Assessment of soil-soil solution distribution coefficients of global fallout 237Np and 239Pu in Japanese upland soils.

Neptunium-237 and 239Pu are important radionuclides in the safety assessment related to geological disposal of radioactive waste because of the possibility of long-term exposure to humans. Mobilities of these radionuclides in the environment are of particular importance for their radiation dose evaluation; therefore, in this study, we have made the assessment of the soil-soil solution distribution coefficient (Kd, L/kg) using global fallout 237Np and 239Pu in Japanese upland soils. The Kd values were determined by extracting these radionuclides from 24 soil samples using a laboratory batch method. The desorption Kd values of 237Np ranged from 3.3 × 102 to 1.0 × 104 L/kg, and their geometric mean (GM) and arithmetic mean (AM) were 1.7 × 103 L/kg and 2.6 × 103 L/kg, respectively. The desorption Kd values of 239Pu were found to vary from 9.4 × 103 to 7.1 × 104 L/kg, and their GM and AM were 3.3 × 104 L/kg and 4.0 × 104 L/kg, respectively. In Japanese upland soils, the Kd value of 239Pu was one order of magnitude higher than that of 237Np.

Development of a method of evaluating PuO2 particle diameters using an alpha-particle imaging detector.

It is very important to evaluate the diameters (activity median aerodynamic diameter) of plutonium dioxide (PuO2) particles for internal exposure dose evaluation. In this study, a method of evaluating PuO2 particle diameters using an alpha-particle imaging detector was developed. PuO2 particles with different diameters were modeled by Monte Carlo simulation, and the change in the shape of the energy spectrum for each particle diameter was evaluated. Two different patterns were modeled, namely, the case of 239PuO2 and the case of PuO2 (including isotopic composition of Pu). Multiple regression analysis was performed to determine the PuO2 particle diameter from the obtained parameters. The simulated diameters and the diameters obtained with the regression model were in good agreement. The advantage of using the alpha-particle imaging detector is to measure the alpha energy spectrum for individual particle, and this allows accurate measurement of particle diameter distribution.

Preliminary safety assessment for planning near surface disposal of low-level radioactive waste in Greece.

This paper presents the approach, assumptions, and computational analysis of the preliminary safety assessment, regarding the post-closure period for the disposal of radioactive waste in Greece. The assessment was implemented in the context of the National Program for the disposal of radioactive waste in the country, which is currently in the early stage for the investigation of facility siting. The basis scenario selected for this investigation was the leaching of radionuclides and the exposure in a residence offsite. Moreover, a scenario involving the intrusion in the facility and construction of a dwelling that disturbs the disposal zone is also considered. Due to the significant uncertainties in the current phase, the simulations related to leaching of the waste both in the offsite and intrusion scenarios are based on an uncertainty analysis with 25 site and scenario related parameters. The most important contribution is attributed to Ra-226 with an annual dose equal to about 2 and 3 µSv per MBq disposed, for the offsite and intrusion scenario, respectively. Th-232, Cl-36, C-14, Ag-108m and Pu-239 follow with a dose one order of magnitude less, compared to Ra-226. In the leaching scenarios examined, and for the most dose relevant radionuclides, the exposure related to drinking water from the well and the use of this water to irrigate fruits and vegetables are by far the dominant pathways due to the environmental transfer of the radionuclides and their associated dose coefficients. Th-232 dominates the direct exposure pathways (direct external radiation and plant contamination from the contaminated surface soil) in the intrusion scenario with an annual dose of about 1.4 mSv per Bq/g disposed. Ra-226, Cl-36 and Ag-108m cause exposure levels higher than 0.2 mSv/y per Bq/g disposed in the facility. A wide range was considered for the uncertainty parameters that led to a significant variation in the predicted doses that is expected to envelope the potential exposure for each radionuclide.

Dose Assessment Following a 238 Pu Inhalation Incident at Los Alamos National Laboratory.

