Methods of improving brain dose estimates for internally deposited radionuclides.

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

Mortality among Tennessee Eastman Corporation (TEC) uranium processing workers, 1943-2019.

BACKGROUND: There are few occupational studies of women exposed to ionizing radiation. During World War II, the Tennessee Eastman Corporation (TEC) operated an electromagnetic field separation facility of 1152 calutrons to obtain enriched uranium (235U) used for the Hiroshima atomic bomb. Thousands of women were involved in these operations. MATERIALS AND METHODS: A new study was conducted of 13,951 women and 12,699 men employed at TEC between 1943 and 1947 for at least 90 days. Comprehensive dose reconstruction techniques were used to estimate lung doses from the inhalation of uranium dust based on airborne measurements. Vital status through 2018/2019 was obtained from the National Death Index, Social Security Death Index, Tennessee death records and online public record databases. Analyses included standardized mortality ratios (SMRs) and Cox proportional hazards models. RESULTS: Most workers were hourly (77.7%), white (95.6%), born before 1920 (58.3%), worked in dusty environments (57.0%), and had died (94.9%). Vital status was confirmed for 97.4% of the workers. Women were younger than men when first employed: mean ages 25.0 years and 33.0 years, respectively. The estimated mean absorbed dose to the lung was 32.7 mGy (max 1048 mGy) for women and 18.9 mGy (max 501 mGy) for men. The mean dose to thoracic lymph nodes (TLNs) was 127 mGy. Statistically significant SMRs were observed for lung cancer (SMR 1.25; 95% CI 1.19, 1.31; n = 1654), nonmalignant respiratory diseases (NMRDs) (1.23; 95% CI 1.19, 1.28; n = 2585), and cerebrovascular disease (CeVD) (1.13; 95% CI 1.08, 1.18; n = 1945). For lung cancer, the excess relative rate (ERR) at 100 mGy (95% CI) was 0.01 (-0.10, 0.12; n = 652) among women, and -0.15 (-0.38, 0.07; n = 1002) among men based on a preferred model for men with lung doses <300 mGy. NMRD and non-Hodgkin lymphoma were not associated with estimated absorbed dose to the lung or TLN. CONCLUSIONS: There was little evidence that radiation increased the risk of lung cancer, suggesting that inhalation of uranium dust and the associated high-LET alpha particle exposure to lung tissue experienced over a few years is less effective in causing lung cancer than other types of exposures. There was no statistically significant difference in the lung cancer risk estimates between men and women. The elevation of certain causes of death such as CeVD is unexplained and will require additional scrutiny of workplace or lifestyle factors given that radiation is an unlikely contributor since only the lung and lymph nodes received appreciable dose.

MPS dose reconstruction for internal emitters: some site-specific issues and approaches.

BACKGROUND: As part of the Million Person Study (MPS), dose reconstructions for internal emitters have been performed for several U.S. facilities where large quantities of radionuclides were handled. The main challenges and dominant sources of potential error in retrospective dose estimates for internally exposed workers have been found to vary from site to site. This article discusses some important issues encountered in dose reconstructions performed for selected MPS sites and the approaches used to address those issues. The focus is on some foundational components of retrospective dose assessments that have received little attention in the literature. METHODS: The discussion is built around illustrative exposure data and dose reconstructions for workers at selected facilities addressed in the MPS. Related findings at some non-MPS sites are also discussed. RESULTS: Each of the following items has been found to be a major source of potential error in reconstructed tissue doses for some MPS sites: identification of all dosimetrically important internal emitters; the time pattern of intake; the mode(s) of intake; reliability of bioassay measurements; application of surrogate (coworker) information in lieu of, or in conjunction with, worker-specific monitoring data; the chemical and physical forms of inhaled radionuclides; and the relation of air monitoring data to actual intake. CONCLUSIONS: (1) Much of the dose reconstruction effort for internal emitters should be devoted to development of best feasible exposure scenarios. (2) Coworker data should be used to assign exposure scenarios or dose estimates to workers with missing exposure data only if there is compelling evidence of similar coworker exposure. (3) Bioassay data for some radionuclides and periods of operation at MPS sites are of questionable reliability due to sizable uncertainties associated with contamination, recovery, or background issues. (4) Dose estimates derived solely from air monitoring data should be treated as highly uncertain values in the absence of site-specific information demonstrating that the data are reasonably predictive of intake. (5) For intakes known or assumed to be via inhalation, the uncertainty in lung dose typically is much greater than the uncertainty in dose to systemic tissues, when dose estimates are based on urinary excretion data. (6) The lung dose estimate often can be improved through development of site-specific respiratory absorption parameter values. (7) There is generally insufficient site-specific information to justify development of site-specific systemic models.

