Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Radiation protection”

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 901 records · Page 50
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Analysis of Air-Purifying Respirator (APR) and Powered Air-Purifying Respirator (PAPR) Cartridge Performance Testing on Hanford Tanks BY-108 and BY-110 Volume 1

Washington River Protection Solutions (WRPS) tested four types of chemical cartridges for use in air-purifying respirators (APR) and powered air-purifying respirators (PAPR). These tests were undertaken to determine the period of time that the cartridges would provide adequate performance for APRs and PAPRs used to protect workers when exposed to a mixture of Chemicals of Potential Concern (COPC) from vapors exiting the headspaces of Hanford BY-108 and BY-110 waste storage tanks. The Occupational Safety and Health Administration (OSHA) considers cartridge testing to be a valid approach for establishing cartridge change schedules. Testing commonly is applied in situations where mixtures of COPCs exist, and where other approaches, such as manufacturer recommendations and modeling, are less reliable. The tests were designed and conducted to assure measurement and/or control of the key variables OSHA identified as important to estimate the cartridge service life, including temperature, humidity, COPC concentration, breathing rate, and cartridge adsorption capacity. Cartridge tests were conducted over two days from February 23-25, 2018, using headspace vapors from Hanford tanks BY-108 and BY-110. Headspace vapors from the BY-108 tank were fed to the PAPR respirator cartridge test stand, while vapors from the BY-110 headspace were fed to the APR respirator cartridge test stand. Both the APR and PAPR test stands were developed by WRPS in collaboration with HiLine Engineering (Richland, Washington). Two different cartridges were assessed on each tank. Multipurpose APR cartridges—SCOTT 7422-SD1 and SCOTT 7422-SC1 (SCOTT Safety, Monroe, North Carolina)—were assessed for vapors from BY-110 using the APR cartridge test stand. Multipurpose PAPR cartridges—MSA OptiFilter TL (MSA Safety Inc., Pittsburgh, Pennsylvania) and 3M FR-57 (3M Company, Maplewood, Minnesota)—also were assessed for vapors from BY-108 over the same two days using the PAPR cartridge test stand. Sample media (i.e., sorbent tubes) were used to collect samples of the vapor stream entering and exiting the respirator cartridges, and the samples then were analyzed for COPC concentrations. Pacific Northwest National Laboratory was tasked with conducting an independent analysis of the analytical results and making recommendations based on the results for respiratory cartridge performance and service life. The key conclusions from the analysis are described below.

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

Analysis of Air-Purifying Respirator (APR) and Powered Air-Purifying Respirator (PAPR) Cartridge Performance Testing on a Hanford AP Tank Farm Exhauster Slipstream (Volume 1)

Washington River Protection Solutions (WRPS) tested four types of chemical cartridges for use in air-purifying respirators (APR) and powered air-purifying respirators (PAPR). These tests were undertaken to determine the period of time that the cartridges would provide adequate performance1 for APRs and PAPRs used to protect workers when exposed to a mixture of Chemicals of Potential Concern (COPC) from vapors exiting the Hanford AP tank farm exhauster slipstream. The Occupational Safety and Health Administration (OSHA) considers cartridge testing to be a valid approach for establishing cartridge change schedules. Testing commonly is applied in situations where mixtures of COPCs exist and where other approaches, such as manufacturer recommendations and modeling, are less reliable. The tests were designed and conducted to assure measurement and/or control of the key variables OSHA identified as important to estimate the cartridge service life, including temperature, humidity, COPC concentration, breathing rate, and cartridge adsorption capacity. Cartridge testing on a slipstream from the Hanford AP tank exhauster was conducted on March 23-24, 2018. This testing focused on both APR and PAPR cartridges. Previous testing of APR cartridges was conducted on the AP exhauster in June of 2016. However, an AP exhauster upgrade was completed in September 2016. In the most recent testing, slipstream vapors from the new AP exhauster were fed to two respirator cartridge test stands, one for the PAPR respirator cartridges and the other for the APR respirator cartridges. Both the APR and PAPR test stands were developed by WRPS in collaboration with HiLine Engineering (Richland, Washington). Multipurpose APR cartridges—SCOTT 7422-SD1 and SCOTT 7422-SC1 (SCOTT Safety, Monroe, North Carolina)—were assessed on separate days using the APR cartridge test stand. Multipurpose PAPR cartridges—MSA OptiFilter TL (MSA Safety Inc., Pittsburgh, Pennsylvania) and 3M FR-57 (3M Company, Maplewood, Minnesota)—were also tested consecutively over the same two days as the APR cartridge tests, using the PAPR cartridge test stand. Sample media (i.e., sorbent tubes) were used to collect samples of the vapor stream entering and exiting the respirator cartridges, and the samples then were analyzed for COPC concentrations. Pacific Northwest National Laboratory was tasked with conducting an independent analysis of the analytical results and making recommendations based on the results for respiratory cartridge performance and service life. The key conclusions from the analysis are described below.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

