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At least 433 records · Page 24

Oriented perovskite crystals and methods of making the same

An aspect of the present disclosure is a method that includes combining a first organic salt (A1X1), a first metal salt (M1(X2)2), a second organic salt (A2X3), a second metal salt (M2Cl2), and a solvent to form a primary solution, where A1X1 and M1(X2)2 are present in the primary solution at a first ratio between about 0.5 to 1.0 and about 1.5 to 1.0, and A2X3 to M2Cl2 are present in the primary solution at a second ratio between about 2.0 to 1.0 and about 4.0 to 1.0. In some embodiments of the present disclosure, at least one of A1 or A2 may include at least one of an alkyl ammonium, an alkyl diamine, cesium, and/or rubidium.

Zhu, Kai↗

Luminescent group 1A copper halides and uses thereof

High photoluminescence, high stability, inorganic perovskite compounds comprising an alkali metal selected from potassium (K), rubidium (Rb), and cesium (Cs); copper (Cu); and at least one halogen selected from chlorine (Cl), bromine (Br), and iodine (I). The perovskites may be free of lead (Pb). The inorganic perovskite compound may be used in an optoelectronic device. The optoelectronic device optionally contains a phosphor such as a blue-emitting phosphor. The inorganic perovskite compound may be used as an anti-counterfeiting nanotaggant applied on or within an object that susceptible to counterfeiting to enable confirmation of an authentic object.

Saparov, Bayram↗

Development and application of sorption mass transfer models for fission product transport in TRISO fuel systems using BISON

Fission product mass transfer within and between TRISO particle and compact layers is an important phenomenon. It directly impacts fission product release predictions, which are used as source terms for safety and licensing calculations. Modeling fission product transport within layers and across bonded interfaces is relatively straightforward under the assumptions of isotropic Fickian diffusion, concentration continuity, and flux continuity. Modeling fission product transport across gaps and debonded layers is more difficult. Gaps are known to form between the buffer and inner PyC (IPyC), and debonding may occur between the IPyC and silicon carbide (SiC). A new mass transfer model was developed to provide a more accurate and robust fission product release calculation by accounting for interlayer sorptivity. One importance of the new model is to account for temperature and material changes across the gap based on sorption isotherm. This work presents the development of the sorption behavioral model and a fission product trapping model, and analyses the subsequent fission product diffusion behavior, specifically cesium, through TRISO particles and compact materials using the BISON fuel performance code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Characterization of Scintillating Crystal for High Energy Radiation Detection and Applications

Cesium Hafnium Chloride (Cs2Hfl6) is a promising metal halide scintillating detector due to its excellent characteristics such as non-hygroscopicity, high light yield without doping. Scintillators plays an important role in nuclear security applications as detectors at border monitoring stations. We grew and characterized CHC crystals to be fabricated for gamma detection. Photoluminescence was performed using the Duetta Fluorescence and Absorbance Spectrometer. The energy response to Cs-137 and Co-60 reference sources was recorded for a representative sodium iodide (NaI) scintillating detector.

Reeder, Angel↗

ARC-100 Reactor Security-by-Design Summary

This report applies the security-by-design methodology developed in a previous National Nuclear Security Administration–sponsored work to the Advanced Reactor Concepts 100 (ARC-100) sodium-cooled fast reactor (SFR) design. The report contains no proprietary information specific to the ARC 100 reactor. The insights developed in this report are high-level, and generally applicable to other sodium fast reactor designs. The information presented here is the result of a qualitative safety-based analysis and would not inform any potential adversary beyond what would be found in a docketed safety analysis report. The scope of this present report covers ARC-100’s reactor core, used fuel storage, and used fuel assembly wash station. These systems are also compared to a generic SFR design assumed in the previous study. The security assessment results show changes in structures, systems, and components (SSCs) safety importance relative to the generic SFR SSCs. However, the consequence assessment results are the similar to a previously assessed generic SFR. Several SSCs have higher importance rankings than others, and it is recommended that protection efforts are prioritized for these SSCs. This work will continue in the Fiscal Year 2025 for the remaining ARC-100 systems, including cesium trap, sodium cold trap, noble gas decay tanks (dewar bottles), and used fuel dry storage facility, to provide safety-and-security-by-design insights and recommendations on non-core systems. Results from this work will furnish a technical justification for the feasibility of these solutions for the ARC reactor's design and, where applicable, identify any regulatory benefits conferred by the proactive design aspect within a risk management framework. This initiative will contribute to a more secure design of the ARC reactor and support its licensing process.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Methods for making low bandgap perovskites

The present disclosure relates to a composition that includes a perovskite having a stoichiometry comprising A1-xFAxSn1-yBy(I1-zXz)3, where A is a first cation, B is a second cation, X is a halide, and 0.5≤x≤0.9, 0.5≤y≤0.9, and 0≤z≤1. In some embodiments of the present disclosure, A may include at least one of cesium, guanidinium, and/or methylammonium. In some embodiments of the present disclosure, X may include at least one of bromide and/or chloride. In some embodiments of the present disclosure, z may be equal to zero.