Specific Absorbed Fractions for Spontaneous Fission Neutron Emitters in the ICRP Reference Pediatric Voxel Phantom Series.

ABSTRACT: Specific absorbed fractions (SAFs) are key components in the workflow of internal exposure assessment following the intake of a radionuclide, allowing quick conversion of particle energy released in a source region to the expected absorbed dose in target regions throughout the body. For data completeness, SAFs for spontaneous fission neutron emitters are currently needed for the recently adopted ICRP reference pediatric voxel phantom series. With 77 source regions within each reference individual and 28 radionuclides decaying via spontaneous fission, full Monte Carlo simulation requires significant computation time. In order to reduce this burden, a novel method for neutron SAF estimation was undertaken. The Monte Carlo N-Particle version 6.1 (MCNP6) simulation package was chosen to simulate the 252 Cf Watt fission neutron spectrum originating from 15 source regions in each phantom; dose estimation within 41 target tissues allowed for assessment of the SAF value for each source-target pair. For the remaining source regions, chord length distributions were computed using MATLAB code to determine the separation between the source-target pairs within the pediatric phantom series. These distance distributions were used in conjunction with a 252 Cf neutron dose point kernel calculated in soft tissue, which was modified to account for the source region's depth from the surface of the body. Lastly, the 252 Cf SAF dataset was extended to the other 27 spontaneous fission neutron emitters based on differences in the Watt fission spectrum parameters of each radionuclide. This methodology has been shown to accurately estimate spontaneous fission neutron SAFs to within 20% of the Monte Carlo estimated value for most source-target pairs in the ICRP reference pediatric series.

Radium dial workers: back to the future.

PURPOSE: This paper 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. BACKGROUND: 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. EPIDEMIOLOGIC APPROACH: The mortality experience of 3,276 radium dial painters and handlers employed between 1913 and 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. DISCUSSION: 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.

Mortality among workers at the Los Alamos National Laboratory, 1943-2017.

BACKGROUND: During World War II (WWII), the Manhattan Engineering District established a secret laboratory in the mountains of northern New Mexico. The mission was to design, construct and test the first atomic weapon, nicknamed 'The Gadget' that was detonated at the TRINITY site in Alamogordo, NM. After WWII, nuclear weapons research continued, and the laboratory became the Los Alamos National Laboratory (LANL). MATERIALS AND METHODS: The mortality experience of 26,328 workers first employed between 1943 and 1980 at LANL was determined through 2017. Included were 6157 contract workers employed by the ZIA Company. Organ dose estimates for each worker considered all sources of exposure, notably photons, neutrons, tritium, 238Pu and 239Pu. Vital status determination included searches within the National Death Index, Social Security Administration and New Mexico State Mortality Files. Standardized Mortality Ratios (SMR) and Cox regression models were used in the analyses. RESULTS: Most workers (55%) were hired before 1960, 38% had a college degree, 25% were female, 81% white, 13% Hispanic and 60% had died. Vital status was complete, with only 0.1% lost to follow-up. The mean dose to the lung for the 17,053 workers monitored for radiation was 28.6 weighted-mGy (maximum 16.8 weighted-Gy) assuming a Dose Weighting Factor of 20 for alpha particle dose to lung. The Excess Relative Risk (ERR) at 100 weighted-mGy was 0.01 (95%CI -0.02, 0.03; n = 839) for lung cancer. The ERR at 100 mGy was -0.43 (95%CI -1.11, 0.24; n = 160) for leukemia other than chronic lymphocytic leukemia (CLL), -0.06 (95%CI -0.16, 0.04; n = 3043) for ischemic heart disease (IHD), and 0.29 (95%CI 0.02, 0.55; n = 106) for esophageal cancer. Among the 6499 workers with measurable intakes of plutonium, an increase in bone cancer (SMR 2.44; 95%CI 0.98, 5.03; n = 7) was related to dose. The SMR for berylliosis was significantly high, based on 4 deaths. SMRs for Hispanic workers were significantly high for cancers of the stomach and liver, cirrhosis of the liver, nonmalignant kidney disease and diabetes, but the excesses were not related to radiation dose. CONCLUSIONS: There was little evidence that radiation increased the risk of lung cancer or leukemia. Esophageal cancer was associated with radiation, and plutonium intakes were linked to an increase of bone cancer. IHD was not associated with radiation dose. More precise evaluations will await the pooled analysis of workers with similar exposures such as at Rocky Flats, Savannah River and Hanford.

