Abstract Rapidly identifying individuals who have received internal radiation exposure above action guidelines is crucial for mitigating health risks and addressing public concerns immediately following a radiological event involving the dispersal of radioactive materials. This study describes a novel triage method using a conventional Geiger-Mueller (GM) detector to select those individuals from the large group of persons who may have received an intake of radioactive material at levels corresponding to one Clinical Decision Guide (CDG). The triage method involves placing a portable GM detector against the lower anterior torso of a sitting person as they bend over to surround the detector with their body. The response of the GM detector is evaluated using a new, specially designed anthropometric phantom that simulates combined tissues of the lower thorax and gastrointestinal (GI) tract and is fabricated with a tissue substitute material that matches the overall radiological properties of human tissue present in this body region. The phantom has four separate layers of tissue substitute material with ports to accommodate a single GM detector at the center and one or more sealed radioactive sources that can be arranged to characterize the detector response for a variety of source distributions, including a “hot spot.” In this study, the response of a Ludlum Model 133-4 GM detector was evaluated using sealed sources of 232Th and 137Cs to determine the measurement efficiency for a quantity of activity present in the abdomen within a few hours post-intake equivalent to 1 CDG. Results demonstrate that the Quick Sort triage procedure using a single GM detector placed against the abdomen of a person can reliably detect internal deposition resulting from an intake equivalent to 1 CDG for 232Th or a significantly lower activity of 137Cs within a few hours following a radiological incident. The evaluation was performed over a wide range of photon energies, so the Quick Sort triage procedure is expected to be suitable for most fission products distributed uniformly within the abdomen or as a single “hot spot.”

The absolute intensity for the 803-keV γ ray of 210Po was evaluated by α-γ coincidence technique. A liquid sample with a known amount of 210Po embedded in scintillation fluid was measured in a coincidence-based system that comprises a Liquid Scintillator (LS) detector and a High-Purity Germanium (HPGe) detector. A photo-reflector assembly that contains the 210Po sample provides 100% efficiency for detecting the α particles. The combination between the HPGe and the LS detectors allows to reject non-coincident α-γ events while maintaining high resolution γ spectroscopy. Consequently, the faint 803-keV photopeak from 210Po could be observed in a background-free environment, and its intensity could be evaluated with good accuracy. Sample measurements were carried out over nine months to gather statistics and verify the reliability of the experimental procedure. The absolute intensity of the 803-keV line was found to be (1.22 ± 0.03) × 10-5, in excellent agreement with the adopted value in a recent data compilation and consistent with previous experimental works.

We present first preliminary results of a novel method for measuring independent isotopic fission yields (IIFYs) of spontaneous fission (SF) via direct mass measurements, at the FRS Ion Catcher (FRS-IC) at GSI. Fission products were generated from a 252Cf source installed in a cryogenic stopping cell, and were identified and counted with the multiple-reflection time-of-flight mass spectrometer (MR-TOF-MS) of the FRS-IC, utilizing well-established measurement and data analysis methods. The MR-TOF-MS resolves isobars unambiguously, even with limited statistics, and its non-scanning nature ensures minimal relative systematic uncertainties amongst fission products. The analysis for extracting IIFYs includes isotope-dependent efficiency corrections for all components of the FRS-IC. In particular, we applied a self-consistent technique that takes into account the element-dependent survival efficiencies in the CSC, due to chemical reactions with the buffer gas. Our IIFY results, which cover several tens of fission products in the less-accessible high-mass peak (Z = 56 to 63) down to fission yields at the level of 10−5, are generally similar to those of the nuclear database ENDF/B-VII.0. Nevertheless, they reveal some structures that are not observed in the database smooth trends. These are the first results of a planned campaign to investigate IIFY distributions of spontaneous fission at the FRS-IC. Upcoming experiments will extend our results to wider Z and N ranges, lower fission yields, and other spontaneously-fissioning actinides.

