one important difference with out-of-pile experiments such as VERCORS for which only a combination of high temperature and low oxygen potential induced a significant barium release. Finally another significant difference with analytical out-of-pile experiments comes from the formation of foamy zones due to the fission gas presence in FPT2-type experiments which give an additional possibility for the formation of stable fission product compounds.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  14. Structures and properties of (U,Pu)O2 containing non-active fission products. A simulation of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Schmitz, F.

    1969-01-01

    We have made oxides with the same uranium and plutonium content, the same stoichiometry and the same fission product content as an oxide fuel (U 0,8 PuO 2 )O 1,96 after 2 per cent burn up. We have calculated the stoichiometry changes due to irradiation and checked the calculation by X rays parameters measurements. We have calculated and measured the contraction of the oxide lattice due to fission products in solid solution. Microprobe analysis of precipitates have been made and have lead to the identification of non metallic barium containing compounds and have shown the particular behaviour of molybdenum. Some physical properties have been measured especially the electrical resistivity, the thermal diffusivity and the vapour pressure of zirconium in solid solution. (author) [fr

  15. Transuranics and fission products release from PWR fuels in severe accident conditions. Lessons learnt from VERCORS RT3 and RT4 tests

    International Nuclear Information System (INIS)

    Pontillon, Y.; Ducros, G.; Van Winckel, S.; Christiansen, B.; Kissane, M.P.; Dubourg, R.; Dutheillet, Y.; Andreo, F.

    2006-01-01

    Over the last decades, several experimental programs devoted to the source term of fission products (FP) and actinides released from PWR fuel samples in severe accident (SA) conditions have been initiated throughout the world. In France, in this context, the Institute for Radiological Protection and Safety (IRSN) and Electricite de France (EDF) have supported the analytical VERCORS program which was performed by the Commissariat a l'Energie Atomique (CEA). The VERCORS facility at the LAMA-laboratory (CEA-Grenoble, France) was designed to heat up an irradiated fuel sample - taken from EDF's nuclear power reactors - to fuel relocation, and to capture the fission products released from the fuel and deposited downstream on a series of specific filters (impactors, bead-bed filter). On-line gamma detectors aimed at the fuel position, filters and gas capacity monitored the progress of FP release from the fuel, FP deposition on the filters and the fission gases emitted by the fuel (xenon and krypton). Before and after the test, a longitudinal gamma-scan of the fuel was conducted to measure the initial and final FP inventory in order to evaluate the quantitative fractions of FP emitted by the fuel during the test. All the components of the loop were then gamma-scanned to measure and locate the FPs released during the test and to draw up a mass balance of these FP. 25 annealing tests were performed between 1983 and 2002 on irradiated PWR fuels under various conditions of temperature and atmospheres (oxidising or reducing conditions). The influence of the nature of the fuel (UO 2 versus MOX, burn up) and the fuel morphology (initially intact or fragmented fuel) have also been investigated. This led to an extended data base allowing on the one hand to study mechanisms which promote FP release in SA conditions, and on the other hand to enhance models implemented in SA codes. Because gamma spectrometry is well suited to FP measurement and not to actinides (except neptunium

  16. Modeling of fuel performance and fission product release behavior during HTTR normal operation. A comparative study of the FZJ and JAERI modeling approach

    International Nuclear Information System (INIS)

    Verfondern, Karl; Sumita, Junya; Ueta, Shohei; Sawa, Kazuhiro

    2001-03-01

    For the prediction of fuel performance and fission product release behavior in the High Temperature Engineering Test Reactor, HTTR of the Japan Atomic Energy Research Institute(JAERI), during its normal operation, calculation tools were applied as have been used at the Research Center Juelich (FZJ) in safety analyses for pebble-bed HTGR designs. Calculations were made assuming the HTTR operation with a nominal operation time of 660 efpd including a 110 efpd period with elevated fuel temperatures. Fuel performance calculations by the PANAMA code with given fuel temperature distribution in the core have shown that the additional failure level of about 5x10 -6 is expected which is about twice as much as the as-fabricated through-coatings failure level. Under the extreme safety design conditions, the predicted particle failure fraction in the core increases to about 1x10 -3 in maximum. The diffusive release of metallic fission products from the fuel primarily occurs in the core layer with the maximum fuel temperature (layer 3) whereas there is hardly any contribution from layer 1 except for the recoil fraction. Silver most easily escapes the fuel; the predicted release fractions from the fuel compacts are 10% (expected) and 50% (safety design). The figures for strontium (expected: 1.5x10 -3 ), safety design: 3.1x10 -2 ) and cesium (5.6x10 -4 , 2.9x10 -2 ) reveal as well a significant fraction to originate already from intact particles. Comparison with the calculation based on JAERI's diffusion model for cesium shows a good agreement for the release behavior from the particles. The differences in the results can be explained mainly by the different diffusion coefficients applied. The release into the coolant can not modelled because of the influence of the gap between compact and graphite sleeve lowering the release by a factor of 3 to 10. For the prediction of performance and fission product release behavior of advanced ZrC TRISO particles, more experimental work is

  17. Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

    Science.gov (United States)

    Chang, G. S.; Lillo, M. A.

    2009-08-01

    The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y

  18. Nuclear fission

    International Nuclear Information System (INIS)

    Kodama, T.

    1981-01-01

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

  19. 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

  20. 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.

  1. Amount, disposal and relative toxicity of long-lived fission products and actinides in the radioactive wastes of the nuclear fuel cycles

    International Nuclear Information System (INIS)

    Haug, H.O.

    1975-11-01

    A review is presented on the magnitude of the long-term problems of radioactive wastes from the nuclear power industry of the FRG (and Western Europe). The production of long-lived fission products and actinides has been calculated for several fuel types of the uranium-plutonium and thorium-uranium fuel cycles and related to a prediction of the development and share of LWR, FBR and HTGR. The quantities and concentrations of actinides, the radioactivity and relative toxicity index of the wastes of reprocessing (and fuel refabrication) and their changes by radioactive decay are presented. The radiotoxicity of the nuclide inventory of the solidified high-level wastes have been compared with naturally occuring uranium ores. On the long term (>10 3 years) the radiotoxicity level of the total area of the final repository in deep geological formation does not result in a significantly higher radiotoxicity level than an uranium ore deposit of low uranium content. Also discussed have been the chemical separation of the actinides from high-level wastes and recycling in fission reactors. (orig.) [de

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

    International Nuclear Information System (INIS)

    Krohn, H.; Finken, R.

    1983-06-01

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

  3. 1: the atom. 2: radioactivity. 3: man and radiations. 4: the energy. 5: nuclear energy: fusion and fission. 6: the operation of a nuclear reactor. 7: the nuclear fuel cycle; 1: l'atome. 2: la radioactivite. 3: l'homme et les rayonnements. 4: l'energie. 5: l'energie nucleaire: fusion et fission. 6: le fonctionnement d'un reacteur nucleaire. 7: le cycle du combustible nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    This series of 7 digest booklets present the bases of the nuclear physics and of the nuclear energy: 1 - the atom (structure of matter, chemical elements and isotopes, the four fundamental interactions, nuclear physics); 2 - radioactivity (definition, origins of radioelements, applications of radioactivity); 3 - man and radiations (radiations diversity, biological effects, radioprotection, examples of radiation applications); 4 - energy (energy states, different forms of energy, characteristics); 5 - nuclear energy: fusion and fission (nuclear energy release, thermonuclear fusion, nuclear fission and chain reaction); 6 - operation of a nuclear reactor (nuclear fission, reactor components, reactor types); 7 - nuclear fuel cycle (nuclear fuel preparation, fuel consumption, reprocessing, wastes management). (J.S.)