New Beginnings: Optimising Waste and End-state at UK Nuclear Sites to Enable Environmental Permit surrender (the GRR) - 20193

The decommissioning of nuclear sites involves the management of large volumes of radioactive waste, other conventional waste, and the clean-up of land and groundwater contamination. In the UK, the disposal of radioactive waste from nuclear sites is regulated by the Environment Agency (England), Natural Resources Wales (Wales) and the Scottish Environment Protection Agency (Scotland). All three environment agencies seek to make sure that the generation and disposal of radioactive waste is optimised: ensuring impacts to people are 'as low as reasonably achievable' (ALARA). With a large proportion of UK nuclear sites embarking on decommissioning, the environment agencies have collaboratively developed new guidance: the 'Management of radioactive waste from the decommissioning of nuclear sites: Guidance on the Requirements for Release from radioactive substances regulation' (referred to as the GRR)[1]. The environment agencies worked in liaison with the Office for Nuclear Regulation (ONR), the UK nuclear safety regulator, and the UK government body responsible, the Nuclear Decommissioning Authority (NDA). Prior to finalising the guidance, a 12 month trial was completed at three of the NDA's nuclear sites, enabling the environment agencies to test their draft GRR guidance against 'real world scenarios' and the challenge of incorporating its requirements into decommissioning plans. Experience gained during the trial was incorporated into the current version of the GRR, published in July 2018. The GRR represents the 'new beginning' of this paper's title because for the first time, the UK environmental regulators have expressed in detail their requirements for surrender of environmental permits for nuclear sites. The GRR provides a framework for a UK nuclear site to plan, assess and justify its management of radioactive waste and deliver through decommissioning, an optimal, environmentally safe and sustainable end-state. This paper will describe the requirements and technical standards within the GRR, the establishment of defensible clean-up criteria and the production of a Waste Management Plan (WMP), supported by a Site-Wide Environmental Safety Case (SWESC). It will describe the experience gleaned by operators and regulators during the trial and it will highlight current work by regulators to implement the new approach proportionately. Finally, this paper will look forward and outline planned amendments to legislation in the UK affecting nuclear sites that should enable the full benefits of the GRR to be realised in the future. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Bioaccumulation Factor for Radium in Rio Grande Biota

The Los Alamos National Laboratory (LANL or Laboratory) Environmental Protection and Compliance Division’s Environmental Stewardship Group (EPC-ES) performs the radiological analysis portion of biota dose assessments using the computer program RESRAD-Biota (DOE 2004, DOE 2020). RESRAD-Biota uses a default bioaccumulation factor, B iv , for radium in aquatic animals. This bioaccumulation factor, also listed in Department of Energy (DOE) Standard, DOE-STD-1153-2019, is the ratio of a radionuclide concentration in fresh weight biota to the environment in which it resides, which for aquatic animals is water. DOE-STD-1153-2019 offers conservative values for these ratios that are specific to the radiological contaminant, the biota, and the environment. The B iv in RESRAD-Biota can be used for general screening (Level 1 screening). If the factor proves to be too conservative, then a site-specific B iv may be used. The biota dose assessment of fish from the northern New Mexico region of the Rio Grande fail the general screen, which indicates that further analysis is necessary. Level 2 analysis is site-specific but still has conservative limitations. Level 3 analysis is a more realistic and representative, sitespecific analysis. The purpose of this paper is to use data specific to the Rio Grande to calculate a new aquatic animal radium B iv . This B iv can then be used for Level 3 analysis of aquatic animals specifically in the northern New Mexico region of the Rio Grande. The calculated radium B iv is a useful value for performing detailed, accurate biota dose assessments and calculating dose to fish using RESRAD-Biota when only sediment data are available.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

