Identification of unknown nuclear fuel by impurities and physical parameters

Analytical parameters sufficient for identification of nuclear materialof unknown origin have been assayed for a number of examples of uranium dioxidefuel. These parameters comprise isotopic composition of uranium and the morphologyfeatures of a fuel pellet. However, in practice, it occurs that a sample ofan industrial material corresponds frequently to the specification limitsof several suppliers. In order to narrow further a range for potential sources,additional analytical parameters are necessary, such as fuel pellet impuritylevels and surface roughness. In this study we utilise results from the followingnuclear analytical techniques: thermal-ionization mass spectrometry, glowdischarge mass spectrometry and profilometry. By way of example, we demonstratehow these results serve to differentiate the sources of confiscated, vagabondnuclear materials.     

Some possibilities for the plutonium content determination in an irradiated nuclear fuel

This paper describes the results of the methods used for the plutonium content determination in an irradiated nuclear fuel from the first Czechoslovak atomic power station A1. The main attention was paid to the following methods: mass-spectrometric isotope dilution method, radiometric method and correlation dependence method based on the analysis of the burnt fuel. The principle and the accuracy of the individual methods are discussed.     

Non-destructive determination of nuclear fuel burn-up by measuring the gamma radiation of fission products in irradiated uranium oxide

The burn-up of²³⁵U was determined in two uranium oxide samples (0.713 and 89.9%²³⁵U in mixture) irradiated simultaneously with a cobalt monitor, from the amounts of⁹⁵Zr,¹⁰³Ru,¹³⁷Cs,¹⁴⁰Ba and¹⁴⁴Ce obtained by measuring the intensities of the corresponding gamma radiations. The samples were irradiated for 23 days, and the fission products were measured after cooling for 100 days, nondestructively, by means of a Ge(Li) spectrometer. The integrated neutron flux was determined by measuring the produced⁶⁰Co in the cobalt monitor. The burn-up in both samples was determined by measuring the intensity of eight gamma energies (0.5–1.6 MeV). The determined values are in good agreement. The standard deviation of the mean value ( ) is 5%. The atom per cent fission of²³⁵U in both samples, calculated according to , differs by 1%. The measured σ _f for²³⁵U is in good agreement with the data reported in the literature.     

Migration of fission product137Cs in mud near a nuclear reactor

Due to the mismanagement of nuclear waste as well as heat exchanger degradation for the primary coolant of the one megawatt nuclear research reactor, the fission product¹³⁷Cs has been leaking to the environment ever since 1969. In the past thirty years, the long-lived¹³⁷Cs was accumulated and eventually trapped in the mud of the discharge pond right in front of the waste storage and the reactor facility. The distribution of¹³⁷Cs in mud was measured and contour-mapped to reveal the migration of trace levels of¹³⁷Cs in a period of three decades.     

Two CdZnTe detector-equipped gamma-ray spectrometers for attribute measurements on irradiated nuclear fuel

Summary Some United States Department of Energy-owned spent fuel elements from foreign research reactors (FRRs) are presently being shipped from the reactor location to the US for storage at the Idaho National Engineering and Environmental Laboratory (INEEL). Two cadmium zinc telluride detector-based gamma-ray spectrometers have been developed to confirm the irradiation status of these fuels. One spectrometer is configured to operate underwater in the spent fuel pool of the shipping location, while the other is configured to interrogate elements on receipt in the dry transfer cell at the INEEL’s Interim Fuel Storage Facility (IFSF) Both units have been operationally tested at the INEEL.     

High-pressure liquid chromatography of irradiated nuclear fuel: Separation of neodymium for burn-up determination

Neodymium is separated from solutions of spent nuclear fuel by high-pressure liquid chromatography in methanol-nitric acid-water media using an anion-exchange column. Chromatograms obtained by monitoring at 280 nm, illustrate the difficulties especially with the fission product ruthenium in nuclear chemistry. Preseparation of the rare earths and trivalent actinides using a di(2-ethylhexyl)phosphoric acid/kieselguhr column is described.     

Identification of iodoorganic compounds in kerosene from nuclear fuel reprocessing

In kerosene samples from nuclear fuel reprocessing, iodoalkanes with chain-lengths from C₄ to C₁₃ have been identified. The kerosene samples were purified by means of solid-phase extraction. By this method other fission products like¹²⁵Sb and¹⁰⁶Ru were quantitatively removed from the solution. The only remaining radioactive nuclide was thus¹²⁹I. The iodoorganic compounds in the kerosene from the solvent were enriched from 6000 Bq/L to 100 000 Bq/L¹²⁹I by vacuum distillation. Chromatographic separation by HPLC, fractionation, and -measurement of the fractions showed that at least one polar and one nonpolar iodoorganic compound were present. Derivatisation of the iodoorganic compounds with, 1,4-diazabicyclo-2,2,2-octane to quatermary ammonium salts and²⁵²Cf plasma desorption mass spectrometry of the products revealed that the main iodoorganic constituents in the kerosene were iodobutane as polar and iodododecane as nonpolar compound in approximately equal concentrations.     

