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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Niche Partitioning of Microbial Communities at an Ancient Vitrified Hillfort: Implications for Vitrified Radioactive Waste Disposal

A pre-Viking era vitrified hillfort, Broborg, provides a habitat analogue for disposed radioactive waste glass and shows strong niche partitioning among the organisms involved in glass alteration. Microbes cannot be eliminated from radioactive waste disposal facilities and the consequences of bio-colonization must be understood. We use Broborg as a model system to inform what microbial processes might influence long-term radioactive waste glass durability by examining anthropogenic glass that has been subjected to bio-colonization for over 1,500 years. Scanning electron microscopy (SEM) images reveal the surficial biofilm structure, and chemical/mineralogy analysis in combination with deoxyribonucleic acid (DNA) sequencing of samples from the vitrified substrate, the adjacent soil, and the general topsoil provide insight into niche partitioning. The ancient glass niche supports a unique microbial community of bacteria, fungi, and protists that manifests the species response to local geochemical and mineralogical conditions. Communities from the geochemical niche associated with the glass are distinct and less diverse than soil communities. The microbiome of the glass and adjacent soil are dominated by lichens, lichen-associated microorganisms, and other epilithic, endolithic, and epigeic organisms. Pseudomonads dominate the prokaryotic communities on the vitrified material, but not the adjacent soil. In contrast, the general topsoil communities are enriched in plant rhizosphere organisms. Taxa associated with vitrification have bio-corrosive properties that could be detrimental to glass durability, including silicate mineral dissolution, extraction of essential elements, secretion of geochemically reactive organic acids, and dissolution induced by improved water retention. However, these stable long-term biofilms also possess a homeostatic function that could limit glass alteration. Overall, this study demonstrates the potential impacts that microbial colonization and niche partitioning can have on glass alteration and subsequent release of radionuclides from a disposal facility for vitrified radioactive waste.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2021

Radium Dial Workers: Back to the Future

This report reviews the history of the radium dial workers in the United States, summarizes the scientific progress made since the last evaluation in the early 1990s, and discusses current progress in updating the epidemiologic cohort and applying new dosimetric models for radiation risk assessment. The discoveries of radiation and radioactivity led quickly to medical and commercial applications at the turn of the 20th century, including the development of radioluminescent paint, made by combining radium with phosphorescent material and adhesive. Workers involved with the painting of dials and instruments included painters, handlers, ancillary workers, and chemists who fabricated the paint. Dial painters were primarily women and, prior to the mid to late 1920s, would use their lips to give the brush a fine point, resulting in high intakes of radium. The tragic experience of the dial painters had a significant impact on industrial safety standards, including protection measures taken during the Manhattan Project. The dial workers study has formed the basis for radiation protection standards for intakes of radionuclides by workers and the public. The mortality experience of 3,276 radium dial painters and handlers employed between 1913-1949 is being determined through 2019. The last epidemiologic follow-up was 30 years ago when most of these workers were still alive. Nearly 65% were born before 1920, 37.5% were teenagers when first hired, and nearly 50% were hired before 1930 when the habit of placing brushes in mouths essentially stopped. Comprehensive dose reconstruction techniques are being applied to estimate organ doses for each worker related to the intake of 226Ra, 228Ra, and associated photon exposures. Time dependent dose-response analyses will estimate lifetime risks for specific causes of death. The study of radium dial workers is part of the Million Person Study of low-dose health effects that is designed to evaluate radiation risks among healthy American workers and veterans. Despite being one of the most important and influential radiation effects studies ever conducted, shifting programmatic responsibilities and declining funding led to the termination of the radium program of studies in the early 1990s. Renewed interest and opportunity have arisen. With scientific progress made in dosimetric methodology and models, the ability to perform a study over the entire life span, and the potential applicability to other scenarios such as medicine, environmental contamination and space exploration, the radium dial workers have once again come to the forefront.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Deeper snow increases the net soil organic carbon accrual rate in moist acidic tussock tundra: 210Pb evidence from Arctic Alaska

