Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Radionuclide”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Distribution of plutonium and radium in the human heart

Since 1968, the United States Transuranium and Uranium Registries (USTUR) has studied the biokinetics and tissue dosimetry of uranium and transuranium elements in nuclear workers. As part of the USTUR collaboration with the Million Person Study of Low-Dose Health Effects, radiation dose to different parts of the human heart is being estimated for workers with documented intakes of 239 Pu or 226 Ra. The study may be expanded for workers with intakes of 238 U and other radionuclides. The distribution of radionuclides, expressed in terms of concentration (Bq per kg of tissue) serves as an important parameter for estimating radiation dose. Based on available organs from workers who donated their bodies or tissues for research, nine undissected hearts were selected: seven from USTUR registrants with plutonium exposure (males) and two individuals with radium intakes (female and male). For the plutonium workers, estimated 239 Pu systemic deposition ranged from <74 Bq to 1765 Bq. Estimated 226 Ra ‘initial systemic intakes’ were 10.1 MBq and 14.8 kBq for the female patient and male worker, respectively. Organ dissection was based on a heart model published by Borrego et al (2019 J. Radiol. Prot. 39 950–65). This model includes nine cardiac substructures: aorta, left main coronary artery, left atrium, left anterior descending artery, left circumflex artery, left ventricle, right atrium, right coronary artery, and right ventricle. In addition, heart valves, fat attached to epicardium, fluids, and a coronary bypass graft were collected resulting in 111 samples that are currently undergoing radiochemical analyses and mass-spectrometric measurements. The 239 Pu and 226 Ra evaluations are not completed. The results of this study are intended to support radiation worker health studies by improving associated dosimetric and epidemiological models.

USTUR↗
DOE OSTI2021

Specific absorbed fractions for a revised series of the UF/NCI pediatric reference phantoms: internal photon sources

Assessment of radiation absorbed dose to internal organs of the body from the intake of radionuclides, or in the medical setting through the injection of radiopharmaceuticals, is generally performed based upon reference biokinetic models or patient imaging data, respectively. Biokinetic models estimate the time course of activity localized to source organs. The time-integration of these organ activity profiles are then scaled by the radionuclide S-value, which defines the absorbed dose to a target tissue per nuclear transformation in various source tissues. S-values are computed using established nuclear decay information (particle energies and yields), and a parameter termed the specific absorbed fraction (SAF). The SAF is the ratio of the absorbed fraction—fraction of particle energy emitted in the source tissue that is deposited in the target tissue—and the target organ mass. While values of the SAF may be computed using patient-specific or individual-specific anatomic models, they have been more widely available through the use of computational reference phantoms. In this study, we report on an extensive series of photon SAFs computed in a revised series of the University of Florida and the National Cancer Institute pediatric reference phantoms which have been modified to conform to the specifications embodied in the ICRP reference adult phantoms of Publication 110 (e.g. organs modeled, organ ID numbers, blood contribution to elemental compositions). Following phantom anatomical revisions, photon radiation transport simulations were performed using MCNPX v2.7 in each of the ten phantoms of the series—male and female newborn, 1 year old, 5 year old, 10 year old, and 15 year old—for 60 different tissues serving as source and/or target regions. A total of 25 photon energies were considered from 10 keV to 10 MeV along a logarithm energy grid. Detailed analyses were conducted of the relative statistical errors in the Monte Carlo target tissue energy deposition tallies at low photon energies and over all energies for source–target combinations at large intra-organ separation distances. Based on these analyses, various data smoothing algorithms were employed, including multi-point weighted data smoothing, and log–log interpolation at low energies (1 keV and 5 keV) using limiting SAF values based upon target organ mass to bound the interpolation interval. The final dataset is provided in a series of ten electronic supplemental files in MS Excel format. Finally, the results of this study were further used as the basis for assessing the radiative component of internal electron source SAFs as described in our companion paper (Schwarz et al 2021) for this same pediatric phantom series.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Specific absorbed fractions for a revised series of the UF/NCI pediatric reference phantoms: internal electron sources