Dosimetry and uncertainty approaches for the million person study of low-dose radiation health effects: overview of the recommendations in NCRP Report No. 178.

PURPOSE: Scientific Committee 6-9 was established by the National Council on Radiation Protection and Measurements (NCRP), charged to provide guidance in the derivation of organ doses and their uncertainty, and produced a report, NCRP Report No. 178, Deriving Organ Doses and their Uncertainty for Epidemiologic Studies with a focus on the Million Person Study of Low-Dose Radiation Health Effects (MPS). This review summarizes the conclusions and recommendations of NCRP Report No. 178, with a concentration on and overview of the dosimetry and uncertainty approaches for the cohorts in the MPS, along with guidelines regarding the essential approaches used to estimate organ doses and their uncertainties (from external and internal sources) within the framework of an epidemiologic study. CONCLUSIONS: The success of the MPS is tied to the validity of the dose reconstruction approaches to provide realistic estimates of organ-specific radiation absorbed doses that are as accurate and precise as possible and to properly evaluate their accompanying uncertainties. The dosimetry aspects for the MPS are challenging in that they address diverse exposure scenarios for diverse occupational groups being studied over a period of up to 70 y. Specific dosimetric reconstruction issues differ among the varied exposed populations that are considered: atomic veterans, U.S. Department of Energy workers exposed to both penetrating radiation and intakes of radionuclides, nuclear power plant workers, medical radiation workers, and industrial radiographers. While a major source of radiation exposure to the study population comes from external gamma- or x-ray sources, for some of the study groups, there is also a meaningful component of radionuclide intakes that requires internal radiation dosimetry assessments.

MIRD pamphlet No. 21: a generalized schema for radiopharmaceutical dosimetry--standardization of nomenclature.

The internal dosimetry schema of the Medical Internal Radiation Dose (MIRD) Committee of the Society of Nuclear Medicine has provided a broad framework for assessment of the absorbed dose to whole organs, tissue subregions, voxelized tissue structures, and individual cellular compartments for use in both diagnostic and therapeutic nuclear medicine. The schema was originally published in 1968, revised in 1976, and republished in didactic form with comprehensive examples as the MIRD primer in 1988 and 1991. The International Commission on Radiological Protection (ICRP) is an organization that also supplies dosimetric models and technical data, for use in providing recommendations for limits on ionizing radiation exposure to workers and members of the general public. The ICRP has developed a dosimetry schema similar to that of the MIRD Committee but has used different terminology and symbols for fundamental quantities such as the absorbed fraction, specific absorbed fraction, and various dose coefficients. The MIRD Committee objectives for this pamphlet are 3-fold: to restate its schema for assessment of absorbed dose in a manner consistent with the needs of both the nuclear medicine and the radiation protection communities, with the goal of standardizing nomenclature; to formally adopt the dosimetry quantities equivalent dose and effective dose for use in comparative evaluations of potential risks of radiation-induced stochastic effects to patients after nuclear medicine procedures; and to discuss the need to identify dosimetry quantities based on absorbed dose that address deterministic effects relevant to targeted radionuclide therapy.

The Dosimetric Demise of 97MNB.

Radionuclide differences between ICRP Publication 38 and its succeeding work, ICPR Publication 107, are reviewed. The specific example of the isomer Nb is discussed, examining how dose reporting for this nuclide can be an issue.