The net change in the carbon inventory of arctic tundra remains uncertain as global warming leads to shifts in arctic water and carbon cycles. To better understand the response of arctic tundra carbon to changes in winter precipitation amount, we investigated soil depth profiles of carbon concentration and radionuclide activities ( 7 Be, 137 Cs, 210 Pb, and 241 Am) in the active layer of a twenty-two-year winter snow depth manipulation experiment in moist acidic tussock tundra at Toolik Lake, Alaska. Depth correlations of cumulative carbon dry mass (g cm -2 ) vs. unsupported 210 Pb activity (mBq g -1 ) were examined using a modified constant rate of supply (CRS) model. Results were best fit by two-slope CRS models indicating an apparent step temporal increase in the accumulation rate of soil organic carbon. Most of the best-fit model chronologies indicated that the increase in carbon accumulation rate apparently began and persisted after snow fence construction in 1994. The inhomogeneous nature of permafrost soils and their relatively low net carbon accumulation rates make it challenging to establish robust chronologic records. Nonetheless, the data obtained in this study support a decadal-scale increase in net soil organic carbon accumulation rate in the active layer of arctic moist acidic tussock tundra under conditions of increased winter precipitation.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

A template of information needs for decision-making about delaying remediation on contaminated lands to protect human health

The contamination legacy of industrialization, militarization, and nuclear arms race poses current or future risks to populations and the environment. Responsible parties and regulators make decisions regarding which sites to clean up, how, how much, and when. This study aimed to provide an information needs template to evaluate and reduce risks to human health when considering whether to initiate or delay remediation. This investigation focused on four aspects of timing and prioritization: 1) management, planning and implementation, 2) source terms, pathways, and exposures, 3) risks and receptors, and 4) external drivers. Within each type, issues were identified and described. Management class included personnel, health and safety data, funding, equipment, and structural integrity. Source term included contaminant sources, pathways, initiating events, and barriers to exposure. Risk included types and exposures to workers and general public. External drivers included regulatory framework, stakeholders, Congressional mandates, and economic and social contexts. Risk may increase over time as contamination spreads, enters aquifers, and reaches receptors, and may decline as radionuclides decay, and plumes dissipate. The overall objective was to provide a template of information that is useful to managers and regulators, and might be used by the public to understand the risks and benefits of re-prioritization cleanup.

Environmental Sciences & Ecology↗
DOE OSTI2023

Comprehensive Analysis of Streaming and Shutdown Dose Rate Experiments at JET with ORNL Fusion Neutronics Workflows

Current experimental fusion systems and conceptual designs of fusion pilot plants (FPPs) are growing in complexity and size. Several radiation metrics are crucial to the safe operation of fusion machines, including neutron flux streaming through openings and the shutdown dose rate (SDDR). Most current designs of advanced experimental fusion systems—and the most probable candidates for FPPs—are based on the tokamak concept, which is prone to neutron streaming through the myriad openings needed for diagnostic and support systems. SDDR is caused by decay gamma rays from radionuclides that become activated by neutrons during the operation of a fusion system that use deuterium-deuterium (DD), tritium-tritium, or deuterium-tritium plasma. Because computational tools have become essential for determining these radiation metrics, they must be validated against reliable and applicable experimental data. Experiments at the Joint European Torus (JET) provide a unique source of experimental data for validating computational tools and nuclear data used to determine SDDR and neutron fluxes in streaming-dominated geometries. Here, this paper presents the comprehensive analysis of the high-performance DD JET SDDR, and streaming experiments performed using Oak Ridge National Laboratory (ORNL) fusion workflows. The computational results were compared with experimental results that consist of online SDDR measurements with ionization chambers and neutron fluence streaming measurements using thermoluminescent detectors. The ratio of calculated-to-experimental SDDR values ranges from 0.6 to 2.5, and the streaming results range from 0.5 to 8.0. Future work will include analyzing the JET 2021 DTE2 campaign alongside the integration of the Shift Monte Carlo transport code into all ORNL fusion neutronics workflows.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Implementation of the D1S Methodology for Shutdown Dose Rate Calculations in the OpenMC Monte Carlo Particle Transport Code

We present an implementation of the direct one-step (D1S) methodology for shutdown dose rate (SDR) calculations in the OpenMC Monte Carlo particle transport code. In addition to being the first fully open-source D1S implementation, it is also the first to require no ad hoc source code or nuclear data library modifications. The code can seamlessly switch between production of prompt and decay photons based on a user input parameter, and the decay data needed for decay photon generation are made available through a depletion chain file, which is already used for OpenMC’s built-in depletion/activation solver. A set of Python functions significantly eases the burden of computing and applying time correction factors needed to properly account for the time dependence of radionuclide activity. To assess the accuracy of the D1S implementation, SDR calculations have been carried out for three problems: a prism of iron irradiated by 14-MeV neutrons, the ITER port plug computational benchmark, and the Frascati Neutron Generator (FNG) ITER dose rate benchmark problem from the Shielding INtegral Benchmark Archive and Database (SINBAD). For each of these problems, comparisons were made to calculations using the rigorous two-step (R2S) method. The results on the iron prism problem illustrate how the D1S method achieves superior spatial resolution compared to the R2S method without the need for spatial discretization of the activation regions. The D1S and R2S results for the ITER port plug benchmark agree well with previously reported results in the literature. While the D1S results are 10% to 15% lower than the R2S results, this may be due to stochastic uncertainty and/or spatial discretization in the R2S calculations. On the FNG dose rate benchmark problem, the D1S method produces dose rate estimates that are within 4% of the dose rates predicted using a cell-based R2S workflow. The D1S estimates of the SDR are also in reasonable agreement with the experimental measurements and show the same basic trends that have been observed in previous works. A qualitative analysis of the execution time and uncertainty for the R2S and D1S workflows suggests that the D1S method would attain a higher figure of merit.