MCNP® Code Version 6.3.1 Release Notes

The Monte Carlo N-Particle® (MCNP® ) code is a general-purpose, continuous-energy, generalized-geometry, time-dependent, radiation transport code developed by the MCNP development team. MCNP calculations provide predictive capabilities that can replace expensive or impossible-to-perform experiments. Specific application problems include simulations of experimental diagnostics, intrinsic radiation, radiation detection and measurement, criticality safety, nuclear threat reduction and response, radiation health protection, nuclear weapons effects, and nuclear forensics. This MCNP code, version 6.3.1, follows the MCNP6.3.0 version. Since the release of MCNP6.3.0, a variety of bug fixes and code enhancements have been completed for MCNP6.3.1. A few new features have also been added to this release to support both ongoing research and the release of the latest ENDF/B-VIII.1 nuclear data library. The MCNP code, version 6.3.1, theory and user input information is documented in MCNP® Code Version 6.3.1 Theory & User Manual, the build guidance for various platforms is documented in MCNP® Code Version 6.3.1 Build Guide, and the verification and validation testing for various application benchmark test suites is documented in MCNP® Code Version 6.3.1 Verification & Validation Testing.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Managing Hanford's Direct Feed Low Activity Waste Program - 20002

The Office of River Protection (ORP) is responsible for the management and completion of the River Protection Project (RPP) mission, which comprises both the Hanford Site tank farms operations and the completion and operation of the Waste Treatment and Immobilization Plant Project. The RPP mission is to safely retrieve, treat, and immobilize Hanford's high-level and low-activity tank waste and close the tank farms to protect the Columbia River. The Direct-Feed Low-Activity Waste (DFLAW) Program is a major subset of the overall RPP mission, providing disposition of low-activity waste. The scope is to retrieve, treat, and immobilize the low-activity waste at the Low-Activity Waste Facility. DFLAW will begin operations no later than December 2023. The DFLAW Program overarches a suite of individual projects, activities, and infrastructure upgrades. The major projects and activities include multiple contractors under both ORP and the Richland Operations Office, underscoring the need for effective integration, coordination, and collaboration. The objectives of the DFLAW Program leadership team are to: - Effectively coordinate and integrate the projects that comprise the DFLAW Program - Manage the interfaces between the projects so that the integrated DFLAW Program is completed successfully - Ensure the DFLAW portfolio of projects operate as required without gaps or conflicts at the interfaces. DOE is accomplishing these objectives through a newly established leadership model that integrates the DOE and contractor work streams and promotes a teamwork model to enable collaborative success. Long considered one of the most formidable cleanup challenges at Hanford, the Department of Energy and its contractors are on the verge of achieving a cleanup commitment that has been decades in the making. DOE is preparing to vitrify (turn to glass) Hanford's chemical and radioactive tank waste using the DFLAW process. The Modified Consent Decree milestone for completing hot commissioning of the Hanford Site's Low- Activity Waste Facility is Dec. 31, 2023. While this milestone may seem distant under normal project management circumstances, it is a relatively short period of time for the operational and cultural transformation necessary to successfully begin treating tank waste at Hanford. Direct feed means separating the waste at a tank farm to remove the more radioactive portion (i.e., solids and cesium) so that the resulting low-activity (less radioactive) waste can be fed directly to the Waste Treatment and Immobilization Plant's Low-Activity Waste Facility. Starting DFLAW will require a singular, intensive leadership focus, along with a sustained, collective commitment to excellence and teamwork by the Department and its contractors. Major upgrades to Hanford's infrastructure will occur prior to startup. Supporting DFLAW operations requires significant integration between the DFLAW facilities - the Waste Treatment Plant, Hanford's tank farms, effluent treatment facilities, disposal facilities - and between all of the site contractors to achieve success. (authors)

