Characterization Of Nuclear Materials Research Articles

Uranium measurements in the field using high-resolution cadmium zinc telluride detectors

A new generation of cadmium zinc telluride (CZT) detectors has become available and is being evaluated by the International Atomic Energy Agency (IAEA) for safeguards verifications in the field. The new CZT detector, model M400, is a room temperature spectrometer manufactured by H3D, Inc. The M400 demonstrates superior energy resolution, effective isotope identification capabilities, and convenient usability features when tested in a controlled laboratory environment. These characteristics define the M400 as a potential platform for IAEA field detection applications, which could become suitable for nuclear material characterization (e.g., enrichment verification) and nuclear safeguards missions. The capabilities of gamma spectrometry codes including Fixed energy, Response function Analysis with Multiple efficiencies (FRAM) from Los Alamos National Laboratory, CZT for Uranium (CZTU) from Lawrence Livermore National Laboratory, and Gamma Detector Response and Analysis Software (GADRAS) from Sandia National Laboratories were adapted for M400 spectra, and the performance of the codes has been validated. This was reported in a prior work. To further validate the performance of the high-energy resolution CZT detector and the isotopic analysis codes, a field measurement campaign consisting of uranium hexafluoride (UF6) cylinder measurements was conducted at a fuel fabrication facility. A total of 34 Type 30B cylinders containing UF6 were measured using three different M400 CZT detectors. Each detector was outfitted with a custom rectangular collimator and shield made out T-Flex®, a tungsten-impregnated polymer. Measurements were performed at three different locations of the cylinder, ensuring that the measurement geometry satisfied the infinite thickness criterion. The spectra from the M400 CZT were analyzed using the code General Enrichment Meter (GEM). For analyzing the gamma-ray spectra from UF6 cylinder, the GEM code is the appropriate tool since it relies only on the gamma-ray emissions from 235U and not from other isotopes. Results from the spectral analysis were compared with the known abundance of 235U in the cylinders, as well as with the International Target Values 2020 (ITV2020). The suitability of the different underlying techniques used by the various codes for UF6 analysis is discussed. The challenges of measuring UF6 contained in cylinders and mitigation strategies are highlighted.

Nuclear security measures: A review of selected emerging technologies and strategies

This review article delves into the ever-evolving landscape of nuclear security, presenting an overview of emerging technologies and strategies that have the potential to enhance the safeguarding of nuclear materials, facilities, and information. With the global security environment continuously evolving, staying ahead of potential threats to nuclear assets is paramount. This article explores a wide range of cutting-edge advancements that could revolutionize nuclear security efforts, offering a balanced assessment of their benefits, and challenges. To facilitate an organized presentation and analysis, the extracted articles were categorized into distinct sections based on the identified key topics. These categories included "advanced surveillance and monitoring systems," "cybersecurity and digital protection," "non-destructive evaluation techniques," "physical security enhancements," "nuclear material characterization nuclear forensics," and "international collaboration and policy frameworks." Each section was dedicated to exploring the advancements, challenges, and implications specific to the respective technology or strategy. By offering this synthesis, the article aims to serve as a strategic resource for policymakers, security professionals, and researchers, underlining the pivotal role of technology in safeguarding against nuclear threats in an interconnected world.

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.

Simultaneous isotopic analysis of fission product Sr, Mo, and Ru in spent nuclear fuel particles by resonance ionization mass spectrometry

Fission product Sr, Mo, and Ru isotopes in six 10-μm particles of spent fuel from a pressurized water reactor were analyzed by resonance ionization mass spectrometry (RIMS) and evaluated for utility in nuclear material characterization. Previous measurements on these same samples showed widely varying U, Pu, and Am isotopic compositions owing to the samples’ differing irradiation environments within the reactor. This is also seen in Mo and Ru isotopes, which have the added complication of exsolution from the UO2 fuel matrix. This variability is a hindrance to interpreting data from a collection of particles with incomplete provenance since it is not always possible to assign particles to the same batch of fuel based on isotopic analyses alone. In contrast, the measured 90Sr/88Sr ratios were indistinguishable across all samples. Strontium isotopic analysis can therefore be used to connect samples with otherwise disparate isotopic compositions, allowing them to be grouped appropriately for interpretation. Strontium isotopic analysis also provides a robust chronometer for determining the time since fuel irradiation. Because of the very high sensitivity of RIMS, only a small fraction of material in each of the 10 μm samples was consumed, leaving the vast majority still available for other analyses.