  4. 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

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

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

    International Nuclear Information System (INIS)

    Mogensen, M.; Walker, C.T.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Geiger, Ernesto

    2016-01-01

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

  8. Nuclear Forensics and Radiochemistry: Fission

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-07

    Radiochemistry has been used to study fission since it’ discovery. Radiochemical methods are used to determine cumulative mass yields. These measurements have led to the two-mode fission hypothesis to model the neutron energy dependence of fission product yields. Fission product yields can be used for the nuclear forensics of nuclear explosions. The mass yield curve depends on both the fuel and the neutron spectrum of a device. Recent studies have shown that the nuclear structure of the compound nucleus can affect the mass yield distribution.

  9. 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)

  10. 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.

  11. 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

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

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujita, Misao

    1984-01-01

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

  13. Fission-product source terms

    International Nuclear Information System (INIS)

    Lorenz, R.A.

    1981-01-01

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

  14. Radiochemistry and the Study of Fission

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-14

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

  15. Radiochemistry and the Study of Fission

    International Nuclear Information System (INIS)

    Rundberg, Robert S.

    2016-01-01

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

  16. On the Potential of Nuclear Fission Energy for Effective Reduction of Carbon Emission under the Constraint of Uranium Resources Use without Spent Fuel Reprocessing

    International Nuclear Information System (INIS)

    Knapp, V.; Pevec, D.; Matijevic, M.

    2010-01-01

    Urgency to stop further increase of greenhouse gases emissions and reverse the trends, as stated in the Fourth Intergovernmental Panel on Climate Change (IPPC) Report and in Copenhagen discussions, limits the realistic choice of energy technologies to those available now or in the near future of few decades. In the coming fifty years neither nuclear fusion nor carbon capture and storage (CCS) can be expected to give a significant contribution to world energy production. Two perspective intermittent sources such as wind and solar together with nuclear fission energy covering the base load consumption appears to be a combination with a potential to produce a large share of carbon free energy in the total world energy production. This contribution considers the issues, associated with required large scale deployment of nuclear fission energy. A serious question associated with nuclear energy is nuclear proliferation. Spread of uranium enrichment and spent fuel reprocessing installations in many new countries constructing nuclear reactors would be a major concern in present political environment. We investigate whether uranium resources would be sufficient to support nuclear build-up in next 50-60 years sufficiently large to significantly reduce carbon emission without reprocessing of spent nuclear fuel. A positive answer would mean that 50-60 years can be available to develop effective international control of nuclear fuel cycle installations. Our results show that a maximum nuclear build-up which would consume currently estimated uranium resources by 2065 without reprocessing could reduce by 2065 carbon emission by 39.6% of the total reduction needed to bring the WEO 2009 Reference Scenario prediction of total GHG emissions in 2065 to the level of the WEO 450 Scenario limiting global temperature increase to 2 degrees of C. The less demanding strategy of the nuclear replacement of all non-CCS coal power plants retiring during the 2025-2065 period would reduce emission

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

    International Nuclear Information System (INIS)

    Vathonne, Emerson

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-09-01

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

  19. Studying the fuel burnup of MNSR reactor and estimating the concentrations of main fission products using the codes WIMS-D4 and CITATION

    International Nuclear Information System (INIS)

    Haj Hassan, H.; Ghazi, N.; Hainoun, A.

    2007-01-01

    The codes WIMSD-4 and BORGLES - part of the MTR-PC code package- have been applied to prepare the microscopic cross section library for the main elements of MNSR core for 6 neutron energy groups. The generated library was utilized from the 3D code CITATION to perform the calculation of fuel burn up and depletion including the identification of main fission products and its effects on the multiplication factor. In this regard some modifications have been introduced to the subroutine NUCY in CITATION to incorporate estimating the concentration of the related actinides and fission products. The burn up results indicated that the core life time of MNSR is being mainly estimated by Sm-149 following by Gd-157 and Cd-113. The accumulation of these actinides during 100 continuous operation days caused a reduction of ca. 2 mk for the excess reactivity. This result seems to be in good agreement with the available empirical value of 1.8 mk which relates to the whole discontinuous operation period of the reactor since its start and up to now. The calculation procedure simulates the sporadic operation with an adequate continuous operation period. This approximation is valid for the long lived actinides that mainly dictate the core life time. However, it is an overestimation for the concentration of short lived radioactive products like Xe-135. In the framework of this analysis the possibility of replacement of current MNSR fuel through low enriched fuels has been explored for two the fuel types U02-Mg and U3Si-Al. The results indicate that the first type (UO2-Mg) realize the criticality conditions with low enrichment of 20%, whereas the second type (U3Si-Al) required increasing the uranium enrichment up to 33%. For both fuel types the contribution of plutonium isotopes on the criticality has been also evaluated. Additionally, the influence of mixing burnable absorbers (Gd-113, Cd- 113) with the fresh fuels was investigated to identify their long-term control effect on the

  20. QUEIMAP: a computer routine for punctual analysis of a nuclear fuel depletion with accumulation of fission fragments

    International Nuclear Information System (INIS)

    Couto, R.T.

    1990-01-01

    QUEIMAP is a computer routine for burnup calculation, composed of five FORTRAN-77 subroutines. Its objective is to solve depletion equation of four radionuclides conversion chain, U238, U235, Th232, as well as fission fragments equations. In this paper the burnup is considered punctual and evolutioned under cross section. It presents the solution algorithms employed by QUEIMAP, the validation of its results and the way of use it. (M.I.)

  1. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

    1976-04-01

    Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

  2. Irradiation performance of HTGR fertile fuel in HFIR target capsules HT-12 through HT-15. Part I. Experiment description and fission product behavior

    International Nuclear Information System (INIS)

    Kania, M.J.; Lindemer, T.B.; Morgan, M.T.; Robbins, J.M.

    1977-02-01

    Sixteen types of Biso-coated designs, on ThO 2 kernels, were irradiated in High Flux Isotope Reactor target capsules HT-12 through HT-15. The report addresses the description of the experiment and extensive postirradiation analyses and experiments to determine fertile-particle burnup, fuel coating failures, and fission product behavior. Several low-temperature isotropic (LTI) pyrocarbon coatings, which ''survived'' according to visual inspection, were shown to have developed permeability during irradiation. These particles were irradiated at temperatures approximately equal to 1250 0 C and to burnups equal to or greater than 8 percent fission per initial heavy-metal atom (FIMA). No evidence of permeability was found in similar particles irradiated at temperatures approximately equal to 1550 0 C and burnups approximately equal to 16 percent FIMA. Failures due to permeability were not detectable by visual inspection but required a more extensive investigation by the 1000 0 C gaseous chlorine leaching technique. Maximum particle surface operating temperatures were found to be approximately 300 0 C in excess of design limits of 900 0 C (low-temperature magazines) and 1250 0 C (high-temperature magazines). The extremes of high temperatures and fast neutron fluences up to 1.6 x 10 22 neutrons/cm 2 produced severe degradation and swelling of the Poco graphite magazines and sample holders

  3. 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…

  4. Use of fission track analysis technique for the determination of MicroBequerel level of {sup 239}Pu in urine samples from radiation workers handling MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yadav, J.R., E-mail: yadav_jogendra@rediffmail.co [Health Physics Laboratory, Health Physics Division, BARC, Tarapur 401502 (India); Rao, D.D.; Kumar, Ranjeet [Health Physics Laboratory, Health Physics Division, BARC, Tarapur 401502 (India); Aggarwal, S.K. [Fuel chemistry Division, BARC, Trombay, Mumbai 400085 (India)

    2011-07-15

    Fission track analysis (FTA) technique for the determination of {sup 239}Pu excreted through urine has been standardized using blank samples, tracer and {sup 239}Pu spikes. Double stage anion exchange separation protocol has been applied and an average radiochemical recovery of {sup 239}Pu of 18% was obtained. An average track registration efficiency of 11 tracks per {mu}Bq of {sup 239}Pu, irradiated to 0.35x10{sup 17} neutron fluence was established. Reagent blank urine samples from 11 controlled subjects were analyzed by FTA and an average of 149{+-}14 tracks was obtained. Minimum detectable activity of 34 {mu}Bq L{sup -1} of urine sample was obtained and will be useful for monitoring chronic exposure cases handling MOX fuel.