D1S method↗
DOE OSTI2020

Biokinetic models for group IVB elements

The International Commission on Radiological Protection (ICRP) is updating its biokinetic models in a series of reports titled Occupational Intakes of Radionuclides (OIR series). This paper provides an overview of biokinetic data for the group IVB elements hafnium (Hf) and titanium (Ti), compares these data with findings for the more extensively studied Group IVB element zirconium (Zr), and proposes biokinetic models for systemic Hf and Ti for use in the OIR series. A biokinetic model for systemic Zr recently adopted by the ICRP is proposed for application to Hf in view of the nearly identical chemical and physical properties of these two elements, their closely similar behavior in the environment, and their nearly identical biokinetic properties suggested by available comparative data. Finaly, the model structure applied to Zr and Hf is also applied to Ti, but a separate set of transfer coefficients is proposed for Ti.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Four-decade follow-up of a plutonium-contaminated puncture wound treated with Ca-DTPA

Contaminated wounds are a common route of internal deposition of radionuclides for nuclear and radiation workers. They may result in significant doses to radiosensitive organs and tissues in an exposed individual's body. The United States Transuranium and Uranium Registries' whole-body donor (Case 0303) accidentally punctured his finger on equipment contaminated with plutonium nitrate. The wound was surgically excised and medically treated with intravenous injections of Ca-DTPA. A total of 16 g Ca-DTPA was administered in 18 treatments during the 2 months following the accident. Ninety-three urine samples were collected and analysed over 14 years following the accident. An estimated 239 Pu activity of 73.7 Bq was excreted during Ca-DTPA treatment. Post-mortem radiochemical analysis of autopsy tissues indicated that 40 years post-accident 21.6 ± 0.2 Bq of 239 Pu was retained in the skeleton, 12.2 ± 0.3 Bq in the liver, and 3.7 ± 0.1 Bq in other soft tissues; 1.35 ± 0.02 Bq of 239 Pu was measured in tissue samples from the wound site. To estimate the plutonium intake, late urine measurements, which were unaffected by chelation, and post-mortem radiochemical analysis results were evaluated using the IMBA Professional Plus software. The application of the National Council on Radiation Protection and Measurements wound model with an assumption of intake material as a predominantly strongly retained soluble plutonium compound with a small insoluble fraction adequately described the data (p = 0.46). The effective intake was estimated to be 50.2 Bq of plutonium nitrate and 1.5 Bq of the fragment. Here, the prompt medical intervention with contaminated tissue excision and subsequent Ca-DTPA decorporation therapy reduced 239 Pu activity available for uptake and long-term retention in this individual's systemic organs by a factor of 38.

239Pu↗
DOE OSTI2021

A biokinetic model for systemic sodium

This paper describes an updated biokinetic model for systemic sodium (Na), developed for use in a series of reports by the International Commission on Radiological Protection (ICRP) on occupational intake of radionuclides. In contrast to the ICRP's previous model for intake of radio-sodium by workers, the updated model depicts realistic directions of movement of Na in the body including recycling of activity between blood and tissues. The updated model structure facilitates extension of the baseline transfer coefficients for adults to different age groups and to special exposure scenarios such as transfer of radio-sodium from the mother to the foetus or the nursing infant. Dose coefficients for 22Na and 24Na based on the updated model generally do not differ greatly from those based on the ICRP's previous Na model when both models are connected to the ICRP's latest dosimetry system. The main exception is that the updated model yields roughly twofold higher dose coefficients for endosteal bone surface than does the previous model due to the dosimetrically cautious assumption in the updated model that exchangeable Na in bone resides on bone surface.