Management of carbon dioxide released from spent nuclear fuel through voloxidation

Journal of Radioanalytical and Nuclear Chemistry - The management of radioactive carbon (C-14) from spent nuclear fuel (SNF) in a voloxidation process is vital to prevent radioactive contamination...     

Determination of silver fission product in high-temperature nuclear reactor fuels by ion-exchange separation and γ-counting

The content of silver fission product and its radial distribution in spherical coated fuel particles can be measured by stepwise chemical removal of pyrolytic carbon layers coupled with an anion-exchange separation procedure and subsequent γ-counting. An adsorption efficiency better than 98% ranging over five orders of magnitude in silver concentration was found for a bromide-containing medium. Silver to carbon ratios down to 10^-12 are measurable so that this method is useful for studying the diffusion of silver in pyrolytic carbon.     

Determination of the balance of the nuclear reactor fuel burn-out process by gamma-spectrometry of fission products III

The possibility for the determination of heavy water reactor fuel burn-up on the basis of gamma-spectrometric measurements of the activity quotients¹⁰⁶Ru/¹³⁷Cs and¹³⁴Cs/¹³⁷Cs has been experimentally investigated. The investigation has been carried out on the non-enriched uranium metal fuel of the Czechoslovak Nuclear Power Plant Al. A spectrometer with germanium detector has been used for spectrum analysis of the irradiated fuel gamm-radiation. Burn-up has been determined (1) by the applied here procedure, and (2) from the results of mass-spectrometric determination of the isotopic composition and content of U, Pu and Nd. Two groups of the values obtained have been compared and the influence of the errors of the measured activity quotients on the established deviations has been evaluated.     

Identification of unknown nuclear reaction products using gas phase separation and chemical pattern analysis

A method for the fast separation of short-lived nuclides in the gas phase is described. The attribution of new γ-lines to a certain element is possible by the variation of the chemical and physical separation parameters and the determination of the “yield” of a line, compared with the “yields” of well known nuclides.     

Distribution characteristics of fission gas along the axial direction for an irradiated fuel rod of a pressurized water reactor (PWR)

In this work, the fission gas distribution of an irradiated oxide fuel was investigated by measuring the fission gas release (FGR) and retained gas from and in an oxide fuel along the axial height of the fuel rod. In addition to the fission gas measurements, other destructive and non-destructive post irradiation examination tests such as gamma scanning, eddy current testing, density and oxide layer thickness measurements, and ceramography/metallography were conducted to review their impact on the FGR and gas retention. A lead rod was extracted from a lead test assembly, which was irradiated to 56.9 GWd/tU during three irradiation cycles for a total exposure of 1,406 effective full power days. Considering the water-side oxide layer thickness and local burnup of the fuel rod, the rod was sectioned into four positions along the axial height and four samples were prepared from these positions. A sample fragment of around 0.1–0.2 g was individually melted to measure its retained krypton and xenon concentration. According to the measurement results, the retained krypton and xenon concentration ranges were 0.114–0.139 cc/gU and 1.073–1.338 cc/gU, respectively, and their retention percentage after normalization of the sample local burnup showed a decreasing trend as the axial height of the fuel rod raised. The water-side oxide layer thickness scope of the tested fuels measured by an observation of the optical microscope images was 13–42 μm, and the thickness was increased along the axial height of the rod. The amount of released krypton and xenon into the rod-free volume was measured as 5.7 and 54.6 cc, respectively, by a rod puncturing/collection procedure, which corresponds to a 1.94 % fractional fission gas release referring to fission gas generation by a code calculation.     

Isolation of radiocesium from a fission product mixture

A method of¹³⁷Cs isolation from strongly, acidic solutions of fission products is described, in which vanadyl ferrocyanide is used as a selective ion exchanger for cesium. The effects of the acidity of medium and the carrier concentration on the quantitative yield of separation have been studied and convenient conditions have been found for¹³⁷Cs isolation from the solution of fission products formed after irradiating uranium with neutrons.     