  5. Contribution to the study on the contamination of the primary coolant of pressurized water reactors by fission products. Application for evaluation of failed fuel elements

    International Nuclear Information System (INIS)

    Bourgeois, Pierre.

    1979-01-01

    After describing the radioactive substances present in the primary system, the consequences of this contamination are studied. The many publications relating to the emergence of fission products from the fuel pins are then analyzed. This study reaches the conclusion that none of the proposed methods makes it possible to interpret the activity determinations made in power reactors. This is why the opposite approach is adopted consisting in suggesting a model as from in-reactor measurements. This model will apply in the conditions for which it has been adjusted, i.e. at a steady nominal power in a 900 MWe pressurized water reactor. The originality of this model lies in the determination of a probable emergence of the can-fuel set which reflects the state of the can and the demonstration of a trapping of the iodine in the set, the consequence of which is a lower release than that of the rare gases in the primary system. After comparing the experimental results available in the literature with the results of the model, it is shown how to proceed in order to evaluate the number of failed fuel elements. Finally, on the basis of the experiments performed at the Grenoble Nuclear Study Centre, the mean and maximal predictable activites are calculated for load following working. The study of the transients is completed by a more accurate calculation of the iodine releases when the reactor is shut down which confirms the main hypotheses of the steady state model [fr

  6. 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)

  7. 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

  8. 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

  9. Thermochromatographic investigations of fission product transport and chemistry

    International Nuclear Information System (INIS)

    Growcock, F.B.; Aronson, S.; Friedlander, M.; Skalyo, J. Jr.; Hosseini, A.; Taylor, R.D.

    1978-01-01

    A thermochromatographic technique has been developed to investigate the chemical states of fission products from irradiated fuel as well as in fission product simulation studies. Some recent work on iodine transport and on release of fission products from irradiated fuel kernels will be discussed

  10. Evaluation of the heat transfer in a geological repository concept containing PWR, VHTR and hybrid ads-fission spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jonusan, Raoni A.S.; Pereira, Fernando; Velasquez, Carlos E.; Salome, Jean A.D.; Cardoso, Fabiano; Pereira, Claubia; Fortini, Angela, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    The investigation of the thermal behavior of spent fuel (SF) materials is essential to determining appropriate potential sites to accommodate geological repositories as well as the design of canisters, considering their potential risk to people health and of environmental contamination. This work presents studies of the temperature in a canister containing spent fuels discharged from Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) reactor systems in a geological repository concept. The thermal analyses were performed with the software ANSYS, which is widely used to solve engineering problems through the Finite Element Method. The ANSYS Transient Thermal module was used. The spent nuclear fuels were set as heat sources using data of previous studies derived from decay heat curves. The studies were based on comparison of the mean temperature on a canister surface along the time under geological disposal conditions, for a same amount of each type of spent nuclear fuel evaluated. The results conclude that fuels from VHTR and ADS systems are inappropriate to be disposed in a standardized PWR canister, demanding new studies to determine the optimal amount of spent fuel and new internal canister geometries. It is also possible to conclude that the hypothetical situation of a single type of canister being used to accommodate different types of spent nuclear fuels is not technically feasible. (author)

  11. Extraction chromatography of fission products

    International Nuclear Information System (INIS)

    Bonnevie-Svendsen, M.; Goon, K.

    1978-01-01

    Various cases of using extraction chromatography during analysis of fission products are reviewed. The use of the extraction chromatography method is considered while analysing reprocessed products of nuclear fuel for quantitative radiochemical analysis and control of fission product and actinoide separation during extraction and their chemical state in production solutions. The method is used to obtain pure fractions of typical burnup monitors (neodymium, molybdenum, cerium, cesium, europium, lanthanides) during determination of nuclear fuel burnup degree. While studying the nature of nuclear reactions the method is used to separate quickly short-life isotopes, to purify β-radiator fractions before measuring their half-life periods, to enrich isotopes forming with low output during fission. Examples of using extraction chromatography are given to separate long half-life or stable fission products from spent solutions, to control environment object contamination

  12. ORIGEN-S: scale system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    ORIGEN-S computes time-dependent concentrations and source terms of a large number of isotopes, which are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay, input feet rates and physical or chemical removal rates. The calculations may pertain to fuel irradiation within nuclear reactors, or the storage, management, transportation or subsequent chemical processing of removed fuel elements. The matrix exponential expansion model of the ORIGIN code is unaltered in ORIGEN-S. Essentially all features of ORIGEN were retained, expanded or supplemented within new computations. The primary objective of ORIGEN-S, as requested by the Nuclear Regulatory Commission, is that the calculations may utilize the multi-energy group cross sections from any currently processed standardized ENDF/B data base. This purpose has been implemented through the prior execution of codes within either the SCALE System or the AMPX System, developed at the Oak Ridge National Laboratory. These codes compute flux-weighted cross sections, simulating conditions within any given reactor fuel assembly, and convert the data into a library that can be input to ORIGEN-S. Time-dependent libraries may be produced, reflecting fuel composition variations during irradiation. Presented in the document are: detailed and condensed input instructions, model theory, features available, range of applicability, brief subroutine descriptions, sample input, and I/O requirements. Presently the code is operable on IBM 360/370 computers and may be converted for CDC computers. ORIGEN-S is a functional module in the SCALE System and will be one of the modules invoked in the SAS2 Control Module, presently being developed, or may be applied as a stand alone program. It can be used in nuclear reactor and processing plant design studies, radiation safety analyses, and environmental assessments

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

    International Nuclear Information System (INIS)

    Walker, C.T.; Mogensen, M.

    1987-01-01

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

  14. Parametric study of fission-induced U-Mo fuel creep and structural analysis of fuel plates in view of implications for microstructure evolution

    International Nuclear Information System (INIS)

    Kim, Y.S.; Hofman, G.L.; Choo, Y.S.; Robinson, A.B.

    2010-01-01

    U-Mo fuel deformation during irradiation in U-Mo/Al dispersion plates is investigated by using the irradiation data from the RERTR-3 through -9 tests. The observation of fuel particle sintering during irradiation is also presented and its influence for fuel performance is discussed. Structural analysis was also performed to examine the relationship between the stress distribution in the plate and the location of matrix-pore formation in the plate. (author)

  15. Sequential separation of transuranic elements and fission products from uranium metal ingots in electrolytic reduction process of spent PWR fuels

    International Nuclear Information System (INIS)

    Chang Heon Lee; Kih Soo Joe; Won Ho Kim; Euo Chang Jung; Kwang Yong Jee

    2009-01-01

    A sequential separation procedure has been developed for the determination of transuranic elements and fission products in uranium metal ingot samples from an electrolytic reduction process for a metallization of uranium dioxide to uranium metal in a medium of LiCl-Li 2 O molten salt at 650 deg C. Pu, Np and U were separated using anion-exchange and tri-n-butylphosphate (TBP) extraction chromatography. Cs, Sr, Ba, Ce, Pr, Nd, Sm, Eu, Gd, Zr and Mo were separated in several groups from Am and Cm using TBP and di(2-ethylhexyl)phosphoric acid (HDEHP) extraction chromatography. Effect of Fe, Ni, Cr and Mg, which were corrosion products formed through the process, on the separation of the analytes was investigated in detail. The validity of the separation procedure was evaluated by measuring the recovery of the stable metals and 239 Pu, 237 Np, 241 Am and 244 Cm added to a synthetic uranium metal ingot dissolved solution. (author)

  16. Process for the extraction of fission products

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  17. The fission products palladium and rhodium: Their state in solutions, their behavior in the regeneration of fuel of atomic power stations, and the search for selective extraction techniques

    International Nuclear Information System (INIS)

    Arseenkov, L.V.; Zakharkin, B.S.; Lunichkina, K.P.; Renard, E.V.; Rogozhkin, V.Yu.; Shorokhov, N.A.