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.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

The Auger Radioisotope Microscope: an instrument for characterization of Auger electron multiplicities and energy distributions

We describe a new instrument, the Argonne Auger Radioisotope Microscope (ARM), capable of characterizing the Auger electron (AE) emission of radionuclides, including candidates relevant in nuclear medicine. Our approach relies on event-by-event ion–electron coincidence, time-of-flight, and spatial readout measurement to determine correlated electron multiplicity and energy distributions of Auger decays. We present a proof-of-principle measurement with the ARM using x-ray photoionization of stable krypton beyond the K-edge and identify a bifurcation in the electron multiplicity distribution depending on the emission of K-LX electrons. Extension of the ARM to the characterization of radioactive sources of AE emissions is enabled by the combination of two recent developments: (1) cryogenic buffer gas beam technology to introduce Auger emitters into the detection region with well-defined initial conditions, and (2) large-area micro-channel plate detectors with multi-hit detection capabilities to simultaneously detect multiple electrons emitted in a single decay.

Auger emitters↗
DOE OSTI2025

Successful νp -process in neutrino-driven outflows in core-collapse supernovae

The origin of the solar system abundances of several proton-rich isotopes, especially 92,94 Mo and 96,98 Ru, has been an enduring mystery in nuclear astrophysics. An attractive proposal to solve this problem is the νp-process, which can operate in neutrino-driven outflows in a core-collapse supernova after the shock is launched. Years of detailed studies, however, have cast doubt over the ability of this process to generate sufficiently high absolute and relative amounts of various p-nuclei. The νp-process is also thought to be excluded by arguments based on the long-lived radionuclide 92 Nb. Here, we present explicit calculations, in which both the abundance ratios and the absolute yields of the p-nuclei up to A ≲ 105 are successfully reproduced, even when using the modern (medium enhanced) triple-α reaction rates. The process is also shown to produce the necessary amounts of 92 Nb. The models are characterized by subsonic outflows and by the protoneutron star masses in the ≳ 1.7 M ⊙ range. In conclusion, this suggests that the Mo and Ru p-nuclides observed in the Solar System were made in CCSN explosions characterized by an extended accretion stage.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2025

Metallothionein expression in Eastern mosquitofish ( Gambusia holbrooki ) exposed to 137 Cs and lead: implications for using metallothionein as a metal exposure biomarker

Given the range of threats facing aquatic ecosystems, it is critical we have monitoring tools to quickly and efficiently identify stressors within ecosystems. Metallothionein’s (MTs) are a group of proteins widely used as biomarkers of metal exposure; however, their suitability has been questioned due to their influenceable nature and ability to be induced by various stressors. Through our study, we sought to (1) evaluate induction of MT in eastern mosquitofish (Gambusia holbrooki) exposed to a radionuclide ( 137 Cs) and metal (Pb), (2) assess the influence of in situ 137 Cs pre-exposure on Pb-induced MT expression, and (3) quantify changes in MT concentrations during Pb depuration. We conducted our experiment in three phases using mosquitofish as a model species: Phase I—pre-exposure to 137 Cs in the field, Phase II—exposure to Pb in the lab immediately following 137 Cs exposure, and Phase III—depuration in the laboratory. We quantified MT concentrations over time and assessed differences among treatments. We did not find in situ exposure of mosquitofish to 137 Cs to induce MT, nor did pre-exposure to 137 Cs appear to influence induction of MT following Pb exposure. Metallothionein was rapidly induced in mosquitofish following Pb exposure; however, MT induction was also observed in control fish. Additionally, there was no apparent relationship between MT and Pb body burdens. During depuration, we found MT concentrations to be highly variable among fish demonstrating no clear response. Furthermore, our study demonstrates the high variability exhibited by MT, the need for further validation of MT as a metal exposure biomarker, and the value in additional research on these proteins under varying exposure scenarios, particularly environmentally realistic concentrations.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

URINARY EXCRETION OF PLUTONIUM IN MAYAK WORKERS DURING AND AFTER CA-DTPA ADMINISTRATION

Chelation therapy is sometimes used after potential exposures to plutonium to increase urinary excretion of the radionuclide to improve the accuracy of bioassay measurements. The purpose of this report is to describe the enhancement of urinary excretion of plutonium during and after the administration of the trisodium salt of calcium diethylenetriaminepentaacetate (Ca-DTPA) daily for 3 d to a group of male and female plutonium workers from the Mayak Production Association in Ozyorsk, Russia. One-hundred and two cases (18 females and 84 males) were selected where urinary contents of plutonium, prior to chelation, exceeded the detection threshold. Daily urine collections were obtained during the 3 d of Ca-DTPA treatments. In addition, 58 of these cases had urine bioassays at 1–45 d after chelation. The daily enhancement over baseline values excretion of plutonium was found to be 50.4×/1.4 (geometric mean and geometric standard deviation); 58.9×/1.2; 72.9×/1.4 in the first, second and third days of Ca-DTPA administration. The mean enhancement for the 3-d period was 60.1×/1.7. The rate of plutonium excretion from 1 to 45 d after chelation decreased with a half-period of 3.9 d and the chelation enhancement factor (Кenh-i) is described by the function Кenh-i = (0.79 ± 0.24) + (42.9 ± 1.2) × e-(0.18 ± 0.01) × day.