The measurement of retained fission gas compositions and their isotopic distributions in an irradiated oxide fuel by inert gas fusion-mass spectrometric analysis

In this work, an easy, fast and reliable measurement technique for the quantitative determination of retained fission gases in an irradiated oxide fuel was developed. Many experiments were conducted to determine the optimum conditions for fusion of an oxide fuel, for the quantitative collection and measurements of the released gases. Ion implantation technology was applied to make a krypton or xenon references in a solid matrix. A fragment of oxide fuel, about 0.1 g of an unirradiated SIMFUEL, was completely fused with excess metallic fluxes, 1.0 g of nickel and 1.0 g of tin, in a graphite crucible of a helium atmosphere for 120 s at 850 A as a mixture of metals and alloys. About 96 ± 3 to 98 ± 4% of the krypton and xenon that were injected into the instrument using a standard gas mixture was reproducibly recovered by collecting the releasing gas through the instrument for 120 s. Using the same fusion and collection conditions, it was possible to recover about 97 ± 3% of the injected krypton and xenon by fusing a fragment of SIMFUEL which was wrapped with krypton or xenon implanted aluminum foils. The recovery test results of krypton and xenon using ion planted aluminum foils gave encouraging results suggesting their potential use as a reference specimen. It was confirmed that a fragment of irradiated oxide fuel, 0.051 g, with a code burn-up of 56.9 MWd/MtU, was completely fused as the mixture of metals and alloys through the fusion conditions and more than 99% of the retained fission gases were recovered during the first fusion. Since no cryogenic trap was needed, the collected gas could be measured directly and thus the analysis time could be further reduced. Approximately 7 min was sufficient to finish the measurement of retained fission gases in the irradiated oxide fuel using the developed procedure.     

Determination of americium-243 through its daughter product neptunium-239

The extraction of Np(IV), Zr, Nb, Cs, Ce(III) and Am(III) from nitric acid solutions containing oxalate and phosphate ions by solutions of 1-phenyl-3-methyl-4-benzoylpyrazolon-5 (PMBP) and tri-n-butyl phosphate (TBP) in benzene has been investigated. A solution 0.1M in respect to PMBP and 0.25M in respect to TBP was found to extract 99% of neptunium from aqueous solutions 1M in respect to H₃PO₄ and 0.5M in respect to HNO₃. Under these conditions, the extraction of the other investigated elements does not exceed 0.1%. Based on this finding, a procedure was developed to determine²⁴³Am through its daughter product²³⁹Np in solutions containing large quantities of curium and its fission products. The sensitivity of the procedure is 1·10⁻⁷ mg of²⁴³Am in the sample. The²⁴³Am content is obtained by calculation from measurements of the γ-activity of the extracted²³⁹Np. The purification ratio of²³⁹Np is∼10⁵ from Zr, Nb and Ru, ∼10⁸ from Ce and Cm and >10¹² from Cs.     

The history of the discovery of nuclear fission

Following with the discovery of the electron by J. J. Thomson at the end of the nineteenth century a steady elucidation of the structure of the atom occurred over the next 40 years culminating in the discovery of nuclear fission in 1938–1939. The significant steps after the electron discovery were: discovery of the nuclear atom by Rutherford (Philos Mag 6th Ser 21:669–688, 1911), the transformation of elements by Rutherford (Philos Mag 37:578–587, 1919), discovery of artificial radioactivity by Joliot-Curie and Joliot-Curie (Comptes Rendus Acad Sci Paris 198:254–256, 1934), and the discovery of the neutron by Chadwick (Nature 129:312, 1932a, Proc R Soc Ser A 136:692–708, 1932b; Proc R Soc Lond Ser A 136:744–748, 1932c). The neutron furnished scientists with a particle able to penetrate atomic nuclei without expenditure of large amounts of energy. From 1934 until 1938–1939 investigations of the reaction between a neutron and uranium were carried out by E. Fermi in Rome, O. Hahn, L. Meitner and F. Strassmann in Berlin and I. Curie and P. Savitch in Paris. Results were interpreted as the formation of transuranic elements. After sorting out complex radio-chemistry and radio-physics O. Hahn and F. Strassmann came to the conclusion, beyond their belief, that the uranium nucleus split into smaller fragments, that is nuclear fission. This was soon followed in 1939 by its theoretical interpretation by L. Mietner and O. Frisch.     

Specific activities of radioactive elements from fission product mixtures

Specific activities of radioactive elements at the time of chemical separation from fission product mixtures produced by thermal neutron fission of²³⁵U were computed byBateman's and other equations on an electronic computer. Computations were made for two fission times: fission was assumed to be complete in a few minutes in one case, and over a period of a year in the other case. It was also assumed that each element was separated instantly after allowing the fission products to decay for 1∼10 000 000 hrs (1 140 years). The computations were applied to 12 important elements: Ru, Zr, Nb, Cs, Sr, Pm, Tc, Ba, La, Ce, Kr and Y. Results are given as a diagram for each element. The diagrams are intended to be helpful in the chemical processing of a large quantity of fission products, and industrial or tracer application of these elements.