Dose assessment for ingestion of a 330 kilobecquerel 60Co hot particle

On leaving the irradiated fuel bay at Pickering A nuclear power station, a worker triggered a whole body monitor alarm with activity in or on his head, and despite careful decontamination techniques he subsequently swallowed a hot particle. Over the next 3 d, the radioactivity was tracked through the body. It was then excreted in a single faecal sample and recovered for physical and radiochemical analysis. This analysis demonstrated that the particle contained 330 kBq of 60Co and only traces of other radioactivity. Its dimensions were approximately 50-130 microm and its composition was consistent with that of Stellite 6. A dose assessment was carried out taking into account the residence time of the particle in the mouth and its transit through the body. The estimated committed effective dose was 1.4 mSv, and the equivalent dose to the maximally exposed 1 cm2 of skin, 81 mSv.

Radioecological investigations in shallow bays of the Novaya Zemlya Archipelago in 2002-2005

The results of a survey of underwater disposal sites of potentially hazardous objects in the Kara Sea and Oga, Tsivolky, Stepovoy and Abrosimov Bays are presented. Radionuclide levels were determined, using gamma-spectrometric and radiochemical methods, in zones near to and remote from buried solid radioactive waste in the outer and inner parts of the Bays. At the repository of the solid radioactive waste containers in the inner part of the Stepovoy Bay and Abrosimov Bay, higher than background concentrations of (137)Cs were determined in samples of bottom sediments. At one of the sites located in the Stepovoy Bay (137)Cs and (60)Co were detected.

Power Saturation of ESR Signal in Ammonium Tartrate Exposed to60Co γ-Ray Photons, Electrons and Protons

In this paper we present an investigation of the electron spin resonance (ESR) line shape of ammonium tartrate (AT) dosimeters exposed to radiation with different linear energy transfer (LET). We exposed our dosimeters to gamma-ray photons ((60)Co), 7 MeV and 14 MeV initial energy electrons, and 19.3 MeV initial energy protons. The differences in the power saturation behavior of ESR spectra of AT irradiated with photons, electrons and protons could be correlated to the effective LET of the radiation beams. We analyzed the behavior of peak-to-peak amplitude as a function of microwave power, and we developed a fitting procedure that permits us to obtain the dependence of the homogeneity parameter of the line shape on the LET of the radiation using the Castner saturation theory. This simple procedure allows us to distinguish the LET of the radiation beam.

Management of spent sealed radioactive sources in Turkey

Spent radioactive sources (SRS) have been generated from industrial applications, research, and medicine in Turkey. In this study, management of SRS (Co, Cs) at Cekmece Waste Processing and Storage Facility (CWPSF) is described. Eleven Cs sources (total 851 GBq) and four Co sources (total 27.75 GBq) that had been used as levels and density gauges were conditioned. Reinforced metal drums (200 L in volume) and cement matrix were used for conditioning of these sources. In this way, greater confinement was achieved for long-term storage. Maximum dose rates at the surface of the conditioned waste package were determined. In addition to information about conditioning stages of the sources, various calculations that have been done for shielding are presented. Surface dose rates of the waste packages were 1.60 mSv h for Cs and 1.63 mSv h for Co. Measurements of the final waste packages were presented to fulfill the requirements (<2 mSv h) of transportation according to regulations for the safe transport of radioactive material.

Treatment of radioactive liquid waste by sorption on natural zeolite in Turkey

Liquid radioactive waste has been generated from the use of radioactive materials in industrial applications, research and medicine in Turkey. Natural zeolites (clinoptilolite) have been studied for the removal of several key radionuclides ((137)Cs, (60)Co, (90)Sr and (110m)Ag) from liquid radioactive waste. The aim of the present study is to investigate effectiveness of zeolite treatment on decontamination factor (DF) in a combined process (chemical precipitation and adsorption) at the laboratory tests and scale up to the waste treatment plant. In this study, sorption and precipitation techniques were adapted to decontamination of liquid low level waste (LLW). Effective decontamination was achieved when sorbents are used during the chemical precipitation. Natural zeolite samples were taken from different zeolite formations in Turkey. Comparison of the ion-exchange properties of zeolite minerals from different formations shows that Gordes clinoptilolite was the most suitable natural sorbent for radionuclides under dynamic treatment conditions and as an additive for chemical precipitation process. Clinoptilolite were shown to have a high selectivity for (137)Cs and (110m)Ag as sorbent. In the absence of potassium ions, native clinoptilolite removed (60)Co and (90)Sr very effectively from the liquid waste. In the end of this liquid waste treatment, decontamination factor was provided as 430 by using 0.5 mm clinoptilolite at 30 degrees C.

