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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Evaluation of gamma-ray transmission through rectangular collimator slits for application in nuclear fuel spectrometry

Gamma-ray spectrometry is widely applied in several science fields, and in particular in non-destructive gamma scanning and gamma emission tomography of irradiated nuclear fuel. Usually, a collimator is used in the experimental setup, to selectively interrogate a region of interest in the fuel. For the optimization of instrument design, as well as for planning measurement campaigns, predictive models for the transmitted gamma-ray intensity through the collimator are needed. Commonly, Monte Carlo Radiation Transport tools are used for accurate prediction of gamma-ray transport, however, the long computation time requirements when used in low-efficiency experimental setups present challenges. Here, the full-energy peak intensity transmitted through a rectangular collimator slit was examined. A uniform planar surface source emitting isotropically was considered, and the rate of photons reaching an ideal counter plane on the opposite side of the collimator was evaluated by analytical integration. To find a closed-form primitive function, some idealizations were required, and thereby parametric models were obtained for the optical field of view, dependent on slit dimensions (length, height and width) and source-to-collimator distance. For contributions from outside the optical field of view, where a closed-form expression cannot be found, fast numerical integral methods were instead used. The results were validated using the Monte Carlo code MCNP6 and show an agreement within three percent for the numerical method. For the analytical method, deviations up to tens of percent were obtained, which is deemed to still be sufficient for instrument design and measurement planning, where often the order of magnitude of the count rate is not a priori known. The method is planned for use in iterative optimization routines in the design of Gamma Emission Tomography devices, as well as for the prediction of gamma spectra obtained in the planning of fuel inspections. An application of the proposed method was demonstrated in spectrum prediction for a short cooling-time fuel rod test from the Halden reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Natural attenuation of uranium in a fluvial Wetland: Importance of hydrology and speciation

A nuclear fuel fabrication facility released 43,500 kg of uranium into a riparian wetland located on the Savannah River Site between 1955 and 1988. Studies were undertaken to evaluate hydrological and geochemical processes influencing uranium accumulation in the wetland. Gamma-radiation-mapping surveys were conducted by systematically walking over the contaminated wetland with backpacks equipped with global positioning systems and NaI gamma detectors. Based on maps compiled from >700,000 gamma spectra and eight sediment uranium depth profiles, it was determined that 94% of the released uranium remained in the wetland. The uranium in the wetland is concentrated in five multi-hectare areas along the stream, accounting for ~11% of the land area adjacent to the stream. While land type (upland or wetland) and topography provided a reasonable first approximation of where much of the uranium was deposited, hydrological watershed modeling revealed that the stream velocity was especially slow through many of the hot spots. Here, using autoradiography combined with SEM/EDX measurements of contaminated sediments, surprisingly few hot particles were detected. Instead, uranium was evenly distributed throughout the sampled sediment, suggesting that dissolved uranium had bound to sediment particles that became suspended and later deposited in low energy (low flow velocity) portions of the stream. EXAFS suggested that U atoms were present as individual ions in disordered complexes within the sediment. Furthermore, linear combination analyses suggested that the predominant component of the U(VI) was adsorbed to sediment minerals (~70%) and a minor component (~30%) was associated with organic matter phases. Furthermore, these studies show that wetlands can be extraordinarily effective at binding and retaining uranium, thereby providing a natural barrier to the transport of uranium out of a watershed. However, significant anthropogenic or climatic changes to wetlands, such as those associated with flooding, forest fires, or land use, may disrupt the complex hydrological and biogeochemical balance necessary to maintain long-term immobilization of uranium.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

RAPID QUANTITATION OF URANIUM FROM MIXED FISSION PRODUCT SAMPLES

Chemical similarities between U(VI) and Mo(VI) create challenges for separation and quantification of uranium from a mixed fission product sample. The purpose of this work was to demonstrate the feasibility of using Eichrom’s® UTEVA resin in addition to a tellurium spontaneous deposition to improve the quantitation of 235U using gamma spectroscopy. The optimized method demonstrated a consistent chemical yield of 74 ± 3 % for uranium. This procedure was evaluated using 1.41x1012 fissions produced from an irradiated HEU sample. The uranium was isotopically yielded by HPGe, and the minimum detectable activity (MDA) determined from the gamma spectra. The MDA for 235U, 237U, and 238U was reduced by a factor of two. The chemical isolation of uranium was successfully achieved in less than four hours, with a separation factor of 1.41x105 from molybdenum.

