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DOE OSTI2023

Reconstructed lung doses for the million person study cohort of 26,650 Tennessee Eastman corporation workers employed between 1942 and 1947

Abstract Tennessee Eastman Corporation workers were exposed to uranium dust resulting in high-linear energy transfer (LET) irradiation to lung tissue. In this work, radiation lung doses were reconstructed for 26 650 men and women working at the plant between 1942 and 1947. Site air monitoring data of uranium concentrations and payroll records were used to determine the daily inhaled activities and annualized lung doses. Variations in the activity median aerodynamic diameter of the uranium dust, the solubility of particulate matter in the lungs and the sex-specific breathing rate were investigated as part of a sensitivity analysis. Male and female mean lung doses of 18.9 and 32.7 mGy, respectively, from high-LET alpha irradiation, and there was general agreement with evaluations from previously published epidemiological studies. Annual lung dose estimates and sensitivity analysis for the 26 650 workers in the TEC cohort have been archived on the United States Department of Energy Comprehensive Epidemiologic Data Resource.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2023

An age- and sex-specific biokinetic model for radon *

Publication 137 of the International Commission on Radiological Protection (ICRP) describes a biokinetic model for radon used to derive dose coefficients for occupational intake of radon isotopes. The model depicts transfer of inhaled or ingested radon to blood, exchange of radon between blood and tissues, and gradual loss of radon from the body based on physical laws governing transfer of a non-reactive and soluble gas between materials. Here, this paper describes an age- and sex-specific variation of that model developed for use in an upcoming ICRP series of reports on environmental intake of radionuclides by members of the public titled ‘Dose Coefficients for Intakes of Radionuclides by Members of the Public’. The proposed model modifies the model structure and transfer coefficients presented in Publication 137 to allow more realistic dosimetric treatment of bone marrow and breast and expands the model to address pre-adult ages, based on the physical principles used in the development of the model of Publication 137 together with anatomical and physiological changes occurring during human development.

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DOE OSTI2023

Mathematical solutions in internal dose assessment: A comparison of Python-based differential equation solvers in biokinetic modeling

Abstract In biokinetic modeling systems employed for radiation protection, biological retention and excretion have been modeled as a series of discretized compartments representing the organs and tissues of the human body. Fractional retention and excretion in these organ and tissue systems have been mathematically governed by a series of coupled first-order ordinary differential equations (ODEs). The coupled ODE systems comprising the biokinetic models are usually stiff due to the severe difference between rapid and slow transfers between compartments. In this study, the capabilities of solving a complex coupled system of ODEs for biokinetic modeling were evaluated by comparing different Python programming language solvers and solving methods with the motivation of establishing a framework that enables multi-level analysis. The stability of the solvers was analyzed to select the best performers for solving the biokinetic problems. A Python-based linear algebraic method was also explored to examine how the numerical methods deviated from an analytical or semi-analytical method. Results demonstrated that customized implicit methods resulted in an enhanced stable solution for the inhaled 60 Co (Type M) and 131 I (Type F) exposure scenarios for the inhalation pathway of the International Commission on Radiological Protection (ICRP) Publication 130 Human Respiratory Tract Model (HRTM). The customized implementation of the Python-based implicit solvers resulted in approximately consistent solutions with the Python-based matrix exponential method ( expm ). The differences generally observed between the implicit solvers and expm are attributable to numerical precision and the order of numerical approximation of the numerical solvers. This study provides the first analysis of a list of Python ODE solvers and methods by comparing their usage for solving biokinetic models using the ICRP Publication 130 HRTM and provides a framework for the selection of the most appropriate ODE solvers and methods in Python language to implement for modeling the distribution of internal radioactivity.

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DOE OSTI2024

Transfer functions for Q A /Q B international regulatory limits for the safe transport of radioactive materials

