Fissile Elements Research Articles

β-decay studies across N = 126

With the advent of the first Radioactive Ion Beam facilities of new generation, the investigation of the neutron-rich side of the nuclear chart has experienced an impressive progress. However, the discovery and study of new nuclear species in the region around the heaviest known neutron shell closure, N = 126, is still one of the most coveted experimental challenges. At a slower pace, the exploitation of alternative reaction mechanisms and/or advanced instrumentation has opened the possibility to investigate the isomeric and β decay of new, moderately neutron-rich N ∼ 126 nuclei. These are of relevance for the understanding of the nuclear structure below the doubly-magic 208Pb and for their role in the synthesis of the trans-bismuth fissile elements in the r process. In this contribution, a general overview of the β-decay experiments performed at both sides of N = 126 is provided, with a main focus on the experiments carried out at the fragmentation facilities GSI and RIBF.

Facile Preparation of Macro-Microporous Thorium Oxide via a Colloidal Sol-Gel Route toward Safe MOX Fuel Fabrication.

The identification of new colloidal sol-gel routes for the preparation of actinide oxides, which have a homogeneous and accessible porosity that can easily be impregnated by any concentrated actinide solution, opens new perspectives for the preparation of homogeneous nuclear fuel for minor actinide transmutation. This homogeneity allows us to avoid "hot spot" formation due to the local accumulation of more fissile elements. Here, we report the preparation of macro-microporous ThO2 materials by a colloidal sol-gel route. Using a thorium salt with 6-aminocaproic acid as a complexing agent at a controlled pH, we were able to pilot the condensation of thorium hydroxo species forming colloids of tuned nanometric size and thus the sol stability. After a freeze-drying process to concentrate colloids and a thermal treatment allowing complexing agent removal and macroporosity formation by a brutal gas release during combustion, a loose packing of ThO2 nanoparticles with an ordered distribution of interparticular porosity and a fraction of nanometric crystallites, whose size depends on the initial colloidal size, were obtained. The sols, pastes, and final materials were characterized by small- and wide-angle X-ray scattering to determine the colloidal size and the final structure of the materials, which was also confirmed by transmission electron microscopy. The most promising material was finally successfully impregnated by a simulating minor actinide solution and thermally treated to prepare a mixed actinide oxide material. This safe technology, relying on the colloidal sol-gel process and the formulation of complex fluids forming tunable precursors, opens new perspectives for the reuse of nuclear waste solutions as new fuel.

STUDY OF THE EFFECT OF ZCO CERAMIC PHASE COMPOSITION ON OPTICAL AND THERMOPHYSICAL PROPERTIES

The work is devoted to the study of optical and thermophysical properties of promising materials for nuclear power of a new generation - inert matrices of nuclear dispersed fuel based on ZrO2 – CeO2 (ZCO) doped with Y2O3. Interest in this class of materials is due to great prospects in the field of replacing the traditional type of nuclear fuel with uranium-free dispersed nuclear fuel, which is based on a technological solution using plutonium as a fissile element, placed in a dispersed inert matrix based on oxide or nitride ceramics. The purpose of this work is to study the effect of doping of ZCO ceramics with Y2O3 on the optical and heat-conducting properties. During the study, it was found that an increase in the concentration of the Y2O3 dopant leads to the formation of a highly ordered CeZrO4 phase, the presence of which leads to the appearance of an additional absorption band in the region of 550 – 650 nm, as well as a decrease in the transmission value in the region of 400 – 700 nm. To evaluate the results obtained, the methods of optical UV spectroscopy and heat flux determination were applied. It has been found that doping with Y2O3 leads to the formation of a highly ordered CeZrO4 phase, the formation of which leads to the appearance of additional spectral absorption lines, as well as a shift in the band gap to the region of low energies. During analysis of the thermophysical parameters, it was found that a change in the phase composition leads to an increase in the thermal conductivity coefficient by more than 40 – 60 %.

