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Sample records for residual fission-product solutions

  1. Preparative electrophoresis of industrial fission product solutions

    International Nuclear Information System (INIS)

    Tret, Joel

    1971-07-01

    The aim of this work is to contribute to the development of the continuous electrophoresis technique while studying its application in the preparative electrophoresis of industrial fission product solutions. The apparatus described is original. It was built for the purposes of the investigation and proved very reliable in operation. The experimental conditions necessary to maintain and supervise the apparatus in a state of equilibrium are examined in detail; their stability is an important factor, indispensable to the correct performance of an experiment. By subjecting an industrial solution of fission products to preparative electrophoresis it is possible, according to the experimental conditions, to prepare carrier-free radioelements of radiochemical purity (from 5 to 7 radioelements): 137 Cs, 90 Sr, 141+144 Ce, 91 Y, 95 Nb, 95 Zr, 103+106 Ru. (author) [fr

  2. Vitrification processes for fission product solutions

    International Nuclear Information System (INIS)

    Bonniaud, R.; Jouan, A.; Moncouyoux, J.P.; Sombret, C.

    1982-10-01

    The different processes for fission product vitrification in the world are reviewed. Continuous or discontinuous processes, induction or arc heating, in can melting or casting, tests with radioactive or simulated wastes and industrial realizations are described [fr

  3. Attachment behavior of fission products to solution aerosol

    Energy Technology Data Exchange (ETDEWEB)

    Takamiya, Koichi; Tanaka, Toru; Nitta, Shinnosuke; Itosu, Satoshi; Sekimoto, Shun; Oki, Yuichi; Ohtsuki, Tsutomu [Research Reactor Institute, Kyoto University, Osaka (Japan)

    2016-12-15

    Various characteristics such as size distribution, chemical component and radioactivity have been analyzed for radioactive aerosols released from Fukushima Daiichi Nuclear Power Plant. Measured results for radioactive aerosols suggest that the potential transport medium for radioactive cesium was non-sea-salt sulfate. This result indicates that cesium isotopes would preferentially attach with sulfate compounds. In the present work the attachment behavior of fission products to aqueous solution aerosols of sodium salts has been studied using a generation system of solution aerosols and spontaneous fission source of {sup 248}Cm. Attachment ratios of fission products to the solution aerosols were compared among the aerosols generated by different solutions of sodium salt. A significant difference according as a solute of solution aerosols was found in the attachment behavior. The present results suggest the existence of chemical effects in the attachment behavior of fission products to solution aerosols.

  4. Solidification of residual fission-product solutions; laboratory studies; Solidification de solutions residuaires concentrees de produits de fission: etudes de laboratoire

    Energy Technology Data Exchange (ETDEWEB)

    Boniaud, R; Cohen, P; Sombret, C [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1961-07-01

    This paper describes the results obtained, at laboratory scale, during the study of the incorporation of fission products into glasses and synthetic micas. The rate of leaching of fission products from the glass and their volatility during firing were measured. A hot cell was built to complete these results. (author) [French] Ce rapport resume le resultat des etudes faites en laboratoire (activite de l'ordre du millicurie) sur l'incorporation des produits de fission dans des verres et micas synthetiques. On a mesure le taux de lixiviation des produits de fission et leur volatilisation au cours de la cuisson. Une cellule chaude a ete installee pour completer ces resultats au moyen d'essais realises avec une activite superieure. (auteur)

  5. Waste disposal process on the basis of fission product solutions, and suitable plant

    International Nuclear Information System (INIS)

    Thiele, D.

    1984-01-01

    The nitrous fission product solution containing ruthenium is concentrated in a wiper blade evaporator. After intermediate drying the residue is vitrified adding vitrifying agents and NH4 in an amount of 20 to 300% referred to the nitrate content of the concentrated residue. (orig./PW)

  6. Recovery of fission products from acidic waste solutions thereof

    International Nuclear Information System (INIS)

    Carlin, W.W.; Darlington, W.B.; Dubois, D.W.

    1975-01-01

    Fission products, e.g., palladium, ruthenium and technetium, are removed from aqueous, acidic waste solutions thereof. The acidic waste solution is electrolyzed in an electrolytic cell under controlled cathodic potential conditions and technetium, ruthenium, palladium and rhodium are deposited on the cathode. Metal deposit is removed from the cathode and dissolved in acid. Acid insoluble rhodium metal is recovered, dissolved by alkali metal bisulfate fusion and purified by electrolysis. In one embodiment, the solution formed by acid dissolution of the cathode metal deposit is treated with a strong oxidizing agent and distilled to separate technetium and ruthenium (as a distillate) from palladium. Technetium is separated from ruthenium by organic solvent extraction and then recovered, e.g., as an ammonium salt. Ruthenium is disposed of as waste by-product. Palladium is recovered by electrolysis of an acid solution thereof under controlled cathodic potential conditions. Further embodiments wherein alternate metal recovery sequences are used are described. (U.S.)

  7. Fission products control by gamma spectrometry in purex process solutions

    International Nuclear Information System (INIS)

    Goncalves, Maria Augusta

    1982-01-01

    This paper deals with a radiometric method for fission products analysisby gamma spectrometry. This method will be applied for fission productscontrol at the irradiated material processing facility, under construction inthe Instituto de Pesquisas Energeticas e Nucleares, SP, Brazil. Countinggeometry was defined taking into account the activities of process solutionsto be analysed, the remotely operated aliquotation device of analytical celland the available detection system. Natural and 19,91% enriched uraniumsamples were irradiated at IEAR-1 reactor in order to simulate thecomposition of Purex process solutions. After a short decay time, the sampleswere dissolved with HNO 3 and then, conditioned in standard flasks withdefined geometry. The spectra were obtained by a Ge(Li) semiconductordetector and analysed by the GELIGAM software system, losing a floppy-diskconnected to a PDP-11/05 computer. Libraries were prepared and calibrationswere made with standard sources to fit the programs to the analysis offission products in irradiated uranium solutions. It was possible to choosethe best program to be used in routine analysis with the obtained data.(author)

  8. Extraction of fission product rhodium from nitric acid solutions. 1

    International Nuclear Information System (INIS)

    Gorski, B.; Beer, M.; Russ, L.

    1988-01-01

    The extraction of noble metals from nitric acid solutions represents one problem of separating valueable substances from nuclear wastes in nuclear fuel reprocessing. Results of distribution experiments demonstrate the possibility of solvent extraction of rhodium using tertiary amines in presence of nitrite. Even short mixing times realize high distribution coefficients allowing quantitative separation from aqueous solutions. (author)

  9. Recovery of fission products from waste solutions utilizing controlled cathodic potential electrolysis

    International Nuclear Information System (INIS)

    Carlin, W.W.; Darlington, W.B.

    1975-01-01

    Fission products, e.g., palladium, rhodium and technetium, are recovered from aqueous waste solutions thereof, e.g., aged Purex alkaline waste solutions. The metal values from the waste solutions are extracted by ion exchange techniques. The metals adsorbed by the ion exchange resin are eluted and selectively recovered by controlled cathodic potential electrolysis. The metal values deposited on the cathode are recovered and, if desired, further purified

  10. Radiochemical applications of insoluble sulfate columns. Analytical possibilities in the field of the fission product solutions

    International Nuclear Information System (INIS)

    Barrachina, M.; Sauvagnac, R.

    1962-01-01

    In this paper we go on with our study of the heterogeneous ion-isotopic exchange in column. At present, we apply it to determine the radiochemical composition of the raw solutions used in the industrial recuperation of the long-lived fission products. The separation of the radioelements contained in these solutions is carried out mainly by making use of small columns, 1-3 cm height, of BaSO 4 or SrSO 4 , under selected experimental conditions. These columns behave like a special type of inorganic exchangers, working by absorption or by ion-isotopic exchange depending on the cases,a nd they provide the means for the selective separation of several important fission products employing very small volumes of fixing and eluting solutions. (Author) 11 refs

  11. Removal of fission products from waste solutions using 16 different soil samples

    International Nuclear Information System (INIS)

    Bangash, M.A.; Hanif, J.

    1997-01-01

    Most of the nuclear sites use pits in the surrounding soils for the storage/disposal of low active waste (LAW) solutions. The characteristics of the soil if not suitable for the fixation or adsorption of the radioactive nuclides, may cause migration of these nuclides to hydrosphere. The phenomenon has the risk of radio toxic pollution for the living bodies therefore minerals composing the soil and their adsorption properties need to be investigated. For this purpose 16 different soil samples were collected from all over Pakistan. Mineralogical composition of the soils was determined by X-ray diffraction analysis. It was found that most of the samples contained clay minerals, illite, kaolinite and montmorillonite. Studies for the removal of fission products like, /sup 137/Cs. /sup 60/Sr and activation product /sup 60/CO from solution were carried out on these samples. The sorption experiments were performed by batch technique using radioactive as tracers. Distribution co-efficient were determined by mixing he element solution at pH 3 with the soil at soil solution ratios of 1 to 20. It is revealed from the experimental data that efficient removal of fission products from solutions is achieved by soil samples containing clay mineral montmorillonite, followed by little and kaolinite. These soils thus can be effectively used for the disposal of low level radioactive waste solutions without causing any environmental hazard. (author)

  12. 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.

  13. The present state of research on the vitrification of concentrated solutions of fission products (1962)

    International Nuclear Information System (INIS)

    Bonniaud, R.; Sombret, C.

    1961-01-01

    The present report gives the actual point of studies on vitrification of concentrated solutions of fission products. An active cell, giving glasses in crucibles, permitted to study various glass compositions. The leaching rate from the glass raises 1 to 2 10 -7 g of glass/cm 2 /day. Activity loss by volatility during vitrification remains weak and often below 0.1 per cent of total activity. Off gas cleaning is made easier by presence of filter which is compound of granules including iron oxide. After saturation the content of this filter can be melt. Moreover different processes are in experimentation for a more important production. Daily 72 liters of solution containing tracer activity are treated in a continuous calcination and vitrification plant. The loss in 106 Ru is still important and a modification of installation has been necessary. A pot vitrification plant is in study. In order to reduce cost of processing the possibility to pour glass after melting is actuality in study. A production set of very active glass is also in project. (authors) [fr

  14. 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)

  15. Experiments on the incorporation of concentrated solutions of fission products in glasses and micas

    International Nuclear Information System (INIS)

    Bonniaud, R.; Cohen, P.; Sombret, C.

    1958-01-01

    1) The plants designed for extracting the plutonium of the G1, G2, G3 reactors fuel rods will produce extremely concentrated solutions of fission products. 2) Let us consider a synthetic solution of the 'G2' type: (2N NO 3 H - 2,25 meq/cm 3 AI +++ - 5.10 -3 134 Cs - 137 Cs tracer). We made various glasses and micas by adding and mixing the necessary adjuvants and baking (900 to 1400 deg. C) in a graphite crucible. 3) The products obtained had either the shape of a cylindrical pellet or were reduced into a fine powder. They were mixed with 300 cm 3 of synthetic sea water during variable periods of time in order to study leaching of the activity. 4) Experiments were first carried on caesium because of its solubility. 5) Fabrication of micas on a large scale sets many technological problems more difficult to solve (1400 deg. C) than in the case of glasses (1000 deg. C). A comparative study on both micas and glasses showed that leaching of activity was more important in the micas. (author) [fr

  16. Analytic solutions to linear, time-dependent fission product deposition models for isothermal laminar, slug, or multiregion flow conditions

    International Nuclear Information System (INIS)

    Durkee, J.W. Jr.

    1983-01-01

    The time-dependent convective-diffusion equation with radioactive decay is solved analytically in axisymmetric cylindrical geometry for laminar and slug velocity profiles under isothermal conditions. Concentration dependent diffusion is neglected. The laminar flow solution is derived using the method of separation of variables and Frobenius' technique for constructing a series expansion about a regular singular point. The slug flow multiregion solution is obtained using the method of separation of variables. The Davidon Variable Metric Minimization algorithm is used to compute the coupling coefficients. These solutions, which describe the transport of fission products in a flowing stream, are then used to determine the concentration of radioactive material deposited on a conduit wall using a standard mass transfer model. Fission product deposition measurements for five diffusion tubes in a Fort St. Vrain High-Temperature Gas-Cooled reactor plateout probe are analyzed. Using single region slug and laminar models, the wall mass transfer coefficients, diffusion coefficients, and inlet concentrations are determined using least squares analysis. The diffusion coefficients and inlet concentrations are consistent between tubes. The derived diffusion coefficients and wall mass transfer coefficients are in relative agreement with known literature values

  17. 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

  18. 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

  19. Removal of fission product ruthenium from purex process solutions: thiourea as complexing agent

    International Nuclear Information System (INIS)

    Floh, B.; Abrao, A.

    1980-01-01

    A new method for the treatment of spent uranium fuel is presented. It is based on the Purex Process using thiourea to increase the ruthenium decontamination factor. Thiourea exhibits a strong tendency for the formation of coordination compounds in acidic media. This tendency serves as a basis to transform nitrosyl-ruthenium species into Ru /SC(NH)(NH 2 )/ 2+ and Ru /SC(NH)(NH 2 )/ 3 complexes which are unextractable by TBP-varsol. The best conditions for the ruthenium-thiourea complex formation were found to be: thiourea-ruthenium ratio (mass/mass) close to 42, at 75 0 C, 30 minutes reaction time and aging period of 60 minutes. The ruthenium decontamination factor for a single uranium extraction are ca. 80-100, not interfering with extraction of actinides. These values are rather high in comparison to those obtained using the conventional Purex Process (e.g. F.D. sub(Ru)=10). By this reason the method developed here is suitable for the treatment of spent uranium fuels. Thiourea (100g/l) scrubbing experiments of ruthenium, partially co-extracted with actinides, confirmed the possibility of its removal from the extract. A decontamination greater than 83,5% for ruthenium as fission product is obtained in two stages with this procedure. (Author) [pt

  20. Exact Solution of Fractional Diffusion Model with Source Term used in Study of Concentration of Fission Product in Uranium Dioxide Particle

    International Nuclear Information System (INIS)

    Fang Chao; Cao Jianzhu; Sun Lifeng

    2011-01-01

    The exact solution of fractional diffusion model with a location-independent source term used in the study of the concentration of fission product in spherical uranium dioxide (UO 2 ) particle is built. The adsorption effect of the fission product on the surface of the UO 2 particle and the delayed decay effect are also considered. The solution is given in terms of Mittag-Leffler function with finite Hankel integral transformation and Laplace transformation. At last, the reduced forms of the solution under some special physical conditions, which is used in nuclear engineering, are obtained and corresponding remarks are given to provide significant exact results to the concentration analysis of nuclear fission products in nuclear reactor. (nuclear physics)

  1. Polarographic determination of Iodide and Iodate, in Solutions Coming from Aerosols in Fission Products Containment Studies in Nuclear Power Stations

    International Nuclear Information System (INIS)

    Sanchez, M.; Ballesteros, O.; Fernandez, M.; Clavero, M.A.; Gonzalez, A.M.

    2000-01-01

    A polarographic method is described for the iodine species determination, iodide and iodate in water solutions. the iodate can be determined by differential pulse polarography. Calibration curves and the detection and determination limits have been obtained. Iodides is oxidized to iodate with sodium hypochlorite and the excess of oxidizing agent is destroyed with sodium sulphide. The concentration of iodide is calculated as the difference between the concentration of iodate in the sample before and after the oxidation. As an application, species of iodine in samples coming from the experimental plants GIRS (Gaseous Iodine Removal by Sprays) of Nuclear Fission Department of the CIEMAT, dedicated to fission products containment studies in nuclear power station, were determined. (Author) 10 refs

  2. 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.)

  3. Studies on the effect of Pu and fission products on extraction selectivity of polymeric composite of ammonium phosphomolybdate towards 137Cs from nitric acid solutions

    International Nuclear Information System (INIS)

    Tripathi, S.C.; Pathak, S.K.; Jambunathan, U.; Dakshinamoorthy, A.; Kumar, Manmohan; Bajaj, P.N.

    2007-01-01

    Ultrafine particle size of ammonium phosphomolybdate (AMP) though selective sorbent for Cs form acidic solution, has limitation due to its incompatibility with column operation under practical conditions. A polymeric composite containing ammonium Phospho molybdate(AMP) developed indigenously in granular form, has shown encouraging results for Cs uptake from acidic solutions. Present investigation describes experimental observations carried out to find out the effect of presence of Pu and fission products on the 137 Cs uptake. The polymeric material shown some tendency for sorption of Pu which is found to decrease with increasing acidity. It has great selectivity with respect to other fission products like cerium, zirconium, and ruthenium. (author)

  4. Fission products determination in high activity waste solution by wavelength dispersive X-ray fluorescence spectral interference correction by intensity ratio

    International Nuclear Information System (INIS)

    Sato, I.M.

    1988-01-01

    Fission products Se, Rb, Y, Zr, Mo, Ru, Rh, Pd, Te, Cd, Cs, Ba, La, Ce, Pr, Nd, Sm, Eu and Gd were determined in simulated high activity radioactive waste solution by wavelength dispersive X-ray fluorescence spectrometry without chemical separation. Thin layer technique was employed for the sample preparation. For the L spectral lines, the absorption effect was verified by Rasberry-Heinrich, Lucas Tooth-Pyne and Lachance-Trail relations. This effect was quantified and corrected accordingly. The spectral interferences of Kα and/or Lα lines of Y, Zr, Mo, La, Ce, Pr, Nd, Sm, Eu and Gd elements were eliminated by the intensity ratio method. The overlapping of up to three analytical lines was corrected by applying this method. The concentration influence of the interfering element on the intensity ratio values as well the efficiency of the correction method were investigated in order to assure that no systematic or residual error, resulting from the correction, affect the actual fluorescent intensity determination. The results is compared with the data obtained from measurements of free lines of spectral interference and also with those obtained by the linear equation system. Fission products determination presented a precision in the range of 0.1 to 5.0% and an accuracy of up to ± 7.0% the results are compared with those obtained by neutron activation analysis and inductively coupled plasma - atomic emission spectrometry. Leaching data, when radioactive waste is incorporated in cement matrix, were attempted by X-ray fluorescence technique. For two years leaching period, leaching rate and diffusion coefficient data of cesium were determined. The results obtained agree with those obtained by γ-spectromety. (author) [pt

  5. Process for final storage of high level radioactive fission product solution

    International Nuclear Information System (INIS)

    Ellis, J.R.B.; Fries, B.A.

    1984-01-01

    In this process for the storage of an aqueous solution of radioactive nuclides, the solution is diluted with system water, which is obtained from a reservoir below the bottom of the sea in a porous geological stratum. After dilution, the diluted solution is injected into the same geological stratum under the bottom of the sea. (orig.) [de

  6. 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

  7. Standard test method for gamma energy emission from fission products in uranium hexafluoride and uranyl nitrate solution

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2005-01-01

    1.1 This test method covers the measurement of gamma energy emitted from fission products in uranium hexafluoride (UF6) and uranyl nitrate solution. It is intended to provide a method for demonstrating compliance with UF6 specifications C 787 and C 996 and uranyl nitrate specification C 788. 1.2 The lower limit of detection is 5000 MeV Bq/kg (MeV/kg per second) of uranium and is the square root of the sum of the squares of the individual reporting limits of the nuclides to be measured. The limit of detection was determined on a pure, aged natural uranium (ANU) solution. The value is dependent upon detector efficiency and background. 1.3 The nuclides to be measured are106Ru/ 106Rh, 103Ru,137Cs, 144Ce, 144Pr, 141Ce, 95Zr, 95Nb, and 125Sb. Other gamma energy-emitting fission nuclides present in the spectrum at detectable levels should be identified and quantified as required by the data quality objectives. 1.4 This standard does not purport to address all of the safety concerns, if any, associated with its us...

  8. Experiments for separation and purification of Mo-99 from uranium solution with fission products as tracers

    International Nuclear Information System (INIS)

    Androne, Gabriela Elena; Matei, Ana

    2007-01-01

    Technetium-99 m ( 99m Tc, T 1/2 = 6 hours), one of the most utilised radioisotopes in nuclear medicine, is generated through the beta decay of 99 Mo (T 1/2 = 66 hours) and which will decay through isomer transition to 99 Tc ( T 1/2 = 2.10 5 years) through the emission of a gamma radiation with the energy of 0.140 MeV. The work presents the phases of the process of Mo separation and purification at a tracer level. The tests performed in the laboratory have established the optimum conditions for the separation and purification of Molybdenum. To establish the separation and purification parameters, a synthetic solution which contains the elements which result following the irradiation of a low enriched Uranium foil weighing 10 g (∼20 % 235 U). To mark this solution, about 13 mg of UO 2 10% 235 U was irradiated for 2000 s, at a flux of about 7x10 12 n/cm 2 s. This amount of UO 2 will be added to the above-mentioned solution after dissolution. The method for separating the Molybdenum from irradiated Uranium solution is one of selective precipitation of Mo with α-benzoin-oxyme (α- BO). To purify the Molybdenum solution, two purification columns were utilised. Their role was to absorb the impurities remained in the mass of the precipitate. They and the Molybdenum have passed into the solution simultaneously, allowing the Molybdenum to pass. These columns are: the column with active charcoal (AC)+ active charcoal covered with silver (AgAC); active charcoal column (AC)+active charcoal covered with silver (AgAC)+hydrated zirconium oxide (HZO). Although all the phases of the process are performed with high yields, the final yields of recovery of Mo from U solutions are higher than 80%. (authors)

  9. A Study on the Oxidative-dissolution Leaching of Fission Product Oxides in the carbonate solution

    International Nuclear Information System (INIS)

    Lee, Eil Hee; Kim, Kwang Wook; Lim, Jae Gwan; Chung, Dong Yong; Yang, Han Beom; Joe, Kih Soo; Seo, Heui Seung; Kim, Yeon Hwa; Lee, Se Yoon

    2009-07-01

    This study was carried out to investigate the characteristics of an oxidativedissolution leaching of FP co-dissolved with U in a carbonate solution of Na 2 CO 3 - H 2 O 2 and (NH 4 ) 2 CO 3 -H 2 O 2 , respectively. Simulated FP-oxides which contained 12 components have been added to the solution to examine their oxidative dissolution characteristics. It was found that H 2 O 2 was an effective oxidant to minimize the dissolution of FP in a carbonate solution. In 0.5M Na 2 CO 3 -0.5M H 2 O 2 and 0.5M (NH 4 ) 2 CO 3 -0.5M H 2 O 2 solution, some elements such as Re, Te, Cs and Mo seem to be dissolved together with U. It is revealed that dissolution rates of Re, Te and Cs are high (completely dissolved within 10∼20 minutes) due to their high solubility in Na 2 CO 3 and (NH 4 ) 2 CO 3 solution regardless of the addition of H 2 O 2 , and independent of the concentrations of Na 2 CO 3 and H 2 O 2 . However, Mo was slowly dissolved by an oxidative dissolution with H 2 O 2 . It is found that the most important factor for the oxidative dissolution of FP is the pH of the solution and an effective oxidative dissolution is achieved at a pH between 10∼12 for Na 2 CO 3 and a pH between 9∼10 for (NH 4 ) 2 CO 3 , respectively, in order to minimize the dissolution of FP

  10. 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

  11. Sorption Regularities in Behaviour of Fission-Product Elements during Filtration of their Solutions through Ground

    Energy Technology Data Exchange (ETDEWEB)

    Spitsyn, V. I.; Balukova, V. D.; Gromov, V. V.; Zakharov, S. I.; Zhagin, B. P.; Spiridonov, F. M.

    1960-07-01

    Research on the sorption of radioisotopes under natural conditions employing the controlled filtration process was performed. Radioisotopes were introduced into the solution as soon as filtration had become steady and the process continued for four months. Soil samples were then taken by drilling at different depths and analysed to determine their radioisotope content. Diffusion of radioisotopes was observed at depths of up to 10 m; two distinct boundaries of soil-activity decrease were ascertained: at the surface of the site and at the depth of the solution filtration front. In addition, the radiostrontium absorption by natural sorbents, principally pure minerals widely distributed in soils and subsoils, was investigated separately. The presence of calcium ions, even in small quantities, sharply reduces the degree of radiostrontium sorption. However, other conditions being equal, strontium may be absorbed to a greater extent than calcium, according to the composition of the sorbent. The field investigations of radiostrontium sorption and migration showed that when filtering radioactive solutions two possible variants have to be taken into account. In the first case the solutions are discharged into soil unaffected by any flow of ground water. In this situation the radiostrontium is retained by the soil. In the second case, the radioisotopes proceed directly into the water-bearing horizon. The radiostrontium will then migrate with the ground water flow and through the soil and this migration will be further affected by the sorption and desorption processes occurring. The experiments performed demonstrate the ease with which long-lived radioisotopes migrate under natural conditions and call attention to the need for thorough study of ground water problems in connexion with various methods of disposing of radioactive waste into ground. (author)

  12. 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

  13. Industrial treatment of solutions of fission products. Separation of caesium-137; Traitement industriel de solutions de produits de fission. Separation du cesium-137

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, C; Raggenbass, A [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    Two types of chemical treatment can be considered for the manufacture of solid sources for industrial uses from fission product solutions remaining after plutonium extraction: a) concentration of the solution and preparation of solid sources from the bulk material, without separation, b) separation of one or several fission products from which the sources are made. Examination of the radio-chemical composition of the mixture of fission products that will be available from the Marcoule reactors (G1, G2 and G3) shows that caesium-137 accounts for 30 per cent of the {gamma} energy available immediately after the plutonium separation, 70 per cent two years after and 100 per cent after five years. There is practically no advantage in making sources from bulk fission products, since the separation of caesium-137 is no more complicated and yet it results in a material with more potential uses. The separation of caesium-137 by a method based on the standard phospho-tungstate precipitation method has been considered. Previously, the precipitated caesium phospho-tungstate was dissolved and caesium was recovered from the solution by cation-exchange or by removal of phosphate and tungstate ions by anion-exchange. A study has now been made, of the metathesis of caesium phospho-tungstate to barium phosphate and tungstate by the action of barium hydroxide, the caesium being obtained in solution as the hydroxide. The advantages of this new procedure are: - greater decontamination of caesium-137 without further purification, - possibility of direct transformation to caesium sulphate, - general simplification of the procedure and, consequently, of the equipment. (author) [French] Deux types de traitement chimique peuvent etre envisages pour amener a l'etat de sources solides utilisables industriellement les produits de fission contenus dans les solutions residuaires de l'extraction du plutonium. Ces traitements sont les suivants: a) concentration des solutions et confection de

  14. Industrial treatment of solutions of fission products. Separation of caesium-137; Traitement industriel de solutions de produits de fission. Separation du cesium-137

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, C.; Raggenbass, A. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    Two types of chemical treatment can be considered for the manufacture of solid sources for industrial uses from fission product solutions remaining after plutonium extraction: a) concentration of the solution and preparation of solid sources from the bulk material, without separation, b) separation of one or several fission products from which the sources are made. Examination of the radio-chemical composition of the mixture of fission products that will be available from the Marcoule reactors (G1, G2 and G3) shows that caesium-137 accounts for 30 per cent of the {gamma} energy available immediately after the plutonium separation, 70 per cent two years after and 100 per cent after five years. There is practically no advantage in making sources from bulk fission products, since the separation of caesium-137 is no more complicated and yet it results in a material with more potential uses. The separation of caesium-137 by a method based on the standard phospho-tungstate precipitation method has been considered. Previously, the precipitated caesium phospho-tungstate was dissolved and caesium was recovered from the solution by cation-exchange or by removal of phosphate and tungstate ions by anion-exchange. A study has now been made, of the metathesis of caesium phospho-tungstate to barium phosphate and tungstate by the action of barium hydroxide, the caesium being obtained in solution as the hydroxide. The advantages of this new procedure are: - greater decontamination of caesium-137 without further purification, - possibility of direct transformation to caesium sulphate, - general simplification of the procedure and, consequently, of the equipment. (author) [French] Deux types de traitement chimique peuvent etre envisages pour amener a l'etat de sources solides utilisables industriellement les produits de fission contenus dans les solutions residuaires de l'extraction du plutonium. Ces traitements sont les suivants: a) concentration des solutions et

  15. An analytic solution to the time-dependent first-daughter fission-product plateout problem for multi-region isothermal slug flow

    International Nuclear Information System (INIS)

    Durkee, J.W. Jr.; Lee, C.E.

    1985-01-01

    The time-dependent, axisymmetric, isothermal slug flow convective-diffusion equation with radioactive decay is solved analytically to predict the behavior of a first-daughter fission-product undergoing gaseous transport through multiple materials in a cylindrical pipe. The integration coefficients are determined using the Davidon variable metric minimization method. The behavior of fission-product material deposited on the conduit wall is described by a standard mass-transfer model. The time-dependent plateout rate behavior, determined previously for parent fission-product deposition, is again evident for daughter product plateout. Dominance of the daughter plateout by parent deposition characteristics is apparent. The determination of the daughter wall mass-transfer and diffusion coefficient using a least-squares analysis of measured data depends upon a reasonably low ratio of parent/daughter half-lives. This is illustrated with 137 Cs/ 137 Ba(=2x10 5 ) and 140 Ba/ 140 La(=7.6), where for 137 Cs/ 137 Ba the solution sensitivity to the 137 Ba deposition parameters is small and for 140 Ba/ 140 La a reasonable solution is readily obtained. (author)

  16. Fission product detection

    International Nuclear Information System (INIS)

    Liatard, E.; Akrouf, S.; Bruandet, J.F

    1987-01-01

    The response of photovoltaic cells to heavy ions and fission products have been tested on beam. Their main advantages are their extremely low price, their low sensitivity to energetic light ions with respect to fission products, and the possibility to cut and fit them together to any shape without dead zone. The time output signals of a charge sensitive preamplifier connected to these cells allows fast coincidences. A resolution of 12ns (F.W.H.M.) have been measured between two cells [fr

  17. Fission Product Library and Resource

    Energy Technology Data Exchange (ETDEWEB)

    Burke, J. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Padgett, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-09-29

    Fission product yields can be extracted from an irradiated sample by performing gamma ray spectroscopy on the whole sample post irradiation. There are several pitfalls to avoid when trying to determine a specific isotope's fission product yield.

  18. Fission products in glasses. Pt. 2

    International Nuclear Information System (INIS)

    De, A.K.; Luckscheiter, B.; Malow, G.; Schiewer, E.

    1977-09-01

    Glass ceramics of different composition with high leach and impact resistance can be produced for fission product solidification. In contrast to commercial glass products, they consist of a number of crystalline phases and a residual glass phase. The major crystalline phase allows a classification into celsian, diopside, encryptite, and perovskite ceramics. They all are of special importance as host phases for long-lived fission products. The paper reports on relations between product composition and melting properties, viscosity, crystallization properties, and fixation capability for fission products. Further investigations deal with dimensional stability, impact resistance, thermal expansion, and thermal conductivity. The properties of the ceramics are compared with those of the basic products. The problems still to be solved with regard to further improvement and application of these products are discussed. (RB) [de

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  20. 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)

  1. Chemistry of actinides and fission products

    International Nuclear Information System (INIS)

    Pruett, D.J.; Sherrow, S.A.; Toth, L.M.

    1988-01-01

    This task is concerned primarily with the fundamental chemistry of the actinide and fission product elements. Special efforts are made to develop research programs in collaboration with researchers at universities and in industry who have need of national laboratory facilities. Specific areas currently under investigation include: (1) spectroscopy and photochemistry of actinides in low-temperature matrices; (2) small-angle scattering studies of hydrous actinide and fission product polymers in aqueous and nonaqueous solvents; (3) kinetic and thermodynamic studies of complexation reactions in aqueous and nonaqueous solutions; and (4) the development of inorganic ion exchange materials for actinide and lanthanide separations. Recent results from work in these areas are summarized here

  2. Chemical Separation of Fission Products in Uranium Metal Ingots from Electrolytic Reduction Process

    International Nuclear Information System (INIS)

    Lee, Chang-Heon; Kim, Min-Jae; Choi, Kwang-Soon; Jee, Kwang-Yong; Kim, Won-Ho

    2006-01-01

    Chemical characterization of various process materials is required for the optimization of the electrolytic reduction process in which uranium dioxide, a matrix of spent PWR fuels, is electrolytically reduced to uranium metal in a medium of LiCl-Li 2 O molten at 650 .deg. C. In the uranium metal ingots of interest in this study, residual process materials and corrosion products as well as fission products are involved to some extent, which further adds difficulties to the determination of trace fission products. Besides it, direct inductively coupled plasma atomic emission spectrometric (ICP-AES) analysis of uranium bearing materials such as the uranium metal ingots is not possible because a severe spectral interference is found in the intensely complex atomic emission spectra of uranium. Thus an adequate separation procedure for the fission products should be employed prior to their determinations. In present study ion exchange and extraction chromatographic methods were adopted for selective separation of the fission products from residual process materials, corrosion products and uranium matrix. The sorption behaviour of anion and tri-nbutylphosphate (TBP) extraction chromatographic resins for the metals in acidic solutions simulated for the uranium metal ingot solutions was investigated. Then the validity of the separation procedure for its reliability and applicability was evaluated by measuring recoveries of the metals added

  3. Radiochemical applications of insoluble sulfate columns. Analytical possibilities in the field of the fission product solutions; Aplicaciones radioquimica de las columnas de precipitados de sulfatos insolubles. Contribucion al estudio de las soluciones envejecidas de productos de fision

    Energy Technology Data Exchange (ETDEWEB)

    Barrachina, M; Sauvagnac, R

    1962-07-01

    In this paper we go on with our study of the heterogeneous ion-isotopic exchange in column. At present, we apply it to determine the radiochemical composition of the raw solutions used in the industrial recuperation of the long-lived fission products. The separation of the radioelements contained in these solutions is carried out mainly by making use of small columns, 1-3 cm height, of BaSO{sub 4} or SrSO{sub 4}, under selected experimental conditions. These columns behave like a special type of inorganic exchangers, working by absorption or by ion-isotopic exchange depending on the cases,a nd they provide the means for the selective separation of several important fission products employing very small volumes of fixing and eluting solutions. (Author) 11 refs.

  4. 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

  5. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1984-09-01

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

  6. Review of areas that may require simultaneous coupled solution of the thermal hydraulic and fission product/aerosol behavior equations for source term determination

    International Nuclear Information System (INIS)

    Kress, T.S.

    1984-01-01

    In the determination of the behavior of nuclear aerosols in the reactor coolant system and in the containment for the development of severe accident source terms, present practice generally is to first perform thermal hydraulic calculations for specific plant types and sequences and then to utilize the results as input for separate fission product/aerosol dynamic transport calculations. It is recognized that there are several areas in which the thermal-hydraulics and the fission product/aerosol behavior may be significantly coupled and that it is then basically incorrect to do the analyses in a separated manner. This review paper produces a speculative list of these potentially coupled areas and attempts to assess the importance of the coupling for as many of the specific items that time has allowed before this conference

  7. Separation of short-lived fission products

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  8. HAMCIND, Cell Burnup with Fission Products Poisoning

    International Nuclear Information System (INIS)

    Abe, Alfredo Y.; Dos Santos, Adimir

    2002-01-01

    1 - Description of program or function: HAMCIND is a cell burnup code based in a coupling between HAMMER-TECHNION and CINDER. The fission product poisoning is taken into account in an explicit fashion. 2 - Method of solution: The nonlinear coupled set of equations for the neutron transport and nuclide transmutation equations and nuclide transmutation equations in a unit cell is solved by HAMCIND in a quasi-static approach. The spectral transport equation is solved by HAMMER-TECHNION at the beginning of each time-step while the nuclide transmutation equations are solved by CINDER for every time-step. The HAMMER-TECHNION spectral calculations are performed taking into account the fission product contribution to the macroscopic cross sections (fast and thermal), in the inelastic scattering matrix and even in the thermal scattering matrices. 3 - Restrictions on the complexity of the problem: Restrictions and/or limitations for HAMCIND depend upon the local operating system

  9. 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)

  10. Application of mercury cathode electrolysis to fission-product separation

    International Nuclear Information System (INIS)

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

    1969-01-01

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

  11. Separation of caesium-137 from fission products using phosphotungstic acid

    International Nuclear Information System (INIS)

    Murthy, T.S.; Balasubramaniam, K.R.; Ananthakrishnan, M.; Varma, R.N.

    1977-01-01

    Separation of caesium 137 from fission products using phosphotungstic acid is reported. Phosphotungstate caesium is precipitated as caesium from fission product waste solution in acid medium and subsequently purified. Separation of phosphate and tungstate ions has been done using a typical hydrous oxide like alumina. The exchange capacity of alumina for phosphate and tungstate ions, and the purity of the product are determined. Results are discussed. Based on the findings a procedure is recommended for caesium 137 separation. (A.K.)

  12. SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System

    International Nuclear Information System (INIS)

    Kodaira, Hideki

    1990-01-01

    1 - Description of program or function: SACHET evaluates the dynamic fission product inventories in the multiple compartment system of pressurized water reactor (PWR) plants. 2 - Method of solution: SACHET utilizes a matrix of fission product core inventory which is previously calculated by the ORIGEN code. 3 - Restrictions on the complexity of the problem: Liquid wastes such as chemical waste and detergent waste are not included

  13. NEACRP thermal fission product benchmark

    International Nuclear Information System (INIS)

    Halsall, M.J.; Taubman, C.J.

    1989-09-01

    The objective of the thermal fission product benchmark was to compare the range of fission product data in use at the present time. A simple homogeneous problem was set with 200 atoms H/1 atom U235, to be burnt up to 1000 days and then decay for 1000 days. The problem was repeated with 200 atoms H/1 atom Pu239, 20 atoms H/1 atom U235 and 20 atoms H/1 atom Pu239. There were ten participants and the submissions received are detailed in this report. (author)

  14. RSAC, Gamma Doses, Inhalation and Ingestion Doses, Fission Products Inventory after Fission Products Release

    International Nuclear Information System (INIS)

    Richardson, L.C.

    1967-01-01

    1 - Description of problem or function: RSAC generates a fission product inventory from a given set of reactor operating conditions and then computes the external gamma dose, the deposition gamma dose, and the inhalation-ingestion dose to critical body organs as a result of exposure to these fission products. Program output includes reactor operating history, fission product inventory, dosages, and ingestion parameters. 2 - Method of solution: The fission product inventory generated by the reactor operating conditions and the inventory remaining at various times after release are computed using the equations of W. Rubinson in Journal of Chemical Physics, Vol. 17, pages 542-547, June 1949. The external gamma dose and the deposition gamma dose are calculated by determining disintegration rates as a function of space and time, then integrating using Hermite's numerical techniques for the spatial dependence. The inhalation-ingestion dose is determined by the type and quantity of activity inhaled and the biological rate of decay following inhalation. These quantities are integrated with respect to time to obtain the dosage. The ingestion dose is related to the inhalation dose by an input constant

  15. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, G.

    1976-05-01

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

  16. Fission products stability in uranium dioxide

    International Nuclear Information System (INIS)

    Brillant, G.; Gupta, F.; Pasturel, A.

    2011-01-01

    Fission product stability in nuclear fuels is investigated using density functional theory (DFT). In particular, incorporation and solution energies of He, Kr, Xe, I, Te, Ru, Sr and Ce in pre-existing trap sites of UO 2 (vacancies, interstitials, U-O divacancy, and Schottky trio defects) are calculated using the projector-augmented-wave method as implemented in the Vienna ab initio simulation package. Correlation effects are taken into account within the DFT+U approach. The stability of many binary and ternary compounds in comparison to soluted atoms is also explored. Finally the involvement of FP in the formation of metallic and oxide precipitates in oxide fuels is discussed in the light of experimental results.

  17. Device for measuring fission product density

    International Nuclear Information System (INIS)

    Kaneda, Mitsunori.

    1980-01-01

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

  18. Fission product behaviour in severe accidents

    International Nuclear Information System (INIS)

    Jokiniemi, J.; Auvinen, A.; Maekynen, J.; Valmari, T.

