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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Gepta-EX: a multi-channel germanium detector for X-ray absorption fine structure

Fluorescence-mode X-ray absorption spectroscopy (XAS) at high photon energies requires detectors with high stopping power and excellent energy resolution to measure weak element-specific signals. Traditional silicon-based detectors suffer from poor efficiency above ∼20 keV, while most high-Z materials such as cadmium telluride, cadmium zinc telluride, and gallium arsenide have focused predominantly on imaging applications with limited spectroscopic resolution. In contrast, germanium combines excellent stopping power with superior intrinsic energy resolution due to its low Fano factor and favorable charge transport properties, making it ideal for high-resolution spectroscopy in the 15 keV to 100 keV range. To address these requirements, we report on the development, fabrication, and performance evaluation of Gepta-EX, a compact multi-channel high-purity germanium (HPGe) detector system for fluorescence-mode XAS. The Gepta-EX system features a monolithic seven-channel HPGe pixel array integrated with low-noise CUBE charge-sensitive preamplifiers, housed within a thermally isolated compact cryostat operating near 90 K. The detector achieves energy resolutions of 218 eV at 5.9 keV (from a 55 Fe source) and 373 eV at 59.5 keV (from a 241 Am source), with uniform performance across all channels. By avoiding the escape peak interference commonly associated with silicon detectors and providing stable, high-resolution performance, Gepta-EX represents a powerful tool for high-energy X-ray spectroscopy.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Identify and Assess Technical Challenges in Safeguards Measurements of Spent Advanced Reactor Fuels

Advanced reactor (AR) designs use various nuclear fuel types that can be significantly different than conventional light-water reactor (LWR) fuels, including differences in sizes, compositions, and chemical forms (e.g., oxide, carbide, metal). Nearly all the proposed AR fuels use high-assay low-enriched uranium (HALEU), which will have higher enrichments (5–20 wt% 235 U) than LWR fuels (currently limited to <5 wt% 235 U). In advance of the wide use of these new fuel types around the world, international safeguards organizations such as the International Atomic Energy Agency (IAEA]) are working with some of the AR vendors to formulate safeguards approaches for these AR fuel cycles. As part of the overall safeguards approach, it is important to identify the potential technical challenges in performing safeguards verification measurements of these AR fuels (both fresh and spent fuels) in advance of the widespread adoption of these new fuel types, because new safeguards technologies can take several years to develop, test, and approve for use. This report documents work performed in fiscal year 2024 based on modeling and simulation to assess the performance of the existing safeguards measurement technologies for irradiated or spent AR fuel elements or items. This work is a continuation of the work performed in fiscal year 2023 that focused on fresh AR fuels. Spent AR fuels have a distinct difference from their LWR counterparts: unlike the spent LWR fuels typically stored in a water-filled pool, some spent AR fuels—such as tristructural-isotropic (TRISO)-based fuels—will most likely be stored in air-filled hot cells. Because most safeguards measurements on spent fuel performed to date have been conducted under water, the air-filled hot cell environment could present unique challenges to safeguards measurements. Fork detector (FDET) and Cerenkov viewing device (CVD) systems have been the two primary instruments used by the IAEA for several decades to measure spent LWR fuel assemblies stored in pools for safeguards verification purposes. Because the lower refractive index of air causes Cerenkov light to be of lower intensity in air than in water, existing CVDs are likely unable to perform safeguards verification measurements for spent fuel stored in an air-filled hot cell, as is the case for the TRISO-based spent fuel elements (e.g., pebbles, graphite fuel blocks). Unlike FDET measurements, CVD measurements do not require fuel be moved, so they are a simpler and faster to take than FDET measurements. The inability to perform CVD measurements on the TRISO-based AR fuel types presents a major technical challenge in the effort to use existing technology to perform safeguards measurements on spent AR fuels. This study was mainly conducted through the modeling and simulation of an FDET or an FDET-like system on five spent AR fuel types, including one metallic fuel type and four TRISO-based fuel types in both pebble and graphite block forms in their respective storage configurations and environments. Because the various AR fuel types have significantly different dimensions, FDET systems must be adapted to accommodate them. Partial defect tests were also simulated in this study to assess the FDET’s ability to detect potential fuel diversions. The FDET measures the fuel’s total passive neutron and gamma emissions. The simulated FDET results from spent AR fuel items are compared against results from a typical spent pressurized water reactor (PWR) assembly. High-purity germanium (HPGe) gamma detector measurements were also simulated for the spent AR fuel types and the PWR assembly because the signature photopeaks have been used in LWR safeguards verifications, although HPGe is usually not used to detect diversions because of the fuel’s self-attenuation effects on those photopeaks. The results indicate that these detectors have significant challenges in performing safeguards measurements of the spent AR fuel items, including incompatibilities between AR fuel items and existing FDETs, lower neutron count rates, lower sensitivities to fuel diversions in certain AR fuel items, and significantly higher interference from a neighboring fuel item when the measurement is performed in air. These results suggest that an alternative technology or significant and timely technology development is needed to perform adequate safeguards measurements of some of these AR fuel items.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Development of an In Situ Fission Gas Release Instrument for Fuel Sample Irradiations in the High Flux Isotope Reactor