This paper presents a proposed revision of the International Atomic Energy Agency transport regulations, related to the A 1 and A 2 limit values used to determine the radioactive transport classification. Based on the 'Q system', a novel methodology was introduced to derive Q A and Q B values related to scenarios involving external exposure from a distant source. These values are key parameters that respectively represent the total effective dose and total equivalent dose to the skin, from all primary and secondary particles contributing to radiation exposure. The International Working Group (WG A 1 /A 2 ) is established and associated with the TRANSSC Technical Expert Group on Radiation Protection. A review of the A 1 and A 2 values is performed in response to identified limitations within the existing Q system. The followed approach is based on Monte Carlo simulations that enabled the development of transfer functions aimed at reducing computational time and increasing the flexibility of dose evaluations for any radionuclide with known particle emission spectra. This method allows updating the Q A and Q B values to account for future data evolutions (decay data, fluence-to-dose conversion coefficients) and standardizing the calculation of regulation limits across all referenced radionuclides and scenarios related to external exposure. The transfer functions are established using three Monte Carlo simulation codes—FLUKA, Geant4, and MCNP—and address the previous limitations of the 'Q system', reflecting the latest International Commission for Radiation Protection recommendations and improvements in calculation techniques. The results of the WG show consistent agreement across the codes, with minor discrepancies observed at low primary energies due to statistical uncertainties and different handling of stopping power for electrons/positrons in the codes. This revised approach aligns with current standards and recommendations, ensuring that the radiological consequences of transport accidents are acceptable for the new A 1 and A 2 limits from a radiological protection perspective.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2024

National Council on Radiation Protection (NCRP) 2024 annual meeting: advanced and small modular nuclear power reactors

On 25–26 March 2023, the U.S. National Council on Radiation Protection and Measurements (NCRP) held its 2024 annual meeting in Bethesda, Maryland, USA. The NCRP dates from 1929, and this meeting celebrated the 60th anniversary of receiving a U.S. Congressional Charter. For this annual meeting the NCRP felt it was essential to provide a briefing about advanced and small modular nuclear reactors (SMRs). The Journal of Radiological Protection is delighted to publish the following synopsis of material presented at the U.S. NCRP meeting. This synopsis is divided into five sections. The first section provides an overview of the whole meeting together with summaries of two context setting overview papers. The following four sessions of this synopsis are specific to advanced and small modular nuclear power reactors. The meeting also included keynote presentations by three of NCRP annual award recipients. The meeting topical areas were Technology Overview and Critical Issues. The individual papers laid the groundwork to understanding reactor technologies, terminology, and the fundamental concepts and processes for electrical generation. The perspectives of the U.S. Environmental Protection Agency and states, through the Conference of Radiation Control Program Directors were provided. The papers included a discussion of diverse topics including potential emergency preparedness considerations, radiological survey requirements, an evaluation of the future of nuclear power, the economics of reactors (both large and small), and the critical issues identified by the recent National Academies of Sciences’ study on advanced reactors. Here, the summary papers were developed to briefly document the major points and concepts presented during the oral papers presented at the 2024 NCRP Annual Meeting. The meeting heralded the dawn of a new era for commercial nuclear power.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2026

Mortality among workers at the Rocky Flats Plant, 1951–2017

The Rocky Flats (RFs) Plant operated from 1951–1989 as part of the U.S. Department of Energy (DOE) nuclear complex. Its primary mission was weapons component fabrication, whereby workers were potentially exposed to radioactive and non-radioactive hazards. RF worker mortality was compared to the general population, and dose-response relationships between mortality and radiation organ doses were examined. RF workers first employed between 1951 and 1979 for ⩾30 d were identified (n = 9397). Vital status was determined using national and state death records up to 2017. Organ doses from external photons and neutrons irritation and internalised plutonium (Pu), americium (Am), and uranium (U) were modelled as cumulative lagged total doses per year. Beryllium exposure was evaluated as an effect modifier using data from the DOE Nationwide Beryllium Medical Program. Statistical analyses included standardised mortality ratios (SMRs), Cox proportional hazard models, and excess relative risk (ERR) models. Approximately 53.2% of workers were deceased by the end of the study. Nearly 90% were monitored for radiation exposure, with a mean weighted absorbed dose of 59.0 mGy for the lungs. Nearly 45% of workers had intakes of alpha-particle emitting radionuclides, and 46.7% were monitored for neutrons. Leading causes of death included ischemic heart disease (n = 999) and lung cancer (n = 361). The highest SMRs were observed for berylliosis (SMR: 176.9; 95% CI: 76.2, 348.7; n < 10) and asbestosis (SMR: 4.65; 95% CI: 2.23, 8.55; n = 10). Dose-response analyses showed no statistical increase in risk from low-dose radiation including lung cancer (ERR per 100 mGy: −0.02; 95% CI: −0.11, 0.08; n = 361) and Parkinson’s disease (ERR per 100 mGy: 0.13; 95% CI: −0.26, 0.31; n = 57). Approximately 45% of workers were monitored for beryllium, with a weak non-significant indication of effect modification for lung cancer risk. The RF cohort showed no evidence of a statistically significant increase in mortality from occupational radiation exposure. However, this study was limited by low statistical power, which inhibits the ability to detect effects. Future pooling of Million Person Study (MPS) cohorts will provide further insights, particularly regarding Pu as a carcinogen.