    1998-01-01

    The understanding of fission product (FP) behaviour in severe accidents is important for source term assessment and accident mitigation measures. For example in accident management the operator needs to know the effect of different actions on the behaviour and release of fission products. At VTT fission product behaviour have been studied in different national and international projects. In this presentation the results of projects in EU funded 4th framework programme Nuclear Fission Safety 1994-1998 are reported. The projects are: fission product vapour/aerosol chemistry in the primary circuit (FI4SCT960020), aerosol physics in containment (FI4SCT950016), revaporisation of test samples from Phebus fission products (FI4SCT960019) and assessment of models for fission product revaporisation (FI4SCT960044). Also results from the national project 'aerosol experiments in the Victoria facility' funded by IVO PE and VTT Energy are reported

  19. Transmutation of fission products and actinide waste at Hanford

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-10-01

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

  20. Attachment of gaseous fission products to aerosols

    International Nuclear Information System (INIS)

    Skyrme, G.

    1985-01-01

    Accidents may occur in which the integrity of fuel cladding is breached and volatile fission products are released to the containment atmosphere. In order to assess the magnitude of the subsequent radiological hazard it is necessary to know the transport behaviour of such fission products. It is frequently assumed that the fission products remain in the gaseous phase. There is a possibility, however, that they may attach themselves to particles and hence substantially modify their transport properties. This paper provides a theoretical assessment of the conditions under which gaseous fission products may be attached to aerosol particles. Specific topics discussed are: the mass transfer of a gaseous fission product to an isolated aerosol particle in an infinite medium; the rate at which the concentration of fission products in the gas phase diminishes within a container as a result of deposition on a population of particles; and the distribution of deposited fission product between different particle sizes in a log-normal distribution. It is shown that, for a given mass, small particles are more efficient for fission product attachment, and that only small concentrations of such particles may be necessary to achieve rapid attachment. Conditions under which gaseous fission products are not attached to particles are also considered, viz, the competing processes of deposition onto the containment walls and onto aerosol particles, and the possibility of the removal of aerosols from the containment by various deposition processes, or agglomeration, before attachment takes place. (author)

  1. Downstream behavior of fission products

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  2. Aerosols and fission product transport

    International Nuclear Information System (INIS)

    Megaw, W.J.

    1987-12-01

    A survey is presented of current knowledge of the possible role of aerosols in the consequences of in- and out-of-core LOCAs and of end fitting failures in CANDU reactors. An extensive literature search has been made of research on the behaviour of aerosols in possible accidents in water moderated and cooled reactors and the results of various studies compared. It is recommended that further work should be undertaken on the formation of aerosols during these possible accidents and to study their subsequent behaviour. It is also recommended that the fission products behaviour computer code FISSCON II should be re-examined to determine whether it reflects the advances incorporated in other codes developed for light water reactors which have been extensively compared. 47 refs

  3. Actual point about fission products vitrification

    International Nuclear Information System (INIS)

    Bonniaud, R.

    1982-05-01

    The main characteristics concerning the continuous vitrification process for the confinement of fission product solutions operated at AVM are summarized. The general principle of a vitrification plant is described. The AVM plant efficiency as also its conception of consumable parts interchangeability are satisfying. The evolution of the process and its application developped in two ways: a more spaced installation conception and the improvement of the weak points remarked at AVM, as also the capacity of output. Two industrial units are designed at La Hague. The future evolution of the process aims at manufacturing glass at higher temperatures about 1400 degrees Celsius. Some problems remain to be resolved for the using of ceramic melters associated with a calcination unit. The studies provide for a satisfying behaviour for the material to long-term. The risks of damage by crystallisation, leaching and effects of alpha emission are analysed [fr

  4. 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

  5. The present state of research on the vitrification of concentrated solutions of fission products (1962); Etat des etudes sur la vitrification des solutions concentrees de produits de fission (1962)

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-07-01

    The present report gives the actual point of studies on vitrification of concentrated solutions of fission products. An active cell, giving glasses in crucibles, permitted to study various glass compositions. The leaching rate from the glass raises 1 to 2 10{sup -7} g of glass/cm{sup 2}/day. Activity loss by volatility during vitrification remains weak and often below 0.1 per cent of total activity. Off gas cleaning is made easier by presence of filter which is compound of granules including iron oxide. After saturation the content of this filter can be melt. Moreover different processes are in experimentation for a more important production. Daily 72 liters of solution containing tracer activity are treated in a continuous calcination and vitrification plant. The loss in {sup 106}Ru is still important and a modification of installation has been necessary. A pot vitrification plant is in study. In order to reduce cost of processing the possibility to pour glass after melting is actuality in study. A production set of very active glass is also in project. (authors) [French] Le present rapport fait le point des etudes menees sur la vitrification des solutions concentrees de produits de fission. Une installation active, produisant des verres en creusets, a permis d'etudier plusieurs compositions de verres. Le taux de perte d'activite par lixiviation a l'eau atteint 1 a 2 10{sup -7} gramme de verre/cm{sup 2}/jour. Les pertes d'activite par volatilite au cours de la cuisson restent faibles et souvent inferieures a 0,1 pour cent de l'activite totale. L'epuration des gaz de cuisson est facilitee par la presence d'un filtre a granules riches en oxyde de fer, dont le contenu peut etre fondu apres saturation. Differentes techniques sont, en outre, en experimentation pour une production plus importante: Une installation de calcination et vitrification continue traite 72 litres par jour de solution contenant une activite traceur. La perte en Ru{sup 106} est encore importante

  6. Experiments on the incorporation of concentrated solutions of fission products in glasses and micas; Essais d'incorporation de solutions concentrees de produits de fission dans des verres et des micas

    Energy Technology Data Exchange (ETDEWEB)

    Bonniaud, R; Cohen, P; Sombret, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    1) The plants designed for extracting the plutonium of the G1, G2, G3 reactors fuel rods will produce extremely concentrated tions of fission products. 2) Let us consider a synthetic solution of the 'G2' type: (2N NO{sub 3}H - 2,25 meq/cm{sup 3} AI{sup +++} - 5.10{sup -3} {sup 134}Cs - {sup 137}Cs tracer). We made various glasses and micas by adding and mixing the necessary adjuvants and baking (900 to 1400 deg. C) in a graphite crucible. 3) The products obtained had either the shape of a cylindrical pellet or were reduced into a fine powder. They were mixed with 300 cm{sup 3} of synthetic sea water during variable periods of time in order to study leaching of the activity. 4) Experiments were first carried on caesium because of its solubility. 5) Fabrication of micas on a large scale sets many technological problems more difficult to solve (1400 deg. C) than in the case of glasses (1000 deg. C). A comparative study on both micas and glasses showed that leaching of activity was more important in the micas. (author)Fren. [French] 1) La mise en service de l'Usine d'Extraction du Plutonium destinee en particulier au traitement des barres de combustibles de G1, G2, G3 fournira des solutions residuaires concentrees de produits de fission. 2) Partant d'une solution synthetique 'G2': (2N NO{sub 3}H - 2,25meq/cm{sup 3} AI{sup +++} - 5.10{sup -3} {sup 134}Cs - {sup 137}Cs traceur) nous avens realise differents verres et micas apres adduction et melange des adjuvants necessaires suivis de cuisson (900 a 1400 deg. C) dans un creuset de graphite. 3) Les produits obtenus ont ete, soit sous forme d'eprouvette cylindrique, soit sous forme de poudre finement broyee, mis en contact et agites avec 300 cm{sup 3} d'eau de mer synthetique, durant des temps varies pour etudier la retention de l'activite. 4) Le cesium a ete choisi pour ces essais a cause de sa solubilite. 5) La fabrication de micas a grande echelle pose des problemes technologiques plus difficiles (1400 deg. C) a

  7. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, G.

    1975-01-01

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

  8. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1981-06-01

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

  9. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1982-07-01

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

  10. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1983-08-01

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

  11. Treatment of solutions of fission products - Separation of caesium-137; Traitement des solutions de produits de fission - Separation du cesium-137; Obrabotka rastvorov produktov deleniya - Razdelenie tseziya-137; Tratamiento de soluciones de productos de fision - Separacion del cesio-137

    Energy Technology Data Exchange (ETDEWEB)

    Stiennon-Bovy, R [Centre d' Etude de l' Energie Nucleaire, Mol (Belgium); Gvion, R [Commission Israelienne de l' Energie Atomique, Tel-Aviv (Israel)

    1962-01-15

    For the industrial recovery of caesium-137 from solutions of fission products, the authors utilized the analytical method for determination of caesium by dipicrylamine, adapting it to use on an industrial scale and to the high level of the activities encountered. The process recommended makes it possible both to isolate caesium as a chloride and to recover the precipitation reagent, in one and the same operation. A basic method is suggested. The authors studied the effect of radiation on dipicrylamine and its compounds, this effect proving to be practically nil for solid compounds and negligible for their solutions. The entrainment of caesium by ammonia ion was also studied. The advantages of the proposed process are : high decontamination of the caesium, simple operation and free recycling fo the reagent, high yield for caesium recovery and for dipicrylamine, considerable concentration of caesium activity, operation at room temperature and possibility of continuous operation. By this process caesium can be recovered before certain fission products are eliminated. (author) [French] Pour la recuperation industrielle du cesium-137 des solutions de produits de fission, les auteurs ont adapte le procede analytique de dosage du cesium par la dipicrylamine a l'echelle industrielle et au niveau eleve des activites traitees. Le procede preconise permet en une seule operation d'isoler le cesium sous forme de chlorure et de recuperer aisement le reactif de precipitation; un schema de principe est propose. Les auteurs ont etudie l'effet des radiations sur la dipicrylamine et ses composes; cet effet s'avere pratiquement nul pour les composes solides et negligeable pour leurs solutions. L'entrainement du cesium par l'ion ammoniaque est etudie. Les avantages du procede propose sont la decontamination elevee du cesium, la simplicite du mode operatoire avec recyclage aise du reactif, le rendement eleve pour la recuperation du cesium et pour la dipicrylamine, une concentration

  12. 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)

  13. Evaluation of six decontamination processes on actinide and fission product contamination

    International Nuclear Information System (INIS)

    Conner, C.; Chamberlain, D.B.; Chen, L.

    1995-01-01

    In-situ decontamination technologies were evaluated for their ability to: (1) reduce equipment contamination levels to allow either free release of the equipment or land disposal, (2) minimize residues generated by decontamination, and (3) generate residues that are compatible with existing disposal technologies. Six decontamination processes were selected. tested and compared to 4M nitric acid, a traditional decontamination agent: fluoroboric acid (HBF 4 ), nitric plus hydrofluoric acid, alkaline persulfate followed by citric acid plus oxalic acid, silver(II) plus sodium persulfate plus nitric acid, oxalic acid plus hydrogen peroxide plus hydrofluoric acid, and electropolishing using nitric acid electrolyte. The effectiveness of these solutions was tested using prepared 304 stainless steel couponds contaminated with uranium, plutonium, americium, or fission products. The decontamination factor for each of the solutions and tests conditions were determined; the results of these experiments are presented

  14. Contribution to research on the metabolism of fission product. Studies on the physico-chemical state and the metabolic fate of radio-cerium solution

    International Nuclear Information System (INIS)

    Aeberhardt, A.

    1961-01-01

    This paper describes a study of the physico-chemical state of radio-cerium in dilute solutions on the tracer scale, as a function of the pH of the solution. The way in which this radioelement is transported in the blood is studied in vitro and in vivo, with reference to the ionic or colloidal state of the radio-cerium used. The distribution of cerium amongst the various components of the blood is studied by a new method of blood fractionation and by paper electrophoresis. Evidence of a cerium globulin connection is shown in the case of ionic cerium. A study of the initial distribution of radio-cerium in rats, after intravenous administration of ionic or colloidal solutions, shows considerable differences according to the physico-chemical state of the cerium injected. (author) [fr

  15. 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

  16. 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

  17. Transport properties of fission product vapors

    International Nuclear Information System (INIS)

    Im, K.H.; Ahluwalia, R.K.

    1983-07-01

    Kinetic theory of gases is used to calculate the transport properties of fission product vapors in a steam and hydrogen environment. Provided in tabular form is diffusivity of steam and hydrogen, viscosity and thermal conductivity of the gaseous mixture, and diffusivity of cesium iodide, cesium hydroxide, diatomic tellurium and tellurium dioxide. These transport properties are required in determining the thermal-hydraulics of and fission product transport in light water reactors

  18. Modeling Fission Product Sorption in Graphite Structures

    International Nuclear Information System (INIS)

    Szlufarska, Izabela; Morgan, Dane; Allen, Todd

    2013-01-01

    The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high-temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission products

  19. 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

  20. Quantitative analysis of fission products by γ spectrography

    International Nuclear Information System (INIS)

    Malet, G.

    1962-01-01

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

  1. Leaching behavior of gamma-emitting fission products and Np from neutron-irradiated UO_2-ZrO_2 solid solutions in non-filtered surface seawater

    International Nuclear Information System (INIS)

    Sasaki, Takayuki; Takeno, Yuu; Kobayashi, Taishi; Kirishima, Akira; Sato, Nobuaki

    2016-01-01

    The gamma ray radionuclides Cs-137, Ba-140, I-131, Ce-141, Ru-103, Zr-95, and Np-239 were produced by neutron irradiation of UO_2-ZrO_2 solid solutions that were synthesized as simulated fuel debris under reducing and oxidizing conditions. The leaching ratio of radionuclides was investigated under atmospheric conditions at 25°C for non-filtered natural surface seawater, as well as deionized water after filtration with a membrane of 0.45-μm pore size or that of nominal molecular weight limit of 3 kDa. The uranium molar concentration was affected by the oxidation state in the solid solution samples. The congruent dissolution of Cs, I, and Ba with the hexavalent uranium of U_3O_8 was facilitated in the seawater samples, whereas a lower leaching ratio of nuclides was observed in the deionized water samples. Neptunium-239, originally produced from uranium-238 in U_3O_8, showed behavior that was similar to that of Cs, I, and Ba. However, the dissolution of Np (as a parent nuclide of Pu-239) in the debris of UO_2 and UO_2-ZrO_2 was suppressed in the same manner as Zr(IV) and Ce(IV). The concentration exhibited no filtration dependence after 15 d, which shows that most of the leached nuclides can exist in their ionic form in seawater. (author)

  2. Determination of {sup 90}Sr in uranium fission products

    Energy Technology Data Exchange (ETDEWEB)

    Bajo, S; Tobler, L [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-02-01

    A previously published radiochemical procedure for the determination of {sup 90}Sr in grass and soil has been successfully employed - with minor modifications - for the determination of this nuclide in a solution of uranium fission products. It is suitable for the determination of {sup 90}Sr in environmental materials following a nuclear accident. The procedure is based on tributylphosphate extraction of {sup 90}Y, precipitation of Y-oxalate, and counting in a proportional counter. (author) figs., tabs., 10 refs.

  3. Precipitation behaviors of Cs and Re(/Tc) by NaTPB and TPPCl from a simulated fission products-(Na2CO3-NaHCO3)-H2O2 solution

    International Nuclear Information System (INIS)

    Lee, Eil Hee; Lim, Jae Gwan; Chung, Dong Yong; Yang, Han Beum; Kim, Kwang Wook

    2010-01-01

    In this study, the removal of Cs and Tc from a simulated fission products (FP) solution which were co-dissolved with U during the oxidative-dissolution of spent fuel in a mixed carbonate solution of (Na 2 CO 3 -NaHCO 3 )-H 2 O 2 was investigated by using a selective precipitation method. As Cs and Tc might cause an unstable behavior due to the high decay heat emission of Cs as well as the fast migration of Tc when disposed of underground, it is one of the important issues to removal them in views of the increase of disposal safety. The precipitation of Cs and Re (as a surrogate for Tc) was examined by introducing sodium tetraphenylborate (NaTPB) and tetraphenylphosponium chloride (TPPCl), respectively. Precipitation of Cs by NaTPB and that of Re by TPPCl were completed within 5 minutes. Their precipitation rates were not influenced so much by the temperature and stirring speed even if they were increased by up to 50 .deg. C and 1,000 rpm. However, the pH of the solution was found to have a great influence on the precipitation with NaTPB and TPPCl. Since Mo tends to co-precipitate with Re at a lower pH, especially, it was effective that a selective precipitation of Re by TPPCl was carried out at pH of above 9 without co-precipitation of Mo and Re. Over 99% of Cs was precipitated when the ratio of [NaTPB]/[Cs]>1 and more than 99% of Re, likewise, was precipitated when the ratio of [TPPCl]/[Re]>1.

  4. Fission Product Sorptivity in Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Tompson, Jr., Robert V. [Univ. of Missouri, Columbia, MO (United States); Loyalka, Sudarshan [Univ. of Missouri, Columbia, MO (United States); Ghosh, Tushar [Univ. of Missouri, Columbia, MO (United States); Viswanath, Dabir [Univ. of Missouri, Columbia, MO (United States); Walton, Kyle [Univ. of Missouri, Columbia, MO (United States); Haffner, Robert [Univ. of Missouri, Columbia, MO (United States)

    2015-04-01

    Both adsorption and absorption (sorption) of fission product (FP) gases on/into graphite are issues of interest in very high temperature reactors (VHTRs). In the original proposal, we proposed to use packed beds of graphite particles to measure sorption at a variety of temperatures and to use an electrodynamic balance (EDB) to measure sorption onto single graphite particles (a few μm in diameter) at room temperature. The use of packed beds at elevated temperature is not an issue. However, the TPOC requested revision of this initial proposal to included single particle measurements at elevated temperatures up to 1100 °C. To accommodate the desire of NEUP to extend the single particle EDB measurements to elevated temperatures it was necessary to significantly revise the plan and the budget. These revisions were approved. In the EDB method, we levitate a single graphite particle (the size, surface characteristics, morphology, purity, and composition of the particle can be varied) or agglomerate in the balance and measure the sorption of species by observing the changes in mass. This process involves the use of an electron stepping technique to measure the total charge on a particle which, in conjunction with the measured suspension voltages for the particle, allows for determinations of mass and, hence, of mass changes which then correspond to measurements of sorption. Accommodating elevated temperatures with this type of system required a significant system redesign and required additional time that ultimately was not available. These constraints also meant that the grant had to focus on fewer species as a result. Overall, the extension of the original proposed single particle work to elevated temperatures added greatly to the complexity of the proposed project and added greatly to the time that would eventually be required as well. This means that the bulk of the experimental progress was made using the packed bed sorption systems. Only being able to recruit one

  5. Recovery and use of fission product noble metals

    International Nuclear Information System (INIS)

    Jensen, G.A.; Rohmann, C.A.; Perrigo, L.D.

    1980-06-01

    Noble metals in fission products are of strategic value. Market prices for noble metals are rising more rapidly than recovery costs. A promising concept has been developed for recovery of noble metals from fission product waste. Although the assessment was made only for the three noble metal fission products (Rh, Pd, Ru), there are other fission products and actinides which have potential value

  6. 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

  7. 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

  8. Irradiated uranium reprocessing, Final report I-VI, Part VI - Separation of uranium, plutonium and fission products from HNO3 solution on the zirconium phosphate (part I), Adsorption equilibrium and kinetics

    International Nuclear Information System (INIS)

    Gal, I.; Ruvarac, A.

    1961-12-01

    Separation of uranium, plutonium and long-lived fission products was investigated on a inorganic ion exchanger. Zirconium phospate was chosen for this purpose because its ion exchanger properties were well known. This report deals with the study of equilibrium and kinetics of the adsorption

  9. 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)

  10. 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.

  11. Chemistry of fission products for accident analysis

    International Nuclear Information System (INIS)

    Potter, P.E.

    1985-01-01

    Current knowledge concerning the chemical state of the fission product elements during the development of accidents in water reactor systems is reviewed in this paper. The fission product elements which have been considered are Cs, I, Te, Sr and Ba but aspects of the behavior of Mo, Ru and the lanthanides are also discussed. Some features of the reactions of the various species of these elements with other components of the reactor systems are described. The importance of having an adequate knowledge of thermodynamic data and phase equilibria of relatively simple systems in order to interpret experimental observations on complex multi-component systems is stressed

  12. Transmutation of fission products through accelerator

    International Nuclear Information System (INIS)

    Nakamura, H.; Tani, S.; Takahashi, T.; Yamamura, O.

    1995-01-01

    The transmutation of fission products through particle accelerators has been studied under the OMEGA program. The photonuclear reaction has also been investigated to be applied to transmuting long-lived fission products, such as Cesium and Strontium, which have difficulties on reaction with neutrons due to its so small cross section. It is applicable for the transmutation if the energy balance can be improved with a monochromatic gamma rays in the range of the Giant Dipole Resonance generated through an excellent high current electron linear accelerator. The feasibility studies are being conducted on the transmutation system using it through an electron accelerator. (authors)

  13. Regulatory simplification of fission product chemistry

    International Nuclear Information System (INIS)

    Read, J.B.J.; Soffer, L.

    1986-01-01

    The requirements for design provisions intended to limit fission product escape during reactor accidents have been based since 1962 upon a small number of simply-stated assumptions. These assumptions permeate current reactor regulation, but are too simple to deal with the complex processes that can reasonably be expected to occur during real accidents. Potential chemical processes of fission products in severe accidents are compared with existing plant safety features designed to minimize off-site consequences, and the possibility of a new set of simply-stated assumptions to replace the 1982 set is discussed

  14. 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.)

  15. 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

  16. JNDC nuclear data library of fission products

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Ihara, Hitoshi; Akiyama, Masatsugu; Yoshida, Tadashi; Matumoto, Zyun-itiro; Nakasima, Ryuzo

    1983-10-01

    The JNDC (Japanese Nuclear Data Committee) FP (Fission Product) nuclear data library for 1172 fission products is described in this report. The gross theory of beta decay has been used extensively for estimating unknown decay data and also some of known decay data with poor accuracy. The calculated decay powers of fission products using the present library show excellent agreement with the latest measurements at ORNL (Oak Ridge National Laboratory), LANL (Los Alamos National Laboratory) and UTT (University of Tokyo, Tokai) for cooling times shorter than 10 3 s after irradiation. The calculated decay powers by the existing libraries showed systematic deviations at short cooling times; the calculated beta and gamma decay powers after burst fission were smaller than the experimental results for cooling times shorter than 10 s, and in the cooling time range 10 to 10 3 s the beta-decay power was larger than the measured values and the gamma decay power smaller than the measured results. The present JNDC FP nuclear data library resolved these discrepancies in the short cooling time ranges. The decay power of fission products has been calculated for ten fission types and the results have been fitted by an analytical function with 31 exponentials. This permits the easy application of the present results of decay power calculations to a LOCA (Loss-of-Coolant Accident) analysis of a light water reactor and so on. (author)

  17. 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.)

  18. Model for fission-product calculations

    International Nuclear Information System (INIS)

    Smith, A.B.

    1984-01-01

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

  19. Recent progress in fission product separation

    International Nuclear Information System (INIS)

    Raggenbass, A.

    1964-01-01

    Successful experiments have been done on the method described at Geneva in 1958. The process has been considerably improved: 1 - Initially, the caesium phospho tungstate precipitate was leached barium hydroxide in the centrifuge and this was followed by a distillation of ammonia in a concentrator. The barium hydroxide was then eliminated by carbonate precipitation and centrifugation. It has been proved that the ammonia distillation could be replaced by its evaporation during centrifugation, thus eliminating the need of a concentrator. It was then possible to carry out the carbonation on the solide-liquid mixture produced by the baryte water leaching. 2 - In applying the above process to the treatment of solutions derived from uranium molybdenum fuels, concentrating is to be recommended in order to hold the molybdenum in solution by complexing it with phosphoric acid. This complexing process provides a suspension of zirconium phosphate and ammonium phospho tungstate. These are separated by passing into a basic medium which precipitates the zirconium oxide, then turning back to an acid medium; the end of the treatment remains unchanged. 3 - Studies carried out in several countries on the exchange properties of hetero-polyacid salts have always met with difficulties as a result of the poor mechanical properties of these substances. This difficulty has been overcome by wrapping the ammonium phospho tungstate in a zirconium phosphate matrix. The exchanger obtained possesses: satisfactory mechanical properties, - a capacity of 0.1 milli equivalent per gram in concentrated nitric acid solution. It can be eluted and regenerated by a solution of an ammonium salt. The procedure for recovery of these various fission products is briefly the following: extraction of rare earths by di-2-ethyl hexyl phosphoric acid into dodecane at pH 2, the chemical impurities being complexed by citric acid, extraction of most of the magnesium at pH 4 by the same solvents the solvent being

  20. 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)

  1. 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

  2. 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

  3. Fission product release from SLOWPOKE-2 reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

  4. Release of fission products in transients

    International Nuclear Information System (INIS)

    Christensen, H.; Lundqwist, R.

    1979-07-01

    A station for automatic sampling of coolant has been put in operation at the Oskarshamn-1 reactor. The release of 131 J and other fission products in spikes in connection with reactor trips and scheduled shutdowns has been measured. A model developed at General Electric has been used to predict the spike release in Oskarshamn-1 and the predicted values have been compared with experimental values. Literature data of iodine spikes in BWR and PWR have been reviewed. (author)

  5. Actinide and fission product partitioning and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    The fourth international information exchange meeting on actinide and fission product partitioning and transmutation, took place in Mito City in Japan, on 111-13 September 1996. The proceedings are presented in six sessions: the major programmes and international cooperation, the partitioning and transmutation programs, feasibility studies, particular separation processes, the accelerator driven transmutation, and the chemistry of the fuel cycle. (A.L.B.)

  6. Gas-phase transport of fission products

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  7. Dosimetric measurement of the disintegration rate of fission products

    International Nuclear Information System (INIS)

    Solymosi, J.; Nagy, L.G.; Zagyvai, P.

    1992-01-01

    Investigations on the disintegration rate of fission products of 238 U and 239 Pu are presented. The intensity of the β-and γ-radiation of fission products were measured continously in an interval of 1-1300 hours following the fission, offering the possibility for determining the general and specific characteristics of the individual fission products. A universal measuring procedure was elaborated for the rapid in situ determination of the dosimetric features of fission products, which is suitable for the accurate evaluation and prediction of external absorbed dose even in case of fission products of various origin and unknown composition. (author) 6 refs.; 7 figs.; 1 tab

  8. 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

  9. Actinides and fission products partitioning from high level liquid waste

    International Nuclear Information System (INIS)

    Yamaura, Mitiko

    1999-01-01

    The presence of small amount of mixed actinides and long-lived heat generators fission products as 137 Cs and 90 Sr are the major problems for safety handling and disposal of high level nuclear wastes. In this work, actinides and fission products partitioning process, as an alternative process for waste treatment is proposed. First of all, ammonium phosphotungstate (PWA), a selective inorganic exchanger for cesium separation was chosen and a new procedure for synthesizing PWA into the organic resin was developed. An strong anionic resin loaded with tungstate or phosphotungstate anion enables the precipitation of PWA directly in the resinous structure by adding the ammonium nitrate in acid medium (R-PWA). Parameters as W/P ratio, pH, reactants, temperature and aging were studied. The R-PWA obtained by using phosphotungstate solution prepared with W/P=9.6, 9 hours digestion time at 94-106 deg C and 4 to 5 months aging time showed the best capacity for cesium retention. On the other hand, Sr separation was performed by technique of extraction chromatography, using DH18C6 impregnated on XAD7 resin as stationary phase. Sr is selectively extracted from acid solution and >99% was recovered from loaded column using distilled water as eluent. Concerning to actinides separations, two extraction chromatographic columns were used. In the first one, TBP(XAD7) column, U and Pu were extracted and its separations were carried-out using HNO 3 and hydroxylamine nitrate + HNO 3 as eluent. In the second one, CMP0-TBP(XAD7) column, the actinides were retained on the column and the separations were done by using (NH 4 ) 2 C 2 O 4 , DTPA, HNO 3 and HCl as eluent. The behavior of some fission products were also verified in both columns. Based on the obtained data, actinides and fission products Cs and Sr partitioning process, using TBP(XAD7) and CMP0-TBP(XAD7) columns for actinides separation, R-PWA column for cesium retention and DH18C6(XAD7) column for Sr isolation was performed

  10. Immobilization of fission products in phosphate ceramic waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Singh, D.; Wagh, A. [Argonne National Lab., IL (United States)

    1997-10-01

    Chemically bonded phosphate ceramics (CBPCs) have several advantages that make them ideal candidates for containing radioactive and hazardous wastes. In general, phosphates have high solid-solution capacities for incorporating radionuclides, as evidenced by several phosphates (e.g., monazites and apatites) that are natural analogs of radioactive and rare-earth elements. The phosphates have high radiation stability, are refractory, and will not degrade in the presence of internal heating by fission products. Dense and hard CBPCs can be fabricated inexpensively and at low temperature by acid-base reactions between an inorganic oxide/hydroxide powder and either phosphoric acid or an acid-phosphate solution. The resulting phosphates are extremely insoluble in aqueous media and have excellent long-term durability. CBPCs offer the dual stabilization mechanisms of chemical fixation and physical encapsulation, resulting in superior waste forms. The goal of this task is develop and demonstrate the feasibility of CBPCs for S/S of wastes containing fission products. The focus of this work is to develop a low-temperature CBPC immobilization system for eluted {sup 99}Tc wastes from sorption processes.

  11. Analytical evaluation of fission product sensitivities

    International Nuclear Information System (INIS)

    Sola, A.

    1977-01-01

    Evaluating the concentration of a fission product produced in a reactor requires the knowledge of a fairly large number of variables. Sensitivity studies were made to ascertain the important variables. Analytical formulae were developed sufficiently simple to allow numerical computations. Some simplified formulas are also given and they are applied to the following isotopes: 80 Se, 82 Se, 81 Br, 82 Br, 82 Kr, 83 Kr, 84 Kr, 85 Kr, 86 Kr. Their sensitivities to capture cross sections, fission yields, ratios of activation cross sections, half-lives (during and after irradiation), branching ratios, as well as to the neutron flux and to the time are considered

  12. Separation of fission products using inorganic exchangers

    International Nuclear Information System (INIS)

    Murthy, T.S.; Balasubramanian, K.R.; Rao, K.L.N.; Venkatachalam, R.; Varma, R.N.

    1981-01-01

    This paper describes the separation of long lived fission products like caesium-137, strontium-90 using inorganic exchangers ammonium phosphomolybdate and zirconium antimonate. A revised flow sheet is proposed for the sequential separation of these isotopes using the above two compounds. This is a modification of the earlier scheme developed which involved the use of four inorganic exchangers namely ammonium phosphomolybdate, manganese dioxide, zirconium antimonate and polyantimonic acid. The elution of the adsorbed elements like cerium, strontium, and sodium has been studied and it has been possible to elute these using different eluting agents. (author)

  13. Recovery of noble metals from fission products

    International Nuclear Information System (INIS)

    Jenson, G.A.; Platt, A.M.; Mellinger, G.B.; Bjorklund, W.J.

    1982-11-01

    Scoping studies were started in 1979 to develop a cost-effective, waste-management-compatible process to extract noble metals from fission products. The process, involving the reaction with glassmelting chemicals, a metal oxide (PbO), and a reducing agent (charcoal), was demonstrated for recovering noble metals from simulated high-level waste oxides. The process has now been demonstrated on a laboratory scale (100 g) using irradiated fuels. Recoveries in the recovered lead averaged 80% for Pd, 60% for Rh, and 14% Ru. The resulting glass product was homogeneous in appearance, and the chemical durability was comparable to other waste oxides

  14. Fission Product Experimental Program: Validation and Computational Analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-02-15

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

  15. Simulation of COMEDIE Fission Product Plateout Experiment Using GAMMA-FP

    International Nuclear Information System (INIS)

    Tak, Nam-il; Yoon, Churl

    2014-01-01

    FThis phenomenon is particularly important under a VHTR design with vented low pressure confinement (VLPC), because the vent allows the prompt release of fission products accumulated within the primary circuit to environment during an initial blow-down phase after pipe break accidents. In order to analyze the fission product plateout, an numerical model was developed by Yoo et al. and incorporated into the GAMMA-FP code in the past. The GAMMA-FP model was validated against two experiment data, i.e., VAMPYR-1 and OGL, during the development phase. One of the well-known experiments for fission product plateout is the COMEDIE experiment. In this work, the COMEDIE experiment has been simulated using the GAMMA-FP code to investigate the reliability and applicability of the plateout model of GAMMA-FP. The COMEDIE experiment for fission product plateout was simulated using the GAMMA-FP code in this work. A good agreement was achieved between the measured and predicted plateout activities. The existing solution scheme was modified to allow larger time step size for fission product analysis in order to speed-up the computational time. Nevertheless, the modification of the existing numerical model of GAMMA-FP is necessary when a simulation capability of a long duration of plateout period (e.g., 60 years) is targeted

  16. Polyphase diffusion of fission products in graphite

    International Nuclear Information System (INIS)

    Dannert, V.

    1989-05-01

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

  17. 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

  18. Detector for gaseous nuclear fission products

    International Nuclear Information System (INIS)

    Kobayashi, Yoshihiro; Kubo, Katsumi.

    1979-01-01

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

  19. Resuspension of fission products from sump water

    International Nuclear Information System (INIS)

    Bunz, H.; Koyro, M.; Propheter, B.; Schoeck, W.; Wagner-Ambs, M.

    1992-11-01

    Resuspension of fission products from the boiling sump in the container has long been known as a source of airborne radioactivity. Since this source is very weak, however, not much attention had been paid to it as long as radiological source terms were governed by stronger sources. Recently, the continuous reduction of source terms and the introduction of accident management measures led to a situation where weak but longlasting sources of radioactivity may become important, either as a contribution to the radiological sources term or as an impact to accident filtration systems. Existing data on resuspension from boiling contaminated water all suffered from two deficiencies: they were measured under conditions unlike those in a reactor accident and they scattered over more than two orders of magnitude. In a precursor study this uncertainty was considered to be too large to use the data for source term calculations. A later experimental research programme REST (REsuspension Source Term) was carried out at the Laboratorium fuer Aerosolphysik und Filtertechnik (LAF), Kernforschungszentrum Karlsruhe (KfK). The programme was supported by the Commission of the European Communities Ispra, under Contract No 3009-86-07 ELISPD in the framework of the shared-cost action programme on reactor safety. The investigations started in 1987 and ended in 1990. The objectives of the REST programme were to measure resuspension source characteristics under simulated accident conditions such that an application of the data in fission product transport and depletion models is possible

  20. Radio-chemical applications of columns of precipitated insoluble sulphates - contribution to the study of fission product solutions (1962); Applications en radiochimie de colonnes de precipites de sulfates insolubles - contribution a l'etude des solutions vieillies de produits de fission (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Sauvagnac, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Barrachina, M [Junta de la Energia Nuclear, Madrid (Spain)

    1962-07-01

    After describing the preparation of SrSO{sub 4} and BaS0{sub 4} columns, the authors study the following variables affecting fixation of Sr{sup 90}: - influence of pH, of the SO{sub 4}{sup --} concentration in the solution studied, of the column density, of the flow rate, of the use of the EDTA complex. A mechanism is proposed for the exchange and fixation. Once the column characteristics had been chosen, the fixation of rare-earths in the absence or presence of EDTA was studied. - it was observed that for nitrate or chloride concentrations of over 2 mg/ml, or in the presence of stronger than 1 N HNO{sub 3} or HCl, barium was fixed selectively. - these results were applied to radiochemical separations in fission product solutions: Ba{sup 140} and Sr{sup 90} separation, and that of the ceric earth group and the yttric earth group. - In order to establish a complete scheme for the radio-chemical separation in fission product solution, after at least one year ageing, the techniques of column separation and of solvent extraction using thenoyltrifluoroacetone and bis di-ethylhexylphosphoric acid are combined. (authors) [French] Apres avoir decrit la preparation des colonnes de SrS0{sub 4} et BaS0{sub 4}, on etudie les variables suivantes pour la fixation de {sup 90}Sr: - influence du pH, de la concentration en SO{sub 4}{sup --} dans la solution etudiee, de la densite de la colonne, de la vitesse de passage, de l'emploi du complexant EDTA. On propose une explication du mode d'echange et de fixation. Les caracteristiques de la colonne etant choisies, on a etudie la fixation des terres rares en presence ou non de EDTA. - on a d'autre part constate qu'en milieu chlorure ou nitrate, de concentration superieure a 2 mg/ml, ou en presence de HNO{sub 3} ou HCl superieur a 1 N, on fixait selectivement le baryum. - on a applique ces resultats a des separations radiochimiques dans des solutions de produits de fission: separation de {sup 140}Ba, {sup 90}Sr, separation du groupe des

  1. Ceramic Hosts for Fission Products Immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Peter C Kong

    2010-07-01

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

  2. Diffusion of Fission Product Elements in Compacted Bentonite

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  3. Progress in fission product nuclear data. No. 13

    International Nuclear Information System (INIS)

    Lammer, M.

    1990-11-01

    This is the 13th issue of a report series published by the Nuclear Data Section of the IAEA. The types of activities included are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission), neutron reaction cross-sections of fission products, data related to the radioactive decay of fission products, delayed neutron data of fission products and bumped fission product data (decay heat, absorption, etc.). The first part of the report consists of unaltered original data which the authors have sent to IAEA/NDS. The second part contains some recent references relative to fission product nuclear data, which were not covered by the contributions submitted, and selected papers from conferences. Part 3 contains requirements for further measurements

  4. Progress in fission product nuclear data. No. 14

    International Nuclear Information System (INIS)

    Lammer, M.

    1994-06-01

    This is the 14th issue of a report series on Fission Product Nuclear Data published by the Nuclear Data Section of the IAEA. The types of activities included are measurements, compilations and evaluations of fission product yields, neutron reaction cross sections of fission products, data related to the radioactive decay of fission products, delayed neutron data from neutron induced and spontaneous fission, lumped fission product data. The first part of the report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The second part contains some recent references relative to fission product nuclear data, which were not covered by the contributions submitted, and selected papers from conferences. The third part contains requirements for further measurements

  5. 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)

  6. SEPARATION OF URANIUM, PLUTONIUM, AND FISSION PRODUCTS

    Science.gov (United States)

    Spence, R.; Lister, M.W.

    1958-12-16

    Uranium and plutonium can be separated from neutron-lrradiated uranium by a process consisting of dissolvlng the lrradiated material in nitric acid, saturating the solution with a nitrate salt such as ammonium nitrate, rendering the solution substantially neutral with a base such as ammonia, adding a reducing agent such as hydroxylamine to change plutonium to the trivalent state, treating the solution with a substantially water immiscible organic solvent such as dibutoxy diethylether to selectively extract the uranium, maklng the residual aqueous solutlon acid with nitric acid, adding an oxidizing agent such as ammonlum bromate to oxidize the plutonium to the hexavalent state, and selectlvely extracting the plutonium by means of an immlscible solvent, such as dibutoxy dlethyletber.

  7. 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

  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. 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

  10. Simulation of Fission Product Liftoff Behavior During Depressurization Transients

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  11. Fission product margin in burnup credit analyses

    International Nuclear Information System (INIS)

    Finck, P.J.; Stenberg, C.G.

    1998-01-01

    The US Department of Energy (DOE) is currently working toward the licensing of a methodology for using actinide-only burnup credit for the transportation of spent nuclear fuel (SNF). Important margins are built into this methodology. By using comparisons with a representative experimental database to determine bias factors, the methodology ensures that actinide concentrations and worths are estimated conservatively; furthermore, the negative net reactivity of certain actinides and all fission products (FPs) is not taken into account, thus providing additional margin. A future step of DOE's effort might aim at establishing an actinide and FP burnup credit methodology. The objective of this work is to establish the uncertainty to be applied to the total FP worth in SNF. This will serve two ends. First, it will support the current actinide-only methodology by demonstrating the margin available from FPs. Second, it will identify the major contributions to the uncertainty and help set priorities for future work

  12. Core degradation and fission product release

    International Nuclear Information System (INIS)

    Wright, R.W.; Hagen, S.J.L.