Experimental measurement of gaseous fission product release with respect to temperature and burnup is a critical aspect of understanding nuclear fuel performance, validating predictive models, and qualifying new fuels. To measure this phenomenon in real-time, Oak Ridge National Laboratory has developed an instrument for measuring in situ fission gas release from small-scale fuel samples irradiated in the High Flux Isotope Reactor (HFIR). The instrument uses a continuous flow of Heover the fuel samples to sweep gaseous fission products from a sealed capsule in the HFIR Be reflector to an instrument station adjacent to the reactor. The instrument station houses two high-purity germanium (HPGe) detectors that measure decay gamma rays from fission products passing through a room temperature dwell chamber placed over the detector crystal. The sealed capsules in the reactor are designed to modulate fuel sample temperatures between 700 and 1,100°C by changing the Ar/He gas mixture surrounding the capsules during irradiation. N-type thermocouples are incorporated into the capsule housing to record real-time fuel temperatures. The capsules are heated primarily by prompt gamma rays emitted from the HFIR core with minimal heat contributions from fission in the fuel samples to minimize temperature gradients in the specimens for separate-effects characterization of the material. This paper describes modeling of time-dependent nuclear heating and fission product formation in fuel samples, thermal characteristics of the in-core capsules, and expected gaseous fission product gamma spectra at the HPGe instrument station.

Mulligan, Padhraic L [ORNL] (ORCID:000000025826540↗
DOE OSTI2025

Characterization of the Multi-Coincident Bismuth Germanate Oxide Spectrometer at the Nuclear Counting Facility

A multi-detector γ-ray spectroscopy system including 3 high-purity germanium (HPGe) detectors has been developed at the Nuclear Counting Facility at Lawrence Livermore National Laboratory. This system operates not only in conventional singles γ-ray detection mode but also incorporates features like coincidence and anti-coincidence techniques. These advanced detection modes can significantly enhance the system’s sensitivity, making it a powerful tool for isotope identification and quantification, especially for complex samples with high background and overlapping peaks. In this re port, we present a comprehensive performance characterization of the Multi-Coincident Bismuth Oxide Spectrometer (MCBOS) using a variety of sample types, including 60 Co, 48 Sc, and mixed fission product samples. Each sample provides unique tests for the MCBOS capabilities. Additionally, a GEANT4 model of the MCBOS has been generated to aid with the characterization of the system.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Precise half-life determination of 131 Ba and 125 Xe

Barium-131 and Xenon-125 are valuable radioisotopes with numerous scientific and medical applications. Here, we produced both isotopes via proton irradiation of a cesium iodide (CsI) target at the Brookhaven Tandem Van de Graaff accelerator (BTVG). Following irradiation, we conducted systematic γ-ray spectroscopy measurements at the National Nuclear Data Center (NNDC) decay station using a calibrated High-Purity Germanium (HPGe) detector, collecting data over a 2-month period (spanning approximately 4.5 half-lives for 131 Ba). Through careful analysis of the characteristic γ-ray emissions from both isotopes, we determined the half-life of 131 Ba to be 11.55(6) days, which agrees with the previously evaluated value. For 125 Xe, we measured a half-life of 16.56(8) hours, slightly lower than the evaluated value.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Time-interval distributions in a digital gamma multi-channel analyzer at extreme input rates

Here, this study explores the relationship between experimentally observed inter-event time-interval distributions (TIDs) and measured dead times for pulses generated by a HPGe gamma detector and processed by a digital spectrometer. The system utilizes a fast and slow channel, pile-up rejector, trapezoid filtering, and flash analog-to-digital converter. The experimentally derived TIDs were compared with theory for validation. The results demonstrate that the theoretical model reliably describes measured TIDs up to 40% dead time. However, significant distortion effects become increasingly pronounced at higher input rates. It appears that the deviation between the measured and calculated time-interval distributions take the shape of higher-order convolutions of the TID, which represent the time difference between counts for more than two successive events. In this work, theoretical functions for the TID are expanded to reproduce measurements up to 90%. This refinement in the interpretation and treatment of the measured TIDs provides improved accuracy and precision in the prediction of the true event rate and measured dead time in the counter. Even though it reflects occasional count loss that is specific to the set-up, it is expected that a similar adjustment may be applicable to other detection systems as well.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

