Radioactive Waste Drums Research Articles

Uranium quantification in radioactive waste drums using In-Situ Object Counting System (ISOCS) for safeguards purposes

For the peaceful use of nuclear energy, materials, and technology, various methods have been developed for safeguards purposes. Among them, the In-Situ Object Counting System has demonstrated its usefulness and reliability in various nuclear material verification activities, due to its capability of operating without standard materials. In the present study, the performance of In-Situ Object Counting System for uranium quantification was evaluated using an experimental drum at KEPCO NF as well as low-level radioactive waste drums at the Korea Radioactive Waste Agency. The results showed that In-Situ Object Counting System could quantify uranium in radioactive waste drums, with a relative error ranging from −1 to −67 % compared with destructive assay results. This study also examined, using the Monte Carlo N-particle transport 6.2 code, the factors that affect uncertainty in uranium quantification: geometrical conditions, spectrum quality, heterogeneity, and the self-shielding effect. These findings could prove useful in efforts to optimize the In-Situ Object Counting System as a tool for nuclear material verification in the safeguards context.

Feasibility of a position-sensitive Cherenkov radiation detector using a reflector-coated liquid light guide

In this study, a Cherenkov radiation sensor was developed utilizing a reflector-coated liquid light guide (LLG). In the proposed configuration, a reflector replaced one of the two photomultiplier tubes (PMTs), forming the core sensing element alongside the LLG, with gamma-ray (60Co and 137Cs) and a beta-ray (90Sr) emitters employed to assess its performance as a position-sensitive detector. The time-of-flight method was utilized to determine time differences corresponding to source positions. Position spectra were generated based on these time differences, with the Cherenkov radiation count rates compared across different types of reflectors to identify the optimal choice. We compared position spectra obtained from two PMTs and one PMT with a reflector, confirming the ability to detect multiple sources. Subsequently, each source was positioned perpendicular or parallel to the sensor axis to capture position spectra. The results demonstrate the sensor's potential as a long-distance position-sensitive detector. By leveraging the simplicity of a single PMT system, the proposed sensor facilitates challenging tasks such as pipeline inspection or non-destructive testing of radioactive waste drums.

A SGS efficiency calibration method for measuring the radioactive waste drum based on Monte Carlo simulation and function model

A SGS efficiency calibration method for measuring the radioactive waste drum based on Monte Carlo simulation and function model

Matrix effect correction in active neutron interrogation for residual fissile mass assessment in radioactive waste drums

Metallic waste in the form of shells and nozzles remains following the reprocessing of spent fuel at La Hague plant (France). Before the final disposal, drums containing the waste are transported to the compaction facility where its volume is reduced by a factor of 5. With the objective of controlling the criticality/safety levels, an active neutron interrogation system at the entrance of the facility is used to assess the residual fissile mass remaining in the waste, which relies on prompt fission neutron detection (known as the differential die-away technique, DDA). The measured signal is proportionally linked to the fissile mass by a calibration coefficient. However, two effects can produce an inaccurate prediction. The number of induced fissions for the same fissile mass varies according to the waste matrix composition, particularly the neutron slow-down and absorption ratio. Secondly, the presence of fissile clusters can impact the neutron interrogation due to an increase in self-shielding effect. This work presents a matrix effect correction method based on the use of internal flux monitors (sensitive to the neutron absorption ratio) and the transmission signal (sensitive to the slow-down ratio). It relies on a numerical model of the whole measurement cell developed with the Monte-Carlo N-Particle transport (MCNP) code. The calibration coefficient of 72 different matrices, representative of the waste produced at La Hague, were simulated and a statistical model using a multilinear regression was then established. This simulated-based approach provides a more robust and comprehensive estimation of the residual fissile mass without impacting accuracy thanks to the flux monitors and the transmission signals.

Enhanced radioactive waste drum monitoring: A sensorized LoRa-based network for identification and integrity assessment

A reliable monitoring and oversight system is essential to guarantee the safe operation and eventual closure of radioactive waste disposal facilities. With recent technological advancements, there is now the capability to deploy distributed wireless sensor networks that are cost-effective and low in power consumption. Such networks are versatile, finding applications in several civil, environmental, and industrial sectors. This work presents a novel system and method for identifying and continuously monitoring the integrity of stored radioactive waste packages. The proposed approach, developed by the University of Pisa in the context of the H2020 European Project PREDIS, employs a network of wireless node devices, provided with solid-state micro-power radiation detectors tasked with the measurement of gamma-rays and thermal neutrons coming from within cemented waste drums. Collected data are automatically transmitted across the storage site via Long Range radio technology and forwarded to cloud-based or offline platforms, allowing for further visualization, data processing, and safe storage. The system highlights the feasibility of autonomously and passively gathering data from radioactive waste drums over extended distances (more than 250 m in non-line-of-sight conditions), decreasing the reliance on portable or automated scanning mechanisms. In line with its capability for autonomous, long-range data collection, the system satisfies the requirement for structural integrity, containment, and confinement of packaging throughout both interim and long-term disposal of radioactive wastes, thereby playing a key role in reinforcing the safe management of radioactive waste.

