Dosimetric and microdosimetric analyses for blood exposed to reactor-derived thermal neutrons.

Thermal neutrons are found in reactor, radiotherapy, aircraft, and space environments. The purpose of this study was to characterise the dosimetry and microdosimetry of thermal neutron exposures, using three simulation codes, as a precursor to quantitative radiobiological studies using blood samples. An irradiation line was designed employing a pyrolytic graphite crystal or-alternatively-a super mirror to expose blood samples to thermal neutrons from the National Research Universal reactor to determine radiobiological parameters. The crystal was used when assessing the relative biological effectiveness for dicentric chromosome aberrations, and other biomarkers, in lymphocytes over a low absorbed dose range of 1.2-14 mGy. Higher exposures using a super mirror will allow the additional quantification of mitochondrial responses. The physical size of the thermal neutron fields and their respective wavelength distribution was determined using the McStas Monte Carlo code. Spinning the blood samples produced a spatially uniform absorbed dose as determined from Monte Carlo N-Particle version 6 simulations. The major part (71%) of the total absorbed dose to blood was determined to be from the 14N(n,p)14C reaction and the remainder from the 1H(n,γ)2H reaction. Previous radiobiological experiments at Canadian Nuclear Laboratories involving thermal neutron irradiation of blood yielded a relative biological effectiveness of 26 ± 7. Using the Particle and Heavy Ion Transport Code System, a similar value of ∼19 for the quality factor of thermal neutrons initiating the 14N(n,p)14C reaction in soft tissue was determined by microdosimetric simulations. This calculated quality factor is of similar high value to the experimentally-derived relative biological effectiveness, and indicates the potential of thermal neutrons to induce deleterious health effects in superficial organs such as cataracts of the eye lens.

Sizing particles of natural uranium and nuclear fuels using poly-allyl-diglycol carbonate autoradiography.

Theoretical and experimental methods were developed to assess the size distribution of alpha-emitting particles captured on air-sampler filters. The particle size of oxides of low enriched, depleted and natural uranium and also aged plutonium in mixed oxide reactor fuels of known composition was determined using poly-allyl-diglycol carbonate (PADC) autoradiography, the commercial product TASTRAK((R)), solid-state nuclear track detectors. The exposed PADC was chemically etched to reveal clusters of tracks, radially dispersing from central points. A theoretical model was developed which converted the number of tracks in a track cluster to the hot particle diameter. The diameters of 26 particles of natural uranium oxide were measured (4-130 microm) using an optical microscope. There was a good agreement between these particle size measurements and a theoretical assessment based on the track cluster count.

Monte Carlo simulation of trabecular bone remodelling and absorbed dose coefficients for tritium and 14C.

A Monte Carlo simulation of multiple trabecular bone cavities in adult bone was developed and the absorbed radiation dose factors evaluated for 3H and 14C. The model was developed to assess the dose from radionuclide uptake in quiescent bone, but also the effects of temporal changes in bone turnover by incorporating bone-modelling units (BMU). Absorbed dose fractions were calculated for target regions that include the endosteal layer where radiation-sensitive stem cells in bone marrow are considered to reside preferentially. There were large differences in the absorbed fractions for two types of bone surface, quiescent and forming. Tritium in quiescent bone results in a dose to the endosteum about 20 times that for the same activity in forming bone surface irradiating osteoblasts. When the quiescent bone surface source was extended from an infinitely thin layer to a more realistic 1 microm thick, the tritium absorbed fractions for endosteum and red marrow targets fell by more than 2-fold.

Compartmental analysis of cerebrospinal fluid transfer and absorption in simulated hydrocephalus.

A mathematical model of the cerebrospinal fluid (CSF) flow dynamics found in hydrocephalic infants with myelomeningocele lesions was constructed using criteria obtained from analogous clinical situations where 125I-labelled and 131I-labelled ortho-iodobenzoyl-amino-acetic acid (hippuran) had been employed to measure CSF flow dynamics. The quantitative results from this study allowed clinical data to be assessed and the importance of various CSF transfer mechanisms to be discussed. Our mathematical model indicates that the majority of radiopharmaceutical passes from the cerebral reservoir (the ventricles) into the blood. Experimental evidence indicates that the principal mechanism responsible for this movement is the bulk flow of CSF between its sites of production in the choroid plexus and absorption by the arachnoid villi.