ABSTRACT: A glovebox breach at the plutonium facility at Los Alamos National Laboratory potentially exposed 15 individuals to 238 Pu aerosols. One of the individuals (P0) received two 1-g intravenous DTPA treatments, one on the day of the intake and another the following day. Several urine samples were collected from the individuals involved in the incident. Particle size analysis on the PPE and solubility analysis of the particles on a filter sample were conducted in vitro. The applicability of the results from the in vitro studies for dose assessment was questionable because of the effect of the cloth mask the workers were wearing for COVID-related protection. Based on several considerations, including the effect of cloth masks on the "effective" particle size inhaled and the analysis of fecal-to-urine ratio, the default Type M 1 µm AMAD model was used to estimate intakes and doses. Using the urinary excretion data collected after 100 d post last chelation treatment, the committed effective dose, E(50), for P0 was calculated to be 5.2 mSv. For all others, the bioassay data were consistent with no intakes or very small intakes [corresponding to E(50) less than 0.1 mSv].

Assessment and clinical management of internal contamination.

Internal contamination by radionuclides may occur through inhalation, ingestion and absorption through the skin or subcutaneous tissue. The clinical management of internalized radionuclides requires the integration of clinical signs and symptoms with dose estimates in biological tissues obtained from the face, nose, sputum, urine, faeces and/or skin. The assessment of ingested radionuclides includes bioassays of urine and faeces, and if available, whole body counting for radionuclides that emit penetrating x-rays or gamma-rays. An estimate of intake dose may be made at the time of initial patient evaluation by measuring radioactivity, converting counts/minute to depositions/minute with a specific gamma-ray constant, and comparing the amount to its annual limit on intake, clinical decision guide or derived reference level. Since nobody dies from internal contamination per se, medically unstable patients should be stabilized before addressing internal contamination. Whenever possible, internal contaminants should be physically removed as soon as possible after exposure. For inhaled internal contaminants, radionuclide-specific therapy may include the administration of an ion exchange resin (i.e. Prussian blue, PB) or chelating agent (i.e. diethylenetriamine pentaacetate, DTPA, that binds toradioactiveplutonium, americium, and curium), or the physical removal of insoluble particles with a high activity radionuclide (192Ir,90Sr,210Po) by bronchioalveolar lavage. Decorporation with PB, DTPA and other agents is used to enhance excretion. The treatment of wounds contaminated with an actinide includes gentle irrigation, surgical excision of contaminated tissue and DTPA. The averted dose (i.e. the total effective dose averted by therapy) may be calculated for each exposure route.

Plutonium Systemic Biokinetic Model for Rats.

A baseline compartmental model (relative to modeling decorporation) of the distribution and retention of plutonium (Pu) in the rat for a systemic intake is derived. The model is derived from data obtained from a study designed to evaluate the behavior of plutonium in the first 28 days after incorporation. The model is based on a recently published model of americium (Am) in rats, which incorporated a pharmacokinetic (PK)-front-end modeling approach, which was used to specify transfer to and from the extracellular fluids (ECF) in the various tissues in terms of vascular flow and volumes of ECF. In the americium model, the approach was "cell-membrane limited," meaning that rapid diffusion of americium occurred throughout all the extracellular fluids (i.e., the blood plasma and interstitial fluids), while back-end rates representing transport into and out of the cells were determined empirically. However, this approach was inconsistent with the plutonium dataset. A good fit to the data is obtained by incorporating aspects of the Durbin et al. model structure, with plutonium in plasma separated into "free" and "bound" components. Free plutonium uses a cell-membrane-limited front end as for americium. Bound plutonium uses a capillary-wall-limited front end, where transfer rates from blood plasma into the interstitial fluids are relatively slow, and must be determined either empirically or from a priori knowledge. As in the Durbin et al. model, both free and bound plutonium are available for deposition in bone. In addition, our model has some bound plutonium associated with uptake to the gastrointestinal (GI) tract. Uncertainties in transfer rates were investigated using Markov Chain Monte Carlo (MCMC). It is anticipated that this model structure of plutonium will also be useful in interpreting comparable data from decorporation studies done in experimental animals.

Retrieval and Characterization of a Plutonium-beryllium Source at the University of Utah.