Radiological consequences of the extreme flooding on the lower course of the Rhone valley (December 2003, south east France)

In early December 2003 unusual weather conditions led to major flooding of the lower Rhone valley. When it floods, the Rhone carries large masses of solid matter in suspension, which potentially includes associated artificial (anthropogenic) radioactive contaminants from soil drainage in the catchment area and from re-uptake of sedimentary matter that has been contaminated with low-level radioactive liquid effluents from almost twenty nuclear facilities situated along the Rhone valley. A sampling campaign was carried out to investigate the level and spread of both sediment mass and associated radioactive contamination across the flooded areas. An attempt was made to assess the radiological consequences of such an extreme event on contamination of the food chain. Our results show that almost 700,000 tons of sediment was transported onto the floodplain, of which 80% were coarse and fine sands. These materials transferred 6660 MBq of 137Cs, 93 MBq of (239+240)Pu, 13 MBq of 238Pu and 204 MBq of 60Co over a surface area of 60 km2. More than 90% of deposited sediments are concentrated in a 10 km2 area of agricultural soils, and we estimated that 18% were plowed into the soil. Nevertheless, the level of activity measured in the vegetable crops and milk was not significantly different from the level measured in similar samples from regions that were not affected by the December 2003 floods.

Radiation exposure in mossbauer spectroscopy

This paper deals with the risks involved in routine operations as well as in potentially hazardous management of radioactive sources in a typical Mössbauer spectroscopy research laboratory. It comes from a direct experience at the University of Parma, Italy, and from information obtained from more than 50 different Mössbauer laboratories all over the world. A review of the normal procedures performed by researchers and technicians in the Mössbauer laboratory is presented along with the relative dose evaluations. Radiation doses have been calculated and measured from various situations, ranging from daily operations and maintenance work to radiation risks in case of incorrect procedures and/or device failures.

Changes in optical transmission caused by gamma ray induced coloring in photoluminescence dosimeter

Transmission of visible light and ultraviolet radiation was examined for a phosphate-glass photoluminescence dosimeter irradiated with Co source gamma rays in the dose range of 1-60 Gy (H2O). The transmission for the wavelengths (lambda) less than 600 nm decreased with increasing irradiation dose beginning at 6 Gy. An approximate 20% reduction of transmission was observed for a 60 Gy exposure at the wavelength of ultraviolet radiation used for excitation (lambda = 337 nm). However, no change of transmission was seen in longer wavelength region (lambda > 600 nm), which includes the range of photoluminescence (lambda = 610-710 nm). Relative efficiencies of measured photoluminescence agreed well with estimations that were calculated from the transmission reduction of ultraviolet radiation. This fact indicates that reduction of photoluminescence efficiency induced by high-dose gamma rays is attributable mostly to attenuation of the ultraviolet radiation from an excitation source, rather than saturation of trapping or recombination centers.

A comparison of Australian and Canadian calibration coefficients for air kerma and absorbed dose to water for60Co γ radiation

Australian and Canadian calibration coefficients for air kerma and absorbed dose to water for 60Co gamma radiation have been compared using transfer standard ionization chambers of types NE 2561 and NE 2611A. Whilst the primary standards of air kerma are similar, both being thick-walled graphite cavity chambers but employing different methods to evaluate the Awall correction, the primary standards of absorbed dose to water are quite different. The Australian standard is based on measurements made with a graphite calorimeter, whereas the Canadian standard uses a sealed water calorimeter. The comparison result, expressed as a ratio of calibration coefficients R=N(ARPANSA)/N(NRC), is 1.0006 with a combined standard uncertainty of 0.35% for the air kerma standards and 1.0052 with a combined standard uncertainty of 0.47% for the absorbed dose to water standards. This demonstrates the agreement of the Australian and Canadian radiation dosimetry standards. The results are also consistent with independent comparisons of each laboratory with the BIPM reference standards. A 'trilateral' analysis confirms the present determination of the relationship between the standards, within the 0.09% random component of the combined standard uncertainty for the three comparisons.