Correlation of regional formaldehyde flux predictions with the distribution of formaldehyde-induced squamous metaplasia in F344 rat nasal passages.

Squamous epithelium lines the nasal vestibule of the rat, rhesus monkey, and human. Respiratory, transitional, and olfactory epithelia line most areas posterior to the nasal vestibule. Inhaled formaldehyde gas induces squamous metaplasia posterior to the nasal vestibule and does not induce lesions in the nasal vestibule in rats and rhesus monkeys, indicating that squamous epithelium is resistant to irritant effects of formaldehyde and that squamous metaplasia may be an adaptive response. If squamous metaplasia is determined by formaldehyde dosimetry rather than by tissue-specific factors, squamous epithelium may be protective by absorbing less formaldehyde than other epithelial types. In a previous study, a three-dimensional, anatomically accurate computational fluid dynamics (CFD) model of the anterior F344 rat nasal passages was used to simulate inspiratory airflow and inhaled formaldehyde transport. The present study consisted of two related parts. First, the rat CFD model was used to test the hypothesis that the distribution of formaldehyde-induced squamous metaplasia is related to the location of high-flux regions posterior to squamous epithelium. Regional formaldehyde flux into nonsquamous epithelium predicted by the CFD model correlated with regional incidence of formaldehyde-induced squamous metaplasia on the airway perimeter of one cross-sectional level of the noses of F344 rats exposed to 10 and 15 ppm formaldehyde gas for 6 months. Formaldehyde flux into nonsquamous epithelium was estimated to vary by an order of magnitude depending on the degree of formaldehyde absorption by squamous epithelium. These results indicate that the degree to which squamous epithelium absorbs formaldehyde strongly affects the rate and extent of the progression of squamous metaplasia with continued exposure to formaldehyde. In the second part of this study, the CFD model was used to predict squamous metaplasia progression. Data needs for verification of this model prediction are considered. These results indicate that information on the permeability of squamous epithelium in rats, monkeys, and humans is important for accurate prediction of uptake in regions posterior to the nasal vestibule.

Mortality of iron foundry workers: IV. Analysis of a subcohort exposed to formaldehyde.

In the final phase of the mortality study of workers at an automotive iron foundry, a subset (N = 3929) of the original cohort of 8147 men, consisting of those exposed to formaldehyde during the period from January 1960 through May 1987, was analyzed. In addition to the external US population, an internal population (N = 2032), consisting of men who had worked in the same foundry during the same time period but not in formaldehyde-exposed jobs, was also used as a referent. Follow-up continued through December 31, 1989. Smoking status was ascertained for 65.4% of the exposed and for 55.1% of the unexposed cohorts. Detailed work histories and evaluation of occupational exposures by an industrial hygienist enabled us to categorize cumulative formaldehyde and silica exposures. Standardized mortality ratios were used to compare the mortality experience of the exposed cohort with the US population and, because of concerns about the healthy worker effect, with an occupational referent population. Relative risks for race, formaldehyde exposure status, smoking status, and silica exposure level were estimated by fitting a Poisson regression model to four causes of death: cancers of the buccal cavity and pharynx, lung cancer, diseases of the respiratory system, and emphysema. No association between formaldehyde exposure and deaths from malignant or nonmalignant diseases of the respiratory system was found. Cigarette smoking and silica exposure were found to be significantly associated with deaths attributed to lung cancer and disease of the respiratory system.

Dose to red bone marrow from natural radon and thoron exposure.