ABSTRACT: A shielded neutron irradiation facility has been constructed to contain a 74 GBq plutonium-beryllium neutron source at the University of Utah. Due to lack of information and recent use, the GEANT4 Monte Carlo Toolkit was used to perform dose assessments and validation of the designed shield and beam port for storage and operational geometries where the source was to be placed. Borated polyethylene was used as the primary shielding material with a lead-lined beam port to collimate neutrons and reduce gamma rays within the active beam profile. The lead lining was found to reduce the gamma-to-neutron ratio within the profile to 0.259 ± 0.001 and maintain an average neutron flux of 4,633.7 ± 10.2 n cm-2 s-1. The measured dose rate of the source at 30 cm from its constructed storage configuration was found to be 22.3 ± 1.1 µSv h-1 and in good agreement with simulated values. This was also close to the target rate of 20 µSv h-1, the public limit for whole-body dose accumulation, imposed on the design. Measurements of the source during its retrieval and placement into the shield show good agreement with the simulated results and provide a strong validation for the future use of the source in experiments at the University of Utah.

Estimation of plutonium 241Pu budget in the Gulf of Gdansk and the Gdansk Basin (the southern Baltic Sea).

Presented have been the results of inventory estimation of anthropogenic beta-emitting 241Pu, in different components in the southern Baltic Sea ecosystem. The total 241Pu activity present in the Gulf of Gdansk and the Gdansk Basin for 2021 was estimated at 1.99 TBq and 7.82 TBq, respectively, of which 1.59 TBq and 6.38 TBq in the sediments. The Vistula and the Neman River and the atmospheric fallout were distinguished as the primary sources of plutonium in these basins. In seawater of the Gulf of Gdansk, there was about 0.40 TBq 241Pu (20.1% of total activity) and 1.44 TBq 241Pu (18.4% of total activity) in the Gdansk Basin. The 241Pu accumulated in living organisms in 2021 was 1.13 GBq in the Gulf of Gdansk and 3.96 GBq in the Gdansk Basin. In the Gulf of Gdansk biota, 55.7% was accumulated in zoobenthos, while in the Gdansk Basin, 48.3% was in zoobenthos.

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.

Novel DGT Configurations for the Assessment of Bioavailable Plutonium, Americium, and Uranium in Marine and Freshwater Environments.

Plutonium, americium, and uranium contribute to the radioactive contamination of the environment and are risk factors for elevated radiation exposure via ingestion through food or water. Due to the significant environmental inventory of these radioelements, a sampling method to accurately monitor their bioavailable concentrations in natural waters is necessary, especially since physicochemical factors can cause significant temporal fluctuations in their waterborne concentrations. To this end, we engineered novel diffusive gradients in thin-film (DGT) configurations using resin gels, which are selective for UO22+, Pu(IV + V), and Am(III) among an excess of extraneous cations. In this work, we also report an improved synthesis of our in-house ion-imprinted polymer resin, which we used to manufacture a resin gel to capture Am(III). The effective diffusion coefficients of Pu, Am, and U in agarose cross-linked polyacrylamide were determined in freshwater and seawater simulants and in natural seawater, to calibrate these configurations for environmental deployments.

Assessment of residual radionuclide levels at the Bokak and Bikar Atolls in the northern Marshall Islands.

The Bikar and Bokak Atolls, located in the northern Marshall Islands, are extremely isolated and consist of pristine marine and terrestrial ecosystems. Both atolls may have experienced significant radioactive deposition following the nuclear weapon testing conducted at Bikini and Enewetak proving grounds. Here we report activity concentrations of artificial radionuclides (239Pu, 240Pu, 241Pu, 241Am, 137Cs and 90Sr) in marine and terrestrial samples collected from Bikar and Bokak Atolls. Artificial radionuclides in soil from the Majuro Atoll are also reported and form a radiological baseline against which the levels at the other atolls can be compared. We observed low levels of artificial radionuclides in soil from Majuro and Bokak, but significantly higher levels in soil from Bikar. The residual radioactivity in the Bikar environment is comparable to the levels previously reported for other nearby atolls, including Taka and Utrik, but lower than for Rongerik, Rongelap, Bikini and Enewetak. An analysis of 240Pu/239Pu isotope ratios and estimations of the dates of contamination from 241Am/241Pu activity ratios both indicated that the Bikar Atoll was contaminated mainly by radioactive fallout from the Castle Bravo test in 1954. We compare the results of our measurements at Bikar and Bokak to data from other atolls in the Marshall Islands and to regions of the world affected by both global and regional fallout from atmospheric nuclear weapons testing and nuclear accidents.