The Israeli National Nuclear Forensics Laboratory (INNFL) investigated four unknown samples during the Collaborative Material Exercise 7 (CMX-7). The samples, identified as depleted uranium, were marked ES-1 to ES-4. ES-2 isotopic composition (0.200 ± 0.004

A ground-level prototype system for low-background measurements was developed and tested. The system consists of a high-purity germanium (HPGe) detector used for detecting γ rays and coupled to a liquid scintillator (LS) used for detecting α and β particles. Both detectors are surrounded by shielding materials and anti-cosmic detectors (“veto”) used to suppress background events. The energy and timestamp of detected α, β and γ emissions are recorded event-by-event and analyzed offline. By requiring timing coincidence between the HPGe and LS detectors, background events originating from outside the volume of the measured sample can be effectively rejected. The system performance was evaluated using liquid samples containing known activities of an α emitter (241Am) or a β emitter (60Co) whose decays are accompanied by γ rays. The LS detector was found to provide a solid angle of almost 4π for α and β particles. Compared to the traditional γ-singles mode, operating the system in coincidence mode (i.e., α-γ or β-γ) reduced the background counts by a factor of ∼100. Consequently, the minimal detectable activity for 241Am and 60Co was improved by a factor of 9, being 4 mBq and 1 mBq for an 11-d measurement, respectively. Furthermore, by applying a spectrometric cut in the LS spectrum that corresponds to α emission from 241Am, a background reduction factor of ∼2400 (compared to γ-singles mode) was achieved. Beyond low-background measurements, this prototype exhibits additional compelling features, such as the ability to focus on certain decay channels and study their properties. This concept for a measurement system may be of interest to laboratories that monitor environmental radioactivity, studies involving environmental measurements and/or trace-level radioactivity.

In this study, a high-fidelity Monte Carlo model was utilized to investigate the localized behavior of the power peaking factor in a reactor core. The model considered the effect of water gaps, also known as flux traps, on the local power peaking. The model, developed using the Serpent code, employs a fine spatial mesh to accurately describe the fission power and burnup distribution. The results obtained from the model were validated through comparison with experimental measurements of the power distribution obtained from the IRR1 facility. The study found that the local accumulation of thermal flux caused by enhanced neutron moderation in the flux trap leads to a highly localized increase in power density, affecting the power peaking factor. The quantification of this effect was a key finding of the study. The results obtained from the study were highly dependent on the model's fidelity, with significant differences being observed between the power peaking factor calculated on a fine mesh as opposed to that calculated at the plate or assembly scale. Furthermore, the experimental validation of the model enabled the prediction of the power peaking factor in other regions within the fuel assembly, which are not accessible to measurement instrumentation, with a high degree of confidence. The study's findings may be useful for optimizing reactor design and operation and assessing the safety margins in reactors with similar characteristics.

The absolute intensity for the ‘prompt’ 140.5 keV gamma-ray of 99Mo was evaluated using the β-γ coincidence technique. A liquid sample of 99Mo was prepared from a99Mo/99mTc generator and measured in a 4παβ(LS)-γ(HPGe) system that comprises a Liquid Scintillator (LS) detector and a High-Purity Germanium (HPGe) detector. The sample was introduced into scintillation fluid embedded in a photo-reflector assembly that provides almost 100% efficiency for detecting β particles (in the energy range of intreset). The combination of the HPGe and the LS detectors provided a highly effective rejection mechanism for non-coincident events. Thereby, the distinction between the detected 140.5 keV events originating from decays of 99mTc (IT) and those from transitions bypassing the metastable state could be obtained and the ‘prompt’ intensity was evaluated directly. The system was calibrated for detecting β particles and γ-rays using radioactive sources of known activities and having identical geometry as the sample containing 99Mo. The absolute intensity of the ‘prompt’ 140.5 keV was found to be (5.21 ± 0.02stat±0.16sys)%, in good agreement with results from more recently reported works.

Irradiations of bismuth samples with thermal and epithermal neutrons have been performed for a comparison of activation to the Bi-210g ground state and to the Bi-210m metastable state. The bismuth irradiations were conducted at the Soreq IRR1 research reactor at a near-core location with integrated neutron flux of approximately 2 x 10(18) neutrons/cm(2 )and a gold cadmium ratio of 3.1. Two nearly identical bismuth samples were irradiated, one without and one with a cadmium shield, to enable a comparison between thermal and epithermal neutron capture cross sections. Subsequent gamma spectrometry indicates nearly equal levels of activation to the ground and to the metastable states of Bi-210, both with thermal and epithermal neutrons. Absolute thermal cross sections and resonant integrals were deduced by including previously reported bismuth activation measurements performed conjointly with gold foil activation for normalization. A comparison of the present results with existing data is presented. It is argued that the bismuth resonances lie in the neutron energy range relevant for stellar temperatures, and therefore the neutron capture cross section ratio (sigma(g)/sigma(m)) measured for the epithermal neutrons is relevant also for stellar Bi-210 production. This is of importance since bismuth lies at the end of the chain for the s-process reaction for nucleosynthesis in stars. Precise knowledge of bismuth activation to the Bi-210m metastable state is also vital for the planning of GEN-IV reactors given the 3-million-yr half-life of Bi-210m.