The age-dependent radiation dose to the haematopoietic tissue of bone marrow has been calculated for exposure to radon, thoron and their daughter products. The component of dose due to pure radon is dependent on the fat content of the marrow, since the solubility of radon in fat is about 16 times that in tissue. The mean dose equivalent muSv to the total active marrow is estimated for a range of fat cell diameters from 25 to 200 microns, taking account of the percentage cellularity and distribution of active marrow as a function of age. Similarly, the dose due to the inhalation of short-lived radon daughters was estimated, based on measurements in blood and marrow, modified to allow for the greater deposition of daughter products expected in children. An estimate of the age-dependent dose from long-lived radon daughters was made from uranium miner and natural exposure data. Dose estimates were made for the average UK indoor exposure to radon gas of 20 Bq/m3 and an equilibrium equivalent thoron concentration of 0.3 Bq/m3. The annual radon and thoron derived dose to the active marrow of the newborn was calculated as 30 and 40 muSv, respectively. For a 10-year-old child, the radon and thoron derived annual dose are 70 and 40 muSv, and for a 40-year-old adult 90 and 30 muSv, respectively. The above values exhibit wide range limits due principally to uncertainties in the accumulation of 210Pb in bone, and 210Po in marrow. These data indicate that at the average UK exposure, the alpha-particle dose to active marrow is dominated by that derived from inhaled radon and thoron compared with dietary intake. In infants the dose is dominated by thoron daughters. At the UK radon Action Limit of 200 Bq/m3, the radon and associated thoron derived dose is similar to that from all low LET sources. This work shows that the dose to red bone marrow from radon and thoron is significant, and that the possibility of leukaemia induced by these radiation sources warrants further investigation.

An evaluation of the Meditech M250 and a comparison with other CT scanners.

The Meditech M250 computerised tomography (CT) machine was evaluated during the first half of 1984. Measurements were made of noise, modulation transfer function, slice width, radiation dose profile, uniformity and linearity of CT number, effective photon energy and parameters relating to machine specification, such as pixel size and scan time. All breakdowns were logged to indicate machine reliability. A comparison with the established EMI CT1010 and CT5005 was made for noise, resolution and multislice radiation dose, as well as the dose efficiency or quality (Q) factor for both head and body modes of operation. The M250 was found to perform to its intended specification with an acceptable level of reliability.

Reconstruction of complex passageways for simulations of transport phenomena: development of a graphical user interface for biological applications.

Flow of fluids, such as blood, lymph and air, plays a major role in the normal physiology of all living organisms. Within individual organ systems, flow fields may significantly influence the transport of solutes, including nutrients and chemical toxicants, to and from the confining vessel walls (epithelia and endothelia). Computational fluid dynamics (CFD) provides a potentially useful tool for biologists and toxicologists investigating solute disposition in these flow fields in both normal and disease states. Application of CFD is dependent upon generation of accurate representations of the geometry of the system of interest in the form of a computational reconstruction. The present investigations, which were based on studies of the toxicology of inhaled reactive gases in the respiratory tract of rodents, provide computer programs for the generation of finite element meshes from serial tissue cross-sections. These programs, which interface with a commercial finite element fluid dynamics simulation package (FIDAP 7.05, Fluid Dynamics International, Evanston, IL), permit simulation of fluid flow in the complex geometries and local solute mass flux to the vessel walls of biological systems. The use of these programs and their application to studies of respiratory tract toxicology are described.

A biochemical-based model for the dosimetry of dietary organically bound tritium--Part 1: Physiological criteria.

In this paper the physiological criteria for a novel form of model are described whose biokinetics are governed by the overall metabolic reactions of the principal nutrients: carbohydrates, fats, and proteins. The biokinetics of a particular element are based primarily on the oxidation of glucose, fatty acids, and amino acids and the formation of water, carbon dioxide, and urea. The compartmental models proposed follow the pathways of the major elements including hydrogen and, hence, tritium. The parameters for two models of differing complexity--called the HCNO-S and HCNO-C models--were evaluated here on the basis of biochemical reactions; the results of compartmental analysis are reported in an accompanying paper. The simpler form of the HCNO model has single compartments representing the principal nutrients. The more complex model includes compartments representing the longer-term retention of carbohydrates as glycogen, fats as adipose tissue, and proteins in bone and soft tissues. The pool sizes and hydrogen transfer rates are estimated. The incorporation of biochemical reactions and important metabolic parameters serve to give the models a greater semblance of physiological merit than those currently available. For example, ingestion of carbohydrates results in a respiratory quotient of 1.0 and 100% of the hydrogen content oxidized to water, which are the same as values published in the literature. This form of metabolic model enables development of models for other isotopes, besides 3H, of the major elements of the body, e.g., 14C, 15N, 18O.

