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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 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Implementing an aerosol dynamic shape factor for cerium oxide powder

The shape factor (X B ) of bulk powder CeO 2 (cerium oxide) particles is estimated to be X B = 1.34 (dimensionless). This is defined in relation to the measured aerated bulk density (0.91 g cm -3 , Hosokawa Micron, Summit NJ, 2018) of the CeO 2 powder. This information is used to determine the RRF (respirable release factor), and other parameters during drop test experiments of nuclear material storage containers at Los Alamos National Laboratory. The calculation is based on an experimental shape factor (X P = 2.66 dimensionless) for CeO 2 particles of the same (chain aggregate) morphology studied by other researchers. In that work, the shape factor was defined in terms of the maximum possible (compacted) density of CeO 2 material (7.13 g cm -3 ).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

A comprehensive final summary of the MCNP calculations performed concerning the UW 137 Cs release event

This document combines and supersedes earlier reports (e.g. LA-UR-19-32358 and LA-UR-20- 22216) that discussed MCNP-calculated conversion factors (count rate per unit activity per unit area) for collimated 2”x2” NaI detectors (Ludlum model 44-10 and Thermo model SPA3) in contact with various surfaces of interest to the University of Washington 137 Cs release event. These surfaces include: Exterior surface of round and rectangular ducts, Exterior surface of outdoor brick work, Exterior surface of cast iron pipes, Exterior surface of an O 2 filter assembly For each surface, the calculated conversion factors allowed inferences on 137 Cs activity on interior surfaces based on exterior contact measurements of the 662 keV photopeak count rate. The MCNP calculations included the effect of attached collimators (i.e. either a Pb/Sb-based Ludlum or a W-based “Rees” model) on detector response. As a consequence, the conversion factors only apply to the appropriate collimated 2”x2” NaI detector operated in pulse height analysis (PHA) mode. Other assumptions affecting the calculations and valid application of the conversion factors will be outlined below.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Evaluation report on the Thermo Scientific TruDose (neutron,gamma) EPD3

This report summarizes an evaluation study on the Thermo Scientific TruDose n/γ EPD3 by LANL’s RP-SVS group. Two EPD3s were received for assessment (SNs 7500100 and 7500108) along with an EPD3 desktop reader. The evaluation included intercomparisons with Thermo’s N2 EPD which is in wide circulation at LANL but is no longer commercially available. A favourable assessment of the TruDose EPD3 would help support it as a viable option to replace the N2 EPD. The data and observations included in this report could also serve as the basis of a Technical Basis Document for the use of the EPD3 at LANL.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

RCT: Module 2.07 Respiratory Protection [Slides]

Internal dosimetry controls require the use of engineering controls to prevent the internal deposition of radioactive and non-radiological contaminants. However, when engineering and administrative controls are not available or feasible, respiratory protection may be necessary. The RCT should know and apply the considerations used in determining the respiratory protection equipment that is most appropriate for the job. The inappropriate use of or the use of the wrong respiratory protection equipment may result in undesirable health effects.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Peripheral Blood Transcript Signatures after Internal 131 I-mIBG Therapy in Relapsed and Refractory Neuroblastoma Patients Identifies Early and Late Biomarkers of Internal 131 I Exposures

131 I-metaiodobenzylguanidine ( 131 I-mIBG) is a targeted radiation therapy developed for the treatment of advanced neuroblastoma. We have previously shown that this patient cohort can be used to predict absorbed dose associated with early 131 I exposure, 72 h after treatment. We now expand these studies to identify gene expression differences associated with 131 I-mIBG exposure 15 days after treatment. Total RNA from peripheral blood lymphocytes was isolated from 288 whole blood samples representing 59 relapsed or refractory neuroblastoma patients before and after 131 I-mIBG treatment. We found that several transcripts predictive of early exposure returned to baseline levels by day 15, however, selected transcripts did not return to baseline. At 72 h, all 17 selected pathway-specific transcripts were differentially expressed. Transcripts CDKN1A (P < 0.000001), FDXR (P < 0.000001), DDB2 (P < 0.000001), and BBC3 (P < 0.000001) showed the highest up-regulation at 72 h after 131 I-mIBG exposure, with mean log2 fold changes of 2.55, 2.93, 1.86 and 1.85, respectively. At day 15 after 131I-mIBG, 11 of the 17 selected transcripts were differentially expressed, with XPC, STAT5B, PRKDC, MDM2, POLH, IGF1R, and SGK1 displaying significant up-regulation at 72 h and significant down-regulation at day 15. Interestingly, transcripts FDXR (P = 0.01), DDB2 (P = 0.03), BCL2 (P = 0.003), and SESN1 (P < 0.0003) maintained differential expression 15 days after 131 I-mIBG treatment. These results suggest that transcript levels for DNA repair, apoptosis, and ionizing radiation-induced cellular stress are still changing by 15 days after 131 I-mIBG treatment. Our studies showcase the use of biodosimetry gene expression panels as predictive biomarkers following early (72 h) and late (15 days) internal 131 I exposure. Our findings also demonstrate the utility of our transcript panel to differentiate exposed from non-exposed individuals up to 15 days after exposure from internal 131 I.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2021