UN and U3Si2 composites densified by spark plasma sintering for accident-tolerant fuels

Both UN and U3Si2 are potential candidates for accident tolerant fuels due to their high fissile element density and exceptional thermal conductivity. However, they display a high susceptibility to oxidation and corrosion in steam environments. UN and U3Si2 composites were synthesized by conventional vacuum sintering but suffered significant micro-cracks in the silicide phase due to a mismatch in their thermal expansion. In this work, we report promising results of synthesizing UN-U3Si2 composites by spark plasma sintering and the micro-cracks can be significantly mitigated by controlling the cooling. Composite fuel pellets with a density of over 95% theoretical density and a uniform distribution of nitride and silicide phases can be achieved without micro-crack formation. The composite with 50 wt% UN and 50 wt% U3Si2 displays simultaneously-enhanced strength and fracture toughness, and possesses excellent thermal conductivity. The onset temperatures of all composites tested through dynamic oxidation testing are close to 540 °C, suggesting significantly improved oxidation resistance than the monolithic UN. These results demonstrate the potential of synergizing UN and U3Si2 in composites with enhanced fuel properties.

Cr-incorporated uranium nitride composite fuels with enhanced mechanical performance and oxidation resistance

Cr-incorporated UN (Cr-UN) composite fuels with different Cr amounts up to 10 wt% are fabricated by spark plasma sintering, and their microstructure and phase heterogeneity are analyzed. Highly densified microstructure and homogeneous Cr distribution are identified for the Cr-incorporated UN pellets, demonstrating a liquid sintering characteristic with a Cr enriched phase on the UN grain boundaries. A ternary phase U2CrN3 forms in the Cr-UN matrix with Cr amounts of 5 wt% and 10 wt%. The SPS densified Cr-UN composite pellets display greatly-improved thermal conductivity and simultaneously high hardness and fracture toughness. The fracture toughness of the Cr-incorporated UN pellet is ∼5.5 MPa-m1/2, representing almost 130% enhancement as compared to that of monolithic UN sintered at the same condition. The Cr-incorporated UN pellets also display enhanced oxidation resistance as evidenced by increased onset temperature for oxidation to 450 oC for the 5 wt% Cr-UN. These results highlight that Cr additive and the formation of a ternary phase can be useful to improve the thermal-mechanical properties and oxidation resistance of UN fuels with well-maintained high fissile element density.

Reactor experiments review

Reactors are a copious source of anti-neutrinos and for this reason they have been critical tools to pin down neutrino fundamental properties since their discovery. Depending on the distance between reactor cores and experiments different oscillation parameter regions can be probed: in this review the regimes of very short, short and medium baselines are discussed with highlights on the current results and prospectives. Furthermore the high resolution which can be nowadays achieved allows to understand better the contributions from different fissile elements in the anti-neutrino spectrum. Fuel composition analysis as well as sterile neutrino search are also discussed.

Transition to thorium fuel cycle on a heavy water moderated molten salt reactor by using low enrichment uranium

Lacking fissile element (U-233) in nature is a key issue faced by Th-U fuel cycle, and existing available fuels should be taken as starting fuel to convert Th-232 to U-233 to realize transition to thorium fuel cycle. Low enrichment uranium (LEU, U-235 enrichment < 20%) is the most accessible fuel, and has been widely studied as starting fuel to realize transition to Th-U fuel cycle in graphite moderated molten salt reactors. Due to an inevitable introduction of large amount of fertile element U-238, however, this transition way has proven to be almost impossible. Heavy water moderated molten salt reactor (HWMSR) is an innovative molten salt reactor adopting heavy water as the moderator and heavy metal elements dissolved in fluoride salt as the fuel. A high neutron economy resulting from heavy water paves the way for the transition from LEU to thorium fuel cycle. The results obtained from this study indicated that transition can be achieved by using LEU with enrichment no lower than 5%. Raising enrichment would also enhance the breeding performance, but exert negative/positive influences on the attractiveness of uranium/plutonium contained in the fuel salt, decrease the heavy metal utilization efficiency and impose a small impact on the fuel cycle cost.