Assessment of radiation absorbed dose to internal organs of the body from the intake of radionuclides, or in the medical setting through the injection of radiopharmaceuticals, is generally performed based upon reference biokinetic models or patient imaging data, respectively. Biokinetic models estimate the time course of activity localized to source organs. The time-integration of these organ activity profiles are then scaled by the radionuclide S-value, which defines the absorbed dose to a target tissue per nuclear transformation in various source tissues. S-values are computed using established nuclear decay information (particle energies and yields), and a parameter termed the specific absorbed fraction (SAF). The SAF is the ratio of the absorbed fraction—fraction of particle energy emitted in the source tissue that is deposited in the target tissue—and the target organ mass. While values of the SAF may be computed using patient-specific or individual-specific anatomic models, they have been more widely available through the use of computational reference phantoms. In this study, we report on an extensive series of photon SAFs computed in a revised series of the University of Florida and the National Cancer Institute pediatric reference phantoms which have been modified to conform to the specifications embodied in the ICRP reference adult phantoms of Publication 110 (e.g. organs modeled, organ ID numbers, blood contribution to elemental compositions). Following phantom anatomical revisions, photon radiation transport simulations were performed using MCNPX v2.7 in each of the ten phantoms of the series—male and female newborn, 1 year old, 5 year old, 10 year old, and 15 year old—for 60 different tissues serving as source and/or target regions. A total of 25 photon energies were considered from 10 keV to 10 MeV along a logarithm energy grid. Detailed analyses were conducted of the relative statistical errors in the Monte Carlo target tissue energy deposition tallies at low photon energies and over all energies for source–target combinations at large intra-organ separation distances. Based on these analyses, various data smoothing algorithms were employed, including multi-point weighted data smoothing, and log–log interpolation at low energies (1 keV and 5 keV) using limiting SAF values based upon target organ mass to bound the interpolation interval. The final dataset is provided in a series of ten electronic supplemental files in MS Excel format. Finally, the results of this study were further used as the basis for assessing the radiative component of internal electron source SAFs as described in our companion paper (Schwarz et al 2021) for this same pediatric phantom series.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Imaging and dosimetric characteristics of 67 Cu

In recent years the use of beta-emitting radiopharmaceuticals for cancer therapy has expanded rapidly following development of therapeutics for neuroendocrine tumors, prostate cancer, and other oncologic malignancies. One emerging beta-emitting radioisotope of interest for therapy is 67 Cu (t 1/2 : 2.6 d) due to its chemical equivalency with the widely-established positron-emitting isotope 64 Cu (t 1/2 : 12.7 h). Here we evaluate both the imaging and dosimetric characteristics of 67 Cu, as well as producing the first report of SPECT/CT imaging using 67 Cu. To this end, 67 Cu was produced by photon-induced reactions on isotopically-enriched 68 Zn at the Low-Energy Accelerator Facility (LEAF) of Argonne National Laboratory, followed by bulk separation of metallic 68 Zn by sublimation and radiochemical purification by column chromatography. Gamma spectrometry was performed by efficiency-calibrated high-purity germanium (HPGe) analysis to verify absolute activity calibration and establish radionuclidic purity. Absolute activity measurements corroborated manufacturer-recommended dose-calibrator settings and no radionuclidic impurities were observed. Using the Clinical Trials Network anthropomorphic chest phantom, SPECT/CT images were acquired. Medium Energy (ME) SPECT collimation was found to provide the best image quality from the primary 185 keV gamma emission of 67 Cu. Reconstructed images of 67 Cu were similar in quality to images acquired using 177 Lu. Recovery coefficients were calculated and compared against quantitative images of 99m Tc, 177 Lu, and 64 Cu within the same anthropomorphic chest phantom. Production and clinical imaging of 67 Cu appears feasible, and future studies investigating the therapeutic efficacy of 67 Cu-based radiopharmaceuticals are warranted.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2023