Distribution of Natural Radionuclides at the Nevada National Security Site

According to the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) publications, contributions of terrestrial gamma doses are mainly from the presence of 40 K, and of 238 U and 232 Th together with their progeny in various rocks and soils. A survey of soil distributions of radionuclides 40 K, 238 U, and 232 Th was performed at the Nevada National Security Site (NNSS) using in situ gamma-ray spectrometry with a high-purity germanium (HPGe) detector. The average activity concentrations of 40 K, 238 U, and 232 Th in natural soils at the NNSS are 867 Bq kg -1 (range from 150 ± 8 to 1297 ± 56 Bq kg -1 ), 50 Bq kg -1 (range from 29 ± 3 to 74 ± 8 Bq kg -1 ), and 56 Bq kg -1 (range from 11 ± 2 to 96 ± 10 Bq kg -1 ), respectively. The concentration at each location is significantly associated with its geological lithology. The terrestrial gamma dose rates around the NNSS were estimated from 26 to 144 nSv h -1 with mean value of 93 nSv h -1 . In conclusion, our results provide useful information about the natural background radiation and radiological effects of naturally occurring radionuclides at the NNSS.

environmental assessment↗
DOE OSTI2025

Measurements of short-lived fission product yields from photofission of 238 U using 13.0 MeV monoenergetic photons

Photon-induced fission product yield (FPY) measurements were conducted on the isotope 238U. Fission was induced using Eγ = 13.0 MeV monoenergetic photons produced by the Triangle Universities Nuclear Laboratory’s (TUNL’s) High Intensity γ-ray Source (HIγS) facility. Short-lived FPYs were measured by performing cyclic activation of the sample using a rapid target transfer system. Following activation, the 238U target was rapidly (0.4 s) transferred to a counting station consisting of two well-shielded high-purity germanium (HPGe) detectors. The irradiation-counting cycle was repeated until the summed data had sufficient statistical accuracy. Twenty-eight unique fission products with half-lives ranging from 1 s to 450 s were identified, and their cumulative FPYs determined. Furthermore, the results are compared with previous independent FPY measurements using inverse kinematics. Good agreement between the data sets is found despite the different excitation energy distributions of the fissioning nucleus in the experiments.

Physics - Nuclear physics and radiation physics↗
DOE OSTI2026

Forward modeling approach to nuclear reaction cross sections: Applications in neutron inelastic scattering

The development of nuclear reaction models for the production of evaluated nuclear data has traditionally been performed by comparing measured cross sections with predictions from reaction model codes whose physical input parameters are adjusted to obtain the best agreement between measured and modeled results. To more directly probe reaction model inputs, this work introduces a forward modeling approach to experimental reaction cross-section determination, where the most important physical input parameters to reaction model calculations are obtained via 𝜒 2 minimization between measured and calculated observables. This was demonstrated using data collected by the Gamma Energy Neutron Energy Spectrometer for Inelastic Scattering (GENESIS) at the 88-inch cyclotron at Lawrence Berkeley National Laboratory, a detection array consisting of organic liquid scintillators and high-purity germanium (HPGe) detectors. Using a broad-spectrum neutron beam and a 99.98%-enriched 56 Fe target, GENESIS was used to perform a simultaneous measurement of 56 Fe 𝛾-ray production cross sections and secondary neutron energy and angle distributions. The results of the forward modeling approach to the determination of energy-differential 𝛾-ray production cross sections for the yrast 4 + → 2 + and 6 + → 4 + transitions, as well as eight other off-yrast transitions, were compared against those obtained using conventional techniques, and the results are in good agreement. In addition to discrete 𝛾-ray yield total scattered neutron energy-angular distributions as a function of incident neutron energy were also obtained using forward modeling and found to agree with evaluated data, with the exception of elastic scattering at small angles. The fitted reaction model parameters obtained through forward modeling were also used to calculate the cross section for the unobserved (𝑛, 2⁢𝑛) reaction; excellent agreement with the current evaluation was obtained, providing a validation of the predictive capabilities of the forward model approach. This work bridges the gap between nuclear data experiment and evaluation by providing a new means for extracting inelastic neutron-scattering cross sections and neutron-induced 𝛾-ray production data while directly probing reaction model physics.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Precise absolute 𝛾–ray intensities following the 𝛽 decay of 156 Eu