    1992-01-01

    Experiments on core degradation and melt progression in severe LWR accidents have provided reasonable understanding of the principal processes involved in the early phase of melt progression that extends through core degradation and metallic material melting and relocation. A general but not a quantitative understanding of late phase melt progression that involves ceramic material melting and relocation has also been obtained, primarily from the TMI-2 core examination. A summary is given of the current state of knowledge on core degradation and melt progression obtained from these integral experiments and of the principal remaining significant uncertainties. A summary is also given of the principal results on in-vessel fission product release obtained from these experiments. (author). 8 refs, 5 figs, 3 tabs

  13. Systematics of Fission-Product Yields

    International Nuclear Information System (INIS)

    Wahl, A.C.

    2002-01-01

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

  14. Fission product transport at Three Mile Island

    International Nuclear Information System (INIS)

    Owen, D.E.; Cox, T.E.; Broughton, J.M.

    1985-01-01

    Three Mile Island Unit 2 radionuclide analyses are reviewed and summarized in order to determine how fission products moved to various parts of the reactor system at the time of the accident. Despite high fuel temperatures and major core damage, the core retained a very large fraction of most radionuclides. Reactor coolant, either remaining in the primary system or released to various sumps and tanks, retained significant quantities of cesium and iodine. Noble gases were effectively retained within the containment building with the exception of small releases to the environment. Long-term deposition and retention on vessel, piping, and bulding surfaces were insignificant for all isotopes examined. The measured partitioning of radionuclides within these systems is tabulated and recommendations for additional analyses are presented

  15. Retention of fission products in air filters

    International Nuclear Information System (INIS)

    Sobnack, R.

    1986-01-01

    The plume from the Chernobyl nuclear reactor reached London in the morning of 1st May. Less than two weeks later, the Physics Department, University of Surrey, reported a measurable level of radioactivity in air filters. On 15th May air filters from within the air conditioning plant of the Radioisotope Department at the London Hospital were removed for radiation checks. Crude tests with a geiger counter gave readings of 5-10 times higher than background levels. Gamma-ray spectroscopy of the departmental air filters (AF1) using a 127 mm NaI detector revealed a pattern characteristic of emissions of fission products from a nuclear reactor. Another air filter (AF2), from the home of a member of staff, was much less active. Because of the complexity of the gamma-ray spectrum and the relatively high level of emission from the departmental air filter, a thorough investigation was carried out using a high purity germanium detector. (author)

  16. (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.

  17. Actinide and fission product separation and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1991-07-01

    The first international information exchange meeting on actinide and fission product separation and transmutation, took place in Mito in Japan, on 6-8 November 1990. It starts with a number of general overview papers to give us some broad perspectives. Following that it takes a look at some basic facts about physics and about the quantities of materials it is talking about. Then it proceeds to some specific aspects of partitioning, starting with evolution from today commercially applied processes and going on to other possibilities. At the end of the third session it takes a look at the significance of partitioning and transmutation of actinides before it embarks on two sessions on transmutation, first in reactors and second in accelerators. The last session is designed to throw back into the discussion the main points which need to be looked at when considering future work in this area. (A.L.B.)

  18. Actinide and fission product separation and transmutation

    International Nuclear Information System (INIS)

    1991-01-01

    The first international information exchange meeting on actinide and fission product separation and transmutation, took place in Mito in Japan, on 6-8 November 1990. It starts with a number of general overview papers to give us some broad perspectives. Following that it takes a look at some basic facts about physics and about the quantities of materials it is talking about. Then it proceeds to some specific aspects of partitioning, starting with evolution from today commercially applied processes and going on to other possibilities. At the end of the third session it takes a look at the significance of partitioning and transmutation of actinides before it embarks on two sessions on transmutation, first in reactors and second in accelerators. The last session is designed to throw back into the discussion the main points which need to be looked at when considering future work in this area. (A.L.B.)

  19. Library of data for fission products

    International Nuclear Information System (INIS)

    Blachot, Jean; Devillers, Christian; Tourreil, Roland de; Nimal, Bernadette; Fiche, Charles; Noel, J.-P.

    1975-10-01

    This is the fourth version of the CEA fission products nuclear data library. The third one has been previously published in CEA-N--1526. Data for 635 nuclides ranging from mass A=71 up to A=170 are arranged in increasing order of atomic number. Data are presented in two tables: the first one gives for each nuclide, the half-life, the Q-values and branching ratios for the various decay modes, the energies and intensities of the β - , β + and isomeric transitions and of gamma rays; the second one gives an ordered list of all gamma ray energies, with associated nuclide, half-life and intensity. Bibliographic references and, for most of the data, uncertainties are provided [fr

  20. Systematics of Fission-Product Yields

    Energy Technology Data Exchange (ETDEWEB)

    A.C. Wahl

    2002-05-01

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

  1. Chemical factors affecting fission product transport in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

    1984-10-01

    This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly

  2. Immobilization of fission products in phosphate ceramic waste forms

    International Nuclear Information System (INIS)

    Singh, D.

    1996-01-01

    The goal of this project is to develop and demonstrate the feasibility of a novel low-temperature solidification/stabilization (S/S) technology for immobilizing waste streams containing fission products such as cesium, strontium, and technetium in a chemically bonded phosphate ceramic. This technology can immobilize partitioned tank wastes and decontaminate waste streams containing volatile fission products

  3. Fission product source term research at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1985-01-01

    The purpose of this work is to describe some of the research being performed at ORNL in support of the effort to describe, as realistically as possible, fission product source terms for nuclear reactor accidents. In order to make this presentation manageable, only those studies directly concerned with fission product behavior, as opposed to thermal hydraulics, accident sequence progression, etc., will be discussed

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  5. REGENERATION OF FISSION-PRODUCT-CONTAINING MAGNESIUM-THORIUM ALLOYS

    Science.gov (United States)

    Chiotti, P.

    1964-02-01

    A process of regenerating a magnesium-thorium alloy contaminated with fission products, protactinium, and uranium is presented. A molten mixture of KCl--LiCl-MgCl/sub 2/ is added to the molten alloy whereby the alkali, alkaline parth, and rare earth fission products (including yttrium) and some of the thorium and uranium are chlorinated and

  6. Volatilization and reaction of fission products in flowing steam

    International Nuclear Information System (INIS)

    Johnson, I.; Steidl, D.V.; Johnson, C.E.

    1985-01-01

    The principal risk to the public from nuclear power plants derives from the highly radioactive atoms (fission products) generated as energy is produced in the nuclear fuel. The revolatilization of fission products from reactor system surfaces due to self-heating by radioactive decay has become a complicating factor in the source-term redefinition effort. It has had a major impact on calculations of fission product distributions in accident safety analyses. The focus of this research effort was to investigate the volatilization and transport of fission products and control rod materials in a flowing gaseous steam-hydrogen mixture. Fission product and control rod materials in various combinations were studied including CsI, CsOH, TeO 2 , SrO, Ag, In, Cd and Mn. The vaporization behavior of the deposits were characterized with respect to vaporization rates, chemical species and downstream transport behavior

  7. [Fission product yields of 60 fissioning reactions]. Final report

    International Nuclear Information System (INIS)

    Rider, B.F.

    1995-01-01

    In keeping with the statement of work, I have examined the fission product yields of 60 fissioning reactions. In co-authorship with the UTR (University Technical Representative) Talmadge R. England ''Evaluation and Compilation of Fission Product Yields 1993,'' LA-UR-94-3106(ENDF-349) October, (1994) was published. This is an evaluated set of fission product Yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set

  8. Fission product and aerosol behaviour within the containment

    International Nuclear Information System (INIS)

    Beard, A.M.; Benson, C.G.; Bowsher, B.R.; Dickinson, S.; Nichols, A.L.

    1990-04-01

    Experimental studies have been undertaken to characterise the behaviour of fission products in the containment of a pressurised water reactor during a severe accident. The following aspects of fission product transport have been studied: (a) aerosol nucleation, (b) vapour transport processes, (c) chemical forms of high-temperature vapours, (d) interaction of fission product vapours with aerosols generated from within the reactor core, (e) resuspension processes, (f) chemistry in the containment. Chemical effects have been shown to be important in defining and quantifying fission product source terms in a wide range of accident sequences. Both the chemical forms of the fission product vapours and their interactions with reactor materials aerosols could have a major effect on the magnitude and physicochemical forms of the radioactive emission from a severe reactor accident. Only the main conclusions are presented in this summary document; detailed technical aspects of the work are described in separate reports listed in the annex

  9. Characteristics of fission product release from a molten pool

    International Nuclear Information System (INIS)

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

    2001-01-01

    The volatile fission products are released from the debris pool, while the less volatile fission products tend to remain as condensed phases because of their low vapor pressure. The release of noble gases and the volatile fission products is dominated by bubble dynamics. The release of the less volatile fission products from the pool can be analyzed based on mass transport through a liquid with the convection flow. The physico-numerical models were orchestrated from existing submodels in various disciplines of engineering to estimate the released fraction of fission products from a molten pool. It was assumed that the pool has partially filled hemispherical geometry. For the high pool pressure, the diameter of the bubbles at detachment was calculated utilizing the Cole and Shulman correlation with the effect of system pressure. Sensitivity analyses were performed and results of the numerical calculations were compared with analysis results for the TMI-2 accident. (author)

  10. Status of decay data of fission products

    International Nuclear Information System (INIS)

    Blachot, J.

    1978-01-01

    Fission products (F.P.) are neutron rich isotopes ranging from Zn to Tm. The status of decay data of F.P. was described at the Bologna Panel 1973 by Rudstam. Since then, FPND have improved in general, but still much is valid of what Rudstam said about the accuracies of FPND. The lack of decay data for the short lived F.P. has been considerably reduced, and some of the short lived F.P. have now well studied decay data. The present status of decay data is given in this review, which is composed of six sections. In the first one, the principal new facilities used in decay data measurements are reviewed. The second part is devoted to the total decay energy (Q). In the third Section, the half lives are treated. In the fourth and fifth Sections, beta and gamma energies and intensities, and also average values are discussed. Finally, the last Section considers the different files and compilations devoted to the decay of F.P

  11. ENDF/B fission product decay data

    International Nuclear Information System (INIS)

    Rose, P.F.; Burrows, T.W.

    1976-08-01

    The fission product data have been organized by A-chains in order of ascending A from A = 72 to A = 167. The heading page is followed by more detailed information on the individual members of the chain in order of increasing Z and decreasing metastable state. The detailed information for each member includes the ENDF/B-IV File 1 comments and references if available and applicable to the decay data. Following the comments is a decay scheme of the nuclide tabulating the quantities T/sub 1 / 2 /, Q, branching ratio (BR), (E/sub γ/), (E/sub β/), and (E/sub α/). Uncertainties are given if available in the file. Independent fission yields are given, as well as thermal cross sections and resonance integrals as obtained from ENDF/B-IV. All energies listed in this publication are in keV, and all branching ratios (BR) sum to unity. If there are spectra in the decay data file, the decay scheme is followed by tables of photon, particle, and characteristic radiation. For cases in which the multipolarities could be obtained from the file the tables also contain information on x-rays, conversion electrons, and Auger electrons. Associated with the photon and particle radiation tables are the appropriate average energies per decay for each type of radiation, including neutrino radiation

  12. Polarographic determination of Iodide and Iodate, in Solutions Coming from Aerosols in Fission Products Containment Studies in Nuclear Power Stations; Determinacion Polarografica de Especies de Iodo (Ioduro y Iodato) en Soluciones Procedentes de Aerosoles, para Estudios de Contencion de Productos de Fision en Centrales Nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, M.; Ballesteros, O.; Fernandez, M.; Clavero, M.; Gonzalez, A. M. [Ciemat, Madrid (Spain)

    2000-07-01

    A polarographic method is described for the iodine species determination, iodide and iodate in water solutions. the iodate can be determined by differential pulse polarography. Calibration curves and the detection and determination limits have been obtained. Iodides is oxidized to iodate with sodium hypochlorite and the excess of oxidizing agent is destroyed with sodium sulphide. The concentration of iodide is calculated as the difference between the concentration of iodate in the sample before and after the oxidation. As an application, species of iodine in samples coming from the experimental plants GIRS (Gaseous Iodine Removal by Sprays) of Nuclear Fission Department of the CIEMAT, dedicated to fission products containment studies in nuclear power station, were determined. (Author) 10 refs.

  13. An analysis of the additional fission product release phenomena

    International Nuclear Information System (INIS)

    Takeda, Tsuneo; Nagai, Hitoshi

    1978-09-01

    The additional fission product release behavior through a defect hole on the cladding of fuel rods has been studied qualitatively with a computer program CODAC-ARFP. The additional fission product release phenomena are described as qualitative evaluation. The additional fission product release behavior in coolant temperature and pressure fluctuations and in reactor start-up and shut-down depends on coolant water flow behavior into and from the free space of fuel rods through a defect hole. Based on the results of evaluations, the experimental results with an inpile water loop OWL-1 are described in detail. The estimation methods of fission product quantity in the free space and fission product release ratio (quantity released into the coolant/quantity in the free space before beginning of release) are necessary for analysis of the fission product release behavior; the estimation method of water flow through a defect hole is also necessary. In development of the above estimation methods, outpile and capsule experiments supporting the additional fission product release experiments are required. (author)

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  15. RSAC-6, Gamma doses, inhalation and ingestion doses, fission products inventory after fission products release

    International Nuclear Information System (INIS)

    Wenzel, Douglas R.; Schrader, Brad J.

    2007-01-01

    1 - Description of program or function: RSAC-6 is the latest version of the program RSAC (Radiological Safety Analysis Computer Program). It calculates the consequences of a release of radionuclides to the atmosphere. Using a personal computer, a user can generate a fission product inventory; decay and in-grow the inventory during transport through processes, facilities, and the environment; model the downwind dispersion of the activity; and calculate doses to downwind individuals. Internal dose from the inhalation and ingestion pathways is calculated. External dose from ground surface and plume gamma pathways is calculated. New and exciting updates to the program include the ability to evaluate a release to an enclosed room, resuspension of deposited activity and evaluation of a release up to 1 meter from the release point. Enhanced tools are included for dry deposition, building wake, occupancy factors, respirable fraction, AMAD adjustment, updated and enhanced radionuclide inventory and inclusion of the dose-conversion factors from FOR 11 and 12. 2 - Methods: RSAC6 calculates meteorological dispersion in the atmosphere using Gaussian plume diffusion for Pasquill-Gifford, Hilmeier-Gifford and Markee models. A unique capability is the ability to model Class F fumigation conditions, the meteorological condition that causes the highest ground level concentrations from an elevated release. Doses may be calculated for various pathways including inhalation, ingestion, ground surface, air immersion, water immersion pathways. Dose calculations may be made for either acute or chronic releases. Internal doses (inhalation and ingestion) are calculated using the ICRP-30 model with dose conversion factors from FOR 11. External factors are calculated using FOR 12. 3 - Unusual Features: RSAC6 calculates complete progeny in-growth and decay during all accident phases. The calculation of fission product inventories in particularly useful in the analysis of accidents where the

  16. Study of the fission products fixation in the hydroxyapatite mineral

    International Nuclear Information System (INIS)

    Soriano R, J. M.

    2011-01-01

    In this research work, sorption properties of hydroxyapatite in aqueous solutions were studied using Na + and K + ion behavior. In addition, the fission products 99 Tc and 107 Pd uptake was studied to determine their sorption mechanisms on hydroxyapatite. This research was conducted in two stages. The first stage aimed to identify surface reactive sites of hydroxyapatite surface. This surface study was performed by the radiotracer method using 24 Na and 42 K radionuclides and applying the ion-exchange theory. It provides evidence in terms of the saturation curves of individual behaviour of the Na + and K + cations. Hydroxyapatite reactive sites were identified and quantified from the results and application of the ion-exchange model: a mono-functional site of 0.28 mmol g -1 for the sodium hydroxylate form and a dipr otic site with two saturation curves of 0.14 mmol g -1 each, for the sodium phosphate form. In a second stage, the sorption of fission products, Tc and Pd, on hydroxyapatite was studied. This sorption was expressed in terms of distribution coefficients obtained with equivalent radiotracers: 99m Tc and 109 Pd. Tc presented a low sorption affinity on hydroxyapatite in aqueous medium 0.02 M NaH 2 PO 4 and the results also show that Tc is not sorbed from perchlorate medium (0.01 M Ca(ClO 4 ) 2 ). Sorption behaviour of Pd(II) on hydroxyapatite was studied for different experimental conditions, with parameter such as: ph, aqueous medium (0.01 M NaClO 4 , 0.01 M and 0.025 M Ca(ClO 4 ) 2 , and 0.02 M NaH 2 PO 4 ), the solid solution ratio (10, 4 and 0.020 g/L), and the palladium concentration were studied. Pd sorption was complete at solid-solution ratios 10 and 4 g/L. A strong sorption affinity of hydroxyapatite for palladium was obtained at solid-solution ratio 0.020 g/L. In the interpretation of the results it was considered the aqueous chemistry of palladium, solid dissolution, as well as the existence of reactive sites at the hydroxyapatite surface. The

  17. 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)

  18. 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

  19. Chemistry of fission product iodine under nuclear reactor accident conditions

    International Nuclear Information System (INIS)

    Malinauskas, A.P.; Bell, J.T.

    1986-01-01

    The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs

  20. Interactions of fission product vapours with aerosols

    Energy Technology Data Exchange (ETDEWEB)

    Benson, C G; Newland, M S [AEA Technology, Winfrith (United Kingdom)

    1996-12-01

    Reactions between structural and reactor materials aerosols and fission product vapours released during a severe accident in a light water reactor (LWR) will influence the magnitude of the radiological source term ultimately released to the environment. The interaction of cadmium aerosol with iodine vapour at different temperatures has been examined in a programme of experiments designed to characterise the kinetics of the system. Laser induced fluorescence (LIF) is a technique that is particularly amenable to the study of systems involving elemental iodine because of the high intensity of the fluorescence lines. Therefore this technique was used in the experiments to measure the decrease in the concentration of iodine vapour as the reaction with cadmium proceeded. Experiments were conducted over the range of temperatures (20-350{sup o}C), using calibrated iodine vapour and cadmium aerosol generators that gave well-quantified sources. The LIF results provided information on the kinetics of the process, whilst examination of filter samples gave data on the composition and morphology of the aerosol particles that were formed. The results showed that the reaction of cadmium with iodine was relatively fast, giving reaction half-lives of approximately 0.3 s. This suggests that the assumption used by primary circuit codes such as VICTORIA that reaction rates are mass-transfer limited, is justified for the cadmium-iodine reaction. The reaction was first order with respect to both cadmium and iodine, and was assigned as pseudo second order overall. However, there appeared to be a dependence of aerosol surface area on the overall rate constant, making the precise order of the reaction difficult to assign. The relatively high volatility of the cadmium iodide formed in the reaction played an important role in determining the composition of the particles. (author) 23 figs., 7 tabs., 22 refs.

  1. Interactions of fission product vapours with aerosols

    International Nuclear Information System (INIS)

    Benson, C.G.; Newland, M.S.

    1996-01-01

    Reactions between structural and reactor materials aerosols and fission product vapours released during a severe accident in a light water reactor (LWR) will influence the magnitude of the radiological source term ultimately released to the environment. The interaction of cadmium aerosol with iodine vapour at different temperatures has been examined in a programme of experiments designed to characterise the kinetics of the system. Laser induced fluorescence (LIF) is a technique that is particularly amenable to the study of systems involving elemental iodine because of the high intensity of the fluorescence lines. Therefore this technique was used in the experiments to measure the decrease in the concentration of iodine vapour as the reaction with cadmium proceeded. Experiments were conducted over the range of temperatures (20-350 o C), using calibrated iodine vapour and cadmium aerosol generators that gave well-quantified sources. The LIF results provided information on the kinetics of the process, whilst examination of filter samples gave data on the composition and morphology of the aerosol particles that were formed. The results showed that the reaction of cadmium with iodine was relatively fast, giving reaction half-lives of approximately 0.3 s. This suggests that the assumption used by primary circuit codes such as VICTORIA that reaction rates are mass-transfer limited, is justified for the cadmium-iodine reaction. The reaction was first order with respect to both cadmium and iodine, and was assigned as pseudo second order overall. However, there appeared to be a dependence of aerosol surface area on the overall rate constant, making the precise order of the reaction difficult to assign. The relatively high volatility of the cadmium iodide formed in the reaction played an important role in determining the composition of the particles. (author) 23 figs., 7 tabs., 22 refs

  2. Influence of fission product transport on delayed neutron precursors and decay heat sources in LMFBR accidents

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.

    1981-01-01

    A method is presented for studying the influence of fission product transpot on delayed neutron precursors and decay heat sources during Liquid Metal Fast Breeder Reactor (LMFBR) unprotected accidents. The model represents the LMFBR core as a closed homogeneous cell. Thermodynamic phase equilibrium theory is used to predict fission product mobility. Reactor kinetics behavior is analyzed by an extension of point kinetics theory. Group dependent delayed neutron precursor and decay heat source retention factors, which represent the fraction of each group retained in the fuel, are developed to link the kinetics and thermodynamics analysis. Application of the method to a highly simplified model of an unprotected loss-of-flow accident shows a time delay on the order of 10 ms is introduced in the predisassembly power history if fission product motion is considered when compared to the traditional transient solution. The post-transient influence of fission product transport calculated by the present model is a 24 percent reduction in the decay heat level in the fuel material which is similar to traditional approximations. Isotopes of the noble gases, Kr and Xe, and the elements I and Br are shown to be very mobile and are responsible for a major part of the observed effects. Isotopes of the elements Cs, Se, Rb, and Te were found to be moderately mobile and contribute to a lesser extent to the observed phenomena. These results obtained from the application of the described model confirm the initial hypothesis that sufficient fission product transport can occur to influence a transient. For these reasons, it is concluded that extension of this model into a multi-cell transient analysis code is warranted

  3. Transport of fission products in matrix and graphite

    International Nuclear Information System (INIS)

    Hoinkis, E.

    1983-06-01

    In the past years new experimental methods were applied to or developed for the investigation of fission product transport in graphitic materials and to characterization of the materials. Models for fission product transport and computer codes for the calculation of core release rates were improved. Many data became available from analysis of concentration profiles in HTR-fuel elements. New work on the effect on diffusion of graphite corrosion, fast neutron flux and fluence, heat treatment, chemical interactions and helium pressure was reported on recently or was in progress in several laboratories. It seemed to be the right time to discuss the status of transport of metallic fission products in general, and in particular the relationship between structural and transport properties. Following a suggestion a Colloquium was organized at the HMI Berlin. Interdisciplinary discussions were stimulated by only inviting a limited number of participants who work in different fields of graphite and fission product transport research. (orig./RW)

  4. ENDF/B-5. Fission Product Yields File

    International Nuclear Information System (INIS)

    Schwerer, O.

    1985-10-01

    The ENDF/B-5 Fission Product Yields File contains a complete set of independent and cumulative fission product yields, representing the final data from ENDF/B-5 as received at the IAEA Nuclear Data Section in June 1985. Yields for 11 fissioning nuclides at one or more neutron incident energies are included. The data are available costfree on magnetic tape from the IAEA Nuclear Data Section. (author). 4 refs

  5. Fission product detection by means of photovoltaic cells

    International Nuclear Information System (INIS)

    Liatard, E.; Akrouf, S.; Bruandet, J.F.; Fontenille, A.; Glasser, F.; Stassi, P.; Tsan Ung Chan

    1988-01-01

    The response of photovoltaic cells to heavy ions and fission products have been tested in-beam. Their main advantages are their extremely low price, their low sensitivity to energetic light ions with respect to fission products, and the possibility to cut and fit them together to any shape without dead zone. The time output signals of a charge sensitive preamplifier connected to these cells allows fast coincidences. A resolution of 12 ns (FWHM) has been measured between two cells. (orig.)

  6. ENDF/B-5 Fission Products Library 1979

    International Nuclear Information System (INIS)

    Schwerer, O.; Lemmel, H.D.

    1981-10-01

    This document summarizes contents and documentation of the 1979 version of the Fission Products File of the ENDF/B Library maintained by the National Nuclear Data Center (NNDC) at the Brookhaven National Laboratory, USA. This file contains numerical neutron reaction data and decay data for 877 fission product nuclides. The entire file or selective retrievals from it can be obtained on magnetic tape from the IAEA Nuclear Data Section. (author)

  7. Spray removal of fission products in PWR containments

    International Nuclear Information System (INIS)

    Grist, D.R.

    1982-11-01

    Models and parameters for assessing the rate and extent of removal of various fission product species are described. A range of droplet sizes and of spray additive options is considered and removal of vapour phase inorganic iodine species, of organic iodides and of aerosols containing fission products is discussed. Aerosol removal is assessed in terms of contributing removal mechanisms and the removal rate modelled as a function of the radius of the aerosol particulate species. (author)

  8. Development of glass ceramics for the incorporation of fission products

    International Nuclear Information System (INIS)

    De, A.K.; Luckscheiter, B.; Lutze, W.; Malow, G.; Schiewer, E.

    1976-01-01

    Spontaneous devitrification of fission-product-containing borosilicate glasses can be avoided by controlled crystallization after melting. Glass ceramics have been developed from a vitrified simulated waste and further improvement of product properties was achieved. In particular perovskite, h-celsian, diopside and eucryptite glass ceramics were prepared. These contained leach resistant host phases which exhibited considerable enrichment of long-lived fission products. All products showed increased impact resistance, but the thermal expansion was only slightly improved

  9. ENDF/B-5 Fission Products Library. Rev. 2

    International Nuclear Information System (INIS)

    Schwerer, O.; Pronyaev, V.G.; Lemmel, H.D.

    1984-07-01

    This document summarizes contents and documentation of the 1984 version of the Fission Products Nuclear Data File of the ENDF/B-5 Library (Rev. 2) maintained by the National Nuclear Data Center (NNDC) at the Brookhaven National Laboratory, USA. This file contains numerical neutron reaction data and decay data for 877 fission product nuclides. The entire file or selective retrievals from it can be obtained on magnetic tape from the IAEA Nuclear Data Section. (author)

  10. Analytical measurements of fission products during a severe nuclear accident

    Science.gov (United States)

    Doizi, D.; Reymond la Ruinaz, S.; Haykal, I.; Manceron, L.; Perrin, A.; Boudon, V.; Vander Auwera, J.; tchana, F. Kwabia; Faye, M.

    2018-01-01

    The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d'Investissement d'Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements) is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium) outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

  11. 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

  12. Analytical measurements of fission products during a severe nuclear accident

    Directory of Open Access Journals (Sweden)

    Doizi D.

    2018-01-01

    Full Text Available The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d’Investissement d’Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

  13. Actinide and Fission Product Partitioning and Transmutation

    International Nuclear Information System (INIS)

    2015-06-01

    The benefits of partitioning and transmutation (P and T) have now been established worldwide and, as a result, many countries are pursuing R and D programmes to advance the technologies associated with P and T. In this context, the OECD Nuclear Energy Agency (NEA) has organised a series of biennial information exchange meetings to provide experts with a forum to present and discuss state-of-the-art developments in the field of partitioning and transmutation since 1990. The OECD Nuclear Energy Agency Information Exchange Meeting on Actinides and Fission Products Partitioning and Transmutation is a forum for experts to present and discuss the state-of-the-art development in the field of P and T. Thirteen meetings have been organised so far and held in Japan, the United States, France, Belgium, Spain, the Republic of Korea and the Czech Republic. This 13. meeting was hosted by Seoul National University (Seoul, Republic of Korea) and was organised in co-operation with the International Atomic Energy Agency (IAEA) and the European Community (EC). The meeting covered strategic and scientific developments in the field of P and T such as: fuel cycle strategies and transition scenarios, the role of P and T in the potential evolution of nuclear energy as part of the future energy mix; radioactive waste management strategies; transmutation fuels and targets; advances in pyro and aqueous separation processes; P and T specific technology requirements (materials, spallation targets, coolants, etc.); transmutation systems: design, performance and safety; impact of P and T on the fuel cycle; fabrication, handling and transportation of transmutation fuels. A total of 103 presentations (39 oral and 64 posters) were discussed among the 110 participants from 19 countries and 2 international organisations. The meeting consisted of one plenary session where national and international programmes were presented followed by 5 technical sessions: - Fuel Cycle Strategies and Transition

  14. Measurement of the hydrogen yield in the radiolysis of water by dissolved fission products

    International Nuclear Information System (INIS)

    Sauer, M.C. Jr.; Hart, E.J.; Flynn, K.F.; Gindler, J.E.

    1976-04-01

    Hydrogen from the radiolysis of water by dissolved fission products is stripped from the solution and collected by bubbling CO 2 through the solution. Quantitative measurements of the G value for hydrogen show that the yield is essentially the same as would be obtained by external gamma radiolysis of nonradioactive solutions of the same chemical composition. The hydrogen yield can be enhanced by addition of a hydrogen-atom donor, such as formic acid, to the solution. The yield of hydrogen from fission-waste solutions is discussed with respect to the question of whether it represents a significant energy source

  15. 216-S-1 and S-2 mixed-fission-product crib-characterization study

    Energy Technology Data Exchange (ETDEWEB)

    Van Luik, A. E.; Smith, R. M.

    1982-03-01

    The 216-S-1 and 2 crib is an underground structure that was used for the disposal of radioactively contaminated liquid waste at the Hanford Site. The crib received acidic, intermediate level, mixed fission-product waste solutions from 1952 to 1956. The 1980 status of radioactive contaminants in the sediment beneath the crib was investigated. The results indicate that the radionuclide distributions are stable, with no evidence of significant translocations found since the late 1960's.

  16. 216-S-1 and S-2 mixed-fission-product crib-characterization study

    International Nuclear Information System (INIS)

    Van Luik, A.E.; Smith, R.M.

    1982-03-01

    The 216-S-1 and 2 crib is an underground structure that was used for the disposal of radioactively contaminated liquid waste at the Hanford Site. The crib received acidic, intermediate level, mixed fission-product waste solutions from 1952 to 1956. The 1980 status of radioactive contaminants in the sediment beneath the crib was investigated. The results indicate that the radionuclide distributions are stable, with no evidence of significant translocations found since the late 1960's

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-12-15

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

  18. Partitioning and transmutation of actinides and fission products

    International Nuclear Information System (INIS)

    Baetsle, L.H.

    1993-01-01

    The world's nuclear power plants have a total installed capacity of approximately 340 GWe. They give rise to an annual volume of approximately 9000 t of radioactive waste, which is reprocessed, separated from its plutonium content, contained, and stored in repositories to close the nuclear fuel cycle. Direct disposal is being discussed as an alternative to this procedure. As repositories in suitable types of host rock are not operational, the only viable solution is long-term interim storage above ground. If the volumes of radioactive waste are to be reduced, the longlived actinides and fission products must be partitioned. Isotope partitioning in accelerators, though still sounding like science fiction, may soon be indispensable as the third way of treating radioactive waste. The use of mixed oxide fuel in light water reactors and fast breeder reactors both help to limit waste arisings and protect the long-term continuity of raw materials supply. However, both require public acceptance if they are to succeed. (orig.) [de

  19. Fission-product releases from a PHWR terminal debris bed

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.J.; Bailey, D.G., E-mail: morgan.brown@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    During an unmitigated severe accident in a pressurized heavy water reactor (PHWR) with horizontal fuel channels, the core may disassemble and relocate to the bottom of the calandria vessel. The resulting heterogeneous in-vessel terminal debris bed (TDB) would likely be quenched by any remaining moderator, and some of the decay heat would be conducted through the calandria vessel shell to the surrounding reactor vault or shield tank water. As the moderator boiled off, the solid debris bed would transform into a more homogeneous molten corium pool located between top and bottom crusts. Until recently, the severe accident code MAAP-CANDU assumed that unreleased volatile and semi-volatile fission products remained in the TDB until after calandria vessel failure, due to low diffusivity through the top crust and the lack of gases or steam to flush released fission products from the debris. However, national and international experimental results indicate this assumption is unlikely; instead, high- and medium-volatility fission products would be released from a molten debris pool, and their volatility and transport should be taken into account in TDB modelling. The resulting change in the distribution of fission products within the reactor and containment, and the associated decay heat, can have significant effects upon the progression of the accident and fission-product releases to the environment. This article describes a postulated PHWR severe accident progression to generate a TDB and the effects of fission-product releases from the terminal debris, using the simple release model in the MAAP-CANDU severe accident code. It also provides insights from various experimental programs related to fission-product releases from core debris, and their applicability to the MAAP-CANDU TDB model. (author)

  20. Novel Fission-Product Separation Based on Room-Temperature Ionic Liquids

    International Nuclear Information System (INIS)

    Rogers, Robin D.

    2004-01-01

    This project has demonstrated that Sr2+ and Cs+ can be selectively extracted from aqueous solutions into ionic liquids using crown ethers and that unprecedented large distribution coefficients can be achieved for these fission products. The volume of secondary wastes can be significantly minimized with this new separation technology. Through the current EMSP funding, the solvent extraction technology based on ionic liquids has been shown to be viable and can potentially provide the most efficient separation of problematic fission products from high level wastes. The key results from the current funding period are the development of highly selective extraction process for cesium ions based on crown ethers and calixarenes, optimization of selectivities of extractants via systematic change of ionic liquids, and investigation of task-specific ionic liquids incorporating both complexant and solvent characteristics

  1. 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

  2. Trapping technology for gaseous fission products from voloxidation process

    International Nuclear Information System (INIS)

    Shin, Jin Myeong; Park, J. J.; Park, G. I.; Jung, I. H.; Lee, H. H.; Kim, G. H.; Yang, M. S.

    2005-05-01

    The objective of this report is to review the different technologies for trapping the gaseous wastes containing Cs, Ru, Tc, 14 C, Kr, Xe, I and 3 H from a voloxidation process. Based on literature reviews and KAERI's experimental results on the gaseous fission products trapping, appropriate trapping method for each fission product has been selected considering process reliability, simplicity, decontamination factor, availability, and disposal. Specifically, the most promising trapping method for each fission product has been proposed for the development of the INL off-gas trapping system. A fly ash filter is proposed as a trapping media for a cesium trapping unit. In addition, a calcium filter is proposed as a trapping media for ruthenium, technetium, and 14 C trapping unit. In case of I trapping unit, AgX is proposed. For Kr and Xe, adsorption on solid is proposed. SDBC (Styrene Divinyl Benzene Copolymer) is also proposed as a conversion media to HTO for 3 H. This report will be used as a useful means for analyzing the known trapping technologies and help selecting the appropriate trapping methods for trapping volatile and semi-volatile fission products, long-lived fission products, and major heat sources generated from a voloxidation process. It can also be used to design an off-gas treatment system

  3. Trapping technology for gaseous fission products from voloxidation process

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Jin Myeong; Park, J. J.; Park, G. I.; Jung, I. H.; Lee, H. H.; Kim, G. H.; Yang, M. S

    2005-05-15

    The objective of this report is to review the different technologies for trapping the gaseous wastes containing Cs, Ru, Tc, {sup 14}C, Kr, Xe, I and {sup 3}H from a voloxidation process. Based on literature reviews and KAERI's experimental results on the gaseous fission products trapping, appropriate trapping method for each fission product has been selected considering process reliability, simplicity, decontamination factor, availability, and disposal. Specifically, the most promising trapping method for each fission product has been proposed for the development of the INL off-gas trapping system. A fly ash filter is proposed as a trapping media for a cesium trapping unit. In addition, a calcium filter is proposed as a trapping media for ruthenium, technetium, and {sup 14}C trapping unit. In case of I trapping unit, AgX is proposed. For Kr and Xe, adsorption on solid is proposed. SDBC (Styrene Divinyl Benzene Copolymer) is also proposed as a conversion media to HTO for {sup 3}H. This report will be used as a useful means for analyzing the known trapping technologies and help selecting the appropriate trapping methods for trapping volatile and semi-volatile fission products, long-lived fission products, and major heat sources generated from a voloxidation process. It can also be used to design an off-gas treatment system.

  4. 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

  5. CINDER, Depletion and Decay Chain Calculation for Fission Products in Thermal Reactors

    International Nuclear Information System (INIS)

    England, T.R.; Gorrell, T.C.; Hightower, J.H.

    2001-01-01

    1 - Description of problem or function: CINDER is a four-group, one- point depletion and fission product program based on the evaluation of a general analytical solution of nuclides coupled in any linear sequence of radioactive decays and neutron absorptions in a specified neutron flux spectrum. The desired depletion and fission product chains and all physical data are specified by the problem originator. The program computes individual nuclide number densities, activities, nine energy-group disintegration rates, and macroscopic and barns/fission poisons at each time-step as well as selected summaries of these data. 2 - Method of solution: Time-dependent variations in nuclide cross sections and neutron fluxes are approximated by a user-specified sequential set of values which are considered constant during the duration of the user-specified associated time-increments. When a nuclide concentration is independent of the concentration of any of its progeny, it is possible to resolve the couplings so as to obtain nuclides fed by a single parent. These chains are referred to as linear. 3 - Restrictions on the complexity of the problem: The program is limited to 500 total nuclides formed in up to 240 chains of 20 or fewer nuclides each. Up to 10 nuclides may act as fission product sources, contributing to power, and as many as 99 time-steps of arbitrary length are permitted. All stable nuclides must have a cross section if zero power time-increments are anticipated

  6. Fission product removal from molten salt using zeolite

    International Nuclear Information System (INIS)

    Pereira, C.; Babcock, B.D.

    1996-01-01

    Spent nuclear fuel (SNF) can be treated in a molten salt electrorefiner for conversion into metal and mineral waste forms for geologic disposal. The fuel is dissolved in molten chloride salt. Non-transuranic fission products in the molten salt are ion-exchanged into zeolite A, which is subsequently mixed with glass and consolidated. Zeolite was found to be effective in removing fission product cations from the molten salt. Breakthrough of cesium and the alkaline earths occurred more rapidly than was observed for the rare earths. The effluent composition as a function of time is presented, as well as results for the distribution of fission products along the length of the column. Effects of temperature and salt flow rate are also discussed

  7. 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}.

  8. Map of calculated radioactivity of fission product, (4)

    International Nuclear Information System (INIS)

    Takeda, Tsuneo

    1978-07-01

    The overall radioactivities of fission products depending on irradiation time and cooling time were calculated for 18 different neutron fluxes, which are presented in contour maps and tables. Irradiation condition etc. are the followings: neutron flux (n sub(th)) 1 x 10 12 - 6.8 x 10 14 n/cm 2 /sec, uranium quantity 1 mole (6 x 10 23 atoms, ca. 271 g UO 2 ), U-235 enrichment 2.7%, irradiation time 60. - 6 x 10 7 sec (1 min - 1.9 y), cooling time 0. and 60. - 6 x 10 7 sec (1 min - 1.9 y). The enrichment value represents those for LWRs. To calculate the overall radioactivities, 595 fission product nuclides were introduced. Overall radioactivities calculations were made for 68,000 combinations of irradiation time, cooling time and neutron flux. The many complex decay chains of fission products were treated with CODAC-No.6 computer code. (author)

  9. The release of fission products from uranium metal: a review

    International Nuclear Information System (INIS)

    Minshall, P.C.

    1989-03-01

    The literature on the release of fission products as gaseous species from irradiated uranium metal in oxidising atmospheres has been reviewed. Release of actinides and of fission products as spalled particulate were not considered. Data is given on the release in air, carbon dioxide, steam and mixtures of steam and air. The majority of data discussed lie between 800 and 1200 0 C though some results for xenon, krypton and iodine releases below 800 0 C are given. Two measures of fission product release are discussed: the release fraction, F(tot), which is the ratio of the total release to the initial inventory, and the fractional release, F(ox), which is the fraction released from the oxidised metal. The effect of burn-up, atmosphere and temperature on F(tot) and F(ox) is examined and the conditions under which the release fraction, F(tot) is proportional to the extent of oxidation discussed. (author)

  10. 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.

  11. 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

  12. 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)

  13. Fission product chemistry in severe nuclear reactor accidents

    International Nuclear Information System (INIS)

    Nichols, A.L.