Aluminum-doped U3Si2 composite fuels with enhanced oxidation resistance

Al-doped U3Si2 composite fuels with controlled microstructure were fabricated by spark plasma sintering that display greatly-improved oxidation resistance as compared with monolithic and Al-doped silicides prepared by standard powder metallurgy or arc melting. The effects of Al additives on the thermal-mechanical properties and oxidation resisance of the micron- and nano-sized U3Si2 composites were investigated. A minimal addition of 1.8 at% Al is effective to increase the onset oxidation temperature of as-fabricated U3Si2 pellets to 580 °C, which can be further increased to 610 °C by thermal annealing. The Al-doped U3Si2 composite fuels also display simultaneously higher hardness and fracture toughness than un-doped U3Si2. These results highlight an effective strategy by integrating minimal Al additives, microstructure control and post-thermal annealing to design advanced silicide fuels with excellent oxidation resistance, desired thermal-mechanical properties and maintained high fissile element density.

Optimized core design for small long-life gas cooled fast reactors with natural uranium-thorium-blend as fuel cycle input

Nuclear power is the use of nuclear reactions that release nuclear energy to generate heat, which most frequently is then used in steam turbines to produce electricity. Presently, the vast majority of electricity from nuclear power is produced by nuclear fission of uranium. The alternative fuel was thorium with the key similarity is that both can absorb neutrons and transmute into fissile elements. Thorium is approximately three times as abundant as uranium in the earth’s crusts, reflecting the fact that thorium has a longer half-life. However, thorium is a fertile material that could not react spontaneously. It needs to be mixed with fissile materials such as uranium or plutonium. Optimized core design and fuel management for small long-life gas cooled fast reactors (600 MWt GCFR) has been performed. The fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh natural uranium-thorium blend is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and region-1 filled by fresh natural uranium-thorium blend fuel. This concept is basically applied to all regions in axial core, i.e. shifted the core of (i) region into (i+1) region after the end of 10 years burn-up cycle. For the next cycles, we will add only natural uranium-thorium blend on region-1.The calculation results show that the composition of the fresh fuel for the optimized core consists of 60% thorium-232 and 40% natural uranium.

ON THE MULTIPURPOSE USE OF A PORTABLE NEUTRON SOURCE

The possibility of creating a multipurpose complex for generating a reference field of thermal neutrons based on a portable neutron source (PNS) is considered. It has been shown that our method can be used to detect fissile materials without determining their isotopic composition during an inspection of the hand luggage of passengers, mail, etc. The PNS multipurpose complex will allow to indirectly indicating the possible presence of chemical explosives in the test samples, as well as cadmium and boron, which possibly hide fissile elements from detection. The paper gives recommendations on the use of the most effective instruments and equipment.

U3Si2 and UO2 composites densified by spark plasma sintering for accident-tolerant fuels

This work reports the synthesis and characterization of the U3Si2 and UO2 composites sintered by spark plasma sintering (SPS) with controlled microstructures for accident-tolerant fuel application. The U3Si2 and UO2 composites with different silicide and oxide ratios were synthesized by SPS at temperatures from 1000 to 1300 °C for 5 minutes. The microstructure and phase composition of the SPS densified composite fuels were characterized with scanning electron microscopy, X-ray diffraction (XRD), and energy dispersed spectroscopy (EDS). A systematic study of the thermal and mechanical properties was conducted using microhardness testing and laser flash apparatus, along with oxidation resistance measurements using thermogravimetric analysis (TGA). The results show that the synthesis of composite fuels can be achieved with a 90% theoretical density (TD) at 1000 °C and over 95% TD when sintered at 1300 °C. XRD and EDS results confirmed that the dominant phases in the composites are U3Si2 and UO2. Improved physical density generally leads to improved hardness, fracture toughness, thermal diffusivity, and onset temperature during the oxidation process. U3Si2 was found to play a dominant role in determining the mechanical and oxidation properties of the composite fuels, whereas UO2 had a more important impact on controlling the thermal diffusivity of the composites. The composite with 50 wt% UO2 sintered at 1300 °C displayed the onset oxidation temperature of 500 °C by dynamic oxidation testing using TGA at a ramp degree of 10 °C/min. The composite also achieved a high fracture toughness of ∼3.5 MPa m½. These results highlight the potential of composite fuel forms densified by SPS with simultaneously enhanced fissile element density, fracture toughness, thermal transport properties, and oxidation resistance.