Measurable radiation levels around individuals externally contaminated by nuclear fallout

During the early response to large-scale radioactive contamination events, people who are potentially affected need to be screened for radioactive contamination and public health staff need to triage individuals who may need immediate decontamination. This is typically done by screening individuals for external contamination using ionising radiation detection equipment. In this study, spatially and temporally dependent isotopic compositions from a simulated nuclear detonation and Monte Carlo methods were used to relate contamination activity levels to the measurable radiation levels at select distances away from an individual with whole-body contamination. Radionuclide-specific air kerma rate coefficients and Geiger–Mueller instrument response coefficients at five select distances from contaminated individuals are presented for 662 radionuclides. Temporally and spatially dependent incident-specific coefficients are presented for a hypothetical surface detonation of a 235 U-fueled device.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2026

Statistical Uncertainty of Inhalation Dose Coefficients: Impact of Particle Deposition in ICRP 66 Human Respiratory Tract Model

Inhaled radioactive materials can pose a long-term health concern, as the material can be incorporated into the body’s metabolic pathways and remain in organs and tissues for extended durations. During the retention period, the radioactive material may localize in a source organ and irradiate adjacent target organs and tissues. Distribution of these materials changes over time, requiring biokinetic modeling to evaluate their movement through various tissues and organs. The evolving distribution depends on multiple inputs characterizing the inhaled material, such as particle size and size distribution, particle density, aspect ratio, specific radionuclide, the chemical form, and solubility. In addition, biological parameters such as breathing rate, breathing type (nasal or nasal/oral), respiratory system morphometry, tidal volume, functional residual capacity, and anatomical dead space all influence material transport. These aerosol properties and physiological characteristics of the respiratory tract jointly define a range of initial conditions that influence the time-dependent distribution of radioactive material. To evaluate both uncertainty in the initial conditions of inhalation exposure and the final output (committed effective dose) from biokinetic models, a Python-based software tool, Radiological Exposure Dose Calculator (REDCAL), was developed to propagate uncertainty within the human respiratory tract model. Focusing on deposition fraction uncertainty, the primary objective was to characterize the initial activity distribution across respiratory regions as a function of anticipated particle sizes and distributions. The impact of the deposition fraction uncertainty was propagated to committed effective dose coefficients for selected radionuclides in a companion publication. For each particle size, a lognormal distribution, characterized by its geometric mean as defined within ICRP Publication 66, serves as the basis for introducing uncertainty into the physical processes governing deposition in various lung regions. Finally, this study addresses the deposition process and examines how uncertainty in deposition mechanisms affects activity distribution in the airways, ultimately presenting the expected range and standard deviation of deposited activity as a function of particle size.

International Commission on Radiological Protectio↗
DOE OSTI2022

Decay spectroscopy of the blocked fission product 130 $\mathrm{I}$

We report numerous applications rely on the identification and quantification of fission products with the activation technique, where γ-rays emitted in the decay are used to estimate the initial activity of the radionuclide of interest. 130 I is a so-called blocked fission product, which can be produced only directly through fission, a property that makes it particularly attractive for nuclear forensics. A source of 130 I was produced using a (p,n) reaction on enriched 130 Te at the Brookhaven Tandem Van de Graaff and its decay was studied with Gammasphere at Argonne National Laboratory. Two new levels were identified, and over 25 transitions were added, removed or re-placed in the level scheme, with intensity measurements made down to I γ = 0.00066 per 100 decays. The uncertainty on the intensities of the strongest transitions, those that are commonly used to quantify the activity of the radionuclide, was improved by a factor of 2 compared to the previous best assessment and discrepancies in the literature values were resolved. A detailed angular correlation analysis further permitted the determination of a number of spin assignments for excited levels and mixing ratios for γ-ray transitions

