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Given the need for criteria to control the radiation doses due to radionuclide inhalation, in 1994 the International Commission on Radiological Protection presented a classification for radioactive compounds based on their pulmonary absorption rates. The Commission classified the compounds into fast, moderate, and slow categories and assigned to each material a default absorption class. Nevertheless, the proposed categories do not always resemble the actual behavior of the classified materials in the pulmonary environment. Therefore, the Commission itself suggested the assessment of the inhalation risk of a particulate substance referring to an in vivo study using the same material. Since it is not possible to trace in literature in vivo studies analyzing the physiological behavior of the totality of inhalable radioactive materials, the Commission suggested adopting in vitro systems simulating the pulmonary mechanism. For this reason, in the last 50 y, many simulating setups have been implemented, but none of these seemed to reproduce the lung dissolution dynamics effectively as the results were not comparable with the ones obtained using in vivo techniques. This paper aims to describe an innovative experimental apparatus implemented as an attempt to add another step toward the realization of a gold standard. In particular, the system was validated with BaSO 4 particulate, and the resulting error with respect to the physiological expected value figured as less than 4%. The system was then employed for the lung dissolution tests of radioactive graphite extracted from Politecnico di Milano's nuclear reactor to assess the radiobiological risk due to a slow lung absorption process that workers might run into during the reactor decommissioning.

Animal experiment findings suggest that high doses of ionizing radiation exposure (>1.0 Gy) may cause genetic and epigenetic effects in offspring. However, epidemiological studies of offspring of radiation-exposed parents did not find increased risks of any health effects. Findings of cellular/experimental investigations and studies of human health effects are contradicting, and further investigations are needed to help resolve ambiguities using updated and/or improved data. This paper provides a detailed description of a database of families of workers of the first Russian nuclear facility, Mayak Production Association, located in the Southern Urals in the Chelyabinsk region close to Ozyorsk city, which started its operation in 1948 and today consists of reactors, radiochemical and plutonium production plants, and auxiliary facilities. The Mayak worker cohort includes 22,377 individuals (25% females) who were hired at one of the main Mayak PA facilities between 1948 and 1982 and were externally or internally exposed to ionizing radiation over prolonged periods. Advantages of the cohort include its large size, extensive follow-up period (70 y), individually measured doses from external and internal exposure and the wide range of these doses, heterogeneity by gender/age/ethnicity/initial health status, complete data on vital status and causes of death, available medical information on morbidity and reproduction, available data on non-radiation factors, and stored biological specimens donated by more than one-third of the cohort members. Based on medical and dosimetry database "Clinics" containing raw data on workers of the study cohort, the Mayak workers' family and offspring database was created. To date, it comprises 12,195 family couples (a husband and a wife) and 16,585 offspring. Biological specimens are available for more than 1,000 family triads (a husband, a wife, and their child). Stages of assembling the database and its descriptive characteristics are presented in this paper. Examples of potential applications of the database for investigations of non-targeted and transgenerational radiation effects in offspring of exposed parents are discussed.

The excess relative risk (ERR) of mortality for circulatory disease among nuclear workers was reanalyzed by taking into consideration the annual dose as the dose rate using publicly available epidemiological data of the Hanford site dedicated to the cohort study of nuclear workers in the US, the UK, and Canada. Values of the dose rate (cut-points) were chosen at 2 mSv y intervals from 2 to 40 mSv y, and risk estimates were made for 32,988 workers, considering the doses accumulated below and above each cut-point to have different effects. Similarly to that in the previous study for cancer by Sasaki et al., examinations of sensitivity analysis were also carried out for different risk models, lag periods, and impacts of adjusting the monitoring period to find the effect of the dose rate. As a result, emergence of a statistically significant difference between βL1, which is the ERR for the doses accumulated below the specified cut-point, and βH1, which is the ERR for that above the specified cut-point, was observed for cut-point of the dose rate of 4, 6, 34, 36, and 38 mSv y. While statistically negative values were estimated for βL1 (4 and 6 mSv y) and for βH1 (34, 36, and 38 mSv y), the overall relationship between the ERR and the cut-point of the dose rate was found to be similar to that obtained by the analysis of the mortality for all cancers excluding leukemia.