The US radiation dosimetry standards for Co60 therapy level beams, and the transfer to the AAPM accredited dosimetry calibration laboratories

This work reports the transfer of the primary standard for air kerma from the National Institute of Standards and Technology (NIST) to the secondary laboratories accredited by the American Association of Physics in Medicine (AAPM). This transfer, performed in August of 2003, was motivated by the recent revision of the NIST air-kerma standards for 60Co gamma-ray beams implemented on July 1, 2003. The revision involved a complete recharacterization of the two NIST therapy-level 60Co gamma-ray beam facilities, resulting in new values for the air-kerma rates disseminated by the NIST. Some of the experimental aspects of the determination of the new air-kerma rates are briefly summarized here; the theoretical aspects have been described in detail by Seltzer and Bergstrom ["Changes in the U.S. primary standards for the air-kerma from gamma-ray beams," J. Res. Natl. Inst. Stand. Technol. 108, 359-381 (2003)]. The standard was transferred to reference-class chambers submitted by each of the AAPM Accredited Dosimetry Calibration Laboratories (ADCLs). These secondary-standard instruments were then used to characterize the 60Co gamma-ray beams at the ADCLs. The values of the response (calibration coefficient) of the ADCL secondary-standard ionization chambers are reported and compared to values obtained prior to the change in the NIST air-kerma standards announced on July 1, 2003. The relative change is about 1.1% for all of these chambers, and this value agrees well with the expected change in chambers calibrated at the NIST or at any secondary-standard laboratory traceable to the new NIST standard.

A new approach to the determination of air kerma using primary-standard cavity ionization chambers

A consistent formalism is presented using Monte Carlo calculations to determine the reference air kerma from the measured energy deposition in a primary-standard cavity ionization chamber. A global approach avoiding the use of cavity ionization theory is discussed and its limitations shown in relation to the use of the recommended value for W. The role of charged-particle equilibrium is outlined and the consequent requirements placed on the calculations are detailed. Values for correction factors are presented for the BIPM air-kerma standard for 60Co, making use of the Monte Carlo code PENELOPE, a detailed geometrical model of the BIPM 60Co source and event-by-event electron transport. While the wall correction factor k(wall) = 1.0012(2) is somewhat lower than the existing value, the axial non-uniformity correction k(an) = 1.0027(3) is significantly higher. The use of a point source in the evaluation of k(an) is discussed. A comparison is made of the calculated dose ratio with the Bragg-Gray and Spencer-Attix stopping-power ratios, the results indicating a preference for the Bragg-Gray approach in this particular case. A change to the recommended value for W of up to 2 parts in 10(3) is discussed. The uncertainties arising from the geometrical models, the use of phase-space files, the radiation transport algorithms and the underlying radiation interaction coefficients are estimated.

Patterns and inventories of radioactive contamination of island sites of the Yenisey River, Russia

The distribution of radioactive contamination at three island sites downstream from the Krasnoyarsk Mining and Chemical Combine (KMCC) was studied with the objectives of mapping contamination levels, interpreting radionuclide distributions through consideration of alluvial processes and determining radionuclide inventories. Contamination was measured using in situ gamma spectrometry and landforms characterised using topographic surveying methods. Maximum (137)Cs contamination densities (700 kBq m(-2)) were found on low- and middle-level floodplains and low-lying interconnecting areas of Beriozovy Island (16 km from the KMCC). On Mikhin Island (180 km from the KMCC) maximum total (60)Co, (152)Eu and (154)Eu activity concentrations (30-40 kBq m(-2)) occur in low-lying areas inundated during flooding. Maximum (137)Cs and total (60)Co, (152)Eu and (154)Eu contamination densities on Cheriomukhov Island (250 km from the KMCC) were 390 and 50 kBq m(-2), respectively. Estimated (137)Cs inventories were 145, 148 and 16GBq for Beriozovy, Mikhin and Cheriomukhov Islands, respectively.

Optical absorption and thermoluminescence in single NaCl:Cu crystals exposed to 60Co and UV light

Optical absorption (OA) and thermally stimulated luminescence measurements were performed on NaCl:Cu+(0.04 and 0.08%) crystals blocks grown by the Czochralski technique. The NaCl:Cu+ crystals were exposed to gamma rays from a 60Co source (0.954-30 kGy) as well as UV radiation. The radiation-induced defects were mainly F, Cu- and Cu+ centres, with absorption bands located at 464, 256.7 and 236 nm, respectively. The absorption bands were found to be independent of the Cu impurity concentration. As the gamma-dose irradiation increased, the absorption band at 256.7 nm decreased while the band at 236.3 nm increased highly along with the 256.7 nm band. The F-centres produced at high gamma-radiation dose while thermally bleached showed an increase of the Cu+ OA bands with a simultaneous decrease of Cu- absorption band. The bleaching with F-light showed the participation of the F centre generated by gamma radiation on the ion valence changes of the doping impurity as well as on the TL phenomenon. The TL measurements in NaCl:Cu crystals with both impurity concentrations demonstrated that the Cu+ concentration has a strong influence on the intensity and shape of the glow peaks.

