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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Silicon-Equivalent Neutron Dose Estimations

Radiation transport calculations are often used to estimate dose or exposure to components and personnel surrounding a radiation source. The sources for these calculations are decaying radionuclides within the various nuclear materials. Under long-term radiation exposure, some components may show signs of material damage such as hardening, cracking, discoloration, etc. resulting from the dose received from nearby radioactive sources. Radiation dose quantities are a useful means to correlate material damage or degradation to radiation interactions. The degradation is often dependent on how incident radiation interacts and deposits energy into the material. This dependence implies that radiation dose should be properly defined for the material effect of interest. Generic dose quantities, such as total energy deposited, may not necessarily correlate well for degradation of any given material.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Methods of improving brain dose estimates for internally deposited radionuclides *

The US National Council on Radiation Protection and Measurements (NCRP) convened Scientific Committee 6–12 (SC 6–12) to examine methods for improving dose estimates for brain tissue for internally deposited radionuclides, with emphasis on alpha emitters. This Memorandum summarises the main findings of SC 6–12 described in the recently published NCRP Commentary No. 31, ‘Development of Kinetic and Anatomical Models for Brain Dosimetry for Internally Deposited Radionuclides’. The Commentary examines the extent to which dose estimates for the brain could be improved through increased realism in the biokinetic and dosimetric models currently used in radiation protection and epidemiology. A limitation of most of the current element-specific systemic biokinetic models is the absence of brain as an explicitly identified source region with its unique rate(s) of exchange of the element with blood. The brain is usually included in a large source region called Other that contains all tissues not considered major repositories for the element. In effect, all tissues in Other are assigned a common set of exchange rates with blood. A limitation of current dosimetric models for internal emitters is that activity in the brain is treated as a well-mixed pool, although more sophisticated models allowing consideration of different activity concentrations in different regions of the brain have been proposed. Here case studies for 18 internal emitters indicate that brain dose estimates using current dosimetric models may change substantially (by a factor of 5 or more), or may change only modestly, by addition of a sub-model of the brain in the biokinetic model, with transfer rates based on results of published biokinetic studies and autopsy data for the element of interest. As a starting place for improving brain dose estimates, development of biokinetic models with explicit sub-models of the brain (when sufficient biokinetic data are available) is underway for radionuclides frequently encountered in radiation epidemiology. A longer-term goal is development of coordinated biokinetic and dosimetric models that address the distribution of major radioelements among radiosensitive brain tissues.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Plutonium in Manhattan Project workers: Using autopsy data to evaluate organ content and dose estimates based on urine bioassay with implications for radiation epidemiology

Purpose Radiation dose estimates in epidemiology typically rely on intake predictions based on urine bioassay measurements. The purpose of this article is to compare the conventional dosimetric estimates for radiation epidemiology with the estimates based on additional post-mortem tissue radiochemical analysis results. Methods The comparison was performed on a unique group of 11 former Manhattan Project nuclear workers, who worked with plutonium in the 1940s, and voluntarily donated their bodies to the United States Transuranium and Uranium Registries. Results Post-mortem organ activities were predicted using different sets of urine data and compared to measured activities. Use of urinalysis data collected during the exposure periods overestimated the systemic (liver+skeleton) deposition of 239 Pu by 155±134%, while the average bias from using post-exposure urinalyses was –4±50%. Committed effective doses estimated using early urine data differed from the best estimate by, on average, 196±193%; inclusion of follow-up urine measurements in analyses decreased the mean bias to 0.6±36.3%. Cumulative absorbed doses for the liver, red marrow, bone surface, and brain were calculated for the actual commitment period. Conclusion On average, post-exposure urine bioassay results were in good agreement with post-mortem tissue analyses and were more reliable than results of urine bioassays collected during the exposure.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2023

Dose estimates and their uncertainties for use in epidemiological studies of radiation-exposed populations in the Russian Southern Urals

Many residents of the Russian Southern Urals were exposed to radioactive environmental pollution created by the operations of the Mayak Production Association in the mid- 20 th century. There were two major releases: the discharge of about 1x10 17 Bq of liquid waste into the Techa River between 1949 and 1959; and the atmospheric release of 7.4 * 10 16 Bq as a result an explosion in the radioactive waste-storage facility in 1957. The releases into the Techa River resulted in the exposure of more than 30,000 people who lived in riverside villages between 1950 and 1961. The 1957 accident contaminated a larger area with the highest exposure levels in an area that is called the East Urals Radioactive Trace (EURT). Current epidemiologic studies of the exposed populations are based on dose estimates obtained using a Monte-Carlo dosimetry system (TRDS-2016MC) that provides multiple realizations of the annual doses for each cohort member. These dose realizations provide a central estimate of the individual dose and information on the uncertainty of these dose estimates. In addition, the correlation of individual annual doses over realizations provides important information on shared uncertainties that can be used to assess the impact of shared dose uncertainties on risk estimate uncertainty.This paper considers dose uncertainties in the TRDS-2016MC. Individual doses from external and internal radiation sources were reconstructed for 48,036 people based on environmental contamination patterns, residential histories, individual 90 Sr body-burden measurements and dietary intakes. Dietary intake of 90 Sr resulted in doses accumulated in active bone marrow (or simply, marrow) that were an order of magnitude greater than those in soft tissues. About 84% of the marrow dose and 50% of the stomach dose was associated with internal exposures. The lognormal distribution is well-fitted to the individual dose realizations, which, therefore, could be expressed and easily operated in terms of geometric mean (GM) and geometric standard deviation (GSD). Cohort average GM for marrow and stomach cumulative doses are 0.21 and 0.03 Gy, respectively. Cohort average dose uncertainties in terms of GSD are as follows: for marrow it is 2.93 (90%CI: 2.02–4.34); for stomach and the other non-calcified tissues it is 2.32 (90% CI: 1.78–2.9).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Methods for Estimating Dose from Galactic Cosmic Rays and Jovian Radiation Belts on Reduced-Order Monte Carlo Geometries