A biochemical-based model for the dosimetry of dietary organically bound tritium--Part 2: Dosimetric evaluation.

In this paper the dosimetry for a novel form of physiological model, whose biokinetics are governed by the overall metabolic reactions of the principal nutrients carbohydrates, fats and proteins, is evaluated by compartmental analysis. Two models of differing complexity, called the HCNO-S and HCNO-C models, were developed from parameters evaluated in an accompanying paper. The simpler form has single compartments representing the principal nutrients. The more complex model includes compartments representing the longer-term retention of carbohydrates as glycogen, fats as adipose tissue, and proteins in bone and soft tissues. The effective doses for various tritiated intakes are the same, or similar, as calculated by the two HCNO models, except for tritiated protein. The dose coefficient for an intake of tritiated water is approximately 8% greater than that recommended by the ICRP when the tritium body burden is considered as a homogenous pool. However, when the composition of individual organs is taken into account, the dose coefficient for an HTO intake is approximately 22% greater than the ICRP value. The HCNO-C dose coefficient for OBT in a normal diet is 5.0 x 10(-11) Sv Bq(-1), which is 1.2-fold greater than the ICRP dose coefficient for an OBT intake. The HCNO-C composition model gave organ and tissue doses with the largest range for a tritiated Reference Man dietary intake, the highest dose (red marrow, then breast) being around three-fold the lowest. A property of the HCNO models, important for bioassay analyses, is that a major part (> 90%) of an OBT intake is oxidized and excreted as HTO, which is physiologically more accurate than the current ICRP OBT model. The effective dose of specific tritiated foods, e.g., rice and wheat, was evaluated on the basis of their constituents.

Review of the ICRP tritium and 14C internal dosimetry models and their implementation in the Genmod-PC code.

Biokinetic models for tritium and 14C compounds, as described by various ICRP publications, have been incorporated into the Genmod-PC internal dosimetry code. This work reviews the models for tritium and 14C labeled compounds that the ICRP has formulated over several decades. The ICRP dosimetry prescribed for hydrogen and carbon radionuclides is fundamentally different from that recommended for other elements in that it is based on retention functions for whole body activity instead of compartmental biokinetic models. The ICRP recommends dosimetric methods for tritium and 14C compounds, ten of which are coded in Genmod-PC as compartmental models, namely, five tritium compounds, e.g., tritiated water, tritium gas, and five 14C compounds, e.g., carbon dioxide, carbon-labeled methane. The values of the Genmod-PC calculated dose coefficients were compared with the ICRP's values. It is shown how the dose coefficients for intakes of tritium and 14C compounds are affected by different interpretations of the methods recommended by the ICRP for two of the three classes of vapors and gases. Some aspects of the ICRP models, such as the percent oxidized, would benefit from reconsideration so as to produce tritium and 14C biokinetics that are less dependent on the radionuclide.

Dose to lung from inhaled tritiated particles.

Tritiated particulate materials are of potential hazard in fission, fusion, and other tritium handling facilities. The absorbed fractions (fraction of energy emitted that is absorbed by the target region) are calculated for tritiated particles deposited in the alveolar-interstitial (AI) region of the respiratory tract. The energy absorbed by radiologically sensitive tissue irradiated by tritiated particles, in regions of the lung other than in the AI region, is negligible. The ICRP Publication 71 assumes the absorbed fraction is unity for tritium deposited in the AI region. We employed Monte Carlo methods in a model to evaluate the energy deposition in the wall of the alveolar sac from particles of tritiated beryllium, tritiated graphite, titanium tritide, tritiated iron hydroxide and zirconium tritide. For the five materials examined, the absorbed fraction in alveolar tissue ranged from 0.31 to 0.61 for particles of 1 microm physical diameter and 0.07 to 0.21 for 5 microm diameter particles. The dose to alveolar tissue, for an acute inhalation of tritiated particles by an adult male worker, was calculated based on the ICRP 66 lung model and the particle dissolution model of Mercer (1967). For particles of 5 microm activity median aerodynamic diameter (AMAD), the committed equivalent dose to alveolar tissue, calculated for the five materials, ranged from 32-42%, respectively, of the committed equivalent dose derived assuming the absorbed fractions were unity.