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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 OSTI2020

Coefficients for Estimating Prenatal Dose in Pregnant Workers from Acute Intakes

Inhalation and ingestion dose coefficients for the embryo and fetus from intakes of radionuclides by the mother are provided in the International Commission on Radiological Protection (ICRP) Publication 88 for intake of each of 74 radionuclides. To address the many other possible radionuclides to which workers may be exposed, effective dose coefficients were developed for the embryo/fetus for all additional radionuclides addressed in ICRP Publication 107 with half-life of 10 min or more. The general approach was to use the estimated dose to the mother’s uterus during pregnancy as a scalable proxy for the dose to the embryo/fetus. The set of scaling factors used in the study was derived from analyses of the relationships of the dose to the mother’s uterus and the effective dose to the embryo/fetus for the ~400 cases (considering two intake modes and multiple forms of many of the radionuclides) addressed in Publication 88.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2021

Improved age- and gender-specific radiation risk models applied on cohorts of Swedish patients

The aim of this study is to implement lifetime attributable risk (LAR) predictions for radiation induced cancers for Swedish cohorts of patients of various age and sex, undergoing diagnostic investigations by nuclear medicine methods. Methods: Calculations are performed on Swedish groups of patients with Paget's disease and with bone metastases from prostatic cancer and diagnosed with bone scintigraphy with an administration of 500 MBq 99m Tc-phosphonate. Results: The inclusion of patient survival rates into the calculations lowers the induced radiation cancer risk, as it takes into account that cohorts of patients have shorter predicted survival times than the general population. Conclusion: LAR estimations could be valuable for referring physicians, nuclear medicine physicians, nurses, medical physicists, radiologists, and oncologists and as well as ethical committees for risk estimates for specific subgroups of patients. Caution is however advised with respect to application of LAR predictions to individuals (because of individual sensitivities, circumstances, etc.).

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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.

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DOE OSTI2020

Uncertainty Analysis of Consequence Management Data Products

An interlaboratory effort has developed a probabilistic framework to characterize uncertainty in data products that are developed by the US Department of Energy Consequence Management Program in support of the Federal Radiological Monitoring and Assessment Center. The purpose of this paper is to provide an overview of the probability distributions of input variables and the statistical methods used to propagate and quantify the overall uncertainty of the derived response levels that are used as contours on data products due to the uncertainty in input parameters. Uncertainty analysis results are also presented for several study scenarios. This paper includes an example data product to illustrate the potential real-world implications of incorporating uncertainty analysis results into data products that inform protective action decisions. Data product contours that indicate areas where public protection actions may be warranted can be customized to an acceptable level of uncertainty. Finally, the investigators seek feedback from decision makers and the radiological emergency response community to determine how uncertainty information can be used to support the protective action decision-making process and how it can be presented on data products.

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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 OSTI2020

Compromise of Personal Protective Clothing from Liquid Exposure

Following critiques of multiple personal contamination events from entries into the Oak Ridge National Laboratory’s Spallation Neutron Source transfer bay, it was considered that the most likely causes for contamination were personal protective clothing doffing errors or that moisture (sweat) allowed contamination to wick through the protective clothing. To pursue this issue, radiological protection staff looked more closely, however, at the specific area of the clothing where contamination was highest; under enhanced lighting and photochromic manipulation, there appeared to have been some type of moisture in the area. Recognizing the possibility that moisture may have allowed for migration of contamination through the clothing, further experiments were undertaken to determine under which conditions this transport might have occurred. The objective for this work was to identify the susceptibility of different types of personal protective clothing to various liquids encountered in the workplace.Method: Several tests were performed to determine if perspiration had enabled migration of contamination, and to identify what other liquids might have affected contamination transport. Two layers of personal protective clothing were subjected to static conditions and dynamic conditions, to include active rubbing of the materials while wet. Food dye added to each of the liquids tested enabled visual indications of liquid breakthrough. Additional tests were conducted to see if solid contamination could be transported through the materials along with the liquids. As a result, it was found that all but one type of non-rubberized personal protective clothing in use at Oak Ridge National Laboratory were permanently compromised to some extent by the solvents used for decontamination. It was determined that most common cleaning agents immediately and permanently destroyed the hydrophobic nature of several of the splash-resistant protective clothing materials, allowing for radioactive contamination to penetrate through the material to the worker. Work around wet surfaces or performing wet decontamination will only be performed in protective clothing known to prevent transport of the wetting agent.

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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.