This document outlines the methods and tools developed to estimate dose rates from galactic cosmic rays and the trapped radiation belts surrounding Jupiter using Monte Carlo particle transport software. Furthermore, the techniques described here are specific for translating high-resolution MAVRIC models to analogous but lower resolution Monte Carlo N-Particle Transport (MCNP) code models to estimate charged particle effects on models originally developed to study neutral particles only. This capability was initially developed to predict doses to various components of dynamic radioisotope power systems for spaceflight applications, but the tools and methods can be easily applied to other problem geometries and expanded to other space radiation environments.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2023

Impact to Groundwater All-Pathways Dose Estimates for the Remote-Handled Low-Level Waste Disposal Facility Performance Assessment Using Updated Dose Coefficients from DOE-STD-1196-2022

The Performance Assessment (PA) for the Remote-Handled Low-Level Waste (RHLLW) Disposal Facility at Idaho National Laboratory (INL) was completed in 2018 (DOE-ID 2018) using dose coefficients from U.S. Department of Energy (DOE) Standard DOE-STD-1196-2011 (DOE 2011). Internal and external dosimetry was updated in 2021 and a new technical standard was published in 2022 (DOE-STD-1196-2022) (DOE 2022). This technical memorandum provides a comparison of the ingestion dose coefficients between those published in DOE (2011) and those published in DOE (2022). The dose coefficients in DOE (2022) were then used to calculate the all-pathways dose for the groundwater pathway and the results between the doses published in the 2018 PA and those calculated using the updated dose coefficients in DOE (2022) were compared. Several other issues in the 2018 RHLLW Disposal Facility PA were also addressed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Second Target Station Target Segment PIE Dose Estimates

The purpose of the analysis described in this document is to support post irradiation examination (PIE) future decisions as related to the Second Target Station (STS) Project. The STS Project is discussing plans for future PIE studies, and the radionuclide inventories and dose rates included in this study will aid decision makers as they consider the need for various PIE-related machinery. Three configurations of the lasagna-type target segment are evaluated for this analysis: a single bare tungsten plate, a slice of the segment containing a single tungsten plate, and the upstream half of a segment. The lasagna type target segment used in this analysis has a precipitation hardenable nickel alloy (UNS N07718 or inconel 718) shroud, copper (C10100) cladding, and pure tungsten plates. The target segment is evaluated for 10 years of operation followed by two decay periods: 6 months and 2 years. For each decay period, decay gamma dose rates are tallied 30 cm from the upstream, downstream, top, bottom, and sides of the configurations. Radionuclide inventories are also provided in this report separated by material. The highest dose rates calculated in this study are 2872 rem/hr 30 cm away from the top and bottom of the half-target configuration after 6 months of decay.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Chromosome aberrations among atomic-bomb survivors exposed in utero: updated analysis accounting for revised radiation doses and smoking

A previous study of peripheral blood lymphocyte translocations around age 40 among atomic-bomb survivors exposed in utero revealed no overall association with radiation dose—despite a clear association between translocations and dose among their mothers—but the data suggested an increase at doses below 100 mGy with a definite peak. That analysis of the in utero-exposed survivors did not adjust for their subsequent smoking behavior, an established cause of chromosomal aberrations, or their subsequent exposures to medical irradiation, a potential mediator. In addition, atomic-bomb survivor radiation dose estimates have subsequently been updated and refined. We therefore re-estimated the dose response using the latest DS02R1 dose estimates and adjusting for smoking as well as for city and proximal–distal location at the time of exposure to the atomic bomb. Sex of the survivor, mother’s age around the time of conception, and approximate trimester of gestation at the time of exposure were also considered as explanatory variables and modifiers. Precision of the estimated dose response was slightly lower due to greater variability near zero in the updated dose estimates, but there was little change in evidence of a low-dose increase and still no suggestion of an overall increase across the entire dose range. Adjustment for smoking behavior led to a decline in background number of translocations (the dose–response intercept), but smoking did not interact with dose overall (across the entire dose range). Adjustment for medical irradiation did not alter the association between dose and translocation frequency. Sex, mother’s age, and trimester were not associated with number of translocations, nor did they interact with dose overall. Interactions with dose in the low-dose range could not be evaluated because of numerical instability.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2020