Machine learning in analytical spectroscopy for nuclear diagnostics [Invited

Analytical spectroscopy methods have shown many possible uses for nuclear material diagnostics and measurements in recent studies. In particular, the application potential for various atomic spectroscopy techniques is uniquely diverse and generates interest across a wide range of nuclear science areas. Over the last decade, techniques such as laser-induced breakdown spectroscopy, Raman spectroscopy, and x-ray fluorescence spectroscopy have yielded considerable improvements in the diagnostic analysis of nuclear materials, especially with machine learning implementations. These techniques have been applied for analytical solutions to problems concerning nuclear forensics, nuclear fuel manufacturing, nuclear fuel quality control, and general diagnostic analysis of nuclear materials. The data yielded from atomic spectroscopy methods provide innovative solutions to problems surrounding the characterization of nuclear materials, particularly for compounds with complex chemistry. Implementing these optical spectroscopy techniques can provide comprehensive new insights into the chemical analysis of nuclear materials. In particular, recent advances coupling machine learning methods to the processing of atomic emission spectra have yielded novel, robust solutions for nuclear material characterization. This review paper will provide a summation of several of these recent advances and will discuss key experimental studies that have advanced the use of analytical atomic spectroscopy techniques as active tools for nuclear diagnostic measurements.

Novel nuclear materials characterization workflows enabled by fs-laser milling

Novel nuclear materials characterization workflows enabled by fs-laser milling

Overview of ion chromatographic applications for the analysis of nuclear materials: Case studies

AbstractAccurate, precise, and rapid analytical monitoring of various nuclear materials is essential for the smooth functioning of nuclear reactors. Ion chromatography (IC) has emerged as an effective analytical tool for simultaneous detection of different ions in a wide range of materials used in the nuclear industry. The major advantages over other techniques include superior selectivity and sensitivity for detection of anions and cations, wide dynamic range, and speciation studies of ions. This article provides an overview of different ion chromatographic methodologies developed for the analyses of various nuclear materials such as fuel, control rods, moderator, coolant, and process streams. Comparison of various analytical aspects of IC over the other routine techniques reveals the ease and multidimensional capability of the technique. An insight is given to the modern variations in the field such as coupling of IC with other techniques for the characterization of nuclear matrices, implementation of capillary IC in terms of miniaturization, and so on. The information presented herein will serve as a very useful resource for investigators in the field of characterization of nuclear materials.

Electroplated Tin Absorbers in Microcalorimeter Detectors for Ionizing Radiation

State-of-the-art microcalorimeter gamma-ray detectors 1, 2 show promise for nuclear material characterization because their spectral resolution is substantially better than high purity germanium sensors. Microcalorimeter detectors have been fabricated and tested by National Institute of Standards and Technology (NIST) and Los Alamos National Laboratory (LANL) by combining microfabricated transition edge sensors (TESs) with bulk tin gamma-ray absorbers. The ability to integrate the tin electroplating into the microfabrication process has potential benefits for performance, scalability, reliability, and ease of manufacturing.This work investigates a commercially available tin/silver chemistry traditionally used for 3-D semiconductor packaging. Samples were electroplated on a sacrificial substrate changing chemistry makeup, current density, and time to realize pure tin deposition at thicknesses of 380-400 micron. Characterization was performed using SEM, EDS, XRF, and XPS and compared to bulk tin properties used in past microcalorimeter devices. The electroplated tin samples were diced into individual 1.3x1.3 mm sizes and integrated with TES detectors at NIST and LANL. The assembled devices were tested with gamma radiation sources to evaluate performance. This work will provide an overview of the project, current progress, and future ideas for realizing a fully integrated microcalorimeter fabrication process. D. A. Bennett, R. D. Horansky, D. R. Schmidt, A. S. Hoover, R. Winkler, B. K. Alpert, J. A. Beall, W. B. Doriese, J. W. Fowler, C. P. Fitzgerald, G. C. Hilton, K. D. Irwin, V. Kotsubo, J. A. B. Mates, G. C. O'Neil, M. W. Rabin, C. D. Reintsema, F. J. Schima, D. S. Swetz, L. R. Vale, and J. N. Ullom, Rev Sci Instrum, 83 (9), (2012).D. T. Becker, B. K. Alpert, D. A. Bennett, M. P. Croce, J. W. Fowler, J. D. Gard, A. S. Hoover, Y. I. Joe, K. E. Koehler, J. A. B. Mates, G. C. O'Neil, M. W. Rabin, C. D. Reintsema, D. R. Schmidt, D. S. Swetz, P. Szypryt, L. R. Vale, A. L. Wessels, and J. N. Ullom, Ieee T Nucl Sci, 66 (12), 2355-2363 (2019). Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia, LLC., a wholly owned subsidiary of Honeywell International, Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA-0003525, SAND2020-5481 A . The views expressed in the article do not necessarily represent the views of the U.S. Department of Energy or the United States Government.