In the mid-1970s, American civil defense authorities became increasingly concerned with the potential threat of fission reactor accidents. Research from the Defense Civil Preparedness Agency began to be aimed not just at the ever-present threat of nuclear warfare but also peacetime emergencies as part of a "dual use" philosophy. The Brookhaven National Laboratory received funding to create a prototype radioiodine air sampling system, with multiple publications detailing the creation of the air sampler itself and a special CD V-700 survey meter that accompanied it. In late March 1979, the system found its first operational employment at the reactor accident at Three Mile Island in Pennsylvania. Despite successful use and further study, the BNL Air Sampler was not widely fielded by the federal government. However, lessons learned from its employment and development may be applicable to contemporary radiological emergency response.

A performance test was carried out using five different gamma-ray spectrometers applicable to thyroid monitoring. The energy resolution and efficiency for 7.6-cm × 7.6-cm NaI(Tl), 5.1-cm × 5.1-cm CeBr3, 3.8-cm × 3.8-cm SrI2(Eu), and 2.5-cm × 2.5-cm SrI2(Eu) gamma-ray scintillation spectrometers and a CdTe gamma-ray semiconductor spectrometer were evaluated using the Oak Ridge Institute for Nuclear Studies thyroid-neck phantom with a mock I source for prompt thyroid monitoring after a nuclear emergency. The respective energy resolutions of the full-energy peak for ~0.360 MeV of 3.8-cm × 3.8-cm SrI2(Eu) and 2.5-cm × 2.5-cm SrI2(Eu) scintillation spectrometers were 4.2% and 4.3%, and these values were very close to the value obtained by the CdTe semiconductor spectrometer. The efficiencies of four of the gamma-ray spectrometers were compared based on the efficiency of the 7.6-cm × 7.6-cm NaI(Tl) scintillation spectrometer at 0 cm from the phantom surface, and these values were ~70% for 5.1-cm × 5.1-cm CeBr3, ~30% for 3.8-cm × 3.8-cm SrI2(Eu), 10% for 2.5-cm × 2.5-cm SrI2(Eu) and 2% for 1-cm × 0.1-cm CdTe. Furthermore, the detection limits at various dose rates for the four gamma-ray scintillation spectrometers were evaluated using the method based on International Organization for Standardization publication ISO 11929:2010.

Radiation workers are normally exposed to doses resulting from their day-to-day activities. Besides that, background radiation, such as radon, can contribute to the exposure rates. The radionuclide 222Rn is a noble gas belonging to the uranium series, and its indoor concentration in the air depends on the exhalation from surrounding soil and the exhalation from building materials. Radon exhaling from porous building materials containing high uranium concentrations can become a significant exposure factor in areas with limited ventilation. The objective of this study was to evaluate the 222Rn concentrations in the radiochemistry and radiometric laboratories in the nuclear reactor building of the Nuclear Reactor Center (CERPq) located in the Institute of Nuclear and Energy Research (IPEN), Sao Paulo, Brazil. Measurements were done using a Radon Gas Monitor, model RAD7, equipped with a solid-state alpha detector. A passive method (SSNTD) was also used, consisting of square pieces of C-39 foils (2.5 cm × 2.5 cm) placed within small diffusion chambers. The CR-39 detectors were etched in KOH 30% solution at 80 °C for 5.5 h in a constant-temperature bath. After etching, the detectors were washed, dried, and scanned using a microscope to obtain the track density measurements. The activity concentrations measured with both techniques varied from 52 to 103 Bq m-3 in the studied areas of the CERPq. These values may be compared to the reference level of 100 Bq m-3 established by the World Health Organization to ensure safety environments.

A registry of plutonium-induced lung fibrosis diagnosed in members of a cohort of the first Russian nuclear industry facility Mayak Production Association was established. The registry includes 188 plutonium-induced lung fibrosis diagnoses: 117 (62.23%) in males and 71 (37.77%) in females. This paper describes the structure and detailed characteristics of the registry. Plutonium-induced lung fibrosis was shown to have no association with cumulative lung absorbed dose from external gamma rays as of the date of diagnosis. On the contrary, the plutonium-induced lung fibrosis rate was shown to be associated with cumulative lung absorbed dose from incorporated alpha particles and to increase significantly with increasing dose from internal radiation exposure. This paper discusses potential applications of the registry to scientific investigations in the future.