    1992-01-01

    At the present time many research centers are working on the extraction of noble metals in the form of fission fragments. Consistent data has been obtained on the mass accumulation of noble metals in various forms of processed nuclear fuel. Requirements are noted that must be met for obtaining industrial and economic efficiency in the extraction of noble metals by the Purex process. Presently there is a lack of information on the extraction of noble metals from spent fuel, particularly as far as the nitric acid media of the Purex process are concerned. The authors will discuss individual test observations on simulating systems and real systems with noble metals. The investigations focused on the noble metals of lowest radioactivity, namely palladium and rhodium. The complexity of the chemistry of ruthenium, on the one hand, and the possible selective, clearing distillation of ruthenium tetroxide from nitric acid solutions, on the other hand, make it necessary to focus the attention on the unresolved problems of the extraction of palladium and rhodium. The article further includes discussion on the following topics: noble metals in solutions of purex process, electrochemical operations involving noble metals, extraction systems for rhodium and palladium, separation of palladium from real solutions

  18. Effect of solid fission products forming dissolved oxide(Nd) and metallic precipitate(Ru) on the thermophysical properties of MOX fuel

    International Nuclear Information System (INIS)

    Kim, Dong Joo

    2006-02-01

    This study experimentally investigated the effect of solid fission products on the thermophysical properties of the mixed oxide fuel and evaluated them on the basis of the analytical theory. Neodymium and ruthenium were selected for the experiments to represent the physical states of the solid fission product as a 'dissolved oxide' and 'metallic precipitate', respectively. The state of the additives, crystal structures, lattice parameters, and theoretical densities were investigated with X-ray diffraction (XRD). Thermal diffusivities and thermal expansion rates were measured with laser flash method and dilatometry, respectively. The thermal expansion data were then fitted to obtain an correlation equation of the density variation as a function of the temperature. The specific heat capacity values were determined using the Neumann-Kopp's rule. The thermal expansion of the 'Nd.added' sample linearly increased with the concentration of the neodymium, which is primarily due to the fact that the melting point of Nd 2 O 3 is lower than that of UO 2 . On the other hand, the thermal expansion of the 'Ru.added' sample hardly changed with increasing ruthenium content. Thermal conductivities of the simulated MOX fuel were determined on the basis of the thermal diffusivities, density variation, and specific heat values measured in this study. The effect of additives on the thermal conductivity of the samples was quantified in the form of the thermal resistance equation, the reciprocal of the phonon conduction equation, which was determined from measured data. For 'dissolved oxide' sample in the UO 2 matrix, the effect is mainly attributed to the increase of lattice point defects caused by U 4+ , Ce 4+ , Nd 3+ and O 2- ions, which play the role of phonon scattering centers, that is, mean free path of phonon scattering decreases with the point defects, thus increase the thermal resistance. Also, the mass difference between the host (U) and the substituted atom (Ce and/or Nd) can

  19. The behavior of fission products during nuclear rocket reactor tests

    International Nuclear Information System (INIS)

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

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

    Energy Technology Data Exchange (ETDEWEB)

    Weissfloch, R

    1973-07-15

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

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

    International Nuclear Information System (INIS)

    Weissfloch, R.

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

  2. Ternary fission

    Indian Academy of Sciences (India)

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

  3. 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)

  4. 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

  5. Description and exploitation of benchmarks involving {sup 149}Sm, a fission product taking part of the burn up credit in spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Anno, J.; Poullot, G. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire; Fouillaud, P.; Grivot, P. [CEA Centre d`Etudes de Valduc, 21 - Is-sur-Tille (France)

    1995-12-31

    Up to now, there was no benchmark to validate the Fission Products (FPs) cross sections in criticality safety calculations. The protection and nuclear safety institute (IPSN) has begun an experimental program on 6 FPs ({sup 103}Rh, {sup 133}Cs, {sup 143}Nd, {sup 149}Sm, {sup 152}Sm, and {sup 155}Gd daughter of {sup 155}Eu) giving alone a decrease of reactivity equal to half the whole FPs in spent fuels (except Xe and I). Here are presented the experiments with the {sup 149}Sm and the results obtained with the APOLLO I-MORET III calculations codes. 11 experiments are carried out in a zircaloy tank of 3.5 1 containing slightly nitric acid solutions of Samarium (96,9% in weight of {sup 149S}m) at 0.1048 -0.2148 - 0.6262 g/l concentrations. It was placed in the middle of arrays of UO{sub 2} rods (4.742 % U5 weight %) at square pitch of 13 mm. The underwater height of the rods is the critical parameter. In addition, 7 experiments were performed with the same apparatus with water and boron proving a good experimental representativeness and a good accuracy of the calculations. As the reactivity worth of the Sm tank is between 2000 and 6000 10{sup -5}, the benchmarks are well representative and the cumulative absorption ratios show that {sup 149}Sm is well qualified under 1 eV. (authors). 8 refs., 7 figs., 6 tabs.

  6. Description and exploitation of benchmarks involving 149Sm, a fission product taking part of the burn up credit in spent fuels

    International Nuclear Information System (INIS)

    Anno, J.; Poullot, G.

    1995-01-01

    Up to now, there was no benchmark to validate the Fission Products (FPs) cross sections in criticality safety calculations. The protection and nuclear safety institute (IPSN) has begun an experimental program on 6 FPs ( 103 Rh, 133 Cs, 143 Nd, 149 Sm, 152 Sm, and 155 Gd daughter of 155 Eu) giving alone a decrease of reactivity equal to half the whole FPs in spent fuels (except Xe and I). Here are presented the experiments with the 149 Sm and the results obtained with the APOLLO I-MORET III calculations codes. 11 experiments are carried out in a zircaloy tank of 3.5 1 containing slightly nitric acid solutions of Samarium (96,9% in weight of 149S m) at 0.1048 -0.2148 - 0.6262 g/l concentrations. It was placed in the middle of arrays of UO 2 rods (4.742 % U5 weight %) at square pitch of 13 mm. The underwater height of the rods is the critical parameter. In addition, 7 experiments were performed with the same apparatus with water and boron proving a good experimental representativeness and a good accuracy of the calculations. As the reactivity worth of the Sm tank is between 2000 and 6000 10 -5 , the benchmarks are well representative and the cumulative absorption ratios show that 149 Sm is well qualified under 1 eV. (authors). 8 refs., 7 figs., 6 tabs

  7. Study on uranium metallization yield of spent Pressurized Water Reactor fuels and oxidation behavior of fission products in uranium metals

    International Nuclear Information System (INIS)

    Choi, Ke Chon; Lee, Chang Heon; Kim, Won Ho

    2003-01-01

    Metallization yield of uranium oxide to uranium metal from lithium reduction process of spent Pressurized Water Reactor (PWR) fuels was measured using thermogravimetric analyzer. A reduced metal produced in the process was divided into a solid and a powder part, and each metallization yield was measured. Metallization yield of the solid part was 90.7∼95.9 wt%, and the powder being 77.8∼71.5 wt% individually. Oxidation behaviour of the quarternary alloy was investigated to take data on the thermal oxidation stability necessary for the study on dry storage of the reduced metal. At 600∼700 .deg. C, weight increments of allow of No, Ru, Rh and Pd was 0.40∼0.55 wt%. Phase change on the surface of the allow was started at 750 .deg. C. In particular, Mo was rapidly oxidized and then the alloy lost 0.76∼25.22 wt% in weight

  8. Fission meter

    Science.gov (United States)

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

    2012-04-10

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

  9. Quantitative analysis of the fission product distribution in a damaged fuel assembly using gamma-spectrometry and computed tomography for the Phébus FPT3 test

    International Nuclear Information System (INIS)

    Biard, B.