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

A Large Area Picosecond Photodetector for Neutron Transmission Measurements at DICER at LANSCE

The Device for Indirect Capture Experiments on Radionuclides (DICER) is a new instrument that is being developed at the Los Alamos Neutron Science Center (LANSCE) to study neutron capture reactions on short-lived radionuclides. To achieve that, DICER is using an indirect technique based on measuring the neutron transmission through very small samples (0.12–1.2 mm in diameter, $\mu \text{g}$ -mg in mass). The main detection system of DICER consists of two dual photomultiplier (PMT)-based detectors which are sensitive to the scintillation light produced when neutrons interact with 6 Li glass disks. DICER is developing a new collimation system that includes a 0.1-mm collimator. In such small sizes, the need for neutron beam imaging is crucial and necessary to confirm the appropriate alignment between the irradiated sample and the neutron detectors. The large area picosecond photodetector (LAPPD) is a multichannel plate (MCP)-based photodetector, with picosecond-level timing, single-photon detection capabilities, and spatial resolution of the order of 1 mm. Coupling the LAPPD with a 6 Li glass scintillator results in a neutron imaging detector. The first irradiation of the LAPPD at DICER indicated that apart from imaging the beam, the LAPPD can be used as the main neutron detection system. Furthermore, these results were proof of principle measurements demonstrating the capabilities of the beamline, and the challenge with small samples will be addressed in the future. The first efforts and developments, including irradiations of the LAPPD and the first beam images, will be presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Lithologic controls on microfracturing from legacy underground nuclear explosions

Detection and verification of underground nuclear explosions (UNEs) can be improved with a better understanding of the nature and extent of explosion-induced damage in rock and the effect of this damage on radionuclide migration. Much of the previous work in this area has focused on centimeter- to meter-scale manifestations of damage, but to predict the effect of damage on permeability for radionuclide migration, observations at smaller scales are needed to determine deformation mechanisms. Based on studies of tectonic deformation in tuff, we expected that the heterogeneous tuff layers would manifest explosion-induced damage differently, with welded tuffs showing more fractures and nonwelded tuffs showing more deformation bands. In comparing post-UNE samples with lithologically matched pre-UNE equivalents, we observed damage in multiple lithologies of tuff through quantitative microfracture densities. We find that the texture (e.g., from deposition, welding, alteration, etc.) affects fracture densities, with stronger units fracturing more than weaker units. While we see no evidence of expected deformation bands in the nonwelded tuffs, we do observe, as expected, much larger microfracture densities at close range (<50 m) to the explosive source. We also observe a subtle increase in microfracture densities in post-UNE samples, relative to pre-UNE equivalents, in all lithologies and depths. The fractures that are interpreted to be UNE-induced are primarily transgranular and grain-boundary microfractures, with intragranular microfracture densities being largely similar to those of pre-UNE samples. This work has implications for models of explosion-induced damage and how that damage may affect flow pathways in the subsurface.

58 GEOSCIENCES↗
DOE OSTI2020

Measurement of 139 La(p,x) cross sections from 35–60 MeV by stacked-target activation

A stacked-target of natural lanthanum foils (99.9119% 139 La) was irradiated using a 60 MeV proton beam at the LBNL 88-Inch Cyclotron. 139 La(p,x) cross sections are reported between 35–60 MeV for nine product radionuclides. The primary motivation for this measurement was the need to quantify the production of 134 Ce. As a positron-emitting analogue of the promising medical radionuclide 225 Ac, 134 Ce is desirable for in vivo applications of bio-distribution assays for this emerging radio-pharmaceutical. Finally, the results of this measurement were compared to the nuclear model codes TALYS, EMPIRE and ALICE (using default parameters), which showed significant deviation from the measured values.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Spin assignments for $^{23}\hbox {Mg}$ levels and the astrophysical $^{22}\hbox {Na}(p,\gamma )^{23}\hbox {Mg}$ reaction

