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DOE OSTI2013

Radiological Toolbox User’s Guide

The Radiological Toolbox software developed by Oak Ridge National Laboratory (ORNL) for U. S. Nuclear Regulatory Commission (NRC) is designed to provide electronic access to the vast and varied data that underlies the field of radiation protection. These data represent physical, chemical, anatomical, physiological, and mathematical parameters detailed in various handbooks which a health physicist might consult while in his office. The initial motivation for the software was to serve the needs of the health physicist away from his office and without access to his handbooks; e.g., NRC inspectors. The earlier releases of the software were widely used and accepted around the world by not only practicing health physicist but also those within educational programs. This release updates the software to accommodate changes in Windows operating systems and, in some aspects, radiation protection. This release has been tested on Windows 7 and 8 and on 32- and 64-bit machines. The nuclear decay data has been updated and thermal neutron capture cross sections and cancer risk coefficients have been included. This document and the software’s user’s guide provide further details and documentation of the information captured within the Radiological Toolbox.

59 BASIC BIOLOGICAL SCIENCES↗

DOE OSTI2018

Gamma and Beta Absorbed Dose Conversion Coefficients in the Range from 10 keV to 10 MeV for Accidental Exposures From Point Sources Placed in Clothing in Proximity to the Body [Gamma and Beta Absorbed Dose Conversion Coefficients in the Range from 10 keV to 10 MeV for Accidental Exposures From Point Sources Placed in Clothing in Close Vicinity to the Body]

Here, retrospective dose assessment following acute radiation exposures during radiological incidents can be difficult and inaccurate due to the large uncertainties associated with dose estimation. However, rapid and accurate dose assessment is critical following an incident, such that appropriate treatment can be provided to the patient as early as possible. Incident Dose assessment relies heavily on biological dosimetry with corresponding large uncertainties for inhomogeneous exposures, resulting from the estimates of whole-body doses, while the assessment of absorbed doses to individual tissues might actually be more appropriate for acute radiation exposures.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2025

2025 Review and Revision of Federal Guidance Report 15

Federal Guidance Report No. 15 (FGR 15), External Exposure to Radionuclides in Air, Water and Soil, published in 2019 and referred to below as FGR 2019, provides age-specific effective dose rate coefficients for reference persons externally exposed to each of 1252 radionuclides homogenously distributed in environmental media. Soon after completion of FGR 2019, the International Commission on Radiological Protection (ICRP) published a similar report (ICRP Publication 144, 2020) addressing the same radionuclides and many of the external exposure scenarios addressed in FGR 2019. In 2023 Argonne National Laboratory (ANL) published a review of dose coefficients in FGR 2019, concluding that “The external effective dose from beta radiation is not appropriately accounted for in FGR 15 for both low- and high-energy beta emitters in air, in soil, and on soil surfaces.” That conclusion was based largely on comparisons of dose coefficients in FGR 2019 with values in ICRP Publication 144 but also on consideration of some unexpected patterns of equivalent dose rate coefficients across tissues and of effective dose rate coefficients across different soil depths, for a selected set of low-energy beta emitters. In response to the ANL report, the Center for Radiation Protection Knowledge (CRPK) at Oak Ridge National Laboratory (ORNL) performed an extensive reexamination of the methods and published values of FGR 2019. CRPK found that coding errors had resulted in inaccurate estimates, primarily overestimates, in many of the dose coefficients tabulated in FGR 2019 but that many of the differences between results in FGR 2019 and ICRP Publication 144 could be traced to differences in methodology. CRPK has corrected all errors in FGR 2019 associated with the coding errors and has taken the opportunity to improve a major portion of the remaining dose coefficients in FGR 2019, primarily through additional Monte Carlo calculations resulting in improved statistics for tissue equivalent dose rate coefficients for exposure to monoenergetic sources. This has eliminated the need for extrapolation of results from high and medium energies to low energies, as was done in FGR 2019. In addition, inconsistencies in the methodology identified in the review, such as computational representation of the adult female, were eliminated. The revised effective dose rate coefficients are consistent with values in ICRP Publication 144 except for differences in values clearly arising from differences in methodology; and differences in values for a relatively small set of very low-energy radionuclides with highly uncertain dose coefficients regardless of methodology.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2016