A density correction method for radioactive waste drum based on SRGS technology

The segmented ringed gamma scanning (SRGS) technique represents an advancement in segmented gamma scanning (SGS) technology used for detecting the density of radioactive waste drums, offering enhanced measurement accuracy. However, significant occur errors in the reconstruction of matrix densities due to the non-uniform distribution of density in radioactive waste and the conical beam emitted from the transmission source collimator. This paper proposes a density correction method based on dichotomy to address this issue. The efficacy of this method was verified through both simulations and experiments on a sample containing five different materials, utilizing 137Cs and 60Co for transmission and emission measurements, respectively. The experimental results demonstrate that the errors in the corrected matrix densities are reduced, falling within a margin of 16.8%. Additionally, the corrected reconstruction error of the activity is approximately 25% of the uncorrected results.

A Wireless Gamma-Ray Monitoring System for Cemented Radwaste Drums.

In the framework of the PREDIS EU project, a wireless battery-operated gamma-ray detection system was developed in order to provide a medium-to-long-term monitoring system for radioactive waste drums. It was initially proposed to monitor the gamma radioactivity outcoming from steel drums containing cemented radwaste, even though it could be usefully employed in a wider range of applications. Gamma rays are penetrating and convey information from the drum's internal structure, as the count rate measured on the surface depends on the thickness and density of the crossed materials. A number of sensors arranged around a drum, typically four units, provide indications of the emission anisotropy, and any sensitive change in the measured count rate would hint at some anomaly, thus triggering a suitable inspection by operators.

An efficiency calibration method of segmented gamma scanning in reconstructing radioactive waste drum activity

An efficiency calibration method of segmented gamma scanning in reconstructing radioactive waste drum activity

Application of Machine Learning to MICADO passive and active neutron measurement system for the characterization of radioactive waste drums

Application of Machine Learning to MICADO passive and active neutron measurement system for the characterization of radioactive waste drums

Radiation safety assessments for radioactive concrete recycling workers and residents

Carried out was the assessment of the radiation safety for the recycling workers of the radioactive concrete generated from nuclear power plants and nearby residents. Decommissioning of a nuclear power plant with an electric output of 1000 MW generates approximately 500,000 tons of concrete waste. It is important to reduce the volume of concrete waste to reduce the costs of decommissioning projects. For this purpose, the low-level radioactive cement powder from bioshield concrete is recycled as a solidifying agent. The number of radioactive concrete waste drums to be disposed is reduced by extracting cement from the radioactive concrete and using it as a solidifying agent, rather than using cement from general industrial waste. Various exposure scenarios exist, and it is necessary to evaluate the radiation dose in each scenario. In this study, seven scenarios were selected and evaluated. The radioactive exposure to facility workers, nuclear power plant workers, and landfill management workers is within the annual dose limit, even if they work 2000 h per year. A loader can work close to the drum for 90–93 h per year, and a truck driver can drive 621 one-way trucks per year. Dose evaluations for road users and residents after landfill closure showed that both had values far below the dose limit for the public. Bioshield concrete recycling was confirmed to be radiologically safe in all scenarios if certain conditions were satisfied. Thus, this study showed that cement powder from bioshield concrete could be used safely without exceeding the radiation safety limits for workers and the public when recycling concrete.

An efficiency function model of segmented gamma scanning for measuring radioactive waste drum

Segmented gamma scanning (SGS) is a fast and effective method for measuring radioactive waste drum. The efficiency calibration is directly related to the accuracy of reconstructed radioactivity. An efficiency function model and SGS efficiency calibration method are proposed for solving existing SGS efficiency calibration problems such as time lag, limited by experimental sources or difficult to effectively combine with SGS system. The SGS system model is established by the Geant4 to calculate the segment efficiency under different linear attenuation coefficients of medium and gamma energies. The efficiency calibration function is established with the function model and parameters. Waste drum samples are constructed with the polyethylene and point sources 137Cs/60Co to complete SGS experimental measurement, efficiency calibration and radioactivity reconstruction. The result shows that the relative deviation of the reconstructed activity of a single point source at different locations in the drum is −50.48% to 43.69% and it of multi-point sources in a segment or a drum is −27.88% to 3.57%. Experimental results verify the effectiveness of this efficiency function model and SGS calibration method.