Dose to the cell nucleus from exposure to tritiated pump oil or formaldehyde.

A Monte Carlo simulation of tritium decays in a cell composed of two parts, a nucleus and surrounding cytoplasm, was developed to evaluate the beta-radiation dose to the nucleus. A dose modifying factor (DMF), which is a ratio of the average nuclear dose to the whole-tissue dose, after skin-contact exposure of rats to tritiated pump oil or tritiated formaldehyde was estimated. Biokinetic data characterizing the retention of tritium in liver were available in the form of tritium-specific activities and biological half-times for tritiated water and five macromolecular species (DNA, RNA, acid-soluble fraction, acid-insoluble protein, and lipids). The spatial distribution of tissue-free water and macromolecular species in the nucleus and cytoplasm of rat liver cells was based on published data. In the case of exposure to tritiated pump oil, tritium incorporated into lipids provides the largest percentage (60%) of the absorbed dose to the nucleus. For the tritiated-formaldehyde exposure, the tritium dose to the nucleus is overwhelmingly contributed by tritiated water (58%) and in acid-insoluble proteins (40%). For both these tritiated organic exposures, the tritium-labeled DNA has a negligible effect on the DMF. The DMF for the tritiated pump oil and formaldehyde exposures was estimated as 0.81 and 1.05, respectively: the DMF of both exposures was close to unity. Given the other uncertainties in tritium dosimetry, our results suggest that for these skin-contact exposures a uniform distribution of tritium in tissue is an adequate assumption for dosimetry.

Dose contribution from metabolized organically bound tritium after acute tritiated water intakes in humans.

Urine samples from eight male radiation workers who had an unplanned acute tritiated water intake were measured for tritium-in-urine up to 300 d post-exposure. During the first month or so post-exposure, these individuals increased their fluid intakes to accelerate the turnover rate of tritium in the body for dose mitigation. Their daily fluid intakes reverted to normal levels in the latter period of the study. A non-linear regressional analysis of the tritium-in-urine data showed that the average biological half-life of tritium in body water, with standard deviation, was 6.3 +/- 1.0 d (range, 5.0-8.1 d) and 8.4 +/- 2.0 d (range, 6.2-12.8 d) during the respective periods of increased fluid intake and the later period of normal fluid intake. A longer term component of tritium excretion was also observed with average biological half-life of 74 +/- 18 d (range, 58-104 d), indicating the incorporation of tritium, and its retention, in the organic fractions of the body. A mathematical model was developed and used to estimate the dose increase from the metabolized organically bound tritium on the basis of the kinetics of tritium-in-urine. The model accounts for a change in the rates of urinary excretion caused by variable fluid intakes. The average dose to the body, for the eight male workers, due to the metabolized organically bound tritium was estimated to be 6.2 +/- 1.3% (range, 3.5% to 8.9%) of the committed effective dose due to tritium in the body water. This value for the dose increase from organically bound tritium is in the range of the current recommendations of the International Commission on Radiological Protection, i.e., organically bound tritium incorporated into the body contributes about 10% of the dose to the body water following tritiated water intakes.

Influence of gender differences in the carbon pool on dose factors for intakes of tritium and 14C-labeled compounds.

The ICRP's biokinetic models for five tritium-labeled and five 14C-labeled compounds (not including radiopharmaceutical compounds and excepting carbon monoxide) incorporate a compartment representing the body carbon pool. Using the ICRP models, as coded into the Genmod-PC internal dosimetry code, higher dose coefficients are calculated for females than for ICRP's Reference Man. The ICRP's committed effective dose coefficients for the ingestion of tritiated water and organically bound tritium by the adult male are 1.8 x 10(-11) and 4.2 x 10(-11) Sv Bq(-1), respectively. Using the Genmod-PC code, the corresponding dose coefficients for the adult female are 2.2 x 10(-11) and 6.2 x 10(-11) Sv Bq(-1), which are 25% and 46% greater than the adult male's. Similarly, the ICRP's dose coefficient is 5.8 x 10(-10) Sv Bq(-11) for an intake of organically bound 14C by the adult male, and the estimated dose coefficient using Genmod is 54% greater for the adult female. The carbon retention half-time for an average adult female is calculated as 51 d and that for an average adult male, 38 d; the latter is similar to the carbon half-time of 40 d recommended by International Commission on Radiological Protection (ICRP). The longer turnover time of whole body carbon in females is one factor that causes the dose coefficients for females to be higher than those of males; a second factor is the smaller whole body mass of ICRP's Reference Woman compared to Reference Man.