Plutonium in coral archives: A good primary marker for an Anthropocene type section.

While we officially live in the Holocene epoch, global warming and many other impacts of global change have led to the proposal and wide adoption of the Anthropocene to define the present geological epoch. The Anthropocene Working Group (AWG) established that it should be treated as a formal stratigraphic unit, demonstrated by a reference level commonly known as "golden spike", still under discussion. Here we show that the onset of bomb-derived plutonium recorded in two banded massive corals from the Caribbean Sea is consistent (1955-1956 CE), so sites far from nuclear testing grounds are potentially suitable to host a type section of the Anthropocene. Coastal coral demonstration sites are feasible, could foster economic development, and may serve as focal points for scientific dissemination and environmental education.

Self-powered multilayer radioisotope identification device.

PURPOSE: This is a computational study to develop a rugged self-powered Radioisotope Identification Device (RIID). The principle of operation relies on the High Energy Current (HEC) concept (Zygmanski and Sajo, Med Phys. 43 4-15, 2016) with measurement of fast electron currents between low-Z and high-Z thin-film electrodes separated by nanoporous aerogel films in a multilayer detector structure whose prototypes were previously investigated (Brivio, Albert, Freund, Gagne, Sajo and Zygmanski, Med Phys, 46 4233-4240, 2019), (Brivio, Albert, Gagne, Freund, Sajo and Zygmanski, J Phys D Appl Phys, 53 265303, 2020). Here, we present an optimal detector design that accounts for a wide energy range (keV-MeV) of x-ray-emitting radioisotopes that are of interest to national security and radiation therapy. MATERIALS: We studied numerous multilayer detector geometries with N = 1..24 basic detector elements composed of 3 electrodes: N x (Al-aerogel-Ta-aerogel-Al). The thicknesses of electrodes and their total number were varied depending on the incident x-ray spectra and its ability to penetrate and interact with the different layers, producing fast electrons. We used radiation transport simulations to find a balanced geometry that accounts for all energies from 10 keV to 6 MeV in a single design with relatively few detector elements (N = 24). In the balanced design, the electrodes have increasing thickness as a function of depth in the detector, ranging from 0.5 µm-Ta and 10 µm-Al at the entrance to 10 mm-Ta and 2.5 mm-Al at the exit. Aerogel thickness was fixed at 50 µm. Electron currents forming RIID signals were acquired from all Ta electrodes. A model function M(x, Ei ) representing the detector yield as a function of the cumulative Ta thickness (x) for 70 monoenergetic incident beams (E) was derived. We also investigated the detector response to selected radioactive isotopes (Pd-103, I-125, Pu-239, U-235, Ir-192, Cs-137, Co-60). Additional studies were performed with Bremsstrahlung spectra produced by electron beams in kVp tubes and in MV Linacs used in radiology and radiation therapy departments. We investigated different algorithms for radioisotope identification that would work for unknown unshielded as well as shielded sources. RESULTS: Characteristic features of response functions for monoenergetic beams and radioisotopes were determined and used to develop two inverse algorithms of radioisotope identification. Using these algorithms, we were able to identify the unshielded and shielded sources, quantify the minimum, mean and maximum effective energies of the shielded spectra, and estimate the amount of Compton background in the spectrum. CONCLUSIONS: A multilayer sensor based on fast electron current was optimized and studied in its abilities as RIID. A balanced design permits the identification of radioisotopes with of a wide range of keV-MeV energies. The device is low cost, rugged, self-powered and can withstand very high dose rates, allowing deployment in difficult conditions, including radiation incidents. The algorithm we developed for radioisotope identification and spectral unfolding is robust and it is an important component in practical applications.

Dose Assessment Following a 238Pu-contaminated Wound Case with Chelation and Excision.

