Irradiated Uranium Research Articles

The transfer of irradiated uranium from the Irish Sea coast to the terrestrial environment in Cumbria, UK

Abstract Sea-to-land transfer of irradiated uranium in soil samples was measured along a transect from the Cumbrian coast south of the Windscale Sellafield nuclear site in a NE direction across the Cumbrian coastal plain. Soluble uranium components were extracted by ion exchange chromatography and ratios 236U:238U and 235U:238U were measured using multi-collector ICP-MS. The method employed demonstrated the utility of 236U ratio measurements for the determination of irradiated uranium in environmental samples. The results produced for uranium were like those that have been reported for other fission product and nuclear fuel components that have been shown to be marine in origin. Measured 236U levels were highest close to the seashore and decreased rapidly as an exponential function of distance. The results indicated the absence of depleted uranium, such as might result from the deposition of aerosols generated by the testing of DU munitions at the nearby Eskmeals firing range in all but one sample.

Развитие работ по обращению с РАО на АО «СХК»

The article overviews the experience gained by the Siberian Chemical Combine in dealing with radioactive waste management and nuclear decommissioning challenges. It provides a particularly detailed discussion on the decommissioning efforts implemented at a production facility previously operated at the site No. 3 of its Radiochemical Plant and engaged in the reprocessing of irradiated standard uranium slugs (ISUS). This case study exemplifies SCC’s consistent efforts focused on digital models representing large production facilities, that provided a basis for relevant design specifications and estimates developed in an entirely digital format. The paper also overview the cleanup activities implemented at storage pools B-1 and B-2. The paper considers SCC’s active engagement in the Proryv project as a standalone achievement strengthening its position as a leading nuclear enterprise in Russia.

Thermal conductivity measurement of uranium–plutonium mixed oxide doped with Nd/Sm as simulated fission products

The thermal conductivities of near-stoichiometric (U,Pu,Am)O2 doped with Nd and/or Sm, which are major fission products (FPs) generated during irradiation of uranium–plutonium mixed oxides (MOX) fuels, as simulated fission products are evaluated from 1073 to 1673 K. The thermal conductivities are calculated from the thermal diffusivities that are measured using the laser flash method. To evaluate the thermal conductivity from a homogeneity viewpoint of Nd/Sm cations in MOX, the specimens with different homogeneities are prepared using ball-mill and fusion methods. A homogeneous Nd/Sm distribution decreases the thermal conductivity of MOX with increasing Nd/Sm content, whereas heterogeneous Nd/Sm has no influence. The effect of Nd/Sm on the thermal conductivity is studied using the phonon transport model λ=(A+BT)−1. The dependences of the coefficients A and B on the Nd/Sm content (CNd and CSm, respectively) are evaluated as: A(mK/W)=1.70×10−2+0.93CNd+1.20CSm, B(m/W)=2.39×10−4.

Atomistic study of the properties and behavior under uniaxial tensile stress of some representative UO[formula omitted] grain boundaries

In irradiated polycrystalline uranium dioxide (UO2), the pressure generated at high temperature by the noble gases in the intergranular bubbles as well as the thermomechanical stresses due to temperature gradients cause the fracture of the grain boundaries. In this study, one analyzes through atomistic calculations of static and molecular dynamics type the properties of three UO2 grain boundaries and their behavior under uniaxial tensile loading up to fracture. In these atomistic simulations, the interactions between atoms are described using a many-body variable charge potential. Structural analysis, performed on the three grain boundary structures under uniaxial tensile stresses and over a wide temperature range, reveals no source of plasticity. This suggests a brittle behavior at fracture, in agreement with recent experimental studies carried out on micrometer size specimens containing grain boundaries. A method based on thermodynamic considerations is used to build normal cohesive zone laws for the UO2 grain boundaries using molecular dynamics simulation data. The cohesive zone laws will be employed in a future study in failure simulations at higher scales.