Environmental Sciences & Ecology↗
DOE OSTI2022

Dosimetry and cancer risk estimations for different radiation protection solutions at decommissioning a contaminated nuclear power plant site

Contaminated sediments originating from dredging activities in a nuclear power plant site were placed in a pond, which has to be taken into consideration during the future decommissioning process. The sediments have to be handled to free release the site. The radionuclides Co-60 and Cs-137 were identified and the activity concentrations (Bq/kg) were quantified in the range of 10–6000 and 5–50 Bq kg -1 , respectively. The absorbed dose rate to individuals of various ages and sex present at the site of the dry pond area was estimated. The radiological impact in terms of lifetime attributable risk (LAR) and effective dose were calculated. For a 30-year-old male exposed during one year without any action regarding the sediments in the dried out pond, the LAR was predicted to be 0.0027, which recalculated to effective dose corresponds to 7.6 mSv year -1 . The calculations show that countermeasures will be needed for the contaminated site.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Design and Characterization of an Extremely-Sensitive, Large-Volume Gamma-Ray Spectrometer for Environmental Samples

A large volume gamma spectrometer was designed and constructed to analyze foodstuffs and environmental samples having low radionuclide concentrations. This system uses eight 11-cm × 42.5-cm × 5.5-cm NaI(Tl) detectors, chosen due to their relatively high sensitivity and availability and arranged in an octagonal configuration. The sensitive volume of the system is ~28 cm in diameter and ~42 cm deep. Shielding consists of an 86-cm × 86-cm square, 64-cm-tall lead brick enclosure with 18-cm-thick lead walls lined by 0.3-cm-thick copper plates. An aluminum top was machined to suspend the detectors within this shield. The shielding reduces background counts by 72% at 100 keV and 42% at 1,000 keV. The positional variability in sensitivity of the well was determined by both simulation and experiment. A 2.1-L volume of nearly uniform sensitivity, varying less than 10%, exists in the well's center. Energy resolutions of 14.6% and 7.8% were measured for 241 Am and 137 Cs, respectively. Energy resolution shows a 0.2% variation for both 241 Am and 137 Cs as a function of position within all regions of the well’s central sensitive volume. Dead time was also determined to be less than 35% for all sources measured in the system, the largest of which had an activity of 1,760 kBq. Simulated results for various source geometries show higher counts for smaller samples, especially at lower energies due to less attenuation of low energy photons. In conclusion, minimum detectable activities were determined for all source energies used, less than 5.1 Bq kg -1 for reasonable background and sample counting times.

47 OTHER INSTRUMENTATION↗
DOE OSTI2020

Internal Radiation Dose Evaluation For An Unruptured Post Release Tristructural Istropic Fuel Particle for Advanced and Micro-reactor Applications

There are unique benefits from advanced/micro reactor designs and fuel types that offer safety features in the case of an accident that may reduce environmental consequences compared to conventional reactors and fuels. Several micro reactor designs, for both civilian and military applications, are proposed using tristructural-isotropic (TRISO) fuel. TRISO particles are a robust advanced nuclear fuel type that leads to the unique question of how unruptured, activated TRISO particles will interact with humans. Human exposure to nuclear fuel, or radionuclides, includes all pathways of exposure. However, TRISO particles are 0.9 mm in size, and that particle size restricts internal dose assessment to the ingestion pathway; eliminating the inhalation and dermal absorption pathways. Furthermore, activity of the TRISO particle was established by High Temperature Engineering Test Reactor simulations assuming the TRISO particle was one day post-release from the reactor. International Commission of Radiological Protection publications 100 and 133 were used to develop the internal dosimetry model for an unruptured TRISO particle for the International Commission of Radiological Protection reference male and female dose assessment. The rectosigmoid colon had the highest internal organ dose for both reference male (2.1 Sv) and reference female (2.3 Sv). The internal dose from ingestion of the scenario-defined TRISO particle was 0.25 Sv for the reference male and 0.29 Sv for the refence female, which exceeds the annual occupational effective dose limit of 0.05 Sv in the code of federal regulations, 10.CFR.20 Subpart C.

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↗