The urinary excretion and wound retention data collected after a Pu-contaminated wound were analyzed using Markov Chain Monte Carlo (MCMC) to obtain the posterior distribution of the intakes and doses. An empirical approach was used to model the effects of medical treatments (chelation and excision) on the reduction of doses. It was calculated that DTPA enhanced the urinary excretion, on average, by a factor of 17. The empirical analysis also allowed calculation of the efficacies of the medical treatments-excision and chelation averted approximately 76% and 5.5%, respectively, of the doses that would have been if there were no medical treatment. All bioassay data are provided in the appendix for independent analysis and to facilitate the compartmental modeling approaches being developed by the health physics community.

Development of a New Chelation Model: Bioassay Data Interpretation and Dose Assessment after Plutonium Intake via Wound and Treatment with DTPA.

The administration of chelation therapy to treat significant intakes of actinides, such as plutonium, affects the actinide's normal biokinetics. In particular, it enhances the actinide's rate of excretion, such that the standard biokinetic models cannot be applied directly to the chelation-affected bioassay data in order to estimate the intake and assess the radiation dose. The present study proposes a new chelation model that can be applied to the chelation-affected bioassay data after plutonium intake via wound and treatment with DTPA. In the proposed model, chelation is assumed to occur in the blood, liver, and parts of the skeleton. Ten datasets, consisting of measurements of C-DTPA, Pu, and Pu involving humans given radiolabeled DTPA and humans occupationally exposed to plutonium via wound and treated with chelation therapy, were used for model development. The combined dataset consisted of daily and cumulative excretion (urine and feces), wound counts, measurements of excised tissue, blood, and post-mortem tissue analyses of liver and skeleton. The combined data were simultaneously fit using the chelation model linked with a plutonium systemic model, which was linked to an ad hoc wound model. The proposed chelation model was used for dose assessment of the wound cases used in this study.

Chelation Modeling: The Use of Ad Hoc Models and Approaches to Overcome a Dose Assessment Challenge.

Chelating agents are administered to treat significant intakes of radioactive elements such as plutonium, americium, and curium. These drugs may be used as a medical countermeasure after radiological accidents and terrorist acts. The administration of a chelating agent, such as Ca-DTPA or Zn-DTPA, affects the actinide's normal biokinetics. It enhances the actinide's rate of excretion, posing a dose assessment challenge. Thus, the standard biokinetic models cannot be directly applied to the chelation-affected bioassay data in order to assess the radiation dose. The present study reviews the scientific literature, from the early 1970s until the present, on the different studies that focused on developing new chelation models and/or modeling of bioassay data affected by chelation treatment. Although scientific progress has been achieved, there is currently no consensus chelation model available, even after almost 50 y of research. This review acknowledges the efforts made by different research groups, highlighting the different methodology used in some of these studies. Finally, this study puts into perspective where we were, where we are, and where we are heading in regards to chelation modeling.

Thermal Neutron Characterization and Dose Modeling of a 239PuBe Alpha-Neutron Source.

Determination of neutron dose can be challenging and requires knowledge of neutron flux as a function of energy. The goal of this project was to characterize the thermal neutron flux of a 37 GBq PuBe alpha-neutron source and model the associated neutron dose using version MCNPX of the Monte-Carlo N-Particle transport codes. The PuBe source was placed in a neutron howitzer, and foil activation (dysprosium foils with and without cadmium covers) was used at various distances to determine thermal neutron flux, which was then used to verify the MCNPX model representing the system. The model was then adapted for dosimetric modeling to enable future neutron dose-response studies.

The Influence of Aerosol Density and Shape Factor on the Assessment of Internal Exposure to 239Pu.

Internal exposure due to inhalation of aerosols depends on the ratio of aerodynamic shape factor (χ) to aerosol mass density (ρ). Inhaled aerosol parameters may differ from the default ρ and χ values provided by the International Commission on Radiological Protection, which are adopted for the assessment of internal exposures. This paper focuses on the influences of χ/ρ on the assessment of internal exposure to Pu for reference workers. Regional deposition fractions are found to decrease with increasing χ/ρ, and larger decreases are observed with smaller activity median aerodynamic diameter aerosols, while the slow clearance fractions (fs) in the tracheobronchial region are more sensitive for larger activity median aerodynamic diameter aerosols. Results from biokinetics calculations reveal that both the time-dependent content (excretion) and cumulative activities are determined mainly for particles initially deposited in the alveolar-interstitial region, while fs affects the local cumulative activities in the tracheobronchial region. χ/ρ is proven to have different influences for aerosols with different activity median aerodynamic diameters. The default χ/ρ values can be used when activity median aerodynamic diameters are greater than 1 µm, while one should pay attention to the value of χ/ρ when activity median aerodynamic diameters are less than 1 µm, where significant influence may be anticipated.