    1990-09-01

    A specialist's meeting was held at JRC-Ispra from 15 to 17 January 1990 to review the current understanding of fission-product chemistry during severe accidents in light water reactors. Discussions focussed on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. Recommendations for future chemistry work were made covering the following areas: (a) fuel degradation and fission-product release, (b) transport and attenuation processes in the reactor coolant system, (c) containment chemistry (iodine behaviour and core-concrete interactions)

  14. The chemistry of fission products for accident analysis

    International Nuclear Information System (INIS)

    Potter, P.E.

    1985-01-01

    Current knowledge concerning the chemical state of the fission product elements during the development of accidents in water reactor systems is reviewed in this paper. The fission products elements which have been considered are Cs, I, Te, Sr and Ba but aspects of the behaviour of Mo, Ru and the lanthanides are also discussed. Some features of the reactions of the various species of these elements with other components of the reactor systems are described. The importance of having an adequate knowledge of thermodynamic data and phase equilibria of relatively simple systems in order to interpret experimental observations on complex multi-component systems is stressed

  15. Integral measurement of fission products capture in fast breeder reactors

    International Nuclear Information System (INIS)

    Martin Deidier, Loick.

    1979-12-01

    For the SUPERPHENIX reactor project, it was necessary to know fission products capture with about 10% accuracy in the fast breeder reactor spectra. In this purpose, integral measurements have been carried out on the main separated products by different experimental technics (oscillation, activation and irradiation methods), but particularly on irradiated fuel pins from RAPSODIE and PHENIX reactors in order to directly obtain total effect of fission products. Same tendencies have been observed for both enriched uranium fuel and LMFBR characteristic plutonium fuel. All experimental results have been introduced in CARNAVAL cross section set [fr

  16. Status report on actinide and fission product transmutation studies

    International Nuclear Information System (INIS)

    1997-06-01

    The management of radioactive waste is one of the key issues in today's political and public discussions on nuclear energy. One of the fields that looks into the future possibilities of nuclear technology is the neutronic transmutation of actinides and of some most important fission products. Studies on transmutation of actinides are carried out in various countries and at an international level. This status report which gives an up-to-date general overview of current and planned research on transmutation of actinides and fission products in non-OECD countries, has been prepared by a Technical Committee meeting organized by the IAEA in September 1995. 168 refs, 16 figs, 34 tabs

  17. Status of the French research programme for actinides and fission products partitioning and transmutation

    International Nuclear Information System (INIS)

    Warin, D.

    2003-01-01

    The paper focus on separation and transmutation research and development programme and main results over these ten last years. The massive research programme on enhanced separation, conducted by CEA and supported by broad international cooperation, has recently achieved some vital progress. Based on real solutions derived from the La Hague process, the CEA demonstrated the lab-scale feasibility of extracting minor actinides and some fission products (I, Cs and Tc) using an hydrometallurgical process that can be extrapolated on the industrial scale. The CEA also conducted programmes proving the technical feasibility of the elimination of minor actinides and fission products by transmutation: fabrication of specific targets and fuels for transmutation tests in the HFR and Phenix reactors, neutronics and technology studies for ADS developments in order to support the MEGAPIE, TRADE and MYRRHA experiments and the future 100 MW international ADS demonstrator. Scenarios studies aimed at stabilizing the inventory with long-lived radionuclides, plutonium, minor actinides and certain long-lived fission products in different nuclear power plant parks and to verify the feasibility at the level of the cycle facilities and fuels involved in those scenarios. Three French Research Groups CEA-CNRS carry out partitioning (PRACTIS) and transmutation (NOMADE and GEDEON) more basic studies. (author)

  18. Microbial Transformations of Actinides and Fission Products in Radioactive Waste

    Energy Technology Data Exchange (ETDEWEB)

    Francis, A. J. [Pohang Univ. Science and Technology, Pohang (Korea, Republic of)

    2011-07-01

    The environmental factors that can affect microbial growth and activity include moisture, temperature, ph, Eh, availability of organic and inorganic nutrients, and radiation. The microbial activity in a specific repository is influenced by the ambient environment of the repository, and the materials to be emplaced. For example, a repository in unsaturated igneous rock formations such as volcanic tuff rocks at Yucca Mountain is generally expected to be oxidizing; a repository in a hydrologically expected to be oxidizing; a repository in a hydrologically saturated zone, especially in sedimentary rocks, could be reducing. Sedimentary rocks contain a certain amount of organic matter, which may stimulate microbial activities and, thus maintain the repository and its surrounding areas at reducing conditions. Although the impacts of microbial activity on high-level nuclear waste and the long-term performance of the repository have not fully investigated, little microbial activity is expected in the near-field because of the radiation, lack of nutrients and the harsh conditions. However in the far-field microbial effects could be significant. Much of our understanding of the microbial effects on radionuclides stems from studies conducted with selected transuranic elements and fission products and limited studies with low-level radioactive wastes. Significant aerobic- and anaerobic-microbial activity is expected to occur in the waste because of the presence of electron donors and acceptors. The actinides initially may be present as soluble- or insoluble-forms but, after disposal, may be converted from one to the other by microorganisms. The direct enzymatic or indirect non-enzymatic actions of microbes could alter the speciation, solubility, and sorption properties of the actinides, thereby increasing or decreasing their concentrations in solution.

  19. BIG-10 fission product generation and reaction rates

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1976-01-01

    Fission product generation rates for high quality fission foils and reaction rates of nonfission foils have been measured by gamma ray activation analyses. These foils were irradiated in the BIG-10 facility and the activities were measured by NaI counting techniques

  20. ENDF/B-6 fission-product yield sublibraries

    International Nuclear Information System (INIS)

    Lemmel, H.D.

    1994-01-01

    The contents and the documentation of the ENDF/B-6 fission-product yield sublibraries which were released in 1991 and updated in 1993, are summarized. Copies of the data libraries are available on magnetic tape of PC diskettes from the IAEA Nuclear Data Section, costfree upon request. (author). 1 tab

  1. Fission product retention during faults involving steam generator tube rupture

    International Nuclear Information System (INIS)

    Rodliffe, R.S.

    1983-08-01

    In some PWR fault conditions, such as stuck open safety relief valve in the secondary circuit or main steam line break, the release of fission products to the atmosphere may be increased by the leakage of primary coolant into the secondary circuit following steam generator tube rupture. The release may be reduced by retention either within the primary circuit or within the affected steam generator unit (SGU). The mechanisms leading to retention are reviewed and quantified where possible. The parameters on which any analysis will be most critically dependent are identified. Fission product iodine and caesium may be retained in the secondary side of a SGU either by partition to retained water or by droplet deposition on surfaces and subsequent evaporation to dryness. Two extreme simplifications are considered: SGU 'dry', i.e. the secondary side is steam filled, and SGU 'wet', i.e. the tube bundle is covered with water. Consideration is given to: the distribution of fission products between gaseous and aerosol forms; mechanisms for droplet formation, deposition and resuspension; fission product retention during droplet or film evaporation primary coolant mixing and droplet scrubbing in a wet SGU; and the performance of moisture separators and steam driers. (author)

  2. Applications for fission product data to problems in stellar nucleosynthesis

    International Nuclear Information System (INIS)

    Mathews, G.J.

    1983-10-01

    A general overview of the nucleosynthesis mechanisms for heavy (A greater than or equal to 70) nuclei is presented with particular emphasis on critical data needs. The current state of the art in nucleosynthesis models is described and areas in which fission product data may provide useful insight are proposed. 33 references, 10 figures

  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 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.

  5. Progress in fission product nuclear data. Issue no. 6

    International Nuclear Information System (INIS)

    Lammer, G.; Lammer, M.

    1980-06-01

    This is the sixth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed

  6. Applications of nuclear data on short-lived fission products

    International Nuclear Information System (INIS)

    Rudstam, G.; Aagaard, P.; Aleklett, K.; Lund, E.

    1981-01-01

    The study of short-lived fission products gives information about the nuclear structure on the neutron-rich side of stability. The data are also of interest for various applications both to basic science and to nuclear technology. Some of these applications, taken up by the OSIRIS group at Studsvik, are described in the present contribution. (orig.)

  7. Calculation of vapor pressure of fission product fluorides and oxyfluorides

    International Nuclear Information System (INIS)

    Roux, J.P.

    1976-03-01

    The equilibrium diagrams of the condensed phases - solid and liquid - and vapor phase are collected for the principal fluorides and oxyfluorides of fission product elements (atomic number from 30 to 66). These diagrams are used more particularly in fuel reprocessing by fluoride volatility process. Calculations and curves (vapor pressure in function of temperature) are processed using a computer program given in this report [fr

  8. Reactivity effects of fission product decay in PWRs

    International Nuclear Information System (INIS)

    Aragones, J.M.; Ahnert, C.

    1988-01-01

    The purpose of the work reported in this paper is to analyze the effects of fission product chains with radioactive decay on the reactivity in pressurized water reactor (PWR) cores, calculating their accumulation and absorption rates along fuel burnup at continuous operation and after shutdown periods extending from 1 day to a few months. The authors PWR version of the WIMS-D4 code is first used to obtain the individual number densities, absorption rates, and averaged cross sections for every nuclide of the fission product chains with significant decay rates, as a function of fuel burnup at continuous irradiation. Next, by an auxiliary ad hoc code, these data, have been processed together with the required one for fissile nuclides and boron, also taken from WIMS at each burnup step, to calculate the average or effective values relevant for the analysis and the decay and change in overall absorption after several shutdown times. (1) The reactivity effect of fission product decay changes significantly with the shutdown time. The maximum absorption increase by decay is reached in ∼ 10 days' shutdown. (2) The dependence with fuel type, enrichment, and burnup is slight, but the change with previous power density is nearly linear, which might be significant after coast-down in previous cycles. (3) For long shutdown periods, the overall reactivity effect of decay in the three fission product chains considered is much less than if only the samarium peak due to 149 Nd is considered

  9. The potential for large scale uses for fission product xenon

    International Nuclear Information System (INIS)

    Rohrmann, C.A.

    1983-01-01

    Of all fission products in spent, low enrichment, uranium, power reactor fuels xenon is produced in the highest yield - nearly one cubic meter, STP, per metric ton. In aged fuels which may be considered for processing in the U.S. radioactive xenon isotopes approach the lowest limits of detection. The separation from accompanying radioactive 85 Kr is the essential problem; however, this is state of the art technology which has been demonstrated on the pilot scale to yield xenon with pico-curie levels of 85 Kr contamination. If needed for special applications, such levels could be further reduced. Environmental considerations require the isolation of essentially all fission product krypton during fuel processing. Economic restraints assure that the bulk of this krypton will need to be separated from the much more voluminous xenon fraction of the total amount of fission gas. Xenon may thus be discarded or made available for uses at probably very low cost. In contrast with many other fission products which have unique radioactive characteristics which make them useful as sources of heat, gamma and x-rays and luminescence as well as for medicinal diagnostics and therapeutics fission product xenon differs from naturally occurring xenon only in its isotopic composition which gives it a slightly higher atomic weight, because of the much higher concentrations of the 134 X and 136 Xe isotopes. Therefore, fission product xenon can most likely find uses in applications which already exist but which can not be exploited most beneficially because of the high cost and scarcity of natural xenon. Unique uses would probably include applications in improved incandescent light illumination in place of krypton and in human anesthesia

  10. Repeated radiation injuries by fission products

    International Nuclear Information System (INIS)

    Vasilenko, I.Ya.

    1986-01-01

    Attention is given to repeated radiation injuries during internal irradiation of theoretical and practical interest, particularly in case of the intake into organism of young products of nuclear fission (PNF). The results of experiments with dogs with repeated radioactive iodine injury the isotopes of which (131-135sub(I)) constitute a considerable part of PNF activity are discussed. The blood reaction and protein metabolism state have been studied. Observations for dogs have been continued for about 4 years. The doses for thyroid, gastrointestinal tract and liver subjected to the most intensive irradiation consituted in the first series of experiments after the first intake about 3;0.3;0.05 Gy, after the second - 5;0.5;0.08 Gy and in the second series of experiments - 3;0.3;0.05 Gy and 0.6;0.06;0.01 Gy, respectively. Hematologic factors,thyroid function, changes in exchange and immunologic reactivity have been studied. The dogs have been under observation for 5 years. It is shown in case of repeated intake of Isup(131) PNF into animals organism in quantity which does not cause during the acute period a clinically outlined sickness, substantial differences in the organism reaction as compared with the first intake of radionuclides have not been found. The presence of residual radiation injuries did not cause charging action during the acute period during PNF and repeated intake which in the author's opinion testifies to perfection of compensator mechanisms in case of intake of such quantities of radioactive products. At the remote periods blastomogenic action manifested which is estimated as a result of general biological action of radionuclides administered to the organism. The necessity in subsequent investigations for obtaining the data on organism reactivity, clinic and pathogenesis with the aim of prophylaxis and treatment of such injuries is indicated

  11. Fission-Product Separation as a Final Solution to the Problem of Storing Highly Radioactive Waste; La Separation des Produits de Fission Comme Solution Definitive du Stockage des Dechets Fortement Radioactifs; 0412 042b 0414 ; La Separacion de los Productos de Fision Como Solucion Definitiva del Problema del Almacenamiento de los Desechos de Elevada Radiactividad

    Energy Technology Data Exchange (ETDEWEB)

    Raggenbass, A.; Lefevre, J. [Centre d' Etudes Nucleaires, Saclay (France)

    1963-02-15

    At the 1955 Geneva Conference; it was suggested by Glueckauf that the problems presented by solutions of concentrated fission products could be dealt with by isolating caesium-137 and strontium-90 with a decontamination factor of at least 10{sup 4} so that, after conventional effluent processing, disposal could be effected in 10 to 20 years' time. This paper considers the suggestion in the light of the present situation. Thanks to the progress made in methods of separating caesium and strontium it now seems feasible to obtain this factor of 10{sup 4} under industrial conditions. The following methods are very promising: (a) For caesium-137: precipitation by phosphotungstic acid, exchange using an ammonium salt of a heteropoly-acid, fixation using a zirconium phosphate and (b) For strontium-90: oxalate precipitation, extraction by di-2 ethyl hexylphosphoric acid, exchange using various salts, strontium sulphate, etc. This progress in chemical techniques, backed up by advances in the industrial processing of effluents, indicate that Glueckauf's proposal is technically possible. In addition, the increasing use of large quantities of fission products, particularly in the SNAP programmes, justifies the outlay of capital, which will be recovered at least in part by the sale of the products made. There is no doubt that refractory sources with a very high resistance to attack by all external agents, constitute the most satisfactory means of storage for products so dangerous and of such long half-life. We believe, therefore, that both technically and psychologically, storing by means of concentrated sources with strong cladding is the most reliable method. A comparison of this method with that of fixation in glass is worthy of full and detailed examination. (author) [French] Glueckauf avait formule, a la Conference de Geneve en 1955, l'hypothese qu'une solution definitive aux problemes poses par les solutions de produits de fission concentres pouvait etre trouvee dans l

  12. Volatility of ruthenium during vitrification operations on fission products. part 1. nitric solutions distillation concentrates calcination. part 2. fixation on a steel tube. decomposition of the peroxide; Volatilite du ruthenium au cours des operations de vitrification des produits de fission. 1. partie distillation des solutions nitriques calcination des concentrats 2. partie fixation sur un tube d'acier decomposition du peroxyde

    Energy Technology Data Exchange (ETDEWEB)

    Ortins de Bettencourt, A; Jouan, A [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1969-07-01

    During vitrification of fission product solutions, a high percentage of the ruthenium initially present in these solutions in the form of nitrosyl-ruthenium nitrates is volatilized with the production of the peroxide which itself is decomposed to ruthenium dioxide. The aim of this work has been to study the volatility of the ruthenium during the vitrification processes. During the distillation of the nitric solutions, we have studied in particular the influence on the volatility of the temperature , of the chemical form in which the ruthenium is introduced, of the bubbling of a gas through the solution, of the nitric concentration and of the nitrate concentration. During the calcination, we have observed the influence of the temperature, of the time, of the flow rate and of the nature of the carrier gas, as well as of the action of the ruthenium bi-oxide and the iron oxide on the volatility of the ruthenium. Part 2. This report deals with the study of the thermal decomposition of ruthenium peroxide, RuO{sub 4}, and its deposition on steel tubing. After a brief bibliographic review of the various properties of this substance, a study is made, in the first part, of its deposition on a steel tube. In order to do this, we pass a gas current containing RuO{sub 4} marked with {sup 106}Ru through a stainless steel tube subjected to a temperature gradient which decreases in the direction of the gas flow. The temperature at which RuO{sub 4} is deposited or is fixed on the tube is determined and the influence of the flow rate on this deposit is studied. In the second part, an attempt has been made to study by a static method the kinetics of the decomposition reaction of ruthenium peroxide to give the dioxide: RuO{sub 4} {yields} RuO{sub 2} + O{sub 2}. To do this, we have tried to introduce gaseous RuO{sub 4} into a container placed in an electric oven, and to follow the reaction by {gamma} counting. (author) [French] Au cours de la vitrification des solutions de produits de

  13. Volatility of ruthenium during vitrification operations on fission products. part 1. nitric solutions distillation concentrates calcination. part 2. fixation on a steel tube. decomposition of the peroxide; Volatilite du ruthenium au cours des operations de vitrification des produits de fission. 1. partie distillation des solutions nitriques calcination des concentrats 2. partie fixation sur un tube d'acier decomposition du peroxyde

    Energy Technology Data Exchange (ETDEWEB)

    Ortins de Bettencourt, A.; Jouan, A. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1969-07-01

    During vitrification of fission product solutions, a high percentage of the ruthenium initially present in these solutions in the form of nitrosyl-ruthenium nitrates is volatilized with the production of the peroxide which itself is decomposed to ruthenium dioxide. The aim of this work has been to study the volatility of the ruthenium during the vitrification processes. During the distillation of the nitric solutions, we have studied in particular the influence on the volatility of the temperature , of the chemical form in which the ruthenium is introduced, of the bubbling of a gas through the solution, of the nitric concentration and of the nitrate concentration. During the calcination, we have observed the influence of the temperature, of the time, of the flow rate and of the nature of the carrier gas, as well as of the action of the ruthenium bi-oxide and the iron oxide on the volatility of the ruthenium. Part 2. This report deals with the study of the thermal decomposition of ruthenium peroxide, RuO{sub 4}, and its deposition on steel tubing. After a brief bibliographic review of the various properties of this substance, a study is made, in the first part, of its deposition on a steel tube. In order to do this, we pass a gas current containing RuO{sub 4} marked with {sup 106}Ru through a stainless steel tube subjected to a temperature gradient which decreases in the direction of the gas flow. The temperature at which RuO{sub 4} is deposited or is fixed on the tube is determined and the influence of the flow rate on this deposit is studied. In the second part, an attempt has been made to study by a static method the kinetics of the decomposition reaction of ruthenium peroxide to give the dioxide: RuO{sub 4} {yields} RuO{sub 2} + O{sub 2}. To do this, we have tried to introduce gaseous RuO{sub 4} into a container placed in an electric oven, and to follow the reaction by {gamma} counting. (author) [French] Au cours de la vitrification des solutions de produits de

  14. 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

    generally in the fuel element (pebble or compact) aims at estimating the source term of the fission products release outside the fuel element in normal operation or accidental conditions. In ATLAS, the transport mechanisms are modelled in a single transport law using effective diffusion coefficients for the fission product species in the different constitutive materials. The Verification and Validation of ATLAS code rests on two main steps: - Testing plan on fuel particle thermal mechanical behaviour has been carried out regarding sensitivity on dimensional parameters and physical properties such as kernel diameter, density and layer thicknesses and pyrocarbon layer anisotropy. The obtained results allow justifying and specifying the design for the manufacture. - Regarding fission product release under core heat-up accident conditions, the IAEA Coordinated Research Project 6 on 'Advanced in HTGR Fuel Technology Development' benchmark is the basis of the ATLAS code verification step. The ATLAS results obtained on IAEA benchmark cases with analytical solutions demonstrate that the models used fit the physical, chemical and mathematical laws. Regarding past irradiation tests and heating tests, ATLAS results show good agreement with the experimental database measurements. Comparison between ATLAS code results with analytical and experimental data allows defining confidence zones where ATLAS code gives accurate results and critical limits. These limits show where R and D efforts on models and material properties are needed to refine laws and models. (author)

  15. Study on the calculation method of source term from fission products

    International Nuclear Information System (INIS)

    Zhou Jing; Gong Quan; Qiu Haifeng

    2014-01-01

    As a major part of radioactive nuclides, fission products play an important role in nuclear power plant design. The paper analyzes the calculation model of core activity inventory, the model of fission products releasing from the pellets to RCS, the balance model of fission products in RCS, and then proves them by calculation of the typical pressurized water reactor. The model is proved applicable for calculating fission products of pressurized water reactors. (authors)

  16. Proposal to represent neutron absorption by fission products by a single pseudo-fragment

    International Nuclear Information System (INIS)

    Tsibulya, A.M.; Kochetkov, A.L.; Kravchenko, I.V.; Nikolaev, M.N.

    1991-01-01

    The concentration of fission products during reactor operation is analyzed. The dependence of a composite fission product capture cross-section as a function of time and on the nature of the A of the fissile nuclide are investigated, and the neutron radiative capture in fission products of a thermal reactor is evaluated. It is concluded that neutron absorption by fission products can be described by pseudo-fragments. (author). 18 refs, 2 figs, 3 tabs

  17. Progress in fission product nuclear data. Information about activities in the field of measurements and compilations/evaluations of fission product nuclear data (FPND)

    International Nuclear Information System (INIS)

    Lammer, G.

    1978-07-01

    This is the fourth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); neutron reaction cross sections of fission products; data related to the radioactive decay of fission products; delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.)

  18. Simulation of fission products behavior in severe accidents for advanced passive PWR

    International Nuclear Information System (INIS)

    Tong, L.L.; Huang, G.F.; Cao, X.W.

    2015-01-01

    Highlights: • A fission product analysis model based on thermal hydraulic module is developed. • An assessment method for fission product release and transport is constructed. • Fission products behavior during three modes of containment response is investigated. • Source term results for the three modes of containment response are obtained. - Abstract: Fission product behavior for common Pressurized Water Reactor (PWR) has been studied for many years, and some analytical tools have developed. However, studies specifically on the behavior of fission products related to advanced passive PWR is scarce. In the current study, design characteristics of advanced passive PWR influencing fission product behavior are investigated. An integrated fission products analysis model based on a thermal hydraulic module is developed, and the assessment method for fission products release and transport for advanced passive PWR is constructed. Three modes of containment response are simulated, including intact containment, containment bypass and containment overpressure failure. Fission products release from the core and corium, fission products transport and deposition in the Reactor Coolant System (RCS), fission products transport and deposition in the containment considering fission products retention in the in-containment refueling water storage tank (IRWST) and in the secondary side of steam generators (SGs) are simulated. Source term results of intact containment, containment bypass and containment overpressure failure are obtained, which can be utilized to evaluate the radiological consequences

  19. 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.

  20. Separation and utilization of fission products considering economic aspects

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  1. Calculated apparent yields of rare gas fission products

    International Nuclear Information System (INIS)

    Delucchi, A.A.

    1975-01-01

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

  2. Tracking of fission products release during refueling operations

    International Nuclear Information System (INIS)

    Agarwal, Sharad; Prajapat, M.K.; Vyas, Shyam; Hussain, S.A.

    2001-01-01

    It has been always observed that the release of fission products increase during refueling operations. At RAPP-3 and 4 an attempt has been made to follow-up the change in fission products activity release at each stage of refueling operation and quantification of concentrations of various radionuclides. This exercise was also extended to refueling operation of the channels containing suspected failed fuel. A level of FPNG ( 133 Xe) was observed to increase by a factor of about 10-40 during refueling of failed channel as compared to healthy channel. It can be concluded that by monitoring FPNG levels in exhaust status of the healthiness of spent fuel can be found out. This report discusses in detail the experiment conducted for this purpose. (author)

  3. 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

  4. Release of fission products from contaminated sodium fires

    International Nuclear Information System (INIS)

    Jordan, S.

    1976-01-01

    Leaks in the primary coolant system of a LMFBR and also serious incidents with tank rupture may entail the escape of fission products into the containment of the reactor. For incident analysis it is important to know the retention capability of sodium for the different fission products. The release of cesium and strontium from pools contaminated with 100 to 1000 ppM was investigated by experiments. The cesium content of airborne aerosols depends on oxygen concentration: at 21 percent oxygen concentration the Cs content of sodium-oxide aerosols is 3 times and at 0.5 percent 15 times as high as the initial Cs concentration in the pool. Strontium content of aerosols over burning contaminated sodium pools is 10 3 times smaller than the strontium pool concentration

  5. 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.)

  6. 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

  7. Modification of the fission product inventory program FISPIN

    International Nuclear Information System (INIS)

    Thomas, R.B.

    1977-05-01

    The fission product inventory program FISPIN calculates inventories of fission products, actinides and activation products, during and after irradiation in a nuclear reactor, estimates also being given for heat output and radioactive activity of the isotopes. The program has been developed further by making provision for the simulation of fuel reprocessing and in providing subroutines to make the program compatible with nuclear data in a slightly modified ENDF/B4 format. Continuous development of FISPIN over the years has however involved many program alterations and additions, and this has resulted in a generally untidy and cumbersome program. An attempt has therefore been made to improve the basic structure of the program. The subject is dealt with under the following headings: modularisation, direct access data, override facility, selective output, flowcharts, summary. (U.K.)

  8. Irradiated uranium reprocessing, Final report I-VI, Part VI - Separation of uranium, plutonium and fission products from HNO{sub 3} solution on the zirconium phosphate (part I), Adsorption equilibrium and kinetics; Prerada ozracenog urana. Zavrani izvestaj - I-VI, VI Deo - Odvajanje urana, plutonijuma i fisionih produkata iz rastvora HNO{sub 3} na cirkonijum fosfatu (deo I.), Ravnoteza i kinetika adsorpcije

    Energy Technology Data Exchange (ETDEWEB)

    Gal, I; Ruvarac, A [Institute of Nuclear Sciences Boris Kidric, Odeljenje za eksploataciju nuklearnog goriva, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Separation of uranium, plutonium and long-lived fission products was investigated on a inorganic ion exchanger. Zirconium phospate was chosen for this purpose because its ion exchanger properties were well known. This report deals with the study of equilibrium and kinetics of the adsorption.

  9. 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.

  10. 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)

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

    International Nuclear Information System (INIS)

    Strickert, R.G.

    1976-01-01

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

  12. Fission product source terms and engineered safety features

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1984-01-01

    The author states that new, technically defensible, methodologies to establish realistic source term values for nuclear reactor accidents will soon be available. Although these methodologies will undoubtedly find widespread use in the development of accident response procedures, the author states that it is less clear that the industry is preparing to employ the newer results to develop a more rational approach to strategies for the mitigation of fission product releases. Questions concerning the performance of existing engineered safety systems are reviewed

  13. Process for ultimate storage of radioactive fission products

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  14. Biological effects induced by low amounts of nuclear fission products

    International Nuclear Information System (INIS)

    Vasilenko, I.Ya.; Shishkin, V.F.; Khudyakova, N.V.

    1991-01-01

    The review deals with the problem of biological hazard of low radiation doses for animals and human beings taking into the danger of internal and external irradiation by nuclear fission products under conditions of enhancing anthropogenic radiation contamination of biosphere. An attention is paid to the estimation of life span carcinogenesis, genetic and delayed effects. A conclusion is made on a necessity of multiaspect investigation of biological importance of low radiation doses taking into account modifying effects of other environmental factors

  15. Forced decontamination of fission products deposited on urban areas

    International Nuclear Information System (INIS)

    Warming, L.

    1984-12-01

    Long-lived fission products may be deposited in the environment following a serious reactor accident. Areas of special concern are cities where the collective dose might be high because of the population. An extensive literature list is presented here. Only a few of the references deal with the problem as a whole. Some references deal with non-radiaoctive materials but give us useful information about the behaviour of particles on outdoor surfaces. (author)

  16. Thermochemical data for reactor materials and fission products

    International Nuclear Information System (INIS)

    Cordfunke, E.H.P.; Konings, R.J.M.

    1990-01-01

    This volume presents a collection of critically assessed data on inorganic compounds which are of special interest in nuclear reactor safety studies. Thermodynamic equilibrium calculations are an important and widely used instrument in the understanding of the chemical behavior and release of fission products in the course of nuclear reactor accidents. The reliability of such calculations is, nevertheless, limited by the availability of accurate input data for relevant compounds

  17. The Metabolic Properties of the Fission Products and Actinide Elements

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton M.D., J.G.

    1948-03-01

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

  18. Fission product behaviour in the primary circuit of an HTR

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  19. Radiation research contracts: Distribution of fission products in the biosphere

    International Nuclear Information System (INIS)

    Schoenfeld, T.

    1959-01-01

    According to its Statute the IAEA has to fulfil a dual function - to help individual countries in solving their specific problems and to undertake tasks in the common interest of all its Member States. With this latter aim in mind the Agency has placed a number of research contracts with national research institutes. One of them deals with the distribution of fission products in the biosphere. The Agency has contributed to this work by putting at the institutes' disposal scientists from its own staff apparatus and financial aid. Protection against ionizing radiation given off in nuclear transformations is one of the foremost safety problems in all atomic energy operations. While every effort is being made to prevent reactors, processing plants and all other installations from releasing radioactive materials into the biosphere - air, water and earth - under any foreseeable conditions, small amounts of it are actually released into man's living space. Undoubtedly, this will continue to be so, at least for the time being. For example, low activity liquid wastes from some chemical processing plants are decontaminated in special processes, but traces of fission products remain in the liquids finally discharged on the ground or to nearby waterways. In some installations low and medium activity liquid wastes are even released on the ground or into swamps without prior decontamination. It is also to be expected that in accidents larger amounts of fission products may occasionally be released. (author)

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

    Science.gov (United States)

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

    2014-09-01

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

  1. Radiation research contracts: Distribution of fission products in the biosphere

    Energy Technology Data Exchange (ETDEWEB)

    Schoenfeld, T [Vienna University, First Chemical Institute, Vienna (Austria)

    1959-04-15

    Protection against ionizing radiation given off in nuclear transformations is one of the foremost safety problems in all atomic energy operations. While every effort is being made to prevent reactors, processing plants and all other installations from releasing radioactive materials into the biosphere - air, water and earth - under any foreseeable conditions, small amounts of it are actually released into man's living space. Undoubtedly, this will continue to be so, at least for the time being. For example, low activity liquid wastes from some chemical processing plants are decontaminated in special processes, but traces of fission products remain in the liquids finally discharged on the ground or to nearby waterways. In some installations low and medium activity liquid wastes are even released on the ground or into swamps without prior decontamination. It is also to be expected that in accidents larger amounts of fission products may occasionally be released. To make the routine release of small amounts of fission products safe and to be able to estimate the possible effect of larger releases in accidents, a considerable amount of information is required

  2. Radiation research contracts: Distribution of fission products in the biosphere

    International Nuclear Information System (INIS)

    Schoenfeld, T.

    1959-01-01

    Protection against ionizing radiation given off in nuclear transformations is one of the foremost safety problems in all atomic energy operations. While every effort is being made to prevent reactors, processing plants and all other installations from releasing radioactive materials into the biosphere - air, water and earth - under any foreseeable conditions, small amounts of it are actually released into man's living space. Undoubtedly, this will continue to be so, at least for the time being. For example, low activity liquid wastes from some chemical processing plants are decontaminated in special processes, but traces of fission products remain in the liquids finally discharged on the ground or to nearby waterways. In some installations low and medium activity liquid wastes are even released on the ground or into swamps without prior decontamination. It is also to be expected that in accidents larger amounts of fission products may occasionally be released. To make the routine release of small amounts of fission products safe and to be able to estimate the possible effect of larger releases in accidents, a considerable amount of information is required

  3. ACRR fission product release tests: ST-1 and ST-2

    International Nuclear Information System (INIS)

    Allen, M.D.; Stockman, H.W.; Reil, K.O.; Grimley, A.J.; Camp, W.J.

    1988-01-01

    Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These experiments were designed specifically to investigate the effect of increased total pressure on fission product release; ST-1 was performed at approximately 0.16 MPa and ST-2 was run at 1.9 MPa, whereas other parameters were matched as closely as possible. Release rate data were measured for Cs, I, Ba, Sr, Eu, Te, and U. The release rates were higher than predicted by existing codes for Ba, Sr, Eu, and U. Te release was very low, but Te did not appear to be sequestered by the zircaloy cladding; it was evenly distributed in the fuel. In addition, in posttest analysis a unique fuel morphology (fuel swelling) was observed which may have enhanced fission product release, especially in the high pressure test (ST-2). These data are compared with analytical results from the CORSOR correlation and the VICTORIA computer model

  4. Studies on tin based inorganic ion exchangers for fission products separation

    International Nuclear Information System (INIS)

    Dash, A.; Balasubramanian, K.R.; Murthy, T.S.

    1993-01-01

    Tin(IV) antimonate and hydrous tin(IV) oxide have been prepared and their characteristics are evaluated. A new method has been finalized for the separation of 95 Zr- 95 Nb from irradiated uranium using hydrous tin(IV) oxide. In this process, the irradiated sample is dissolved in concentrated HNO 3 , evaporated to near dryness and taken up in 0.5 M HNO 3 . The solution is passed over tin(IV) oxide column and the isotope eluted with 10 M HNO 3 . The product is obtained in pure nitrate form which is generally preferred for different applications. A method has been finalized for the separation of 106 Ru from fission product solution using tin(IV) antimonate. In this method fission product solution is adjusted to 2 M with respect to nitric acid, 137 Cs is separated on a column of ammonium phosphomolybdate, the effluent after adjustment of acidity to 0.2 M is then passed over a column of tin(IV) antimonate where the effluent contains pure 106 Ru. (author). 14 refs., 6 figs., 2 tabs

  5. Study of the behaviour of tetracycline as fission products extracting agent

    International Nuclear Information System (INIS)

    Cunha, I.I.L.

    1983-01-01

    Both spectrophotometric and potentiometric titration techniques were used to show the formation of complexes between tetracycline and the elements: zirconium, uranium, molybdenum, strontium, barium and ruthenium. It has been verified that tetracycline does not form complexes with cesium, tellurium and iodine. Those techniques have also been used to determine the sites on the tetracycline molecule at which ions may be bound. The behaviour of tetracycline as an extracting agent for those elements, as well as for niobium and technetium has been studied and the influence of the acidity of the aqueous phase upon extraction of the elements mentioned has been considered. Extraction experiments were carried out in the presence of chloride, perchlorate, nitrate and sulfate ions. Studies have been made to determine whether or not the complex extracted into organic phase is really the complex formed between tetracycline and the elements considered as well as to determine the time of shaking necessary so that the equilibrium between the phases is attained. Based on all information obtained from extraction experiments made for uranium and the fission products Zr-95, Nb-95, Ce-141, La-140, Ru-103, Ba-140 and Cs-137, the possibility of using tetracycline for separating those fission products from each other and from uranium has been studies and a scheme for simultaneous separation of those elements has been proposed. The same study has been made for I-131, Tc-99m, Mo-99, Te-132, Np-239 and uranium. The method described is applicable to the separation of some fission products existing in solutions at tracer levels, and not to be used in nuclear fuel reprocessing or any other industrial application. (Author) [pt

  6. 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

  7. 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)

  8. Determination of technetium-99 in mixed fission products by neutron activation analysis

    International Nuclear Information System (INIS)

    Bate, L.C.

    1980-01-01

    A method has been developed for analysis of 99 Tc in fission product mixtures. The analysis consists of a chemical separation of 99 Tc, neutron irradiation of the isolated 99 Tc, and gamma-ray spectrometric determination of the induced 100 Tc radioactivity. Technetium-99 is chemically separated from most fission products by a cyclohexanone extraction from basic carbonate solution. Technetium-99 is stripped into water by addition of carbon tetrachloride to the cyclohexanone phase. A final step in the separation procedure is adsorption of 99 Tc on an anion exchange column which provides additional decontamination and places the 99 Tc in a concentrated form for neutron activation analysis. Neutron irradiations of the isolated 99 Tc were made in the pneumatic tube facility at the High Flux Isotope Reactor at a flux of 5 x 10 14 n/cm 2 /sec for 11 seconds. Induced 100 Tc radioactivity was determined immediately after irradiation using gamma-ray spectrometry to measure the 540 and 591 keV lines. Sensitivity of the analysis under these conditions is approximately 5 ng, and samples of up to about 100 ml volume can be easily processed. The method has been successfully applied to reactor fuel solutions and off-gas traps containing 6.5 x 10 -4 to 240 μg 99 Tc/ml

  9. Removal of heavy metals from aqueous solution by Carrot residues

    International Nuclear Information System (INIS)

    Eslamzadeh, T.; Nasernejad, B.; Bonakdar Pour, B.; Zamani, A.; Esmaail-Beygi, M.

    2004-01-01

    The removal of Copper(II), Zinc(II), and Chromium (III) from wastewater by carrot residues was investigated to evaluate cation exchange capacity. The effects of solution P H and co-ions were studied in batch experiments. Adsorption equilibria were initially rapidly established, and then decreased markedly after 10 min. Column experiments were carried out in a glass column filled with carrot residues to evaluate the metal removal capacity. The influences of the feed concentration and feed rate were also studied in order to compare the dynamic capacity for metal binding in different feed concentrations

  10. Production of Plutonium Metal from Aqueous Solutions

    Energy Technology Data Exchange (ETDEWEB)

    Orth, D.A.

    2003-01-16

    The primary separation of plutonium from irradiated uranium by the Purex solvent extraction process at the Savannah River Plant produces a dilute plutonium solution containing residual fission products and uranium. A cation exchange process is used for concentration and further decontamination of the plutonium, as the first step in the final preparation of metal. This paper discusses the production of plutonium metal from the aqueous solutions.

  11. Thermoradiation treatment of sewage sludge using reactor waste fission products

    International Nuclear Information System (INIS)

    Reynolds, M.C.; Hagengruber, R.L.; Zuppero, A.C.

    1974-06-01

    The hazards to public health associated with the application of municipal sewage sludge to land usage are reviewed to establish the need for disinfection of sludge prior to its distribution as a fertilizer, especially in the production of food and fodder. The use of ionizing radiation in conjunction with mild heating is shown to be an effective disinfection treatment and an economical one when reactor waste fission products are utilized. A program for researching and experimental demonstration of the process on sludges is also outlined

  12. Sensitivity and uncertainty analysis for fission product decay heat calculations

    International Nuclear Information System (INIS)

    Rebah, J.; Lee, Y.K.; Nimal, J.C.; Nimal, B.; Luneville, L.; Duchemin, B.

    1994-01-01

    The calculated uncertainty in decay heat due to the uncertainty in basic nuclear data given in the CEA86 Library, is presented. Uncertainties in summation calculation arise from several sources: fission product yields, half-lives and average decay energies. The correlation between basic data is taken into account. The uncertainty analysis were obtained for thermal-neutron-induced fission of U235 and Pu239 in the case of burst fission and irradiation time. The calculated decay heat in this study is compared with experimental results and with new calculation using the JEF2 Library. (from authors) 6 figs., 19 refs

  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. Transient fission product release during reactor shutdown and startup

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  15. 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)

  16. Fission-product burnup chain model for research reactor application

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-12-01

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

  17. Crystallization study of a glass used for fission product storage

    International Nuclear Information System (INIS)

    Morlevat, J.-P.; Uny, Gisele; Jacquet-Francillon, Noel.