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

Analysis of Respirator Cartridge Performance Testing on Hanford Tank A-101

Washington River Protection Solutions (WRPS) conducted tests on two types of chemical cartridges for use in air purifying respirators (APR) to determine the period of time that the cartridges would provide adequate performance1 to protect workers when exposed to a mixture of Chemicals of Potential Concern (COPCs) from vapors emanating from the headspace of tank A-101 on the Hanford Site. The Occupational Safety and Health Administration (OSHA) identifies cartridge testing as a valid approach for establishing a cartridge change schedules.[3] Testing is commonly applied in situations where mixtures of COPCs exist, and where other approaches, such as manufacturer recommendations and modeling, are less reliable. The tests were designed and conducted to ensure measurement and/or control of the key variables OSHA identified as important to estimate cartridge service life, including temperature, humidity, COPC concentration, breathing rate, and cartridge adsorption capacity. Testing was conducted over a period from July 22-24, 2016, using headspace vapors from Hanford tank A-101 under static conditions2 fed to a respirator cartridge test stand developed by WRPS in collaboration with HiLine Engineering (Richland, Washington). Multipurpose respirator cartridges, SCOTT 7422-SD1 and 7422-SC1 (SCOTT Safety, Monroe, North Carolina) were assessed on separate days with A-101 headspace vapors. Sample media (sorbent tubes) were used to collect samples of the vapor stream entering and exiting the respirator cartridge, and were subsequently analyzed for COPC concentrations. Pacific Northwest National Laboratory was tasked with conducting an independent analysis of the analytical results and making recommendations based on the results for respiratory cartridge performance and service life. Key conclusions from the assessment of the 59 COPCs in this study are described below: • Based on d measured cartridge inlet vapor concentrations from tank A-101 headspace, both ammonia and N-Nitrosodimethylamine (NDMA) exceeded their Occupational Exposure Limits (OEL).3 These measurements were generally consistent with maximum A-101 headspace measurements previously obtained for these two compounds. • Ammonia had respirator cartridge outlet concentrations that exceeded 10% of the OEL for both cartridges tested, indicating breakthrough for each. For the SCOTT 7422-SD1 cartridge, ammonia appeared to breakthrough above 10% of its OEL after 2 hours. For the SCOTT 7422-SC1 cartridge, ammonia breakthrough appeared to occur after 8 hours. 1 “Adequate performance” refers to the breakthrough criteria used in this analysis—that is, sustained cartridge outlet concentrations above 10% of the OEL. For some COPCs, an alternate threshold has been applied when necessary due to higher detection limit (DL)/reporting limit (RL) values for specific compounds. Ultimately, Industrial Hygiene professionals will use these results along with specific hazard assessments to determine service life and cartridge use that provides the necessary performance. 2 These tests were conducted under static conditions absent waste-disturbing activities in the subject tank or tank farm. 3 Occupational Exposure Limits accepted for Hanford Tank Farm use are based on OELs established by a U.S. governmental agency or national professional organization (e.g., OSHA, National Institute for Occupational Safety and Health, American Conference of Governmental Industrial Hygienists), or if no U.S. OEL exists, standard toxicological practices are applied to develop OELs using non-U.S. exposure limits, other established OELs for chemical surrogates when available, or other standard procedures. The OEL for NDMA was established in 2005 based on the MAK (Maximale Arbeitsplatzkonzentration) Commission standard adopted in Europe. iv • Despite respirator inlet measurements for NDMA that were in excess of its OEL, all corresponding outlet measurements, from both respirator cartridges, were below analytical DL 1 for the