Advanced FIB/SEM Characterization of Nuclear Materials in the Irradiated Materials Characterization Lab

Advanced FIB/SEM Characterization of Nuclear Materials in the Irradiated Materials Characterization Lab

Microscopy and spectroscopy of plutonium dioxide aging under ambient and near-ambient conditions

Characterization of nuclear materials under controlled conditions promotes a better understanding of their chemistry and provides insight into their stability. Of particular importance is identifying how environmental and storage conditions affect both bulk and sub-bulk scale chemical and physical properties. Such data may provide insight for nuclear forensic applications and can help address fundamental safety and storage concerns. In this work, plutonium dioxide (PuO2) and thorium dioxide (ThO2) aging was studied as a function of temperature and relative humidity (RH) while simultaneously observing the effects of fluoropolymer containment. Morphological changes were monitored using scanning electron microscopy and interpreted utilizing a previously-published lexicon of descriptive imaging terminology. Changes in phase were observed using X-ray diffraction measured on a single-crystal X-ray diffractometer to permit sub-bulk scale (< 10 mg) analysis. Ancillary techniques, including energy-dispersive X-ray, infrared, Raman, and X-ray photoelectron spectroscopy, were used to further probe and interpret the observed aging effects. ThO2 materials exhibited no change in macroscopic appearance and microstructure, whereas PuO2 underwent transitions to mixed plutonium dioxide/plutonium fluoride hydrate and mixed plutonium dioxide/ammonium plutonium fluoride phases under ambient temperature at 81% RH and 45 °C at 81% RH, respectively. These results highlight important stability differences between actinide oxides and underpin the belief that further research is needed to elucidate the fundamental chemical and physical properties of many actinide compounds, like PuO2 and ThO2, near ambient temperatures and humidities.

Proposal of a prototype plant based on the exfoliation process for the treatment of irradiated graphite

Abstract Most of irradiated graphite that should be disposed comes from moderators and reflectors of nuclear power plants. The quantity of irradiated graphite could be higher in the future if high-temperature reactors (HTRs) will be deployed. In this case noteworthy quantities of fuel pebbles containing semi-graphitic carbonaceous material should be added to the already existing 250,000 tons of irradiated graphite. Industry graphite is largely used in industrial applications for its high thermal and electrical conductivity and thermal and chemical resistance, making it a valuable material. Irradiated graphite constitutes a waste management challenge owing to the presence of long-lived radionuclides, such as 14C and 36Cl. In the ENEA Nuclear Material Characterization Laboratory it has been successfully designed a procedure based on the exfoliation process organic solvent assisted, with the purpose of investigate the possibility of achieving graphite significantly less toxic that could be recycled for other purpose [1]. The objective of this paper is to evaluate the possibility of the scalability from laboratory to industrial dimensions of the exfoliation process and provide the prototype of a chemical plant for the treatment of irradiated graphite.