Evolving Adjustments to External (Gamma) Slope Factors for CERCLA Risk and Dose Assessments: The MCNP Years - 16307

To model the external exposure pathway in risk and dose assessments of radioactive contamination at Superfund sites, the U.S. Environmental Protection Agency (EPA) uses slope factors (SFs), also known as risk coefficients, and dose conversion factors (DCFs). Without any adjustment these external radiation exposure pathways effectively assumes that an individual is exposed to a source geometry that is effectively an infinite slab. The concept of an 'infinite slab' means that the thickness of the contaminated zone and its aerial extent are so large that it behaves as if it were infinite in its physical dimensions. EPA has been making increasingly complex adjustments to account for the extent of the contamination and its corresponding radiation field to provide more accurate risk and dose assessment modeling when using its calculators. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Applicability of ICRP 116 dose conversion coefficients to Korean pediatric population exposed to external photons - Paper 25

In the event of a radiological accident or attack, the estimation of organ dose is crucial to determine the possible health effect to general public. To that end, International Commission on Radiological Protection (ICRP) published a comprehensive set of organ dose conversion coefficients derived from the adult stylized (or mathematical) computational phantoms. More recently, ICRP updated the conversion coefficients based on their new reference computational phantoms, adult male and female voxel phantoms, exposed to a variety of radiation types and irradiation geometries. The adult voxel phantoms represent the organ mass and body dimensions of the Caucasian population. It is known that children generally receive greater radiation dose to organs compared to adults for external irradiation due to their smaller body size. that may provide less shielding to internal organs than adults. However, only a few studies investigated if the organ dose conversion coefficients developed for adults can be used for children and how much dose discrepancy needs to be expected. In this study, we adopted a Korean 7-year-old whole body voxel phantom to calculate a comprehensive organ dose conversion coefficients when the phantom is exposed to external photon irradiation geometries. We also compared the results with the data provided in the ICRP Publication 116, which is based on the ICRP adult voxel phantoms. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2013

Site Specific Reference Person Parameters and Derived Concentration Standards for the Savannah River Site

The purpose of this report is twofold. The first is to develop a set of behavioral parameters for a reference person specific for the Savannah River Site (SRS) such that the parameters can be used to determine dose to members of the public in compliance with Department of Energy (DOE) Order 458.1 “Radiation Protection of the Public and the Environment.” A reference person is a hypothetical, gender and age aggregation of human physical and physiological characteristics arrived at by international consensus for the purpose of standardizing radiation dose calculations. DOE O 458.1 states that compliance with the annual dose limit of 100 mrem (1 mSv) to a member of the public may be demonstrated by calculating the dose to the maximally exposed individual (MEI) or to a representative person. Historically, for dose compliance, SRS has used the MEI concept, which uses adult dose coefficients and adult male usage parameters. Beginning with the 2012 annual site environmental report, SRS will be using the representative person concept for dose compliance. The dose to a representative person will be based on 1) the SRS-specific reference person usage parameters at the 95th percentile of appropriate national or regional data, which are documented in this report, 2) the reference person (gender and age averaged) ingestion and inhalation dose coefficients provided in DOE Derived Concentration Technical Standard (DOE-STD-1196-2011), and 3) the external dose coefficients provided in the DC_PAK3 toolbox.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2017

Organ and effective dose rate coefficients for submersion exposure in occupational settings

External dose coefficients for environmental exposure scenarios are often computed using assumption on infinite or semi-infinite radiation sources. For example, in the case of a person standing on contaminated ground, the source is assumed to be distributed at a given depth (or between various depths) and extending outwards to an essentially infinite distance. In the case of exposure to contaminated air, the person is modeled as standing within a cloud of infinite, or semi-infinite, source distribution. However, these scenarios do not mimic common workplace environments where scatter off walls and ceilings may significantly alter the energy spectrum and dose coefficients. In this study, dose rate coefficients were calculated using the International Commission on Radiological Protection (ICRP) reference voxel phantoms positioned in rooms of three sizes representing an office, laboratory, and warehouse. For each room size calculations using the reference phantoms were performed for photons, electrons, and positrons as the source particles to derive mono-energetic dose rate coefficients. Since the voxel phantoms lack the resolution to perform dose calculations at the sensitive depth for the skin, a mathematical phantom was developed and calculations were performed in each room size with the three source particle types. Coefficients for the noble gas radionuclides of ICRP Publication 107 (e.g., Ne, Ar, Kr, Xe, and Rn) were generated by folding the corresponding photon, electron, and positron emissions over the mono-energetic dose rate coefficients. Finally, results indicate that the smaller room sizes have a significant impact on the dose rate per unit air concentration compared to the semi-infinite cloud case. For example, for Kr-85 the warehouse dose rate coefficient is 7% higher than the office dose rate coefficient while it is 71% higher for Xe-133.