Autonomous Neutron Facility for Detecting Fissile Nuclear Materials

The prospect of creating an autonomous neutron facility for the detection of fissile nuclear materials in samples, including those in confined volume, is discussed. It is proposed to obtain a reference field of thermal neutrons on the basis of a polyethylene moderator ball and a portable fast neutrons source, developed at NRC “Accelerator” NSC KIPT based of a continuous electrostatic accelerator of deuterons. The developed source of thermal neutrons is planned to be used to activate small objects and goods in order to identify substances prohibited for movement containing to find 233U, 235U and 239Pu in their composition. The prompt finding of fissile elements will indicate about an attempt to transport them illegally. A more thorough inspection can be carried out using special equipment after the detention of suspicious goods, citizens or vehicles. The possibility of detecting prompt fission neutrons is considered not only in the traditional way using a polyethylene moderator and proportional 3He detector, but also without application of any moderator using oxide or semiconductor scintillators. For detection fissile materials the method based on using the high-energy part of the γ-spectrum of fission fragments (greater than 4900 keV), as well, as the approach applied in the passive non-destructive analysis by the γ-line with Eg = 185.7 keV from 235U, are substantiated. It is shown that the proposed facility for the detection of fissile nuclear materials is able to determine the presence of isotopes 233U, 235U and 239Pu in tested objects and goods with the using non-destructive testing method.

Analysis of the Recent Nuclear Data through a Fast Benchmarks Highly Enriched Uranium Study Using the Mcnp6.1 Code

In this work, we present our MCNP6.1 modeling of some critical fast experimental benchmarks, aiming to qualify our cross section libraries derived from ENDF/B-VII.1, ENDF/B-VII, JEFF-3.1, JENDL-3.3, and JENDL-4.0. The analyzed benchmarks are characterized by simple geometries which helps to have taken precise results, and concerning the type: HEU-MET-FAST (highly enriched uranium). Those benchmarks are extracted from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE) published by the Nuclear Energy Agency [1]. A detailed comparison of the results of our simulation was made in order to highlight the influence of these nuclear data types on our calculations, due to its importance for the stability of nuclear reactors. We interpreted the difference between calculation and experiment ($$C{-}E$$) for the principal parameter $$k_{\textrm{eff}}$$ through a comparison of the fission and capture rates of the major fissile elements. The different spectral indices F28/F25, F49/F25, F37/F25, C28/F25, and F23/F25 at the cores center are also calculated. For the majority of the studied HEU (highly enriched uranium) benchmark cases, The ENDF/B-VII and JEFF-3.1 have a good agreement with the experimental ones, concerning the $$k_{\textrm{eff}}$$ results. The average discrepancy from the experimental values for ENDF/B-VII is 0.42$$\%$$, and 0.39$$\%$$ for JEFF-3.1. An overestimation was observed for most evaluations concerning benchmarks with tungsten carbonate reflectors. The best results were obtained by JENDL-3.3, with a maximal discrepancy $$C{-}E$$ estimated 0.47$$\%$$ concerning fission rate, and 4.25$$\%$$ for capture rate. In analyzing the spectral indices, for GODIVA and FLATTOP-25, the best results were obtained by JENDL-4.0 with a maximal discrepancy of 2.66$$\%$$.