    1981-06-01

    The vitreous matrix used in France is a borosilicate glass of low melting point allowing introduction of volatil fission products and of good chemical stability. However, like any glass, if storage temperature is higher than transformation temperature a partial crystallization can occur. Before final storage, it is important to determine of leaching by water eventually occuring on the choosen site is modified by crystalline phases. The aim of this study is the determination of the leaching rate and the identification of crystalline phases formed during thermal treatment and evaluation of its volumic fraction [fr

  18. Equilibrium Temperature Profiles within Fission Product Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Kaminski, Michael D. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-10-01

    We studied waste form strategies for advanced fuel cycle schemes. Several options were considered for three waste streams with the following fission products: cesium and strontium, transition metals, and lanthanides. These three waste streams may be combined or disposed separately. The decay of several isotopes will generate heat that must be accommodated by the waste form, and this heat will affect the waste loadings. To help make an informed decision on the best option, we present computational data on the equilibrium temperature of glass waste forms containing a combination of these three streams.

  19. Estimated effects of interfacial vaporization on fission product scrubbing

    International Nuclear Information System (INIS)

    Moody, F.J.; Nagy, S.G.

    1983-01-01

    When bubbles containing non-condensible gas rise through a water pool, interfacial evaporation causes a flow of vapor into the bubbles. The inflow reduces the outward particle motion toward the bubble wall, diminishing the effectiveness of fission product particle removal. This analysis provides an estimate of evaporation on pool scrubbing effectiveness. It is shown that hot gas, which boils water at the bubble wall, reduces the effective scrubbing height by less than five centimeters. Although the evaporative humidification in a rising bubble containing non-condensible gas has a diminishing effect on scrubbing mechanisms, substantial decontamination is still expected even for the limiting case of a saturated pool

  20. Geochemistry of actinides and fission products in natural aquifer systems

    International Nuclear Information System (INIS)

    Kim, J.I.

    1989-06-01

    The progress in the research area of the community project MIRAGE: 'Geochemistry of actinides and fission products in natural aquatic systems' has been reviewed. This programme belongs to a specific research and technical development programme for the European Atomic Energy Community in the field of management and storage of radioactive waste. The review summarizes research progresses in subject areas: complexation with organics, colloid generation in groundwater and basic retention mechanisms in the framework of the migration of radionuclides in the geosphere. The subject areas are being investigated by 23 laboratories under interlaboratory collaborations or independent studies. (orig.)

  1. Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2005-01-01

    U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k eff ) to determine the net importance of cross sections to k eff . The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: 151 Sm, 103 Rh, 155 Eu, 150 Sm, 152 Sm, 153 Eu, 154 Eu, and 143 Nd

  2. Exact solutions for chemical bond orientations from residual dipolar couplings

    International Nuclear Information System (INIS)

    Wedemeyer, William J.; Rohl, Carol A.; Scheraga, Harold A.

    2002-01-01

    New methods for determining chemical structures from residual dipolar couplings are presented. The fundamental dipolar coupling equation is converted to an elliptical equation in the principal alignment frame. This elliptical equation is then combined with other angular or dipolar coupling constraints to form simple polynomial equations that define discrete solutions for the unit vector(s). The methods are illustrated with residual dipolar coupling data on ubiquitin taken in a single anisotropic medium. The protein backbone is divided into its rigid groups (namely, its peptide planes and C α frames), which may be solved for independently. A simple procedure for recombining these independent solutions results in backbone dihedral angles φ and ψ that resemble those of the known native structure. Subsequent refinement of these φ-ψ angles by the ROSETTA program produces a structure of ubiquitin that agrees with the known native structure to 1.1 A C α rmsd

  3. Modeling of molten core-concrete interactions and fission-product release

    International Nuclear Information System (INIS)

    Norkus, J.K.; Corradini, M.L.

    1991-09-01

    The study of molten core-concrete interaction is important in estimating the possible consequences of a severe nuclear reactor accident. CORCON-Mod2 is a computer program which models the thermal, chemical, and physical phenomena associated with molten core-concrete interactions. Models have been added to extend and improve the modeling of these phenomena. An ideal solution chemical equilibrium methodology is presented to predict the fission-product vaporization release. Additional chemical species have been added, and the calculation of chemical equilibrium has been expanded to the oxidic layer and to the mixed layer configuration. Recent experiments performed at Argonne National Laboratory are compared to CORCON predictions of melt temperature, erosion depth, and release fraction of fission products. The results consistently underpredicted the melt temperatures and erosion rates. However, the predictions of release of Te, Ba, Sr, and U were good. A sensitivity study of the effects of initial temperature, concrete type, use of the mixing option, degree of zirconium oxidation, cavity size, and amount of control material on erosion, gas production, and release of radioactive materials was performed for a PWR and a BWR. The initial melt temperature had the greatest effect on the results of interest. Concrete type and cavity size also had important effects. 78 refs., 35 figs., 40 tabs

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

    International Nuclear Information System (INIS)

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

    1989-05-01

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

  5. Radiochemical determination of Beryllium-7 in a fission-product mixture containing many inorganic salts

    International Nuclear Information System (INIS)

    Prigent, Y.; Van Kote, F.

    1969-01-01

    A radiochemical method is described for analysing beryllium-7 in a mixture of fission products containing many inorganic salts. By studying the influence of various parameters it has been possible to speed up the decontamination on an anionic resin using an HCl isopropanol mixture, as proposed by KORKISCH- and al. Be(OH) 2 is first precipitated in the presence of E.D.T.A.; the main contaminants are then fixed on Dowex 1 x 10 in 12 M HCl and on Dowex 1 x 8 in a 3 M HCl (20 per cent)-isopropanol (80 per cent) (vol/vol) mixture. The Be, which is not fixed, is precipitated by NH 4 H 2 PO 4 in the presence of E.D.T.A., ignited as Be 2 P 2 O 7 , filtered, weighed, and analyzed by gamma spectrometry. The method makes it possible to dose 4 samples in 16 hours with a chemical yield of 80 per cent, using a 4 day-old fission product solution. The overall decontamination factor, exceeds 10 8 . (authors) [fr

  6. Advanced system for separation of rare-earth fission products

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  7. 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

  8. COMEDIE BD1 experiment: Fission product behaviour during depressurization transients

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  9. High flux transmutation of fission products and actinides

    International Nuclear Information System (INIS)

    Gerasimov, A.; Kiselev, G.; Myrtsymova, L.

    2001-01-01

    Long-lived fission products and minor actinides accumulated in spent nuclear fuel of power reactors comprise the major part of high level radwaste. Their incineration is important from the point of view of radwaste management. Transmutation of these nuclides by means of neutron irradiation can be performed either in conventional nuclear reactors, or in specialized transmutation reactors, or in ADS facilities with subcritical reactor and neutron source with application of proton accelerator. Different types of transmutation nuclear facilities can be used in order to insure optimal incineration conditions for radwaste. The choice of facility type for optimal transmutation should be based on the fundamental data in the physics of nuclide transformations. Transmutation of minor actinides leads to the increase of radiotoxicity during irradiation. It takes significant time compared to the lifetime of reactor facility to achieve equilibrium without effective transmutation. High flux nuclear facilities allow to minimize these draw-backs of conventional facilities with both thermal and fast neutron spectrum. They provide fast approach to equilibrium and low level of equilibrium mass and radiotoxicity of transmuted actinides. High flux facilities are advantageous also for transmutation of long-lived fission products as they provide short incineration time

  10. A novel method for fission product noble gas sampling

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  11. Tables of RCN-2 fission-product cross section evaluation

    International Nuclear Information System (INIS)

    Gruppelaar, H.

    1979-05-01

    This report (continuation of ECN-13 and ECN-33) describes the third part of the RCN-2 evaluation of neutron cross sections for fission product nuclides in KEDAK format. It contains evaluated data for nine nuclides, i.e. 142 Nd, 143 Nd, 144 Nd, 145 Nd, 146 Nd, 147 Nd, 148 Nd, 150 Nd and 147 Pm. Most emphasis has been given to the evaluation of the radiative capture cross section, in order to provide a data base for adjustment calculations using results of integral measurements. Short evaluation reports are given for this cross section. The evaluated capture cross sections are compared with recent experimental differential and integral data. Graphs are given of the capture cross sections at neutron energies above 1 keV, in which also adjusted point cross sections, based upon integral STEK and CFRMF data have been plotted. Moreover, the results are compared with those of the well-known ENDF/B-IV evaluation for fission product nucleides. Finally, evaluation summaries are given, which include tables of other important neutron cross sections, such as the total, elastic scattering and inelastic scattering cross sections

  12. Research and Development on Coatings for Retaining Fission Product Iodine

    Energy Technology Data Exchange (ETDEWEB)

    Genco, J. M.; Berry, D. A.; Rosenberg, H. S.; Cremeans, G. E.; Morrison, D. L. [Battelle Memorial Institute, Columbus Laboratories, Columbus, OH (United States)

    1968-12-15

    It is well known that elemental iodine has the propensity for forming charge-transfer complexes with amines. These complexes are stable at ambient temperatures and retain much of this stability at elevated temperatures. Amines also react with methyl iodide and hydrogen iodide to form the quaternary ammonium salts and amine salts, respectively. These chemical properties of amines provide the basis for the development of retentive coatings for fission product iodine. Various amine-containing polymers were studied in steam-air environments at elevated temperatures using dilute quantities of tagged iodine. Both non-condensing and condensing steam conditions were investigated. Several of the polymers showed sorption rates and capacities that would be adequate for the chemical removal of accident-released fission-product iodine and were several times more effective than commercial protective coatings currently being used. The removal capabilities for amine polymers also could be enhanced by impregnating the reactant on a matrix material such as asbestos mat, presumably because the impregnation technique leads to enhanced surface area and porosity. The two most promising coating systems found were 1:10-phenanthroline impregnated upon asbestos and a three component composite film of the co-polymer of Genamid 2000 and Epon 828 as a reactive binder with 1,10-phenanthroline as a reactive filler. The use of a reactive coating as a passive safety system should reduce appreciably the airborne iodine half-life and the hazards associated with iodine release during a nuclear reactor accident. (author)

  13. 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)

  14. Very-long-term storage of fission products

    International Nuclear Information System (INIS)

    Sousselier, Y.; Pradel, J.; Cousin, O.

    The large majority of the fission products, with 99.9 percent of the radioactivity content, do not pose actual problems in storage in a geological formation for which we can guarantee total confinement. Safety of storage in a geological formation for the radionuclides of long half-life is based in particular on the absorption capacity of the geological formations and the example of the Oklo fossil reactor and the retention of Pu which is produced is a striking example. But the problems are not the same, and the properties that we look for in the terrain are not the same. We could thus be led to storage in different geological formations for the fission products and the long-half-life emitters. Their separation is interesting from this point of view, but the date at which the separation is made will not be necessarily that of reprocessing. But there is a question of one or the other, and these storages will offer a very high level of insurance and will present only the potential hazards that are very comparable with those presented by natural conditions

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  16. Fission product data for thermal reactors. Final report. Part 2. Users manual for EPRI-CINDER code and data

    International Nuclear Information System (INIS)

    England, T.R.; Wilson, W.B.; Stamatelatos, M.G.

    1976-12-01

    A four-group fission-product absorption chain library using ENDF/B-IV decay data and cross sections processed with a typical light water reactor spectrum for a modified version of the original CINDER code has been developed as described in Part 1. CINDER is a general point-depletion and fission product code based on an analytical solution of the equations describing nuclides coupled in any linear sequence of radioactive decays and neutron absorptions. The basic code has been in wide use for a number of years. Previously, the user was required to specify all physical data. This report describes the chain library in detail and a modified version of the basic CINDER code (EPRI-CINDER) that is still compatible with existing libraries

  17. 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)

  18. Separation of uranium, plutonium and fission products on zirconium phosphate, Part 1 - Adsorption equilibria and kinetics

    International Nuclear Information System (INIS)

    Gal, I.; Ruvarac, A.

    1963-01-01

    The distribution coefficients of UO 2 ++ , PuO 2 ++ , Pu 3+ , Pu 4+ , Fe 3+ , 137 Cs + , 90 Sr ++ , 95 Zr + + 95 Nb 5+ , 106 Ru and 144 Ce 3+ were determined in the system zirconium phosphate-aqueous solution of HNO 3 . As for the exchange reation Cs + /H + and Sr ++ /2H + , it has been shown that the mass action law can be applied. For these reactions the corresponding equilibrium constants were calculated. The rates of adsorption of Cs + , Sr ++ , Fe 3+ and Pu 4+ from solutions of a fixed HNO 3 concentration were studied, and empirical rate equations were derived. The experimental data confirm that UO 2 ++ can be separated from Pu 4+ . Among the fission products, 90 Sr, 106 Ru and 144 Ce mainly follow the fraction of uranium, while 137 Cs, 95 Zr and 95 Nb follow the plutonium fraction. Separations within the fractions are possible (author)

  19. A fundamental study of fission product deposition on the wall surface

    International Nuclear Information System (INIS)

    Ishiguro, R.; Sakashita, H.; Sugiyama, K.

    1987-01-01

    Deposition of soluble matters on wall surfaces is studied in the present report for the purpose to understand a mechanism of fission product deposition on the wall surface in a molten salt reactor. Calcium carbonate solution is used to observe the fundamental mechanism of deposition. The experiments are performed under conditions of turbulent flow of the solution over a heated wall. According to the experimental results a model is proposed to estimate deposition rate. The model consists of two parts, one is the initial nucleus formation on a clean wall surface and the other is the constant increase of deposition succeeding to the first stage. The model is assessed by comparing it with the experimental results. Both results coincide well in some parameters, but not so well in others. (author)

  20. 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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  2. Fission product yield evaluation for the USA evaluated nuclear data files

    International Nuclear Information System (INIS)

    Rider, B.F.; England, T.R.

    1994-01-01

    An evaluated set of fission product yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set

  3. FISPRO: a simplified computer program for general fission product formation and decay calculations

    International Nuclear Information System (INIS)

    Jiacoletti, R.J.; Bailey, P.G.

    1979-08-01

    This report describes a computer program that solves a general form of the fission product formation and decay equations over given time steps for arbitrary decay chains composed of up to three nuclides. All fission product data and operational history data are input through user-defined input files. The program is very useful in the calculation of fission product activities of specific nuclides for various reactor operational histories and accident consequence calculations

  4. Analytical solutions for peak and residual uplift resistance of pipelines

    Energy Technology Data Exchange (ETDEWEB)

    Nixon, J.F. [Nixon Geotech Ltd., Calgary, AB (Canada); Oswell, J.M. [Naviq Consulting Inc., Calgary, AB (Canada)

    2010-07-01

    Frost heave can occur on cold pipelines that traverse unfrozen, non permafrost terrain. The stresses experienced by the pipeline are partly a function of the strength of the soil on the non heaving side of the frozen-unfrozen interface. This paper proposed three analytical solutions to estimate the soil uplift resistance by considering the pipeline and soil to act similar to a strip footing, a punching shear failure, and by considering the formation of horizontal crack emanating from the spring line of the pipe. Peak uplift resistance and residual uplift resistance were discussed. Results for full scale pipe and for laboratory scale model pipes were presented, with particular reference to cover depth, temperature and crack width; and limits to residual uplift resistance. It was concluded that the peak uplift resistance and the residual uplift resistance are generally independent and controlled by different factors. The peak resistance is related directly to pipe diameter, and less strongly dependent on springline depth. It is also strongly dependent on soil temperature. However, the residual uplift resistance is strongly dependent on burial depth, weakly dependent on pipe displacement rate and also on soil temperature. 15 refs., 19 figs.

  5. 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

  6. Preparation of multigroup lumped fission product cross-sections from ENDF/B-VI for FBRs

    International Nuclear Information System (INIS)

    Devan, K.; Gopalakrishnan, V.; Mohanakrishnan, P.; Sridharan, M.S.

    1997-01-01

    Multigroup pseudo fission product cross-sections were computed from the American evaluated nuclear data library ENDF/B-VI, corresponding to various burnups of the proposed 500 MWe prototype fast breeder reactor (PFBR), in India. The data were derived from the cross-sections of 111 selected fission products that account for almost complete capture of fission products in an FBR. The dependence of burnup on the pseudo fission product cross-sections, and comparison with other data sets, viz. JNDC, ENDF/B-IV and ABBN, are discussed. (author)

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

    International Nuclear Information System (INIS)

    Mo, S.C.

    1995-01-01

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

  8. Implementation of a Thermodynamic Solver within a Computer Program for Calculating Fission-Product Release Fractions

    Science.gov (United States)

    Barber, Duncan Henry

    benchmark calculation demonstrates the improvement in agreement of the total inventory of those chemical elements included in the RMC fuel model to an ORIGEN-S calculation. ORIGEN-S is the Oak Ridge isotope generation and depletion computer program. The Gibbs energy minimizer requires a chemical database containing coefficients from which the Gibbs energy of pure compounds, gas and liquid mixtures, and solid solutions can be calculated. The RMC model of irradiated uranium dioxide fuel has been converted into the required format. The Gibbs energy minimizer has been incorporated into a new model of fission-product vaporization from the fuel surface. Calculated release fractions using the new code have been compared to results calculated with SOURCE IST 2.0P11 and to results of tests used in the validation of SOURCE 2.0. The new code shows improvements in agreement with experimental releases for a number of nuclides. Of particular significance is the better agreement between experimental and calculated release fractions for 140La. The improved agreement reflects the inclusion in the RMC model of the solubility of lanthanum (III) oxide (La2O3) in the fuel matrix. Calculated lanthanide release fractions from earlier computer programs were a challenge to environmental qualification analysis of equipment for some accident scenarios. The new prototype computer program would alleviate this concern. Keywords: Nuclear Engineering; Material Science; Thermodynamics; Radioactive Material, Gibbs Energy Minimization, Actinide Generation and Depletion, FissionProduct Generation and Depletion.

  9. Transient fission-product release during reactor shutdown and startup

    International Nuclear Information System (INIS)

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

    1997-12-01

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

  10. Early history of soil contamination with fission products

    International Nuclear Information System (INIS)

    Nagel, E.

    1985-01-01

    Underneath a balloon and instrument hut demolished today and on the natural pasture next to it, a series of 2 to 4-layer soil samples was carried out to determine the contents in Cs-137 and Sr-90. The results showed a sudded decrease of the Cs concentration behind the walls of the hut, but no further decrease towards the centre of the hut. As expected, the decrease of Sr-90 concentration was slower, both in a horizontal line and in the depth. The results reveal that the soil underneath the hut has not received further fission products since it was built in 1956 from depositions of later nuclear weapon tests. Furthermore, the radionuclides were not transported much further into lower strata so that by taking into consideration of the half-life of the two nuclides their concentration in the soil can be traced back to the year 1956 for comparison with the results measured at the time. (orig./HP) [de

  11. ENDF/B-5 fission product cross section evaluations

    International Nuclear Information System (INIS)

    Schenter, R.E.; England, T.R.

    1979-12-01

    Cross section evaluations were made for the 196 fission product nuclides on the ENDF/B-5 data files. Most of the evaluations involve updating the capture cross sections of the important absorbers for fast and thermal reactor systems. This included updating thermal values, resonance integrals, resonance parameter sets, and fast capture cross sections. For the fast capture results generalized least-squares calculations were made with the computer code FERRET. Input for these cross section adjustments included nuclear models calculations and both integral and differential experimental data results. The differential cross sections and their uncertainties were obtained from the CSIRS library. Integral measurement results came from CFRMF and STEK Assemblies 500, 1000, 2000, 3000, 4000. Comparisons of these evaluations with recent capture measurements are shown. 15 figures, 10 tables

  12. Modeling of fission product release in integral codes

    International Nuclear Information System (INIS)

    Obaidurrahman, K.; Raman, Rupak K.; Gaikwad, Avinash J.

    2014-01-01

    The Great Tohoku earthquake and tsunami that stroke the Fukushima-Daiichi nuclear power station in March 11, 2011 has intensified the needs of detailed nuclear safety research and with this objective all streams associated with severe accident phenomenology are being revisited thoroughly. The present paper would cover an overview of state of art FP release models being used, the important phenomenon considered in semi-mechanistic models and knowledge gaps in present FP release modeling. Capability of FP release module, ELSA of ASTEC integral code in appropriate prediction of FP release under several diversified core degraded conditions will also be demonstrated. Use of semi-mechanistic fission product release models at AERB in source-term estimation shall be briefed. (author)

  13. Evaluation of fission product neutron cross sections for JENDL

    International Nuclear Information System (INIS)

    1984-01-01

    The recent activities on the evaluation of fission product (FP) neutron cross sections for JENDL (Japanese Evaluated Nuclear Data Library) are presented briefly. The integral test of JENDL-1 FP cross section file was performed using the CFRMF sample activation data and the STEK sample reactivity data, and the ratio of experiment to calculation was nearly constant for all the samples in the STEK measurement. Therefore, a tentative analysis was performed by applying the correction to the calculated scattering reactivity component. Better agreement with the experiment was obtained after applying this correction in most cases. The evaluation work on the JENDL-2 FP neutron cross sections is now in progress. The improvement of the data evaluation is presented in an itemized form. The JENDL-2 FP file will contain the evaluated data for 100 nuclides from Kr to Tb. The improvement and the future scope of the integral test for JENDL-2 FP data are summarized. (Asami, T.)

  14. 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)

  15. A revised ANS standard for decay heat from fission products

    International Nuclear Information System (INIS)

    Schrock, V.E.

    1978-01-01

    The draft ANS 5.1 standard on decay heat was published in 1971 and given minor revision in 1973. Its basis was the best estimate working curve developed by K. Shure in 1961. Liberal uncertainties were assigned to the standard values because of lack of data for short cooling times and large discrepancies among experimental data. Research carried out over the past few years has greatly improved the knowledge of this phenomenon and a major revision of the standard has been completed. Very accurate determination of the decay heat is now possible, expecially within the first 10 4 seconds, where the influence of neutron capture in fission products may be treated as a small correction to the idealized zero capture case. The new standard accounts for differences among fuel nuclides. It covers cooling time to 10 9 seconds, but provides only an ''upper bound'' on the capture correction in the interval 10 4 9 seconds. (author)

  16. Monitoring of fission products through on-line gamma spectrometry

    International Nuclear Information System (INIS)

    Montagnon, F.; Warlop, R.

    1989-01-01

    Under normal operating conditions, the monitoring of the possible deterioration of the pressurized water reactor core fuel rods is achieved through analysis of the radioactive fission products carried by the primary system. For acquiring results of spectrometric analyses in real time, and avoiding risks of errors linked to manual operations, CEA/DMG and EDF/SEPTEN have jointly developed an entirely automatic system. This system allows measuring permanently the primary system activity of two coupled units, with no human operation nor any handling of active coolant specimens. The PIGAL facility has been set up in the nuclear auxiliary building, common to the two units, and it is used on a demonstration basis for units 2 and 3 of the BUGEY site. This device has been patented

  17. Method of processing liquid waste containing fission product

    International Nuclear Information System (INIS)

    Funabashi, Kiyomi; Kawamura, Fumio; Matsuda, Masami; Komori, Itaru; Miura, Eiichi.

    1988-01-01

    Purpose: To prepare solidification products of low surface dose by removing cesium which is main radioactive nuclides from re-processing plants. Method: Liquid wastes containing a great amount of fission products are generated accompanying the reprocessing for spent nuclear fuels. After pH adjustment, the liquid wastes are sent to a concentrator to concentrate the dissolved ingredients. The concentrated liquid wastes are pumped to an adsorption tower in which radioactive cesium contributing much to the surface dose is removed. Then, the liquid wastes are sent by way of a surge tank to a mixing tank, in which they are mixed under stirring with solidifying agents such as cements. Then, the mixture is filled in a drum-can and solidified. According to this invention, since radioactive cesium is removed before solidification, it is possible to prepare solidification products at low surface dose and facilitate the handling of the solidification products. (Horiuchi, T.)

  18. Fission product poisoning in KS-150 reactor operation

    International Nuclear Information System (INIS)

    Rana, S.B.

    1978-01-01

    A three-dimensional model of the KS-150 reactor was used to study reactivity changes induced by reactor poisoning with fission products Xe 135 and Sm 149 . A comparison of transients caused by the poisoning showed the following differences: (1) the duration of the transient Xe poisoning (2 days) is shorter by one order of magnitude than the duration of Sm poisoning (20 days); however, the level of Xe poisoning is greater approximately by one order than the level of the Sm poisoning; (2) the level of steady-state Xe poisoning depends on the output level of the reactor; steady-state Sm poisoning does not depend on this level; (3) following reactor shutdown Xe poisoning may increase to the maximum value of up to Δrhosub(Xe)=20% and will then gradually decrease; Sm poisoning may reach maximum values of up to Δrhosub(Sm)=2% and does not decrease. (J.B.)

  19. The Phebus Fission Product and Source Term International Programmes

    International Nuclear Information System (INIS)

    Clement, B.; Zeyen, R.

    2005-01-01

    The international Phebus FP programme, initiated in 1988 is one of the major research programmes on light water reactors severe accidents. After a short description of the facility and of the test matrix, the main outcomes and results of the first four integral tests are provided and analysed. Several results were unexpected and some are of importance for safety analyses, particularly concerning fuel degradation, cladding oxidation, chemical form of some fission products, especially iodine, effect of control rod materials on degradation and chemistry, iodine behaviour in the containment. Prediction capabilities of calculation tools have largely been improved as a result of this research effort. However, significant uncertainties remain for a number of phenomena, requiring detailed physical analysis and implementation of improved models in codes, sustained by a number of separate-effect experiments. This is the subject of the new Source Term programme for a better understanding of the phenomenology on important safety issues, in accordance with priorities defined in the EURSAFE project of the 5 th European framework programme aiming at reducing the uncertainties on Source Term analyses. It covers iodine chemistry, impact of boron carbide control rods degradation and oxidation, air ingress situations and fission product release from fuel. Regarding the interpretation of Phebus, an international co-operation has been established since over ten years, particularly helpful for the improvement and common understanding of severe accident phenomena. Few months ago, the Phebus community was happy to welcome representatives of a large number of organisations from the following new European countries: the Czech republic, Hungary, Lithuania, Slovakia, Slovenia and also from Bulgaria and Romania. (author)

  20. Evaluation and compilation of fission product yields 1993

    International Nuclear Information System (INIS)

    England, T.R.; Rider, B.F.

    1995-01-01

    This document is the latest in a series of compilations of fission yield data. Fission yield measurements reported in the open literature and calculated charge distributions have been used to produce a recommended set of yields for the fission products. The original data with reference sources, and the recommended yields axe presented in tabular form. These include many nuclides which fission by neutrons at several energies. These energies include thermal energies (T), fission spectrum energies (F), 14 meV High Energy (H or HE), and spontaneous fission (S), in six sets of ten each. Set A includes U235T, U235F, U235HE, U238F, U238HE, Pu239T, Pu239F, Pu241T, U233T, Th232F. Set B includes U233F, U233HE, U236F, Pu239H, Pu240F, Pu241F, Pu242F, Th232H, Np237F, Cf252S. Set C includes U234F, U237F, Pu240H, U234HE, U236HE, Pu238F, Am241F, Am243F, Np238F, Cm242F. Set D includes Th227T, Th229T, Pa231F, Am241T, Am241H, Am242MT, Cm245T, Cf249T, Cf251T, Es254T. Set E includes Cf250S, Cm244S, Cm248S, Es253S, Fm254S, Fm255T, Fm256S, Np237H, U232T, U238S. Set F includes Cm243T, Cm246S, Cm243F, Cm244F, Cm246F, Cm248F, Pu242H, Np237T, Pu240T, and Pu242T to complete fission product yield evaluations for 60 fissioning systems in all. This report also serves as the primary documentation for the second evaluation of yields in ENDF/B-VI released in 1993

  1. Nuclearization of ionic chromatography system for fission products analysis

    International Nuclear Information System (INIS)

    Dimeglio, Remi

    1996-06-01

    The accident at Tchernobyl in 1986 had entailed the release in the atmosphere of different products coming from the splitting of the fuel. It is to better understand, and also to warn this type of catastrophe that the CEA (Commissariat a L'Energie Atomique) develops many programs of researches, aiming to characterize these fission products and to study their mechanisms of relaxation. Thus, the LESC (Laboratoire d'Etude de la Surete du Combustible) takes part, since several years, in many nuclear safety experiences, and in particular to the project PHEBUS PF, that is a reconstitution, in reduced scale, of an accident entailing the fusion of the reactor core. The aim of the researches that have been led during this training period was to the nuclearization of an HPIC (High Performance Ion Chromatography) system, dedicated to the analysis of the PHEBUS PF fission products analysis. The first step was to develop HPIC lines already settled, so as to reduce the quantity of wastes. Indeed, those one are very difficult to process in a radioactive area. For this purpose, we have implanted a column cationic more effective, so as to decrease analysis times, and, by there even, the quantity of sewage generated. We have equally replaced, on lines cationic and anionic, the system of suppression of the eluent conductivity, to make it thriftier in fluid. But the radioactive products characterization necessitates that all analyses are led within a special box with gloves. The second step of the project was therefore to adapt the system to this type of cell, and to its automation. It has been necessary to modify the system of sample injection, the system of detection, and to put in place a supplementary box with gloves, connected by sieve to the first, for the active products dilution. (author) [fr

  2. Analysis of Dust and Fission Products in PBMR Turbine

    International Nuclear Information System (INIS)

    Stempniewicz, M.M.; Wessels, D.

    2014-01-01

    A 400 MWth direct cycle Pebble Bed Modular reactor was under development in South Africa. The work performed included design and safety analyses. In HTR/PBMR, graphite dust is generated during normal reactor operation due to pebble-to-pebble scratching. This dust will be deposited throughout the primary system. Furthermore, the dust will become radioactive due to sorption of fission products released, although in very small quantities, during normal operation. This paper presents a model and analyses of the PBMR turbine with the SPECTRA code. The purpose of the present work was to estimate the amount and distribution of deposited dust and the fission products, namely cesium, iodine, and silver, during plant life-time, which was assumed to be 40 full-power years. The performed work showed that after 40 years of plant life-time deposited layers are very small. The largest deposition is of course observed on the dust filters. Apart from the dust filters, the largest dust deposition is observed on the: • Outer Casing (inner walls) • Turbine Rotor Cooling Cavity (inner walls) • HPC Cold Cooling Gas Header (inner walls) This is caused by relatively low gas velocities in these volumes. The low velocities allow a continuous build-up of the dust layer. About 90% of cesium, 40% of iodine, and 99.9% of silver is adsorbed on the metallic structures of the turbine. The sorption rate increases along the turbine due to decreasing temperatures. In case of cesium and iodine the highest concentrations are observed in the last stage (stage 12) of the turbine. In the case of silver the sorption is so large that the silver vapor is significantly depleted in the last stages of the turbine. This is a reason for having a maximum in silver concentration in the stage 10. In the following stages the concentration decreases due to very small silver vapor fraction in the gas. (author)

  3. Evaluation and compilation of fission product yields 1993

    Energy Technology Data Exchange (ETDEWEB)

    England, T.R.; Rider, B.F.

    1995-12-31

    This document is the latest in a series of compilations of fission yield data. Fission yield measurements reported in the open literature and calculated charge distributions have been used to produce a recommended set of yields for the fission products. The original data with reference sources, and the recommended yields axe presented in tabular form. These include many nuclides which fission by neutrons at several energies. These energies include thermal energies (T), fission spectrum energies (F), 14 meV High Energy (H or HE), and spontaneous fission (S), in six sets of ten each. Set A includes U235T, U235F, U235HE, U238F, U238HE, Pu239T, Pu239F, Pu241T, U233T, Th232F. Set B includes U233F, U233HE, U236F, Pu239H, Pu240F, Pu241F, Pu242F, Th232H, Np237F, Cf252S. Set C includes U234F, U237F, Pu240H, U234HE, U236HE, Pu238F, Am241F, Am243F, Np238F, Cm242F. Set D includes Th227T, Th229T, Pa231F, Am241T, Am241H, Am242MT, Cm245T, Cf249T, Cf251T, Es254T. Set E includes Cf250S, Cm244S, Cm248S, Es253S, Fm254S, Fm255T, Fm256S, Np237H, U232T, U238S. Set F includes Cm243T, Cm246S, Cm243F, Cm244F, Cm246F, Cm248F, Pu242H, Np237T, Pu240T, and Pu242T to complete fission product yield evaluations for 60 fissioning systems in all. This report also serves as the primary documentation for the second evaluation of yields in ENDF/B-VI released in 1993.

  4. First-principles study of defects and fission product behavior in uranium diboride

    Science.gov (United States)

    Jossou, Ericmoore; Oladimeji, Dotun; Malakkal, Linu; Middleburgh, Simon; Szpunar, Barbara; Szpunar, Jerzy

    2017-10-01

    A Systematic study of defects and incorporation of xenon (Xe) and zirconium (Zr) fission products in uranium diboride (UB2) has been investigated using density functional theory (DFT) calculations as implemented in Quantum ESPRESSO code. The incorporation and solution energies show that both FPs (Xe and Zr) are most stable in U vacancies with Zr being more stable than Xe. A volume expansion is observed as the concentration of Xe increases in the fuel matrix while Zr incorporation leads to a contraction. Bader charge analysis is used to establish the formation of Zr-B ionic/covalent bond due to large electron transfer observed while there is only a weak electronic interaction between Xe and the UB2 lattice. Finally, using climbing-image nudged elastic band calculation, we found that the energy barrier of U in UB2 is 0.08 eV higher than B migration energy.

  5. Continuous radiochemical analysis of fission products in a nuclear reactor water coolant

    International Nuclear Information System (INIS)

    Moskvin, L.N.; Zakharov, L.K.; Leont'ev, G.G.; Mel'nikov, V.A.; Orlenkov, I.S.; Slutskij, G.K.

    1975-01-01

    Method for continuous radiochemical analysis of I, Cs, Ba, Sr and Ce isotopes in a reactor water heat-transfer agent was developed. A continuous two-dimensional chromatographic process of complex mixtures separation of substances proved to be feasible on several parallel sorbent layers, which moved at constant velocities and separated by stationary intermediate collectors. Tests on model solutions containing I, Ce, Cs and Ba isotopes and on heat-carrier samples showed quantitative separation of elements. The results were indicative of a basic possibility of using multisorbent chromatographs for continuous control of multicomponent mixtures, particularly for control of radioactive fission product compositions in water heat-transfer agents in nuclear power plants. A diagram is shown for a two-dimensional chromatographic separation of a multicomponent mixture. Also shown is a flow chart of an installation for continuous control of iodine and cesium isotope activities

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

    International Nuclear Information System (INIS)

    Fang Chao

    2013-01-01

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

  7. NEANDC specialists meeting on yields and decay data of fission product nuclides

    International Nuclear Information System (INIS)

    Chrien, R.E.; Burrows, T.W.

    1983-01-01

    Separate abstracts were prepared for the 29 papers presented. Workshop reports on decay heat, fission yields, beta- and gamma-ray spectroscopy, and delayed neutrons are included. An appendix contains a survey of the most recent compilations and evaluations containing fission product yield, fission product decay data, and delayed neutron yield information

  8. Energy distribution of antineutrinos originating from the decay of fission products in a nuclear reactor

    International Nuclear Information System (INIS)

    Rudstam, G.; Aleklett, K.

    1979-01-01

    The energy spectrum of antineutrinos around a nuclear reactor has been derived by summing contributions from individual fission products. The resulting spectrum is weaker at energies above approx. 8 MeV than earlier published antineutrino spectra. The reason may be connected to the strong feeding of high-lying daughter states in the beta decay of fission products with high disintegration energies

  9. Role of fission product in whole core accidents: research in the USA

    International Nuclear Information System (INIS)

    Jackson, J.F.; Deitrich, L.W.

    1977-01-01

    The techniques being developed in the United States for analyzing postulated whole-core accidents in LMFBRs are briefly reviewed. The key mechanistic analysis methods are discussed in detail. Important research projects in the area of fission product effects are examined. Some typical results on the role of fission products in whole-core accidents are presented

  10. NEANDC specialists meeting on yields and decay data of fission product nuclides

    Energy Technology Data Exchange (ETDEWEB)

    Chrien, R.E.; Burrows, T.W. (eds.)

    1983-01-01

    Separate abstracts were prepared for the 29 papers presented. Workshop reports on decay heat, fission yields, beta- and gamma-ray spectroscopy, and delayed neutrons are included. An appendix contains a survey of the most recent compilations and evaluations containing fission product yield, fission product decay data, and delayed neutron yield information. (WHK)

  11. Characteristic relation for the mass and energy distribution of the nuclear fission products

    International Nuclear Information System (INIS)

    Alexandru, G.

    1977-01-01

    The dispersion relation for nuclear fission is written in the two part fragmentation approach which allows to obtain the characteristic relation for the mass and energy distribution of the nuclear fission products. One explains the resonance approximation in the mass distribution of the fission products taking into account the high order resonances too. (author)

  12. The universal library of fission products and delayed neutron group yields

    International Nuclear Information System (INIS)

    Koldobskiy, A.B.; Zhivun, V.M.

    1997-01-01

    A new fission product yield library based on the Semiempirical method for the estimation of their mass and charge distribution is described. Contrary to other compilations, this library can be used with all possible excitation energies of fissionable actinides. The library of delayed neutron group yields, based on the fission product yield compilation, is described as well. (author). 15 refs, 4 tabs

  13. Utilization of fast reactor excess neutrons for burning long-lived fission products

    International Nuclear Information System (INIS)

    Kawashima, K.; Kobayashi, K.; Kaneto, K.

    1995-01-01

    An evaluation is made on a large MOX fuel fast reactor's capability of burning long lived fission product Tc-99, which dominates the long term radiotoxicity of the high level radioactive waste. The excess neutrons generated in the fast reactor core are utilized to transmute Tc-99 to stable isotopes due to neutron capture reaction. The fission product target assemblies which consist of Tc-99 are charged to the reactor core periphery. The fission product target neutrons are moderated to a great deal to pursue the possibility of enhancing the transmutation rate. Any impacts of loading the fission product target assemblies on the core nuclear performances are assessed. A long term Tc-99 accumulation scenario is considered in the mix of fission product burner fast reactor and non-burner LWRs. (author)

  14. 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

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

    International Nuclear Information System (INIS)

    Klonk, H.

    1976-01-01

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

  16. Plutonium and surrogate fission products in a composite ceramic waste form

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  17. Photofission observations in reactor environments using selected fission-product yields

    International Nuclear Information System (INIS)

    Gold, R.; Ruddy, F.H.; Roberts, J.H.