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.

Improving Bayesian radiological profiling of waste drums using Dirichlet priors, Gaussian process priors, and hierarchical modeling

We present three methodological improvements of our recently proposed approach for Bayesian inference of the radionuclide inventory in radioactive waste drums, from radiological measurements. First we resort to the Dirichlet distribution for the prior distribution of the isotopic vector. The Dirichlet distribution possesses the attractive property that the elements of its vector samples sum up to 1. Second, we demonstrate that such Dirichlet priors can be incorporated within an hierarchical modeling of the prior uncertainty in the isotopic vector, when prior information about isotopic composition is available. Our used Bayesian hierarchical modeling framework makes use of this available information but also acknowledges its uncertainty by letting to a controlled extent the information content of the indirect measurement data (i.e., gamma and neutron counts) shape the actual prior distribution of the isotopic vector. Third, we propose to regularize the Bayesian inversion by using Gaussian process (GP) prior modeling when inferring 1D spatially-distributed mass or, equivalently, activity distributions. As of uncertainty in the efficiencies, we keep using the same stylized drum modeling approach as proposed in our previous work to account for the source distribution uncertainty across the vertical direction of the drum. A series of synthetic tests followed by application to a real waste drum show that combining hierarchical modeling of the prior isotopic composition uncertainty together with GP prior modeling of the vertical Pu profile across the drum works well. We also find that our GP prior can handles both cases with and without spatial correlation. Of course, our GP prior modeling framework only makes sense in the context of spatial inference. Furthermore, the computational times involved by our approach are on the order of a few hours, say about 2, to provide uncertainty estimates for all variables of interest in the considered inverse problem. This warrants further investigations to speed up the inference.

A novel 3D-imaging and characterisation technique for special nuclear materials in radioactive waste

A novel technique for non-destructive assay (NDA) of radioactive waste called ARCTERIX (Advanced Radwaste Characterisation based on Tomographically Enhanced Radiation Imaging without X-rays) is presented. The concept is based on a 3D-tomographic imaging technique for special nuclear materials – neutron-gamma emission tomography (NGET). ARCTERIX takes the NGET principle from its original application area of nuclear security systems into the realm of radioactive waste assay with its special characteristics and challenges. By adding localisation and imaging of SNM inside shielded waste containers to the array of existing techniques used for radioactive waste characterisation, ARCTERIX complements the state of the art in passive and active NDA interrogation methods. It is aimed primarily at the class of mixed, long-lived radioactive waste that is commonly called “legacy” or “historic” waste which has special safety, security and safeguards concerns due to its mixed composition, commonly poor documentation, and the frequent presence of SNM. The ARCTERIX concept provides rapid imaging and characterisation of nuclear materials in radioactive waste with a high degree of automation and high throughput capabilities, making it possible to quickly scan large radioactive waste inventories for the presence of special nuclear materials with minimal manual intervention. The first ARCTERIX prototype system has demonstrated a high technological readiness for the implementation of the technique in a commercial stand-alone system for rapid assessment of radioactive waste drums or in a system operating in conjunction with established techniques.

Radioactive waste repackaging simulation based on MATLAB/Simulink

This study implemented a packaging and drum generation process for radioactive waste using Matlab/Simulink. The simulation was based on the regulations of field work, and aimed to derive the optimal repackaging method by adjusting the constraints. We analyzed the radioactive waste nuclide concentration, weight, and small package amount as constraints and also analyzed the number of drums for each radioactive level generated according to the scenarios. The radioactive waste repackaging process simulation was designed to analyze cases according to the confidence interval, and the optimization algorithm to reduce the number of radioactive waste drums was analyzed while checking the change in the number of drums according to the repackaging scenario. Simulation analysis can be used for radioactive waste management and analysis, decision-making, and generation prediction, and through systematic management, the safety and economic efficiency of radioactive waste management can be expected.