Monte Carlo determination of age-dependent steady-state dose to red bone marrow and bone from 14C exposure.

Monte Carlo simulations were carried out to calculate the age-dependent dose from the beta decay of 14C to marrow and bone on the basis of a steady-state specific-activity model. A model of the trabecular cavity containing spherical fat cells in a square lattice surrounded by haemopoietic tissue was employed. The age-dependent 14C dose to haemopoietic (red) marrow was estimated taking account of the change in the fat cell size with fat fraction. Allowances were made for the change in the percentage cellularity and distribution of active marrow in the whole skeleton as a function of age. Age-dependent changes in trabecular cavity size and bone composition were found to have only a small effect on dose. Dose rates were estimated under steady-state conditions, for food ingested with a 14C specific activity of 1 Bq g(-1) of C. The equivalent dose rate to the haemopoietic tissue of a 20 year-old adult is 77 microSv a(-1), and 39 microSv a(-1) for a 3-month-old infant. Similarly, the equivalent dose rate to the bone surfaces of an adult is 48 microSv a(-1), and 38 microSv a(-1) for an infant. Therefore, the equivalent dose rate to marrow and bone stem cells increases with age under steady state conditions.

Sizing alpha emitting particles of aged plutonium on personal air sampler filters using CR-39 autoradiography.

Methods have been developed to assess the size distribution of alpha emitting particles of reactor fuel of known composition captured on air sampler filters. The sizes of uranium oxide and plutonium oxide particles were determined using a system based on CR-39 solid-state nuclear track detectors. The CR-39 plastic was exposed to the deposited particles across a 400 microm airgap. The exposed CR-39 was chemically etched to reveal clusters of tracks radially dispersed from central points. The number and location of the tracks were determined using an optical microscope with an XY motorised table and image analysis software. The sample mounting arrangement allowed individual particles to be simultaneously viewed with their respective track cluster. The predicted diameters correlated with the actual particle diameters, as measured using the optical microscope. The efficacy of the technique was demonstrated with particles of natural uranium oxide (natUO2) of known size, ranging from 4 to 150 microm in diameter. Two personal air sampler (PAS) filters contaminated with actinide particles were placed against CR-39 and estimated to have size distributions of 0.8 and 1.0 microm activity median aerodynamic diameter (AMAD).

Retention and excretion of inhaled 3H and 14C radiolabeled methane in rats.

A radiological concern for workers at heavy water reactor nuclear facilities is the hazard presented by tritium (H) and C. Radioactive methane is one of many potential H and C containing chemicals to which Nuclear Energy Workers (NEWs) may be exposed. Current dosimetric models for H- and C-methane, recommended by the International Commission on Radiological Protection (ICRP), are based on the assumption that 1% of methane is absorbed following its inhalation. Of this 1%, all H is converted immediately to tritiated water and C is converted immediately to CO2 (50%) and organically bound carbon (50%). In the study, rats were exposed to methane standards (H-methane and C-methane) mixed with breathing air to give a final concentration of 0.27% methane and resulting in final activity concentrations of 4.2 GBq m and 0.88 GBq m for H and C, respectively. This corresponds to exposure estimates of 580 kBq g and 120 kBq g. Simultaneous exposure to H- and C-methane allowed for the direct comparison of the retention of these radionuclides and removed uncertainties concerning their relative uptake and retention. The results demonstrate that the total methane uptake from the inhaled dose was threefold less than the 1% methane uptake predicted by the ICRP dosimetric models for H- and C-methane, with the H concentration being substantially higher than anticipated in the liver. This study provided data suggesting that current ICRP dosimetric methane models overestimate the fraction of H- and C-methane that is absorbed following inhalation and assisted in providing information to better understand the metabolism of inhaled H and C radiolabeled methane.

Toxicity of irradiated advanced heavy water reactor fuels.

The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.