    1982-01-01

    A new method for the observation of photofission in reactor environments is advanced. It is based on the in-situ observation of fission product yield. In fact, at a given in-situ reactor location, the fission product yield is simply a weighted linear combination of the photofission product yield, Y/sub gamma/, and the neutron induced fission product yield, Y/sub n. The weight factors arising in this linear combination are the photofission fraction and neutron induced fission fraction, respectively. This method can be readily implemented with established techniques for measuring in-situ reactor fission product yield. For example, one can use the method based on simultaneous irradiation of radiometric (RM) and solid state track recorder (SSTR) fission monitors. The sensitivity and accuracy and current knowledge of fission product yields. Unique advantages of this method for reactor applications are emphasized

  18. Evaluation of Neutron Induced Reactions for 32 Fission Products

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeong Il

    2007-02-15

    Neutron cross sections for 32 fission products were evaluated in the neutron-incident energy range from 10{sup -5} eV to 20 MeV. The list of fission products consists of the priority materials for several applications, extended to cover complete isotopic chains for three elements. The full list includes 8 individual isotopes, {sup 95}Mo, {sup 101}Ru, {sup 103}Rh, {sup 105}Pd, {sup 109}Ag, {sup 131}Xe, {sup 133}Cs, {sup 141}Pr, and 24 isotopes in complete isotopic chains for Nd (8), Sm (9) and Dy (7). Our evaluation methodology covers both the low energy region and the fast neutron region.In the low energy region, our evaluations are based on the latest data published in the Atlas of Neutron Resonances. This resource was used to infer both the thermal values and the resolved resonance parameters that were validated against the capture resonance integrals. In the unresolved resonance region we performed the additional evaluation by using the averages of the resolved resonances and adjusting them to the experimental data.In the fast neutron region our evaluations are based on the nuclear reaction model code EMPIRE-2.19 validated against the experimental data. EMPIRE is the modular system of codes consisting of many nuclear reaction models, including the spherical and deformed Optical Model, Hauser-Feshbach theory with the width fluctuation correction and complete gamma-ray emission cascade, DWBA, Multi-step Direct and Multi-step Compound models, and several versions of the phenomenological preequilibrium models. The code is equipped with a power full GUI, allowing an easy access to support libraries such as RIPL and CSISRS, the graphical package, as well the utility codes for formatting and checking. In general, in our calculations we used the Reference Input Parameter Library, RIPL, for the initial set model parameters. These parameters were properly adjusted to reproduce the available experimental data taken from the CSISRS library. Our evaluations cover cross

  19. Evaluation of Neutron Induced Reactions for 32 Fission Products

    International Nuclear Information System (INIS)

    Kim, Hyeong Il

    2007-02-01

    Neutron cross sections for 32 fission products were evaluated in the neutron-incident energy range from 10 -5 eV to 20 MeV. The list of fission products consists of the priority materials for several applications, extended to cover complete isotopic chains for three elements. The full list includes 8 individual isotopes, 95 Mo, 101 Ru, 103 Rh, 105 Pd, 109 Ag, 131 Xe, 133 Cs, 141 Pr, and 24 isotopes in complete isotopic chains for Nd (8), Sm (9) and Dy (7). Our evaluation methodology covers both the low energy region and the fast neutron region.In the low energy region, our evaluations are based on the latest data published in the Atlas of Neutron Resonances. This resource was used to infer both the thermal values and the resolved resonance parameters that were validated against the capture resonance integrals. In the unresolved resonance region we performed the additional evaluation by using the averages of the resolved resonances and adjusting them to the experimental data.In the fast neutron region our evaluations are based on the nuclear reaction model code EMPIRE-2.19 validated against the experimental data. EMPIRE is the modular system of codes consisting of many nuclear reaction models, including the spherical and deformed Optical Model, Hauser-Feshbach theory with the width fluctuation correction and complete gamma-ray emission cascade, DWBA, Multi-step Direct and Multi-step Compound models, and several versions of the phenomenological preequilibrium models. The code is equipped with a power full GUI, allowing an easy access to support libraries such as RIPL and CSISRS, the graphical package, as well the utility codes for formatting and checking. In general, in our calculations we used the Reference Input Parameter Library, RIPL, for the initial set model parameters. These parameters were properly adjusted to reproduce the available experimental data taken from the CSISRS library. Our evaluations cover cross sections for almost all reaction channels

  20. Transport and release of fission products during nuclear reactor accident

    International Nuclear Information System (INIS)

    Lee, K.W.; Kuhlman, M.R.; Gieseke, J.A.

    1984-01-01

    This study represents the identification and formulation of a systematic, mechanistic approach to estimating source terms and the implementation of this approach through calculations of fission products release to the environment for a large PWR reactor under a selected set of accident conditions. The development and improvement of calculational procedures is an evolutionary process and in the long term must be verified through experimental studies. It is anticipated that as additional information is obtained the accuracy of predictions can be improved and uncertainties reduced. Transport and deposition of radionuclides were found to be quite dependent on the accident sequences and the corresponding thremal hydraulic conditions. Reduced temperatures led to increased deposition of vapor species, and reduced flow rates to increased aerosol deposition. It is to be recognized that the estimates of release fractions are subject to uncertainties in the data and computer models employed in the calculations and are expected to have been influenced by assumptions regarding plant geometry, thermal hydraulics, deposition mechanisms, and sequence events. The effects of these assumptions will be investigated as this study continues. (Author)

  1. Geochemistry of long lived transuranic actinides and fission products

    International Nuclear Information System (INIS)

    1992-01-01

    The IAEA initiated in 1987 a new Co-ordinated Research Programme (CRP) on geochemistry of long lived transuranic actinides and fission products for a duration of 5 years. The framework of the CRP consists of three main components: (1) development of a working hypothesis with focus on laboratory studies; (2) testing of the working hypothesis with the focus on the field studies; and (3) transport modelling. The contents of this document reflect the results reported on by a number of Member States who participated in this Co-ordinated Research Programme which investigated the geochemical processes and mechanisms which affect rock-water interactions and migration of the chemical elements in geological media as scientific background in support of safety assessments of repositories for high level radioactive wastes. Studies conducted considered the migration of the long lived radionuclides of Tc, I, Np and Pu in both the near and far field. The programme investigated natural occurrences and geochemical processes and mechanisms which may affect migration of the chemical elements under consideration in geological media which may be used for disposal of radioactive wastes. 47 refs, 9 figs, 1 tab

  2. Fission product yield measurements using monoenergetic photon beams

    Science.gov (United States)

    Krishichayan; Bhike, M.; Tonchev, A. P.; Tornow, W.

    2017-09-01

    Measurements of fission products yields (FPYs) are an important source of information on the fission process. During the past couple of years, a TUNL-LANL-LLNL collaboration has provided data on the FPYs from quasi monoenergetic neutron-induced fission on 235U, 238U, and 239Pu and has revealed an unexpected energy dependence of both asymmetric fission fragments at energies below 4 MeV. This peculiar FPY energy dependence was more pronounced in neutron-induced fission of 239Pu. In an effort to understand and compare the effect of the incoming probe on the FPY distribution, we have carried out monoenergetic photon-induced fission experiments on the same 235U, 238U, and 239Pu targets. Monoenergetic photon beams of Eγ = 13.0 MeV were provided by the HIγS facility, the world's most intense γ-ray source. In order to determine the total number of fission events, a dual-fission chamber was used during the irradiation. These irradiated samples were counted at the TUNL's low-background γ-ray counting facility using high efficient HPGe detectors over a period of 10 weeks. Here we report on our first ever photofission product yield measurements obtained with monoenegetic photon beams. These results are compared with neutron-induced FPY data.

  3. Cerenkov Detectors for Fission Product Monitoring in Reactor Coolant Water

    Energy Technology Data Exchange (ETDEWEB)

    Strindehag, O

    1967-09-15

    The expected properties of Cerenkov detectors when used for fission product monitoring in water cooled reactors and test loops are discussed from the point of view of the knowledge of the sensitivity of these detectors to some beta emitting isotopes. The basic theory for calculation of the detector response is presented, taking the optical transmission in the sample container and the properties of the photomultiplier tube into account. Special attention is paid to the energy resolution of this type of Cerenkov detector. For the design of practical detectors the results from several investigations of various window and reflector materials are given, and the selection of photomultiplier tubes is briefly discussed. In the case of optical reflectors and photomultiplier tubes reference is made to two previous reports by the author. The influence of the size and geometry of the sample container on the energy resolution follows from a separate investigation, as well as the relative merits of sample containers with transparent inner walls. Provided that the energy resolution of the Cerenkov detector is sufficiently high, there are several reasons for using this detector type for failed-fuel-element detection. It seems possible to attain the desired energy resolution by careful detector design.

  4. 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

  5. Fission product separation from seawater by electrocoagulation method

    International Nuclear Information System (INIS)

    Kitagaki, T.; Hoshino, T.; Sambommatsu, Y.; Yano, K.; Takeuchi, M.; Igarashi, T.; Suzuki, T.

    2013-01-01

    At the Fukushima Daiichi nuclear power station, seawater was urgently injected into the reactor core. Therefore a large amount of seawater containing highly radioactive fission products (FP) accumulated and its treatment has been a serious problem. FP such as Cs, Sr and I in water are generally removed by an ion exchanger such as zeolite and separated with column or chemical precipitation methods. An alternative electrocoagulation method, which efficiently separates fine particles from the liquid phase without a chemical reagent is expected to be part of a useful separation system that can reduce the amount of waste, decrease processing time and simplify the process. In this study, powdered adsorbents, such as ferrocyanide and zeolite, were added to seawater containing simulated FP, and the electrocoagulation effect with Al alloy electrodes were investigated. More than 99 % of Cs and 90 % of I were removed by potassium nickel hexacyanoferrate(II) and silver zeolite, respectively. Sedimentation was promoted by electrocoagulation and addition of an inorganic cohesion promoter further increased the sedimentation rate. Moreover, rapid dissolution reaction with heating of the aggregation substance was not observed, so the thermal risk of aqueous processing of it would be low. In addition, thermal analyses showed that the electrocoagulation process did not lead to thermal decomposition. Therefore, if the electrocoagulation method is applied to a decontamination system, it has the potential to thermally stabilize and reduce waste. (author)

  6. Preliminary treatment of chlorinated waste streams containing fission products

    Energy Technology Data Exchange (ETDEWEB)

    Hudry, Damien; Bardez, Isabelle; Bart, Florence [CEA Marcoule DTCD/SECM/LM2C, BP 17171, 30207 Bagnols sur Ceze (France); Deniard, Philippe; Jobic, Stephane [Institut des Materiaux Jean Rouxel, Universite de Nantes, CNRS, BP 32229, 44322 Nantes cedex 3 (France); Rakhmatullin, Aydar [Conditions Extremes et Materiaux: Hautes Temperatures et Irradiations, CEMHTI-CNRS, 45071 Orleans cedex 2 (France); Bessada, Catherine [Conditions Extremes et Materiaux: Hautes Temperatures et Irradiations, CEMHTI-CNRS, 45071 Orleans cedex 2 (France); Universite d' Orleans, Faculte des Sciences, BP 6749, 45067 Orleans cedex 2 (France)

    2008-07-01

    Separating actinides from fission products (FP) by electrolytic techniques in a molten chloride medium produces high-level waste which, because of its high chlorine content, cannot be directly and quantitatively loaded in a glass matrix and therefore requires the development of new management methods. In this regard the strategy of submitting chlorinated waste streams to a preliminary treatment consists in separating the various types of FP from the solvent to minimize the ultimate high-level waste volume. Selective precipitation of the rare earth elements by NH{sub 4}H{sub 2}PO{sub 4} was investigated in a LiCl-KCl medium, and could constitute the first step in the purification process. Unlike the use of alkali orthophosphate, this method provides similar conversion factors with the simple addition of stoichiometric phosphorus (P:rare-earth = 1) and does not require excess phosphate (P:rare-earth = 5). This prevents the formation of a secondary Li{sub 3}PO{sub 4} phase. Moreover, NH{sub 4}H{sub 2}PO{sub 4} also allows chlorine bound to rare earth elements to be eliminated as NH{sub 4}Cl. The formation of HCl is highly probable.

  7. Fission product release from core-concrete mixtures

    International Nuclear Information System (INIS)

    Roche, M.F.; Settle, J.; Leibowitz, L.; Johnson, C.E.; Ritzman, R.L.

    1988-01-01

    The objective of this research is to measure the amount of strontium, barium, and lanthanum that is vaporized from core-concrete mixtures. The measurements are being done using a transpiration method. Mixtures of limestone-aggregated concrete, urania doped with a small amount of La, Sr, Ba, and Zr oxides, and stainless steel were vaporized at 2150 K from a zirconia crucible into flowing He-6% H 2 -0.06% H 2 O (a partial molar free energy of oxygen of -420 kJ). The amounts that were vaporized was determined by weight change and by chemical analyses on condensates. The major phases present in the mixture were inferred from electron probe microanalysis (EPM). They were: (1) urania containing calcia and zirconia, (2) calcium zirconate, (3) a calcium magnesium silicate, and (4) magnesia. About 10% of the zirconia crucible was dissolved by the concrete-urania mixture during the experiment, which accounts for the presence of zirconia-containing major phases. To circumvent the problem of zirconia dissolution, we repeated the experiments using mixtures of the limestone-aggregate concrete and the doped urania in molybdenum crucibles. These studies show that thermodynamic calculations of the release of refractory fission products will yield release fractions that are a factor of sixteen too high if the effects of zirconate formation are ignored

  8. An application program for fission product decay heat calculations

    International Nuclear Information System (INIS)

    Pham, Ngoc Son; Katakura, Jun-ichi

    2007-10-01

    The precise knowledge of decay heat is one of the most important factors in safety design and operation of nuclear power facilities. Furthermore, decay heat data also play an important role in design of fuel discharges, fuel storage and transport flasks, and in spent fuel management and processing. In this study, a new application program, called DHP (Decay Heat Power program), has been developed for exact decay heat summation calculations, uncertainty analysis, and for determination of the individual contribution of each fission product. The analytical methods were applied in the program without any simplification or approximation, in which all of linear and non-linear decay chains, and 12 decay modes, including ground state and meta-stable states, are automatically identified, and processed by using a decay data library and a fission yield data file, both in ENDF/B-VI format. The window interface of the program is designed with optional properties which is very easy for users to run the code. (author)

  9. Baseline Glass Development for Combined Fission Products Waste Streams

    International Nuclear Information System (INIS)

    Crum, Jarrod V.; Billings, Amanda Y.; Lang, Jesse B.; Marra, James C.; Rodriguez, Carmen P.; Ryan, Joseph V.; Vienna, John D.

    2009-01-01

    Borosilicate glass was selected as the baseline technology for immobilization of the Cs/Sr/Ba/Rb (Cs), lanthanide (Ln) and transition metal fission product (TM) waste steams as part of a cost benefit analysis study.(1) Vitrification of the combined waste streams have several advantages, minimization of the number of waste forms, a proven technology, and similarity to waste forms currently accepted for repository disposal. A joint study was undertaken by Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL) to develop acceptable glasses for the combined Cs + Ln + TM waste streams (Option 1) and Cs + Ln combined waste streams (Option 2) generated by the AFCI UREX+ set of processes. This study is aimed to develop baseline glasses for both combined waste stream options and identify key waste components and their impact on waste loading. The elemental compositions of the four-corners study were used along with the available separations data to determine the effect of burnup, decay, and separations variability on estimated waste stream compositions.(2-5) Two different components/scenarios were identified that could limit waste loading of the combined Cs + LN + TM waste streams, where as the combined Cs + LN waste stream has no single component that is perceived to limit waste loading. Combined Cs + LN waste stream in a glass waste form will most likely be limited by heat due to the high activity of Cs and Sr isotopes.

  10. Map of calculated radioactivity of fission product, 3

    International Nuclear Information System (INIS)

    Takeda, Tsuneo

    1977-02-01

    In this work, the radioactivities of fission products were calculated and summarized in contour maps and tables depending on irradiation and cooling times. The irradiation condition and other parameters used for the present calculation are shown in the followings. Neutron flux (N sub(th)): 3x10 13 n/sec/cm 2 Atom number of uranium: 1 mole (6x10 23 , ca. 271 gUO 2 ) Enrichment of U-235: 2.7% Range of irradiation time: 60-6x10 7 sec (ca. 1.9 y) Range of cooling time: 60-6x10 7 sec (ca. 1.9 y). Values of the neutron flux and the enrichment treated here are representative for common LWRs. The maps and tables of 560 nuclides are divided and compiled into the following three volumes. Vol. I: Maps of radioactivity of overall total, element total and each nuclide (Ni - Zr), Vol. II: Maps of radioactivity of each nuclide (Nb - Sb), Vol. III: Maps of radioactivity of each nuclide (Te - Tm). (auth.)

  11. Map of calculated radioactivity of fission product, (1)

    International Nuclear Information System (INIS)

    Takeda, Tsuneo

    1977-02-01

    In this work, the radioactivities of fission products were calculated and summarized in contour maps and tables depending on irradiation and cooling times. The irradiation condition and other parameters used for the present calculation are shown in the followings. Neutron flux (N sub(th)): 3x10 13 n/sec/cm 2 Atom number of uranium: 1 mole (6x10 23 , ca. 271 gUO 2 ) Enrichment of U-235: 2.7% Range of irradiation time: 60-6x10 7 sec (ca. 1.9 y) Range of cooling time: 60-6x10 7 sec (ca. 1.9 y). Values of the neutron flux and the enrichment treated here are representative for common LWRs. The maps and tables of 560 nuclides are divided and compiled into the following three volumes. Vol. I Maps of radioactivity of overall total, element total and each nuclide (Ni - Zr) Vol. II Maps of radioactivity of each nuclide (Nb - Sb) Vol. III Maps of radioactivity of each nuclide (Te - Tm) (auth.)

  12. Map of calculated radioactivity of fission product, 2

    International Nuclear Information System (INIS)

    Takeda, Tsuneo

    1977-02-01

    In this work, the radioactivities of fission products were calculated and summarized in contour maps and tables depending on irradiation and cooling times. The irradiation condition and other parameters used for the present calculation are shown in the followings. Neutron flux (N sub(th)): 3x10 13 n/sec/cm 2 Atom number of uranium: 1 mole (6x10 23 , ca. 271 gUO 2 ) Enrichment of U-235: 2.7% Range of irradiation time: 60-6x10 7 sec (ca. 1.9 y) Range of cooling time: 60-6x10 7 sec (ca. 1.9 y). Values of the neutron flux and the enrichment treated here are representative for common LWRs. The maps and tables of 560 nuclides are divided and compiled into the following three volumes. Vol. I: Maps of radioactivity of overall total, element total and each nuclide (Ni - Zr), Vol. II: Maps of radioactivity of each nuclide (Nb - Sb), Vol. III: Maps of radioactivity of each nuclide (Te - Tm). (auth.)

  13. A Covariance Generation Methodology for Fission Product Yields

    Directory of Open Access Journals (Sweden)

    Terranova N.

    2016-01-01

    Full Text Available Recent safety and economical concerns for modern nuclear reactor applications have fed an outstanding interest in basic nuclear data evaluation improvement and completion. It has been immediately clear that the accuracy of our predictive simulation models was strongly affected by our knowledge on input data. Therefore strong efforts have been made to improve nuclear data and to generate complete and reliable uncertainty information able to yield proper uncertainty propagation on integral reactor parameters. Since in modern nuclear data banks (such as JEFF-3.1.1 and ENDF/BVII.1 no correlations for fission yields are given, in the present work we propose a covariance generation methodology for fission product yields. The main goal is to reproduce the existing European library and to add covariance information to allow proper uncertainty propagation in depletion and decay heat calculations. To do so, we adopted the Generalized Least Square Method (GLSM implemented in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation, developed at CEA-Cadarache. Theoretical values employed in the Bayesian parameter adjustment are delivered thanks to a convolution of different models, representing several quantities in fission yield calculations: the Brosa fission modes for pre-neutron mass distribution, a simplified Gaussian model for prompt neutron emission probability, theWahl systematics for charge distribution and the Madland-England model for the isomeric ratio. Some results will be presented for the thermal fission of U-235, Pu-239 and Pu-241.

  14. Fission product yield measurements using monoenergetic photon beams

    Directory of Open Access Journals (Sweden)

    Krishichayan

    2017-01-01

    Full Text Available Measurements of fission products yields (FPYs are an important source of information on the fission process. During the past couple of years, a TUNL-LANL-LLNL collaboration has provided data on the FPYs from quasi monoenergetic neutron-induced fission on 235U, 238U, and 239Pu and has revealed an unexpected energy dependence of both asymmetric fission fragments at energies below 4 MeV. This peculiar FPY energy dependence was more pronounced in neutron-induced fission of 239Pu. In an effort to understand and compare the effect of the incoming probe on the FPY distribution, we have carried out monoenergetic photon-induced fission experiments on the same 235U, 238U, and 239Pu targets. Monoenergetic photon beams of Eγ = 13.0 MeV were provided by the HIγS facility, the world's most intense γ-ray source. In order to determine the total number of fission events, a dual-fission chamber was used during the irradiation. These irradiated samples were counted at the TUNL's low-background γ-ray counting facility using high efficient HPGe detectors over a period of 10 weeks. Here we report on our first ever photofission product yield measurements obtained with monoenegetic photon beams. These results are compared with neutron-induced FPY data.

  15. Brief description of out-of-pile test facilities for study in corrosion and fission product behaviour in flowing sodium

    International Nuclear Information System (INIS)

    Iizawa, K.; Sekiguchi, N.; Atsumo, H.

    1976-01-01

    The experimental methods to perform tests for study in corrosion and fission products behaviour in flowing sodium are outlined. Flow diagrams for the activated materials and fission products behaviour test loop are given

  16. DFT study of the hexagonal high-entropy alloy fission product system

    Energy Technology Data Exchange (ETDEWEB)

    King, D.J.M., E-mail: daniel.miks@live.com [School of Electrical Engineering, University of New South Wales, Kensington, 2052, NSW (Australia); Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW (Australia); Burr, P.A.; Obbard, E.G. [School of Electrical Engineering, University of New South Wales, Kensington, 2052, NSW (Australia); Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW (Australia); Middleburgh, S.C. [Westinghouse Electric Sweden AB, SE-721 63, Västerås (Sweden); Department of Materials, Imperial College London, South Kensington, London, SW7 2AZ (United Kingdom); KTH Royal Institute of Technology, Reactor Physics, 106 91 Stockholm (Sweden)

    2017-05-15

    The metallic phase fission product containing Mo-Pd-Rh-Ru-Tc can be described as a hexagonal high-entropy alloy (HEA) and is thus investigated using atomic scale simulation techniques relevant to HEAs. Contrary to previous assumptions, the removal of Tc from the system to form the Mo-Pd-Rh-Ru analog is predicted to reduce the stability of the solid solution to the point that σ-Mo{sub 5}Ru{sub 3} may precipitate out at typical fuel operating temperatures. The drive for segregation is attributed to the increased stability of the solid solution with the ejection of Mo and Ru. When Tc is included in the system, a single phase hexagonal solid solution is expected to form for a wider range of compositions. Furthermore, when cooled below 700 °C, this single phase solid solution is predicted to transition to a partially ordered structure. Future studies using the Tc-absent analogue will need to take these structural and chemical deliberations into consideration.

  17. Migration studies of fission product nuclides in rocks. Pt.5: Diffusion and permeability of nuclide 125I in marble

    International Nuclear Information System (INIS)

    Wen Ruiyuan; Gao Hongcheng; Wang Xiangyun

    1996-01-01

    The migration behaviour of nuclide 125 I, as a simulation of the long lived fission product 129 I, in marble is studied in self-designed cells. A series of the most important parameters of diffusion and permeability (e.g., intrinsic diffusion coefficient, dispersion coefficient and interstitial flow velocity, etc.) are determined. Based on the differential equation of the nuclide migration, the distribution function and numerical solution of 125 I in marble are presented. The results show that the migration velocity of 125 I in marble is fast, indicating that it is not suitable to dispose nuclear waste in marble

  18. 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

  19. Deep Atomic Binding (DAB) Approach in Interpretation of Fission Products Behavior in Terrestrial and Water Ecosystems

    International Nuclear Information System (INIS)

    Ajlouni, Abdul-Wali M.S.

    2006-01-01

    A large number of studies and models were established to explain the fission products (FP) behavior within terrestrial and water ecosystems, but a number of behaviors were non understandable, which always attributed to unknown reasons. According to DAB hypothesis, almost all fission products behaviors in terrestrial and water ecosystems could be interpreted in a wide coincidence. The gab between former models predictions, and field behavior of fission products after accidents like Chernobyl have been explained. DAB represents a tool to reduce radio-phobia as well as radiation protection expenses. (author)

  20. Tables and figures from JNDC Nuclear Data Library of fission products, version 2

    International Nuclear Information System (INIS)

    Ihara, Hitoshi

    1989-11-01

    The content of JNDC (Japanese Nuclear Data Committee) FP (Fission Product) Nuclear Data Library version 2 for 1227 fission products is presented in the form of tables and figures. The library is inclusive of evaluated decay data such as decay constant, Q-value, average energies of beta, gamma and internal conversion electron, spin-parity, branching ratio of each decay mode and fission yield. The neutron capture cross-sections are also contained for 166 nuclides. The mass number of the fission product nuclides ranges from A = 66 to A = 172. (author)

  1. Temperature dependent fission product removal efficiency due to pool scrubbing

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Shunsuke, E-mail: suchida@iae.or.jp [Institute of Applied Energy, 1-14-2, Nishi-Shimbashi, Minato-ku, Tokyo 105-0003 (Japan); Itoh, Ayumi; Naitoh, Masanori; Okada, Hidetoshi; Suzuki, Hiroyuki [Institute of Applied Energy, 1-14-2, Nishi-Shimbashi, Minato-ku, Tokyo 105-0003 (Japan); Hanamoto, Yukio [KAKEN, Inc., 1044, Hori-machi, Mito 310-0903 (Japan); Osakabe, Masahiro [Tokyo University of Marine Science & Technology, Koutou-ku, Tokyo 135-8533 (Japan); Fujikawa, Masahiro [Japan Broadcasting Corporation, 2-2-1, Jinnan, Shibuya-ku, Tokyo 150-8001 (Japan)

    2016-03-15

    Highlights: • Pool temperature effects on the FP removal were not clearly concluded in the previous publications. • It was confirmed that the removal efficiency decreased with temperature around the boiling point. • A modified empirical formula for FP removal was proposed as a function of sub-cooling temperature. • DF could be predicted with an accuracy within a factor of 2 with the proposed formula. - Abstract: The wet-well of boiling water reactors plays important roles not only to suppress the pressure in the primary containment vessel due to steam scrubbing effects during severe accidents but also to mitigate release of radioactive fission products (FP), aerosols and particulates, into the environment. The effects of steam scrubbing in the wet-well on FP removal have been well studied and reported by changing major parameters determining the removal efficiencies, e.g., aerosol diameters, submergence (depth of scrubbing nozzles) and steam/non-condensable gas volume fraction. Unfortunately, the effects of pool temperature on the FP removal were not clearly concluded in the previous publications, though it would be easily expected that boiling in the pool resulted in reduced aerosol removal efficiency. In order to determine the temperature effects on FP removal efficiency, amounts of cesium in aerosols released from scrubbing pool were measured by changing pool temperature in mini and medium scale scrubbing experiments, and then, it was confirmed that the removal efficiency clearly decreased with temperature around the boiling point. Then, a modified empirical formula to express the FP removal around the boiling point temperature was proposed as a function of sub-cooling temperature by applying the effective steam volume fraction, which was designated as the volume ratio of condensed steam in the pool versus the sum of input steam and non-condensable gas. By comparing the measured removal efficiency with the calculated, it was validated that the

  2. Temperature dependent fission product removal efficiency due to pool scrubbing

    International Nuclear Information System (INIS)

    Uchida, Shunsuke; Itoh, Ayumi; Naitoh, Masanori; Okada, Hidetoshi; Suzuki, Hiroyuki; Hanamoto, Yukio; Osakabe, Masahiro; Fujikawa, Masahiro

    2016-01-01

    Highlights: • Pool temperature effects on the FP removal were not clearly concluded in the previous publications. • It was confirmed that the removal efficiency decreased with temperature around the boiling point. • A modified empirical formula for FP removal was proposed as a function of sub-cooling temperature. • DF could be predicted with an accuracy within a factor of 2 with the proposed formula. - Abstract: The wet-well of boiling water reactors plays important roles not only to suppress the pressure in the primary containment vessel due to steam scrubbing effects during severe accidents but also to mitigate release of radioactive fission products (FP), aerosols and particulates, into the environment. The effects of steam scrubbing in the wet-well on FP removal have been well studied and reported by changing major parameters determining the removal efficiencies, e.g., aerosol diameters, submergence (depth of scrubbing nozzles) and steam/non-condensable gas volume fraction. Unfortunately, the effects of pool temperature on the FP removal were not clearly concluded in the previous publications, though it would be easily expected that boiling in the pool resulted in reduced aerosol removal efficiency. In order to determine the temperature effects on FP removal efficiency, amounts of cesium in aerosols released from scrubbing pool were measured by changing pool temperature in mini and medium scale scrubbing experiments, and then, it was confirmed that the removal efficiency clearly decreased with temperature around the boiling point. Then, a modified empirical formula to express the FP removal around the boiling point temperature was proposed as a function of sub-cooling temperature by applying the effective steam volume fraction, which was designated as the volume ratio of condensed steam in the pool versus the sum of input steam and non-condensable gas. By comparing the measured removal efficiency with the calculated, it was validated that the

  3. Application of dynamic pseudo fission products and actinides for accurate burnup calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Kloosterman, J.L.

    1996-09-01

    The introduction of pseudo fission products for accurate fine-group spectrum calculations during burnup is discussed. The calculation of the density of the pseudo nuclides is done before each spectrum calculation from the actual densities and their cross sections of all nuclides to be lumped into a pseudo fission product. As there are also many actinides formed in the fuel during its life cycle, a pseudo actinide with fission cross section is also introduced. From a realistic burnup calculation it is demonstrated that only a few fission products and actinides need to be included explicitly in a spectrum calculation. All other fission products and actinides can be accurately represented in the pseudo nuclides. (author)

  4. Fission product and actinide data evaluations for ENDF/B--V

    International Nuclear Information System (INIS)

    Schenter, R.E.

    1978-05-01

    The planned content and performance of fission product and actinide nuclide evaluations for the ENDF/B-V collection of data are reviewed. Representative values of parameters for a few nuclides are shown. 10 figures, 5 tables

  5. User's manual for computer code RIBD-II, a fission product inventory code

    International Nuclear Information System (INIS)

    Marr, D.R.

    1975-01-01

    The computer code RIBD-II is used to calculate inventories, activities, decay powers, and energy releases for the fission products generated in a fuel irradiation. Changes from the earlier RIBD code are: the expansion to include up to 850 fission product isotopes, input in the user-oriented NAMELIST format, and run-time choice of fuels from an extensively enlarged library of nuclear data. The library that is included in the code package contains yield data for 818 fission product isotopes for each of fourteen different fissionable isotopes, together with fission product transmutation cross sections for fast and thermal systems. Calculational algorithms are little changed from those in RIBD. (U.S.)

  6. Shielding calculation of a hot cell for the processing of fission products

    International Nuclear Information System (INIS)

    Rocha, A.C.S. da; Pina, J.L.S. de; Silva, J.J.G. da.

    1986-12-01

    A dose rate estimation is made for an operator of a lead wall, fission products processing hot cell, in a distance of 50 cm from the emission source, at Brazilian Institute of Nuclear Engineering (IEN). (L.C.J.A.)

  7. 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)

  8. 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

  9. 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.

  10. The LANL C-NR counting room and fission product yields

    Energy Technology Data Exchange (ETDEWEB)

    Jackman, Kevin Richard [Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)

    2015-09-21

    This PowerPoint presentation focused on the following areas: LANL C-NR counting room; Fission product yields; Los Alamos Neutron wheel experiments; Recent experiments ad NCERC; and Post-detonation nuclear forensics

  11. Interaction of noble-metal fission products with pyrolytic silicon carbide

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  12. Fission product model for lattice calculation of high conversion boiling water reactor

    International Nuclear Information System (INIS)

    Iijima, S.; Yoshida, T.; Yamamoto, T.

    1988-01-01

    A high precision fission product model for boiling water reactor (BWR) lattice calculation was developed, which consists of 45 nuclides to be treated explicitly and one nonsaturating pseudo nuclide. This model is applied to a high conversion BWR lattice calculation code. From a study based on a three-energy-group calculation of fission product poisoning due to full fission products and explicitly treated nuclides, the multigroup capture cross sections and the effective fission yields of the pseudo nuclide are determined, which do not depend on fuel types or reactor operating conditions for a good approximation. Apart from nuclear data uncertainties, the model and the derived pseudo nuclide constants would predict the fission product reactivity within an error of 0.1% Δk at high burnup

  13. Libraries of decay data and fission product yields in the ABBN-93 constant set

    International Nuclear Information System (INIS)

    Zabrodskaya, S.V.; Nikolaev, M.N.; Tsibulya, A.M.

    2001-01-01

    This paper describes three new libraries in the Abb. constant set which are essential for calculating radioactivity: basic decay data, radioactive decay photon spectra and fission product yields. (author)

  14. TMI-2 [Three Mile Island] fission product inventory program: FY-85 status report

    International Nuclear Information System (INIS)

    Langer, S.; Croney, S.T.; Akers, D.W.; Russell, M.L.

    1986-11-01

    This report presents the status of the TMI-2 fission product inventory program through May 1985. The fission product inventory program is an assessment of the location of fission products distributed in the plant as a result of the TMI-2 accident. Included in this report are principal results of samples from the reactor building where most of the mobile fission products (i.e., radiocesium and iodine) are expected to be found. The data are now complete enough for most reactor components; therefore, it is possible to direct the balance of the examination and sampling program to areas and components where it is likely to be most productive. Those areas are the reactor core and the reactor building basement, with emphasis on the currently unsampled portions of the core

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

    International Nuclear Information System (INIS)

    1985-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  18. Decommissioning of the Fission Product Development Laboratory at Holifield National Laboratory

    International Nuclear Information System (INIS)

    Schaich, R.W.

    1975-01-01

    The decontamination of the Fission Product Development Laboratory was initiated in FY 1975 after 17 years of processing fission product waste streams to produce commercial quantities of 90 Sr, 137 Cs, 144 Ce, and 147 Pm. The objective of the decommissioning program is the removal of all radiation and contamination areas in the facility to a level which will be compatible with the environment in the foreseeable future

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

    International Nuclear Information System (INIS)

    Harnden, A.M.C.

    1995-09-01

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

  20. Cross sections of the lumped fission products for the AMZ library

    International Nuclear Information System (INIS)

    Ono, S.; Corcueca, R.P.; Nascimento, J.A.

    1985-01-01

    The preparation of the lumped fission product cross section for the AMZ library is described. For this purpose 100 nuclides were selected. The cross sections for each nuclide were generated by the NJOY code with evaluated nuclear data from ENDF/B-V, complemented with ENDF/B-IV data. A comparison is performed between the data obtained and the lumped fission product cross section of JFS-II [pt

  1. TRAFIC, a computer program for calculating the release of metallic fission products from an HTGR core

    International Nuclear Information System (INIS)

    Smith, P.D.

    1978-02-01

    A special purpose computer program, TRAFIC, is presented for calculating the release of metallic fission products from an HTGR core. The program is based upon Fick's law of diffusion for radioactive species. One-dimensional transient diffusion calculations are performed for the coated fuel particles and for the structural graphite web. A quasi steady-state calculation is performed for the fuel rod matrix material. The model accounts for nonlinear adsorption behavior in the fuel rod gap and on the coolant hole boundary. The TRAFIC program is designed to operate in a core survey mode; that is, it performs many repetitive calculations for a large number of spatial locations in the core. This is necessary in order to obtain an accurate volume integrated release. For this reason the program has been designed with calculational efficiency as one of its main objectives. A highly efficient numerical method is used in the solution. The method makes use of the Duhamel superposition principle to eliminate interior spatial solutions from consideration. Linear response functions relating the concentrations and mass fluxes on the boundaries of a homogeneous region are derived. Multiple regions are numerically coupled through interface conditions. Algebraic elimination is used to reduce the equations as far as possible. The problem reduces to two nonlinear equations in two unknowns, which are solved using a Newton Raphson technique

  2. Separation of uranium, plutonium and fission products on zirconium phosphate, Part 1 - Adsorption equilibria and kinetics

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-02-15

    The distribution coefficients of UO{sub 2}{sup ++}, PuO{sub 2}{sup ++}, Pu{sup 3+}, Pu{sup 4+}, Fe{sup 3+}, {sup 137}Cs{sup +}, {sup 90}Sr{sup ++}, {sup 95}Zr{sup +}+{sup 95}Nb{sup 5+}, {sup 106}Ru and {sup 144}Ce{sup 3+} were determined in the system zirconium phosphate-aqueous solution of HNO{sub 3}. As for the exchange reation Cs{sup +}/H{sup +} and Sr{sup ++}/2H{sup +}, it has been shown that the mass action law can be applied. For these reactions the corresponding equilibrium constants were calculated. The rates of adsorption of Cs{sup +}, Sr{sup ++}, Fe{sup 3+} and Pu{sup 4+} from solutions of a fixed HNO{sub 3} concentration were studied, and empirical rate equations were derived. The experimental data confirm that UO{sub 2}{sup ++} can be separated from Pu{sup 4+}. Among the fission products, {sup 90}Sr, {sup 106}Ru and {sup 144}Ce mainly follow the fraction of uranium, while {sup 137}Cs, {sup 95}Zr and {sup 95}Nb follow the plutonium fraction. Separations within the fractions are possible (author)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

    Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the pool. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool was estimated by product of the mass concentration and energy conversion factor of each fission product. For the calculation of heat generation rate in the pool, twenty-nine elements were chosen and classified by their chemical properties. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis was performed for the TMI-2 accident. The pool is assumed to be a partially filled hemispherical geometry and the change of pool geometry during the numerical calculation was neglected. Results of the numerical calculation revealed that the peak temperature of the molten pool significantly decreased and most of the volatile fission products were released from the molten pool during the accident. (authors)

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  5. Use of dwell time concept in fission product inventory assessment for CANDU reactors

    International Nuclear Information System (INIS)

    Bae, C.J.; Choi, J.H.; Hwang, H.R.; Seo, J.T.

    2003-01-01

    A realistic approach in calculating the initial fission product inventory within the CANFLEX-NU fuel has been assessed for its applicability to the single channel event safety analysis for CANDU reactors. This approach is based on the dwell time concept in which the accident is assumed to occur at the dwell time when the summation of fission product inventory for all isotopes becomes largest. However, in the current conservative analysis, the maximum total inventory and the corresponding gap inventory for each isotope are used as the initial fission product inventories regardless of the accident initiation time. The fission product inventory analysis has been performed using ELESTRES code considering power histories and burnup of the fuel bundles in the limiting channel. The analysis results showed that the total fission product inventory is found to be largest at 20% dwell time. Therefore, the fission product inventory at 20% dwell time can be used as the initial condition for the single channel event for the CANDU 6 reactors. (author)

  6. 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)

  7. Radiation Damage and Fission Product Release in Zirconium Nitride

    Energy Technology Data Exchange (ETDEWEB)

    Egeland, Gerald W. [New Mexico Inst. of Mining and Technology, Socorro, NM (United States)

    2005-08-29

    Zirconium nitride is a material of interest to the AFCI program due to some of its particular properties, such as its high melting point, strength and thermal conductivity. It is to be used as an inert matrix or diluent with a nuclear fuel based on transuranics. As such, it must sustain not only high temperatures, but also continuous irradiation from fission and decay products. This study addresses the issues of irradiation damage and fission product retention in zirconium nitride through an assessment of defects that are produced, how they react, and how predictions can be made as to the overall lifespan of the complete nuclear fuel package. Ion irradiation experiments are a standard method for producing radiation damage to a surface for observation. Cryogenic irradiations are performed to produce the maximum accumulation of defects, while elevated temperature irradiations may be used to allow defects to migrate and react to form clusters and loops. Cross-sectional transmission electron microscopy and grazing-incidence x-ray diffractometry were used in evaluating the effects that irradiation has on the crystal structure and microstructure of the material. Other techniques were employed to evaluate physical effects, such as nanoindentation and helium release measurements. Results of the irradiations showed that, at cryogenic temperatures, ZrN withstood over 200 displacements per atom without amorphization. No significant change to the lattice or microstructure was observed. At elevated temperatures, the large amount of damage showed mobility, but did not anneal significantly. Defect clustering was possibly observed, yet the size was too small to evaluate, and bubble formation was not observed. Defects, specifically nitrogen vacancies, affect the mechanical behavior of ZrN dramatically. Current and previous work on dislocations shows a distinct change in slip plane, which is evidence of the bonding characteristics. The stacking-fault energy changes dramatically with

  8. 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

  9. Contribution to research on the metabolism of fission product. Studies on the physico-chemical state and the metabolic fate of radio-cerium solution; Contribution a l'etude du metabolisme des produits de fission. Recherches sur l'etat physico-chimique et le devenir metabolique des solutions de radiocerium

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-07-01

    This paper describes a study of the physico-chemical state of radio-cerium in dilute solutions on the tracer scale, as a function of the pH of the solution. The way in which this radioelement is transported in the blood is studied in vitro and in vivo, with reference to the ionic or colloidal state of the radio-cerium used. The distribution of cerium amongst the various components of the blood is studied by a new method of blood fractionation and by paper electrophoresis. Evidence of a cerium globulin connection is shown in the case of ionic cerium. A study of the initial distribution of radio-cerium in rats, after intravenous administration of ionic or colloidal solutions, shows considerable differences according to the physico-chemical state of the cerium injected. (author) [French] Ce travail presente une etude de l'etat physico-chimique du radiocerium en solutions diluees a l'echelle des indicateurs, en fonction du pH de la solution. En fonction de l'etat ionique ou colloidal du radiocerium utilise, le mode de transport de ce radioelement dans le sang est etudie in vitro et in vivo. La distribution du cerium entre les differents constituants du sang est etudiee par une methode nouvelle de fractionnement du sang et par electrophorese sur papier. L'existence d'une liaison cerium-globuline est mise en evidence dans le cas du cerium ionique. L'etude de la distribution initiale du radiocerium, chez le rat apres administration par voie veineuse de solutions ionique ou colloidale, montre des differences importantes en fonction de l'etat physico-chimique du cerium injecte. (auteur)

  10. Utilization of ruthenium volatilization at heating of residue containing phosphates and nitrates for ruthenium separation and for its qualitative proof

    International Nuclear Information System (INIS)

    Holgye, Z.