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DOE OSTI2022

Radiological HEPA Filter 10-year Lifetime Evaluation in Research Facilities

High-efficiency particulate air (HEPA) filters are widely employed by nuclear facilities to remove radiological particulate matter from their effluent exhaust streams. The purpose of this study is to evaluate the relationships between the 10-year HEPA filter lifetime deployment and its other performance indicators. This 10-year-long endeavor to collect and analyze data regarding the service life of HEPA filters at the Pacific Northwest National Laboratory began in 2010. A set of HEPA filters were selected and have been surveyed and analyzed at least annually to verify compliance with permit conditions. The study suggests the frequency of filter replacement should be based on the actual operational requirements, such as fume hood face velocity and/or efficiency test results, instead of on the prescribed filter “age limit” of 10 years from the date of manufacture (e.g., birth date) when operating under dry conditions. The study has now been completed, and over the past decade all the HEPA filters have been replaced, due to either technical issues as listed in this report or the previously recommended filter “age limit” of 10 years as prescribed by the oversight bodies. Experimentally determined failure rates are also determined from the data set and can be used to estimate the chances of HEPA filters surviving 15, 20, or even 30 years.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2022

Radon-222 Charcoal Canister Steady State Model Calibrations Performed in a Highly Controlled Environmental Chamber and a Natural Indoor Environment

Charcoal canisters are a common method of 222 Rn screening. The calibrations of different batches of activated charcoal used in different canister designs are typically performed in large volume controlled environmental chambers with known and controlled radon concentration, temperature, and humidity. Radon screening could be facilitated in locations without ready access to environmental chambers if canisters could be accurately calibrated and undergo quality control in less controlled environments. This study compares charcoal canister calibrations from a highly controlled radon chamber with calibrations from a basement storage area experiencing temporally varying radon. In addition, the impacts of exposure time and spectral region of interest selection on calibration accuracy are investigated by comparing calibrations calculated using three different choices. Approximately 30 mo after calibrations were completed, groups of canisters were exposed for different durations of time to investigate measurement accuracy and calibration validity over an extended time. A digital simulation of charcoal canister kinetics was also performed to establish limits on how stable radon must be in a space for equilibrium-based calibration to be performed there. Overall, the accuracy of measurements using calibrations from each space differed by less than 10% after 2 d exposure time, showing that carefully controlled conditions are not necessary for the accurate calibration of charcoal canisters. Furthermore, measurement accuracy differed by less than 2% for different spectral region selections. Accuracy improved slightly with canister exposure duration. Simulations suggest that radon instability is most tolerable near the beginning of canister calibration exposures, but this merits further experimental study.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2022

Alternate Analysis of an In Vitro Solubility Study on the Lung Dissolution Rate of 238PuO2 Material Involved in the 2020 Incident at Los Alamos National Laboratory

Here this work presents an alternate analysis of an in vitro solubility study on the lung dissolution rate of 238 PuO 2 material involved in a recent inhalation incident at Los Alamos National Laboratory (LANL). The original dataset used in this work was retrieved from a recently published report. The present work shows an analysis of the same dataset by modeling the dissolution in separate time intervals rather than modeling the cumulative dissolution.

238Pu↗

DOE OSTI2023

Methods to Account for CAP-88 PC-Omitted Nuclides in Radioactive Air Emissions From DOE Facilities

Routine research and development activities at US Department of Energy facilities can result in the release of radioactive emissions, potentially exposing the public and the environment. Such emissions are subject to certain Clean Air Act regulations, specifically those outlined in 40 CFR Part 61, Subpart H, which are enforced by the US Environmental Protection Agency. Compliance is determined in part with a dose standard. Doses to members of the public from most US Department of Energy facility airborne emissions are modeled using CAP-88 PC software, the latest version of which contains more than 1,200 radionuclides. Radioactive effluents from US Department of Energy facilities may contain radioisotopes that are not available in CAP-88 PC, and those radioisotopes must still be evaluated to determine whether doses to members of the public are below established limits. CAP-88 PC-omitted radionuclides can be accounted for using alternate methods, including the use of surrogate radionuclides. Here, this paper elucidates the process of accounting for CAP-88 PC-omitted radionuclides by providing criteria and rationale for surrogate selection, a compilation of surrogate radionuclides used in the past by US Department of Energy facilities, a discussion of alternate methods used to account for CAP-88 PC-omitted radionuclides and a comparison of methods and impacts on receptor doses. Ultimately, this paper aims to aid in the process of surrogate selection and consequently to simplify and expedite compliance with Clean Air Act regulations.

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