Depleted Uranium Samples Research Articles

Photofission delayed gamma-ray measurements in a large cemented radioactive waste drum during LINAC irradiation

Nuclear materials characterization in 870 L cemented waste drums is challenging due to significant neutron and gamma ray attenuation. Among the existing non-destructive approaches, photon active interrogation is of particular interest. A LINAC accelerates electrons to energies greater than 10 MeV, leading to highly penetrating interrogating bremsstrahlung photons that can reach nuclear materials in the depth of the 870 L waste drum and allow the detection of delayed gamma rays emitted by photofission products. Because the interrogating photon and neutron fluxes are extremely high during irradiation, measurements with high purity germanium detectors are usually performed after a long irradiation time, for example by moving the detector or the waste drum. Therefore, such measurements are only sensitive to delayed gamma rays associated to photofission products having sufficient long half-life of at least tens of seconds, which cuts a valuable signal coming from fission products with a shorter half-life. In this paper, we report gamma rays measurements of very short lived photofission products using a 3 in × 3 in cylindrical LaBr3(Ce) fast scintillation detector. A depleted uranium sample was irradiated by a pulsed bremsstrahlung source with maximum energy 17.5 MeV in a macro pulsing mode (1 s irradiation, 2 s cooling). This allowed to acquire a strong delayed gamma signal within a time window of 2 s.A clear signature of nuclear materials was observed even in the depth of a concrete mock-up, with an excellent signal-to-noise ratio beyond 3.5 MeV where the background of nonnuclear activated materials is negligible. These results are compared to a semi analytical model that qualitatively agrees but underestimates the measurements by a constant factor 1.7, probably caused by a wrong beam intensity normalization. Indeed, in this work the LINAC pulse time structure (macro pulsing) was different from its nominal working point used to characterize the beam (usual pulsing mode). Since it allows direct calculations of the detector response to delayed gamma-ray photofission measurements, this model is a fast calculation alternative to time-consuming Monte Carlo simulations, in view of further studies that will follow this feasibility demonstration.

High-fidelity photoneutron detection via neutron activation analysis

To interdict illicit radioactive sources and prevent nuclear terror, governments have deployed portal monitor systems at ports of entry. Such portal monitors rely upon the detection of spontaneously-emitted radiation signatures, however, detection scenarios become more challenging for highly enriched uranium, which emits relatively few spontaneous radiation signatures. Photon active interrogation techniques can augment current cargo inspection scenarios and improve detection capabilities for illicit nuclear material. A high-energy photon interrogation source can induce photofission in nuclear material, producing strong radiation signatures for detection. Prompt fission signals are of specific interest because prompt neutrons are greater than 99% of the fission signal. When using detectors capable of spectroscopy, e.g., pulse shape discrimination-capable organic scintillators, the intense interrogating source often will cause pulse pile-up, degrading particle classification. In this work, we demonstrate photoneutron detection via neutron activation of aluminum and iron. The experiments rely upon nuclear reactions that can only be induced by neutrons, and are robust against pile-up, thereby providing high-confidence neutron rates. In this work, we apply neutron activation analysis for the detection of high-energy neutrons produced by high-energy photon irradiation; the depleted uranium sample was irradiated with a 9-MeV bremsstrahlung beam. We find that the irradiated depleted uranium produced an aluminum activation flux of 1.77 × 105 n cm −2 s−1 and an iron-56 activation flux of 1.05 × 105 n cm −2 s−1. The ability to detect radioactivity induced by neutrons created in (γ, n) and photofission reactions demonstrates a new approach for photon active interrogation to identify nuclear materials.

Methods and Influencing Factors for the Simple and Rapid Identification of Depleted Uranium Weapon Use under Battlefield Conditions.

The purpose of this paper is to explore how to rapidly and easily identify depleted uranium (DU) samples under battlefield conditions and to study the factors that influence their measurement. The air-absorbed dose rate and surface contamination levels for DU samples of 2-330 g were measured using a patrol instrument and portable energy spectrometer. The results were analyzed in accordance with IAEA standards for judging radioactive substances. The energy spectra of 5-g quantities of DU samples were analyzed using a high-purity germanium gamma spectrometer, and the uranium content of 100 mg DU samples was determined with an inductively coupled plasma mass spectrometer to clarify the type and composition of the uranium. The same batches of DU samples were identified using a portable gamma-ray spectrometer. We added 0-5 g environmental soil powders at different proportions. After sealing, the spectra were collected with a detection distance of 1-5 cm for 10 min. The activities of U and U nuclides in the samples were detected with an NaI(TI) scintillation detector. The U and U mass abundances in samples were calculated from measured specific activities. The sample was determined to contain DU if the U to U ratio was below 0.00723. It is found that for detecting DU materials with a low activity, surface contamination level measurements are more effective than calculating the air-absorbed external irradiation dose rate. Hence, for low-activity samples suspected to be radioactive, a radiometer with a high sensitivity for surface contamination is recommended, and the optimal measurement distance is 1-3 cm. Under all detection conditions, U can be identified using a portable gamma spectrometer, whereas U can only be detected under certain conditions. If these nuclides can be detected simultaneously, a U to U ratio of below 0.00723 indicates the presence of DU. The main factors affecting this identification include the sample mass, sample purity, measurement distance, and measurement time. For the rapid identification of DU with a portable gamma-ray spectrometer, the mass of uranium in the sample must be more than 1 g, the measuring distance needs to be less than 1 cm, and the measuring time must be 1-10 min. It is feasible to use a portable gamma-ray spectrometer to rapidly identify the types and composition of nuclides in DU samples. The detection of U activity is a precondition for the identification of DU.