    2013-01-01

    Highlights: • The FP quantitative distribution in the fuel bundle is measured by gamma-spectrometry. • The FP location is obtained with emission tomograms and other experiment results. • X-ray tomograms provide the material and density mapping of the degraded bundle. • The self-attenuation may then be computed for each isotope at its key line energy. • Results are consistent with other FPT3 measurements, with acceptable uncertainties. -- Abstract: The international Phébus FP programme, initiated in 1988 by the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN), in cooperation with the European Commission (EC) and with financial support from USNRC, Canada, PSI/HSK (Switzerland), Japan and Korea, was aimed at studying severe accident phenomena: the fuel degradation, the release of fission products (FPs) and their transport through the reactor coolant system to the containment building. The FPT3 test, conducted in 2004, was the last of the five light water reactor core meltdown accident tests performed on irradiated fuel rods. After the experiment, the test device was recovered and analysed through a full set of non-destructive examinations performed over the fuel bundle zone, including gamma-scanning, gamma emission tomography, X-ray radiography and X-ray transmission tomography. The gamma-scanning was specifically devoted to the location, identification and amount quantification of the FPs remaining in the bundle. Since the fuel bundle became highly degraded during the experiment, the geometry was different at each level examined, and did not correspond to the well-known initial state. The self-attenuation of the test device and consequently the efficiency correction could then not be estimated by classical means that need to know the geometry of the object. Using the results of the other non-destructive examinations, specific computational tools and methods have therefore been developed to compute the self-attenuation of the bundle

  10. Quantitative analysis of the fission product distribution in a damaged fuel assembly using gamma-spectrometry and computed tomography for the Phébus FPT3 test

    Energy Technology Data Exchange (ETDEWEB)

    Biard, B., E-mail: bruno.biard@irsn.fr

    2013-09-15

    Highlights: • The FP quantitative distribution in the fuel bundle is measured by gamma-spectrometry. • The FP location is obtained with emission tomograms and other experiment results. • X-ray tomograms provide the material and density mapping of the degraded bundle. • The self-attenuation may then be computed for each isotope at its key line energy. • Results are consistent with other FPT3 measurements, with acceptable uncertainties. -- Abstract: The international Phébus FP programme, initiated in 1988 by the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN), in cooperation with the European Commission (EC) and with financial support from USNRC, Canada, PSI/HSK (Switzerland), Japan and Korea, was aimed at studying severe accident phenomena: the fuel degradation, the release of fission products (FPs) and their transport through the reactor coolant system to the containment building. The FPT3 test, conducted in 2004, was the last of the five light water reactor core meltdown accident tests performed on irradiated fuel rods. After the experiment, the test device was recovered and analysed through a full set of non-destructive examinations performed over the fuel bundle zone, including gamma-scanning, gamma emission tomography, X-ray radiography and X-ray transmission tomography. The gamma-scanning was specifically devoted to the location, identification and amount quantification of the FPs remaining in the bundle. Since the fuel bundle became highly degraded during the experiment, the geometry was different at each level examined, and did not correspond to the well-known initial state. The self-attenuation of the test device and consequently the efficiency correction could then not be estimated by classical means that need to know the geometry of the object. Using the results of the other non-destructive examinations, specific computational tools and methods have therefore been developed to compute the self-attenuation of the bundle

  11. Theories of fission gas behaviour

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-01-01

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

  12. Structures and properties of (U,Pu)O{sub 2} containing non-active fission products. A simulation of irradiated nuclear fuel; Structure et proprietes de (U, Pu)O{sub 2} contenant des produits de fission sous forme inactive. Une simulation de combustible nucleaire irradie

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-07-01

    We have made oxides with the same uranium and plutonium content, the same stoichiometry and the same fission product content as an oxide fuel (U{sub 0,8}PuO{sub 2})O{sub 1,96} after 2 per cent burn up. We have calculated the stoichiometry changes due to irradiation and checked the calculation by X rays parameters measurements. We have calculated and measured the contraction of the oxide lattice due to fission products in solid solution. Microprobe analysis of precipitates have been made and have lead to the identification of non metallic barium containing compounds and have shown the particular behaviour of molybdenum. Some physical properties have been measured especially the electrical resistivity, the thermal diffusivity and the vapour pressure of zirconium in solid solution. (author) [French] Nous avons fabrique des oxydes dont la composition en uranium et plutonium, la stoechiometrie et la teneur en produit de fission, sont identiques a celles d'un oxyde (U{sub 0,8}PuO{sub 2})O{sub 1,96} ayant subi 2 pour cent de combustion. Nous avons calcule les changements de stoechiometrie entraines par l'irradiation et controle ces calculs par des mesures de parametre. Nous avons calcule et mesure la contraction du reseau due aux produits de fissions solubles dans la matrice. Des analyses a la microsonde des precipites de produits de fission insolubles ont ete faites et ont conduit a l'identification de composes non metalliques contenant du baryum et a la mise en evidence du role particulier du molybdene. Certaines proprietes physiques ont ete mesurees sur ces composes, en particulier la resistivite electrique, la diffusivite thermique et la tension de vapeur du zirconium dissout dans la matrice. (auteur)

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

    International Nuclear Information System (INIS)

    Waterbury, G.R.

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Waterbury, G.R. (comp.)

    1978-01-01

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

  15. Ideological Fission

    DEFF Research Database (Denmark)

    Christiansen, Steen Ledet

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

  16. The importance of mobile fission products for long-term safety in the case of disposal of vitrified high-level waste and spent fuel in a clay formation

    International Nuclear Information System (INIS)

    Marivoet, J.; Weetjens, E.

    2009-01-01

    In Belgium, the possibility to dispose of high-level radioactive waste in clay formations is studied since 1976. In the PAGIS report, which was the first performance assessment of the disposal of vitrified high-level waste in a clay formation and which was published in 1988, the most important contributors to the total dose via a water well pathway were 237 Np, 135 Cs and 99 Tc. Since 1988, several elements that strongly influence the calculated doses have evolved:?the inventory of long-lived mobile fission and activation products in vitrified high-level waste has been improved; the half-life of 79 Se has been re-estimated; substantial progress has been made in the determination of migration parameters of the main fission and activation products and actinides. In recent performance assessments, the actinides and 135 Cs do not significantly contribute to the total dose, as they remain confined in the host clay formation during several millions of years due to sorption on clay minerals. Consequently, the total dose resulting from the disposal of vitrified high-level waste or spent fuel is essentially due to releases of mobile fission and activation products. On the basis of recent waste inventory data and parameter values, the most important contributors to the total dose via a water well are: in the case of disposal of spent fuel: 79 Se, 129 I, 126 Sn, 36 Cl, and 99 Tc; in the case of disposal of vitrified HLW: 79 Se, 126 Sn, 36 Cl, 129 I, and 99 Tc. Important remaining uncertainties are the transfer factors of volatile fission and activation products into the vitrified waste during reprocessing and migration parameters of Se. (author)

  17. Mica fission detectors

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  18. Considerations for a sustainable nuclear fission energy in Europe

    International Nuclear Information System (INIS)

    Cognet, G.; Ledermann, P.; Cacuci, D.