Chemical Stability of Zirconolite for Proliferation Resistance under Conditions Typically Required for the Leaching of Highly Refractory Uranium Minerals

In this study, synthetic zirconolite samples with a target composition Ca0.75Ce0.25ZrTi2O7, prepared using two different methods, were used to study the stability of zirconolite for nuclear waste immobilisation. Particular focus was on plutonium, with cerium used as a substitute. The testing of destabilisation was conducted under conditions previously applied to other highly refractory uranium minerals that have been considered for safe storage of nuclear waste, brannerite and betafite. Acid (HCl, H2SO4) leaching for up to 5 h and alkaline (NaHCO₃, Na2CO3) leaching for up to 24 h was done to enable comparison with brannerite leached under the same conditions. Ferric ion was added as an oxidant. Under these conditions, the synthetic zirconolite dissolved much slower than brannerite and betafite. While the most intense conditions were observed previously to result in near complete dissolution of brannerite in under 5 h, zirconolite was not observed to undergo significant attack over this timescale. Fine zirconolite dissolved faster than the coarse material, indicating that dissolution rate is related to surface area. This data and the long term stability of zirconolite indicate that it is a good material for long-term sequestration of radioisotopes. Besides its long term durability in the disposal environment, a wasteform for fissile material immobilisation must demonstrate proliferation resistance such that the fissile elements cannot be retrieved by leaching of the wasteform. This study, in conjunction with the previous studies on brannerite and betafite leaching, strongly indicates that the addition of depleted uranium to the wasteform, to avert long term criticality events, is detrimental to proliferation resistance. Given the demonstrated durability of zirconolite, long term criticality risks in the disposal environment seem a remote possibility, which supports its selection, above brannerite or betafite, as the optimal wasteform for the disposition of nuclear waste, including of surplus plutonium.

Characterization Study of Algerian Bentonite Samples Using Nuclear Techniques Analysis for Environment Applications

Bentonite material was synthesized and processed based on a raw matrix from the M'Zila deposit, W. Mostaganem, Algeria, to evaluate their efficiencies in the treatment of radioactive waste in solution. These BMB and (α-Fe2O3 / BMP-Na) materials have been subjected to elemental chemical analysis using two nuclear techniques which are the Instrumental Neutron Activation Analysis (INAA) and the Energy Dispersive X-ray fluorescence analysis (EDXRF) technique. These techniques were multi-elementary, non-destructive and have a good precision and allowed us to identify the overall chemical composition of solid samples, in major elements such as iron (Fe), and Potassium (K) which form the composition of the soil matrix, as well as minor elements such as zirconium (Zr), Barium (Ba), and Sodium (Na) in particular rare earth elements (REEs), which are indicators of the presence of fissile elements such as uranium and thorium. The trace elements also have been detected such as Chromium (Cr), Zinc (Zn), Cobalt (Co), Galium (Ga), Scandium (Sc) and Cesium (Cs), Arsenic (As), and Bromine (Br) which are only trace elements in the ppm range and not metals. we recorded also the absence of Mercury (Hg), Selenium (Se) and Antimony (Sb) which indicates that our sampling deposit of the studied bentonite was not exposed to any external source pollution. These techniques allowed us also to confirm and quantify the insertion of hematite (α-Fe2O3) during doping and after sodium homoionization of the soil matrix.

Radiation-induced grain subdivision and bubble formation in U3Si2 at LWR temperature

U3Si2, an advanced fuel form proposed for light water reactors (LWRs), has excellent thermal conductivity and a high fissile element density. However, limited understanding of the radiation performance and fission gas behavior of U3Si2 is available at LWR conditions. This study explores the irradiation behavior of U3Si2 by 300 keV Xe+ ion beam bombardment combining with in-situ transmission electron microscopy (TEM) observation. The crystal structure of U3Si2 is stable against radiation-induced amorphization at 350 °C even up to a very high dose of 64 displacements per atom (dpa). Grain subdivision of U3Si2 occurs at a relatively low dose of 0.8 dpa and continues to above 48 dpa, leading to the formation of high-density nanoparticles. Nano-sized Xe gas bubbles prevail at a dose of 24 dpa, and Xe bubble coalescence was identified with the increase of irradiation dose. The volumetric swelling resulting from Xe gas bubble formation and coalescence was estimated with respect to radiation dose, and a 2.2% volumetric swelling was observed for U3Si2 irradiated at 64 dpa. Due to extremely high susceptibility to oxidation, the nano-sized U3Si2 grains upon radiation-induced grain subdivision were oxidized to nanocrystalline UO2 in a high vacuum chamber for TEM observation, eventually leading to the formation of UO2 nanocrystallites stable up to 80 dpa.