Reference document for LANL stack sampling and ANSI N13.1 (Article) Gielow RL and McNamee MR 1993. Numerical Flue Gas Flow Modeling for Continuous Emissions Monitoring Applications. EPRI CEM Users Group Meeting. Baltimore. RP1961-13

American National Standard N13.1 “sets forth guidelines and performance criteria for sampling the emissions of airborne radioactive substances in the air discharge ducts and stacks of nuclear facilities. Emphasis is on extractive sampling from a location in a stack or duct where the contaminant is well mixed. At such a location, sampling may be conducted at a single point. This standard provides performance-based criteria for the use of air sampling probes, transport lines, sample collectors, sample monitoring instruments, and gas flow measuring methods. This standard also covers sampling program objectives, quality assurance issues, developing air sampling action levels, system optimization, and system performance verification. Workplace, containment, and environmental air monitoring are not addressed. Specific sample analysis methods and the reporting or interpreting of results are also not addressed.” (HPS 2011).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Aerosol Engineering Facility 2021-22 summary sheet [Slides]

The Aerosol Engineering Facility solves problems that involve radioactive particulates, including HEPA air filters, storage containers, exhaust stacks, and continuous air monitors. Other work includes silica dust sampling, bioaerosols, and insecticide sprays. Aerosol technology science describes the behavior of microscopic and nano-sized particles, in both molecular and continuum gas flow regimes.

36 MATERIALS SCIENCE↗
DOE OSTI2020

General Employee Radiological Refresher Training (Self-Study 47968)

This General Employee Radiological Training (GERT) course is designed to refresh knowledge of basic radiological principles and terms, the risks of exposure to ionizing radiation, LANL radiological hazards and postings, ALARA tools, emergency response responsibilities, and how to obtain personal radiological dose records.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

RCT Continuing Training: Item Removal

Welcome to RCT Continuing Training. This quarter's topic will review the fundamental elements of removing items from controlled areas to include: evaluation if an item may be released and the actions necessary to perform a release from a controlled area. To complete this training, you will be required to view the online presentation and independently complete its associated quiz.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Chelation Modeling: The Use of Ad Hoc Models and Approaches to Overcome a Dose Assessment Challenge

Here, chelating agents are administered to treat significant intakes of radioactive elements such as plutonium, americium, and curium. These drugs may be used as a medical countermeasure after radiological accidents and terrorist acts. The administration of a chelating agent, such as Ca-DTPA or Zn-DTPA, affects the actinide's normal biokinetics. It enhances the actinide's rate of excretion, posing a dose assessment challenge. Thus, the standard biokinetic models cannot be directly applied to the chelation-affected bioassay data in order to assess the radiation dose. The present study reviews the scientific literature, from the early 1970s until the present, on the different studies that focused on developing new chelation models and/or modeling of bioassay data affected by chelation treatment. Although scientific progress has been achieved, there is currently no consensus chelation model available, even after almost 50 y of research. This review acknowledges the efforts made by different research groups, highlighting the different methodology used in some of these studies. Finally, this study puts into perspective where we were, where we are, and where we are heading in regards to chelation modeling.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Applying ALARA Principles in the Design of New Radiological Facilities

The application of ALARA (As Low As Reasonably Achievable) principles to the design of new radiological facilities at Argonne National Laboratory provides a consistent radiation safety basis for future facility operations. The Department of Energy Rule 10 CFR 835 specifies design objectives to be adopted during the design of new facilities for controlling personal radiation exposure. One is to keep exposure levels below 20% of the applicable standards in 10 CFR 835.202. For a radiation worker 20% of the standard corresponds to 5 µSv hr -1 for a 2,000 hour work year. For a member of the public the DOE design objective corresponds to 0.2 mSv in a calendar year. For a new facility worst case radionuclides and their source strengths are chosen. Local shielding is specified to reduce exposure rates to less than 50 µSv hr -1 at 30 cm from the shielding. The current version MCNP6 of the Los Alamos radiation shielding computer program MCNP is then used to calculate the exposure rates elsewhere. Design modifications are made to meet the criteria. The calculations and resulting facility design modifications are discussed for two new radiological facilities.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Biokinetic and Dosimetric Models [Slides]