Haney, Morgan M.↗
DOE OSTI2019

Detector performance and defect densities in CdZnTe after two-step annealing

Defects both micro-scale and nano-scale are plays an important role in CdZnTe(CZT) device performance. Typical micro-scale defect, Te inclusions, was removed by two-step annealing and its concentration were analyzed by IR transmission microscopy. In addition, transmission electron microscopy (TEM) measurement was employed to investigate the evolution of nano-scale defects after annealing process. Micro-twins and stacking faults were commonly observed defects in CZT as-grown and annealed CZT. The line shape defects which possibly related to the stress field around dislocations, were combine and disappear in in-situ the annealing at 200-220 °C. Frisch-grid CZT detector made of two-step annealed one exhibit better energy resolution and low backscattering counts in Cs-137 gamma spectra.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Semi-empirical simulation of in-motion radiation detection systems

The Replicative Assessment of Spectroscopic Equipment (RASE) is an open-source software that uses experimental data as the basis to simulate the response of commercial radiation detectors to sources in various situations, particularly in the context of nuclear security and safeguards applications. Dynamic RASE introduces the capability to simulate scenarios where sources and detector are in relative motion. Position-dependent experimentally acquired gamma spectra are ingested by Dynamic RASE to build maps that describe the detector response over all space. These response maps are used to replicate the time-dependent energy spectra collected as sources move on a path near the detector. Here, a Gaussian process is used to build each map, incorporating a novel kernel adapted to the special case of radiation detection. The approach has been validated against experimental data acquired using a NaI-based detector for 137 Cs and 54 Mn sources. The capability to create accurate simulations using either long-dwell static measurements or dynamic pass-by measurements as source data has been demonstrated. Quantitative relative performance, benefits, and shortcomings are discussed.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Assessment of alternative radionuclides for use in a radioisotope thermoelectric generator

238 Pu is the current fuel source for MMRTGs. This is due to favorable properties, such as high decay heat and low gamma emissions. 238 Pu supplies are dwindling rapidly, necessitating the review of potential replacement isotopes for deep space missions. This work considers nine potential replacements, Ac-227, Am-241, Cf-252, Cm-244, Cs-137, Po-210, Sr-90, Th-228, and Th-170. We have assessed performance of alternatives to 238 Pu by modeling the specific decay-heat and gamma spectra of several potential radioisotopes in SCALE-ORIGEN, and comparing these with 238 Pu. An analysis of heat output, heat stability, safety, availability and required mass, baselined against 238 Pu demonstrates that of those radioisotopes studied, 241 Am has properties that make it most conducive for 238 Pu replacement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Exploring diversion-pathway analysis of a generic molten-salt fast reactor using multiphysics informed signatures

Molten salt reactors are being explored by multiple commercial ventures due to their inherent safety features, flexibility in fuel sources, and high fuel utilization and thermal efficiency. The continual flow of fuel salt, large fissile quantities present, and ability to add or divert material due to the liquid nature introduces new challenges for international safeguards. To understand how international safeguards should be applied, it is important to capture the inherent multi-physics nature of a molten salt reactor. This work examines a generic molten salt fast reactor to understand how potential diversion scenarios would affect the concentration of radionuclides in the primary and auxiliary systems. Three types of diversion were examined: a slow drip of fuel salt, gaseous plutonium extraction, and uranium metal plating. The analysis determined that several key isotopes become statistically significant once diversion begins, indicating that detection of such diversion cases would be possible through measuring specific signatures such as gamma spectra.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Test of nuclear decay rate variation due to an antineutrino flux

Unexplained variations of the decay rate parameter for weak interaction decays such as β ± decay, electron capture, as well as strong interaction α decay have been reported. Because these variations have been presented by a number of groups, at various locations, using various types of detectors, different isotopes, and over extended periods of time, some researchers have interpreted the source of these variations as not from ambient environmental factors such as temperature, pressure, and humidity but via an unexplained fundamental interaction. The state of decay rate parameter variations experiments is reviewed, and the reported results are placed into a common comparable context by defining a phenomenological cross section. After that we make decay parameter measurements as a function of time at the level of 10 -5 in the presences of an antineutrino flux with an on-off cycle time of ≈ 30 days. This level of precision requires a detailed understanding of both systematic and statistical errors; otherwise, systematic errors in the measurement may mimic the decay events of fundamental interactions. The experiment searched for variation of the $^{54}_{25}$Mn, e - capture decay rate parameter and $^{137}_{55}$Cs, β - decay rate parameter both to a level of precision of 1 part in ≈10 5 by comparing the difference between the decay rate in the presence of an antineutrino flux ≈ 3 x 10 12 $\bar{ν}$ cm -2 s -1 and no flux measurements. The experiment was located 6.53 m from the reactor core of the High Flux Isotope Reactor at the Oak Ridge National Laboratory. Two weak interaction decays, one via electron capture and the other via β - decay, were selected because the final state and the time reverse state each contain a neutrino and antineutrino, covering arguments that the antineutrino flux may interact differently or not at all in one of the cases. The $\gamma$ spectra from both decays were collected and analyzed independently. The measured variation in the decay rate parameters are found to be δλ/λ =(0.034 ± 1.38) x10 -5 for $^{54}_{25}$Mn and δλ/λ =(0.67±1.56) x10 -5 for $^{137}_{55}$Cs. These results are consistent with no measurable decay rate parameter variation due to an antineutrino flux, yielding a 68% confidence level upper limit sensitivity for $^{54}_{25}$Mn δλ/λ ≤ 1.31 x 10 -5 or σ ≤ 1.29 x10 -25 cm 2 in cross section and for $^{137}_{55}$Cs δλ/λ ≤ 2.23 x 10 -5 or σ ≤ 5.69 x10 -27 cm 2 . The cross-section upper limit obtained in these null or no observable effect measurements are ≈ 10 4 times more sensitive than past experiments reporting positive results in 54 Mn and 137 Cs. . .