Mortality among radiation workers at Rocketdyne (Atomics International), 1948-1999.

A retrospective cohort mortality study was conducted of workers engaged in nuclear technology development and employed for at least 6 months at Rocketdyne (Atomics International) facilities in California, 1948-1999. Lifetime occupational doses were derived from company records and linkages with national dosimetry data sets. International Commission on Radiation Protection (ICRP) biokinetic models were used to estimate radiation doses to 16 organs or tissues after the intake of radionuclides. Standardized mortality ratios (SMRs) compared the observed numbers of deaths with those expected in the general population of California. Cox proportional hazards models were used to evaluate dose-response trends over categories of cumulative radiation dose, combining external and internal organ-specific doses. There were 5,801 radiation workers, including 2,232 monitored for radionuclide intakes. The mean dose from external radiation was 13.5 mSv (maximum 1 Sv); the mean lung dose from external and internal radiation combined was 19.0 mSv (maximum 3.6 Sv). Vital status was determined for 97.6% of the workers of whom 25.3% (n = 1,468) had died. The average period of observation was 27.9 years. All cancers taken together (SMR 0.93; 95% CI 0.84-1.02) and all leukemia excluding chronic lymphocytic leukemia (CLL) (SMR 1.21; 95% CI 0.69-1.97) were not significantly elevated. No SMR was significantly increased for any cancer or for any other cause of death. The Cox regression analyses revealed no significant dose-response trends for any cancer. For all cancers excluding leukemia, the RR at 100 mSv was estimated as 1.00 (95% CI 0.81-1.24), and for all leukemia excluding CLL it was 1.34 (95% CI 0.73-2.45). The nonsignificant increase in leukemia (excluding CLL) was in accord with expectation from other radiation studies, but a similar nonsignificant increase in CLL (a malignancy not found to be associated with radiation) tempers a causal interpretation. Radiation exposure has not caused a detectable increase in cancer deaths in this population, but results are limited by small numbers and relatively low career doses.

Revisions to the ORNL series of adult and pediatric computational phantoms for use with the MIRD schema.

The age-dependent series of stylized computational phantoms developed at the Oak Ridge National Laboratory in the late 1970's to early 1980's has found wide applicability in dosimetry studies ranging from dose coefficient compilations for external and internal photon emitters, simulations of patient radiological exams, and dose reconstruction activities. In the present study, we report on a series of revisions to the Oak Ridge National Laboratory series for their intended use within the MIRD schema of medical internal dosimetry. These revisions were made to (1) incorporate recent developments in stylized models of the head, brain, kidneys, rectosigmoid colon, and extra-pulmonary airways; (2) incorporate new models of the salivary glands and the mucosa layer of the urinary bladder, alimentary tract organs, and respiratory airways; (3) adopt reference values of elemental tissue compositions and mass densities from ICRP Publication 89 and ICRU Report 46; (4) provide for explicit treatment of left and right organs within organ pairs; (5) provide for a systematic tabulation of electron absorbed fractions as a function of energy and subject age for all internal organs; and (6) provide for methods of deriving patient-specific values of the specific absorbed fraction for both electrons and photons through interpolation/extrapolation of their phantom-derived values. While tomographic computational phantoms provide improved anatomic realism given the CT or MR image sets used in their construction, there does not yet exist a comprehensive series of reference pediatric tomographic phantoms, nor the ability to simulate very fine anatomic structures as can be modeled via mathematical approximation. Consequently, stylized pediatric phantoms will continue to fill this data need in medical dosimetry.

A comprehensive dose reconstruction methodology for former rocketdyne/atomics international radiation workers.