Progress of the Analytical Methods and Application of Plutonium Isotopes in the Environment

Since the beginning of the use of nuclear energy in the twentieth century, atmospheric nuclear weapons testing, nuclear accidents, and spent fuel reprocessing have released large amounts of radioisotopes into the environment, especially plutonium isotopes. As an important anthropogenic radionuclide, plutonium is mainly produced by neutron irradiation of uranium. Plutonium isotopes and their ratios from different sources have a specific ‘fingerprint’, which is determined by irradiation time and intensity. Accordingly, several methods have been developed for the analysis of plutonium isotopes. As a popular new tracer in recent years, plutonium isotopes could be used for environmental tracer applications and source identification of regional environmental contamination sources. This paper summarizes and compares the pretreatment of plutonium isotopes in various samples and the detection and analysis methods in recent years. Plutonium isotopes in various environmental samples were enriched by chemical treatments, such as ashing, acid digestion, and alkali fusion. Then it was purified and separated by extraction resins to remove interfering nuclides for the subsequent mass spectrometry. The practical applications of plutonium isotopes and their related ratios in the environment are summarized, such as the determination of plutonium isotopes and their related ratios in the environment near two representative nuclear power plant accidents and the monitoring of the safe operation of nuclear power plants; the establishment of a plutonium isotope database in the vicinity of Chinese waters; and the traceability of plutonium isotopes in environmental regions through the global atmosphere and sand and dust. Finally, the outlook for subsequent research in plutonium isotope detection and applications is presented.

Measurement of Mass‐Independent Perturbations of Cadmium in Nuclear Debris Samples

The isotopic composition of cadmium in nuclear debris was measured by multi‐collector inductively coupled plasma‐mass spectrometry (MC‐ICP‐MS). Mass‐independent perturbations due to fission product decay on the isotopes of Cd were observed. Relative fission yields for masses 111, 112, 114 and 116 were determined in each sample of nuclear debris and compared with reported values found in irradiated plutonium and uranium fuels. Mass spectrometry measurements of the valley fission product endpoints are extremely challenging due to the low cumulative fission yields and small amounts of each isotope that are produced. To overcome these challenges, a new purification method for Cd was developed and validated by examination of a number of geological reference materials.

Thermodynamic predictions of CANDU fuel volatilization and fission product behaviour under severe accident conditions

A comprehensive thermodynamic investigation has been performed on irradiated uranium dioxide CANDU fuel under representative severe accident conditions. Two cases have been explored in this work: Case I) the fuel is in contact with varying combinations of hydrogen and steam (i.e., mainly reducing), and Case II) the fuel is in contact with varying combinations of air and steam (i.e., oxidizing). The system thermochemistry has been computed for a wide range of gas mixtures for both cases, fuel to gas ratios, temperatures, and hydrostatic pressures. The calculations predict fission gas speciation and fuel volatilization based on equilibrium thermodynamics, including the formation of several secondary phases, changes in UO2±x stoichiometry, fission product solubility in various phases, melting, and vaporization. The overall objective of this work was to inform source term analyses in the industry standard toolset SOURCE under severe accident conditions of a CANDU station.

Using Fission-Track Radiography Coupled with Scanning Electron Microscopy for Efficient Identification of Solid-Phase Uranium Mineralogy at a Former Uranium Pilot Mill (Grand Junction, Colorado)

At a former uranium pilot mill in Grand Junction, Colorado, mine tailings and some subpile sediments were excavated to various depths to meet surface radiological standards, but residual solid-phase uranium below these excavation depths still occurs at concentrations above background. The combination of fission-track radiography and scanning electron microscope energy-dispersive X-ray spectroscopy (SEM-EDS) provides a uniquely efficient and quantitative way of determining mineralogic associations of uranium that can influence uranium mobility. After the creation of sample thin sections, a mica sheet is placed on those thin sections and irradiated in a nuclear research reactor. Decay of the irradiated uranium creates fission tracks that can be viewed with a microscope. The fission-track radiography images indicate thin section sample areas with elevated uranium that are focus areas for SEM-EDS work. EDS spectra provide quantitative elemental data that indicate the mineralogy of individual grains or grain coatings associated with the fission-track identification of elevated uranium. For the site in this study, the results indicated that uranium occurred (1) with coatings of aluminum–silicon (Al/Si) gel and gypsum, (2) dispersed in the unsaturated zone associated with evaporite-type salts, and (3) sorbed onto organic carbon. The Al/Si gel likely formed when low-pH waters were precipitated during calcite buffering, which in turn retained or precipitated trace amounts of Fe, As, U, V, Ca, and S. Understanding these mechanisms can help guide future laboratory and field-scale efforts in determining long-term uranium release rates to groundwater.