    1979-01-01

    The volatility of ruthenium during the heating of a residue after evaporation of a solution containing ruthenium, phosphates and nitrates may be utilized for the separation of ruthenium from various substances. Sup(103,106) Ru may be rapidly, selectively, and quantitatively separated from fission products mixture. Ruthenium may be also separated in this way from various inorganic salts or from biological material. The volatility of ruthenium may be used also for its qualitative proof. (author)

  11. A Research Program for Fission Product/Dust Transport in HTGR's

    International Nuclear Information System (INIS)

    Loyalka, Sudarshan

    2016-01-01

    High and Very High Temperatures Gas Reactors (HTGRs/VHTRs) have five barriers to fission product (FP) release: the TRISO fuel coating, the fuel elements, the core graphite, the primary coolant system, and the reactor building. This project focused on measurements and computations of FP diffusion in graphite, FP adsorption on graphite and FP interactions with dust particles of arbitrary shape. Diffusion Coefficients of Cs and Iodine in two nuclear graphite were obtained by the release method and use of Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS) and Instrumented Neutron Activation Analysis (INAA). A new mathematical model for fission gas release from nuclear fuel was also developed. Several techniques were explored to measure adsorption isotherms, notably a Knudsen Effusion Mass Spectrometer (KEMS) and Instrumented Neutron Activation Analysis (INAA). Some of these measurements are still in progress. The results will be reported in a supplemental report later. Studies of FP interactions with dust and shape factors for both chain-like particles and agglomerates over a wide size range were obtained through solutions of the diffusion and transport equations. The Green's Function Method for diffusion and Monte Carlo technique for transport were used, and it was found that the shape factors are sensitive to the particle arrangements, and that diffusion and transport of FPs can be hindered. Several journal articles relating to the above work have been published, and more are in submission and preparation.

  12. Fission product behavior in high-temperature water: CsI vs MoO4

    Science.gov (United States)

    Kanjana, K.; Silva, K.; Channuie, J.

    2017-09-01

    Fission product behaviors of Cs, a major element released in a severe nuclear accident, still remain unclear. The question frequently addressed is whether Cs released will be in the form of Cs2MoO4 or CsOH. This is a challenging issue since it has been demonstrated that the reaction between Cs2MoO4 and water leading to CsOH production is thermodynamically favored. The present research aims at investigation of CsOH generation through this chemical channel. A high-temperature setup with a flow system based on the cooling system of a water-cooled nuclear reactor has been assembled. The reaction between aqueous solutions of CsI and Na2MoO4 in a high-corrosion-resistant hot cell (Hastelloy) has been studied up to 80 °C in deoxygenated system. The products have been characterized using FTIR and XRD. The results have shown that there is no reaction between CsI and Na2MoO4 under the experimental conditions.

  13. Actinide, lanthanide and fission product speciation and electrochemistry in high and low temperature ionic melts

    Energy Technology Data Exchange (ETDEWEB)

    Bhatt, Anand I.; Kinoshita, Hajime; Koster, Anne L.; May, Iain; Sharrad, Clint A.; Volkovich, Vladimir A.; Fox, O. Danny; Jones, Chris J.; Lewin, Bob G.; Charnock, John M.; Hennig, Christoph

    2004-07-01

    There is currently a great deal of research interest in the development of molten salt technology, both classical high temperature melts and low temperature ionic liquids, for the electrochemical separation of the actinides from spent nuclear fuel. We are interested in gaining a better understanding of actinide and key fission product speciation and electrochemical properties in a range of melts. Our studies in high temperature alkali metal melts (including LiCl and LiCl-KCl and CsCl-NaCl eutectics) have focussed on in-situ species of U, Th, Tc and Ru using X-ray absorption spectroscopy (XAS, both EXAFS and XANES) and electronic absorption spectroscopy (EAS). We report unusual actinide speciation in high temperature melts and an evaluation of the likelihood of Ru or Tc volatilization during plant operation. Our studies in lower temperature melts (ionic liquids) have focussed on salts containing tertiary alkyl group 15 cations and the bis(tri-fluor-methyl)sulfonyl)imide anion, melts which we have shown to have exceptionally wide electrochemical windows. We report Ln, Th, U and Np speciation (XAS, EAS and vibrational spectroscopy) and electrochemistry in these melts and relate the solution studies to crystallographic characterised benchmark species. (authors)

  14. Actinide, lanthanide and fission product speciation and electrochemistry in high and low temperature ionic melts

    International Nuclear Information System (INIS)

    Bhatt, Anand I.; Kinoshita, Hajime; Koster, Anne L.; May, Iain; Sharrad, Clint A.; Volkovich, Vladimir A.; Fox, O. Danny; Jones, Chris J.; Lewin, Bob G.; Charnock, John M.; Hennig, Christoph

    2004-01-01

    There is currently a great deal of research interest in the development of molten salt technology, both classical high temperature melts and low temperature ionic liquids, for the electrochemical separation of the actinides from spent nuclear fuel. We are interested in gaining a better understanding of actinide and key fission product speciation and electrochemical properties in a range of melts. Our studies in high temperature alkali metal melts (including LiCl and LiCl-KCl and CsCl-NaCl eutectics) have focussed on in-situ species of U, Th, Tc and Ru using X-ray absorption spectroscopy (XAS, both EXAFS and XANES) and electronic absorption spectroscopy (EAS). We report unusual actinide speciation in high temperature melts and an evaluation of the likelihood of Ru or Tc volatilization during plant operation. Our studies in lower temperature melts (ionic liquids) have focussed on salts containing tertiary alkyl group 15 cations and the bis(tri-fluor-methyl)sulfonyl)imide anion, melts which we have shown to have exceptionally wide electrochemical windows. We report Ln, Th, U and Np speciation (XAS, EAS and vibrational spectroscopy) and electrochemistry in these melts and relate the solution studies to crystallographic characterised benchmark species. (authors)

  15. Method to Reduce Long-lived Fission Products by Nuclear Transmutations with Fast Spectrum Reactors.

    Science.gov (United States)

    Chiba, Satoshi; Wakabayashi, Toshio; Tachi, Yoshiaki; Takaki, Naoyuki; Terashima, Atsunori; Okumura, Shin; Yoshida, Tadashi

    2017-10-24

    Transmutation of long-lived fission products (LLFPs: 79 Se, 93 Zr, 99 Tc, 107 Pd, 129 I, and 135 Cs) into short-lived or non-radioactive nuclides by fast neutron spectrum reactors without isotope separation has been proposed as a solution to the problem of radioactive wastes disposal. Despite investigation of many methods, such transmutation remains technologically difficult. To establish an effective and efficient transmutation system, we propose a novel neutron moderator material, yttrium deuteride (YD 2 ), to soften the neutron spectrum leaking from the reactor core. Neutron energy spectra and effective half-lives of LLFPs, transmutation rates, and support ratios were evaluated with the continuous-energy Monte Carlo code MVP-II/MVP-BURN and the JENDL-4.0 cross section library. With the YD 2 moderator in the radial blanket and shield regions, effective half-lives drastically decreased from 106 to 102 years and the support ratios reached 1.0 for all six LLFPs. This successful development and implementation of a transmutation system for LLFPs without isotope separation contributes to a the ability of fast spectrum reactors to reduce radioactive waste by consuming their own LLFPs.

  16. Biosorption and retention of several actinide and fission-product elements by biomass from Mycobacterium phlei

    International Nuclear Information System (INIS)

    Bouby, M.; MacCordick, H.J.; Billard, I.

    1996-01-01

    The properties of mobile, 5% w/w cell suspensions of Mycobacterium phlei have been examined for their capacity to adsorb and retain uranyl(VI) and neptunyl(V) cations from nitrate-buffered solutions at pH 1. Equilibrium conditions of sorption were attained after 3 hours for concentrations (C) in the range 0.015-18 mM cation and indicated a maximum specific adsorption capacity (Qe max ) of 182 μmol/g dry biomass for C ≥ 10 mM. NpO 2 + generally showed higher Qe values than UO 2 2+ at corresponding concentrations. Lixiviation tests with cation-loaded biomass in neutral and acidic media indicated that the extent of desorption did not vary extensively between pH 7 and pH 1 and did not exceed 3% for U and 1% for Np ions at pH 7 during 7-day periods of treatment. Analogous experiments with U-loaded biomass subjected to neutron activation prior to lixiviation enabled retention measurements for various fission-product isotopes produced in situ and showed that retention of 239 Np formed within the cellular matrix was >99% at pH 7 and ≥94% at pH 1. (author). 13 refs., 5 figs., 3 tabs

  17. A Research Program for Fission Product/Dust Transport in HTGR’s

    Energy Technology Data Exchange (ETDEWEB)

    Loyalka, Sudarshan [Univ. of Missouri, Columbia, MO (United States)

    2016-02-01

    High and Very High Temperatures Gas Reactors (HTGRs/VHTRs) have five barriers to fission product (FP) release: the TRISO fuel coating, the fuel elements, the core graphite, the primary coolant system, and the reactor building. This project focused on measurements and computations of FP diffusion in graphite, FP adsorption on graphite and FP interactions with dust particles of arbitrary shape. Diffusion Coefficients of Cs and Iodine in two nuclear graphite were obtained by the release method and use of Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS) and Instrumented Neutron Activation Analysis (INAA). A new mathematical model for fission gas release from nuclear fuel was also developed. Several techniques were explored to measure adsorption isotherms, notably a Knudsen Effusion Mass Spectrometer (KEMS) and Instrumented Neutron Activation Analysis (INAA). Some of these measurements are still in progress. The results will be reported in a supplemental report later. Studies of FP interactions with dust and shape factors for both chain-like particles and agglomerates over a wide size range were obtained through solutions of the diffusion and transport equations. The Green's Function Method for diffusion and Monte Carlo technique for transport were used, and it was found that the shape factors are sensitive to the particle arrangements, and that diffusion and transport of FPs can be hindered. Several journal articles relating to the above work have been published, and more are in submission and preparation.

  18. 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

  19. 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

  20. Heat and fission product transport in molten core material pool with crust

    International Nuclear Information System (INIS)

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

    2005-01-01

    Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the reactor vessel during a severe accident. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool is estimated by product of the mass concentration and energy conversion factor of each fission product. Twenty-nine elements are chosen and classified by their chemical properties to calculate heat generation rate in the pool. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis is performed for heat and fission product transport in a molten core material pool during the Three Mile Island Unit 2 (TMI-2) accident. The pool is assumed to be a partially filled hemisphere, whose change in geometry is neglected during the numerical calculation. Calculated results indicate that the peak temperature in the molten pool is significantly lowered, since a substantial amount of the volatile fission products is released from the molten pool during progression of the accident. The results may directly be applied to the existing severe accident analysis codes to more mechanistically determine the thermal load to the reactor vessel lower head during the in-vessel retention

  1. Considerations on the influence of fission products in whole core accidents

    International Nuclear Information System (INIS)

    Meyer Heine, A.; Pattoret, A.; Schmitz, F.

    1977-01-01

    If the hypothetical Whole Core Accidents which are taken into account in reactor safety analysis can change from one country to another, there is nevertheless a general agreement over their description and main phases. Furthermore the important parameters have also been identified by every laboratory. During the development of such core accidents the role of the fission products in essential. It is not the purpose of this paper to give an exhaustive description of the phases which can be influenced by the fission products, we will try however to focus this study on the most important ones. In a second step we will discuss the equation of state of irradiated fuels; here again one principal preoccupation being to quantify the influence of fission products on reactor accidents. It is not our purpose to enter on the fundamental aspects of the equation of state. The studies and the experimental program launched at the CEA will then be described. Special attention will be directed towards the eventual role of fission products in molten fuel-coolant interactions (MFCls) or the events leading to the initiation of whole core accidents. This paper will be limited to oxide fuels. Whether the whole core accident is initiated by a reactivity defect or a coolant coast-down, one has to deal with four great categories of phenomena. Loss of flow: the power is around the nominal value, while the coolant flow has been reduced by a factor of 5 to 10. This induces boiling and clad weakening. Will the plenum pressure lead to a clad rupture? In case of a rupture, what will be the effect on the voiding of the channel? Transient over power: influence of gases from gaseous and volatile fission products on the fuel movements? MFCIs: Influence of the fission products in the mode of contact between fuel and coolant? Influence on the fuel characteristics. Sodium vapour bubble expansion: influence of the fission products on the heat transfer and eventual condensation of the bubble?

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

    International Nuclear Information System (INIS)

    Kolarik, Z.

    1991-11-01

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

  3. Improvement and Validation of an Aerosol Deposition Model in the GAMMA-FP, a Fission Product Analysis Module for VHTRs

    International Nuclear Information System (INIS)

    Yoon, Churl; Lim, Hong Sik

    2013-01-01

    GAMMA-FP (GAs Multicomponent Mixture Analysis-Fission Products module), consists of gaseous and aerosol fission product analysis modules. The aerosol FP module adopts a multi-component and multi-sectional aerosol analysis model that has been developed based on the MAEROS model. For the first work of FP module development, the MAEROS model has been implemented and examined against some analytic solutions and experimental data by Yoo et al. An aerosol transport model was developed and implemented in the GAMMA-FP code, and verified. In this study, the aerosol deposition model in the GAMMA-FP code was improved by adopting recent achievements, and was validated against an experimental data available. The aerosol deposition model in the GAMMA-FP code has been improved and successfully validated against the STORM SR-11 deposition test. The simulation with the improved deposition model predicted the matched results with the experimental data well. For future studies, the aerosol deposition model by flow irregularities will be implemented and validated against the TRANSAT bend effect test

  4. Improvement and Validation of an Aerosol Deposition Model in the GAMMA-FP, a Fission Product Analysis Module for VHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Churl; Lim, Hong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    GAMMA-FP (GAs Multicomponent Mixture Analysis-Fission Products module), consists of gaseous and aerosol fission product analysis modules. The aerosol FP module adopts a multi-component and multi-sectional aerosol analysis model that has been developed based on the MAEROS model. For the first work of FP module development, the MAEROS model has been implemented and examined against some analytic solutions and experimental data by Yoo et al. An aerosol transport model was developed and implemented in the GAMMA-FP code, and verified. In this study, the aerosol deposition model in the GAMMA-FP code was improved by adopting recent achievements, and was validated against an experimental data available. The aerosol deposition model in the GAMMA-FP code has been improved and successfully validated against the STORM SR-11 deposition test. The simulation with the improved deposition model predicted the matched results with the experimental data well. For future studies, the aerosol deposition model by flow irregularities will be implemented and validated against the TRANSAT bend effect test.

  5. Sorption Regularities in Behaviour of Fission-Product Elements during Filtration of their Solutions through Ground; Lois de l'Absorption des Elements Radioactifs Lors du Filtrage des Solutions a Travers les Terrains; 0417 0410 041a 041e 041d 0414 ; Leyes de la Absorcion de los Elementos Radiactivos Cuando se Filtran Soluciones a Traves de las Formaciones Geologicas

    Energy Technology Data Exchange (ETDEWEB)

    Spitsyn, V. I.; Balukova, V. D.; Gromov, V. V.; Zakharov, S. I.; Zhagin, B. P.; Spiridonov, F. M.

    1960-07-01

    Research on the sorption of radioisotopes under natural conditions employing the controlled filtration process was performed. Radioisotopes were introduced into the solution as soon as filtration had become steady and the process continued for four months. Soil samples were then taken by drilling at different depths and analysed to determine their radioisotope content. Diffusion of radioisotopes was observed at depths of up to 10 m; two distinct boundaries of soil-activity decrease were ascertained: at the surface of the site and at the depth of the solution filtration front. In addition, the radiostrontium absorption by natural sorbents, principally pure minerals widely distributed in soils and subsoils, was investigated separately. The presence of calcium ions, even in small quantities, sharply reduces the degree of radiostrontium sorption. However, other conditions being equal, strontium may be absorbed to a greater extent than calcium, according to the composition of the sorbent. The field investigations of radiostrontium sorption and migration showed that when filtering radioactive solutions two possible variants have to be taken into account. In the first case the solutions are discharged into soil unaffected by any flow of ground water. In this situation the radiostrontium is retained by the soil. In the second case, the radioisotopes proceed directly into the water-bearing horizon. The radiostrontium will then migrate with the ground water flow and through the soil and this migration will be further affected by the sorption and desorption processes occurring. The experiments performed demonstrate the ease with which long-lived radioisotopes migrate under natural conditions and call attention to the need for thorough study of ground water problems in connexion with various methods of disposing of radioactive waste into ground. (author) [French] En declenchant un processus de filtration, il a ete possible d'etudier l'absorption des elements radioactifs dans

  6. Absorption by plants of unseparated fission products derived from the hydrogen bomb detonated in the spring of 1954 at Bikini Atoll

    Energy Technology Data Exchange (ETDEWEB)

    Yatazawa, M; Ishihara, T

    1955-01-01

    In a radiochemical survey on the contamination of white clover grown in a field, sample plants were obtained from the same grass land at 3 different times. The ash of each sample was analyzed. It was concluded that radioactive alkaline earths, especially /sup 89/Sr and /sup 90/Sr were selectively accumulated in plants. The selective absorption of Bikini ash by rice plants was also studied. Noncontaminated rice plants were cultivated in the radioactive solution produced from Bikini ash for 20 days. Then the absorption by plants of radioactive elements was examined by chromatographic exchange. From the elution curve and ratio of radioactivity of each separation group, it has become clear that rice plants accumulated larger parts of fission products in their roots and selectively absorbed and translocated radioactive alkaline earths in their shoots even if the absorption ratio of Bikini fission products was comparatively small.

  7. Determination of procedures for transmutation of fission product wastes by fusion neutrons. Volume 2. Final report

    International Nuclear Information System (INIS)

    Lang, G.P.

    1980-12-01

    This study is concerned with the engineering aspects of the transmutation of fission products utilizing neutrons generated in fusion reactors. It is assumed that fusion reactors, although not yet developed, will be available around the turn of the century. Therefore, early studies of this type are appropriate as a guide to the large amount of further investigations that will be needed to fully evaluate this concept. Not all of the radioactive products from light water reactors can be economically transmuted, but it appears that the most hazardous can. This requires that fission-product wastes must first be separated into a number of fractions, and in some instances this must be accomplished with extremely high separation factors. A review of current commercial separation processes and of promising methods that are now in the laboratory stage indicate that the necessary processes can most likely be developed but will require an active and sustained development program. Current fusion reactor concepts were examined as to their suitability for transmuting the separated fission wastes. It was concluded that the long-lived fission products were most amenable to transmutation. The medium-lived fission products, Cs-137 and Sr-90, require higher neutron fluxes than are available in the most developed fusion reactor concepts. Concepts which are less developed may eventually be adaptable as transmuters of these fission products

  8. Prediction of fission product and aerosol behaviour during a postulated severe accident in a LWR

    International Nuclear Information System (INIS)

    Guentay, S.; Aeby, F.; Raguin, M.; Passalacqua, R.

    1990-02-01

    Lack of appropriate energy removal causes fuel elements in a reactor core to overheat and may eventually cause core to degrade. Fission products will be emitted from a degraded reactor core. Aerosols are generated when the vapours of various fuel and structural materials reach a cold environment and nucleate. In addition to the fission products release and aerosol generation taking place in the reactor vessel, some more fission products release and aerosol generation will occur when the molten core debris leaves the pressure vessel bottom head and comes in contact with the pedestal concrete floor. Fission products, if they are released to environment from the containment boundary, exert a great danger to public health. A source term is defined as the quantity, timing, and characteristics of the release of radionuclide material to the environment following a postulated severe accident. At PSI a considerable effort hase been spent in investigating and establishing a source term assessment methodology in order to predict the source term for a given Light Water Reactor (LWR) accident scenario. This report introduces the computer programs and the methods associated with the release of the fission products, generation of the aerosols and behaviour of the aerosols in LWR compartments used for a source term assessment analysis at PSI. (author) 4 figs., 5 tabs., 28 refs

  9. 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)

  10. The effect of stimulated fission products on the structure and the mechanical properties of zircaloy

    International Nuclear Information System (INIS)

    Holub, F.

    1982-01-01

    The objective of investigation was to study the long-term effects of individual simulated fission products on the mechanical properties and the structure of Zircaloy. Tensile Test specimens of Zircaloy were annealed with important simulated fission products at 350 0 C up to 10,000 hours and at higher temperatures (500, 700 0 C) up to 2,000 hours. The principal methods of investigation on annealed Zircaloy specimens were tension tests at room temperature and at 400 0 C, scanning electron microscopy and microprobe technique, X-ray diffraction, X-ray fluorescence, optical metallography. The action of fission products at normal temperatures of reactor operation will give rise to a small enhancement of strength and a small drop of ductility of the fuel cladding material only. At high fuel pin temperatures which may be realized under abnormal operation conditions, some of the fission products potentially will produce detrimental consequences on the integrity of fuel pins. The most effective fission products will be: lanthanum oxide, followed by the earth alkaline oxides and the other rare earth oxides, molybdenum, iodine and cadmium

  11. Analysis of fission product release from HTGR core during transient temperature excursion

    International Nuclear Information System (INIS)

    Saito, Takao; Yamatoya, Naotoshi; Onuma, Mamoru

    1978-01-01

    The computer program ''FRANC'' was developed to calculate the release activity of fission products from a high-temperature gas cooled reactor (HTGR) core during transient temperature excursions such as a hypothetical loss of forced circulation combined with design basis depressurization. The program utilizes a segmented cylindrical core spatial model with the associated values of the prior fuel irradiation history and temperature conditions. The fission product transport and decay chain behavior is expressed by a set of differential equations. This set of equations describes the entire core inventory of fission products by means of calculated parameters based on the detailed spatial core conditions. The program determines the time-dependent amounts of fission product nuclides escaping from the core into the coolant. Coded in Continuous System Simulation Language (CSSL) with double precision, FRANC showed appropriate results for both short- and long-lived fission product nuclides. The sample calculation conducted by applying the program to a large HTGR indicated that it would take about one hour for noble gases and volatile nuclides to be released to the coolant, and several hours for metalic nuclides. (auth.)

  12. 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)

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Malet, G

    1962-07-01

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

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

    International Nuclear Information System (INIS)

    Breederland, D.G.

    1982-01-01

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

  16. Simulating fission product transients via the history-based local-parameter methodology

    International Nuclear Information System (INIS)

    Jenkins, D.A.; Rouben, B.; Salvatore, M.

    1993-01-01

    This paper describes the fission-product-calculation capacity of the history-based local-parameter methodology for evaluating lattice properties for use in core-tracking calculations in CANDU reactors. In addition to taking into account the individual past history of each bundles flux/power level, fuel temperature, and coolant density and temperature that the bundle has seen during its stay in the core, the latest refinement of the history-based method provides the capability of fission-product-drivers. It allows the bundle-specific concentrations of the three basic groups of saturating fission products to be calculated in steady state or following a power transient, including long shutdowns. The new capability is illustrated by simulating the startup period following a typical long-shutdown, starting from a snapshot in the Point Lepreau operating history. 9 refs., 7 tabs

  17. Continuous fission-product monitor system at Oyster Creek. Final report

    International Nuclear Information System (INIS)

    Collins, L.L.; Chulick, E.T.

    1980-10-01

    A continuous on-line fission product monitor has been installed at the Oyster Creek Nuclear Generating Station, Forked River, New Jersey. The on-line monitor is a minicomputer-controlled high-resolution gamma-ray spectrometer system. An intrinsic Ge detector scans a collimated sample line of coolant from one of the plant's recirculation loops. The minicomputer is a Nuclear Data 6620 system. Data were accumulated for the period from April 1979 through January 1980, the end of cycle 8 for the Oyster Creek plant. Accumulated spectra, an average of three a day, were stored on magnetic disk and subsequently analyzed for fisson products, Because of difficulties in measuring absolute detector efficiency, quantitative fission product concentrations in the coolant could not be determined. Data for iodine fission products are reported as a function of time. The data indicate the existence of fuel defects in the Oyster Creek core during cycle 8

  18. 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

  19. 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.

  20. Chemical aspects of fission product transport in the primary circuit of a light water reactor

    International Nuclear Information System (INIS)

    Bowsher, B.R.; Dickinson, S.; Nichols, A.L.; Ogden, J.S.; Potter, P.E.

    1985-01-01

    The transport and fission products in the primary circuit of a light water reactor are of fundamental importance in assessing the consequences of severe accidents. Recent experimental studies have concentrated upon the behaviour of simulant fission product species such as caesium iodide, caesium hydroxide and tellurium, in terms of their vapour deposition characteristics onto metals representative of primary circuit materials. An induction furnace has been used to generate high-density/structural materials aerosols for subsequent analysis, and similar equipment has been incorporated into a glove-box to study lightly-irradiated UO/sub 2/ clad in Zircaloy. Analytical techniques are being developed to assist in the identification of fission product chemical species released from the fuel at temperatures from 1000 to 2500 0 C. Matrix isolation-infrared spectroscopy has been used to identify species in the vapour phase, and specific data using this technique are reported

  1. Chemical aspects of fission product transport in the primary circuit of a light water reactor

    International Nuclear Information System (INIS)

    Bowsher, B.R.; Dickinson, S.; Nichols, A.L.; Ogden, J.S.; Potter, P.E.

    1985-01-01

    The transport and deposition of fission products in the primary circuit of a light water reactor are of fundamental importance in assessing the consequences of severe accidents. Recent experimental studies have concentrated upon the behavior of simulant fission product species such as cesium iodide, cesium hydroxide and tellurium, in terms of their vapor deposition characteristics onto metals representative of primary circuit materials. An induction furnace has been used to generate high density/structural materials aerosols for subsequent analysis, and similar equipment has been incorporated into a glove-box to study lightly-irradiated UO 2 clad in Zircaloy. Analytical techniques are being developed to assist in the identification of fission product chemical species released from the fuel at temperatures from 1000 to 2500 0 C. Matrix isolation-infrared spectroscopy has been used to identify species in the vapor phase, and specific data using this technique are reported

  2. 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)

  3. 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

  4. 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

  5. Status of recent fast capture cross section evaluations for important fission product nuclides

    International Nuclear Information System (INIS)

    Gruppelaar, H.

    1982-01-01

    A comparison is made between recent evaluations of fission-product cross sections as given in the CNEN/CEA, ENDF/B-IV, ENDF/V-V, JENDL-1, RCN-2 and RCN-3 data libraries. The intercomparison is restricted to 24 important fission products in a fast power reactor. The evaluation methods used to obtain the various data files are reviewed and possible shortcomings are indicated. A survey is given of the experimental data based used in the various evaluations. Some graphs are included showing the new ENDF/B-V and RCN-3 fastcapture cross-section evaluations. Further intercomparisons are made by means of multi-group and one-group cross sections. It is shown that lumped fission-product cross sections calculated from the most recent versions of the data files are in quite good agreement with each other. This review concludes with a discussion on observed discrepancies and requests for new measurements. 78 references

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  7. 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)

  8. Immobilization of fission products arising from pyrometallurgical reprocessing in chloride media

    Science.gov (United States)

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

    2005-12-01

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

  9. Comparison of actinides and fission products recycling scheme with the normal plutonium recycling scheme in fast reactors

    Directory of Open Access Journals (Sweden)

    Salahuddin Asif

    2013-01-01

    Full Text Available Multiple recycling of actinides and non-volatile fission products in fast reactors through the dry re-fabrication/reprocessing atomics international reduction oxidation process has been studied as a possible way to reduce the long-term potential hazard of nuclear waste compared to that resulting from reprocessing in a wet PUREX process. Calculations have been made to compare the actinides and fission products recycling scheme with the normal plutonium recycling scheme in a fast reactor. For this purpose, the Karlsruhe version of isotope generation and depletion code, KORIGEN, has been modified accordingly. An entirely novel fission product yields library for fast reactors has been created which has replaced the old KORIGEN fission products library. For the purposes of this study, the standard 26 groups data set, KFKINR, developed at Forschungszentrum Karlsruhe, Germany, has been extended by the addition of the cross-sections of 13 important actinides and 68 most important fission products. It has been confirmed that these 68 fission products constitute about 95% of the total fission products yield and about 99.5% of the total absorption due to fission products in fast reactors. The amount of fissile material required to guarantee the criticality of the reactor during recycling schemes has also been investigated. Cumulative high active waste per ton of initial heavy metal is also calculated. Results show that the recycling of actinides and fission products in fast reactors through the atomics international reduction oxidation process results in a reduction of the potential hazard of radioactive waste.

  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. 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

  12. Fission Product Transport Models Adopted in REFPAC Code for LOCA Conditions in PWR and WWER NPPS

    International Nuclear Information System (INIS)

    Strupczewski, A.

    2003-01-01

    The report presents assumptions and physical models used for calculations of fission product releases from nuclear reactors, their behavior inside the containment and leakages to the environment after large break loss of coolant accident LB LOCA. They are the basis of code REFPAC (RElease of Fission Products under Accident Conditions), designed primarily to represent significant physical processes occurring after LB LOCA. The code describes these processes using three different models. Model 1 corresponds to established US and Russian practice, Model 2 includes all conservative assumptions that are in agreement with the actual state-of-the-art, and Model 3 incorporates formulae and parameter values actually used in EU practice. (author)

  13. Fission product inventory calculation by a CASMO/ORIGEN coupling program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)

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

    International Nuclear Information System (INIS)

    Bailly, H.W.

    1988-01-01

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

  15. FISP5 - an extended and improved version of the fission product inventory code FISP

    International Nuclear Information System (INIS)

    Tobias, A.

    1978-05-01

    In order to accommodate the UKFPDD-1 fission product data library the CEGB fission product inventory code FISP4 has been modified and extended. The opportunity was taken to revise the algorithm used for calculating the nuclide concentrations during irradiation in order to reduce the problem of rounding errors which arise as a result of the computer limitation to a finite word length. The resulting code FISP5 is shown in addition to offer considerable improvement in execution time in comparison with FISP4. Details of the revised algorithm are given together with a brief users' guide to FISP5. (author)

  16. 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

  17. Fission product inventory calculation by a CASMO/ORIGEN coupling program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)

  18. Transmutation of fission products in reactors and accelerator-driven systems

    International Nuclear Information System (INIS)

    Janssen, A.J.

    1994-01-01

    Energy flows and mass flows in several scenarios are considered. Economical and safety aspects of the transmutation scenarios are compared. It is difficult to find a sound motivation for the transmutation of fission products with accelerator-driven systems. If there would be any hesitation in transmuting fission products in nuclear reactors, there would be an even stronger hesitation to use accelerator-driven systems, mainly because of their lower energy efficiency and their poor cost effectiveness. The use of accelerator-driven systems could become a 'meaningful' option only if nuclear energy would be banished completely. (orig./HP)

  19. Enrichment of fission products in ionic salt bath by countercurrent electromigration

    International Nuclear Information System (INIS)

    Matsuura, Haruaki; Takagi, Ryuzo; Okada, Isao; Fujita, Reiko.

    1997-01-01

    We have proposed to apply a countercurrent electromigration method to enrichment of fission products in ionic melts. In the test runs, for this purpose, we have enriched Cs, Sr and Gd from their dilute melts. All of Cs, Sr and Gd were much concentrated at the area near the anode in the migration tubes. Gd and Sr were more concentrated than Cs. It was found that the electromigration method can be applied to the salt bath refleshing process after an electrorefining process, which removes fission products of multivalent cations. (author)

  20. An assessment of fission product data for decay power calculation in fast reactors

    International Nuclear Information System (INIS)

    Sridharan, M.S.; Murthy, K.P.N.

    1987-01-01

    A review of our present capability at IGC, Kalpakkam to predict fission product decay power in fast reactors is presented. This is accomplished by comparing our summation calculations with the calculations of others and the reported experimental measurements. Our calculations are based on Chandy code developed at our Centre. The fission product data base of Chandy is essentially drawn from the yield data compiled by Crouch (1977) and the data on halflives etc. compiled by Tobias (1973). In general, we find good agreement amongst the different calculations (within ±5%) and our calculations also compare well with experimental measurements of AKIAMA et al and MURPHY et al

  1. Calculated leaching of certain fission products from a cylinder of French glass

    International Nuclear Information System (INIS)

    Blomqvist, G.

    1977-07-01

    The probable total leaching of the most important fission products and actinides have been tabulated for a cylinder of French HLW glass with approximately 9 percent fission products. The calculations cover the period between 30 and 10000 years after removal from the reactor. The cylinder is of the type planned for the introduction of the HLW into Swedish crystalline rocks. All the components are supposed to have the same leach rate. The calculations also include the probable thickness of eroded glass layer/year. (author)

  2. Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

  3. Direct and preequilibrium effects in the fission-product mass range

    International Nuclear Information System (INIS)

    Gruppelaar, H.; Hogenbirk, A.

    1992-07-01

    Until recently inelastic scattering did not gain the proper attention in fission-product cross section evaluations. In many existing evaluations global spherical optical models have been used, neglecting direct and pre-equilibrium effects. There are also few experimental data relevant to inelastic scattering in fission products. This paper is focussed on the anomalously high inelastic scattering cross sections observed in even-mass nuclei near mass A=100 at low energies. Both more data and more refined theoretical analyses are required. A number of suggestions for relevant coupled-channel calculations is made. (author). 29 refs., 5 figs., 1 tab

  4. 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.

  5. Mass resolved angular distribution of fission products in 20Ne + 232Th reaction

    International Nuclear Information System (INIS)

    Tripathi, R.; Sodaye, S.; Sudarshan, K.; Kumar, Amit; Guin, R.

    2011-01-01

    Mass resolved angular distribution of fission products was measured in 20 Ne + 232 Th reaction at beam energy of 120 MeV. A preliminary analysis of the angular distribution data of fission products shows higher average anisotropy compared to that calculated using statistical theory. A signature of rise in anisotropy near symmetry, as reported in earlier studies in literature, is also seen. Further study is in progress to get more detailed information about the contribution from non-compound nucleus fission and dependence of angular anisotropy on asymmetry of mass division

  6. 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

  7. Preparation of lumped fission product (FP) cross sections for a multigroup library

    International Nuclear Information System (INIS)

    Ono, S.; Corcuera, R.P.

    1984-01-01

    A method for the calculation of lumped Fission Product (FP) cross sections has been developed. The group constants fo each nuclide are generated by NJOY code, based on ENDF/B-V data. In this first version, cross section of 28 nuclides are lumped for typical characteristics of Binary Breeder Reactor (BBR). One energy group calculations are made for a 1000 MWe fast reactor to verify the influence of burnup, number of FP and fuel composition on the lumped fission product cross sections. (Author) [pt

  8. Contribution to the study of the behaviour, in the urban environment, during the runoff of rainwater, of the fission products emitted during a nuclear accident

    International Nuclear Information System (INIS)

    Pioch, M.

    1993-01-01

    In the context of research into the environmental consequences of a serious accident occurring on a pressurized water reactor, this paper concerns the experimental study of behaviour of five fission products (caesium, strontium, iodine, ruthenium and tellurium) in the urban environment under the action of rainwater. Stable or radioactive multiple-element aerosols were produced. Their physicochemical characteristics and their solubility in rainwater were studied. Caesium and rubidium forms solutions totally and quickly, while strontium is partially soluble (approximately 50 %) and iodine is only slightly soluble. The behaviour of fission products on five urban surfaces was then studied. Batch experiments showed that the retention of dissolved forms of radioelements varied according to the material. The reactions involved are ion exchange reactions. The presence of certain ions in water (in particular NH 4 + ) increase the desorption of radioelements. Using a laboratory rainfall simulator, the re-entrainment of fission products by rainwater was examined. Two modes of deposition and two intensities of rainfall were simulated. The desorption of radioelements is greater after wet deposition and remobilization is reduced by an increase in intensity of rainfall. An addition of NH 4 + in water is especially effective in the case of wet depositions. Suggestions are made in order to improve experimental protocols and continue the research. (author). 75 refs., 51 figs., 69 tabs., 14 appends

  9. Development and validation of the aerosol transport module gamma-FP for evaluating radioactive fission product source terms in a VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Churl; Lim, Hong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-12-15

    Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model, and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data, which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.

  10. Ranking multiple docking solutions based on the conservation of inter-residue contacts

    KAUST Repository

    Oliva, Romina M.; Vangone, Anna; Cavallo, Luigi

    2013-01-01

    ) conformations in the top positions is still an open problem. Herein we present CONSRANK, a simple and effective tool to rank multiple docking solutions, which relies on the conservation of inter-residue contacts in the analyzed decoys ensemble. First

  11. 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)

  12. The LOMOsup(R) process: a solution for residual monomers

    International Nuclear Information System (INIS)

    Derbyshire, R.L.

    1979-01-01

    Regulatory activity over the last several years has addressed the potential problems associated with the migration of residual monomers from a number of commodity food packages. Regardless of the outcome of current debates, it will always be desirable to reduce monomer levels to as low a level as economically practicable so that they do not become indirect additives. The LOMO process is a body of technology inclusive of an ionizing radiation treatment which can result in sharp reduction of residual monomer levels in commodity plastic resins. The process may be applicable to factory intermediates, raw resins, or finished articles. Depending upon the individual system and its monomers, LOMO treatment can result in reductions to levels which press today's analytical test capability. Industrial radiation processing is normally accomplished with electron beam accelerators. Electron beam processing continues to gain in understanding and acceptance as one of the very few basic methods by which energy can be imparted to an industrial process system. Typically, whole factories are constructed around one accelerator. (author)

  13. 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

  14. Combining extractant systems for the simultaneous extraction of transuranic elements and selected fission products

    International Nuclear Information System (INIS)

    Horwitz, E.P.

    1993-01-01

    The popularity of solvent extraction (SX) stems from its ability to operate in a continuous mode, to achieve high throughputs and high decontamination factors of product streams, and to utilize relatively small quantities of very selective chemical compounds as metal ion complexants. The chemical pretreatment of nuclear waste for the purpose of waste minimization will probably utilize one or more SX processes. Because of the diversity and complexity of nuclear waste, perhaps the greatest difficulty for the separation chemist is to develop processes that remove not only actinides but also selected fission products in a single process. A stand alone acid-side SX process (TRUEX) for removal of uranium and transuranic elements (Np, Pu, Am) from nuclear waste has been widely reported. Recently, an acid-side SX process (SREX) to extract and recover 90 Sr from high-level nuclear waste has also been reported. Both the TRUEX and SREX processes extract Tc to a significant extent although not as efficiently as they extract transuranics and Sr. Ideally one would like to have a process that can extract and recover all actinides as well as 99 Tc, 90 Sr, and 137 Cs. A possible solution to multielement extraction is to mix two extractants with totally different properties into a single process solvent formulation. For this approach to be successful, both extractants must be essentially the same type, either neutral, liquid cationic, or liquid anionic. Experimental work has been carried out on mixed TRUEX and SREX processes, for synthetically created waste, and demonstrates the combined solvent formulation is effective at extracting both the actinides and Tc, as well as Sr. There is no evidence for the presence of either synergistic or antagonistic effects between the two extractants. This demonstates the feasibility of at least part of a combined solvent extraction scheme

  15. Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

    International Nuclear Information System (INIS)

    Wagner, J.C.