This study aimed to identify immunological biomarkers for prolonged occupational radiation exposure and thus studied a random sample of the Mayak Production Association worker cohort (91 individuals). The control group included 43 local individuals never employed at the Mayak Production Association. To identify biomarkers, two groups of workers were formed: the first one included workers chronically exposed to external gamma rays at cumulative doses of 0.5-3.0 Gy (14 individuals); the second one included workers exposed to combined radiation-external gamma rays at doses ranging from 0.7 to 5.1 Gy and internal alpha radiation from incorporated plutonium with 0.3-16.4 kBq body burden (77 individuals). The age range of the study individuals was 66-91 y. Peripheral blood serum protein concentrations of cytokines, immunoglobulins, and matrix metalloproteinase-9 were analyzed using enzyme-linked immunoassay following the manufacturer's protocol. Flow cytometry was used to analyze levels of various lymphocyte subpopulations. The findings of the current study demonstrate that some immunological characteristics may be considered as biomarkers of prolonged chronic radiation exposure for any radiation type (in the delayed period after the exposure) based on fold differences from controls: M immunoglobulin fold differences were 1.75 ± 0.27 (p = 0.0001) for external gamma-ray exposure and 1.50 ± 0.27 (p = 0.0003) for combined radiation exposure; matrix metalloproteinase-9 fold differences were 1.5 ± 0.22 (p = 0.008) for external gamma-ray exposure and 1.69 ± 0.24 (p = 0.00007) for combined radiation exposure; A immunoglobulin fold differences were 1.61 ± 0.27 (p = 0.002) for external gamma-ray exposure and 1.56 ± 0.27 (p = 0.00002) for combined radiation exposure; relative concentration of natural killer cell fold differences were 1.53 ± 0.23 (p = 0.01) for external gamma-ray exposure and 1.35 ± 0.22 (p = 0.001) for combined radiation exposure; and relative concentration of T-lymphocytes fold differences were 0.89 ± 0.04 (p = 0.01) for external gamma-ray exposure and 0.95 ± 0.05 (p = 0.003) for combined radiation exposure. Based on fold differences from controls, interferon-gamma (3.50 ± 0.65, p = 0.031), transforming growth factor-beta (2.91 ± 0.389, p = 0.026), and relative blood serum levels of T-helper cells (0.90 ± 0.065, p = 0.02) may be used as immunological markers of chronic external gamma-ray exposure. Moreover, there was a significant inverse linear association of relative concentration of T-helper cells with dose from external gamma rays accumulated over an extended period.

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 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. Evaluation and adjustment for individual cumulative measures of pitchblende dust inhalation were made for lung and kidney diseases.

Tritium that is released from nuclear facilities, especially from nuclear reactor units, may be the main origination of tritium in the environment. Atmospheric tritiated water (HTO) is the main chemical form of tritium that is released from nuclear reactor units. HTO in the air, drinking water and foods should be monitored routinely to protect public from the radiological hazards of tritium. Here, concentrations of HTO in the air, drinking water and foods in the vicinity of Qinshan Nuclear Power (QNP) base were measured from 2012 to 2016. And based on the concentrations of HTO measured, annual dose equivalent and collective dose equivalent for adults were calculated to evaluate the radiological hazards derived from HTO. The annual dose equivalent for adults in Qinshan and Wuyuan communities were both far below the dose limit for public exposure, and both showed an increasing trend, indicating that great attention still should be payed to the releasing of tritium from nuclear reactor units in the QNP base.

The U.S. National Oceanic and Atmospheric Administration's Field Research Division uses the HYRad-HYSPLIT dispersion model to assess hypothetical accidental releases of airborne radioactive materials at the Idaho National Laboratory in southeastern Idaho. The State of Idaho Department of Environmental Quality Idaho National Laboratory Oversight Program provides independent assessment of these releases using a different model, RASCAL, which is the U.S. Nuclear Regulatory Commission's primary reactor-emergency response code. To confirm RASCAL is a reasonable independent assessment tool, the Oversight Program compared the two models' output for typical meteorological cases encountered at the Idaho National Laboratory. RASCAL results were also compared to National Oceanic and Atmospheric Administration experimental SF6 tracer data from 2013 at the Idaho National Laboratory. Both RASCAL and HYRad predicted very similar plume shapes and paths for the different meteorological cases. For typical daytime conditions, HYRad predicted slightly higher integrated air concentrations by up to a factor of two at downwind distances of less than about 40 km, then decreased below the RASCAL concentrations. The opposite was true for a nighttime release, with RASCAL giving significantly higher concentrations (by one to two orders of magnitude) at a distance of 20 km. For all the runs, RASCAL predicted significantly less total deposition, except at the outer edges of the plume during a nighttime release. Most of the discrepancies are believed to be due to differences in the models' simulation algorithms and the hard-wired input parameter values used by each model (e.g., deposition velocities). For tracer data comparisons, RASCAL's straight-line Gaussian plume model calculated maximum 2 h predicted-to-observed concentration ratios of 0.8 to 1.8 for unstable conditions and 0.4 to 0.9 for neutral conditions.