Middle-term thorium strategy for PWR fleets

Thorium constitutes a natural resource alternative to uranium to fuel nuclear fission power plants. This study explores an original way to benefit from thorium with pressurized water reactors (PWR). It relies on a versatile reactor EPR™ core concept which well circumscribes thorium fuel reprocessing and re-fabricationFleets representative of what can be deployed at the scale of a small country were simulated using the scenario software COSI6. The cornerstone of this study is here the possibility to multi-recycle plutonium into Pu/Th fuel as plutonium degradation is softened by mixing it with high-grade fissile matters. Plutonium content inside Pu/Th fuel remains then far below the limit existing in PWR for safety reasons. This synergistic effect between U/Pu and U/Th cycles enables a better valorization of fissile elements present in spent fuels.Improved plutonium management makes possible to stabilize spent fuel stocks and the plutonium inventory at the cycle back-end. Thanks to multi-recycling allowed by the U/Pu/Th studied scheme, uranium resource savings exceed 25%. Minor actinides production is also reduced even though production of curium alone increases. These results reveal that thorium introduction in standard PWR may be a sustainable middle-term option, assuming that U/Pu/Th fuel reprocessing and re-fabrication would become available.

Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

Abstract Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th) O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

Investigation of neutronic behavior in a CANDU reactor with different (Am, Th, 235U)O2 fuel matrixes

Abstract Recently thorium-based fuel matrixes are taken into consideration for nuclear waste incineration because of thorium proliferation resistance feature moreover its breeding or convertor ability in both thermal and fast reactors. In this work, neutronic influences of adding Am to (Th-235U)O2 on effective delayed neutron fraction, reactivity coefficients and burn up of a fed CANDU core has been studied using MCNPX 2.6.0 computational code. Different atom fractions of Am have been introduced in the fuel matrix to evaluate its effects on neutronic parameters of the modeled core. The computational data show that adding 2% atom fraction of Am to thorium-based fuel matrix won't noticeably change reactivity coefficients in comparison with the fuel matrix containing 1% atom fraction of Am. The use of 2% atom fraction of Am resulted in a higher delayed neutron fraction. According to the obtained data, 32.85 GWd burn up of the higher Americium-containing fuel matrix resulted in 55.2%, 26.5%, 41.9% and 2.14% depletion of 241Am, 243Am, 235U and 232Th respectively. 132.8 kg of 233U fissile element is produced after the burn up time and the nuclear core multiplication factor increases in rate of 2390 pcm. The less americium-containing fuel matrix resulted in higher depletion of 241/ 243Am, 235U and 232Th while the nuclear core effective multiplication factor increases in rate of 5630 pcm after the burn up time with 9.8 kg additional 233U production.

Analytical applications of delayed and instrumental neutron activation analysis

Delayed neutron activation analysis (DNAA) is a rapid and sensitive analytical method for the determination of fissile elements in a variety of samples. The present work describes two different analytical applications of delayed neutron activation for the analysis of biological and environmental samples, respectively. In the first application, DNAA was utilized to determine the natural uranium content in NIST standard reference materials (SRM) 1547 peach leaves and 1573a tomato leaves. Measured uranium mass fractions are comparable to the non-certified values listed on the certificates for these materials. In the second application, delayed neutron activation is coupled with instrumental neutron activation analysis (INAA) for the measurement of rare earth elements (REE) (cerium, lanthanum, neodymium, and ytterbium) in NIST SRM 2586, Trace Elements in Soil Containing Lead from Paint. DNAA was utilized to determine the uranium mass fraction in SRM 2586 for the subsequent application of a correction factor to account for cerium, lanthanum, and neodymium produced as part of the INAA irradiation. Measured and corrected mass fractions for the REEs described here are all within the uncertainty limits provided on the NIST certificate for SRM 2586. These results and the demonstrated sensitivity of the DNAA system establish and validate the use of this method for the determination of REEs and for potential nuclear forensics applications.