The most recent 𝛽-decay data for 156 Eu is over 40 years old, and the 𝛾-ray intensities are only known with a fractional precision of about 10% or worse. Here, in this work, we present new results for the absolute intensities of more than 60 𝛾 rays emitted following the 𝛽 decay of 156 Eu with many measured to ≲1% accuracy, representing an order-of-magnitude improvement in precision. This was achieved by leveraging the fission-product beams produced at the CARIBU facility to create radio-pure samples and detecting the radiation with a 4⁢𝜋 𝛽-particle counter coupled with a precisely efficiency-calibrated HPGe detector at Texas A and M University.

Physics - Nuclear physics and radiation physics↗
DOE OSTI2026

Optimization of an Energy Tuning Assembly for High Explosives Detection

The Portable Isotopic Neutron Spectroscopy (PINS) system, employs neutron-induced gamma-ray spectroscopy and provides a nondestructive method for high explosives detection. In standard operation it uses Californium-252 as a neutron source. Operating PINS with a deuterium-tritium (DT) neutron generator has some advantages over Cf-252, including lifetime and ability to produce high-energy inelastic scattering gamma rays. However, current systems using DT neutron generators suffer from a high environmental background and reduced ability to induce neutron capture, reducing spectral quality and limiting nitrogen sensitivity. Here, this study presents the development of an energy-tuning assembly (ETA) designed to optimize the DT neutron energy spectrum to increase nitrogen reaction rates in a target, thereby improving high explosive detection capabilities. A metaheuristic optimization framework, MultiGNOWEE, coupled with MCNP, was employed to generate two ETA configurations: a single-objective ETA, which maximizes nitrogen capture reactions, and a multi-objective ETA, which balances neutron capture and inelastic scattering. Simulations demonstrated the optimized configurations achieved up to a 10-fold improvement in nitrogen capture rates compared to the bare configuration. Experimental validation was conducted using a DT neutron generator and a high-purity germanium (HPGe) detector. Two prototype ETAs were constructed and assessed on a melamine simulant. Measurements demonstrated improved nitrogen detection for both prototype ETA configurations when compared to the standard system.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2026

Evaluation of Compton suppression for enhancing trace element identification in neutron activation analysis of reference materials

Neutron activation analysis (NAA) is a powerful technique for identifying and quantifying trace elements in materials. However, challenges such as high dead times, spectral interferences, and high Compton continuum often arise. This study employs a Compton suppression system (CSS) to enhance NAA sensitivity by reducing the Compton continuum, thereby improving the peak-to-Compton ratio. National Institute of Standards and Technology certified reference materials 1632d, 1633c, and Canadian National Research Council TORT-1 were irradiated in a thermal and epithermal neutron flux under various irradiation, decay and counting times and analyzed using high-resolution gamma-ray spectroscopy with and without Compton suppression. The reduction factor was calculated to evaluate the effectiveness of the CSS, demonstrating significant background reduction and improved detection limits for trace elements. Additional experiments with a 137Cs point source demonstrated the impact of detector-source geometry on system performance, showing a decrease in the reduction factor as the source was moved further from the NaI detector. An optimum distance between the source and HPGe detector was observed, yielding the highest peak-to-Compton ratio. The results highlight the CSS's ability to minimize spectral interference and enhance elemental identification.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Gamma-ray Exportable Adjustable Spectrum Evolution

An initial number of fissions on a fissile/fissionable material, defined by ENDF library availability, is used to produce an initial inventory of fission products (FPs). The FPs are decayed using full, analytical solutions to the Bateman equations to calculate the FP inventory at a user-specified time. The radiation detector, a high-purity germanium (HPGe) detector, has been modeled using every gamma-ray in the ENDF/TENDL libraries individually, and the modeled response can be multiplied by the activity of each fission product to determine a full, energy dependent detector response, including dead time. The detector response can be output as a PeakEasy (.spe) file, or as a text file. Since each radioisotope's inventory is known, the proportion to the total may be shown for each isotope, as well as many radionuclides in a user-specified energy window. Full execution and display of results takes 1-3 seconds, depending on the complexity of the input parameters.