The purpose of this paper is to present an overview of ongoing work on the Million Worker Study (MWS), highlighting some of the key methods and progress so far as exemplified by the study of workers at the Mallinckrodt Chemical Works (MCW). The MWS began nearly 25 y ago and continues in a stepwise fashion, evaluating one study cohort at a time. It includes workers from U.S. Department of Energy (DOE) Manhattan Project facilities, U.S. Nuclear Regulatory Commission (NRC) regulated nuclear power plants, industrial radiographers, U.S. Department of Defense (DoD) nuclear weapons test participants, and physicians and technologists working with medical radiation. The purpose is to fill the major gap in radiation protection and science: What is the risk when exposure is received gradually over time rather than briefly as for the atomic bomb survivors? Studies published or planned in 2018 include leukemia (and dosimetry) among atomic veterans, leukemia among nuclear power plant workers, mortality among workers at the MCW, and a comprehensive National Council on Radiation Protection and Measurements (NCRP) Report on dosimetry for the MWS. MCW has a singular place in history: the 40 tons (T) of uranium oxide produced at MCW were used by Enrico Fermi on 2 December 1942 to produce the first manmade sustained and controlled nuclear reaction, and the atomic age was born. Seventy-six years later, the authors followed the over 2,500 MCW workers for mortality and reconstructed dose from six sources of exposure: external gamma rays from the radioactive elements in pitchblende; medical x rays from occupationally required chest examinations; intakes of pitchblende (uranium, radium, and silica) measured by urine samples; radon breath analyses and dust surveys overseen by Robley Evans and Merril Eisenbud; occupational exposures received before and after employment at MCW; and cumulative radon concentrations and lung dose from the decay of radium in the work environment. The unique exposure reconstructions allow for multiple evaluations, including estimates of silica dust. The study results are relevant today. For example, NASA is interested that radium, deposited in the brain, releases high-LET alpha particles - the only human analogue, though limited, for high energy, high-Z particles (galactic cosmic rays) traveling through space that might affect astronauts on Mars missions. Don't discount the past; it's the prologue to the future!

The Savannah River National Laboratory (SRNL) with the assistance of Georgia Regents University, completed a comparison of the U.S. Environmental Protection Agency's (U.S. EPA) environmental dosimetry code CAP88 PC V3.0 with the recently developed V4.0. CAP88 is a set of computer programs and databases used for estimation of dose and risk from radionuclide emissions to air. At the U.S. Department of Energy's Savannah River Site, CAP88 is used by SRNL for determining compliance with U.S. EPA's National Emission Standards for Hazardous Air Pollutants (40 CFR 61, Subpart H) regulations. Using standardized input parameters, individual runs were conducted for each radionuclide within its corresponding database. Some radioactive decay constants, human usage parameters, and dose coefficients changed between the two versions, directly causing a proportional change in the total effective dose. A detailed summary for select radionuclides of concern at the Savannah River Site (60Co, 137Cs, 3H, 129I, 239Pu, and 90Sr) is provided. In general, the total effective doses will decrease for alpha/beta emitters because of reduced inhalation and ingestion rates in V4.0. However, for gamma emitters, such as 60Co and 137Cs, the total effective doses will increase because of changes U.S. EPA made in the external ground shine calculations.

The capabilities of indirect radiobioassay by urine and fecal sample analysis were compared with the direct radiobioassay methods of whole body counting and lung counting for the most common radionuclides and inhalation exposure scenarios encountered by Hanford workers. Radionuclides addressed by in vivo measurement included 137Cs, 60Co, 154Eu, and 241Am as an indicator for plutonium mixtures. The same radionuclides were addressed using gamma energy analysis of urine samples, augmented by radiochemistry and alpha spectrometry methods for plutonium in urine and fecal samples. It was concluded that in vivo whole body counting and lung counting capability should be maintained at the Hanford Site for the foreseeable future, however, urine and fecal sample analysis could provide adequate, though degraded, monitoring capability for workers as a short-term alternative, should in vivo capability be lost due to planned or unplanned circumstances.