    2005-01-01

    Presented is the global energy perspectives and and sustainable development fission vision scenario. Described are the innovative concepts with technological breakthroughs concerning the fuel cycle and evolution of the spent fuel radiotoxic contents

  19. 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)

  20. 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.

  1. Fission product release mechanisms and groupings

    International Nuclear Information System (INIS)

    Iglesia, F.C.; Brito, A.C.; Liu, Y.

    1995-01-01

    During CANDU postulated accidents the reactor fuel is estimated to be exposed to a variety of conditions. These conditions are dynamic and, during the course of an accident, the fuel may experience a wide range of temperatures and conditions from highly oxidizing to mildly reducing environments. The exposure of the reactor fuel to these environments and temperatures may affect its stoichiometry and release performance. In this paper a review of the important fission product release mechanisms is presented, the results of three out-of-pile experimental programs are summarized, and fission product release groups, for both oxidizing and reducing conditions are proposed. (author)

  2. Fission product release mechanisms and groupings

    Energy Technology Data Exchange (ETDEWEB)

    Iglesia, F C; Brito, A C; Liu, Y [Ontario Hydro, Toronto, ON (Canada); and others

    1996-12-31

    During CANDU postulated accidents the reactor fuel is estimated to be exposed to a variety of conditions. These conditions are dynamic and, during the course of an accident, the fuel may experience a wide range of temperatures and conditions from highly oxidizing to mildly reducing environments. The exposure of the reactor fuel to these environments and temperatures may affect its stoichiometry and release performance. In this paper a review of the important fission product release mechanisms is presented, the results of three out-of-pile experimental programs are summarized, and fission product release groups, for both oxidizing and reducing conditions are proposed. (author) 92 refs., 6 tabs.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  4. 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

  5. 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.

  6. Nuclear fission and reactions

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

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

  7. Laser driven fusion fission hybrids

    International Nuclear Information System (INIS)

    Hansen, L.F.; Maniscalco, J.A.

    1977-11-01

    The role of the fusion-fission hybrid reactor (FFHR) as a fissile fuel and/or power producer is discussed. As long range options to supply the world energy needs, hybrid-fueled thermal-burner reactors are compared to liquid metal fast breeder reactors (LMFBR). A discussion of different fuel cycles (thorium, depleted uranium, and spent fuel) is presented in order to compare the energy multiplication, the production of fissile fuel, the laser efficiency and pellet gain requirements of the hybrid reactor. Lawrence Livermore Laboratory (LLL) has collaborated with Bechtel Corporation and with Westinghouse in two engineering design studies of laser fusion driven hybrid power plants. The hybrid designs which have resulted from these two studies are briefly described and analyzed by considering operational parameters, such as energy multiplication, power density, burn-up and plutonium production as a function time

  8. Measurements of fission yields

    International Nuclear Information System (INIS)

    Denschlag, H.O.

    2000-01-01

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

  9. Radiochemical studies on fission

    Energy Technology Data Exchange (ETDEWEB)

    None

    1973-07-01

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

  10. 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

  11. Thermodynamic analysis of volatile organometallic fission products

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  12. ELSA: A simplified code for fission product release calculations

    International Nuclear Information System (INIS)

    Manenc, H.; Notley, M.J.

    1996-01-01

    During a light water reactor severe accident, fission products are released from the overheated core as it progressively degrades. A new computer module named ELSA is being developed to calculate fission product release. The authors approach is to model the key phenomena, as opposed to more complete mechanistic approaches. Here they present the main features of the module. Different release mechanisms have been identified and are modeled in ELSA, depending on fission product volatility: diffusion seems to govern the release of the highly volatile species if fuel oxidation is properly accounted for, whereas mass transport governs that of lower volatility fission products and fuel volatilization that of the practically involatile species

  13. Fission-suppressed hybrid reactor: the fusion breeder

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Coops, M.S.

    1982-12-01

    Results of a conceptual design study of a 233 U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed

  14. 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)

  15. 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.)

  16. Proportioning of {sup 79}Se and {sup 126}Sn long life radionuclides in the fission products solutions coming from spent fuels processing; Dosage des radionucleides a vie longue {sup 79}Se et {sup 126}Sn dans les solutions de produits de fission issues du traitement des combustibles nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Comte, J

    2001-11-01

    The determination of radionuclides present in waste resulting from the nuclear fuel reprocessing is a request from the regulatory authorities to ensure an optimal management of the storage sites. Long-lived radionuclides (T{sub 1/2} > 30 years) are particularly concerned owing to the fact that their impact must be considered for the long term. Safety studies have established a list of long-lived radionuclides (LLRN) whose quantification is essential for the management of the disposal site. Among these, several are pure {beta} emitters, present at low concentration levels in complex matrices. Their determination, by radiochemical method or mass spectrometry, involves selective chemical separations from the others {beta}/{gamma} emitters and from the measurement interfering elements. The work undertaken in this thesis relates to the development of analytical methods for the determination of two long-lived radionuclides: selenium 79 and tin 126, in acid solutions of fission products present in nuclear fuel reprocessing plant. For selenium 79, a {beta} emitter with a half live estimated to be 10{sup 6} years, the bibliography describes different chemical separation methods including precipitation, liquid-liquid extraction and chromatography on ionic resins. After optimisation on a synthetic solution, two of these techniques, precipitation by potassium iodine and separation with ion exchange resins were applied to a genuine solution of fission products at Cogema La Hague. The results showed that only the ion exchange method allows us to obtain a solution sufficiently decontaminated (FD{beta}{gamma} = 250) with a significant selenium recovery yield (85%). This separation allows the measurement of the {sup 79}Se by electrothermal vaporization coupled with inductively coupled plasma mass spectrometry (ETV-ICP/MS), after transfer of the samples to CEA/Cadarache. The concentration of {sup 79}Se measured is 0,42 mg/L in the solution of fission products with an isotopic ratio

  17. Actinide and fission product separation and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-07-01

    The second international information exchange meeting on actinide and fission product separation and transmutation, took place in Argonne National Laboratory in Illinois United States, on 11-13 November 1992. The proceedings are presented in four sessions: Current strategic system of actinide and fission product separation and transmutation, progress in R and D on partitioning processes wet and dry, progress in R and D on transmutation and refinements of neutronic and other data, development of the fuel cycle processes fuel types and targets. (A.L.B.)

  18. Migration of fission products in UO2. Final report

    International Nuclear Information System (INIS)

    Prussin, S.G.; Olander, D.R.

    1995-01-01

    Results of an experimental and calculational effort to examine the fundamental mechanisms of fission product migration in and release from polycrystalline uranium dioxide are reported. The experiments were designed to provide diffusion parameters for the representative fission products tellurium, iodine, xenon, molybdenum and ruthenium under both reducing and oxidizing conditions. The calculational effort applied a new model of fission product release from reactor fuel that incorporates grain growth as well as grain boundary and lattice diffusion

  19. Safety device for nuclear fission reactors

    International Nuclear Information System (INIS)

    Brownlee, M.L.

    1982-01-01

    A plurality of radially arranged and neutron absorbing baffles are stacked in vertical sets under the fuel core assemblies, and the whole enclosed in a bottle shaped containment vessel. The radially arranged baffles of each set extend vertically, and each set has double the number of baffles as the set above it in the stack. A melt-down of a fuel core assembly drops the fissioning nuclear fuel into the stacked sets of baffles, there, as it passes through, to be progressively divided, redivided and dispersed in smaller and smaller masses between the doubling number of baffles in safe fuel pellet size. Neutron absorbing containment prevents contamination of the environment and together with cooling means stops fissioning of fuel

  20. Fission Research at IRMM

    Directory of Open Access Journals (Sweden)

    Al-Adili A.