Incomplete radiation exposure histories, inadequate treatment of internally deposited radionuclides, and failure to account for neutron exposures can be important uncertainties in epidemiologic studies of radiation workers. Organ-specific doses from lifetime occupational exposures and radionuclide intakes were estimated for an epidemiologic study of 5,801 Rocketdyne/Atomics International (AI) radiation workers engaged in nuclear technologies between 1948 and 1999. The entire workforce of 46,970 Rocketdyne/AI employees was identified from 35,042 Kardex work histories cards, 26,136 electronic personnel listings, and 14,189 radiation folders containing individual exposure histories. To obtain prior and subsequent occupational exposure information, the roster of all workers was matched against nationwide dosimetry files from the Department of Energy, the Nuclear Regulatory Commission, the Landauer dosimetry company, the U.S. Army, and the U.S. Air Force. Dosimetry files of other worker studies were also accessed. Computation of organ doses from radionuclide intakes was complicated by the diversity of bioassay data collected over a 40-y period (urine and fecal samples, lung counts, whole-body counts, nasal smears, and wound and incident reports) and the variety of radionuclides with documented intake including isotopes of uranium, plutonium, americium, calcium, cesium, cerium, zirconium, thorium, polonium, promethium, iodine, zinc, strontium, and hydrogen (tritium). Over 30,000 individual bioassay measurements, recorded on 11 different bioassay forms, were abstracted. The bioassay data were evaluated using ICRP biokinetic models recommended in current or upcoming ICRP documents (modified for one inhaled material to reflect site-specific information) to estimate annual doses for 16 organs or tissues taking into account time of exposure, type of radionuclide, and excretion patterns. Detailed internal exposure scenarios were developed and annual internal doses were derived on a case-by-case basis for workers with committed equivalent doses indicated by screening criteria to be greater than 10 mSv to the organ with the highest internal dose. Overall, 5,801 workers were monitored for radiation at Rocketdyne/AI: 5,743 for external exposure and 2,232 for internal intakes of radionuclides; 41,169 workers were not monitored for radiation. The mean cumulative external dose based on Rocketdyne/AI records alone was 10.0 mSv, and the dose distribution was highly skewed with most workers experiencing low cumulative doses and only a few with high doses (maximum 500 mSv). Only 45 workers received greater than 200 mSv while employed at Rocketdyne/AI. However, nearly 32% (or 1,833) of the Rocketdyne/AI workers had been monitored for radiation at other nuclear facilities and incorporation of these doses increased the mean dose to 13.5 mSv (maximum 1,005 mSv) and the number of workers with >200 mSv to 69. For a small number of workers (n=292), lung doses from internal radionuclide intakes were relatively high (mean 106 mSv; maximum 3,560 mSv) and increased the overall population mean dose to 19.0 mSv and the number of workers with lung dose>200 mSv to 109. Nearly 10% of the radiation workers (584) were monitored for neutron exposures (mean 1.2 mSv) at Rocketdyne/AI, and another 2% were monitored for neutron exposures elsewhere. Interestingly, 1,477 workers not monitored for radiation at Rocketdyne/AI (3.6%) were found to have worn dosimeters at other nuclear facilities (mean external dose of 2.6 mSv, maximum 188 mSv). Without considering all sources of occupational exposure, an incorrect characterization of worker exposure would have occurred with the potential to bias epidemiologic results. For these pioneering workers in the nuclear industry, 26.5% of their total occupational dose (collective dose) was received at other facilities both prior to and after employment at Rocketdyne/AI. In addition, a small number of workers monitored for internal radionuclides contributed disproportionately to the number of workers with high lung doses. Although nearly 12% of radiation workers had been monitored for neutron exposures during their career, the cumulative dose levels were small in comparison with other external and internal exposure. Risk estimates based on nuclear worker data must be interpreted cautiously if internally deposited radionuclides and occupational doses received elsewhere are not considered.

39th Lauriston S. Taylor Lecture: Dosimetry of Internal Emitters: Contribution of Radiation Protection Bodies and Radiological Events.