Determination of the uranium enrichment without calibration standards using a 2 × 2 inch LaBr3(Ce) room temperature detector and Monte Carlo sampling approach for uncertainty assessment

In recent years room temperature medium resolution scintillation devices, such as LaBr3(Ce), have attracted much interest as possible alternatives to traditional spectrometers based on HPGe and NaI detectors, for the determination of the uranium enrichment in safeguards applications. This paper focuses on the investigation of possibilities and limits of a net peak area based methodology used for the determination of the uranium enrichment without use of calibration standards and introduces the isotopic code MCSIGMA for LaBr3(Ce) scintillators. Tests are conducted with a room temperature, medium resolution spectrometer based on a 2×2 inch LaBr3(Ce) scintillator using which spectra of different statistical quality are obtained from certified uranium standards. Gamma peaks in the 143-1001keV energy range are used as uranium gamma-ray signatures. Results indicate a promising performance of the applied methodology with a room temperature medium resolution scintillator of the LaBr3(Ce) type, however at a cost of significantly higher uncertainty budget on the derived enrichment compared to HPGe, especially for natural and depleted uranium samples. This uncertainty budget is primarily influenced by the statistical quality of the measured spectra. Implemented algorithms and analysis routines are described in detail and presented.

Morphology study on the depleted uranium as hydriding/dehydriding cycles

Depleted Uranium (DU) is one of the strongest candidates as a getter material of hydrogen isotopes in the nuclear fusion reactor. In this work, small DU lump specimen with 99.8% purity was prepared for observation of morphology variation as hydriding/dehydriding cycles. Hydriding/dehydriding of DU was carried out more than 10 cycles for powder preparation. The pulverized DU specimen was safely handled in the glove box under Argon gas condition to minimize contact with oxygen and humidity. The morphology change according to hydriding/dehydriding cycles was observed by visual cell reactor, optical microscope and scanning electron microscope. The first hydriding of the small DU sample has progressed slowly with surface enlargement and volume expansion as time passes. After third hydriding/dehydriding cycles, most of DU was pulverized. The powder fineness of DU developed as hydriding/dehydriding cycle progresses. But the agglomerates of fine DU particles were observed. It was confirmed that the DU particles exist as porous agglomerates. And the particle agglomerate shows poor fluidity and even has the cohesive force.

A novel inductively coupled plasma atomic emission spectrometry method for uranium isotope ratio measurements using chemometric techniques

Inductively coupled plasma atomic emission spectrometry was applied for uranium isotope ratio measurement. Even high resolution spectrometers cannot resolve small isotopic shift between 235U and 238U emission lines. Multivariate calibration techniques are useful tools for overcoming spectral overlap. The emission spectra of natural and depleted uranium samples were modelled by the laboratory written PLS-2 program (according to PLS algorithm). Four factors were selected as the optimum number of factors. The PLS model was applied on an independent test set. The results were validated by thermal ionization mass spectrometry (TIMS). A good agreement between the predicted results from the PLS-2 model and the TIMS results were obtained.

CALCULATING CAPSTONE DEPLETED URANIUM AEROSOL CONCENTRATIONS FROM BETA ACTIVITY MEASUREMENTS

Beta activity measurements were used as surrogate measurements of uranium mass in aerosol samples collected during the field testing phase of the Capstone Depleted Uranium (DU) Aerosol Study. These aerosol samples generated by the perforation of armored combat vehicles were used to characterize the DU source term for the subsequent Human Health Risk Assessment (HHRA) of Capstone aerosols. Establishing a calibration curve between beta activity measurements and uranium mass measurements is straightforward if the uranium isotopes are in equilibrium with their immediate short-lived, beta-emitting progeny. For DU samples collected during the Capstone study, it was determined that the equilibrium between the uranium isotopes and their immediate short-lived, beta-emitting progeny had been disrupted when penetrators had perforated target vehicles. Adjustments were made to account for the disrupted equilibrium and for wall losses in the aerosol samplers. Values for the equilibrium fraction ranged from 0.16 to 1, and the wall loss correction factors ranged from 1 to 1.92. This paper describes the process used and adjustments necessary to calculate uranium mass from proportional counting measurements.