61 RADIATION PROTECTION AND DOSIMETRY↗

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 OSTI2018

Dosimetry is Key to Good Epidemiology: Workers at Mallinckrodt Chemical Works had Seven Different Source Exposures

Mallinckrodt Chemical Works was the earliest uranium processing facility in the Manhattan Project, beginning in 1942. Even then, concern existed about possible health effects resulting from exposure to radiation and pitchblende dust. This concern was well founded as the facility processed Belgian Congo pitchblende ore that was up to 60% pure uranium with high 235 U content and up to 100 mg of radium per ton. Workers were exposed to external gamma radiation plus internal radiation from inhalation and ingestion of pitchblende dust (uranium, radium, and silica). Multiple sources of exposure were available for organ dose reconstruction to a degree unique for an epidemiologic study. Personal film badge measures available from 1945 captured external exposures. Additional external exposure included 15,518 occupational medical x-rays and 210 radiation exposure records from other facilities outside of Mallinckrodt employment. Organ dose calculations considered organ-specific coefficients that account for photon energy and job-specific orientation of workers to the radiation source during processing. Intakes of uranium and radium were based on 39,451 uranium urine bioassays and 2,341 breath radon measurements, and International Commission on Radiological Protection (ICRP) Publication 68 biokinetic models were used to estimate organ-specific radiation absorbed dose. Estimates of exposure to airborne radon and its short-lived progeny were based on radon measurements in work areas where radium-containing materials were handled or stored, together with estimated exposure times in these areas based on job titles. Dose estimates for radon and its short-lived progeny were based on models and methods recently recommended in ICRP Publication 137. This comprehensive dosimetric approach follows methods outlined by the National Council on Radiation Protection Scientific Committee 6-9 for the Million Worker Study. Annual doses were calculated for six organs: lung, brain, heart, kidney, colon and red bone marrow. Finally, evaluation and adjustment for individual cumulative measures of pitchblende dust inhalation were made for lung and kidney diseases.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2018

Comparison of neutron organ and effective dose coefficients for PIMAL stylized phantom in bent postures in standard irradiation geometries

Neutron dose coefficients for standard irradiation geometries have been reported in International Commission on Radiological Protection (ICRP) Publication 116 for the ICRP Publication 110 adult reference phantoms. In the present paper, organ and effective dose coefficients have been calculated for a receptor in both upright and articulated (bent) postures representing more realistic working postures exposed to a mono-energetic neutron radiation field. This work builds upon prior work by Dewji and co-workers comparing upright and bent postures for exposure to mono-energetic photon fields. Simulations were conducted using the Oak Ridge National Laboratory’s articulated stylized adult phantom, “Phantom wIth Moving Arms and Legs” (PIMAL) software package, and the Monte Carlo N-Particle (MCNP) version 6.1.1 radiation transport code. Organ doses were compared for the upright and bent (45° and 90°) phantom postures for neutron energies ranging from 1 × 10 -9 to 20 MeV for the ICRP Publication 116 external exposure geometries—antero-posterior (AP), postero-anterior (PA), and left and right lateral (LLAT, RLAT). Using both male and female phantoms, effective dose coefficients were computed using ICRP Publication 103 methodology. The resulting coefficients for articulated phantoms were compared to those of the upright phantom. Computed organ and effective dose coefficients are discussed as a function of neutron energy, phantom posture, and source irradiation geometry. For example, it is shown here that for the AP and PA irradiation geometries, the differences in the organ coefficients between the upright and bent posture become more pronounced with increasing bending angle. In the AP geometry, the brain dose coefficients are expectedly higher in the bent postures than in the upright posture, while all other organs have lower dose coefficients, with the thyroid showing the greatest difference. Finally and overall, the effective dose estimated for the upright phantom is more conservative than that for the articulated phantom, which may have ramifications in the estimation or reconstruction of radiation doses.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2014