Estimated radiation dose to breast feeding infant following maternal administration of 57Co labelled to vitamin B12

Administration of a radiopharmaceutical may result in a radiation dose to an infant due to ingestion of the radiopharmaceutical secreted in the breast milk. Following a maternal administration of Co labelled to vitamin B12 (cyanocobalamin) as part of a Schilling test an estimate of the absorbed dose to a breast feeding infant was calculated. Milk samples were collected from every feed in the first 24 h, and at approximately 48 and 72 h post-administration. The absorbed dose to the infant's liver (the organ receiving the highest dose) was calculated to be 0.23 mGy. The effective dose to the infant was calculated to be 0.025 mSv, which is considerably lower than the current regulatory limit of 1 mSv. The Administration of Radioactive Substances Advisory Committee advise that the first feed, at approximately 4 h after administration, be discarded. The data show that this was unwarranted, and that the peak concentration of Co in the breast milk occurred at around 24 h.

Shift in absorbed dose for megavoltage photons when changing to TRS-398 in Australia

Australian primary standards of air kerma and absorbed dose are realized in 60Co gamma rays. To calibrate the megavoltage photon beams from linear accelerators, radiotherapy centres have their ionization chamber calibrated in a 60Co beam and then use a protocol to transfer this calibration to the higher energy. The radiotherapy community is in the process of changing from the ACPSEM Protocol (Second Edition 1998) based on an air kerma calibration to the IAEA's TRS-398 Code of Practice, based on an absorbed dose to water calibration. To evaluate the shift in absorbed dose resulting from the new protocol, the absorbed dose should be determined using both protocols and compared. We present a formula for this shift which can be used to check the result. To use this formula the centre needs to measure a displacement correction and know the ratio of the air kerma to absorbed dose to water calibration factors at 60Co. We calculate the change they should expect by using the average ratio of the air kerma and absorbed dose to water calibration factors for NE2571 and NE2561 chambers, based on Australian standards, and by estimating the displacement correction from published depth dose data. We find the absorbed dose in a megavoltage photon beam to increase by between 0.1 and 0.6% for NE2571 chambers and between 0.7 and 1.1% for NE2561 chambers, for beams up to 35 MV. The dose measured using TRS-398 is always higher.

Radiation effectiveness factors for use in calculating probability of causation of radiogenic cancers

This paper presents so-called radiation effectiveness factors that are intended to represent the biological effectiveness of different radiation types, relative to high-energy Co gamma rays, for the purpose of estimating cancer risks and probability of causation of radiogenic cancers in identified individuals. Radiation effectiveness factors are expressed as subjective probability distributions to represent uncertainty that arises from uncertainties in estimates of relative biological effectiveness obtained from radiobiological studies of stochastic endpoints, limited data on biological effectiveness obtained from human epidemiological studies, and other judgments involved in evaluating the applicability of available information to induction of cancers in humans. Primarily on the basis of reviews and evaluations of available data by experts, probability distributions of radiation effectiveness factors are developed for the following radiation types: neutrons of energy less than 10 keV, 10-100 keV, 0.1-2 MeV (including fission neutrons), 2-20 MeV, and greater than 20 MeV; alpha particles of any energy emitted by radionuclides; photons of energy 30-250 keV and less than 30 keV; and electrons of energy less than 15 keV. Photons of energy greater than 250 keV and electrons of energy greater than 15 keV are assumed to have the same biological effectiveness as reference Co gamma rays and are assigned a radiation effectiveness factor of unity, without uncertainty. For neutrons and alpha particles, separate probability distributions of radiation effectiveness factors are developed for solid tumors and leukemias, and small corrections to represent an inverse dose-rate effect are applied to those distributions in cases of chronic exposure. A radiation effectiveness factor different from unity for 15-60 keV electrons is discussed but is not adopted due to a lack of relevant radiobiological data. Radiation effectiveness factors presented in this paper are incorporated in the Interactive RadioEpidemiological Program and were developed for use by The National Institute for Occupational Safety and Health and U.S. Department of Labor in evaluating claims for compensation for radiogenic cancers by workers at U.S. Department of Energy facilities.