    2001-01-01

    This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses

  16. 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

  17. Influence of solvent radiolysis on extraction, scrubbing and stripping of uranium and some fission products

    International Nuclear Information System (INIS)

    Gawlowska, W.; Nowak, M.

    1978-01-01

    Radiolytically degraded TBP-n-paraffins solvent was used in the laboratory flow-sheet to study the influence of radiation exposure on decontamination of uranium. The influence of accumulated doses on extraction, scrubbing and stripping of uranium and some fission products has been discussed. (author)

  18. Release of fission products during controlled loss-of-coolant accidents and hypothetical core meltdown accidents

    International Nuclear Information System (INIS)

    Albrecht, H.; Malinauskas, A.P.

    1978-01-01

    A few years ago the Projekt Nukleare Sicherheit joined the United States Nuclear Regulatory Commission in the development of a research program which was designed to investigate fission product release from light water reactor fuel under conditions ranging from spent fuel shipping cask accidents to core meltdown accidents. Three laboratories have been involved in this cooperative effort. At Argonne National Laboratory (ANL), the research effort has focused on noble gas fission product release, whereas at Oak Ridge National Laboratory (ORNL) and at Kernforschungszentrum Karlsruhe (KfK), the studies have emphasized the release of species other than the noble gases. In addition, the ORNL program has been directed toward the development of fission product source terms applicable to analyses of spent fuel shipping cask accidents and controlled loss-of-coolant accidents, and the KfK program has been aimed at providing similar source terms which are characteristic of core meltdown accidents. The ORNL results are presented for fission product release from defected fuel rods into a steam atmosphere over the temperature range 500 to 1200 0 C, and the KfK results for release during core meltdown sequences

  19. Thermodynamics of soluble fission products cesium and iodine in the Molten Salt Reactor

    NARCIS (Netherlands)

    Capelli, E.; Beneš, O.; Konings, R.J.M.

    2018-01-01

    The present study describes the full thermodynamic assessment of the Li,Cs,Th//F,I system. The existing database for the relevant fluoride salts considered as fuel for the Molten Salt Reactor (MSR) has been extended with two key fission products, cesium and iodine. A complete evaluation of all

  20. Spread of fission products after a nuclear melt-down accident

    International Nuclear Information System (INIS)

    Andriesse, C.D.; Tanke, R.H.J.

    1987-01-01

    In this article experimental investigation is described into the spread of fission products within a nuclear power plant, which after an accident involving melting of the nucleus, will be possible in spite of prohibiting constructions for the case of severe unbalancing of generated and carried-off energy. 6 refs.; 4 figs

  1. 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)

  2. Thermochemical data for reactor materials and fission products: The ECN database

    International Nuclear Information System (INIS)

    Cordfunke, E.H.P.; Konings, R.J.M.

    1993-02-01

    The activities of the authors regarding the compilation of a database of thermochemical properties for reactor materials and fission products is reviewed. The evaluation procedures and techniques are outlined and examples are given. In addition, examples of the use of thermochemical data for the application in the field of Nuclear Technology are given. (orig.)

  3. Nuclear data project in Korea and resonance parameter evaluation of fission products

    International Nuclear Information System (INIS)

    Chang, Jonghwa; Oh, Soo-Youl

    2000-01-01

    Nuclear data activities in the fields of evaluation, processing, measurement, and service in Korea are presented in this paper. As one of the current activities, the neutron resonance parameters for stable or long-lived nineteen fission products have been evaluated and the results are presented here. (author)

  4. Report on the Behavior of Fission Products in the Co-decontamination Process

    International Nuclear Information System (INIS)

    Martin, Leigh Robert; Riddle, Catherine Lynn

    2015-01-01

    This document was prepared to meet FCT level 3 milestone M3FT-15IN0302042, 'Generate Zr, Ru, Mo and Tc data for the Co-decontamination Process.' This work was carried out under the auspices of the Lab-Scale Testing of Reference Processes FCT work package. This document reports preliminary work in identifying the behavior of important fission products in a Co-decontamination flowsheet. Current results show that Tc, in the presence of Zr alone, does not behave as the Argonne Model for Universal Solvent Extraction (AMUSE) code would predict. The Tc distribution is reproducibly lower than predicted, with Zr distributions remaining close to the AMUSE code prediction. In addition, it appears there may be an intricate relationship between multiple fission product metals, in different combinations, that will have a direct impact on U, Tc and other important fission products such as Zr, Mo, and Rh. More extensive testing is required to adequately predict flowsheet behavior for these variances within the fission products.

  5. Fission product behavior in the Peach Bottom and Fort St. Vrain HTGRs

    International Nuclear Information System (INIS)

    Hanson, D.L.; Baldwin, N.L.; Strong, D.E.

    1980-11-01

    Actual operating data from Peach Bottom and Fort St. Vrain were compared with code predictions to assess the validity of the methods used to predict the behavior of fission products in the primary coolant circuit. For both reactors the measured circuit activities were significantly below design values, and the observations generally verify the codes used for large HTGR design

  6. On-line nuclear half life and spectroscopic measurements on mass-separated fission product nuclei

    International Nuclear Information System (INIS)

    McDonald, J.; Fogelberg, B.; Baecklin, A.

    1979-01-01

    A description is given of the methods and equipment employed for nuclear spectroscopy studies of short lived fission product nuclei at the OSIRIS ISOL facility in Studsvik, Sweden. Furthermore a table of new nuclear half-lives measured with this equipment is presented. (author)

  7. Calculating the mass distribution of heavy nucleus fission product by neutrons

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-09-01

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

  9. 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.

  10. Release and transport of fission product cesium in the TMI-2 accident

    International Nuclear Information System (INIS)

    Lorenz, R.A.; Collins, J.L.

    1986-01-01

    Approximately 50% of the fission product cesium was released from the overheated UO 2 fuel in the TMI-2 accident. Steam that boiled away from a water pool in the bottom of the reactor vessel transported the released fission products throughout the reactor coolant system (RCS). Some fission products passed directly through a leaking valve with steam and water into the containment structure, but most deposited on dry surfaces inside of the RCS before being dissolved or resuspended when the RCS was refilled with water. A cesium transport model was developed that extended measured cesium in the RCS back to the first day of the accident. The model revealed that ∼62% of the released 137 Cs deposited on dry surfaces inside of the RCS before being slowly leached and transported out of the RCS in leaked or letdown water. The leach rates from the model agreed reasonably well with those measured in the laboratory. The chemical behavior of cesium in the TMI-2 accident agreed with that observed in fission product release tests at Oak Ridge National Laboratory (ORNL)

  11. 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

  12. The Kinetics of Fission Products Release from Microfuel Taking into Account the Trapped Fraction and Limited Solubility Effects

    International Nuclear Information System (INIS)

    Ivanov, A.S.; Rusinkevich, A.A.

    2014-01-01

    In this paper the effect of the oxygen getter on fission products release from the coated particle was studied by the “FP Kinetics” code. Trapped fraction and limited solubility effects taken into consideration. It was shown that these effects have a significant impact on the concentration profile and integral release of fission products. (author)

  13. Transmutation analysis considering and explicit fission product treatment based on a coupled Hammer-Technion and Cinder-2 system

    International Nuclear Information System (INIS)

    Abe, A.Y.

    1989-01-01

    This work presents a study about neutron absorption in a typical PWR cell by considering an explicit treatment for the fission products. The proposed methodology to treat fission product neutron absorption in a lattice calculation combines the HAMMER-TECHNION and CINDER-2 codes. The fission product chain treatment considers nearly 99% of all original CINDER-2 neutron absorption chain treatment. Parallel to the explicit treatment, a cross section library in the HAMMER-TECHNION code multigroup structure for the fission products was generated using the ENDF/B-V fission product library and processed by NJOY and AMPX-II processing codes. The methodology validation was investigated against two available benchmarks and it was obtained excellent results for the K-Infinity (IAEA-TECDOC-233) as function of burnup and enrichment and for the aggregate quantity sup(σ)2200 in units of barns/fission cross sections (OKAZAKI and SOKOLOWSKI). This work contributed for a better understanding of the fission product neutron absorption in a typical PWR cell and showed that the explicit fission product treatment can be successfully achieved. Besides that the performance of the ENDF/B-V fission product library was accessed. (author)

  14. Corrosion mechanisms of containment glasses for fission products

    International Nuclear Information System (INIS)

    Nogues, J.L.

    1984-01-01

    After a review of nuclear energy production and waste vitrification principles, the aqueous corrosion mechanisms of the containment glasses and the various parameters affecting the corrosion are studied: effects of glass composition, temperature, lixiviation agent pH, lixiviation duration and mode. Conventional mass loss measurement and solution analyses are coupled to sophisticated surface analysis techniques. The hydrolyzed layer formation and the solubility limits are discussed. 87 figs., 30 tabs., 144 refs

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

    International Nuclear Information System (INIS)

    Holub, F.

    1987-08-01

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

  16. Study on mitigation of in-vessel release of fission products in severe accidents of PWR

    International Nuclear Information System (INIS)

    Huang, G.F.; Tong, L.L.; Li, J.X.; Cao, X.W.

    2010-01-01

    Research highlights: → In-vessel release of fission products in severe accidents for 600 MW PWR is analyzed. → Mitigation effect of primary feed-and-bleed on in-vessel release is investigated. → Mitigation effect of secondary feed-and-bleed on in-vessel release is studied. → Mitigation effect of ex-vessel cooling on in-vessel release is evaluated. - Abstract: During the severe accidents in a nuclear power plant, large amounts of fission products release with accident progression, including in-vessel and ex-vessel release. Mitigation of fission products release is demanded for alleviating radiological consequence in severe accidents. Mitigation countermeasures to in-vessel release are studied for Chinese 600 MW pressurized water reactor (PWR), including feed-and-bleed in primary circuit, feed-and-bleed in secondary circuit and ex-vessel cooling. SBO, LOFW, SBLOCA and LBLOCA are selected as typical severe accident sequences. Based on the evaluation of in-vessel release with different startup time of countermeasure, and the coupling relationship between thermohydraulics and in-vessel release of fission products, some results are achieved. Feed-and-bleed in primary circuit is an effective countermeasure to mitigate in-vessel release of fission products, and earlier startup time of countermeasure is more feasible. Feed-and-bleed in secondary circuit is also an effective countermeasure to mitigate in-vessel release for most severe accident sequences that can cease core melt progression, e.g. SBO, LOFW and SBLOCA. Ex-vessel cooling has no mitigation effect on in-vessel release owing to inevitable core melt and relocation.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  18. 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

  19. An evaluation of nodalization/decay heat/ volatile fission product release models in ISAAC code

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yong Mann; Park, Soo Yong; Kim, Dong Ha

    2003-03-01

    An ISAAC computer code, which was developed for a Level-2 PSA during 1995, has developed mainly with fundamental models for CANDU-specific severe accident progression and also the accident-analyzing experiences are limited to Level-2 PSA purposes. Hence the system nodalization model, decay model and volatile fission product release model, which are known to affect fission product behavior directly or indirectly, are evaluated to both enhance understanding for basic models and accumulate accident-analyzing experiences. As a research strategy, sensitivity studies of model parameters and sensitivity coefficients are performed. According to the results from core nodalization sensitivity study, an original 3x3 nodalization (per loop) method which groups horizontal fuel channels into 12 representative channels, is evaluated to be sufficient for an optimal scheme because detailed nodalization methods have no large effect on fuel thermal-hydraulic behavior, total accident progression and fission product behavior. As ANSI/ANS standard model for decay heat prediction after reactor trip has no needs for further model evaluation due to both wide application on accident analysis codes and good comparison results with the ORIGEN code, ISAAC calculational results of decay heat are used as they are. In addition, fission product revaporization in a containment which is caused by the embedded decay heat, is demonstrated. The results for the volatile fission product release model are analyzed. In case of early release, the IDCOR model with an in-vessel Te release option shows the most conservative results and for the late release case, NUREG-0772 model shows the most conservative results. Considering both early and late release, the IDCOR model with an in-vessel Te bound option shows mitigated conservative results.

  20. Irradiation effects and behaviour of fission products in zirconia and spinel

    International Nuclear Information System (INIS)

    Gentils, A.

    2003-10-01

    Crystalline oxides, such as zirconia (ZrO 2 ) and spinel (MgAl 2 O 4 ), are promising inert matrices for the transmutation of plutonium and minor actinides. This work deals with the study of the physico-chemical properties of these matrices, more specifically their behaviour under irradiation and their capacity to retain fission products. Irradiations at low energy and incorporation of stable analogs of fission products (Cs, I, Xe) into yttria-stabilized zirconia and magnesium-aluminate spinel single crystals were performed by using the ion implanter IRMA (CSNSM-Orsay). Irradiations at high energy were made on several heavy ion accelerators (GANIL-Caen, ISL-Berlin, HIL-Warsaw). The damage induced by irradiation and the release of fission products were monitored by in situ Rutherford Backscattering Spectrometry experiments. Transmission electron microscopy was also used in order to determine the nature of the damage induced by irradiation. The results show that irradiation of ZrO 2 and MgAl 2 O 4 with heavy ions (about hundred keV and about hundred MeV) induces a huge structural damage in crystalline matrices. Total disorder (amorphization) is however never reached in zirconia, contrary to what is observed in the case of spinel. The results also emphasize the essential role played by the concentration of implanted species on their retention capacity. A dramatic release of fission products was observed when the concentration exceeds a threshold of a few atomic percent. Irradiation of implanted samples with medium-energy noble-gas ions leads to an enhancement of the fission product release. The exfoliation of spinel crystals implanted at high concentration of Cs ions is observed after a thermal treatment at high temperature. (author)

  1. A Study on Fission Product Model Comparison between MAAP4 and MAAP5

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Tae-young; Seo, Mi Ro [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The newly added safety goal required that the sum of the accident frequency that the release of the radioactive nuclide Cs-137 to environment exceeds the 100TBq should be less than 1.0E-6/RY. This requirement is known to be come from the provision for preventing the long term ground contamination due to the release of radioactive material. Validation of this standard was performed by many researchers recently. In the outlook of Cs-137, the mass of Cs-137 correspondent with the 100TBq is calculated as 32g. However, during the severe accident, if the containment has been failed, it is generally expected that the mass of Cs-137 released to the environment is more than 1kg for most accident sequences. The purpose of this study compare fission product model in MAAP4 and MAAP5. So the same accident will be simulated as MAAP4 and MAAP5. And will compare fission product release fraction. This will help to improvements obtained to meet the regulatory requirements of Cs-137. This paper was a comparison of MAAP4's fission product models with those of MAAP5. And this paper simulated the station blackout accident to compare MAAP4 and MAAP5 fission product release fraction. So far Level 2 PSA analysis used MAAP4. And this result failed to meet the regulatory requirements of Cs-137 up to now. Fission product release fraction calculated by MAAP5 is more conservative than that calculated by MAAP4. Therefore, using MAAP5 is more difficult to meet the requirements of Cs-137. Thus, Level 1 PSA analysis must find ways to reduce CDF and Level 2 PSA analysis must find ways to reduce CFF in order to meet regulatory requirements. Not only, it seems to be required a study on the possible safety systems to alleviate the containment failure after the core damage.

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

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

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

  3. Potential for large-scale uses for fission-product Xenon

    International Nuclear Information System (INIS)

    Rohrmann, C.A.

    1983-03-01

    Of all fission products in spent, low-enrichment-uranium power-reactor fuels, xenon is produced in the highest yield - nearly one cubic meter, STP, per metric ton. In aged fuels which may be considered for processing in the US, radioactive xenon isotopes approach the lowest limits of detection. The separation from accompanying radioactive 85 Kr is the essential problem; however, this is state-of-the-art technology which has been demonstrated on the pilot scale to yield xenon with pico-curie levels of 85 Kr contamination. If needed for special applications, such levels could be further reduced. Environmental considerations require the isolation of essentially all fission-product krypton during fuel processing. Economic restraints assure that the bulk of this krypton will need to be separated from the much-more-voluminous xenon fraction of the total amount of fission gas. Xenon may thus be discarded or made available for uses at probably very low cost. In contrast with many other fission products which have unique radioactive characteristics which make them useful as sources of heat, gamma and x-rays, and luminescence - as well as for medicinal diagnostics and therapeutics - fission-product xenon differs from naturally occurring xenon only in its isotopic composition which gives it a slightly hgiher atomic weight, because of the much higher concentrations of the 134 Xe and 136 Xe isotopes. Therefore, fission-product xenon can most likely find uses in applications which already exist but which can not be exploited most beneficially because of the high cost and scarcity of natural xenon. Unique uses would probably include applications in improved incandescent light illumination in place of krypton and in human anesthesia

  4. Substituted polyethylene glycols in extraction of fission products and actinoids

    International Nuclear Information System (INIS)

    Smirnov, I.V.; Efremova, T.I.; Shadrin, A.Yu.; Baulin, V.E.; Tsvetkov, E.N.

    1993-01-01

    Europium and americium extraction from nitric acid media by solutions of phosphoryl-containing podands in o-nitroethylbenzene was studied. Extractive capacity of podands is determined by the length of polyether chain and the structure of terminal groups. Podand Ph 2 P(O)CH 2 CH 2 OCH 2 xCH 2 P(O)PH 2 as regards its extractive, capacity is similar to carbamoylphosphineoxides. Introduction of rigid o-phenylene fragments to polyether chain brigs, about a sharp decrease in extractive capacity of the podands

  5. 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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

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

    International Nuclear Information System (INIS)

    Benedict, U.

    1977-01-01

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

  9. Formation and characterization of fission-product aerosols under postulated HTGR accident conditions

    International Nuclear Information System (INIS)

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

    1982-07-01

    The paper presents the results of an experimental investigation on the formation mechanism and physical characterization of simulated nuclear aerosols that could likely be released during an HTGR core heat-up accident. Experiments were carried out in a high-temperature flow system consisting essentially of an inductively heated release source, a vapor deposition tube, and a filter assembly for collecting particulate matter. Simulated fission products Sr and Ba as oxides are separately impregnated in H451 graphite wafers and released at elevated temperatures into a dry helium flow. In the presence of graphite, the oxides are quantitatively reduced to metals, which subsequently vaporize at temperatures much lower than required for the oxides alone to vaporize in the absence of graphite. A substantial fraction of the released material is associated with particulate matter, which is collected on filters located downstream at ambient temperature. The release and transport of simulated fission product Ag as metal are also investigated

  10. Fission product thermodynamic data: technical progress report 1 January-30 June 1991

    International Nuclear Information System (INIS)

    Ball, R.G.J.; Yates, A.D.; Bowsher, B.R.; Dickinson, S.; Freemantle, N.E.; Day, T.; Ogden, J.S.

    1991-07-01

    Thermodynamic data are being determined for a number of compounds formed from specific fission products and reactor materials. The compounds selected for experimental study and critical assessment were chosen because their thermodynamic data were inadequate or did not exist, as assessed and recommended at a specialists' meeting. These data can be used in the appropriate computer code so that the speciation and transport properties of the fission products can be predicted during a severe reactor accident. Experimental studies have focussed on the vaporization of tellurium dioxide, caesium ruthenate, strontium borate and indium hydroxide. Critical evaluations have begun for a number of tellurides of importance in severe accident assessments, and preliminary analyses have been made of the Fe-Te, Ni-Te and Cr-Te systems. (author)

  11. Thermophysical modeling of volatile fission product release from a debris pool

    International Nuclear Information System (INIS)

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

    1999-01-01

    A model is described for fission product release from the debris pool in the lower plenum of the reactor pressure vessel. In the pool, turbulent natural convection flow is formed due to homogeneous internal heat generation. Using the best-known correlations, heat transfer at the curved bottom and the top of the pool may be calculated. Volatile fission product gases in the pool nucleate and diffuse to bubbles. Both the homogeneous nucleation and heterogeneous nucleation are considered. The bubble nucleation, growth, coalescence and loss due to rise is modeled pursuant to bubble dynamics. If the pressure and temperature of the pool are very high, homogeneous nucleation that accounts for effect of decrease in the pool pressure can occur. The effect of the bubble-to-pool interfacial tension and the pool pressure on the nucleation rate is investigated in this work

  12. Mobility of gamma-emitting fission products in typical black and laterite soils

    International Nuclear Information System (INIS)

    D'Souza, T.J.; Mistry, K.B.

    1981-01-01

    Studies on the movement of the gamma-emitting fission products 106 Ru, 125 Sb, 137 Cs and 144 Ce, surface deposited on columns of the black and laterite soils revealed that the mobility of these nuclides in the two soils depended mainly on the ionic form of the nuclide, soil reaction and predominant clay mineral type in the soil. Results showed a rapid breakthrough of 106 Ru in the black soil and considerable migration to the lower layers in the laterite soil on leaching with rain water, or moderately high calcium or sodium irrigation waters. While near complete fixation of 144 Ce and 137 Cs in the top contaminated layer of both soils was observed, a slight movement of 125 Sb only in the black soil was noticed. Incorporation of organic matter had no effects on the movement of fission products in both soil types, whereas ethylenediaminetraacetic acid (EDTA) induced variable movement of the different radionuclides in the two soil types. (author)

  13. High-power proton linac for transmuting the long-lived fission products in nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, G.P.

    1991-01-01

    High power proton linacs are being considered at Los Alamos as drivers for high-flux spallation neutron sources that can be used to transmute the troublesome long-lived fission products in defense nuclear waste. The transmutation scheme being studied provides a high flux (> 10{sup 16}/cm{sup 2}{minus}s) of thermal neutrons, which efficiently converts fission products to stable or short-lived isotopes. A medium-energy proton linac with an average beam power of about 110 MW can burn the accumulated Tc99 and I129 inventory at the DOE's Hanford Site within 30 years. Preliminary concepts for this machine are described. 3 refs., 5 figs., 2 tabs.

  14. Fission product transport and behavior during two postulated loss of flow transients in the air

    International Nuclear Information System (INIS)

    Adams, J.P.; Carboneau, M.L.

    1991-01-01

    This document discusses fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradation, respectively) in the Advanced Test Reactor (ATR). These transients are designated ATR Transient LCPI5 (high-pressure) and LPP9 (low-pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be analyzed. A probabilistic risk assessment indicated that the probability of occurrence for these two transients is of the order of 10 -5 and 10 -7 per reactor year for LCP15 and LPP9, respectively

  15. The ASIND-MEPhI library of independent actinide fission product yields

    International Nuclear Information System (INIS)

    Bogomolova, E.S.; Grashin, A.F.; Efimenko, A.D.; Lukasevich, I.B.

    1997-01-01

    This data base of independent fission product yields has been set up at the Moscow Engineering Physics Institute on the basis of theoretical calculations within the framework of the super-nonequilibrium thermodynamic model. The database consists of independent yield sets for 1163 fission products in the wide range of fissile nuclides from thorium-229 to fermium-257 with excitation energies up to 20 MeV. The use of the theoretical model made it possible to raise the accuracy of prediction for poorly explored fission reactions. The number of yield sets is larger than in the ENDF/B. For example, photofission product yields are included in the ASIND-MEPhI database as virtual sets. (author). 14 refs, 17 figs, 2 tabs

  16. Fractionation of gamma-emitting fission products absorbed by red kidney beans (Phaseolus vulgaris L.)

    International Nuclear Information System (INIS)

    D'Souza, T.J.; Mistry, K.B.

    1980-01-01

    The gamma-emitting fission product nuclides 106 Ru, 125 Sb, 137 Cs and 144 Ce that accumulated in the edible pods of bean (Phaseolus vulgaris L.) plants grown in nutrient culture were subjected to chemical fractionation. The results indicated that the largest fraction of 106 Ru, 125 Sb and 144 Ce was associated with ionic forms including salts of organic acids, phosphates, carbonates and some protein-bound forms extracted with dilute mineral acids (acid fraction). The association of these radionuclides with lipids including lipophyllic pigments, free amino acids and amino sugars (ethanol fraction) was next in significance. The association of 137 Cs was, however, greater with the ethanol fraction than with the acid fraction. Considerably reduced amounts of the fission products were present in the pectates, proteins, polysaccharides and nucleic acids. (U.K.)

  17. Fission product nuclear data obtained by use of an on-line mass spectrometer

    International Nuclear Information System (INIS)

    Reeder, P.L.; Wright, J.F.; Anderl, R.A.

    1975-01-01

    A Spectrometer for On-Line Analysis of Radionuclides (SOLAR) has been installed at a 1 MW TRIGA reactor at Washington State University. Fission product ions from a combination target/ion source located within the thermal column are brought out to a 60 0 magnetic sector mass spectrometer. Surface ionization provides copious beams of Rb + and Cs + ions and less intense beams of Br - and I - ions with negligible contamination by other elements. About 40 fission product nuclides can thus be chemically and physically separated in times of less than 1 second. Past results on independent and cumulative fission yields along with measurements of half-lives of some very neutron-rich nuclides are presented. Current work on delayed-neutron emission probabilities and energy spectra of delayed neutrons from individual nuclides is described. (7 tables, 2 figures) (U.S.)

  18. Study of fission product γ spectra in the band 2-500 keV

    International Nuclear Information System (INIS)

    Rousseau, A.

    1959-01-01

    In the study of the γ spectrum of uranium fission products, particular attention has been given in this note to the part of the spectrum ranging between 0 and 500 keV after a given pile operating programme and the evolution of this spectrum with time after a pile shutdown has been followed. The study be related to the fission products which appear in the pile as a whole or on those produced in a uranium sample assumed to have been placed in the pile. The latter case has been envisaged here. The spectrum determination is based partly on theory and partly on experiment. The pile operating conditions are different in the two cases, which widens the range of validity of the spectra traced here. (author) [fr

  19. 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

  20. FITPULS: a code for obtaining analytic fits to aggregate fission-product decay-energy spectra

    International Nuclear Information System (INIS)

    LaBauve, R.J.; George, D.C.; England, T.R.

    1980-03-01

    The operation and input to the FITPULS code, recently updated to utilize interactive graphics, are described. The code is designed to retrieve data from a library containing aggregate fine-group spectra (150 energy groups) from fission products, collapse the data to few groups (up to 25), and fit the resulting spectra along the cooling time axis with a linear combination of exponential functions. Also given in this report are useful results for aggregate gamma and beta spectra from the decay of fission products released from 235 U irradiated with a pulse (10 -4 s irradiation time) of thermal neutrons. These fits are given in 22 energy groups that are the first 22 groups of the LASL 25-group decay-energy group structure, and the data are expressed both as MeV per fission second and particles per fission second; these pulse functions are readily folded into finite fission histories. 65 figures, 11 tables

  1. Curves and tables of neutron cross sections of fission product nuclei in JENDL-3

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Tsuneo [ed.

    1992-06-15

    Neutron cross sections of 172 nuclei in the fission product region stored in JENDL-3 are shown in graphs and tables. The evaluation work of these nuclei was made by the Fission Product Nuclear Data Working Group of the Japanese Nuclear Data Committee, in the neutron energy region from 10{sup {minus}5} eV to 20 MeV. Almost of the cross section data reproduced in graphs in this report. The cross section averaged over 38 energy intervals are listed in a table. Shown in order tables are thermal cross sections, resonance integrals, Maxwellian neutron flux average cross sections, fission spectrum average cross sections, 14-MeV cross sections, one group average cross sections in neutron flux of typical types of fission reactors and average cross sections in the 30-keV Maxwellian spectrum.

  2. 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.

  3. Burn-Up Determination by High Resolution Gamma Spectrometry: Fission Product Migration Studies

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R S; Blackadder, W H; Ronqvist, N

    1967-04-15

    The migration of solid fission products, in particular caesium and ruthenium, in high temperature oxide fuel can create a severe problem during the application of non-destructive burn-up methods employing gamma spectrometry, since caesium-137 is otherwise the most convenient long-lived burn-up monitor and ruthenium-106 can be used to distinguish between fissions in U-235 and Pu-239. As part of an experimental programme to develop burn-up methods, gamma scanning experiments have been performed on slices of irradiated UO{sub 2} pellets using a lithium-drifted germanium detector. The usefulness of the technique for migration studies has been demonstrated by comparing the fission product distribution curves across the specimen diameters with the microstructure of the specimens after polishing and etching.

  4. Diffusion modeling of fission product release during depressurized core conduction cooldown conditions

    International Nuclear Information System (INIS)

    Martin, R.C.

    1991-01-01

    A simple model for diffusion through the silicon carbide layer of TRISO particles is applied to the data for accident condition testing of fuel spheres for the High-Temperature Reactor program of the Federal Republic of Germany (FRG). Categorization of sphere release of 137 Cs based on fast neutron fluence permits predictions of release with an accuracy comparable to that of the US/FRG accident condition fuel performance model. Calculations are also performed for 85 Kr, 90 Sr, and 110m Ag. Diffusion of cesium through SiC suggests that models of fuel failure should consider fuel performance during repeated accident condition thermal cycling. Microstructural considerations in models of fission product release are discussed. The neutron-induced segregation of silicon within the SiC structure is postulated as a mechanism for enhanced fission product release during accident conditions. As oxygen-enhanced SiC decomposition mechanism is also discussed. (author). 12 refs, 11 figs, 2 tabs

  5. Influence of corium oxidation on fission product release from molten pool

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V., E-mail: bechta@sbor.spb.s [Alexandrov Scientific-Research Institute of Technology (NITI), Sosnovy Bor (Russian Federation); Krushinov, E.V.; Vitol, S.A.; Khabensky, V.B.; Kotova, S.Yu.; Sulatsky, A.A. [Alexandrov Scientific-Research Institute of Technology (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V.; Almyashev, V.I. [Grebenschikov Institute of Silicate Chemistry of the Russian Academy of Sciences (ISC RAS), St. Petersburg (Russian Federation); Ducros, G.; Journeau, C. [CEA, DEN, Cadarache, F-13108 St. Paul lez Durance (France); Bottomley, D. [Joint Research Centre Institut fuer Transurane (ITU), Karlsruhe (Germany); Clement, B. [Institut de Radioprotection et Surete Nucleaire (IRSN), St. Paul lez Durance (France); Herranz, L. [CIEMAT, Madrid (Spain); Guentay, S. [PSI, Wuerenlingen (Switzerland); Trambauer, K. [GRS, Muenchen (Germany); Auvinen, A. [VTT, Espoo (Finland); Bezlepkin, V.V. [SPbAEP, St. Petersburg (Russian Federation)

    2010-05-15

    Qualitative and quantitative determination of the release of low-volatile fission products and core materials from molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. The experiments carried out in a cold crucible with induction heating and RASPLAV test facility are described. The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidation kinetics, critical influence of melt surface temperature and oxidation index on the fission product release rate, aerosol particle composition and size distribution. The relevance of measured high release of Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimental data with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions from IVTANTHERMO and GEMINI/NUCLEA codes are made. Recommendations for further investigations are proposed following the major observations and discussions.

  6. Ex-vessel water-level and fission-product monitoring for LWR

    International Nuclear Information System (INIS)

    DeVolpi, A.; Markoff, D.

    1988-01-01

    Given that the need for direct measurement of reactor coolant inventory under operational or abnormal conditions remains unsatisfied, a high-energy gamma-ray detection system is described for ex-vessel monitoring. The system has been modeled to predict response in a PWR, and the model has been validated with a LOFT LOCA sequence. The apparatus, situated outside the pressure vessel, would give relative water level and density over the entire vessel height and distinguish differing levels in the downcomer and core. It would also have significant sensitivity after power shutdown because of high-energy gamma rays from photoneutron capture, the photoneutrons being the result of fission-product decay in the core. Fission-products released to the coolant and accumulated in the top of a PWR vessel would also be theoretically detectable

  7. The study of radiochemical separation methods on gaseous Fission product krypton-88

    International Nuclear Information System (INIS)

    Yang Zhihong; Zhang Shengdong; Yang Lei; Ding Youqian; Sun Hongqing; Ma Peng

    2012-01-01

    Half-life of krypton-88 is 2.84 hours, high fission yields and a relatively large gamma branching ratio is had. The gas is short-lived fission products in burnup measurements. Only New fission products can extract from extraction in gas of fissile irradiation target. But krypton-88 with krypton-85, krypton-87, xenon -135, and xenon-138 is coexisted together, thus radiochemical separation must quickly taken. selected the irradiation time is 1-2 hours and cooling time is best 2 hours for sample preparation, krypton and xenon were separated using activated carbon adsorption, the ratio of krypton and xenon were measured by gamma spectroscopy. Then according to the ratio of krypton-85 and xenon-125 count rate coefficient around separation were calculated yield of krypton and decontamination factor of xenon and the final the yield of krypton-85 is calculated. (authors)

  8. 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

  9. Rapid separation of individual rare-earth elements from fission products

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  10. Solidification of highly active fission products by a thermite reaction. Pt. 1

    International Nuclear Information System (INIS)

    Rudolph, G.; Hild, W.

    1976-07-01

    To solidify high-level fission products a process was developed according to which a high-melting ceramic product is obtained as a solidification matrix in a thermite reaction. With a constant content of fission product oxides reaction mixtures consisting of 35 to 55 wt.% of manganese dioxide, 24 to 32 wt.% of aluminum shot and 17 to 36 wt.% of sand give suitable products. In the thermite reactiom some components contained in the reactic mixture volatilize partly by evaporation (alkali oxides, manganese oxide, and others) and partly by the formation of volatile oxides having lower valencies (silicon and aluminum oxide). The smoke generated can be easily collected in filters made of glass wool fibers. (orig./HR) [de

  11. Evaluations of fission product capture cross sections for ENDF/B-V

    International Nuclear Information System (INIS)

    Schenter, R.E.; Johnson, D.L.; Mann, F.M.; Schmittroth, F.

    1979-01-01

    Capture cross section evaluations were made for the 36 most important fission product absorbers in a fast reactor system. These evaluations were obtained by use of a generalized least-squares approach with calculations being performed with the computer code FERRET. These results will provide the major revisions to the ENDF/B-IV Fission Product Cross Section File which will be released as part of ENDF/B-V. Input for the cross section adjustment calculations included both integral and differential experimental data results. The differential cross sections and their uncertainties were obtained from the CSIRS library. Integral measurement results came from CFRMF and STEK Assemblies 500, 1000, 2000, 3000, and 4000. Comparisons of these evaluations with recent capture measurements are presented. 14 figures

  12. Influence of corium oxidation on fission product release from molten pool

    International Nuclear Information System (INIS)

    Bechta, S.V.; Krushinov, E.V.; Vitol, S.A.

    2009-01-01

    Release of low-volatile fission products and core materials from molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. The experiments carried out in cold crucible with induction heating and RASPLAV test facility are described. The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidation kinetics, critical influence of melt surface temperature and oxidation index on the fission product release rate and aerosol particle composition. The relevance of measured high release of Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimental data with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions from IVTANTHERMO and GEMINI/NUCLEA are set. (author)

  13. Curves and tables of neutron cross sections of fission product nuclei in JENDL-3

    International Nuclear Information System (INIS)

    Nakagawa, Tsuneo

    1992-06-01

    Neutron cross sections of 172 nuclei in the fission product region stored in JENDL-3 are shown in graphs and tables. The evaluation work of these nuclei was made by the Fission Product Nuclear Data Working Group of the Japanese Nuclear Data Committee, in the neutron energy region from 10 -5 eV to 20 MeV. Almost all the cross section data are reproduced in graphs in this report. The cross section averaged over 38 energy intervals are listed in a table. Shown in other tables are thermal cross sections, resonance integrals, Maxwellian neutron flux average cross sections, fission spectrum average cross sections, 14-MeV cross sections, one group average cross sections in neutron flux of typical types of fission reactors and average cross sections in the 30-keV Maxwellian spectrum. (author)

  14. Development of assessment technology for hydrogen burn and fission product behavior in containment

    International Nuclear Information System (INIS)

    Kim, S. B.; Kim, J. T.; Ha, K. S.; Hong, S. W.; Song, Y. M.; Park, J. H.; Cho, Y. R.; Kang, H. S.

    2012-04-01

    Analysis tools for hydrogen burn was established to resolve the hydrogen issues in containment. To validate CFX commercial CFD(computational fluid dynamics) code, the hydrogen combustion experiments such as FLAME and ENACEFF for reactor containment were analyzed. And OpenFOAM hydrogen combustion code was developed and validated. Experiments for the flame propagation characteristics in IRWST and the run-up-distance for DDT(Deflagration to detonation transition) were performed and analytical model was evaluated to evaluation of the performance of hydrogen mitigation system, that is, PAR(Passive auto-catalistic re-combiner) To improvement of the fission product modelling in containment, separate analysis module for Iodine behavior and its application tool of K-IODIP (Korea IODIne Package) were developed. PHEBUS FPT-3 analysis was performed to validate MELCOR code. And also the characteristics of fission product behaviors in Future Reactors(GEN-IV) were compared

  15. RELOS.MOD2: a code system for the determination of instationary fission product releases from molten pools

    International Nuclear Information System (INIS)

    Kortz, Ch.; Koch, M.K.; Unger, H.; Funke, F.

    1999-01-01

    For the assessment of molten corium pool source terms, a mechanistic model has been developed to describe the transport of fission products from liquid corium pool surfaces into a colder gas atmosphere. Modelling is based on an approach for diffusive and convective transport processes coupled with thermochemical equilibrium considerations enabling detailed speciation analyses of the fission products released. Both have been implemented into the code system RELOS.MOD2. RELOS.MOD2 sensitivity calculations on possible effects of anticipated uncertainties in the thermo-chemical data on the fission product release predictions are presented. (author)

  16. Fission product chemistry and aerosol behaviour in the primary circuit of a pressurised water reactor under severe accident conditions

    International Nuclear Information System (INIS)

    Bowsher, B.R.

    1985-09-01

    Three key accident sequences are considered covering a representative range of different environments of pressure, flow, temperature history and degree of zircaloy oxidation, and their principle thermal hydraulic and physical characteristics affecting chemistry behaviour are identified. Inventories, chemical forms and timing of fission product release are summarized together with the major sources of structural materials and their release characteristics. Chemistry of each main fission product species is reviewed from available experimental and/or theoretical data. Studies modelling primary circuit fission product behaviour are reviewed. Requirements for further study are assessed. (UK)

  17. Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents

    International Nuclear Information System (INIS)

    Ellison, P.G.; Monson, P.R.; Mitchell, H.A.

    1990-01-01

    This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises

  18. The evaluated neutron cross sections and resonance integrals of fission products with Z = 57-62

    International Nuclear Information System (INIS)

    Fedorova, A.F.; Pisanko, Zh.I.; Novoselov, G.M.

    1976-01-01

    Neutron cross sections at a neutron velocity of V=2200 m/s, and resonance integrals for fission products with Z=57-71 are estimated. In obtaining the recommended values the results of the neutron cross sections and resonance integrals for elements used as references were normalized in accordance with the latest adjusted values. In the course of estimation, preference was given to the more accurate methods for obtaining the measured values and to the more recent investigations

  19. French practice for assessing the fission product releases from the containment during a PWR severe accident

    International Nuclear Information System (INIS)

    Duco, J.; Dufresne, J.; L'homme, A.

    1988-10-01

    French safety philosophy as concerns severe PWR accidents has already been outlined by the Director of CEA/IPSN in an article published in ''Nuclear Safety''. Therefore the present paper will focus on: a) the French reference source terms, as used for elaborating ultimate emergency procedures on PWRs and for emergency planning; b) the methods currently developed for more realistic assessments of the release of fission products during a severe accident

  20. Enhancement of ASTEC and COCOSYS regarding fission product release during MCCI

    Energy Technology Data Exchange (ETDEWEB)

    Agethen, Kathrin [Bochum Univ. (Germany). Reactor Simulation and Safety Group

    2016-10-15

    The focus in this paper is on the enhancement of the fission product release model during molten core concrete interaction in the severe accident analysis codes ASTEC and COCOSYS. After both codes are harmonised and the model interaction as well as the input parameters are adapted, extended model approaches are implemented. These lead to an improvement of the release rates for selected semi-volatile species validated against the ACE tests under ex-vessel conditions.