The 22 Na(p,γ) 23 Mg reaction is responsible for destruction of the long-lived radionuclide 22 Na produced during nova explosions. Since the reaction proceeds through resonances from levels in 23 Mg above the proton threshold at 7.581 MeV, the properties of these levels such as excitation energies, spins, and parities are crucial ingredients to deter- mine the 22 Na(p,γ) 23 Mg reaction rate. Despite recent studies of these levels, their spins are not well constrained in many cases. We have measured the 24 Mg(p,d) 23 Mg transfer reaction to determine spectroscopic properties of these levels at the Holifield Radioactive Ion Beam Facility at Oak Ridge National Laboratory. The spin of the E x = 7.788 MeV level in 23 Mg is constrained to be J π = (3/2 + , 5/2 + ) through the present work. Here, the astrophysical 22 Na(p,γ) 23 Mg reaction rate at nova temperatures is updated accordingly. Nova nucleosynthesis model calculations using the newly updated 22 Na(p,γ) 23 Mg reaction rate shows that the final weighted abundance of the radionuclide 22 Na is increased by 42% compared to that obtained by using the previous 22 Na(p,γ) 23 Mg reaction rate of Sallaska et al. for a 1.35 M ⊙ ONeMg white dwarf.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

SCALE Non-LWR Models for NRC Volume 3

This dataset contains input and result files of computational simulations with the SCALE code system. The simulations cover radionuclide inventory and reactivity analyses of various advanced reactors. Users wanting to reproduce results from this dataset are required to obtain a license to the SCALE code system for which details on the distribution can be found here: https://www.ornl.gov/scale/releases

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

SCALE Non-LWR Models for NRC Volume 5

This dataset contains input and result files of computational simulations with the SCALE code system. The simulations cover radionuclide inventory generation, criticality calculations, and dose rate/shielding analyses of various advanced reactors. Users wanting to reproduce results from this dataset are required to obtain a license to the SCALE code system for which details on the distribution can be found here: https://www.ornl.gov/scale/releases

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Analysis of near-field and far-field aerosol dispersion for microreactors

The current paper presents a simulation-based analysis of aerosol dispersion in the near-field and far-field of a generic, conceptual microreactor operating at pressures close to the ambient pressure; therefore, in the event of an accident that causes radionuclide leakage from the microreactor containment, the radionuclide particles are less likely to travel too far from the reactor, as opposed to conventional reactors. Accordingly, the presented work provides estimates of average and 95-percentile values of the relative effluent concentration. A parametric study is then performed to narrow down the parameters which affect the aerosol dispersion characteristics most significantly. Simulations were performed in the computer code ARCON96, and the parameters found to affect aerosol dispersion characteristics are the atmospheric stability class, and the distance between the release point and the receptor. It is recommended that the computer code RADTRAD be used to calculate actual dosage over distance, using the outputs from ARCON96 as inputs, along with reactor-specific core term inventories. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Stochastic parametric skeletal dosimetry model for humans: General approach and application to active marrow exposure from bone-seeking beta-particle emitters

The objective of this study is to develop a skeleton model for assessing active marrow dose from bone-seeking beta-emitting radionuclides. This article explains the modeling methodology which accounts for individual variability of the macro- and microstructure of bone tissue. Bone sites with active hematopoiesis are assessed by dividing them into small segments described by simple geometric shapes. Spongiosa, which fills the segments, is modeled as an isotropic three-dimensional grid (framework) of rod-like trabeculae that “run through” the bone marrow. Randomized multiple framework deformations are simulated by changing the positions of the grid nodes and the thickness of the rods. Model grid parameters are selected in accordance with the parameters of spongiosa microstructures taken from the published papers. Stochastic modeling of radiation transport in heterogeneous media simulating the distribution of bone tissue and marrow in each of the segments is performed by Monte Carlo methods. Model output for the human femur at different ages is provided as an example. The uncertainty of dosimetric characteristics associated with individual variability of bone structure was evaluated. An advantage of this methodology for the calculation of doses absorbed in the marrow from bone-seeking radionuclides is that it does not require additional studies of autopsy material. The biokinetic model results will be used in the future to calculate individual doses to members of a cohort exposed to 89,90 Sr from liquid radioactive waste discharged to the Techa River by the Mayak Production Association in 1949–1956. Further study of these unique cohorts provides an opportunity to gain more in-depth knowledge about the effects of chronic radiation on the hematopoietic system. In addition, the proposed model can be used to assess the doses to active marrow under any other scenarios of 90 Sr and 89 Sr intake to humans.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2025