Estimation of 235U concentration in some depleted uranium samples by high resolution gamma-ray spectrometry using 185 keV and 1001 keV gamma-energies of 235U and 234mPa

The identification of isotopic composition of depleted uranium obtained after the reprocessing of spent fuel is very much important as far as their further applications are concerned. The concentration of 235U in the reprocessed uranium will be lower and their depletion depends mainly on the type of the reactor and burn up. The depleted uranium has a variety of applications in nuclear fuel cycle operations and industries warranting quick assessment of its isotopic composition. This paper describes the methodology developed and used for the estimation of 235U concentration in reprocessed uranium in the form of uranyl nitrate solution. A non destructive technique by high resolution gamma-ray spectrometry with HPGe detector has been used for the analysis. The activity ratio of 238U/235U, obtained from the absolute activity measurement on the 185 keV gamma-line of 235U and 1001 keV gamma-line of 234mPa, has been used for the estimation of 235U isotopic content in the sample using a mathematical formula. This method offers rapid and reliable estimate of the 235U concentration in samples comparable with that of mass spectrometry measurements. The results show that 235U concentration in the sample can be determined within 5% error for 10000 seconds counting.

The applicability of MGA method for depleted and natural uranium isotopic analysis in the presence of actinides ( 232Th, 237Np, 233Pa and 241Am)

The multi-group analysis (MGA) method for the determination of uranium isotopic abundances in depleted uranium (DU) and natural uranium (NU) samples is applied in this study. A set of non-destructive γ-ray measurements of DU and NU samples were performed using a planar Ge detector. The relative abundances of 235U and 238U isotopes were compared with the declared values of the standards. The relative abundance for 235U obtained by MGA for a “clean” DU or NU sample with a content of uranium >1 wt% is determined with an accuracy of about ±5%. However, when several actinides such as 232Th, 237Np, 233Pa and 241Am are present along with uranium isotopes simulating “dirty” DU or NU, it has been observed that MGA method gives erroneous results. The 235U abundance results for the samples were 6–25 times higher than the declared values in the presence of above-mentioned actinides, since MGA is utilized the X-ray and γ-ray peaks in the 80–130 keV energy region, covering XK α and XK β regions. After the least-squares fitting of the spectra, it is found that the increases in the intensities of the X-ray and γ-ray peaks of uranium are remarkably larger in the complex 80–130 keV region. On the other hand, it is observed that the interferences of the actinide peaks are relatively less dominant in the higher γ-ray region of 130–300 keV. The results imply the need for dirty DU and NU samples that the MGA method should utilize the higher energy γ-rays (up to 1001 keV of 234mPa) combined with lower energies of the spectra, which may be collected in a two detector mode (a planar Ge and a high efficient coaxial Ge).

Influence of energetic-driven “Taylor-Wave” shock-wave prestraining on the structure/property response of depleted uranium

The influence of shock prestraining, via direct energetic Taylor-wave (triangular wave) loading, on the post-shock structure/property behavior of depleted uranium (DU) was studied. Samples were shock prestrained within a soft shock recovery fixture composed of momemtum traps and a spall plate to assure 1-dimensional loading. The DU samples exhibit roughly a 30% increase in yield strength following shock prestraining to ∼45 GPa. The texture evolution in DU was quantified using electron-backscatter diffraction (EBSD). Detailed quantification of the substructure evolution following shock prestraining revealed high volume fractions of {130), '{172}', and {112) deformation twins. The volume fraction of (130) twins was found to be ∼10x the volume fraction of '{172}' and {112} twins. Details of the twin system activation and volume fraction relative to the local Schmidt factor within grains are presented. The influence of HE-driven shock prestraining on the structure/property response of DU is compared and contrasted to that seen in 304SS and 316SS subjected to Taylor-wave shock prestraining.