Study on the performance of different uranium isotopic codes used in nuclear safeguards activities

Accurate characterization of Nuclear Materials (NMs) for nuclear safeguards purposes requires studying the optimal conditions over a wide range of experimental conditions. Gamma spectroscopy is commonly used to measure the 235U enrichment in NMs bearing samples. Several commercials Codes such as the Multi Group Analysis (MGA++), the Multi Group Analysis for Uranium (MGAU) and the “Fixed energy, Response function Analysis with Multiple efficiencies” (PC/FRAM) are used to measure 235U enrichment. Many factors, such as sample enrichment, density, homogeneity, geometry, and container type may affect the results of measurement. This work studies the effect of these factors on the performance of each of the three codes to estimate uranium enrichment. The results are compared to determine the optimum conditions and best performance among the three codes for achieving the most accurate estimation of 235U enrichment in the measured samples.

Analytical chemistry of nuclear material: case studies from Los Alamos National Laboratory

Presented here are three case studies that demonstrate the breadth of bulk actinide analytical chemistry at Los Alamos National Laboratory (LANL). This laboratory was established during the Manhattan Project and is still considered one of the premier analytical laboratories for special nuclear materials characterization around the globe. First, the analytical chemistry of 238Pu is presented, showcasing the unique challenges of this high activity isotope. Next, the analysis of 241AmO2 is discussed from the perspective of supporting nascent production at LANL. Finally, the analytical chemistry support of the production of plutonium non-destructive assay standards is discussed. These examples paint a picture of the analytical capabilities needed to meet the dynamic challenges of LANL’s nuclear programs.

Nuclear recoil spectroscopy of levitated particles

We propose a new method for the detection and characterization of nuclear decay processes. Specifically, we describe how nuclear decay recoil can be observed within small particles levitated in an optical trap with high positional resolution. Precise measurements of the magnitude of each recoil as well as their rate of occurrence can provide accurate information about the isotopic composition of a radioactive sample. We expect that this new technique for nuclear material characterization will be especially useful in the area of nuclear forensic analysis.

Determination of the instantaneous strain rate during small punch testing of 316 L stainless steel

The Small Punch Test, SPT, has been widely used by scientists for the mechanical characterization of nuclear materials and different applications. This technique is able to provide an approximation of mechanical properties from a very small amount of material. However, the inhomogeneous distributions of local stresses and strains, coupled to the response of the material hinder accurate predictions of the mechanical properties in terms of tensile tests parameters. Additionally, the determination of the instantaneous strain rate, as the test proceeds, remained a pending issue. This article analyses the predictions of SPT on 316 L steels by a FEM model, validated by an additively manufactured 316 L SLM (Selective Laser Melting) stainless steel, in terms of thickness evolution, local deformation and instantaneous strain rate during SPT testing. This alternative analysis shows that the deformation behaviour and, thus, the final form of the sample is affected by the strain hardening behaviour, which it has an influence on the instantaneous local strain rate. The instantaneous local strain rate was calculated for two dissimilar 316 L austenitic stainless steel, one “hardened”, characterized by low ductility but very high yield strength (YS) and ultimate tensile strength UTS, and another “annealed”, characterized by very low YS and large ductility, typical of annealed 316 L steel. The calculated central and minimum thicknesses, strain and instantaneous local strain rate, are not linear as SPT testing proceeds. Therefore, they are fitted to polynomial equations for both annealed and hardened 316 L steels. These functions allow, for instance, LVDT measurement corrections for appropriate SPT curve analysis, and proper true strain rate measurements for creep (SPCT) data analysis.

A Photofission Delayed γ-ray Spectra Calculation Tool for the Conception of a Nuclear Material Characterization Facility

The photon interrogation analysis is a nondestructive technique allowing to identify and quantify fissile materials in nuclear waste packages. This paper details an automatic procedure which has been developed to simulate the delayed γ-ray spectra for several actinide photofissions. This calculation tool will be helpful for the fine conception (collimation, shielding, noise background optimizations, etc.) and for the on-line analysis of such a facility.

Retrospective Imaging and Characterization of Nuclear Material.