A survey of IAEA/WHO 60Co TLD postal dose intercomparison exercises during 1985-2003

Results of the survey of IAEA/WHO Co postal dose intercomparison exercises in which the Secondary Standard Dosimetry Laboratory participated during the period 1985-2003 are presented. In 14 exercises spread over a period of 19 years, the deviation in therapy dose level measurements of the Secondary Standard Dosimetry Laboratory has never exceeded the International Atomic Energy Agency Dosimetry Laboratory acceptable levels of accuracy. The IAEA/WHO thermoluminescent dosimeter postal dose intercomparison service, the procedure for irradiation of thermoluminescent dosimeters, and the water absorbed dose determination are briefly described.

The effective dose equivalent and effective dose for hot particles on the skin

Whole body exposure from photon-emitting hot particles is a relatively new problem. Until recently, the U.S. Nuclear Regulatory Commission required the use of deep dose equivalent (DDE) to estimate and report whole body exposures from hot particles. In this study, effective dose equivalent (EDE) and effective dose (ED) were calculated for point sources with photon energies between 0.1 MeV to 2.0 MeV for 74 locations covering the entire body surface, using the MCNP code and the MIRD-type stylized phantoms. Tabulated data show that the sources located near the upper chest and the lower waist have the highest EDE and ED, while sources near the top of head and feet yielded the smallest. The calculated DDE values are much higher than the EDE values. For an exposure of 75 microCi h(-1) to a 60Co source located at the center upper chest area, the EDE is 36.5 microSv (3.65 mrem), which is a factor of 240 smaller than the corresponding DDE. EDE and ED data are tabulated for quick reference by users in nuclear power plants.

Neutron-activation revisited: the depletion and depletion-activation models

The growth of a radioactive daughter in neutron activation is commonly described with the saturation model that ignores the consumption of parent nuclei during the radio-activation process. This approach is not valid when radioactive sources with high specific activities are produced or when the particle fluence rates used are very high. Assuming a constant neutron fluence rate throughout the activation target, a neutron-activation model that accounts for the depletion in parent nuclei is introduced. This depletion model is governed by relationships similar to those describing the parent-daughter-granddaughter decay series, and, in contrast to the saturation model, correctly predicts the practical limit of the daughter specific activity, irrespective of the particle fluence rate. Also introduced is a neutron-activation model that in addition to parent depletion accounts for the neutron activation of daughter nuclei in situations where the cross section for this effect is high. The model is referred to as the depletion-activation model and it provides the most realistic description for the daughter specific activity in neutron activation. Three specific neutron activation examples of interest to medical physics are presented: activation of molybdenum-98 into molybdenum-99 described by the saturation model; activation of cobalt-59 into cobalt-60 described by the depletion model; and activation of iridium-191 into iridium-192 described by the depletion-activation model.

Field and model investigations of external gamma dose rates along the Cumbrian coast, NW England

A survey of the contribution to external dose from gamma rays originating from intertidal sediments in the vicinity of the British Nuclear Group Sellafield site showed that the major anthropogenic contributions were due to (137)Cs and (60)Co. At some sites, traces of other anthropogenic radionuclides were detected, namely (106)Ru, (125)Sb, and (154)Eu. The proportions of fine grained material (<63 microm) were used to improve model predictions of dose contribution due to external exposure to gamma rays, using the CUMBRIA77/DOSE77 model. Model dose predictions were compared to those directly measured in the field. Using the new proportions of fine grained material (1-17.5%) in conjunction with field gamma-ray spectra, model predictions were improved considerably for most sites. Exceptions were at Drigg Barn Scar and Whitehaven Coal Sands sites, which had their own unique characteristics. The highest (60)Co activity concentrations in this study were detected at Drigg Barn Scar. These relatively high activity concentrations of (60)Co were due to the presence of (60)Co in mussels and barnacles, hence upsetting the fine sediment relationships used in previous dose calculations. Whitehaven Coal Sands was unusual in that it contained higher levels of radionuclides than would be expected in sandy sediment. The mineralogy of these sediments was the controlling factor on (137)Cs binding, rather than the proportion of fine grained material. By adjusting the effective fine grained sediment proportions for calculations involving (60)Co and (137)Cs at Drigg Barn Scar and Whitehaven Coal Sands respectively, the CUMBRIA77/DOSE77 model predictions could be improved upon significantly for these sites. This work highlights the influence of particle size and sediment composition on external dose rate calculations, as well as the potential for external dose contributions from biota.