Health physics is a recognized safety function in the holistic context of the protection of workers, members of the public, and the environment against the hazardous effects of ionizing radiation, often generically designated as radiation protection. The role of the health physicist as protector dates back to the Manhattan Project. Nuclear security is the prevention and detection of, and response to, criminal or intentional unauthorized acts involving or directed at nuclear material, other radioactive material, associated facilities, or associated activities. Its importance has become more visible and pronounced in the post 9/11 environment, and it has a shared purpose with health physics in the context of protection of workers, members of the public, and the environment. However, the duties and responsibilities of the health physicist in the nuclear security domain are neither clearly defined nor recognized, while a fundamental understanding of nuclear phenomena in general, nuclear or other radioactive material specifically, and the potential hazards related to them is required for threat assessment, protection, and risk management. Furthermore, given the unique skills and attributes of professional health physicists, it is argued that the role of the health physicist should encompass all aspects of nuclear security, ranging from input in the development to implementation and execution of an efficient and effective nuclear security regime. As such, health physicists should transcend their current typical role as consultants in nuclear security issues and become fully integrated and recognized experts in the nuclear security domain and decision making process. Issues regarding the security clearances of health physics personnel and the possibility of insider threats must be addressed in the same manner as for other trusted individuals; however, the net gain from recognizing and integrating health physics expertise in all levels of a nuclear security regime far outweighs any negative aspects. In fact, it can be argued that health physics is essential in achieving an integrated approach toward nuclear safety, security, and safeguards.

The current Boron Neutron Capture Therapy (BNCT) experiments performed at the University of Pavia, Italy, are focusing on the in vivo irradiations of small animals (rats and mice) in order to evaluate the effectiveness of BNCT in the treatment of diffused lung tumors. After the irradiation, the animals are manipulated, which requires an evaluation of the residual radioactivity induced by neutron activation and the relative radiological risk assessment to guarantee the radiation protection of the workers. The induced activity in the irradiated animals was measured by high-resolution open geometry gamma spectroscopy and compared with values obtained by Monte Carlo simulation. After an irradiation time of 15 min in a position where the in-air thermal flux is about 1.2 × 10(10) cm(-2) s(-1), the specific activity induced in the body of the animal is mainly due to 24Na, 38Cl, 42K, 56Mn, 27Mg and 49Ca; it is approximately 540 Bq g(-1) in the rat and around 2,050 Bq g(-1) in the mouse. During the irradiation, the animal body (except the lung region) is housed in a 95% enriched 6Li shield; the primary radioisotopes produced inside the shield by the neutron irradiation are 3H by the 6Li capture reaction and 18F by the reaction sequence 6Li(n,α)3H → 16O(t,n)18F. The specific activities of these products are 3.3 kBq g(-1) and 880 Bq g(-1), respectively.

Three different 90Y internally administered radionuclide therapies are currently used in both standard-of-care and clinical trial procedures atMD Anderson Cancer Center. TheraSphere and SIR-Spheres therapies utilize 90Y-labeled microspheres, while Zevalin is an 90Y-labeled radioimmunotherapeutic agent. Several publications have indicated radionuclidic impurities resulting from 90Y production methods. The 90Y in SIR-Spheres and Zevalin are produced from a 90Sr/90Y generator, which leaves measurable quantities of 90Sr in the final product. TheraSphere 90Y is produced in a nuclear reactor which results in a large number of impurities, most notably 88Y and 91Y. Product information sheets reference these impurities with specific limits given. These limits represent a tiny fraction of the total product activity, and in the case of TheraSphere and SIR-Spheres gamma-emitting impurities, this has been verified in the literature. An analysis of 90Sr impurities in SIR-Spheres and Zevalin is presented in this paper. Impurity quantities were found to be within the vendors’ documented limits.

The increased occupational doses resulting from the Chernobyl nuclear reactor accident that occurred in Ukraine in April 1986, the reactor accident of Fukushima that took place in Japan in March 2011, and the early operations of the Mayak Production Association in Russia in the 1940s and 1950s are presented and discussed. For comparison purposes, the occupational doses due to the other two major reactor accidents (Windscale in the United Kingdom in 1957 and Three Mile Island in the United States in 1979) and to the main plutonium-producing facility in the United States (Hanford Works) are also covered but in less detail. Both for the Chernobyl nuclear reactor accident and the routine operations at Mayak, the considerable efforts made to reconstruct individual doses from external irradiation to a large number of workers revealed that the recorded doses had been overestimated by a factor of about two.Introduction of Increased Occupational Exposures: Nuclear Industry Workers. (Video 1:32, http://links.lww.com/HP/A21).