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

On Feature, Classifier and Detector Fusers for 235U Signatures Using Gamma Spectral Counts

Three types of information fusion strategies are studied to assess the performance of classifiers for detecting low-level 235 U radiation sources, using features obtained from gamma spectra of NaI detectors. These three strategies are based on using two spectral region features, fusing eight classifiers of diverse designs, and fusing multiple detectors located at different positions around the source. The inner, middle and outer groups of detectors, within a formation of two concentric circles and a spiral of 21 detectors, are identified based on their distance to the source, which is located at the center. This study provides two main qualitative insights into this classification task. First, the fusion of detectors leads to an overall improved classification performance, least in the inner group, most in the outer group, and in between for the middle group. Second, several classifiers and fusers achieve lower training error which does not translate to lower generalization error, indicating their over-fitting to training data.

Rao, Nageswara↗
DOE OSTI2018

Reactor Facility Operational State Classification Using Gas Effluents and Radiation Measurements

Inferring the operational state of a reactor facility, using measurements from an independent in-situ monitoring system, is critical to the assessment of its compliance to agreements. We consider the problem of inferring the on/off operational state of a facility by using gas effluents and radiation measurements collected at the reactor's off-gas ventilation stack and within its surrounding area. Effluent release concentrations of three noble gases, namely, Ar-41, Cs-138, and Xe-138, are collected inside the ventilation stack and provided every four hours. Gamma spectra is recorded every second by a radiation detector located about 100 meters offset from the stack, which is affected by weather and nearby conditions. We study classifiers based on the correlation coefficient (CC) method [1] applied to individual modalities and their combinations. We implement CC classifiers using ground truth measurements collected over 2 operational cycles of High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) in year 2017.

Ramirez Aviles, Camila A.↗
DOE OSTI2020

Performance of High Stopping Power Bismuth-Loaded Plastic Scintillators for Radiation Portal Monitors

Plastic scintillators are widely used in radiation portal monitors because of their low cost and availability in large sizes. However, due to their low density and low effective atomic number (Z), they offer low intrinsic efficiency and little spectroscopic information. The addition of high-Z constituents to these plastics can greatly increase both their total stopping power and the amount of photoelectric absorption, leading to full-energy deposition and thus useful gamma spectra. Herein, we present the performance of the latest formulation of Bi-loaded plastic scintillators showing their useful spectroscopic information up to relatively high energy (~1 MeV) due to their high stopping power compared to the current commercially available plastics. These Bi-loaded plastics use 20 weight percent (wt%) Bi-pivalate (8 wt% elemental Bi) dissolved in polyvinyltoluene (PVT) matrix and conventional fast fluors (~10 ns decay time). These Bi-loaded plastics achieve up to approximately 6000 photons/MeV and have been produced in sizes up to 17 in 3 . The performance of these Bi-loaded plastics is also demonstrated in the existing portal monitor hardware (Rapiscan Model TSA Trainer 770) showing the possibility to provide improved sensitivity as a drop-in replacement with continued scale-up.