Since the early days of the Manhattan Engineer District, Oak Ridge National Laboratory (ORNL) has served to advance the dosimetry models used to set protection standards for radionuclides taken into the body. Throughout the years, this effort benefited significantly from ORNL staff's active participation in national and international scientific bodies. The first such interaction was in 1946 with the National Committee on Radiation Protection (NCRP), chaired by L.S. Taylor, which led to the 1949 to 1953 series of tripartite conferences of experts from Canada, the United Kingdom, and the United States. These conferences addressed the need for standardization of dosimetry models and led to the establishment of an anatomic and physiologic model called "Standard Man," a precursor of the reference worker defined in Publication 23 of the International Commission on Radiological Protection (ICRP). Standard Man was used in setting the maximum permissible concentrations in air and water published in NBS Handbook 52 and subsequent reports by NCRP and ICRP. K.Z. Morgan, then director of the Health Physics Division at ORNL, participated in the tripartite conferences and subsequently established ORNL as a modeling and computational resource for development of radiation protection standards. ORNL's role expanded with participation in the work of the Medical Internal Radiation Dose (MIRD) Committee of the Society of Nuclear Medicine. Results of interactions with the MIRD Committee are evident in the radiation protection guidance for internal emitters in ICRP Publication 30. The annual limit on intake and derived air concentration values tabulated in Publication 30 were computed by an ORNL-based task group of ICRP Committee 2. A few years after the appearance of Publication 30, the Chernobyl nuclear reactor accident made clear the need to develop standard dosimetry models for pre-adult ages as members of the public. In the late 1980s, ICRP began an effort to extend its reference worker concept to a reference family and develop dosimetric models for application to intake of radionuclides by members of the public. However, the modeling approach underlying the ICRP Publication 30 computational framework was not amenable to age and gender considerations. With support of U.S. federal agencies, ORNL had begun efforts in the early 1980s to develop age- and gender-specific dosimetric models, including physiologically informed biokinetic models and age-specific dosimetric phantoms. ORNL's models and methods became the starting point for the ICRP's series of reports on dose coefficients for radionuclide intake by the public. Currently ICRP Committee 2 is overseeing development of a second generation of post-Chernobyl models and methods, with updates of Publications 30 and 68 soon to appear and new models for members of the public in preparation. The focus of this Lauriston S. Taylor Lecture is to chronicle advancements in the dosimetry of internal emitters with some discussion of models and methods but with due deference to decisions within scientific bodies and stimulated by radiological events.

MIRD Pamphlet No. 26: Joint EANM/MIRD Guidelines for Quantitative 177Lu SPECT Applied for Dosimetry of Radiopharmaceutical Therapy.

The accuracy of absorbed dose calculations in personalized internal radionuclide therapy is directly related to the accuracy of the activity (or activity concentration) estimates obtained at each of the imaging time points. MIRD Pamphlet no. 23 presented a general overview of methods that are required for quantitative SPECT imaging. The present document is next in a series of isotope-specific guidelines and recommendations that follow the general information that was provided in MIRD 23. This paper focuses on (177)Lu (lutetium) and its application in radiopharmaceutical therapy.

Dose reconstruction for the million worker study: status and guidelines.

The primary aim of the epidemiologic study of one million U.S. radiation workers and veterans [the Million Worker Study (MWS)] is to provide scientifically valid information on the level of radiation risk when exposures are received gradually over time and not within seconds, as was the case for Japanese atomic bomb survivors. The primary outcome of the epidemiologic study is cancer mortality, but other causes of death such as cardiovascular disease and cerebrovascular disease will be evaluated. The success of the study is tied to the validity of the dose reconstruction approaches to provide realistic estimates of organ-specific radiation absorbed doses that are as accurate and precise as possible and to properly evaluate their accompanying uncertainties. The dosimetry aspects for the MWS are challenging in that they address diverse exposure scenarios for diverse occupational groups being studied over a period of up to 70 y. The dosimetric issues differ among the varied exposed populations that are considered: atomic veterans, U.S. Department of Energy workers exposed to both penetrating radiation and intakes of radionuclides, nuclear power plant workers, medical radiation workers, and industrial radiographers. While a major source of radiation exposure to the study population comes from external gamma- or x-ray sources, for some of the study groups, there is a meaningful component of radionuclide intakes that requires internal radiation dosimetry assessments. Scientific Committee 6-9 has been established by the National Council on Radiation Protection and Measurements (NCRP) to produce a report on the comprehensive organ dose assessment (including uncertainty analysis) for the MWS. The NCRP dosimetry report will cover the specifics of practical dose reconstruction for the ongoing epidemiologic studies with uncertainty analysis discussions and will be a specific application of the guidance provided in NCRP Report Nos. 158, 163, 164, and 171. The main role of the Committee is to provide guidelines to the various groups of dosimetrists involved in the MWS to ensure that certain dosimetry criteria are considered: calculation of annual absorbed doses in the organs of interest, separation of low and high linear-energy transfer components, evaluation of uncertainties, and quality assurance and quality control. It is recognized that the MWS and its approaches to dosimetry are a work in progress and that there will be flexibility and changes in direction as new information is obtained with regard to both dosimetry and the epidemiologic features of the study components. This paper focuses on the description of the various components of the MWS, the available dosimetry results, and the challenges that have been encountered. It is expected that the Committee will complete its report in 2016.