Site-Specific Reference Person Parameters and Derived Concentration Standards for the Savannah River Site

The U.S. Department of Energy Order 458.1 states that the compliance with the 1 mSv annual dose constraint to a member of the public may be demonstrated by calculating dose to the maximally exposed individual (MEI) or to a representative person. Historically, the MEI concept was used for dose compliance at the Savannah River Site (SRS) using adult dose coefficients and adult male usage parameters. For future compliance, SRS plans to use the representative person concept for dose estimates to members of the public. The representative person dose will be based on the reference person dose coefficients from the U.S. DOE Derived Concentration Technical Standard and on usage parameters specific to SRS for the reference and typical person. Usage parameters and dose coefficients were determined for inhalation, ingestion and external exposure pathways. The parameters for the representative person were used to calculate and tabulate SRS-specific derived concentration standards (DCSs) for the pathways not included in DOE-STD-1196-2011.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2014

Photon Organ and Effective Dose Coefficients for Cranial and Caudal Irradiation Geometries - Paper 148

With the introduction of new recommendations of the ICRP in Publication 103 the methodology for determining the protection quantity effective dose has been modified. The modifications include changes in organ/tissue weighting factors as well as some changes in radiation weighting factors. Computations of protection dose coefficients now require the use of reference male and female voxel phantoms, see ICRP Publication 110. Computations of equivalent doses in organs and tissues are now performed in each phantom and the average is used to compute the effective dose. The latest set of external dose coefficients using the new reference phantoms and the latest guidance from the ICRP were performed by the ICRP DOCAL task group and reported in ICRP Publication 116, the revision of ICRP Publication 74 and ICRU Report 57. The coefficients were determined for the following irradiation geometries: anterior-posterior (AP), posterior-anterior (PA), right and left lateral (RLAT and LLAT), rotational (ROT), and isotropic (ISO). For all geometries, except the ISO geometry, parallel beams are used to compute the dose coefficients. In this paper, the methodology of ICRP Publication 116 was used to compute dose coefficients for photons for irradiations of the body with parallel beams directed upward from below the feet (caudal) and directed downward from above the head (cranial). These geometries may be encountered in the workplace for personnel standing on contaminated surfaces or volumes and from overhead sources. In this work calculations of organ and tissue kerma and absorbed doses for caudal and cranial exposures to photons ranging in energy from 10 keV to 10 GeV have been performed using the MCNP6.1 radiation transport code and the adult reference phantoms of ICRP Publication 110. Dose coefficients for caudal and cranial irradiation geometries have been calculated in accordance with ICRP Publication 103. The coefficients differ significantly from AP values for both organ and tissues of ICRP Publication 116, as would be expected. For workers in these scenarios organ and effective doses can differ significantly from AP values. For common photon energies (i.e. 100 keV to a few MeV) the AP effective dose overestimates the caudal and cranial values by about a factor of ten. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2020

Updated External Exposure Dose Coefficients

In 2018, the Environmental Protection Agency (EPA) updated and expanded the Federal Guidance Report (FGR) No. 12 to No. 15; tabulating age-specific external exposure effective dose rate coefficients for 1,252 radionuclides in the air, water, and soil (EPA 2018). However, the EPA discovered errors in the tables of dose coefficients for soil which led to the report being retracted for correction. This led to the adoption of “a uniform method of data-smoothing and extrapolation of monoenergetic organ doses at low energies across all exposure pathways; the originally-published dose coefficients utilized different methods for each media type” (EPA 2019). This resulted in all the external exposure dose coefficients being updated and FGR 15 being republished in August 2019 with corrected tables.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Enhanced Monte Carlo Simulation of the Voxel Phantom Lattice Submersed in a Contaminated Air Environment