    2010-03-01

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

  1. 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

  2. On the safety of conceptual fusion-fission hybrid reactors

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  3. Equilibrium fission model calculations

    International Nuclear Information System (INIS)

    Beckerman, M.; Blann, M.

    1976-01-01

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

  4. Irradiated uranium reprocessing, Final report I-VI, IV Deo IV - Separation of uranium, plutonium and fission products from the irradiated fuel of the reactor in Vinca; Prerada ozracenog urana. Zavrsni izvestaj - I-VI, IV Deo - Odvajanje urana, plutonijuma i fisionih produkata iz isluzenog goriva reaktora u Vinci

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-12-15

    This study describes the technology for separation of uranium, plutonium and fission products from the radioactive water solution which is obtained by dissolving the spent uranium fuel from the reactor in Vinca. The procedure should be completed in a hot cell, with the maximum permitted activity of 10 Ci.

  5. Fission-product release during accidents

    International Nuclear Information System (INIS)

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

    1991-09-01

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

  6. Energy from nuclear fission an introduction

    CERN Document Server

    De Sanctis, Enzo; Ripani, Marco

    2016-01-01

    This book provides an overview on nuclear physics and energy production from nuclear fission. It serves as a readable and reliable source of information for anyone who wants to have a well-balanced opinion about exploitation of nuclear fission in power plants. The text is divided into two parts; the first covers the basics of nuclear forces and properties of nuclei, nuclear collisions, nuclear stability, radioactivity, and provides a detailed discussion of nuclear fission and relevant topics in its application to energy production. The second part covers the basic technical aspects of nuclear fission reactors, nuclear fuel cycle and resources, safety, safeguards, and radioactive waste management. The book also contains a discussion of the biological effects of nuclear radiation and of radiation protection, and a summary of the ten most relevant nuclear accidents. The book is suitable for undergraduates in physics, nuclear engineering and other science subjects. However, the mathematics is kept at a level that...

  7. 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

  8. Radiation shielding for fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

    2000-03-01

    Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

  9. The fusion-fission hybrid

    International Nuclear Information System (INIS)

    Teller, E.

    1985-01-01

    As the history of the development of fusion energy shows, a sustained controlled fusion reaction is much more difficult to produce than rapid uncontrolled release of fusion energy. Currently, the ''magnetic bottle'' technique shows sufficient progress that it might applied for the commercial fuel production of /sup 233/U, suitable for use in fission reactors, by developing a fusion-fission hybrid. Such a device would consist of a fusion chamber core surrounded by a region containing cladded uranium pellets cooled by helium, with lithium salts also present to produce tritium to refuel the fusion process. Successful development of this hybrid might be possible within 10 y, and would provide both experience and funds for further development of controlled fusion energy

  10. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 3. Alternatives for interim storage and transportation

    International Nuclear Information System (INIS)

    1976-05-01

    Volume III of the five-volume report contains information on alternatives for interim storage and transportation. Section titles are: interim storage of spent fuel elements; interim storage of chop-leach fuel bundle residues; tank storage of high-level liquid waste; interim storage of solid non-high-level wastes; interim storage of solidified high-level waste; and, transportation alternatives

  11. Estimation of penetration depth of fission products in cladding Hull

    International Nuclear Information System (INIS)

    Kim, Hee Moon; Jung, Yang Hong; Yoo, Byong Ok; Choo, Yong Sun; Hong, Kwon Pyo

    2005-01-01

    A disposal and a reprocessing for spent fuel rod with high burnup need de-cladding procedure. Pellet in this rod has been separated from a cladding hull to reduce a radioactivity of hull by chemical and mechanical methods. But fission products and actinides(U,Pu) still remain inside of cladding hull by chemical bonding and fission spike, which is called as 'contamination'. More specific removal of this contamination would have been considered. In this study, the sorts of fission products and penetration depth in hull were observed by EPMA test. To analyze this behavior, SRIM 2000 code was also used as energies of fission products and an oxide thickness of hull

  12. Fuel assemblies

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi.

    1979-01-01

    Purpose: To prevent scattering of gaseous fission products released from fuel assemblies stored in an fbr type reactor. Constitution; A cap provided with means capable of storing gas is adapted to amount to the assembly handling head, for example, by way of threading in a storage rack of spent fuel assemblies consisting of a bottom plate, a top plate and an assembly support mechanism. By previously eliminating the gas inside of the assembly and the cap in the storage rack, gaseous fission products upon loading, if released from fuel rods during storage, are stored in the cap and do not scatter in the storage rack. (Horiuchi, T.)

  13. Actinide and fission product partitioning and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The third international information exchange meeting on actinide and fission product partitioning and transmutation, took place in Cadarache France, on 12-14 December 1994. The proceedings are presented in six sessions : an introduction session, the major programmes and international cooperation, the systems studies, the reactors fuels and targets, the chemistry and a last discussions session. (A.L.B.)

  14. Fission level densities

    International Nuclear Information System (INIS)

    Maslov, V.M.

    1998-01-01

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

  15. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 2. Alternatives for waste treatment

    International Nuclear Information System (INIS)

    1976-05-01

    Volume II of the five-volume report is devoted to the description of alternatives for waste treatment. The discussion is presented under the following section titles: fuel reprocessing modifications; high-level liquid waste solidification; treatment and immobilization of chop-leach fuel bundle residues; treatment of noncombustible solid wastes; treatment of combustible wastes; treatment of non-high-level liquid wastes; recovery of transuranics from non-high-level wastes; immobilization of miscellaneous non-high-level wastes; volatile radioisotope recovery and off-gas treatment; immobilization of volatile radioisotopes; retired facilities (decontamination and decommissioning); and, modification and use of selected fuel reprocessing wastes

  16. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 2. Alternatives for waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Volume II of the five-volume report is devoted to the description of alternatives for waste treatment. The discussion is presented under the following section titles: fuel reprocessing modifications; high-level liquid waste solidification; treatment and immobilization of chop-leach fuel bundle residues; treatment of noncombustible solid wastes; treatment of combustible wastes; treatment of non-high-level liquid wastes; recovery of transuranics from non-high-level wastes; immobilization of miscellaneous non-high-level wastes; volatile radioisotope recovery and off-gas treatment; immobilization of volatile radioisotopes; retired facilities (decontamination and decommissioning); and, modification and use of selected fuel reprocessing wastes. (JGB)

  17. Measurement of the fission cross section induced by fast neutrons of the {sup 232}Th/{sup 233}U nuclei within the innovating fuel cycles framework; Mesure de la section efficace de fission induite par neutrons rapides des noyaux {sup 232}Th/{sup 233}U dans le cadre des cycles de combustible innovants

    Energy Technology Data Exchange (ETDEWEB)

    Grosjean, C

    2005-03-15

    The thorium-U{sup 233} fuel cycle might provided safer and cleaner nuclear energy than the present Uranium/Pu fuelled reactors. Over the last 10 years, a vast campaign of measurements has been initiated to bring the precision of neutron data for the key nuclei (Th{sup 232}, Pa{sup 233} and U{sup 233}) at the level of those for the U-Pu cycle. This is the framework of these measurements, the energy dependent neutron induced fission cross section of Th{sup 232} and U{sup 233} has been measured from 1 to 7 MeV with a target accuracy lesser than 5 per cent. These measurements imply the accurate determination of the fission rate, the number of the target nuclei as well as the incident neutron flux impinging on the target, the latter has been obtained using the elastic scattering (n,p). The cross section of which is very well known in a large neutron energy domain ({approx} 0,5 % from 1 eV to 50 MeV) compared to the U{sup 235}(n,f) reaction. This technique has been applied for the first time to the Th{sup 232}(n,f) and U{sup 233}(n,f) cases. A Hauser-Feshbach statistical model has been developed. It consists of describing the different decay channels of the compound nucleus U{sup 234} from 0,01 to 10 MeV neutron energy. The parameters of this model were adjusted in order to reproduce the measured fission cross section of U{sup 233}. From these parameters, the cross sections from the following reactions could be extracted: inelastic scattering U{sup 233}(n,n'), radiative capture U{sup 233}(n,{gamma}) and U{sup 233}(n,2n). These cross sections are still difficult to measure by direct neutron reactions. The calculated values have allowed us to fill the lack of experimental data for the major fissile nucleus of the thorium cycle. (author)