Zhu, Kai↗

Status of the Mu2e calorimeter readout electronics

The Mu2e experiment [1] at Fermilab will search for the neutrino-less coherent conversion of a muon into an electron in the field of a nucleus. Mu2e detectors comprise a straw tracker, an electromagnetic calorimeter and a veto for cosmic rays. The calorimeter employs 1348 Cesium Iodide crystals readout by silicon photomultipliers and fast front-end and digitization electronics. The front-end electronics consists of two discrete readout circuits (AMP-HV) for each crystal. These provide the amplification, shaping stage and linear regulation of the SiPM bias voltage and monitoring. The SiPM and front-end control electronics is implemented in a battery of mezzanine boards each equipped with an ARM processor that controls a group of 20 Amp-HV circuits distributing the low voltage and the high-voltage. The electronic is hosted in crates located on the external surface of calorimeter disks. The crates also host the waveform digitizer board (DIRAC) that performs digitization of the front end signals and transmit the digitized data to the Mu2e DAQ. Calorimeter electronic is hosted inside the cryostat and must sustain very high radiation and magnetic field so it was necessary to fully qualify it. The system design and quality assurance procedures will be reviewed.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

Removal and Deactivation of Bond Sodium from Fast Reactor Blanket Materials

The disposition of sodium-bonded spent nuclear fuel and blanket materials in a repository is complicated by the presence of sodium metal that is used as a thermal bond between the uranium metal fuel and blanket slugs and their cladding. The concern is that the metallic sodium could react with water, producing explosive hydrogen gas, or could exhibit a pyrophoric character. Thus, experimental studies were performed to investigate and demonstrate the removal and deactivation of bond sodium from blanket material in a dry environment. Specifically, bond sodium was removed from unirradiated Fermi-1 blanket elements and an assembly via a melt-drain-evaporate process using elevated temperature and reduced pressure. The effectiveness of sodium metal removal from the blanket materials and their associated cladding was =99.9998%, based on post-test quantitative analyses. The separated sodium metal was collected and subsequently deactivated by reacting it in a molten state with a controlled addition of ammonium chloride particles atop a molten salt medium. In this process sodium chloride is formed and assimilates into the salt pool. The subsequent deactivation of the bond sodium produced a solid ingot of sodium chloride, potassium chloride, lithium chloride, and cesium chloride that was devoid of sodium metal based on post-test analyses of the salt product. Both the sodium removal and deactivation operations were conducted within a dry inert atmosphere enclosure. The results of this study substantiate a path forward for the disposition of sodium-bonded blanket materials, including 34 metric tons heavy metal in irradiated Fermi-1 blanket material currently stored at Idaho National Laboratory.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

AGR TRISO Fuel Performance Modeling in FY-25

This report summarizes the activities performed in FY-25 to support fuel performance modeling for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. This includes implementation of a new uranium oxycarbide (UCO) kernel swelling rate model, inner pyrolytic carbon (IPyC) cracking behavior and failure predictions, development of new UCO and silicon carbide (SiC) cesium diffusion parameters, and the thermomechanical behavior of particle layers using experimental data from micro-tensile strength testing.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development and Testing of a new Current-Regulated Arc Modulator for the LINAC

In this study, the performance of a current-regulated arc modulator was investigated with a focus on its role in initiating and sustaining plasma discharge within the Magnetron Body of the LINAC system. The analysis centered on how switching components, circuit topology, and feedback loop architecture influence critical factors such as energy efficiency, discharge stability, and long-term plasma containment. Particular attention was given to variations in pulse termination behavior, as observed through oscilloscope traces, which revealed inconsistencies affecting the duty factor and cathode temperature. These fluctuations have downstream effects on the cesium-coated cathode surface, thereby impacting H⁻ ion production and beam reliability. Simulation-based testing in LTspice was used to evaluate noise suppression techniques and arc current regulation schemes, revealing how optimized snubber networks, improved pulse shaping, and feedback stability can mitigate modulator-induced noise. The results identified hardware level parameters that significantly enhance discharge repeatability and improve overall plasma performance under operational conditions.