Study of new routes for purification of fission 99Mo

99mTc is the most applied medical radioisotope in the world, especially for cancer diagnosis procedures. It is provided by 99Mo radioactive decay, which is one of the fission products from the uranium irradiation in nuclear reactors. At the main production plants, the 99Mo chemical processing may be lined up in several steps to separate it from other fission products according to the features of the used targets or the local requirements as well. In this work, two routes of 99Mo purification, MR1, and MR2, were purposed as an alternative to be set up in the Brazilian Multipurpose Reactor project (BMR). The MR1 route was performed by three chromatographic columns packed with Dowex 1x8 resin, Chelex resin, and alumina, respectively. The route MR2 was carried out also by chromatography applying two columns filled with Dowex 1x8 and alumina respectively, but including a sublimation process performed in a tubular oven under programmed conditions. The final yield for the MR1 route was 84.4 % and the overall time process was about 7 hours, whereas the MR2 route reached 75.3 % in 9 hours.

Recovery of high specific activity molybdenum-99 from accelerator-induced fission on low-enriched uranium for technetium-99m generators

A new process was developed to recover high specific activity (no carrier added) 99Mo from electron-accelerator irradiated U3O8 or uranyl sulfate targets. The process leverages a novel solvent extraction scheme to recover Mo using di(2-ethylhexyl) phosphoric acid following uranium and transuranics removal with tri-n-butyl phosphate. An anion-exchange concentration column step provides a final purification, generating pure 99Mo intended for making 99Mo/99mTc generators. The process was demonstrated with irradiated uranium targets resulting in more than 95% 99Mo recovery and without presence of fission products or actinides in the product.

Alteration of organic macerals by uranium irradiation in lower Paleozoic marine shales

Measuring the degree of organic matter (OM) alteration caused by uranium (U) irradiation is important in the effective evaluation of the hydrocarbon potential and thermal maturity of U-rich source rocks. This study investigates OM alteration in the Baltoscandian U-rich, lower Paleozoic Alum shale obtained from three research wells in the thermally immature parts of eastern Sweden and western Estonia. The results show a marked increase in the proportion of refractory, non-generative organic carbon (NGOC) in the U-rich shale. Fluorescence spectrometry of unicellular alginites and lamalginites show a reduction of up to 95.6% in fluorescence relative intensity (RI) with increase in U content from 29 to 401 ppm. Fluorescence spectra of the liptinites exhibit a marked “red shift” as expressed by increase in red(R)/green(G) quotients. This change in fluorescence properties of the liptinites is associated with up to 77.6% decrease in hydrogen index (HI) and hence loss of hydrocarbon generation potential. Furthermore, organic molecules show significant loss of aromatic moieties as well as aliphatic constituents resulting in a more condensed macromolecular structure. The measured random reflectance of solid bitumen (BRo) appears to be significantly elevated in micro scale proximity to the U-containing minerals. The heterogeneous distribution of U-containing minerals and the contact with solid bitumen attributes to the wider range of BRo values and lead to the overall increase in mean BRo in samples with high U contents. In contrast, Ro values of zooclast macerals (GRo, e.g., graptolite, chitinozoans and vitrinite-like fragments) appear to be less influenced by U irradiation and hence are a more reliable maturity indicator in U- rich lower Paleozoic shales.

Rapid Separation of Photofissioned Uranium Products via a Single-Pass Multiplexed Chromatographic Fission Product Separation System.

Current state-of-the-art fission product separations frequently involve multiple independent separation columns and sample manipulation processes; to couple these processes together, multiple evaporation and transposition steps are often required. The addition of these steps results in lengthy separation times, increased analysis costs, the potential for sample loss, and release of radioactive contamination. We report a new semiautomated method for the rapid separation of U, Zr, Mo, Ba, Sr, Te, and lanthanide fission products from irradiated uranium samples. Chemical yields for U, Zr, Ba, Sr, Te and the lanthanides from less than 3-day old uranium fission product samples are consistently greater than 90%, while those of Mo are greater than 70%. This method minimizes the use and addition of oxidation and reduction reagents that often cause issues with retention and separation. Uranium dissolution and fission product separations using this single-pass method are achievable in under 2 h, representing a significant improvement over traditional gravimetric uranium fission product separation procedures.

Supply of I-131 in a 2 MW molten salt reactor with different production methods

Four I-131 production methods including irradiated TeO2 target and uranium target in the irradiation channel, batch-wise extracted iodine from the fuel salt, and online extracted solid tellurium through the by-pass loop system have been assessed in a 2 MW molten salt reactor. The latter method can produce a large annual yield of I-131 (about 155,000 Ci). The radioactivity shielding demand of the latter method is much smaller than the other I-131 production methods under the identical annual yield of I-131.