Production and purification of research scale 161 Tb using cation-exchange semi-preparative HPLC for radiopharmaceutical applications

Terbium-161 ( 161 Tb) is emerging as a promising radionuclide for cancer therapy due to its favorable nuclear properties that are similar to clinically established lutetium-177 ( 177 Lu) along with its therapeutic edge arising from the higher number of Auger and conversion electrons per decay. These low energy electrons result in higher cytotoxicity within a short range of the decaying nuclei to enhance therapeutic efficacy. Despite these promising characteristics, a significant challenge remains in the lack of a domestic 161 Tb supply in the United States, which poses an obstacle to the advancement of 161 Tb-based radiopharmaceutical research and development. Here, this study developed a reliable cation-exchange high-performance liquid chromatography-based method for purification of reactor-produced 161 Tb at quantities suitable to support research and preclinical studies. The purified 161 Tb product showed high radionuclidic purity with excellent radiochemical purity, and the successful labeling studies with the DOTA chelator and DOTA-TATE peptide demonstrated the effective incorporation of the purified 161 Tb into radiopharmaceuticals designed for targeted cancer therapy.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2020

Results for the Fourth Quarter Calendar Year 2018 Tank 50 Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the Fourth Quarter Calendar Year 2018 (CY18) sample of Tank 50 salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Remediation (SRR) for the transfer of aqueous waste from Tank 50 to the Saltstone Production Facility (SPF), where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) Limits and Targets. The chemical and radionuclide contaminant results from the characterization of the Fourth Quarter CY18 sampling of Tank 50 were requested by SRR personnel via a Task Technical Request (TTR) and details of the testing are presented in the Savannah River National Laboratory (SRNL) Task Technical and Quality Assurance Plan (TTQAP). This Technical Report is part of Deliverable 2 relating to Task 1 from the SRR request. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the TTQAP for the Tank 50 Saltstone task.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Analysis of the sludge batch 9 (Macrobatch 10) DWPF pour stream glass sample

A pour stream (PS) sample taken during processing of Sludge Batch 8 (SB8, Macrobatch 10) was analyzed by Savannah River National Laboratory (SRNL) to verify compliance with the Defense Waste Processing Facility (DWPF) Glass Product Control Program. The following conclusions were drawn from these analyses; The sum of oxides for the SB8 PS glass (97.41 wt%) is within the Product Composition Control System (PCCS) limits (100 ± 5 wt%); The average Waste Dilution Factor (WDF) calculated from the Tank 40 dried sludge analysis results for SB8 is 1.9 ± 0.1. In general, the radionuclide concentration of the SB8 PS glass sample calculated using this WDF is in good agreement with the measured radionuclide content; The Product Consistency Test (PCT) results indicate that the SB8 PS glass meets the Waste Acceptance Product Specifications (WAPS) for durability, with normalized release rates for boron, lithium and sodium all at least two standard deviations lower than those of the Environmental Assessment (EA) glass; The average measured density of the SB8 PS glass is 2.72 g/cm 3 , consistent with previous PS samples; and, The average measured Fe 2+ /ΣFe ratio of the SB8 PS glass is 0.15, within the target range of 0.09- 0.33, and closely resembling the predicted REDOX value of 0.15.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