The Savannah River National Laboratory's Environmental Dosimetry Group performs dosimetry assessments for Savannah River Site (SRS) radionuclide air emissions utilizing the Clean Air Act Assessment Package-1988 (CAP88) code (CAP88 PC Ver. 3.0) and the MAXDOSE-SR Ver. 2011 code, which is an SRS-specific version of the Nuclear Regulatory Commission's MAXIGASP code. CAP88 PC and MAXDOSE-SR are used at SRS for demonstrating compliance with Environmental Protection Agency dose standards for radionuclide emissions to the atmosphere and Department of Energy Order 458.1 dose standards, respectively. During a routine comparison of these two assessment models, it was discovered that CAP88 PC Ver. 3.0 was not producing the expected results when using multiple co-located stacks in a single run. Specifically, if the stack heights are considered separately, the results for several radionuclides (but not all) differ from the combined run [i.e., 1 + 2 does not equal (1+2)]. Additionally, when two or more stack heights are considered in a run, the results depend on the order of the selected stack heights. For example, for a two stack-height run of 0 meter and 61 m input produces different results from a 61 m and 0 m input run. This study presents a comparison of CAP88 PC Ver. 3.0 and MAXDOSE-SR Ver. 2011 based on SRS input data and on two-stack release scenarios. The selected radionuclides for this study included gases/vapors (H, C, Kr, and I) and particulates (Sr, Cs, Pu, and Am) commonly encountered at SRS.

Lawrence Berkeley National Laboratory developed a simple, efficient, and cost-effective replacement for the traditional glass column system used to monitor H and C emissions from rooftop stacks. The primary goals in developing a replacement (the modified jar system) were to 1) maintain or improve collection efficiency while keeping leakage to less than 5%, 2) simplify the set-up process, and 3) reduce costs. Both the traditional glass column assembly and the modified jar system were operated in tandem for a 13-mo period. Results showed that the modified sample jar system provided equivalent or improved collection efficiency for both H and C. Additional advantages included reduced leak-test errors, quicker and simpler set-up, and material costs that were reduced by nearly an order of magnitude.

Rapid measurement of radioactivity from 89,90Sr in aqueous solutions is performed using a technique combining a strontium rad disk and a picobeta spectrometer. Identification of 89,90sr radionuclides is accomplished in as little as 90min in a radiation-tainted solution that contains more highly radioactive cesium. It is possible to perform triage by assessing skin exposure doses in this short time. This simple technique could be used in mobile laboratories. 89,90Sr having 1 Bq radioactivities are measured in 10 kBq 137Cs in aqueous solution. The radioactivity contained in rinse water used to decontaminate the feet of workers who stepped into highly contaminated water in the basement of the turbine building of Unit 3 at the Fukushima Daiichi nuclear power station was measured. The amount of 90Sr radioactivity in rinse water using the authors'rapid measurement technique (0.29Bq mL-1)and a traditional method agree well, with 3.6% difference. Based on this agreement, this technique is confirmed to be useful for rapid measurement of 89,90Sr radioactivities.

Sandia National Laboratories' Z-Machine is the world's most powerful electrical device, and experiments have been conducted that make it the world's most powerful radiation source. Because Z-Machine is used for research, an assortment of materials can be placed into the machine; these materials can be subjected to a range of nuclear reactions, producing an assortment of activation products. A methodology was developed to provide a systematic approach to evaluate different materials to be introduced into the machine as wire arrays. This methodology is based on experiment specific characteristics, physical characteristics of specific radionuclides, and experience with Z-Machine. This provides a starting point for bounding calculations of radionuclide source-term that can be used for work planning, development of work controls, and evaluating materials for introduction into the machine.

For emergency preparedness and response with respect to nuclear power plant accidents, the concept of Emergency Care Centers has been developed in Germany. This setup aims at monitoring contamination, to decontaminate if needed, assess the dose, and perform an initial medical evaluation of people who might have been affected by the accident. The concept has been tested in many exercises. In response to a terrorist attack involving a dirty bomb, this concept may prove useful for attending contaminated people who are not severely injured.

The overall objective of this study was to demonstrate that a new technology, known as RadBall®, could locate submerged radiological hazards. RadBall® is a novel, passive, radiation detection device that provides a 3-D visualization of radiation from areas where measurements have not been previously possible due to lack of access or extremely high radiation doses. This technology has been under development during recent years, and all of its previous tests have included dry deployments. This study involved, for the first time, underwater RadBall® deployments in hot cells containing 137CsCl capsules at the U.S. Department of Energy's Hanford Site. RadBall® can be used to characterize a contaminated room, hot cell, or glovebox by providing the locations of the radiation sources and hazards, identifying the radionuclides present within the cell, and determining the radiation sources' strength (e.g., intensities or dose rates). These parameters have been previously determined for dry deployments; however, only the location of radiation sources and hazards can be determined for an underwater RadBall® deployment. The results from this study include 3-D images representing the location of the radiation sources within the Hanford Site cells. Due to RadBall®'s unique deployability and non-electrical nature, this technology shows significant promise for future characterization of radiation hazards prior to and during the decommissioning of contaminated nuclear facilities.