The application of synchrotron micro-computed tomography to characterize the three-dimensional microstructure in irradiated nuclear fuel

This research demonstrated the first time that synchrotron X-ray micro-computed tomography (μ-CT) was utilized to analyze the three-dimensional (3-D) morphology of the microstructure of irradiated nuclear fuel. Both absorption contrast and propagation-based phase contrast enhanced μ-CT techniques were evaluated. Porosity and phase region volumes were determined in neutron irradiated uranium 10 wt% zirconium (U–10Zr) using propagation-based phase contrast enhanced μ-CT. Volumes were calculated using image processing software and algorithms, with the porosity volume subsequently compared to the ASTM E562 method. The 3-D porosity of a small region (∼8 × 105 μm3) located in the intermediate fuel redistribution zone that was removed via a focused ion beam was determined to be 7.2% via image processing. A detailed 3-D analysis of the porosity with respect to volume and morphology was provided. Five growth stages of pore evolution were observed including: nucleation, growth, coalescence, interconnected porosity, and extended/interconnected porosity. Four distinct phase regions were identified, including the pores, via both absorption contrast and propagation-based phase contrast enhanced μ-CT. A quantitative comparison of fuel swelling between this localized microscopic U–10Zr fuel region that was previously irradiated in the Fast Flux Test Facility, and historic U–10Zr fuels irradiated in the Experimental Breeder Reactor-II (EBR-II), was conducted. It was determined that for similar burnups, the localized swelling in this microscopic fuel region was 7.7%, which was approximately one-half of the value of the macroscopic swelling reported in EBR-II irradiated fuel. With the advent of new specimen preparation tools for irradiated materials, as well as advances in characterization techniques, researchers can now obtain previously inaccessible microstructural insight into irradiated fuels. This possibility naturally lends itself to improved fuel performance models for the nuclear community.

Approximate model for evaluation of thermal-hydraulic transients associated with rapid power increase in research nuclear reactor

A simple model, for the estimation of changes in the nuclear fuel element cladding temperature as well as the conditions of the formation of the onset of nucleate boiling, is proposed. The results of this estimation are sufficient to assess the effect of the transient with the peak of the reactor's power on the device's safety, without the need for time-consuming thermal calculations. The basic parameters determined using the proposed model are the maximum wall temperature of the device in a hot spot, the time constant of the wall temperature change, and the course of changes in the onset of nucleate boiling ratio in time. The model may be used for investigating the thermal behavior of thin heat-generating and water-cooled elements (such as fuel elements or uranium irradiation targets) during rapid power rise. The results of the temperature estimation with the proposed model were tested considering the hot spot in the MR-6 type nuclear fuel. The SN code with coupled kinetics and thermal-hydraulics, developed in the MARIA reactor was used to validate the results. The maximum cladding temperature, the thermal time constant and the onset of nucleate boiling ratio parameter simulated by the SN code and the proposed scheme appeared to be very consistent.

Changes in physicochemical properties of organic matter by uranium irradiation: A case study from the Ordos Basin in China

Carbonaceous debris (CD) are exclusively disseminated in sandstones of uranium-bearing strata from the Dongsheng uranium ore field of the northern Ordos Basin, northern China. Physicochemical properties of CD were investigated through a series of tests including element analyses, Fourier transform infrared spectroscopy (i.e., FTIR) and organic carbon isotope (i.e., δC13) as well as specific surface area. The results show that δC13 of the samples from uranium-bearing strata range from −22.86‰ to −25.82‰, indicating that the CD is from the same origin. The average values of C, H, O and N are 54.54%, 4.55%, 39.15% and 0.49%, respectively, and H/C as well as O/C atomic ratios vary from 0.77 to 1.26 and 0.36 to 1.72, respectively, implying that CD is mostly composed of C and O, followed by H, and is grouped into Type Ⅲ kerogen at an immature to low mature stage. Compared with the samples without uranium enrichment, C, H, N and δ13C of the uranium-bearing samples increase by 13.66%, 5.32%, 12.50% and 0.98‰, respectively, while decrease by 18.50%, 7.84% and 33.33%, respectively, for O, H/C as well as O/C atomic ratios. As the intensity of uranium enrichment increases, H/C atomic ratios and aliphaticity decrease, and aromaticity increases. Besides, uranium-bearing samples are of smaller specific surface area than those without uranium enrichment. Alpha irradiation energy released by uranium and its daughters could cause the crosslinking in n-alkanes and long-chain polymers through a free-radical mechanism, which is supported by a decrease in O, aliphaticity and an increase in aromaticity. Hence, it results in polycondensation of molecular structure with a decrease in side chain length, function groups as well as specific surface area, and changes in physicochemical properties of organic matter, which is instructive in prospecting.