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Monte Carlo simulation of shielding designs for a cabinet form factor preclinical MV–energy photon FLASH radiotherapy system

Purpose: A preclinical MV-energy photon FLASH radiotherapy system is being designed at Stanford and SLAC National Accelerator Laboratory. Because of the higher energy and dose rate compared to conventional kV-energy photon laboratory-scale irradiators, adequate shielding in a stand-alone cabinet form factor is more challenging to achieve. We present a Monte Carlo simulation of multilayered shielding for a compact self-shielding system without the need for a radiation therapy vault. Methods: A multilayered shielding approach using multiple alternating layers of high-Z and low-Z materials is applied to the self-shielded cabinet to effectively mitigate the secondary radiation produced and to allow the device to be housed in a Controlled Radiation Area outside of a radiation vault. The multilayered shielding approach takes advantage of the properties of high-Z and low-Z radiation shielding materials such as density, cross-section, atomic number of the shielding elements, and products of radiation interactions within each layer. The Monte Carlo radiation transport code, FLUKA, is used to simulate the total effective dose produced by the operation. Results: The multilayered shielding designs proposed and simulated produced effective dose rates significantly lower than monolayer designs with the same total material thickness at the regulatory boundary; this is accomplished through the manipulation of the locations where secondary radiation is produced and reactions due to material properties such as neutron back reflection in hydrogen. Borated polyethylene at 5 wt% significantly increased the shielding performance as compared to regular polyethylene, with the magnitude of the reduction depending upon the order of the shielding material. Conclusions: The multilayered shielding provides a path for shielding preclinical FLASH systems that deliver MV-energy bremsstrahlung photons. Furthermore, this approach promises to be more efficient with respect to the shielding material mass and space claim as compared to shielded vaults typically required for clinical radiation therapy with MV photons.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2023

X-ray Inspection Model Validation with Physical Dosimetry

A non-functional printed circuit board assembly was developed using ULTIboard and a simple computational transport model was created for use with the Monte Carlo software MCNP. EBT 3 film radiography was used to compare with Monte Carlo simulation to validate the board representation. The computational transport model was then revised to include connection pins, solder balls, highly attenuating internal structures, and copper trace distribution. Results from the revised model were compared to the EBT 3 films to observe improvements to dose profiles. It was found that this method was useful in verifying the placement of components, as the dose profiles were observed to follow the same trends. The experiment was then repeated using XRQA 2 films to achieve the same level of contrast with 1% the dose of EBT 3 films. It was found that high contrast may be achieved using these films to identify major issues with the model geometry, at a cost of dose profile accuracy. A second validation method was applied to the model using 37 CaF 2 thermoluminescent dosimeters (TLDs). TLD measurements were compared with the simplified and complex transport models to identify the features that have the greatest impact on simulation accuracy. The TLD calibration to CaF 2 was found to be accurate within 5.6%, while calibration to dose in Si was found to be accurate within 4.7%. Finally, it was observed that the accurate representation of solder balls and proper modeling of highly attenuating internal structures had the greatest impact on simulation accuracy.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Characterization of 244 Cm Neutron Sources

There is a need to find a replacement for the 252 Cf sources currently used in Standards-based performance testing of neutron instrumentation. This is a result of its relatively short half-life, concerns about future availability and recent increases in cost. A potential replacement source, 244 Cm is evaluated in this study where two commercially available sources have been acquired and their neutron emission spectra measured using a pair of spectrometers. Both instruments showed the 244 Cm spontaneous neutron spectrum to be fully consistent with a MCNP6®-calculated spectrum using published Watt fission parameters for 244 Cm. In addition, the emission rate of the weaker 244 Cm source was established through a direct count rate comparison against a calibrated and similarly encapsulated 252 Cf source. The 244 Cm source emission rate was found to be in excellent agreement with value stated on the manufacturer's source certificate. It is concluded that 244 Cm would be an ideal replacement for 252 Cf based on its longer half-life (18.1 y) and its essentially identical neutron emission spectrum. In addition, 244 Cm sources are commercially available at reasonable cost.

