Nuclear Waste Drums Research Articles

Reconstruction of finer voxel grid transmission images in Tomographic Gamma Scanning

Tomographic Gamma Scanning (TGS) is a technique used to assay the nuclide distribution and radioactivity in nuclear waste drums. Limited to few transmission measurements, the drum is divided into large voxels and thus leads to the inhomogeneity of the voxels and negatively affects the results. A new algorithm is presented to reconstruct finer voxel grid transmission images with the same number of measurements. The small voxel size decreases the effect of the inhomogeneity and makes the result more accurate. The algorithm employs total variation minimization and precisely describes the attenuation process. The influences of the different scan modes are discussed with Monte Carlo simulations. Experiments are performed to verify the effectiveness of our method.

An Attenuated Projector for Iterative Reconstruction Algorithm of a High Sensitivity Tomographic Gamma Scanner

Tomographic Gamma Scanners are tools for nondestructive assay and characterization of nuclear waste drums. In these scanners, a three dimensional image of the activity distribution of every radioisotope stored in the drum is obtained by performing a single-photon emission tomography. AR-TGS is a novel architecture of tomographic gamma scanners that combines an HPGe detector with six NaI(Tl) detectors in order to achieve high-sensitivity. In this work, a projector for a 2D MLEM reconstruction algorithm of AR-TGS is presented. This projector models the geometry of the system, the collimator's response and the attenuation in the field of view by performing a ray-tracing with several lines of response per detector. The projector was evaluated with Monte Carlo simulations of different phantoms and with experimental measurements. The algorithm proved to be an accurate model of the acquisition process and was used to reconstruct data sets with different strategies, such as utilising matched and unmatched projector/backprojector pairs. The results showed that the use of this projector in image reconstruction considerably improved spatial resolution and image quality compared with an attenuated Siddon projector. The quantification properties of the algorithm for homogeneous and heterogeneous drums matrices were also analyzed.

Collimated LaBr3 detector response function in radioactivity analysis of nuclear waste drums

The properties of a Lanthanum bromide(LaBr3)detector and its response functions were investigated via experiments and simulations in this paper.The LaBr3detector had good relative energy resolution and higher efficiency than a high-purity germanium detector.Monte Carlo and other numerical methods were used to calculate the efficiencies of a LaBr3detector with a square collimation window.A model of the numerical method was established based on a pure geometric model that was consistent with the experimental situation.The results showed that the detector response functions calculated by these methods were in great agreement with experimental results.

Analytical calculation of the collimated detector response for the characterization of nuclear waste drums by segmented gamma scanning

Improved methods for the reconstruction of the isotope specific activity content in nuclear waste drums with data obtained by a gamma scanning system developed at Shanghai Jiao Tong University require an analytical function of the detector response. In this work we derive an analytical detector response function for a collimated HPGe detector with a square collimation window. The model is based on a purely geometric model respecting the configuration of the collimated detector system, the positions of radioactive point sources and the absorption of γ-rays in the matrix as well as in the HPGe crystal. We show that the derived analytical detector response function is in good agreement with data simulated by MCNP5.

Reconstruction of the isotope activity content of heterogeneous nuclear waste drums

Radioactive waste must be characterized in order to verify its conformance with national regulations for intermediate storage or its disposal. Segmented gamma scanning (SGS) is a most widely applied non-destructive analytical technique for the characterization of radioactive waste drums. The isotope specific activity content is generally calculated assuming a homogeneous matrix and activity distribution for each measured drum segment. However, real radioactive waste drums exhibit non-uniform isotope and density distributions most affecting the reliability and accuracy of activities reconstruction in SGS. The presence of internal shielding structures in the waste drum contributes generally to a strong underestimation of the activity and this in particular for radioactive sources emitting low energy gamma-rays independently of their spatial distribution.In this work we present an improved method to quantify the activity of spatially concentrated gamma-emitting isotopes (point sources or hot spots) in heterogeneous waste drums with internal shielding structures. The isotope activity is reconstructed by numerical simulations and fits of the angular dependent count rate distribution recorded during the drum rotation in SGS using an analytical expression derived from a geometric model. First results of the improved method and enhancements of this method are shown and are compared to each other as well as to the conventional method which assumes a homogeneous matrix and activity distribution. It is shown that the new model improves the accuracy and the reliability of the activity reconstruction in SGS and that the presented algorithm is suitable with respect to the framework requirement of industrial application.