External dose coefficient calculations reported for submersions in semi-infinite clouds, as in the Environmental Protection Agency: Federal Guidance Report 12, assume a person standing on the ground submerged in a radioactive cloud. In the scope of the work presented here, Oak Ridge National Laboratory's Center for Radiation Protection Knowledge has computed dose coefficients for monoenergetic photons, electrons, and positrons for worker exposures in three rooms of finite sizes representing an office, laboratory, and warehouse. In the process of simulating these exposures, International Commission on Radiological Protection (ICRP) Publication 110 reference adult male and female voxel phantom lattices were placed in the rooms filled with contaminated air. It was apparent that additional simulations were need to treat the non-tissue voxels surrounding the phantoms as containing the contaminated air. Dose coefficients for three room sizes representing an office, laboratory, and warehouse were computed. The coefficients for electrons exhibit little dependence on room size for most energies. However, for organs within the range of electron transport, the dose from source activity within the non-tissue voxels of the lattice surrounding the phantom becomes increasingly important. The photon dose coefficients vary based on room size because of longer mean free paths. It is important to include the air surrounding the voxel phantom within the voxel lattice in the calculations, when simulating a voxel phantom submersed in air for source particle energies below 200-300 keV, otherwise, the organ dose coefficient would be underestimated. Organ and effective dose coefficients for each gender, energy, and room size will be tabulated on the Oak Ridge National Laboratory - Center for Radiation Protection Knowledge website: http://ornl.gov/crpk.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2019

Radionuclide Risk Coefficient Uncertainty Distributions for Use in Performance Assessments - 19156

Traditionally, radiation dose has been used as the regulatory criterion for decision making in performance assessments (PAs), despite the concept of cancer risk arguably being a more easily understood endpoint than dose for the public, and a preferable metric for decision analysis. However, the United States Environmental Protection Agency (EPA) has developed risk parameters for use in risk assessments and PAs in the form of radionuclide-specific risk coefficients, or RCs, which are published in various Federal Guidance Reports (FGRs). RCs are employed as scalars to convert environmental radionuclide exposure (via ingestion, inhalation, and external exposure) estimates to incremental lifetime cancer risk (ILCR). ILCR is the additional cancer risk averaged over a lifetime, over-and-above all-cause cancer risk, associated with carcinogen exposure. Uncertainties associated with RCs are discussed in FGR 13 and in a joint report published by EPA and Oak Ridge National Laboratory, but statistical distributions for the RCs were not directly developed. The uncertainty assessments in the EPA/ONRL report were based on dosimetric modeling and elicitation of nine experts. From these sources, five uncertainty categories were defined based on ratios of the 5. and 95. quantiles of uncertainty. Thus, the representations of uncertainty defined by EPA/ORNL are not distributions in a traditional sense. Furthermore, the authors only addressed uncertainty in RCs for inhalation and ingestion pathways. We develop a methodology for assigning statistical distributions to EPA's published mortality RC values (i.e. to estimate the probability of dying from radiation-induced cancers) for the inhalation (slow, medium, and fast absorption types), ingestion (water, food, and soil), and external exposure pathways associated with all cancer types. We employed the same dosimetric concepts and assumptions as EPA/ORNL, including cancer risk (dose-response) distributional information from a previous expert elicitation, which was also employed by EPA/ORNL in their work. We assumed that uncertainty in RCs can be described by lognormal distributions based on the modeling conducted previously by EPA/ORNL, and that point estimates tabulated in FGR 13 represent median values of the distributions. We also extended EPA/ORNL's uncertainty ratios to statistical distributions for external exposure RCs. The results are a set of radionuclide-specific distributions that are applicable in PAs that focus on cancer mortality. These distributions are currently and will subsequently be employed in PAs at various sites, in addition to the required regulatory dose limits. Quantification of ILCRs and associated uncertainty in PAs will provide regulators and the public with additional information for good decision making. Future directions include further refinement of the distributions and development of statistical distributions for morbidity RCs, which are used to estimate the probability of developing, but not necessarily dying as a result of, radiation-induced cancer. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

SU-D-202-04: Validation of Deformable Image Registration Algorithms for Head and Neck Adaptive Radiotherapy in Routine Clinical Setting