244Cm↗
DOE OSTI2022

Public Health Response and Medical Management of Internal Contamination in Past Radiological or Nuclear Incidents: A Review

Following a radiological or nuclear emergency, workers, responders and the public may be internally contaminated with radionuclides. Screening, monitoring and assessing any internal contamination and providing necessary medical treatment, especially when a large number of individuals are involved, is challenging. Experience gained and lessons learned from the management of previous incidents would help to identify gaps in knowledge and capabilities on preparedness for and response to radiation emergencies. In this paper, eight largescale and five workplace radiological and nuclear incidents are reviewed cross 14 technical areas, under the broader topics of emergency preparedness, emergency response and recovery processes. The review findings suggest that 1) new strategies, algorithms and technologies are explored for rapid screening of large populations; 2) exposure assessment and dose estimation in emergency response and dose reconstruction in recovery process are supported by complementary sources of information, including ‘citizen science’; 3) surge capacity for monitoring and dose assessment is coordinated through national and international laboratory networks; 4) evidence-based guidelines for medical management and follow-up of internal contamination are urgently needed; 5) mechanisms for international and regional access to medical countermeasures are investigated and implemented; 6) long-term health and medical follow up programs are designed and justified; and 7) capabilities and capacity developed for emergency response are sustained through adequate resource allocation, routine nonemergency use of technical skills in regular exercises, training, and continuous improvement.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Spatial and temporal prediction of radiation dose rates near Fukushima Daiichi Nuclear Power Plant

In this paper, we have developed a methodology to estimate the spatiotemporal distribution of radiation air dose rates around the Fukushima Daiichi Nuclear Power Plant (FDNPP). In our exploratory data analysis, we found that (1) the temporal evolution of dose rates is composed of a log-linear decay trend and fluctuations of air dose rates that are spatially correlated among adjacent monitoring posts; and (2) the slope of the log-linear environmental decay trend can be represented as a function of the apparent initial dose rates, coordinate position, land-use type, and soil type. From these observations, we first estimated the log-linear decay trend at each location based on these predictors, using the random forest method. We then developed a modified Kalman filter coupled with a Gaussian process model to estimate the dose-rate time series at a given location and time. We applied this method to the Fukushima evacuation zone (as of March 2017), which included 17 monitoring post locations (with monitoring datasets collected between 2014 and 2018) and generated a time series of dose-rate maps. Our results show that this approach allows us to produce accurate spatial and temporal predictions of radiation dose-rate maps using limited spatiotemporal measurements.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Conditional Point Sampling: A Monte Carlo Method for Radiation Transport in Stochastic Media.

Current methods for stochastic media transport are either computationally expensive or, by nature, approximate. Moreover, none of the well-developed, benchmarked approximate methods can compute the variance caused by the stochastic mixing, a quantity especially important to safety calculations. Therefore, we derive and apply a new conditional probability function (CPF) for use in the recently developed stochastic media transport algorithm Conditional Point Sampling (CoPS), which 1) leverages the full intra-particle memory of CoPS to yield errorless computation of stochastic media outputs in 1D, binary, Markovian-mixed media, and 2) leverages the full inter-particle memory of CoPS and the recently developed Embedded Variance Deconvolution method to yield computation of the variance in transport outputs caused by stochastic material mixing. Numerical results demonstrate errorless stochastic media transport as compared to reference benchmark solutions with the new CPF for this class of stochastic mixing as well as the ability to compute the variance caused by the stochastic mixing via CoPS. Using previously derived, non-errorless CPFs, CoPS is further found to be more accurate than the atomic mix approximation, Chord Length Sampling (CLS), and most of memory-enhanced versions of CLS surveyed. In addition, we study the compounding behavior of CPF error as a function of cohort size (where a cohort is a group of histories that share intra-particle memory) and recommend that small cohorts be used when computing the variance in transport outputs caused by stochastic mixing.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Dose conversion factors for absorbed dose in a mobile phone to absorbed dose in critical organs in an anthropomorphic phantom for emergency dosimetry applications: OSL and TL experimental results, and Monte Carlo simulations