Quality Assurance Plan for Federal Guidance Report 16

Federal Guidance Report No. 16 will provide risk coefficients for use in projecting the probability of cancer morbidity or mortality from exposure to environmental radionuclides. The derivation of nuclide-specific risk coefficients involves multiple steps, with each step relying on the accuracy of the previous step as well as the numerical data utilized within each step. Thus, to ensure the quality of the derived risk coefficients, it is necessary to evaluate and confirm the quality and proper implementation of input data at each step of the derivation. The Oak Ridge National Laboratory (ORNL) dosimetry research team currently has two largely independent software codes for deriving radiation dose estimates, dose coefficients, and cancer risk coefficients for internally deposited radionuclides: DCAL and QCAL. The agreement of derived values based on alternate codes is an important step in quality assurance (QA) of the risk coefficients produced in this project. DCAL is an integration of the dosimetric software and numerical databases developed at ORNL over the past 25 years for the EPA’s Office of Radiation and Indoor Air (ORIA). QCAL was developed in 2015 to serve as a substitute for DCAL while DCAL was undergoing revision and to provide QA for the updated version of DCAL. The complex methodology employed in the derivation of cancer risk coefficients presents a challenge for QA, and as such, this report describes the QA steps used by the ORNL dosimetry team in this process as implemented in DCAL and QCAL for Federal Guidance Report No. 16.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Evaluation of Attila and MCNP computational methods for dose and exposure estimation

Radiation transport calculations are often used to estimate dose or exposure to components and personnel surrounding a radiation source. The sources for these calculations are decaying radionuclides within various nuclear materials. Historically, dose calculations use MCNP (Monte Carlo N-Particle) transport code as the primary particle transport tool without a secondary computational tool to validate the results from the MCNP simulations [1]. The goal of this study is to make an independent check of the Monte Carlo solution from MCNP6 Version 6.2.1 with the discrete ordinates solution from Attila 10.2.0 Beta 3. As an example problem for this study, water-filled, stainless-steel vessels, modeled with an unstructured mesh (UM) with both MCNP and Attila [2], are exposed to 252Cf and 60Co point sources. This report also includes a discussion of the limitations of unstructured mesh in a MCNP calculation.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Evaluation of neutron dosimetry capabilities with the MC-15 portable multiplicity counter

This work proposes a preliminary neutron dose rate estimation method for a neutron multiplicity detector through measurement- and simulation-based analyses. Uncharacterized neutron-emitting sources may be encountered in situations such as nuclear emergency response, safeguards, and treaty verification. These circumstances may present irradiation risk to personnel conducting field assay, search, and characterization measurements. It is therefore of interest to provide a field neutron dosimetry capability with the existing neutron multiplicity counting (NMC) capabilities. To date, no commercially-available neutron detection systems are capable of both accurate NMC and real-time neutron dosimetry. This work will focus on estimating dose rate using input from a single fielded NMC called the MC-15. The energy-dependent neutron detection efficiency response of the MC-15 was quantified in monoenergetic neutron simulations and evaluated in response to two neutron-emitting sources and to a polyethylene-moderated source. The results were compared to existing neutron dosimeters and established the proof of concept for further investigation of the MC-15 for dose estimation. Measurement results were also replicated in simulations; additional simulations were then conducted to expand upon the limited empirical data. The initial empirical results provided a conversion factor appropriate for use when measuring 252 Cf neutrons that is independent of polyethylene shielding presence and thickness. The simulated data sets were then used to evaluate a fit equation allowing estimation of the neutron dose rate for less restricted geometric configurations and dependent only on the distance between the source and the detector. Additionally, the energy dependence of the efficiency response indicates that further empirical evaluations could provide energy-dependent conversion factors for broader neutron dosimetry capabilities with a wider range of neutron-emitting sources.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Reconstructing Hanford worker external doses from photons for epidemiology

The accurate reconstruction of external photon doses is essential for credible radiation epidemiology. This article presents the methodology used to derive dose estimates for 37 012 Hanford Site workers included in the Million Person Study. The approach employs historical dose records from the Hanford Radiation Exposure database and a previous epidemiology study. Bias correction factors specific to dosimeter type and period of use were applied and missing annual doses were estimated using a hierarchical nearby method to estimate deep dose equivalent for each worker. For early years with limited detection sensitivity, missed doses were quantified based on expected time-period-specific, low-dose statistical distributions. The revised dose estimates resulted in lower median and mean career doses than unadjusted data, while increasing the number of person-years with nonzero dose. Sensitivity analyses assessed the influence of bias in dosimetry measurements, missed doses and gap years on dose estimates. Differences in cumulative dose estimates between unadjusted and revised annual estimates are most prominent in the early operational years due to the highest bias during that time period.

dose reconstruction↗