An experimental and Monte Carlo investigation of the energy dependence of alanine/EPR dosimetry: II. Clinical electron beams

The energy dependence of alanine/EPR dosimetry for 8, 12, 18 and 22 MeV clinical electron beams was investigated by experiment and by Monte Carlo simulations. Alanine pellets in a waterproof holder were irradiated in a water phantom using an Elekta Precise linear accelerator. The dose rates at the reference point were determined following the TG-51 protocol using an NACP-02 parallel-plate chamber calibrated in a (60)Co beam. The EPR spectra of irradiated pellets were measured using a Bruker EMX 081 EPR spectrometer. Experimentally, we found no significant change in alanine/EPR response to absorbed dose-to-water over the energy range 8-22 MeV at an uncertainty level of 0.6%. However, the response for high-energy electrons is about 1.3 (+/-1.1)% lower than for (60)Co. The EGSnrc Monte Carlo system was used to calculate the ratio of absorbed dose-to-alanine to absorbed dose-to-water and it was shown that there is 1.3 (+/-0.2)% reduction in this ratio from the (60)Co beam to the electron beams, which confirms the experimental results. Alanine/EPR response per unit absorbed dose-to-alanine was also investigated and it is the same for high-energy electrons and (60)Co gamma-rays.

Monte Carlo estimation of the dose and heating of cobalt adjuster rods irradiated in the CANDU 6 reactor core

The present work is a part of a more complex project related to the replacement of the original stainless steel adjuster rods with cobalt assemblies in the CANDU 6 reactor core. The 60Co produced by 59Co irradiation could be used extensively in medicine and industry. The paper will mainly describe some of the reactor physics and safety requirements that must be carried into practice for the Co adjuster rods. The computations related to the neutronic equivalence of the stainless steel adjusters with the Co adjuster assemblies, as well as the estimations of the activity and heating of the irradiated cobalt rods, are performed using the Monte Carlo codes MCNP5 and MONTEBURNS 2.1. The activity values are used to evaluate the dose at the surface of the device designed to transport the cobalt adjusters.

Energy and angular anisotropy optimisation of a p-type diode for in vivo dosimetry in photon-beam radiotherapy

We present simulation work using the Monte Carlo code MCNPX that shows that there is a possibility of improving the silicon p-type diode as a radiation dosemeter, by altering the construction of the diode. Altering the diode die thickness can reduce the inherent angular anisotropy of the diode, with little effect on its energy response. Conversely, the contact material and geometry have a large impact on the energy response with little effect on the inherent angular anisotropy. By correct choice of contact material, the typical over-response -100 keV relative to the response at 60Co energy can be reduced from approximately 20 to 4. It is expected that further enhancements may be made with different geometries and materials.

Application of a sitting MIRD phantom for effective dose calculations

In typical realistic scenarios, dose factors due to 60Co contaminated steel, used in consumer products, cannot be approximated by standard exposure geometries. It is then necessary to calculate the effective dose using an appropriate anthropomorphic phantom. MCNP calculations were performed using a MIRD human model in two settings. In the first, a male office worker is sitting in a chair containing contaminated steel, surrounded by contaminated furniture. In the second, a male driver is seated inside an automobile, the steel of which is uniformly contaminated. To accurately calculate the dose to lower body organs, especially the gonads, it was essential to modify the MIRD model to simulate two sitting postures: chair and driving position. The phantom modifications are described, and the results of the calculations are presented. In the case of the automobile scenarios, results are compared to those obtained using an isotropic fluence-to-dose conversion function.

MCNP simulation of a Theratron 780 radiotherapy unit

A Theratron 780 (MDS Nordion) 60Co radiotherapy unit has been simulated with the Monte Carlo code MCNP. The unit has been realistically modelled: the cylindrical source capsule and its housing, the rectangular collimator system, both the primary and secondary jaws and the air gaps between the components. Different collimator openings, ranging from 5 x 5 cm2 to 20 x 20 cm2 (narrow and broad beams) at a source-surface distance equal to 80 cm have been used during the study. In the present work, we have calculated spectra as a function of field size. A study of the variation of the electron contamination of the 60Co beam has also been performed.