Purpose: To evaluate the ROI contours and accumulated dose difference using different deformable image registration (DIR) algorithms for head and neck (H&N) adaptive radiotherapy. Methods: Eight H&N cancer patients were randomly selected from the affiliated hospital. During the treatment, patients were rescanned every week with ROIs well delineated by radiation oncologist on each weekly CT. New weekly treatment plans were also re-designed with consistent dose prescription on the rescanned CT and executed for one week on Siemens CT-on-rails accelerator. At the end, we got six weekly CT scans from CT1 to CT6 including six weekly treatment plans for each patient. The primary CT1 was set as the reference CT for DIR proceeding with the left five weekly CTs using ANACONDA and MORFEUS algorithms separately in RayStation and the external skin ROI was set to be the controlling ROI both. The entire calculated weekly dose were deformed and accumulated on corresponding reference CT1 according to the deformation vector field (DVFs) generated by the two different DIR algorithms respectively. Thus we got both the ANACONDA-based and MORFEUS-based accumulated total dose on CT1 for each patient. At the same time, we mapped the ROIs on CT1 to generate the corresponding ROIs on CT6 using ANACONDA and MORFEUS DIR algorithms. DICE coefficients between the DIR deformed and radiation oncologist delineated ROIs on CT6 were calculated. Results: For DIR accumulated dose, PTV D95 and Left-Eyeball Dmax show significant differences with 67.13 cGy and 109.29 cGy respectively (Table1). For DIR mapped ROIs, PTV, Spinal cord and Left-Optic nerve show difference with −0.025, −0.127 and −0.124 (Table2). Conclusion: Even two excellent DIR algorithms can give divergent results for ROI deformation and dose accumulation. As more and more TPS get DIR module integrated, there is an urgent need to realize the potential risk using DIR in clinical.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2014

Estimated External Doses to Members of the Public from Patients with {sup 131}I Treatment - Paper 143

To protect individuals from the radiation emitted by a patient receiving {sup 131}I treatment a licensee must estimate the maximum likely effective dose to members of the public. Under current guidance by the United States Nuclear Regulatory Commission (NRC), a licensee may authorize the release of the patient if the effective dose is unlikely to exceed 5 mSv. NRC provides guidance to its licensees, in NRC Regulatory Guide 8.39, on acceptable methods that may be used to calculate the dose to a member of the public that may result from releasing the patient. To further assist licensees in implementing the regulation, NRC used conservative assumptions to estimate the activity of radionuclide in a patient, and the dose rate at 1 meter from the patient, that would be indicators of a potential dose to a member of the public of 5 mSv. Licensees may use these tables as operational criteria for releasing a patient if the conservative assumptions are considered acceptable. Licensees may also perform their own calculations, using case-specific assumptions, to show compliance with 5 mSv release criterion. The scope of this work is to provide dose coefficients for a variety of plausible public exposure scenarios that could be used by licensees to to show compliance. In the patient-specific calculations, the licensee is free to consider factors such as occupancy factor, effective half-life of the radionuclide in the patient's body, and tissue shielding. A conservative method of calculating the maximum likely dose to members of the public is provided in NRC Reg. Guide 8.39, which assumes no tissue attenuation of by the patient's body, no biological component that reduces the effective half-life of the radionuclide, and the patient's body burden is a point source. The assumption of no biological removal from the body introduces some conservatism into dose estimates for {sup 131}I. Also, it has been shown that the assumption of a point source with no tissue attenuation substantially overestimates the dose rate to nearby persons. The current method for calculating the maximum dose rate to members of the public from a nuclear medicine patient provided in US-NRC Regulatory Guide 8.39 assumes the patient's body burden is a point-source. Lack of consideration of tissue attenuation by the patient's body and biological removal of the radionuclide leads to a substantial overestimate of the actual external dose rate to a nearby member of the public. However, use of these assumptions in estimating potential dose to a member of the public simplifies the calculations considerably and, although generally conservative, the calculations serve as a useful screening tool. Many, if not most patients, could be released on that basis. If not, then refined calculations may be used, and the coefficients described in this paper provide one of the means for refining the calculations in cases where the screening calculations were not adequate. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2023

Findings during Incorporation of the Federal Guidance Report No. 15 External Dose Coefficients into the RESRAD Family of Codes

Argonne National Laboratory was tasked by the U.S. Department of Energy (DOE) to incorporate the recently published Federal Guidance Report No. 15 (FGR 15) (EPA 2019) external dose rate coefficients into the database of the RESRAD family of codes. FGR 15 is one of two publications that contain age-specific external effective dose rate coefficients that are based on the International Commission on Radiological Protection (ICRP) Publication 103 (ICRP-103) methodology (ICRP 2007) and the ICRP-107 (ICRP 2008) nuclear decay data; ICRP-144 (ICRP 2020) is the other publication.

61 RADIATION PROTECTION AND DOSIMETRY↗