ADVANTG hybrid code application for the analysis of neutron field depression in uranium foils activated in the MARK III converter of the BR1 reactor

This paper deals with the self-shielding correction calculation for natural and 40% enriched uranium foils activated in a fast-epithermal neutron field in the MARK III converter of the BR1 research reactor. Although the irradiated uranium foils are thin (~70 μm to 90 μm), neutron self-shielding is not negligible. The calculation of the neutron field depression in the foils was challenged by unreliable results and long computing times using the MCNP6.2 Monte Carlo code. These simulations require the use of the variance reduction techniques to compute a detector response with low uncertainties within an acceptable computing time. An additional problem was the convergence of the spectrum averaged cross section of the 238U(n,γ) reaction, even with a high number of particle histories (e.g. 1 × 1010), even when the average neutron flux has low statistical uncertainty. To speed up the calculations and to test the reliability of variance reduction techniques, a comparison was made with the tally results and statistical uncertainties of the MCNP6.2 weight window technique, ADVANTG-accelerated MCNP6.2 and default MCNP6.2 calculations. To generate variance reduction parameters with the ADVANTG code for the final calculations, the consistent adjoint driven importance sampling (CADIS) method is used. The neutron self-shielding has been found considerable in the 238U(n,γ) reaction due to three strong resonances (at 7 keV, 21 keV and 37 keV) leading to a correction factor of 1.57 for the natural uranium foil. The sensitivity study has shown a negligible impact from the selected nuclear data libraries (JEFF and ENDF/B) on the 238U(n,γ) and 235U(n,f) spectrum averaged cross sections in the MARK III converter.

Performance testing of a Compton suppressed coincidence measurements using the Advanced Radionuclide Gamma-spectrOmeter

A Compton-suppressed multidimensional gamma-spectrometer is being developed within the Shallow Underground Laboratory at Pacific Northwest National Laboratory (PNNL, USA). This Advanced Radionuclide Gamma-spectrOmeter system consists of two Broad Energy Germanium detectors, surrounded by a 12-detector sodium iodide Compton suppression shield, situated inside a low-background lead and copper shield, fitted with a cosmic veto background reduction system. The detector has advanced functionality, including operation in single or combined detector mode, coincidence mode and Compton suppressed coincidence mode. Performance testing experiments measuring a neutron irradiated natural uranium sample, have demonstrated order of magnitude improvements in gamma detection sensitivity compared to conventional gamma-spectrometry techniques. Importantly this research demonstrates the feasibility of processing and analysing high-multiplicity datasets and the potential for high-detection sensitivity radionuclide measurements.

A novel chronometry technique for dating irradiated uranium fuels using Cm isotopic ratios

A novel chronometry method is developed to date irradiated nuclear fuels. It is based on the measurement of two Cm isotopic ratios (244Cm/246Cm and 245Cm/246Cm) and does not require the addition of a yield tracer. The method relies on the decay of 244Cm while the 245Cm/246Cm ratio is considered constant after irradiation. Feasibility is demonstrated using different types of irradiated fuels with known irradiation history. A precision of 5 year is reached so far but clear potential for improvement has been identified. The novel chronometry method represents a promising alternative to existing dating techniques for nuclear materials.

Targeted alpha therapy with bismuth-213 and actinium-225: Meeting future demand.

Targeted alpha therapy (TAT) is a promising approach for the treatment of cancer. The use of alpha emitters for cancer therapy has two distinct advantages over conventional therapies. The short range of alpha radiation in human tissue (less than 0.1 mm), corresponding to only a few cell diameters, allows selective killing of targeted cancer cells while sparing surrounding healthy tissue. At the same time, the high energy (several MeV) of alpha radiation and its associated high linear energy transfer leads to highly effective cell kill. Consequently, alpha radiation can destroy cells which otherwise exhibit resistance to treatment with beta or gamma irradiation or chemotherapeutic drugs, and can thus offer a therapeutic option for tumors resistant to conventional therapies. Recent results demonstrating the remarkable therapeutic efficacy of alpha emitters to treat various cancers have underlined the clinical potential of TAT. This paper describes the recent clinical experience with 213 Bi and 225 Ac. In view of the enormous benefit of targeted cancer treatment with alpha emitters, their production will have to be considerably increased beyond current supply capabilities. Alternative production methods based on the irradiation of uranium, thorium, or radium targets at reactors or accelerator facilities have the potential to meet future demand.