Reconstruction of the activity of point sources for the accurate characterization of nuclear waste drums by segmented gamma scanning

This work improves the reliability and accuracy in the reconstruction of the total isotope activity content in heterogeneous nuclear waste drums containing point sources. The method is based on χ(2)-fits of the angular dependent count rate distribution measured during a drum rotation in segmented gamma scanning. A new description of the analytical calculation of the angular count rate distribution is introduced based on a more precise model of the collimated detector. The new description is validated and compared to the old description using MCNP5 simulations of angular dependent count rate distributions of Co-60 and Cs-137 point sources. It is shown that the new model describes the angular dependent count rate distribution significantly more accurate compared to the old model. Hence, the reconstruction of the activity is more accurate and the errors are considerably reduced that lead to more reliable results. Furthermore, the results are compared to the conventional reconstruction method assuming a homogeneous matrix and activity distribution.

An Analytical Simulation to Calibrate Gamma-Ray Cylindrical Detectors for Environmental and Radioactive Waste Measurements

The calibration of gamma-ray cylindrical detectors is often required in the analysis of high or low environmental samples and the homogenously distributive nuclear waste drums. Therefore, a new analytical simulation method is proposed in the present work to calculate the full-energy peak efficiencies of high-purity germanium cylindrical detectors using extended sources of low and high volumes. The sources were mounted at three different positions with respect to the detector’s axis (coaxial, parallel, and perpendicular), labeled as Position 1, Position 2, and Position 3, respectively. The self-attenuation and the coincidence summing effects at low source-detector distance are also included in the algorithm. A remarkable agreement between the measured and the calculated efficiencies is achieved with discrepancies <4% for the first two positions and between 5 and 7% for the last one.

Photofission for the detection of actinides

Nuclear waste management cost at CEA is a heavy budget share. Therefore, it is important, in order to minimize that cost, to select the most suitable repository for the waste. Significant cost savings are obtained, for example, whenever a medium activity waste drum initially filed to end up in a geological repository can be measured accurately enough to be accepted by the Centre de Stockage de l’Aube (CSA). Photofission can be used as a non-destructive technique to measure the actinide content embedded in a nuclear waste drum. It can even be applied when the waste is packaged in its heavy concrete shielding. It is equally sensitive to non-fissile nuclear material. However, the present state of art suffers from a lack of sensitivity to be able to help redirecting the waste to the proper repository. A new program called INPHO is proposed at CEA Saclay aiming at enhancing the sensitivity sufficiently to be applicable for waste characterization.

A Stochastic Problem Arising in the Storage of Radioactive Waste

Nuclear waste drums can contain a collection of radioactive components of uncertain activity and randomly dispersed in position. This implies that the dose-rate at the surface of different drums in a large assembly of similar drums can have significant variations according to the physical makeup and configuration of the waste components. The present paper addresses this problem by treating the drum, and its waste, as a stochastic medium. It is assumed that the sources in the drum contribute a dose-rate to some external point. The strengths and positions are chosen by random numbers, the dose-rate is calculated and, from several thousand realizations, a probability distribution for the dose-rate is obtained. It is shown that a very close approximation to the dose-rate probability function is the log-normal distribution. This allows some useful statistical indicators, which are of environmental importance, to be calculated with little effort.As an example of a practical situation met in the storage of radioactive waste containers, we study the problem of “hotspots.” These arise in drums in which most of the activity is concentrated on one radioactive component and hence can lead to the possibility of large surface dose-rates. It is shown how the dose-rate, the variance, and some other statistical indicators depend on the relative activities on the sources. The results highlight the importance of such hotspots and the need to quantify their effect.

Nondestructive waste-drum assay for transuranic content by gamma-ray active and passive computed tomography

Abstract A gamma-ray-based, active (A) and passive (P) computed tomography (CT) technology has been developed that locates, identifies and quantifies gamma-ray emitting isotopes, transuranic (TRU) and others, in nuclear waste drums. ACT uses a collimated external source and a HPGe detector to measure selected mono-energetic gamma-rays that are attenuated by waste-drum contents; a separate PCT measurement uses the HPGe detector to record the spectra of gamma-rays emitted from within a drum. The ACT attenuation images and the PCT emission spectra are coupled to quantitatively assay drum contents for ∼0.1–200 g of TRU isotopes. Calibration requires a single measurement of a known radioactive standard; construction of waste-drum surrogates is not required. Fixed and mobile systems demonstrated compliance with a DOE quality assurance program via several independent blind tests.