Database of virtual spectrum of artificial radionuclides for education and training in in-situ gamma spectrometry

As the field of application of in-situ gamma spectroscopy is diversified, proficiency is required for consistent and accurate analysis. In this study, a program was developed to virtually create gamma energy spectra of artificial nuclides, which are difficult to obtain through actual measurements, for training. The virtual spectrum was created by synthesizing the spectra of the background radiation obtained through actual measurement and the theoretical spectra of the artificial radionuclides obtained by a Monte Carlo simulation. Since the theoretical spectrum can only be obtained for a given geometrical structure, representative major geometries for in-situ measurement (ground surface, concrete wall, radioactive waste drum) and the detectors (HPGe, NaI(Tl), LaBr3(Ce)) were predetermined. Generated virtual spectra were verified in terms of validity and harmonization by gamma spectrometry and energy calibration. As a result, it was confirmed that the energy calibration results including the peaks of the measured spectrum and the peaks of the theoretical spectrum showed differences of less than 1 keV from the actual energies, and that the calculated radioactivity showed a difference within 20% from the actual inputted radioactivity. The verified data were assembled into a database and a program that can generate a virtual spectrum of desired condition was developed.

A detailed design for a radioactive waste safety management system using ICT technologies

A Radioactive Waste Information Management System (RAWINGS) currently in operation mainly manages the inventory and history of the operating waste. The system has the disadvantages of the entered information needing to be transferred manually from the site to the system, information getting incorrectly entered during the process or information going missing. Recently, the Nuclear Safety and Security Commission (NSSC) and Korea Radioactive Waste Agency (KORAD) called for the development of a digital system that can show information transparently in real-time regarding the preliminary inspections of RAdioactive Waste (RAW) and the assessment of its suitability for disposal before the radioactive waste is delivered to the disposal site. A Digital Twin (DT)system is being developed for the safety management of radioactive waste to address the problems that these systems have and meet the needs of disposal operators. This paper introduces the DT technology that uses Augmented Reality (AR) technology enabling users to check the contents of small-packaged wastes in radioactive waste drums without opening them, Internet of Things (IoT) sensor technology that checks the status of the drums in the radioactive waste storage and the RAWINGS system. Based on the performance of a prototype Digital Twin consisting of three modules (AR, IoT and RAWINGS), the augmented reality enables users to see the shape information and filling rate of small-packaged wastes in the radioactive waste drums and includes Quick Response (QR) code management. The basic data of the radioactive waste used in the augmented reality, as well as small packaged wastes and repackaged drums, were processed in conjunction with RAWINGS. In addition, real-time monitoring of radioactive waste drums loaded in the designated space (Y zone: an area where combustible waste is loaded within radioactive waste storage and TEST area: a section where drums scheduled to be transported to the disposal site are loaded) of the radioactive waste storage was possible by transmitting IoT sensor signals attached to the drum to the digital twin. Currently, augmented reality has an important role in enhancing the visibility and intuitiveness of radioactive waste information for radioactive waste managers and workers by overlapping digital information about radioactive waste storage. Due to the nature of radioactive waste, it is difficult to know what waste is inside the enclosed drum. However, the results of this study confirmed that waste contained in radioactive waste drums can be identified in real time in the Digital Twin rather than in the radioactive waste storage.This technology will be useful in determining the conformity of the radioactive waste acceptance criteria required by KORAD before the delivery of radioactive waste drums to disposal sites.

Detection of Fission Coincidences With Plastic Scintillators for the Characterization of Radioactive Waste Drums

This work addresses the use of plastic scintillators as an alternative to ³He neutron detectors for radioactive waste drum characterization. The time response of scintillators is three orders of magnitude faster than that of gas proportional counters, and they offer similar neutron detection efficiency at lower cost. However, they are sensitive to gamma rays and the neutron-gamma pulse shape discrimination (PSD) is not possible with standard PVT scintillators. The proposed approach uses the neutron and gamma times of flight in triple coincidences recorded with ²⁵²Cf, AmBe, and ⁶⁰Co sources. A 2-D histogram of time delays between the second and first detected pulses, on the <inline-formula> <tex-math notation="LaTeX">$x$ </tex-math></inline-formula>-axis, and between the third and second pulses, on the <inline-formula> <tex-math notation="LaTeX">$y$ </tex-math></inline-formula>-axis, evidences a specific region of interest for spontaneous fission coincidences. MCNPX-PoliMi simulations are performed, which are in good agreement with previous experiments and allow for investigating the types of coincidences (<inline-formula> <tex-math notation="LaTeX">$\gamma \gamma \gamma $ </tex-math></inline-formula>, nnn, <inline-formula> <tex-math notation="LaTeX">$\gamma $ </tex-math></inline-formula>nn, <inline-formula> <tex-math notation="LaTeX">$\gamma \gamma \text{n}$ </tex-math></inline-formula>) and optimizing the rejection of neutron and gamma scattering crosstalk. The method was also experimentally tested with a 118-L mock-up drum filled with a metallic or an organic matrix, showing a correct estimation of the net fission signal up to an &#x201C;alpha ratio&#x201D; of 12 between fission and (<inline-formula> <tex-math notation="LaTeX">$\alpha $ </tex-math></inline-formula>,n) neutron emissions. Matrix and localization effects were also measured, showing a sensitivity (useful signal per gram of ²⁴⁰Pu_eq) for a homogeneous distribution of plutonium in the iron matrix about three times larger than in the wood matrix, due to neutron slowing-down in the latter. This difference can be taken into account by the prior knowledge of the matrix characteristics or by neutron and gamma attenuation measurements. In contrast, in case of a point source with an unknown position in the matrix, the relative localization uncertainty is 26% for the metallic drum, and 41% for the organic one.