The absolute intensity for the 'prompt' 140.5 keV gamma-ray of 99Mo was evaluated using the beta-gamma coincidence technique. A liquid sample of 99Mo was prepared from a99Mo/99mTc generator and measured in a 4 pi alpha beta(LS)gamma(HPGe) system that comprises a Liquid Scintillator (LS) detector and a High-Purity Germanium (HPGe) detector. The sample was introduced into scintillation fluid embedded in a photo-reflector assembly that provides almost 100% efficiency for detecting beta particles (in the energy range of intreset). The combination of the HPGe and the LS detectors provided a highly effective rejection mechanism for non-coincident events. Thereby, the distinction between the detected 140.5 keV events originating from decays of 99mTc (IT) and those from transitions bypassing the metastable state could be obtained and the 'prompt' intensity was evaluated directly. The system was calibrated for detecting beta particles and gamma-rays using radioactive sources of known activities and having identical geometry as the sample containing 99Mo. The absolute intensity of the 'prompt' 140.5 keV was found to be (5.21 +/- 0.02stat +/- 0.16sys)%, in good agreement with results from more recently reported works.

A rapid method for determining low activity concentrations of 210Pb in drinking water was developed and tested. The method consists of a few stages for sample preparation that involve passing 12 L of water through a column with acrylic fibers implanted with MnO2 (used to adsorb 210Pb). The MnO2 fibers are oven-dried, compressed and measured by a broad-energy germanium detector used to quantify 210Pb via its characteristic 46.5 keV γ-ray. The time taken for sample preparation is approximately 4 h and recovery factors for lead in tap water of 87 ± 3% were achieved. After a measurement duration of 4 h, the minimum detectable activity concentration reaches 0.02 Bq/L for 210Pb, being well below the respective limit for drinking water in Israel (0.2 Bq/L) as well as the value recommended by the World Health Organization (0.1 Bq/L). Furthermore, a measurement duration of 48 h provides a minimum detectable activity concentration of ∼0.006 Bq/L, which is similar in magnitude to other, well-established methods that rely on lengthy and rather complex procedures. Thus, the combination of MnO2 fibers and gamma-ray spectrometry may be attractive for routine use by analytical laboratories that monitor radioactivity in drinking water.

A new method has been developed to identify and localize a single hot particle in the lungs using an array of four high-purity germanium detectors. The method is based upon calculating a set of three count rate ratios (generated by each individual detector in the array) that are evaluated in sequence to designate whether the measured deposition can be associated with a hot particle rather than the default assumption of a uniform activity distribution. Identification and localization of the hot particle are determined from a single in vivo measurement in which detectors are positioned above and below the thorax. The method was tested using an anthropomorphic thorax phantom in which point sources of 241Am, 137Cs and 60Co were individually inserted in the lungs at 15 different locations and were measured using a scanning bed whole-body counter. Depending upon source location and photon energy, a bias of -35% up to +76% could be introduced by falsely assuming a uniform activity distribution in the lungs. This bias would directly translate to an erroneous dose estimate to the lungs. It was demonstrated that by using the appropriate detector efficiencies for the single hot particle, the bias associated with the activity determination is reduced to <10% and ~2% in average.