The Mayak Production Association was the first site for the production of weapons-grade plutonium in Russia. Early operations led to the waterborne release of radioactive materials into the small Techa River. Residents living downstream used river water for drinking and other purposes. The releases and subsequent flooding resulted in deposition of sediments along the shoreline and on floodplain soil. Primary routes of exposure were external dose from the deposited sediments and ingestion of 90Sr and other radionuclides. Study of the Techa River Cohort has revealed an increased incidence of leukemia and solid cancers. Epidemiologic studies are supported by extensive dose-reconstruction activities that have led to various versions of a Techa River Dosimetry System (TRDS). The correctness of the TRDS has been challenged by the allegation that releases of short-lived radionuclides were much larger than those used in the TRDS. Although the dosimetry system depends more upon measurements of 90Sr in humans and additional measurements of radionuclides and of exposure rates in the environment, a major activity has been undertaken to define more precisely the time-dependent rates of release and their radionuclide composition. The major releases occurred during 1950-1951 in the form of routine releases and major accidental releases. The reevaluated amount of total release is 114 PBq, about half of which was from accidents that occurred in late 1951. The time-dependent composition of the radionuclides released has also been reevaluated. The improved understanding presented in this paper is possible because of access to many documents not previously available.

Bioassay for individual radionuclides is an essential and first step in estimation of radiation risk to nuclear facilities workers and people who are exposed to the contaminated environment in the event of a nuclear accident or radiological attack. Urine is a frequently used biological sample for this purpose. Tritium and (14)C are important radionuclides for workers in nuclear reactors and radiopharmaceutical laboratories. A method for the determination of tritium and (14)C in organic and inorganic forms in urine has been developed. It involves activated charcoal absorption of organic matter followed by combustion to separate tritiated water from organically-bound tritium. Inorganic (14)C from organically-bound (14)C, the separated tritium and (14)C were measured using liquid scintillation counting. Iodine-129, a long-lived beta emitter, is normally released to the atmosphere during the operation of nuclear facilities, especially in reprocessing plants. The high concentration of iodine in the thyroid makes this radionuclide an important source of exposure to exposed populations. A simple method has been developed in this work for the determination of (129)I in urine by anion exchange preconcentration, extraction purification and liquid scintillation counting. Using accelerator mass spectrometry, urine samples can be analyzed for low level (129)I in both organic and inorganic forms after active charcoal adsorption and solvent extraction separation. Condensed water collected daily from the reactor hall in a Danish research reactor and monthly urine samples from the staff working in the reactor building were collected from 2003-2010 and analyzed using this method, and the results are presented and discussed.

This paper discusses the current status of the ongoing Advanced Fuel Cycle Initiative (AFCI) program in the U.S. Department of Energy that is investigating the potential for using the processing and recycling of used nuclear fuel to improve radioactive waste management, including used fuel. A key element of the strategies is to use nuclear reactors for further irradiation of recovered chemical elements to transmute certain long-lived highly-radioactive isotopes into less hazardous isotopes. Both thermal and fast neutron spectrum reactors are being studied as part of integrated nuclear energy systems where separations, transmutation, and disposal are considered. Radiotoxicity is being used as one of the metrics for estimating the hazard of used fuel and the processing of wastes resulting from separations and recycle-fuel fabrication. Decay heat from the used fuel and/or wastes destined for disposal is used as a metric for use of a geologic repository. Results to date indicate that the most promising options appear to be those using fast reactors in a repeated recycle mode to limit buildup of higher actinides, since the transuranic elements are a key contributor to the radiotoxicity and decay heat. Using such an approach, there could be much lower environmental impact from the high-level waste as compared to direct disposal of the used fuel, but there would likely be greater generation of low-level wastes that will also require disposal. An additional potential waste management benefit is having the ability to tailor waste forms and contents to one or more targeted disposal environments (i.e., to be able to put waste in environments best-suited for the waste contents and forms).