Workshop Report on Atomic Bomb Dosimetry--Review of Dose Related Factors for the Evaluation of Exposures to Residual Radiation at Hiroshima and Nagasaki.

Groups of Japanese and American scientists, supported by international collaborators, have worked for many years to ensure the accuracy of the radiation dosimetry used in studies of health effects in the Japanese atomic bomb survivors. Reliable dosimetric models and systems are especially critical to epidemiologic studies of this population because of their importance in the development of worldwide radiation protection standards. While dosimetry systems, such as Dosimetry System 1986 (DS86) and Dosimetry System 2002 (DS02), have improved, the research groups that developed them were unable to propose or confirm an additional contribution by residual radiation to the survivor's total body dose. In recognition of the need for an up-to-date review of residual radiation exposures in Hiroshima and Nagasaki, a half-day technical session was held for reports on newer studies at the 59 th Annual HPS Meeting in 2014 in Baltimore, MD. A day-and-a-half workshop was also held to provide time for detailed discussion of the newer studies and to evaluate their potential use in clarifying the residual radiation exposure to atomic bomb survivors at Hiroshima and Nagasaki. The process also involved a re-examination of very early surveys of radioisotope emissions from ground surfaces at Hiroshima and Nagasaki and early reports of health effects. New insights were reported on the potential contribution to residual radiation from neutron-activated radionuclides in the airburst's dust stem and pedestal and in unlofted soil, as well as from fission products and weapon debris from the nuclear cloud. However, disparate views remain concerning the actual residual radiation doses received by the atomic bomb survivors at different distances from the hypocenter. The workshop discussion indicated that measurements made using thermal luminescence and optically stimulated luminescence, like earlier measurements, especially in very thin layers of the samples, could be expanded to detect possible radiation exposures to beta particles and to determine their significance plus the extent of the various residual radiation areas at Hiroshima and Nagasaki. Other suggestions for future residual radiation studies are included in this workshop report.

Web-access to a Lung Deposition Model for radiation protection and physiological modeling.

A C++ implementation of the ICRP Lung Deposition Model was validated and made accessible by creating a web interface using Perl. Considerable understanding of this lung deposition model was obtained and a small number of errors identified in the code during this process. Documentation was developed using MathML and scalable vector graphics (SVG).

ICRP Publication 116--the first ICRP/ICRU application of the male and female adult reference computational phantoms.

ICRP Publication 116 on 'Conversion coefficients for radiological protection quantities for external radiation exposures', provides fluence-to-dose conversion coefficients for organ-absorbed doses and effective dose for various types of external exposures (ICRP 2010 ICRP Publication 116). The publication supersedes the ICRP Publication 74 (ICRP 1996 ICRP Publication 74, ICRU 1998 ICRU Report 57), including new particle types and expanding the energy ranges considered. The coefficients were calculated using the ICRP/ICRU computational phantoms (ICRP 2009 ICRP Publication 110) representing the reference adult male and reference adult female (ICRP 2002 ICRP Publication 89), together with a variety of Monte Carlo codes simulating the radiation transport in the body. Idealized whole-body irradiation from unidirectional and rotational parallel beams as well as isotropic irradiation was considered for a large variety of incident radiations and energy ranges. Comparison of the effective doses with operational quantities revealed that the latter quantities continue to provide a good approximation of effective dose for photons, neutrons and electrons for the 'conventional' energy ranges considered previously (ICRP 1996, ICRU 1998), but not at the higher energies of ICRP Publication 116.