Campos, Nathan [Unlisted, US; Fermilab]↗

Post-irradiation 1600°C Heating Test of AGR-1 UCO Fuel Kernels

Five bare kernels were selected from the U.S. Advanced Gas Reactor (AGR) 1 irradiation experiment Compact 5-3-1 to perform post-irradiation safety tests. The safety test involved heating the kernels in the Fuel Accident Condition Simulator (FACS) furnace in the inert atmosphere to a peak temperature of 1600°C and isothermally holding at this temperature for about 47 hours while collecting fission products released. This test was to assess the retention of fission products in bare kernels without the effects of the other TRISO layers (buffer, IPyC, SiC, and OPyC) or the graphitic matrix material. The bare kernels released nearly 100% of cesium and antimony, while they were able to maintain about 30% of europium, 50% of strontium, and the majority of cerium and ruthenium. In addition, about 48% of the calculated inventory of Kr-85 released during the test, indicating that kernels were capable of retaining a considerable fraction of fission gas K-85 during irradiation.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

ARC-100 Reactor Security-by-Design Summary 2025

This report applies the security-by-design methodology developed in a previous National Nuclear Security Administration–sponsored work to the ARC-100, a sodium-cooled fast reactor (SFR) being developed by ARC Clean Technology, Inc (ARC). The report contains no proprietary information specific to the ARC 100 reactor. The insights developed in this report are high-level, and generally applicable to other sodium fast reactor designs. The information presented here is the result of a qualitative safety-based analysis and would not inform any potential adversary beyond what would be found in a docketed safety analysis report. The scope of this present report covers ARC-100’s reactor core, used fuel storage, used fuel assembly wash station, cesium trap, sodium cold trap, noble gas decay tanks, used fuel dry storage facility, damaged fuel storage facility, and radioactive waste building. These systems are also compared to a generic SFR design assumed in the previous study. The security assessment results show changes in structures, systems, and components (SSCs) safety importance relative to the generic SFR SSCs. Several SSCs have higher importance rankings than others, and it is recommended that protection efforts are prioritized for these SSCs. Results from this work will furnish a technical justification for the feasibility of these solutions for the ARC reactor's design and, where applicable, identify any regulatory benefits conferred by the proactive design aspect within a risk management framework. This initiative will contribute to a more secure design of the ARC reactor and support its licensing process.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

FGMS-poster

Idaho National Laboratory (INL) performs post irradiation examination (PIE) of tri-structural isotropic (TRISO)-coated particle fuel to help qualify it for high temperature gas cooled reactors. TRISO fuel compacts are re-irradiated in the Neutron Radiography Reactor (NRAD) to generate the short lived fission products needed for fission product release testing. The Fuel Accident Condition Simulator (FACS) furnace and newly added Screen Neutron Irradiated Fuel for Failure (SNIFF) furnace heat the compacts in helium to temperatures of up to 2,000°C, prompting fission product release—predominantly gaseous xenon and krypton isotopes and condensable products such as cesium—from failed particles. These released isotopes are transported to a fission gas monitoring system (FGMS 1 or FGMS 3), where they accumulate in cryogenic cold traps and are quantified using high-purity germanium (HPGe) detectors. The addition of SNIFF and FGMS 3 increases throughput by enabling simultaneous testing of multiple compacts. Furthermore, automated INL developed software provides continuous, near-real time monitoring of fission product inventories and manages the liquid nitrogen cooling of the traps. These system enhancements improve the efficiency, data quality, and testing capacity of TRISO fuel performance evaluations.

07 - ISOTOPES AND RADIATION SOURCES↗

Application of Nuclear Technology to Art Identification Problems: First Annual Report

Throughout the centuries, as man has become increasingly affluent, his interest in antiquities and objects of art, and the price he has been willing to pay for them, has increased. Inevitably, one result has been to attract the unscrupulous to the production and sale of counterfeit paintings, sculptures, and other works of art. The skill of forgers varies but often is highly sophisticated. However, dealers, museum curators, and other experts have also become more skillful in the detection of forgeries. Scientific tools of increasing sensitivity and sophistication have gradually been applied to the examination of the materials of art and archaeology. Such tools frequently support and render more reliable the judgment of the experts. The quality of forgeries may improve further as their makers become acquainted with the new methods of detection and in turn learn to circumvent or confound the methods. Therefore, the development of still more advanced methods of examination has great utility in the art world. This is particularly true if the ultimate effect is to make circumvention of the methods of examination so costly that the economic incentives for producing forgeries may be substantially reduced, if not eliminated.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Comparison and investigation of production of specific nuclides by different fission yield data