Boron-Coated Straws Imaging Panel Capability for Neutron Emission Computed Tomography for Source Localization Inside Radioactive Drums

The evaluation of fissile mass inside radioactive waste drums by neutron measurements is essential for radioactive waste management, nuclear safety, and criticality issues. However, passive and active neutron measurements can be strongly impacted by the uncertainty on the neutron source position within the drum and by the matrix effects. Therefore, an imaging panel proposed by Proportional Technologies Inc., and composed of seven boron-coated straw (BCS) detectors has been tested to localize neutron interactions, in view to reduce uncertainties associated with plutonium or uranium position inside radioactive waste drums. In a previous work, a numerical model of the imaging panel has been developed and validated from a comparison with experimental profiles obtained with a ²⁵²Cf source. In the first section, the feasibility of neutron emission tomography by a setup composed of five extended imaging BCS panels is demonstrated by numerical Monte Carlo simulation. The second section details the experimental validation of the neutron emission tomography. Measurements are carried out with AmBe and ²⁵²Cf sources located inside an empty 118 L drum by rotating the BCS imaging panel around it. Afterward, the Richardson&#x2013;Lucy deconvolution algorithm is applied to provide 2-D neutron source images for each angle. Finally, the 3-D images are reconstructed using the reconstruction toolkit (RTK) backprojection projection routine. The results demonstrate the capability of the BCS imaging to provide the 3-D location, i.e., axial and radial positions of one and two neutron sources. Furthermore, the first tests with this passive neutron measurement system show a satisfactory 3-D reconstruction for ²⁵²Cf and AmBe sources separated by 20 cm. Consequently, BCS imaging panels open interesting prospects to reduce the uncertainty associated with plutonium or uranium localization in neutron measurements. Work is undergoing to assess the capability of this system for 118 L drums filled with organic and metallic matrices. Additionally, further prospects concern the performance of other deconvolution and reconstruction algorithms.

Transmission reconstruction algorithm by combining maximum-likelihood expectation maximization and a convolutional neural network for radioactive drum characterization

One of the most critical factors that affect reconstructions of the activity of a radioactive drum is the accuracy of tomographic gamma scanning transmission reconstruction. The traditional algorithms applied for reconstructing the density map, such as maximum-likelihood expectation maximization (MLEM), algebraic reconstruction technique, or filtered back-projection, produce a grid distribution with severe grid artifacts and a high level of noise, which significantly degrade the detail of the transmission image, thereby increasing the reconstruction error for both the density map and the activity. Thus, we propose a novel algorithm for transmission reconstruction by combining MLEM and a deep convolutional neural network (CNN). The CNN is trained using a supervised learning approach with image pairs obtained from the drum's ground truth (target) and the result constructed using MLEM (input). Our experimental results indicate that the proposed reconstruction algorithm can significantly improve the spatial resolution while also removing grid artifacts. In addition, the algorithm is sufficiently robust when dealing with a noisy input image. When the input image contained 20% noise, the mean squared error of the output image decreased by 78.51% compared with the conventional reconstruction method, and the peak signal-to-noise ratio and structural similarity index measure for the output image improved by 81.19% and 71.74%, respectively. The new algorithm improves the accuracy of the radioactive drum density map reconstruction as well as decreasing the measurement time required. We consider that the proposed algorithm is an effective new method for use in the field of radioactive waste drum transmission image reconstruction.