The production and oxidation of uranium nanoparticles produced via pulsed laser ablation

Depleted uranium samples were ablated using five nanosecond pulses from a Nd:YAG laser and produced films of ∼1600 Å thickness that were deposited with an angular distribution typical of a completely thermal ablation (cos 1 θ). The films remained contiguous for many months in vacuum but blistered due to tensile stress induced in the films several days after being brought into air. While under vacuum (2 × 10 −10 Torr base pressure) the films were allowed to oxidize from the residual gases, of which water vapor was found to be the primary oxidizer. During the oxidation, the samples were monitored with both X-ray and ultraviolet photoemission spectroscopy (XPS and UPS) and were found to oxidize following Langmuir kinetics. That a 2D-surface growth model described the oxidation indicates that, even at these low pressures, oxygen accumulation on the surface is a much faster process than diffusion into the bulk. While bulk diffusion did occur, the oxygen present at the surface saturated the measurements taken using photoemission and diffusion was difficult to observe. A method for determining oxide concentration via photoemission from the valence level, as opposed to the more conventional core levels, is also presented.

Oxidation of uranium nanoparticles produced via pulsed laser ablation

AbstractAn experimental apparatus designed for the synthesis, via pulsed laser deposition, and analysis of metallic nanoparticles and thin films of plutonium and other actinides was tested on depleted uranium samples. Five nanosecond pulses from a Nd:YAG laser produced films of ∼1600 Å thickness that were deposited showing an angular distribution typical of thermal ablation. The films remained contiguous for many months in vacuum but blistered due to induced tensile stresses several days after exposure to air. The films were allowed to oxidize from the residual water vapor within the chamber (2×10-10 Torr base pressure). The oxidation was monitored by in-situ analysis techniques including x-ray photoelectron spectroscopy (XPS), ultraviolet photoelectron spectroscopy (UPS), and scanning tunneling microscopy (STM) and followed Langmuir kinetics.

Quantification of Large Scale Micro-X-Ray Fluorescence Elemental Images

Niobium is commonly alloyed with uranium to prevent surface oxidation, and determining how the niobium concentration is distributed throughout a sample is useful in explaining observed material properties. The niobium concentration distribution was determined across the surface of depleted uranium samples using micro-X-ray fluorescence (MXRF). To date, MXRF has been employed primarily as a qualitative tool for determining relative differences in elemental concentrations across a sample surface. Here, a process was developed to convert qualitative MXRF niobium distribution images from depleted uranium samples into images displaying concentration values. Thus, MXRF was utilized to determine elemental concentrations across a surface in a manner similar to that of the established method of electron microprobe X-ray analysis (EMPA). However, MXRF can provide such information from relatively large sample areas many cm2 in size that are too large to examine by the higher spatial resolution technique of EMPA. Although the sample surfaces were polished to the same degree as the standards, little or no sample preparation should be necessary for sample systems where a high energy analyte XRF line can be used for imaging.

Determination of actinide isotope ratios using glow discharge optogalvanic spectroscopy

Diode laser-excited optogalvanic spectroscopy (OGS) of a glow discharge has been utilized to measure U-235/U-235+U-238 isotope ratios. This `optical mass spectrometric' measurement has been demonstrated for a number of samples including uranium oxide, fluoride, and metal. Various diode laser-accessible atomic transitions in the 775–835 nm region have been evaluated; these transitions were chosen by considering OGS sensitivity and isotope shift. Using the 831.84 nm uranium line, for example, it was possible to measure the U-235/U-235+U-238 isotope ratio (0.0026) of depleted uranium samples. A prototypical field instrument to make these measurements has been assembled and demonstrated. A U-236 spectral line was identified in an enriched uranium sample, and an abundance was measured.

A rapid method for determination of the isotopic composition of uranium samples by alpha spectrometry

A simple method of analyzing alpha spectra from natural and enriched or depleted uranium samples is developed. The procedure is non-iterative, and takes into consideration low-energy tail and branching-ratio corrections to accurately calculate the area corresponding to each uranium isotope (234U, 235U, 236U, 238U) in the spectrum, and then the isotopic composition of the sample. A BASIC computer program, called ENURA, has been developed to perform all the necessary calculations to give the results together with their uncertainties. Several samples were prepared with different uranium concentrations made from standard solutions with known compositions, and the method was checked against the experimental measurements from these samples. Other series of uranium spectra were theoretically constructed using a given line shape in order to cover the required range of enriched or depleted uranium.

Application of the fission track registration technique in the estimation of fissile materials: 235U-content in natural and depleted uranium samples and total uranium in solutions

The results of an investigation on the estimation of the 235U content in natural and depleted uranium samples and total uranium in solutions including highly active fuel reprocessing waste streams using the solid state track registration technique are described. Two different methods were used: the “dry” method in which the detector (Lexan strip) was in contact with the fissile material in a 2 π geometry and the “wet” method in which the Lexan detector strips were kept immersed in the solution of the fissile material. The relative merits and demerits and also the accuracy and precision attainable by the two methods are discussed. The feasibility of using a portable neutron source for the estimation of the 235U content was demonstrated and some of the practical problems encountered and the way in which they were overcome are discussed.