Holschuh, ThomasV [Idaho National Laboratory (INL)↗
DOE OSTI2025

Deuteron-induced reactions on natural Zr from threshold to 50 MeV: production of 86g Y

Two stacks of thin Zr foils were irradiated with 30 and 50 MeV deuterons, respectively, using the Lawrence Berkeley National Laboratory 88-Inch Cyclotron, and 19 excitation functions for nat Zr(d,x) reactions were measured over a beam energy range of 6.3–47.64 MeV, where the independent cross sections for nat Zr(d,x) 88 Nb and nat Zr(d,x) 86m,g Y were measured for the first time. The well-characterized nat Fe(d,x) 56 Co, nat Ni(d,x) 56 Co, nat Ni(d,x) 58 Co, nat Ni(d,x) 61 Cu, nat Ti(d,x) 46 Sc and nat Ti(d,x) 48 V monitor reactions were used to determine the deuteron beam current throughout the stacks. All cross sections were determined using High Purity Germanium (HPGe) detector γ-ray spectroscopy. A variance minimization technique was employed to simultaneously constrain the deuteron beam currents with multiple monitor reactions, thus reducing systematic uncertainties. An additional 16 channels are reported for reactions on the nickel, titanium, and iron monitor foils, leading to a total of 35 excitation functions, with seven reaction channels reported for the first time in this work. The measured excitation functions are compared to calculations provided by the reaction modeling codes TALYS – 2.0, ALICE – 2020, CoH – 3.5.3 and EMPIRE – 3.2.3, as well as the TENDL – 2023 data library. The degree of agreement between theory and experiments is discussed. The possible production of the important PET radionuclide 86g Y via the nat Zr(d,x) route was critically examined. The physical yields for nat Zr(d,x) 86 Y and other yttrium isotopes produced were calculated and compared to other production pathways. Due to high-level of associated radionuclide impurities, this route cannot deliver 86g Y suitable for medical applications.

86gY↗
DOE OSTI2025

Analysis of Radioactive Waste and Contamination using 3D Position-Sensitive CdZnTe Detectors

INTRODUCTION The US has surplus plutonium. This material will be diluted in gloveboxes at SRS [Ref 1 and 2]. This process will create holdup. Holdup is material in an unexpected or unwanted location. Because this holdup material is primarily PuO2, it must be accounted for (measured). This measurement is typically done using a liquid-nitrogen cooled, HPGe detector. To enable more rapid processing, advanced holdup technologies are being explored [Ref 3 and 4]. METHODS For the surplus plutonium dilute and dispose program, the traditional detector system could slow down the optimal operational tempo of the program. A multiple-detector system, providing complete glovebox coverage, based on CdZnTe crystals, 3D-positioning electronics, and a coded aperture mask, is in the process of being developed by ORNL and SRNL [Ref 4].

Whiteside, Tad [Savannah River National Laboratory↗
DOE OSTI2025

FRAM v.7.1

Fixed-energy Response-function Analysis with Multiple efficiency (FRAM) is a software code designed primarily for plutonium and uranium isotopic analysis. It is widely used in both the domestic and international safeguards community. FRAM can quickly and accurately determine the isotopic compositions of plutonium, uranium, and mixed oxides (MOX) when measuring with a high-purity germanium (HPGe), cadmium zinc telluride (CZT), or lanthanum bromide (LaBr3) detector. The capabilities of FRAM have been enhanced to analyze the data of the pixelated CZT detector (made by H3D) and to measure the mass of plutonium, uranium, and MOX. Both the isotopic composition and mass of the item can be quickly determined with one measurement using a gamma detector with FRAM v.7.1.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Assessment of Naturally Occurring Radioactive Material (NORM) in Illicit Drugs Using High-Resolution Gamma Spectroscopy

This report documents results assessing the presence of Naturally Occurring Radioactive Material (NORM) in illicit drugs using high-resolution gamma spectroscopy. Specifically, high-purity germanium (HPGe) measurements were conducted at the Drug Enforcement Administration (DEA) South Central Laboratory on 11 samples (referred to as exhibits) including cocaine, fentanyl, heroin, ketamine, LSD, marijuana, methamphetamines, psilocybin mushrooms, and synthetic cannabinoids.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

GADRAS-DRF Validation for Safeguards and Custom Peak Fit Enhancements

In previous years, SGTech funded enhancements to the isotopics routine in the software called Gamma Detector Response and Analysis Software-Detector Response Function (GADRAS-DRF), including the addition of peak fit customization capabilities. A project was also funded that focused on implementing a peak-based model fitting routine, allowing model fitting to be performed without dependence on export-controlled cross-sections. In FY25 significant improvements were made to the custom peak fitting interface, accompanied by several validation studies within GADRAS-DRF. These studies encompassed IsotopeID performance, distributed source analysis, isotopics validation, and activity estimation. Additionally, the peak-only model fitting option was validated using an HPGe measurement of a rotating drum with line sources.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