History of Dose, Risk, and Compensation Assessments for US Veterans of the 1966 Plutonium Cleanup in Palomares, Spain.

In 1966, about 1,600 US military men-mostly Air Force-participated in a cleanup of plutonium dispersed from two nuclear bombs in Palomares, Spain. As a base for future analyses, we provide a history of the Palomares incident, including the dosimetry and risk analyses carried out to date and the compensation assessments made for veterans. By law, compensation for illnesses attributed to ionizing radiation is based on maximum estimated doses and standard risk coefficients, with considerable benefit of the doubt given to claimants when there is uncertainty. In the Palomares case, alpha activity in urine fell far faster than predicted by plutonium biokinetic excretion models used at the time. Most of the measurements were taken on-site but were disqualified on the grounds that they were "unreasonably high" and because there was a possibility of environmental contamination. Until the end of 2013, the Air Force used low dose estimates derived from environmental measurements carried out well after the cleanup. After these estimates were questioned by Congress, the Air Force adopted higher dose estimates based on plutonium concentration measurements in urine samples collected from 26 veterans after they left Palomares. The Air Force assumed that all other cleanup veterans received lower doses and therefore assigned to them maximum organ doses based on the individual among the 26 with the lowest urine measurements. These resulting maximum organ doses appear to be sufficient to justify compensation to all Palomares veterans with lung and bone cancer and early-onset liver cancer and leukemia but not other radiogenic cancers.

Construction, Validation and Sensitivity Analyses of a Job Exposure Matrix for Early Plutonium Workers at the Sellafield Nuclear Site, United Kingdom.

Plutonium is a radiologically significant alpha-particle emitter. The potential for adverse health effects from internal exposures due to plutonium intakes has been recognized since the 1940s. The workforce of the Sellafield nuclear facility (Cumbria, UK), includes one of the world's most important groups of plutonium-exposed workers for studying the potential health risks of this internal exposure. However, for several hundred workers employed at the start of plutonium work at the facility (1952-1963), historical monitoring records based on measurements of urinary excretion of plutonium are not sufficiently reliable to provide the accurate and unbiased exposure assessments needed for epidemiological studies. Consequently, these early workers have had to be excluded from such studies, significantly reducing their power. We constructed a population-specific quantitative job exposure matrix (JEM) to estimate the average intakes of "typical plutonium workers" in this period, from 1952-1963, and assessed its validity and sensitivity to exposure assessment decisions. We conducted internal cross-validation using an a priori 10% extracted sample to evaluate reliability of estimates, explored JEM sensitivity to assumptions in the exposure assessment, and assessed the impact of uncertainty in urinalysis measurements on the precision of annual intake estimates using Markov Chain Monte Carlo (MCMC) methodology. Pairwise correlations ( RP) of estimated (JEM) and measured (10% sample) annual intakes were moderate to high ( RP > 0.4) for 10 out of 13 JEM groups, while absolute differences were <20% for 11 out of 13 JEM groups. There was little evidence of a temporal trend in correlations ( P = 0.13) or absolute differences ( P = 0.34). The median JEM-derived cumulative intake of 95.2 (IQR, 55.0-130.0) Bq was comparable to those based on alternative assumptions in the exposure assessment (median range, 95.2-100.0 Bq; 75th percentiles, 130.0-146.0 Bq). Measurement error simulation resulted in a 40-60% reduced median cumulative intake but higher maximum cumulative intakes. The JEM finds a balance between reliability and precision that makes it useful for epidemiological purposes and is relatively insensitive to specific choices in the exposure assessment. This JEM will allow the inclusion of workers with longest follow-up and who could not be included up until now in epidemiological studies without introducing significant bias.