Presentation describes internal dosimetry as an intersection of disciplines involving Physiology, Anatomy, Physics, Mathematics and Computer Science, which some people call "The art of Internal Dosimetry". The basic components for internal dose calculations are: Biokinetic (metabolic) models describing the intake, distribution, retention and excretion of radionuclides in the body, Dosimetric models describing the interaction of radiation within the several body organs and tissues, and System of dose limitation.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Chelation Model Validation: Modeling of a Plutonium-238 Inhalation Incident Treated with DTPA at Los Alamos National Laboratory

Accidental inhalation of plutonium at the workplace is a non-negligible risk, even when rigorous safety standards are in place. The intake and retention of plutonium in the human body may be a source of concern. Thus, if there is a suspicion of a significant intake of plutonium, medical countermeasures such as chelation treatment may be administered to the worker. The present work aimed to interpret the bioassay data of a worker involved in an inhalation incident due to a glovebox breach at Los Alamos National Laboratory’s plutonium facility. The worker was treated with intravenous injections of calcium salts of diethylenetriaminepentaacetic acid (DTPA) in an attempt to reduce the amount of plutonium from the body and therefore reduce the internal radiation dose. It is well known in the internal dosimetry field that the administration of chelation treatment poses additional challenges to the dose assessment. Hence, a recently developed chelation model was used for the modeling of the bioassay data. The objectives of this work are to describe the incident, model the chelation-affected and non-affected bioassay data, estimate the plutonium intake, and assess the internal radiation dose.

238Pu↗
DOE OSTI2023

Early-Stage Radiation Safety Analysis for the Spallation Neutron Source Second Target Station Bunker Operations

The Second Target Station project at Oak Ridge National Laboratory will develop a cold neutron source to meet growing experimental needs. This paper describes calculations of the residual dose rates associated with the monolith shield plug and the beamline bunker, two key conventional operations and radiation safety features. While neutron production is active, the instrument hall outside the bunker must be generally accessible with dose rates of less than 0.25 mrem/h. When neutron production is halted, the bunker must be accessible for hands-on maintenance operations. Further, these two requirements form the cause for the assessments reported herein of residual dose rates caused by the monolith shield plug and residual dose rates in the bunker. The monolith shield plug was shown to not produce significant dose rates inside the bunker after a 20-year lifetime, and the residual dose rates inside the bunker for the case of an operating beamline were shown to reasonably allow for hands-on maintenance. These calculations are based on preliminary design models of the relevant systems. Additionally, an example showing the significance of considering neutron supermirror physics in transport calculations that track nuclide production and destruction rates to produce gamma sources for residual dose rate calculations is included. The example shows that if neutron supermirror physics is not considered, dose rate fields may be significantly underpredicted.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Modeling Plutonium Decorporation in a Female Nuclear Worker Treated with Ca-DTPA after Inhalation Intake

The present work models plutonium (Pu) biokinetics in a female former nuclear worker. Her bioassay measurements are available at the US Transuranium and Uranium Registries. The worker was internally exposed to a plutonium-americium mixture via acute inhalation at a nuclear weapons facility. She was medically treated with injections of 1 g Ca-DTPA on days 0, 5, and 14 after the intake. Between days 0 and 20, fecal and urine samples were collected and analyzed for 239 Pu and 241 Am. Subsequently, she was followed up for bioassay monitoring over 14 y, with additional post-treatment urine samples collected and analyzed for 239 Pu. The uniqueness of this dataset is due to the availability of: (1) both early and long-term bioassay data from a female with plutonium intake; (2) data on chelation therapy for a female; and (3) fecal measurement results. Chelation therapy with Ca- and/or Zn-salts of DTPA is known to aid in reducing the internal radiation dose by enhancing the excretion of plutonium and americium from the body. Such enhancement affects plutonium biokinetics in the human body, posing a challenge to the internal dose assessment. The current radiation dose assessment practice is to exclude the data affected by Ca-DTPA from the analysis. The present analysis is the first to explicitly model the chelation-affected bioassay data in a female by using a newly developed chelation model. Thus, the bioassay data collected during and after the Ca-DTPA administrations were used for biokinetic modeling and dose assessment. The Markov Chain Monte Carlo method was used to investigate model parameter uncertainty, based on the bioassay data and assumed prior probability distributions. A χ 2 /nData (number of data points) ≈ 1 was observed in this study, which indicates self-consistency of the data with the model. Results of this study show that the worker’s 239 Pu intake was 12 Bq, with a committed effective dose to the whole-body of 1.2 mSv and a committed equivalent dose to the bone surfaces, liver, and lungs of 37.8, 9.1, and 0.8 mSv, respectively. This study also discusses the worker’s dose reduction due to chelation treatment.

61 RADIATION PROTECTION AND DOSIMETRY↗