  18. Release of fission products from a fuel rod with an artificial hole through cladding irradiated in an in-pile water loop, (2)

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi

    1978-11-01

    To make clear the iodine spiking phenomenon from a defective fuel rod into the primary coolant, the fuel rod (UO 2 pellets, with stainless steel sheath) with an artificial pin hole was irradiated in the inpile test section of water loop JMTR.OWL-1. Experimental conditions were depressurization and temperature drop of the primary loop coolant and diameter and position of the pin hole. Iodine 131 and cesium 137 in loop coolant were measured under various coolant conditions. The inventory and translation rate of iodine 131 in fuel rod related to irradiation histories were calculated. The levels of I-131 and Cs-137 released to loop coolant from fuel rod were compared. Comparison of the results with LWRs was made by way of the spiked amount and release rate of iodine 131. (author)

  19. The role of fission products in whole core accidents

    Energy Technology Data Exchange (ETDEWEB)

    Baker, A R [FRSD, UKAEA, RNPDE, Risley, Warrington (United Kingdom); Teague, H J [SRD, UKAEA, Culcheth, Warrington (United Kingdom)

    1977-07-01

    The review of the role of fission products in whole-core accidents falls into two parts. Firstly, there is a discussion of the hypothetical accidents usually considered in the UK and how they are dealt with. Secondly, there is a discussion of individual topics where fission products are known to be important or might be so. There is a brief discussion of the UK work on the establishment of an equation of state for unirradiated fuel and how this might be extended to incorporate fission product effects. The main issue is the contribution of fission products to the effective vapour pressure and the experimental programme on the pulsed reactor VIPER investigates this. Fission products may influence the probability of occurrence and the severity of MFCIs. Finally, the fission product effects in the pre-disassembly, disassembly and recriticality stages of an accident are discussed. (author)

  20. Design, Fabrication, and Testing of a Laboratory-Scale Voloxidation System for Removal of Tritium and Other Volatile Fission Products from Used Nuclear Fuel

    International Nuclear Information System (INIS)

    Spencer, Barry B; DelCul, Guillermo D; Bradley, Eric Craig; Jubin, Robert Thomas; Hylton, Tom D; Collins, Emory D

    2008-01-01

    Advanced nuclear fuel processing methodologies are being demonstrated at the Oak Ridge National Laboratory (ORNL) as part of the Global Nuclear Energy Partnership (GNEP) program. A coupled end-to-end (CETE) research and development (R and D) capability is being installed to provide all primary processing operations, ranging from spent fuel receipt to production of products and waste forms. This R and D capability is designed for small, laboratory-scale throughput and will permit conduct of experiments in the range of 20 kg of spent fuel per year. The head-end processing segment includes single-pin shearing, voloxidation to remove tritium from the fuel before it enters the aqueous based separations systems, cleanup of the cladding hulls for disposition, and transfer of the fuel powder to the dissolution process. This paper describes the voloxidation system design and presents results from the cold checkout of the hardware. Preliminary results of the initial processing campaign with spent fuel is presented as well

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

    International Nuclear Information System (INIS)

    Ikusawa, Yoshihisa; Ozawa, Takayuki

    2008-03-01

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

  2. 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

  3. Nuclear fuel

    International Nuclear Information System (INIS)

    D Hondt, P.

    1998-01-01

    The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

  4. Fission product behavior in the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

    1975-10-01

    Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with 235 U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233 U fuel over a period of about 15 months (more than 5100 effective full-power hours)

  5. 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

  6. Materials specific work at Forschungszentrum Karlsruhe and in cooperation with the industrial partners ALKEM and Interatom for the development of nuclear oxide fuels for fission reactors

    International Nuclear Information System (INIS)

    Kleykamp, H.; Muehling, G.

    2005-09-01

    The fabrication of uranium-plutonium oxide fuel started in Forschungszentrum Karlsruhe and at ALKEM company to begin for the criticality experiments in the SNEAK reactor and subsequently for stationary fuel pin irradiations in the FR2, BR2, DFR, Rapsodie, Phenix and KNK II reactors. The production methods comprised first the mechanical blending of UO2 and PuO2 followed by direct pressing and sintering of the pellets, later the advanced methods such as optimized comilling and ammonium uranyl plutonyl coprecititation. The fabrication of pellets was described in the main, further the alternative fuel pin manufacturing processes by vibrational compaction and hot-impact densification were discussed. The first capsule and pin irradiations in the FR2 and BR2 reactors contributed to the assessment of the maximum operation parameters within the fuel pin development such as linear heat rating, cladding temperature and burnup. Subsequently, small-bundle and largebundle irradiations were made in fast reactors in cooperation with Interatom company in order to verify the specifications for the commercial fast reactor SNR 300. Milestones were the maximum burnup of 175 GWd/t metal, corresponding 18.6 % of the heavy atoms, obtained in one of the KNK II fuel pin assemblies, and the displacement rates in the cladding materials of 140 dpa NRT attained in the Phenix reactor. Higher implications gained later the stationary irradiations of defected mixed-oxide pins, the mild fuel pin transient operations, the local blockage experiments and the severe hypothetic accidents in the respective Siloe, HFR, BR2 and CABRI reactors. These experiments were made solely in international partnership. Further activities were the chemical analyses of solid residues and coprecipitations of irradiated mixed-oxide fuels in the head-end of the reprocessing. All these actions were coordinated in the then fast breeder project. Furthermore, irradiated fuels and fuel pins of other reactor types were

  7. Status of fission power

    International Nuclear Information System (INIS)

    Levenson, M.

    1977-01-01

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

  8. 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)

  9. Fission neutron multiplicity calculations

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  10. Intermediate energy nuclear fission

    International Nuclear Information System (INIS)

    Hylten, G.

    1982-01-01

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

  11. Material synergism fusion-fission

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  12. 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

  13. Background and derivation of ANS-5.4 standard fission product release model. Technical report

    International Nuclear Information System (INIS)

    1982-01-01

    ANS Working Group 5.4 was established in 1974 to examine fission product releases from UO2 fuel. The scope of ANS-5.4 was narrowly defined to include the following: (1) Review available experimental data on release of volatile fission products from UO2 and mixed-oxide fuel; (2) Survey existing analytical models currently being applied to lightwater reactors; and (3) Develop a standard analytical model for volatile fission product release to the fuel rod void space. Place emphasis on obtaining a model for radioactive fission product releases to be used in assessing radiological consequences of postulated accidents

  14. Geology behind nuclear fission technology

    International Nuclear Information System (INIS)

    Dhana Raju, R.

    2005-01-01

    Geology appears to have played an important role of a precursor to Nuclear Fission Technology (NFT), in the latter's both birth from the nucleus of an atom of and most important application as nuclear power extracted from Uranium (U), present in its minerals. NFT critically depends upon the availability of its basic raw material, viz., nuclear fuel as U and/ or Th, extracted from U-Th minerals of specific rock types in the earth's crust. Research and Development of the Nuclear Fuel Cycle (NFC) depends heavily on 'Geology'. In this paper, a brief review of the major branches of geology and their contributions during different stages of NFC, in the Indian scenario, is presented so as to demonstrate the imp