Kernel-Integration Scatter Model for the Characterization of Nuclear Waste Drums by Gamma Emission Tomography

The objective of this study is the determination of the distribution of the activity of radio-elements contained in radioactive waste packages by means of single photon emission computed tomography (SPECT). A three- dimensional (3D) projector simulator for a parallel hole collimator and NaI(Tl) scintillator, based on the point kernel integration method, is proposed. The model takes into account the attenuation and scatter of gamma rays. Primary scattered photons are treated by the Compton process and Klein-Nishina formula, and the multiple scattering is accounted for by means of the dose buildup factor normally used in shielding problems. An advantage of the proposed model is that it offers the possibility of generating a full two-dimensional point-spread function (PSF) that can be used for 3D reconstruction. The developed model is convenient in those situations where more exact techniques (such as Monte Carlo simulation) are not economical. The model has been evaluated for the package having homogenous density with the Cs-137 point source inside.

Estimation of the matrix attenuation in heterogeneous radioactive waste drums using dual-energy computed tomography

Gamma spectroscopy measurements of the activity of radionuclides in nuclear waste drums must be corrected for the attenuation due to the non-homogeneous waste matrix. The attenuation factors depend on the matrix local density and effective atomic number, and on the energy of the gamma rays emitted by the radionuclides. The requirements for the system presented in this paper are to estimate the attenuation in low-density (<0.4 g/cm 3), 120 l drums containing radionuclides emitting in the (59.5 keV, 1.4 MeV) energy range. A series of three-dimensional (3D) attenuation maps of the drum are computed using a dual-energy computerized tomography (DE-CT) system with an external, polychromatic X-ray source. The system successively records low-energy (mean energy about 62 keV) and high-energy (about 300 keV) projections using different tube voltages, anode current, and filtration. Each projection is acquired by 22 BGO scintillators – PM detectors in fan-beam geometry. The drum is rotated and elevated in a helical scan. A DE calibration transforms pairs of DE projections into pairs of “equivalent basis materials (BM)” projections. This non-linear transformation allows to correct for polychromaticity. After reconstruction, the two “equivalent BM” 3D maps are used, together with tabulated attenuation data of the BMs, in order to extrapolate the 3D attenuation map at any energy peak. Maps of the mass density and of the effective atomic number can also be computed. The total examination time is less than 5 min. Experimental images are shown.

Development of a Numerical Experiment Technique to Solve Inverse Gamma-Ray Transport Problems with Application to Nondestructive Assay of Nuclear Waste Barrels

A numerical experimental technique is presented to find an optimum solution to an undetermined inverse gamma-ray transport problem involving the nondestructive assay of radionuclide inventory in a nuclear waste drum. The method introduced is an optimization scheme based on performing a large number of numerical simulations that account for the counting statistics, the nonuniformity of source distribution, and the heterogeneous density of the self-absorbing medium inside the waste drum. The simulation model uses forward projection and backward reconstruction algorithms. The forward projection algorithm uses randomly selected source distribution and a first-flight kernel method to calculate external detector responses. The backward reconstruction algorithm uses the conjugate gradient with nonnegative constraint or the maximum likelihood expectation maximum method to reconstruct the source distribution based on calculated detector responses. Total source activity is determined by summing the reconstructed activity of each computational grid. By conducting 10 000 numerical simulations, the error bound and the associated confidence level for the prediction of total source activity are determined.The accuracy and reliability of the simulation model are verified by performing a series of experiments in a 208-l waste barrel. Density heterogeneity is simulated by using different materials distributed in 37 egg-crate-type compartments simulating a vertical segment of the barrel. Four orthogonal detector positions are used to measure the emerging radiation field from the distributed source. Results of the performed experiments are in full agreement with the estimated error and the confidence level, which are predicted by the simulation model.

Application of nuclear analytical techniques using long-life sealed-tube neutron generators.

The new range of sealed-tube neutron generators developed by SODERN appears to be appropriate for the industrial environment. The main characteristics are the high emission stability during the very long lifetime of the tube, flexible pulsed mode capability, safety in operation with no radiation in "off" state, and the easy transportation of equipment. Some applications of the neutron generators, called GENIE, are considered: high-sensitivity measurement of transuranic elements in nuclear waste drums, bulk material analysis for process control, and determination of the airborne pollutants for environmental monitoring.

Industrial analysis possibilities using long-life sealed-tube neutron generators

Long-life sealed tube neutron generators appear very attractive for use in industrial systems, because of their modulation and their high neutron output capabilities. SODERN is now able to put on the market different types of transportable neutron generators, designed for intensive use, called GENIE. The main potential applications are: high sensitivity measurement of transuranic elements in nuclear waste drums, in situ measurements for oil and mineral logging, process control for bulk material or in metallurgy, hidden explosives detection, neutron radiography, etc. ... The present paper shows some examples of how to use such generators most effectively.