42 ENGINEERING↗
DOE OSTI2016

SU-F-I-56: High-Precision Gamma-Ray Analysis of Medical Isotopes

Purpose: Advanced, time-resolved, Compton-suppressed gamma-ray spectroscopy with germanium detectors is implemented for assaying medical isotopes to study the radioactive decay process leading to a more accurate appraisal of the received dose and treatment planning. Lowell’s Array for Radiological Assay (LARA), a detector array that is comprised of six Compton-suppressed high-purity germanium detectors, is currently under development at UMass-Lowell which combines Compton-suppression and time-and-angle correlations to allow for highly efficient and highly sensitive measurements. Methods: Two isotopes produced Brookhaven Linac Isotope Producer (BLIP) were investigated. {sup 82}Sr which is the parent isotope for producing {sup 82}Rb is often used in cardiac PET. {sup 82}Sr gamma-ray spectrum is dominated by the 511keV photons from positron annihilation which prevent precise measurement of co-produced contaminant isotopes. A second project was to investigate the production of platinum isotopes. Natural platinum was bombarded with protons from 53MeV to 200MeV. The resulting spectrum was complicated due to the large number of stable platinum isotopes in the target, the variety of open reaction channels (p,xn), (p,pxn), (p,axn). Results: By using face-to-face NaI(Tl) counters 90-degrees to the Compton-suppressed germaniums to detect the 511keV photons, a much cleaner and more sensitive measurement of {sup 85}Sr and other contaminants was obtained. For the platinum target, we identified the production of {sup 188–189–191–195}Pt, {sup 191–192–193–194–195–196}Au and {sup 186–188–189–190–192–194–189–190–192–194}Ir. For example, at the lower energies (53 and 65MeV), we measured {sup 191}Pt production cross-sections of 144mb and 157mb. Considerable care was needed in following the process of dissolving and diluting the samples to get consistent results. The new LARA array will help us better ascertain the absolute efficiency of the counting system and more reliable production cross-sections. Conclusion: Modern HPGe spectroscopic techniques provide enhanced sensitivity, promising precise quantification of the quality of radioisotopes used in medical physics. Using new decay information may have non-trivial impact on treatment planning and dose-assessment.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Code for BALDR Study 07.04

SAND2024-11256O The Code for BALDR Study 07.04 software reproduces results from the BALDR study concerning "Multilabel Proportion Prediction and Out-of-Distribution Detection on Gamma Spectra of Short-Lived Fission Products." This code can reproduce a scientific study following the step numbers present in the file names. The scientific study uses synthetic and measured data to find the best model for the radioisotope proportion estimation task of interest and generates results. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525.

Morrow, Tyler↗
DOE OSTI2025

Full Rod Length Gamma Spectroscopy Measurements of LWR Used Fuel Rods

Twenty-five high-burnup pressurized water reactor used fuel rods were scanned in the Irradiated Fuels Examination Laboratory at Oak Ridge National Laboratory with a high-resolution gamma spectrometer. A planar high-purity germanium detector was used across an energy range from 60 keV to 3 MeV. Gamma spectra of individual fuel rods were collected in 4.5 mm steps along the entire length of each rod. Additional spectra were collected on select rods in rotational increments of 30 degrees to provide data on the distribution of fission products along the rod perimeter.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Special Nuclear Material Gamma-Ray Signatures for Reachback Analysts

These are slides on special nuclear material gamma-ray signatures for reachback analysts for an LSS Spectroscopy course. The closing thoughts for this presentation are the following: SNM materials have definite spectral signatures that should be readily recognizable to analysts in both bare and shielded configurations. One can estimate burnup of plutonium using certain pairs of peaks that are a few keV apart. In most cases, one cannot reliably estimate uranium enrichment in an analogous way to the estimation of plutonium burnup. The origin of the most intense peaks from some SNM items may be indirect and from ‘associated nuclides.' Indirect SNM signatures sometimes have commonalities with the natural gamma-ray background.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

INITIAL UC Davis Reactor Analysis Context

The INITIAL reactor experiment pitched to UC Davis and agreed upon by the professor teaching the course is a natural uranium neutron activation experiment in the McClellan Nu- clear Research Center (MNRC) reactor. Both non-cadmium covered (Ex:1 – bare) and cadmium covered (Ex:2 – Cd) experiments will be performed and analysis will ensure on gamma spectra collected from both experiments. The following provides necessary information for completing the experiment analysis. This information includes definitions, assumptions, and plots tailored to the specifics of this experiment. The content is for the students to be completing the laboratory for reference.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Annual IPCA2 Performance Report for JFY18

This report summarizes the results of monthly control measurements of IPCA2 performed over the period of October 2018 through March 2019 and represents an annual performance overview for JFY18. Note that the start of the measurements in October 2018 was defined by the start of the contract agreement. Monthly measurements of Plutonium neutron detection efficiency, AmLi stability, Curium stability and HPGe gamma spectra of Plutonium standards were performed and analyzed. All the results are shown with respect to original control bounds established from 2013- 2017 data in. Updated control bounds established based on this JFY18 data for use during JFY19 measurements are summarized in Appendix A. Based on Pu efficiency measurements, the performance of the IPCA2 during this reporting period was stable within 0.6% at 1 σ level. Measurements were compared to room temperature and humidity and no dependence was found for any detector performance. The end of this report contains summary of all IPCA2 measurements performed since October 2018 and includes discussion of troubleshooting and repair activities performed over the reporting period.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