The objective of this paper is to describe nuclear forensics material characterization, as conducted in the Israeli National Nuclear Forensic Laboratory (INNFL) during an international exercise Collaborative Material Exercise 6 (CMX6) organized by the Nuclear Forensics international technical working group (ITWG). The characterized materials are uranium and cerium metal ingots. The goal was characterization of physical parameters, elemental and isotopic composition, morphology and chemical composition of the surface. Electron microscopy and Raman spectroscopy techniques were combined for morphological, structural and elemental study of the surface. The information from these analyses was used to answer the exercise questions, comparing the materials (labeled ES-1 and ES-2) to each other and to other samples described in the scenario. The INNFL investigation identified a connection between ES-1 and ES-2 based on several indicators, including the isotopic composition of the uranium present in both samples, as well as other surface contaminations, mainly yttria and graphite. Based on these finding the INNFL established a connection between ES-1 and ES-2 to ES-3B (theoretical sample with uranium traces) and has not established a connection to ES-3A (theoretical sample of plutonium powder). ES1 and ES-2 have been suggestively connected to their theoretical source, by their isotopic ratio. The INNFL conclusion was that the theoretical source materials (oxide and fluoride) were converted into the actual materials.

An orphan radium-beryllium (Ra–Be) neutron source (Nuclear Chicago Corporation) detected inside a scrap metal shipping container, was seized and subjected to nuclear forensic analysis. Physical and chemical characterization methods were used before and after source dismantling. Non-Destructive Testing (NDT) proved to be most useful for initial analysis of the source. Dissolution of the Ra-Be capsule was followed with elemental analysis by Inductively Coupled Plasma (ICP) Atomic Emission Spectroscopy (AES) and Mass Spectrometry (MS). Model age determination by ICP-MS (81.6 y ± 4.08) was found to be sensitive to insufficient accuracy in measuring daughters present in the parent solution.

The detection efficiency of a broad-energy germanium detector was determined empirically and compared to predictions by the Monte-Carlo software GEANT4 and FLUKA . The considered source geometries were point-like, filter paper and a water beaker. Calculations were performed in the photon energy range of 30–1836 keV . Simulated detection efficiencies obtained from both Monte-Carlo software were found to be in a good agreement (typically <5%) with experimental results for the considered source geometries. A comparison between the detection efficiencies obtained with GEANT4 and FLUKA codes exhibited a good agreement with a robust average of approximately 1%.

The transfer of 226Ra from irrigation water to basil crops was studied in field conditions. A dedicated basil plot was established and divided into test and control subplots irrigated with water having high (2.1 Bq L-1) and low (0.05 Bq L-1) activity concentrations of 226Ra, respectively. The experiment was performed over a period of 18 months during the autumn, winter and spring seasons, altogether eight cycles of growth and harvest. The activity concentration of 226Ra in basil grown in the test subplots was found to increase from a value of 0.6 Bq kg-1 up to 5.1 Bq kg-1 with successive cycles, compared to a mean value of 0.2 Bq kg-1 for basil grown in the control subplots. The increase in activity concentration of 226Ra in basil grown in the test subplots is mainly attributed to its build-up in the soil in which the level of 226Ra was found to increase by ~ 40%. The effective uptake of 226Ra from the irrigation water (via soil) by the basil plants was found to be approximately 0.4%. The maximal radiation dose following consumption of basil crops grown in the test subplots is negligible (~3 μSv/y).

A numerical model was developed that fully reconstructs a high-resolution gamma-ray spectrum of a fission sample. The code combines GEANT4 Monte Carlo simulations (used to calculate the response of a germanium detector) with the in-house Koala package (used to calculate the fission product inventory). In order to validate the model against experimental data, a highly enriched uranium sample was irradiated in a reactor and measured for up to 31 days later. A comparison between the simulated spectra and the experimental spectra exhibited a good quantitative agreement, demonstrating the high reliability of the model, especially for early times. While a simple source geometry was used for benchmarking the developed code, it can be easily extended to include complex geometries. The synthetic fission spectra can be used in several applications, e.g. assessing the capability of detector systems in identifying certain radionuclides, verification of gamma spectrometry analysis software and proficiency tests for analytical laboratories.

This paper demonstrates that that nuclear track detectors made of CR39 plates are suitable for detecting low energy protons and alpha-particles. The authors present evidence for a background line that arises from the interaction of neutrons with ${}^{17}$O inside the CR39 plate. This can be used for self-calibration of the CR39 detector

The cross sections of the Co-59(d, p)(CO)-C-60m,g,V- 51(d, p)(52) V, and V-nat(d, xn)(51) Cr reactions were accurately measured at the Soreq Applied Research Accelerator Facility in the energy range 2.7-5.4 MeV. The obtained results together with other data, available in the same region of the nuclear chart, were used for verifying the theoretical models. The consistent theoretical model calculation successfully reproduces the experimental results.