Modern techniques for detection of covert nuclear material requires some combination of real time measurement and/or sampling of the material. More common is real time measurement of the ionizing emission caused by radioactive decay or through the materials measured in response to external interrogation radiation. One can expose the suspect material with various radiation types, including high energy photons such as x rays or with larger particles such as neutrons and muons, to obtain images or measure nuclear reactions induced in the material. Stand-off detection using imaging modalities similar to those in the medical field can be accomplished, or simple collimated detectors can be used to localize radioactive materials. In all such cases, the common feature is that some or all of the nuclear materials have to be present for the measurement, which makes sense; as one might ask, "How you can measure something that is not there?" The current work and results show how to do exactly that: characterize nuclear materials after they have been removed from an area leaving no chemical trace. This new approach is demonstrated to be fully capable of providing both previous source spatial distribution and emission energy grouping. The technique uses magnetic resonance for organic insulators and/or luminescence techniques on ubiquitous refractory materials similar in theory to the way the nuclear industry carries out worker personnel dosimetry. Spatial information is obtained by acquiring gridded samples for dosimetric measurements, while energy information comes through dose depth profile results that are functions of the incident radiation energies.

Source Correlated Prompt Neutron Activation Analysis for Material Identification and Localization

This paper investigates the energy spectrum of photon signatures from an associated particle imaging deuterium tritium (API-DT) neutron generator interrogating shielded uranium. The goal is to investigate if signatures within the energy spectrum could be used to indirectly characterize shielded uranium when the neutron signature is attenuated. By utilizing the correlated neutron cone associated with each pixel of the API-DT neutron generator, certain materials can be identified and located via source correlated spectrometry of prompt neutron activation gamma rays. An investigation is done to determine if fission neutrons induce a significant enough signature within the prompt neutron-induced gamma-ray energy spectrum in shielding material to be useful for indirect nuclear material characterization. The signature deriving from the induced fission neutrons interacting with the shielding material was slightly elevated in polyethylene-shielding depleted uranium (DU), but was more evident in some characteristic peaks from the aluminum shielding surrounding DU.

Vibro‐acoustic imaging development for microstructure characterization and metrology

Vibro-acoustics (VA) is being developed into a portable scanning infrastructure for a novel material characterization technique focused on nuclear applications to characterize fuel, cladding materials, and structures at the Idaho National Laboratory (INL). The proposed VA technology is based on ultrasound and acoustic waves; however, it provides information beyond what is available from the traditional ultrasound techniques and can expand the knowledge on nuclear material characterization and microstructure evolution. VA is a three‐dimensional (3D) imaging modality based on ultrasound‐stimulated acoustic emission. VA uses the force caused by the beating of two frequencies to generate an acoustic emission signal. Due to absorption or reflection, the energy density in the object at an acoustic focal point changes to produce a “radiation force.” This force locally vibrates the object, which results in an acoustic field that depends on the characteristics of the object at that point. This acoustic field is detected for every point in the object; the resulting data is used to make an image of the object’s mechanical properties. This paper will focus on the development of a portable scanning system to image in situ the microstructure of materials used in nuclear applications.

Sensitivity and Uncertainty Analysis of Plutonium and Cesium Isotopes in Modeling of BR3 Reactor Spent Fuel

Nuclear forensics has a broad task to characterize recovered nuclear or radiological material and interpret the results of investigation. One approach to isotopic characterization of nuclear material obtained from a reactor is to chemically separate and perform isotopic measurements on the sample and verify the results with modeling of the sample history, for example, operation of a nuclear reactor. The major actinide plutonium and fission product cesium are commonly measured signatures of the fuel history in a reactor core. This study investigates the uncertainty of the plutonium and cesium isotope ratios of a fuel rod discharged from a research pressurized water reactor when the location of the sample is not known a priori. A sensitivity analysis showed overpredicted values for the 240Pu/239Pu ratio toward the axial center of the rod and revealed a lower probability of the rod of interest (ROI) being on the periphery of the assembly. The uncertainty analysis found the relative errors due to only the rod position and boron concentration to be 17% to 36% and 7% to 15% for the 240Pu/239Pu and 137Cs/135Cs ratios, respectively. This study provides a method for uncertainty quantification of isotope concentrations due to the location of the ROI. Similar analyses can be performed to verify future chemical and isotopic analyses.