Here, an anthropomorphic phantom was equipped with three smartphones and irradiated in five separate exposure geometries in air using a 137 Cs source. The phantom was fitted internally with thermoluminescence dosimeters (TLD-700 chips) to measure the absorbed dose to the individual organs, while the absorbed doses in the phones were evaluated using optically stimulated luminescence (OSL) from the phones’ surface mount resistors (SMRs), along with reference dosimeters, namely Luxel OSL films and TLD-700 chips. The results indicate a strong dependence on the phone location with respect to the source due to shielding and scattering from the body. Dose conversion factors were calculated for several plausible exposure scenarios in order to correlate the absorbed doses measured in the phones to the average body dose received by the phantom. The measured conversion factors were compared with calculated factors using MCNP6.2 Monte Carlo simulations and an ADAM voxel phantom. Experiments were conducted to assess how the phone orientation and angle impacts the dose received by the phone components. The overall results indicate that the doses measured can vary significantly depending upon exposure geometry (including phone position and angle). However, in rotational irradiation symmetry, the phone doses lead to an over-estimate of the average whole-body dose by an average of ~10%, independent of the phone locations tested in this study. This result, if confirmed in future studies, may be acceptable for triage.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

A review of criticality dosimetry at the Y-12 National Security Complex and practical importance of dose accuracy in emergency response

A nuclear criticality results in the emission of both neutron and gamma radiation and can produce doses to personnel near the event that exceed 0.1 Gy (10 rad). The primary purpose of nuclear accident dosimetry is to rapidly identify affected personnel in need of prompt medical treatment and to reassure personnel who have been only minimally exposed. While accurate dosimetry is desired, it must be recognized that dose determinations made from whole-body dosimeters or simple triage methods are very rough estimates and contain significant uncertainties. Even when accounting for factors like varying neutron energy spectra, mean photon energies, body orientation within the radiation field, and transient effects on dosimeter response, etc., the end value is a dosimetric quantity defined for very specific radiological conditions and determined within a simple phantom usually at a single depth. Of more importance is the biological response to the radiation, which will vary by person and can be affected by the individual’s radiation sensitivity, age, gender, mass, and underlying health conditions. The overall biological, person-specific response to a given dose cannot be precisely determined except by patient symptom observation and individual biological dosimetry (e.g. chromosome analysis, lymphocyte ratios, etc.). This work describes and discusses the criticality accident dosimetry program at the Y-12 National Security Complex, a United States Department of Energy National Nuclear Security Administration facility. In conclusion, the primary goals of the Y-12 accident dosimetry program are, among others, the rapid identification of significantly exposed persons, prompt routing of exposed workers for medical evaluation and treatment, and the ultimate processing of dosimeters to assign doses to personnel.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2026

Fading study of Harshaw LiF:Mg,Ti thermoluminescence dosimeters exposed to neutron doses

Sensitivity and signal fading were tested in Harshaw LiF:Mg,Ti thermoluminescence dosimeters (TLDs) exposed to neutron and beta/photons doses. A significant difference in the signal fading rate was observed for TLDs exposed to neutrons compared with TLDs exposed to betas. This difference may have been related to differences in glow curves for dosimeters exposed to different beams and to differences in glow curve evolution after exposure. A fading-correction algorithm was developed to correct the cumulative sensitivity + signal fading of TL doses for the fading time t using the formulae: y = 0.147∗exp(-t/4.36) + 0.308∗exp(-t/34.1) + 0.73 for neutrons, and: y = 0.132∗exp(-t/10.72) + 0.174∗exp(-t/31.5) + 0.8 for betas/photons. The formula was validated using the dose recovery test, where the average deviation of fading-corrected doses from delivery doses was within 1%. In conclusion, the proposed fading-correction approaches may significantly improve the accuracy of Harshaw TL dosimetry with LiF:Mg,Ti cards for both neutron and beta/gamma irradiation.

61 RADIATION PROTECTION AND DOSIMETRY↗