The purpose of this study was to obtain quantitative data on plutonium microdistribution in different structural elements of human bone tissue for local dose assessment and dosimetric models validation. A sample of the thoracic vertebra was obtained from a former Mayak worker with a rather high plutonium burden. Additional information was obtained on occupational and exposure history, medical history, and measured plutonium content in organs. Plutonium was detected in bone sections from its fission tracks in polycarbonate film using neutron-induced autoradiography. Quantitative analysis of randomly selected microscopic fields on one of the autoradiographs was performed. Data included fission fragment tracks in different bone tissue and surface areas. Quantitative information on plutonium microdistribution in human bone tissue was obtained for the first time. From these data, the quantitative relationships of plutonium decays in bone volume to decays on bone surface in cortical and trabecular fractions were defined as 2.0 and 0.4, correspondingly. The measured quantitative relationship of decays in bone volume to decays on bone surface does not coincide with recommended models for the cortical bone fraction by the International Commission on Radiological Protection. Biokinetic model parameters of extrapulmonary compartments might need to be adjusted after expansion of the data set on quantitative plutonium microdistribution in other bone types in humans as well as other cases with different exposure patterns and types of plutonium.

A 1985 plutonium puncture wound resulted in the initial deposition of 48 kBq of transuranic alpha activity, primarily 239Pu and 241Am, in a worker’s right index finger. Surgical excisions in the week following reduced the long-term residual wound activity to 5.4 kBq, and 164 DTPA chelation therapy administrations over a 17-month period resulted in urinary excretion of about 7 kBq. The case was published in 1988, but now 20 additional years of follow-up data are available. Annual bioassay measurements have included wound counts, skeleton counts, liver counts, lung counts, axillary lymph node counts, and urinalyses for plutonium and 241Am. These measurements have shown relatively stable levels of 241Am at the wound site, with gradually increasing amounts of 241Am detected in the skeleton. Liver counts has shown erratic detection of 241Am, and lung counts indicate 241Am as shine from the axillary lymph nodes and skeleton. Urine excretion of 239Pu since termination of chelation therapy has typically ranged from 10 to 20 mBq d-1, with 241Am excretion being about 10% of that for 239Pu. In addition, the worker has undergone annual routine medical exams, which have not identified any adverse health effects associated with the intake.

Neutron and photon radiation survey records have been used to evaluate and develop a neutron to photon (NP) ratio to reconstruct neutron doses to workers around Hanford's single pass reactors that operated from 1945 to 1972. A total of 5,773 paired neutron and photon measurements extracted from 57 boxes of survey records were used in the development of the NP ratio. The development of the NP ratio enables the use of the recorded dose from an individual's photon dosimeter badge to be used to estimate the unmonitored neutron dose. The Pearson rank correlation between the neutron and photon measurements was 0.71. The NP ratio best fit a lognormal distribution with a geometric mean (GM) of 0.8, a geometric standard deviation (GSD) of 2.95, and the upper 95 th % of this distribution was 4.75. An estimate of the neutron dose based on this NP ratio is considered bounding due to evidence that up to 70% of the total photon exposure received by workers around the single pass reactors occurs during shutdown maintenance and refueling activities when there is no significant neutron exposure. Thus when this NP ratio is applied to the total measured photon dose from an individual film badge dosimeter, the resulting neutron dose is considered bounded.

After nearly three decades of medical product sterilization, 3M launched a major new project to build and license an irradiator facility. 3M Corporate Health Physics was responsible for the licensing aspect of this project. The licensing process consisted of six amendments, over 30 submissions to the U.S. Nuclear Regulatory Commission (U.S. NRC) and four U.S. NRC site visits. It took approximately 22 months to complete. The six license amendments are reviewed and several of the submissions are discussed. These include 3M's response to the U.S. NRC's interest in the shielding calculations used for the bioshield, the development of a protocol of radiation safety system test methods, and an analysis to show that a dropped cask during loading operations would not fall on sealed sources. A number of lessons were learned during the course of licensing the new irradiator. Among these were the importance of understanding the U.S. NRC license reviewer's perspective, the need to thoroughly review the irradiator manufacturer's licensing package during project negotiations, the benefits of leaving the Health Physics Office and meeting with the non-health physicists involved in the project, and the necessity of maintaining the solid relationships that already existed with the site Radiation Safety Officer and Sterilization Engineer.

In the course of the decommissioning of the ASTRA research reactor at the site of the Austrian Research Centers at Seibersdorf, the operator and licensee, Nuclear Engineering Seibersdorf, conducted an extensive site survey and characterization to demonstrate compliance with regulatory site release criteria. This survey included radiological characterization of approximately 400,000 m(2) of open land on the Austrian Research Centers premises. Part of this survey was conducted using a mobile large-area gas proportional counter, continuously recording measurements while it was moved at a speed of 0.5 ms(-1). In order to set reasonable investigation levels, two alarm levels based on statistical considerations were developed. This paper describes the derivation of these alarm levels and the operational experience gained by detector deployment in the field.