Generally, passive gamma spectroscopy (PGS) is known as a useful technique for nondestructive measurement of burnup and nuclide production in normal spent nuclear fuel. However, PGS can be used not only for normal spent fuel, but also for molten fuel generated in severe accidents involving core meltdowns. In such cases, it is estimated by the radioactivity ratios of specific nuclides. At the time of the decommissioning of the Three Mile Island's unit 2 reactor, fission products (FPs) such as Cs{sup 134}, Cs{sup 137}, Ce{sup 144}, Eu{sup 154}, Ru{sup 106}, and Sb{sup 125} were concerned. The radioactivity ratios are estimated by numerical simulations. However, there are various uncertainties in numerical simulations. In this study, we evaluated the effect of the difference in fission yield data. We compared the productions of concerned nuclides of Three Mile Island's unit 2 reactor by using Tokyo-tech and JENDL FP Fission Yields Data File 2011(FPY- 2011). Tokyo-tech is newly created by Tsubakihara et al. at Tokyo Institute of Technology. FPY-2011 is currently the latest fission yield data in Japan. In the concerned nuclides, the relative difference (Tokyo-tech/FPY-2011 -1) in the production of Sb{sup 125} is 4.5% and that of Eu{sup 154} is -5.6%. The relative difference in the production of Sb{sup 125} is found to be caused by the difference in the independent yields of Sn{sup 125}, In{sup 125}, Cd{sup 125}, Ag{sup 125}, and Pd{sup 125}. The relative difference in Eu{sup 154} is found to be caused by the relative difference in the production of Sm{sup 152} and the relative difference in the cumulative yield of Eu{sup 153}.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Enhanced prediction of Cs removal by CST from Hanford tank waste with K accountability

Abstract The treatment of Hanford tank waste is one of the most challenging environmental cleanup activities to date. To expedite the processing of liquid waste stored in underground tanks in Washington State it is necessary to remove the significant dose contributor, 137 Cs. Crystalline silicotitanate ion exchanger is currently used to remove 137 Cs from the aqueous phase of Hanford tank wastes in preparation for vitrification at the Waste Treatment and Immobilization Plant (WTP). Improving the understanding of potassium impacts on ion exchange behavior of Cs will help in the operation of a critical component of one of the most complex treatment processes in the world today. Optimization of this process can result in significant cost savings and less waste production. Toward this effort, a series of batch contact tests varied in potassium concentration were conducted to look at the impact of potassium concentration on Cs distribution. Experimental distribution ratios ( K d ) were compared to the distribution ratios predicted using the ZAM model. A significant underprediction of Cs capacity in the presence of potassium was seen with the existing model. A revision of the equilibrium constants was determined and provided a statistically better fit for determining the Cs K d values in tank waste matrices.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Identification, isolation, and characterization of a novel type of Fukushima-derived microparticle

Abstract In the course of the Fukushima nuclear accident, radionuclides were released in various forms, including so-called radiocesium-bearing microparticles (CsMP). So far, four types of CsMP were described: Type A is smaller in size (< 10 μm), Types B, C, and D are larger (> 100 μm). In this work, we present a novel type of CsMP (proclaimed Type E). Three particles of Type E were extracted from a contaminated blade of grass that was sampled 1.5 km from the Fukushima Daiichi nuclear power plant in late 2011. They were located using autoradiography, isolated using an optical microscope and micromanipulator, and characterized using scanning electron microscopy, energy dispersive x-ray spectroscopy, and low-level gamma-ray spectrometry. Type E CsMPs are 10–20 μm in size and exhibit an unusually low and barely detectable 137 Cs activity of only ≤ 10 mBq per particle. Their brittle and fragile character may indicate a high surface tension.

137Cs↗

Penetration of Fission Products Ions into Complex Solids and the Effect of Ionic Wash Methods

During washing of radiologically impacted building surfaces, penetration of radionuclide ions into complex solids associated with these surfaces may occur. This study investigates the penetration of Cs-137, Sr-85, and Eu-152 solutions into numerous common building materials and radionuclide behavior when these materials were exposed to a static bath or low-pressure flow of tap water, 0.1 M potassium chloride (KCl), and 0.5 M KCl. The decontamination efficacy and the depth profile for residual contamination were measured to determine the conditions under which applying a wash solution has benefit compared to physically removing the surface material. On asphalt, 70-80% of the radionuclides were found to be within 0.02 mm of the surface. Concrete is more porous than asphalt, and 80% of the radionuclides were within 0.2 mm of the surface for Cs-137 and Eu-152 and 50-80% for Sr-85. Water effectively removed all contaminants from hard nonporous surfaces. Finally, this paper illustrates that a wash penalty factor concept-defined as ratio of the depth at which 50% of the radioactivity is found in the washed sample divided by the depth at which 50% of radioactivity is found in the control-can serve as a way to quantify whether the wash method increases the depth at which contamination penetrates into the material and thus the material becomes more difficult to decontaminate.

building materials↗